Text

Exam Rept 50-395/92-301 Administered During Wk of 921207. Exam Results:All Applicants Passed Exam
	ML20127G877
Person / Time
Site:	Summer
Issue date:	01/05/1993
From:	Lawyer L, Edwin Lea; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20127G820	List: IR 05000395/1992301; ML20127G816;
References
	50-395-92-301, NUDOCS 9301220103
	Download: ML20127G877 (247)
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pt # U N UNITED STATES e4 4'o NUCLEAR HEGULATORY COMMISSION

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'4 4 . . . . . ,o ENCLOSURE 1 EXAMINATION REPORT - 50-395/92-301 Facility Licensee: South Carolina Electric and Gas Company Facility Name: V. C. Summer Nuclear Station Facility Docket No.: 50-395 Facility License No.: DPR-12 The NRC administered examinations at the V. C. Summer Nuclear Plant near Jenkinsville, S Chief Examiner: [Ad Edwin Lea'

o .> / - 5- ff Date Signed Approved By: M % ,r * y I /- A 93 Lawrence'Lf' Lawyer, C6ief Date Signed Operator Licensing Section 1 Division of Reactor Safety SUMMARY Scope: The NRC conducted initial examinations during the week of December 7, 1992. Written and operating examinations were administered to one Senior Reactor Operator (SRO) applicant and eleven Reactor Operator (RO) applicant Results: All applicants passed the examination Weaknesses were identified in the following areas:_ (1) knowledge of how plant systems respond to steam flow instrumentation failures (paragraph "3.c"), and (2) the adequacy of procedure A0P 401.3 (paragraph "3.d.2").

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REPORT DETAILS 1. Facility Employees Attending Exit '

S. Furstenberg, Associate Manager of Operations T. Matlosz, Supervisor of Nuclear Training K. Nettles, General Manager, Station Support R. Quick, Senior Instructor G. Taylor, General Manager, Nuclear Operations- .

K. Woodward, Manager of Nuclear Training 2. Examiners G. Hopper, Region II E. Lea, Chief Examiner, Region II K. Faris, PNL L. Sherfey, PNL 3. Discussion Examinatio. .esults During the week of December 7, 1992, the NRC administered written examinations and operating tests to one Senior Reactor Operator (SRO) applicant and eleven Reactor Operator (RO) applicants. All-applicants passed the examination The facility's training staff had no post-examination comment The licensee attributed the absence of post-exam comments to the extensive preexamination reviews, Reference material The licensee submitted all material requested in the 90 day letter, and supplied additional materials as requested. The materials submitted were well organized and notably aidM d- the development of the examinations. The examiners noted that a number of the Job Performance Measures (JPMs) submitted had not been revised to reflect changes in plant- procedures or -

modifications. The license; stated that JPMs are only revised when they are used in a current training cycle, Operator Performance The overall performance of the candidates was satisfactor However, it was' noted that the candidates had difficulties in performing several of the JPMs.

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Report Details 2 JPN 6A, Failure of Ste. Flow Channel FT-474, presented the candidates with a scenario in which they had to respond to the failure of a selected steam-line flow transmitter failing lo The failure of this transmitter caused the indicated steam flow to fail downscale. The downscale failure of the steam-line flow transmitter resulted in a steam flow / feed flow mismatch. The feedwater system reduced feed flow in an attempt to match steam flow. The reduced feed flow caused a reduction in S/G leve Level was reduced in all S/G due to the reduction in total feed flo The candidates responded to the event with AOP 401.3, Steam Flow -

Feedwater Flow Protection Channel f ailur This procedure did not adequately address the events of the scenario. The JPM was given to eight candidates. Two of the candidates allowed the reactor to trip, and several of the candidates wanted to trip the reactor because they believed that some type of line break had occurre The failure of numerous candidates to understand and properly analyze the plant response to a steam flow transmitter failing downscale is a generic knowledge deficienc The JPM r.nd the adequacy of A0P 401.3, addressed in the JPM, were discussed with the licensee. The licensee agreed that A0P 40 could be enhanced to give operators better guidance. Additional details concerning the adequacy of A0P 401.3 are discussed in paragraph "3.d.2". Procedures During the course of the examination several procedural problems were encountered by the examination team. Several of the problems identified are discussed belo (1) E0P-2.2, Transfer to Cold leg Recirculation Step seven directs the operator to " Switch the CCW Pump supplying the non-essential loads to fast speed". This requires the operator to stop the running pump, have the Intermediate Building Operator shift the pump to fast speed,.

and then restart the pump. However, the candidates 1)

shifted the standby pump to fast speed, 2) started the standby pump, and 3) secured the running pump last. The candidates stated that they had been trained to perform this step in this manner because cooling water flow to the nonessential loads is not interrupted. The procedure.does not agree with the actual practiced method employed by the operator .

f Report Details 3 (2) A0P 401.3, Steam Flow - Feedwater Flow Protection Channel Failure This procedure did not provide sufficient guidance for any failure mode resulting in a controlling steam flow signal failing low. At high power levels the operators might not be able to restore S/G 1evels to within i 5 percent of programmed level due to a reduction in the Main Feed Pump (MFP) Master Speed Controller setpoint causing a reduction

'in MFP spee The procedure directed the operators to take manual control of the affected S/G's Flow Control Valve and restore leve However, due to the reduction in the MFP's speed, steam flow could still exceed feed flow even with the FCVs in the full open position. In this situation the operators could not-restore level to within the program band. The procedure did not direct the operators to select the operable steam flow channel until level was restored to within 5 percent of the programmed level. The procedure also'did not direct the operators to take manual control of the MFP Master Speed Controller if necessary to adjust MFP speed. If the procedure is followed verbatim in this situation, the reactor will trip on a low S/G-level conditio During the performance of a JPM in which FT-474 failed low, two of eight candidates allowed the reactor to trip on a-low S/G level condition. Several candidates wanted to trip the reactor because they thought that a feed line break of some sort existed. The candidates that were able to avoid a reactor trip either took manual control.of the MFP Master .-

Speed Controller, or disregarded the. procedural guidance and shifted the input signal into the Steam Generator Water Level Control system to the operable channel, without regard to S/G leve (3) A0P 600.1, Control Room Evacuation Step 3.3.d directed the operators to manually trip the turbine from the Turbine Front Standard if the reactor could not be tripped from the Main Control Board (MCB). The candidates were observed to have tripped the turbine from the MCB in lieu of this' step during a Control Room Evacuation JPM in which the Manual Reactor Trip Switches failed to operat In addition, several-candidates commenced an emergency boration from the MCB in response to the ATWS. This step is not contained in the immediate actions of the procedure. The candidates were not certain as to whether they should perform A0P-600.1 or enter E0P -

13.0. It was apparent to the examiners that the candidates were confused as to the rules of usage in this situatio .

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I Report Octails 4 Material Condition Of The Plant The licensee's attention to housekeeping was noteworthy. All areas of the plant observed were clean and well organized. In most irstances plant equipment was adequately labeled and access-ibl The examiners did notice components which had either missing labels or were improperly labeled. The source range power switch on Control Room Evacuation Panel "B" had no label. Reactor Coolant Pump A Seal Supply Isolation Valve (8102A) identification label (metal tag) was mounted backwards on the valve bod . Exit Meeting At the conclusion of the site visit, the examiners met with those ,

representatives of the plant staff indicated in paragraph 1, to discuss the results of the examinations and inspection findings. Support offered by the training staff was both helpful and appreciated. The licensee did not identify as proprietary any material provided to or reviewed by the examiner The licensee's cooperation and assistance during the preparation and administration of the examination was noted and appreciate _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ __ .

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ENCLOSURE 2 SIMtJLATOR FIDEllTY REPOR_T Facility Licensee: South Carolina Electric and Gas Company Facility Name: V. C. Summer Nuclear Plant Facility Docket No.: 50-395 Operating Tests Administered On: December 7 - 11, 1992 This form is to be used only to report observations. These observations do not constitute, in and of themselves, audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the- simulation facility other than to provide information which may be used in future evaluations. No licensee action is required solely in response to these observation During the conduct of the simulator portion of the operating tests, no simulator deficiencies were identified.

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U. S.-NUCLEAR REGULATORY-COMMISSION SITE SPECIFIC EXAMINATION

' SENIOR OPERATOR LICENSE REGION 2 CANDIDATE'S NAME:

FACILITY: V. C. Summer 1

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REACTOR TYPE: PWR-WEC3 -

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DATE ADMINISTERED: 92/12/07 INSTRUCTIONS TO CANDIDATE:

-Use the answer sheets _provided to document your answers. Staple thisLcover 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 60%. Examination papers w'ill'be picked up four (4) hoursJafter the examination start CANDIDATE'S TEST VALUE SCORE %

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100.00 TOTALS

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FINAL GRADE All" work done on this examination is my ow I have neither given nor-

-received ai Candidate's Signature

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SENZOR REACTOR" OPERATOR Page 2-ANSWER SHEET

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Multiple Choice ~ -(Circle or X your choice)-

If.you change your answer,~ write your selection in the blan MULTIPLE CHOICE 023 a b .c ~d-001 a b c d 024 a b c d-

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003 a b c d 026 a b c d 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 .,

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A N S W E:R_ S-H E-E T-

-Multiple Choice (Circle-or X your choice).

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HIf you change _your answer,_ write your selection in the blank.-

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047 a b c d 070 a b c d U

'048 a b c d 071 a b- -c d 049 a b c d 072 a b c d J050- a b c d 073 a b c d

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Mult1ple Choice (Circle or X your choice)

If.you change your answer, write your selection in the blan a b c d 093 a b c d 094 a b c d 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-100 a b c d _

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Page 5 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS j During the administration of this examination the following rules apply:

1. Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . After the examination has been-completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not ,

received or given assistance-in completing the examination. This must be *

done after you complete the examinatio .

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S. Restroom trips are to be limited and only one applicant at a time may f

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leav You must avoid all contacts with anyone outside the examination !

room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction !

5. Print your name in the blank provided in the upper right-hand corner of

the examination cover shoot and each answer shee ,

6. Mark your answers on the answer sheet provide USE ONLY THE PAPER PROVIDED ,

AND DO NOT WRITE ON THE BACK SIDE OF THE PAG ,

7. Before you turn in your examination, consecutively number each answer sheet, !

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including any additional pages iraserted when writing your answers on the examination question pag l 8. Use abbreviations only if they are commonly used in facility literature.

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Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect ar;swe Write it ou . The point value for each question is indtrated in parentheses after the questio . Show all calculations, methods, or assumptions used to obtain an answer to any-short answer question . . Partial credit may be given except on multipio cnoice questions.= Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN .. Proportional grading willibe applie Any additional wrong information that is provided may count against you. F example, if a question-is worth one point and asks for four responst each of.which is worth 0,25 points, and you give five responses, each v. your responses will-be worth- ,

0.20 point If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80-instead of 1.00 even though you got the four correct answer ;

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13. If-the-intent of a question is unclear, ask~ questions of the examinor onl r

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Page 6 14. When turning in your examination, assemble the completed examination with examination questions, examinat' ion aids and answer sheet In addition, turn in all scrap paper.

15. Ensure all information you wish to have evaluated as part of your answer is on your answer sheet. Scrap paper will be disposed of immediately following the examination.

16. To pass the examination, you must achieve a grade of 80% or greater, 17. There is a time limit of four (4) hours for completion of the examination.

18. When you are done and have turned in your examination, leave the examination arek (EXAMINER WILL DEFINE THE AREA). If you are found in this area while-the examination is still in progress, your license may be denied or revoke !

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SENIOR REACTOR OPERATOR- Page 7 OUESTION: 001 (1.00)

WHICH ONE (1) of the following 's the source of the Tref signal? It is calculated from auctioneered high T-av It is calculated from NI normalized power level, It is calculated from averaged T-ho It is calculated from turbine first stage pressur _

QUESTION: 002 (1.00)

WHICH-ONE (1) of the following rod control system conditions will cause a NON URGENT failure alarm at the power cabinet?

- Loose circuit card

- Loss of multiplexer signal Loss of the main power supply Control thyrister fault QUESTION: 003 (1.00)

WHICH ONE (1) of the following is the reason that all the RCP breakers open on underfrequency?- To-protect the RCP anti-rotation device from damage due to abnormal coastdown, To preserve the RCP flywheel kinetic energy, To avoid water hammer transients in the RCS induced by rapid RCP speed chang To reduce the probability of a stress induced RCP sheared

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OUESTION: 004 (1.00)

The following plant conditions exist:

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The reactor is at 18% powe L

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"A" RCP has the following parameters: !

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Vibration of 3.5 mil '

Seal inlet temperature of 130 degrees '

- Radial bearing temperature of 228 degrees Seal injection flow is 9 gp !

- #1 seal delta-p is 187 psi #1 seal leakoff flow of 5.8 gp Normal thermal barrier cooling flow and temperatur .

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- VCT pressure is 20 psi WHICH ONE (1) of the following actions is the FIRST corrective action per applicable plant procedures? Trip the Reactor, Trip the "A" RC Increase VCT pressur Close the "A" RCP #1 seal leakoff isolation valv i

= QUESTION: 005 (1.00) ,

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-WHICH ONE (1) of the following is the reason for. requiring RCS '

' pressure to be above 325 psig. prior to STARTING a Reactor Coolant-pump?- -- Guarantees that the delta pressure across theinumber:one (1) seal is greater than 220 psi , Guarantees flow through the number two (2) sea Assures not positive suction pressure (NPSH).to the RC Assures stability of Reactor Vessel components when the pump is STARTED.-

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f QUESTION: 006 (1.00)

The following plant conditions exist:

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The Unit is in MODE RCS temperature is-210 degrees A bubble exists in the Pressurize ;

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RCP "A" is-RUNNIN The Component Cooling Water System supply to the Reactor '

Coolant Pumps (RCPs) has been isolated for repair of a lea WHICH ONE (1) of the following is the reason Seal Water injection is REQUIRED _ to be in service to the Reactor Coolant pumps (RCPs)

per SOP-101, " Reactor Coolant System"? The RCS is FUL An RCP is RUNNIN A bubble-is established in the Pressurize RCS temperature is 210 degrees QUESTION: 007 (1.00)

The following plant conditions exist:

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The Volume Control Tank (VCT) level is at 39%.

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Automatic make-up-is:in progres _

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A leak develops in the reference leg associated with the automatic level controller (LT-112).

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Assume no operator actio WHICH ONE (1) of the following describes the'FIRST system response? LCV-115A begins-to diver ' Automatic makeup to_the VCT will-STO Indicated VCT (LT-112) level will-DECREAS ^ . Charging pumps will automatically shift suction to the RWS ,

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QUESTION .008 (1.00) *

WHICH ONE (1) of the.following leaks could result in a DILJTION of the'RCSY .; Regenerative Heat Exchanger leak $ Letdown Heat Exchanger leak Reactor. Coolant Drain Tank Heat Exchanger leak Seal Water Heat Exchanger leak

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3-QUESTION: 009 (1.00)

The following plant conditions exist:

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The reactor is in MODE It has been determined that BOTH SI trains automatic actuation logic and actuation relays did NOT meet the-surveillance acceptance. criteria for the test previously conducted in MODE 5, but the manual-initiations passed the surveillance WHICH ONE (1) of the following actions should be taken? Stay in MODE 4 and do NOT' allow transition to MODE J Continue with the plant startup, no. restraints apply, Enter and comply with' Technical Specification l3, Within i hour verify the actuation relays are in the required state for the existing plant conditio ,

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-SENIOR REACTOR OPERATOR Page 11 QUESTION: 010 (1.00)

WHICH ONE (1) of the following systems receives an input from the Power Range Nuclear Instrumentation? Rod Position Indication System Audio Count Rate System Steam Generator Water Level Control System High Flux At Shutdown Alarm System QUESTION: 011 (1.00)

Each of the following is an input to the Core Subcooling Monitor-EXCEPT: Wide range hot leg RTD Wide range cold leg RTD Narrow range Pressurizer pressure Average Core Exit Thermocouple temperature

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QUESTION: 012 (1.00) .

WHICH-ONE (1) of the following co'nditions will result in CLOSING all Feedwater Isolation valves? Feedwater temperature is 220 degrees All Steam Generators indicate 13% leve Feedwater Forward Flushing valve-(XVG-1689A) is 50% OPE HIGH-HIGH Reactor Building ~ sump level.

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QUESTION: 013 (1.00)

WHICH ONE (1) of the following conditions will AUTOMATICALLY start a Motor Driven Emergency Feedwater pump with the reactor at 50%

power? "

A" Steam Generator level is 15%.

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A", "B", AND "C" Steam Generator levels are 18%.

r Zero (0) Volts on Bus 1EA , Trip of "A", "B", OR "C" Main Feedwater pumps QUESTION: 014 (1.00)

The following plant conditions exist:-

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The Unit has tripped from 100% powe Emergency Feedwater flow (EFW) CANNOT be establishe Steam Generator levels indicate 5% wide range leve Steam Generator pressures indicate 500 psi ,

-- All Reactor Coolant pumps have been TRIPPE WHICH ONE (1) of the following methods of RCS cooling is IMMEDIATELY available? Natural circulation cooling established by low pressure '

steaming through the Steam Generator Power-Operated Relief valves (PORVs), Natural circulation cooling established'by Condensate System makeup to the_ Steam Generators, c ._ Natural circulation cooling established by Main.Feedwater makeup to the Steam Generators, _ Bleed and feed cooling of the RCS using the CVCS System and_ two (2) _ opened Pressurizer Power-Operated Relief valves (PORVs).

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SENIOR REACTOR' OPERATOR Page 13-QUESTION: 015 (1.00)

The following plant conditions exist:

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A waste gas release was in progres Waste Gas Discharge Radiation Monitor, RM-A10, is in servic HCV00014, Vent Stack Hand Control Valve has tripped shu A new analysis and release permit have been requeste WHICH ONE (1) of the following is the MINIMUM action necessary for the waste gas release to continue? HCV-00014, Vent Stack. Hand Control Valve, must-be reset

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locally at the valv HCV-014, Waste Gas Discharge Control Valve controller must be taken to ZERO then the valve re-opene HCV-014, Waste Gas Discharge Control Valve selector switch must be cycled to closed then re-opene HCV-014, Waste Gas Discharge Control Valve selector switch must be cycled to CLOSED then re-opened AND HCV-0014, Waste Gas Discharge Control Valve controller must be taken to ZER QUESTION: 016 (1.00)

WHICH.ONE (1) of the following AUTOMATIC actions result from a HIGH radiation. condition on Liquid Waste-Effluent Monitor, RM-L5? Liquid radionctive-waste discharge control valve (RCV-018) CLOSE Waste. Monitor Tank pumps TRIP, Liquid waste flow is DIVERTED to the Nuclear Blowdown Monitor tank,

' ' Liquid Effluent To Fairfield Penstocks' valve-(PVD-6910).

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QUESTION: 017 (1.00)

'. WHICH ONE (1) of the following combinations of oxygen and hydrogen REQUIRES immediate suspension of addition of waste gas and a .'

reduction of the oxygen concentration in the waste gas holdup system, per Technical Specifications 3.11.2, " Gaseous Effluent"?

Hydrogen Oxygen % 7% % 5% % 3 %- % 1%

-QUESTION: 018 (1.00)

The following plant conditions exist:

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Control Rods are being withdrawn for startu A stable startup rate of one decade per minute is being maintaine Tavs is 545 degrees WHICH ONE (1) of the following MINIMUM actions is required by Technical Specification 3.1.1.4, " Minimum Temperature For Criticality"?

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~ Restore Tavg to greater than or equal _to 547 degrees F within one hour or be in HOT STANDBY within-the next:

hour, Restore Tavg to greater-than or equal to 551 degrees F within one hour or be in HOT-STANDBY within=the next hou :. Restore Tavg to greater.than or equal to 547 degrees _F within 15 minutes or-be-in HOT STANDBY within the next 15-minutes.

l l Restore Tavgoto greater-than:or equal to 551; degrees F within 15 minutes'or be in~ HOT STANDBY'within-the next 15-minutes.

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' SENIOR REACTOR OPERATOR Page 15 QUESTION: 019 (1.00)

The following plant conditions exist:

- The Unit has trippe Offsite power has been los Upper range Reactor Vessel Level Indication System (RVLIS) indicates 65%.

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Narrow range RVLIS indicates 50%.

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Wide range RVLIS indicates 15%.

WHICH ONE'(1) of the following is the current level / condition in the Reactor Vessel? - % % % (full) but voids are presen Four (4) inches above the Hot Legs.

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QUESTION: 020 (1.00)

l The-following plant conditions exist: l

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The Unit has tripped from 100% powe l

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Safety Injection has actuate l

- Train "A" and "B" Safety Injection Reset switches have- I been placed in the RESET positio RWST level is 20%. '

- Reactor Building sump level indicates two (2) fee AUTOMATIC swap over of RHR suction to the Reactor Building sump has NOT occurre WHICH ONE (1) of the following is the reason AUTOMATIC swap over of RHR suction to the Reactor Building sump has NOT occurred? RWST level must be less than eighteen (18) percen Train "A" and "B" Safety Injection has been RESE ' NORMAL / RESET switches-are in the RESET positio Reactor Building sump level must be five (5) fee QUESTION: 021 (1.00)

WHICH ONE (1) of the following "A" Accumulator parameters needs to be corrected prior to declaring the "A" Accumulator OPERABLE while the reactor is operating at 100% power? . ,

- ' Volume is 7500 gallon Boric acid concentration is 2015 pp f Pressure is 650 ps19 Isolation valve (MVG-8808) is OPEN and the RED indicating ~

light is O . T I

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SENIOR REACTOR OPERATOR Page 17 s :

QUESTION: 022 (1.00) j

-To WHICH ONE (1) of-the following systems / locations is an ECCS Accumulator vented when lowering pressure? Reactor Building atmosphere

' Reactor Coolant Drain Tank (RCDT) Pressurizer Relief Tank (PRT) [ Reactor Building Purge System

,-

,

h QUESTION: 023 (1.00)

.

The following plant conditions exist The reactor is at 100% powe '

,

-

Pressurizer heaters are OF Normal pressurizer spray valves are' MODULATIN *

-

Pressurizer-PORVs-are CLOSE WHICH ONE (1) of the'following RCS pressures is appropriate for the- !

above conditions? psig , psig

, psig_ psig

..+

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.

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-!

-

_

.,

SENIOR REACTOR OPERATOR Page la QUESTION: 024 (1.00)

The following plant condition exists:-

-

A MANUAL reactor trip has been initiated by use of the spring loaded control switche The red indicating lights below the control switches are DE-ENERGIZE The green indicating lights below the control switches-are ENERGIZE WHICH ONE (1) of the following conditions exists?

_ The shunt trip coil is DE-ENERGIZED and all Reactor Trip breakers are CLOSE The undervoltage coil i~s ENERGIZED and all Reactor Trip breakers are OPE _ The shunt trip coil is ENERGIZED and all-Reactor Trip breakers are CLOSE The undervoltage trip coil is DE-ENERGIZED and all Reactor Trip breakers are OPE QUESTION: 025 (1.00)

WHICH ONE (1) of the following is the system accuracy for the Rod

~

Position Indication System when the COIL _B output is_ lost? +10, -10 steps +4, -4 steps +4, -10 steps , +10 steps i

. . . . . . .

l SENIOR REACTOR OPERATOR Page 19 j

!

QUESTION: 026 (1.00) f Each one (1) of the following is a concern when level is reduced i BELOW minimum in the Spent Fuel Cooling System EXCEPT: Minimize radiation levels at the pool surfac Minimize the heat load on the Spent Fuel Pool Heat I Exchanger (s). . '

Minimize velocity of a dropped spent fuel' shipping cas . Minimize the amount of iodine released f rom a- ruptured l fuel assembl t QUESTION: 027 (1.00)

WHICH ONE (1) of the following is the basis for maintaining a-MINIMUM of 9.5 feet of water between operators on the SFP bridge and a fuel assembly being moved? I Level is sufficient to remove 99% of the assumed 10%

-iodine gap activity released from a ruptured irradiated fuel assembl :

'

- Level is sufficient to limit the maximum dose rate at the pool surface to 2.5 mR/HR during transfer operation Level _is sufficient _.to limit the maximum velocity of.the spent fuel shipping cask to an impact velocity of 44 ft/se ' Level is sufficient to' ensure the purification skimming t.. "':h is covered to remove any particulate matter released during fuel movemen ,

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SENIOR REACTOR OPERATOR Page 20 QUESTION: 028 (1.00)

WHICH ONE (1) of the following describes the FAILURE. mode of the Steam Generator Power-Operated Relief valves, PORV-2000, 2010, and 2020, on a loss of the selector control signal from the Steamline Power Relief Mode (PWR RLP/ AUTO)

switch? OPEN CLOSED , Steam Dump Mode Overpressure Mode

QUESTION: 029 (1.00)

The following plant conditions exist:

- An SI occurs with a loss of off-site powe The Emergency Diesel Generators start and commence '

loadin Reactor Building Spray signal activates at 25 seconds after Emergency Diesel Generator start WHICH ONE (1)- of the following describes the response of the-Containment spray pumps? -;

_ The pumps require a-manual start _following automatic- -!

repositioning of the pump discharge. valve . The pumps start immediately upon Reactor Building Spray j actuation signa The pumps start approximately 35 seconds after Emergency Diesel Generator star z: The pumps will automatically start ONLY after the pump discharge valves are manually repositione !

i

,

_ , - - - - . - - - . . _ _ _ _ _ . - _ . . . - . . . _ - _ . _ . - . -

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SENIOR. REACTOR OPERATOR Page 21 QUESTION: 030 (1.00)

WHICH ONE (1) of the following AUTOMATIC actions results from a HIGH radiation condition on Component Cooling Water radiation monitor, RM-L2A? Reactor Building isolation valves, MVG-9605 and MVG 9606, CLOS Reactor Building isolation valves, MVG-9625 and MVG-9626, CLO6 Component Coolir.3 Water surge tank vent valve, PVV-7096, CLOSE Reactor Building isolation valve, MVG-9600, CLOSE QUESTION: 031 (1.00)

-The following plant conditions exist:

-

The plant is operating at 100% powe The Circulating Water System is in its normal-full power lineu Circulating Water inlet temperature is 64 degrees WHICH ONE (1) of the following occurs if ONE (1) Circulating Water pump TRIPS?

a .- The operating pumps RUN OUT.- A net loss of electrical generating capacity will occu AUTOMATIC closure of Discharge Isolation valve MVA-807 Circulating Water-discharge temperature limits are VIOLATED.

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-. _ - . _ ._ __ _ _ _ .

_ _ ___ _ _ _.___ - _____ _______ _ _ _ _ _ _ _ _ _ _ .

SENIOR' REACTOR OPERATOR Page 22 QUESTION: 032 (1.00)

.

The following plant conditions exist:

- A fire has occurred requiring flow from either the electric fire pump or the diesel fire pum The electric fire pump has failed to star .

-

IPS 4911, Diesel Fire Pump Discharge Pressure Switch, is INOPERABL WHICH ONE (1) of the following actions will start the diesel fire pump from the local control panel? Place the mode selector switch in the TEST position to-open the discharge drain valv Place the mode selector switch in OFF/ RESET and depress the START pushbutton, Place the mode selector switch in MAN 1 and depress the START pushbutto Ensure the mode selector switch is in OFF/ RESET, then cycle the feeder breaker to XSW-1C QUESTION: 033 (1.00)

WHICH ONE (1) of the1following would-occur on a tube leak in the Component Cooling Water (CCW) Heat Exchanger?

l Automatic CCW system makeup from the Reactor Makeup Wate system would occur, providing the necessary NPSH to the CCW pump CCW system liquid inventory would increase, thus-increasing the CCW flowrate to components cooled by th CCW~ system, CCW would leak into the Service Water (SW) System, potentially contaminating the SW' syste . CCW surge tank level wouldLincrease, which would cause the-vent valve.to lift due_'to. increasing CCW system pressur I r w e en ,L- * ww .-m - n A . - r+ w -ows , 1sY -

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SENIOR REACTOR OPERATOR Page 23 QUESTION: 034 (1.00)

Wi!ICH ONE (1) of the following containment hydrogen concentrations is the MM.IMUM allowable prior to placing Hydrogen Recombiners in service while performing EOP-14.0, " Response To Inadequate Core Cooling"? .5% .5% .5%

_ .5%

,

[t QUESTION: 035 (1.00)

'

. .

The following plant conditions exist:

- A Loss of Coolant Accident (LOCA) has c; curre The liydrogen Recombiners are INOPERABL W11ICll ONE (1) of the following is the preferred metbod to REDUCE hydrogen concentration in the Reactor Building per EOP-14.0,

" Response To Inadequate Core Cooling"? Add liydrazine to the Reactor Building Spray, START the Reactor Building Alternate Purge Syste Vent the hydrogen to the Atmospher Purge the Reactor Building with nitrogen ga _ _ _ _ _ - _ - _ _ _ - _ _ _ _ _ - - .

SENIOR REACTOR OPERATOR Page 24 QUESTION: 036 (1.00)

WHICH ONE (1) of the following conditions is indicated by the red OVERLOAD BYPASS light on the Reactor Cavity Manipulator Crane control panel being LIT 7 The 2700 POUND OVERLOAD interlock is BYPASSE The 3200 POUND OVERLOAD interlock is BYPASSE The 2700 POUND OVERLOAD interlock is NOT BYPASSE The 3200 POUND OVERLOAD interlock is NOT BYPASSED QUESTION: 037 (1.00)

WHICH ONE (1) of the following describes Main Turbine operation when the Main Generator output breaker is CLOSED? Steam flow is controlled by the Stop valves to control turbine speed, Steam flow is controlled by the Stop valves to control generator loa Steam flow is controlled by the Control valves to control turbine spee d.- Steam flow is controlled by the Control valves to control generator load.,

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SENIOR REACTOR OPERAIOR Page 25 f

1

t

'

QUESTION: 038 (1. 00)

!

WHICH ONE (1) of the following supplies cooling water PIRECTLY to i the Reactor Building cooling units during a Loss of teclant _

Accident (LOCA)?' ;

. Component Cooling Water pumps : Component Cooling Water booster pumps ,

i Service Water pumps ; Service Water booster pumps

!

I

'

QUESTION: 039 (1.00) :

>

WHICH ONE (1) of the following supplies air to the_ Breathing Air System? The Reactor Building Air System through Service Air !

Isolation valve IPV-832 b.- The Instrument Air centrifugal air compressor through IA Backup System Sup Hdr Isolation valve IPV XVB'263 , The Service Air centrifugal air compressor through Service Air Isolation valve IPV-832 , .The Supplemental air co'mpressor through IA Backup System f Sup Hdr Isolation valve XVB-263 ,

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T k

k

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. _ _ _ _ . . - _ _ _ . - . _ . _ _ . _ _ _ _ _ _ . _ . . _ _ _ . _. _ r

!

I SENIOR REACTOR OPERATOR Page 26 ;

!

,

QUESTION: 040 (1.00) .

!

The following plant conditions exist: !

r

- The Reactor has tripped from 100% power due to a small !

break loss of coolant accident (LOCA). l

-

A Safety Injection Signal has been generate :

- All electrical buses are energized as designe WHICH ONE (1) of the following air compressors is supplying the

Instrument Air System? Supp Instr Air Compressor (XAC 12) powered from bus XSW1A1 ,

!

' Supp Instr Air Compressor (XAC-12) powered from bus

XSW1DB1 Instr Air Comp "B" (XAC-3B) powered from XSW1A1 . Instr Air Comp "B" (XAC 3B) powered from XSW1DB1

QUESTION: 041 (1.00)

The following plant conditions exist -

-

- The reactor is at 80% powe Control rods are in AUTOMATIC.- .

-- A Tave-Tref-deviation of-7 degrees F exist "

- Control rods are NOT movin WHICH ONE (1) of the following actions should be taken? Adjust Tave to within 5 degrees of Tref by adjusting .

-turbine load and do NOT attempt to move rods until troubleshooting-is complete, Place the rod control in manual and verify. control _ rod operability by moving rods in.5 steps, then-cut 5-step _

, _ Trip the reactoriandienter EOP-1.0, " Reactor Trip / Safety-Injection Actuation". Place rod control in manual and adjust rods to match Tave-to within 1.5 degree-F of Tre ,

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SENIOR REACTOR OPERATOR Page-27 QUESTION: 042 (1.00)

The following plant conditions exists

- The reactor is at 80% powe control rods were found misaligned from their group step counter by greater than 12 step The rods were determined to be TRIPPABLE, but otherwise IMMOVABL WHICH ONE (1) of the following actions should be taken? The rods must be declared inoperable and the SHUTDOWN -

MARGIN recalculated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , THERMAL POWER must be reduced to less than or equal to 75% within the next hour, c Action must be taken within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place-the unit in a MODE in which the specificttion does not apply, The remainder of the rods in the group with the inoperable. rods must be aligned within +/- 12 steps of the inoperable rods within i hou _ _

_ _ -- . ,.

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_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ ___ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ .

,

SENIOR REACTOR OPERATO Page 30 i

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t

'

QUESTION: 043 (1.00)

The following Reactor Coolant Pump conditions exists

-

An inadvertent valve manipulation has resulted in loss of Component Cooling Water (CCW) to Reactor Coolant Pump 6 (RCP) "A". l

-

Delta Pressure across the No. 1 seal 10 250 paid, i

-

Motor bearing. temperature is 200 degrees RCP shaft peak to peak composite vibration is 15 mils and stabl RCP frame composite vibration is 4 mils and stable.

. WHICH ONE (1) of the following is the reason the "A" RCP should be TRIPPED immediately? RCP frame vibration is HIG Delta pressure across the No. 1 seal is HIG , Motor bearing temperature is HIG ~ RCP shaft vibration is HIG .

QUESTION: 044 (1.00)

WHICH ONE (1) of the following conditions requires EMERGENCY BORATION.per AOP-106.1, " Emergency Boration"? " Shutdown Margin of 2.00% delta k/k while in MODE 1 One (1) shutdown bank rod remaining out of the core following a reactor tri A 100% Load Rejection from the Main Turbine' Generator L ppm boron concentration in the RCS while.in MODE 6 l

l U

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-

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SENIOR REACTOR OPERATOR Page 29

!

- QUESTION: 045 (1.' 0 0 )

WHICH ONE (1) of the following is an ALTERNATE method to accomplish '

EMERGENCY BORATION should the normal flow path be unavailable while i operating at 100% power per AOP-106.1, " Emergency Boration"?- [ Charging pump suction from the RWS Boric Acid tank to the VCT via gravity fee j Boric Acid Transfer pump discharge to Charging pump

-

,

'

discharg Gravity feed from the RWST to the VC i t

QUESTION: 046 (1.00) i The following plant conditions exist:

- The Unit is operating at 100% powe Pressurizer pressure control transmitter PT-444 has failed HIG !

WHICH ONE (1) of the following IMMEDIATE actions should be taken-per AOP-401.5, " Pressurizer Pressure Control Channel Failure"?

, Close PORV'444 Close PORV 444B Block valv E Close PORV 445 Close PORV 445B Block valv i

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r* P ut.UMN P True pp-#w----

- .--.- .. . - . . . - - - - . - . - . . . . _ - . - . - . - - . . . , _ _

,

SENIOR REACTOR OPERATOR Page 30 ;

,

QUESTION: 047 (1.00) i WHICH ONE (1) of the following is an example of a Steam Generator depressurizing in an UNCONTROLLED manner? (Assume all plant !

systems operate as designed.) A steam break exists on the inlet to the llP Turbine l causing a rapid decrease in Steam Generator pressur ; An Atmospheric Steam Dump has failed OPEN causing a rapid :

decrease in Steam Generator pressur A steam break exists on the steam supply line to the EFW turbine (between the Main Steamline and XVG-2802A)

causing a rapid decrease in Steam Generator pressure, A feedwater flow controller failure has resulted in over feeding the "A" Steam Generator causing a rapid decrease r in Steam Generator pressur l

,

QUESTION: 048 (1.00) i WHICH ONE (1) of the following satisfies criteria for MANUALLY actuating Safety Injection per EOP-1.0, " Reactor Trip / Safety Injection Actuation"?

Pressurizer pressure is 1900 psig and DECREASIN Reactor Building pressure is 3.3 psig and STABL i l

Steam Generator "B" pressure is 760 psig and' Steam Generator "C" pressure'is 850 psi ,

, Pressurizer level is 4% and DECREASIN .;

.

I

I SENIOR REACTOR OPERATOR Page 31 QUESTION: 049 (1.00)

The following plant conditions exist:

-

The reactor is at 100% powe Megawatt output begins a DECREASING tren Condenser pressure is 20 in Hg. Absolute and INCREASIN WHICH ONE (1) of the following is the FIRST action required? Trip the Reactor, the turbine, and enter EOP- START the standby Main Condenser Vacuum pump and REDUCE turbine load to 70% at 5% per minut Notify the Technical Staff to trend operating time at increased backpressur START additional Circulating Water pumps and observe condenser pressure trend to determine if additional ,

measures are necessar ;

QUESTION: 050 (1.00)

The following plant conditions exist:

- Operators are attempting a MANUAL start of Diesel Generator during a Station Blackou The speed signal generator output failed HIGH when the DG reached ten (10) RP The Low Speed Relay (115 RPM) has now LOCKED OUT power from both solenoid operated start valve WHICH ONE (1) of the following methods can be used to START D/G A? PRESS and HOLD the EMERGENCY START pushbutton, DEPRESS the GEN RELAY RESET pushbutto 'l MANUALLY override the main air start valv ,

d .- DISCONNECT the leads to the shutdown solenoids, m .._. _. _ _ _ _ _ - . . ~...~. _ _ . _ _. _-.. _ .__,_ _ _c

_ . _ .

SENIOR REACTOR OPERATOR Page 32 QUESTION: 051 (1.00)

The following plant conditions exist:

-

The reactor was operating at 100% powe The Unit tripped and a station blackout occurre EOP-6.0, " Loss of All ESP AC Power", was entered and Step 1 " Verify Reactor Trip" could NOT be verified in the Control Room due to no available indicatio A manual trip of the Reactor _wca attempted using BOTH Reactor Trip Switche WHICH ONE (1) of the following is the NEXT required action? - LOCALLY verify Reactor Trip and Bypass breakers are open, MA!IUALLY insert control rods at 48-steps per minut Transition to EOP-13, " Response to Abnormal Nuclear Powe Generation". LOCALLY restore 125 VDC power to the DC. vital distribution-panels in the ESF battery room QUESTION: 052 (1.00)

The following plant conditions exist:

-

A normal electrical all'gnment exists when AC power is-LOST to Vital Bus 1D WHICH ONE (1) of the following will provide power to Safeguards Bus APN-5903? V Vital AC Bus APN-1FB Inverter XIT-5901

. Inverter XIT-5908 "B" Battery Bus DPN-1HB

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,

SEWIOR REACTOR OPERATOR - Page 33 QUESTION: 053 (1.00)

i The following plant. conditions exist

-

A fire has been burning out of control 4.n the area surrounding Diesel Generator DGA-for i rty (30) minutes, q

<

i

,

WHICH ONE (1) of the'following is the Emergency Classification level for this event? ! Unusual Event

, Alert Site Area Emergency General Emergency QUESTION: 054 (1.00)

WHICH ONE (1) of the following is the-basis for NOT tripping thel _

"A" RCP when the Control Room is; evacuated per-AOP-600.1,." Control Room Evacuation"? Assists in RCS pressure-control-from the~ Control-Room Evacuation Panels (CREPs). ___ Ensures Emergency Boration flow-gets evenly. distributed-

. through the RCS.

.or ' n s in establishing natural circulation cooldown by-M s'.c+ the shutdown of the pumpiuntil control 1is-esicolished from the CREPs.

s - Ensures decay neat removal until CREPs are OPERABL f

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. _ . . . _ . _ _ . - - . . . _ _ _ _ _ , . ~ .._ - -_ . _._ ..~. _

-

_ - _, .. ._ .. _ _ . - . . .

.

SENIOR REACTOR OPERATOR - Page 341

,

-Q ESTION: 055 (1.00)

The following plant conditions exit:

-

The Unit Control room has been evacuate Reactor Coolant Pumps have been TRIPPED and operators _are _

attempting-to verify natural circulation coolin ~

-

RCS pressure is 2000 psi RCS Thot (Th) is_604 degrees F. and. STABL RCS Tcold (Tc) is 540 degrees F. and STABL '

-

Steam Generator pressure is 1000 psig and STABL Each of the following parameters is an indication that natural circulation' cooling has been established per AOP-600,1, " Control ;

-Room Evacuation", EXCEPT: RCS subcooling RCS Thot (Th) RCS Tcold (Tc)

Steam Generator pressure QUESTION: 056 (1. 0 0 ).

Each.of=the following is a Major Action Category of_EOP-17.0,

- " Response to High Reactor Building Pressure" EXCEPT: - Verify containment isolatio . Check for excessive hydrogen concentrationL and determine appropriate actio , Reduce heat' input to containment _by shutting down RCPs

,

and closing primary PORV Verifyfcontainment heat remova .

- -- - -t-,- -w r- --w -

v c

SENIOR REACTOR OPERATOR Page 35 QUESTION: 057 (1.00)

WHICH ONE (1) of the following constitutes a LOSS of containment integrity per Technical Specifications 3/4.6, " Containment Systems"?

a, while at 100% power, an electrician opens the outer Reactor Building airlock door to perform maintenance activities on the CLOSED INOPERABLE inner Reactor Building door without prior approval, While performing an operability test of two normally open, redundant Reactor Building isolation valves at 100% ---

power, one of the valves fails to clos While in MODE 6, the equipment door is closed but held in place by only four (4) bolt The Integrated Containment Leakage Rate Test results indicate a leakage rate of 0.4 of the maximum allowable leakage (La).

_

- -__._____ .

, _ - _ _ _ . - . . _ . .. , - . _ - _.-_ __ -

. . . . . - _ . - _ , . - _ , _ . - . _ _ _ _ . .

SENIOR REACTOR OPERATORD Page_36 QUESTION: 058 4 (1.00)-

.TheLfollowing plant conditions exist:

-

The Unit has tripped from 100% powe Critical Safety Functions are being monitored'with the following indications:

- Intermediate SUR is negativ Pressurizer level is 0%.

-

Pressurizer pressure is O psi RCS Teold (Tc) is 350 degrees Wide range Steam' Generator levels are 20 percen Steam Generator pressures are 1240 psi Reactor Building pressure is 59 psi cmergency Feedwater flow to each Steam Generator _is 150 gpm.

1 WHICH ONE (1) of the following-Critical Safety Functions.must be corrected FIRST per EOP-12, " Monitoring Of Critical Safety-Functions"? Heat Sink Inventory Containment Subcriticality

.

b

, . . - . . . . . . . . - .-. -__ . .~ -. . . --- SENIOR' REACTOR OPERATOR ,

Page 37'

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LQUESTION: 059 (1.00).

~

WHICH ONEl (1). of' the following is the reason for feeding ONLY one steam; generator 11n.the event that ALL steam generators have dried out_during a loss of heat 1 sink event?

Prevents excess thermal stress in-multiple steam generator b.- Prevents excess thermal stress in the reactor vessel, Ensures adequate feedwater inventory 1for complete .

restoration of normal coolin Ensures that only one reactor coolant pump must be returned to servic QUESTION: 060 (1.00)

The following plant conditions exist:

-

Following a LOCA the. core exit thermocouples. indicate 710'

degrees RCS pressure is 350-psi RVLIS indicates 30%.

'WHICH ONE (1) of the following conclusions can be drawn from the above information?

a, Limited core' melting has occurre Core'uncovery:is occurrin Core cooling mode is reflu Core exit thermocouples.have failed.

.

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. _ ._ _- _ . . _ _ . . . . . . . . _ __ .

SENIOR REACTOR OPERATOR Page[38-

' QUESTION: 061 -(1.00)

The=-following plant conditions exist:

-

The reactor has been at 100% power for 30 day Chemistry reports ~that RCS activity is 1.5 microCuries per_ gram DOSE EQUIVALENT I-13 WHICH ONE (1) of the following actions will reduce RCS activity?

' Vent the Volume Control Tank (VCT) to the Waste Gas System, Place cation demineralizer in service and maximize letdow Maximize letdown through the mixed bed demineralizer, Reduce letdown to minimum and establish an RCS pn-between 6.5~and 7.5 by chemical injectio .

A QUESTION: 062 (1.00)

,The_following plant conditions-exist:

- . Reactor is 30% increasing 3%/ hou Control rods are in AUTOMATI Bank "D" rods are at 140 steps and stepping out with-N demand signa WHICH ONE (1) .of the following actions are required to be: performed'

FIRST per AOP-403.3, " Continuous Rod Withdrawal"?! > ' Place the_ ROD CNTRL-DANK-SEL switch to the CBD_ positio Place the ROD ~CNTRL BANK'SEL switch to the MAN positio TRIP the reactor and enter EOP-1.0, " Reactor. Trip / Safety Injection Actuation".. LEnter EOP-13.0, " Response To. Abnormal Nuclear Power Generation", to ensure the reactor is TRIPPE !

._ . . _ _ _ . .- . .~ _ . _ _ _,

e SENIOR ~ REACTOR OPERATt Page:39 QUESTION: 063 (1.00)' ,

The following plant conditions exist:

-

The Unit is operating at 100%. ,

-

Two (2)- Bank "A" rods- have dropped into the cor WHICH ONE (1) of the following actions is required to be'taken IMMEDIATELY per AOP-403.6, " Dropped Control Rod"?- Place the ROD CNTRL EANK SEL switch to the CBA positio Place the-ROD CNTRL BANK SEL switch to the MAN positio TRIP the reactor and enter-EOP-1.0, " Reactor Trip / Safety Injection Actuation", REDUCE Turbine load and stabilice-Tavg with Tre .i-QUESTION: 064 (1.00)

WHICH ONE (1) of the following is a reason for the time delay between a turbine trip and a generator trip? RCP overspeed can occur-during a major LOCA resulting in o a high enough RCS flowrate to damage the reactor internal ,

t A loss of RCS flow due to a failure of auto bus transfer would not be as serious because.the reactor would have been shutdown for the duration of the time delay, The thermal stresses imposed on the RCP shafts.are far greater if the RCP's are stopped _immediately (due to high RCS delta t) after turbine tri ., The--time' delay-allows circulating transformer currents induced by the trip-transient to stabilize following the plant trip, reducing the probability of damage to the RCP motors.

D

, , -- _ .-

, , . . - - - _ . - - - . .. - -. ..

SENIOR REACTOR ~ OPERATOR Page 40-QUESTION: 065 (1.00)

WHICH-ONE (1) of the following conditions may result from the FAILURE of oneL (1) Combined Intermediate valve to CLOSE on a turbine-trip? , Moisture Separator Reheater (MSR) level would increase,- ,

causing water induction into the LP turbin Expansion of steam in the MSR resulting in an overspeed of the turbin Depressurization of the shell side of the MSR~by blowdown '

to the condenser, causing a MSR tube ruptur Unbalanced torque between the "A" and "B" LP turbines resulting in turbine shaft failure.

.

QUESTION: 066 (1.00)

At WHICH ONE (1) of the following points following a Reactor Trip and Safety Injection shall monitoring of the Critical Safer-Function Status Trees for IMPLEMENTATION begin? Upon transition-from EOP- As soon as possible following the Reactor Trip, h Following verification of Safety Injection flow, Following completion of all Immediate-Actions.

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SENIOR' REACTOR OPERATOR Page 41

.

OUESTION: 067 -(1. 00) ~

'The: following plant conditions exist:

_

--

The Unit has-tripped from-100% power at 0800 hour0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months Safety. Injection Actuat' ion occurred at 0805 hour0.00932 days 0.224 hours 0.00133 weeks 3.063025e-4 months RCS pressure is-1300 psig and DECREASING slowly.-

-

Pressurizer level is 100 %.

-

RCS Thot (Th) is 557 degrees Containment pressure is 1 psig and INCREASING slowl WHICH ONE (1) of the-following events has occurred? Main Steam line break inside the Reactor Building Inadvertent Safety Injection RCS cold leg break Pressurizer vapor space break-QUESTION: 068 (1.00)

WHICH ONE (1) of the following' conditions requires ALL Reacto !

Coolant Pumps (RCPs)-to be TRIPPED? RCS pressure is 500 psig and SI flow has NOT been verifie RCS pressure is 1200 psig.and SI flow HAS been verifie RCS pressure is 500 psig and Reactor Building pressure is 8 psig, RCS. pressure is 1200 psig andLReactor Buildingipressure is 11 psi .

% - , -.-_ - , . , - - . _ , . - ,

-

~.. --. .- . . - . . . . _ . , - .

'SENIORiREACTOR10PERATOR Page142-

- LQUESTION: 069' (1.00)

-The following plant conditions exist:

-

_

-

The reactor has been shutdown for 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months .-

-

The RCS temperature is 140 degree r!

-

The_-RCS is at mid-loo A total: loss of'RHR occur .

No core-cooling is re-establishe .WHICH ONE- (1) of the following is the MINIMUM time requiredffor the RCS-to reach saturation? , minutes minutes minutes _ minutes QUESTION: 070 (1.00)

The following plant conditions exist:

~

-

The Unit is_. operating in MODE 5.-

--

.RHR-is in service.

,2 -

The operating'RHR pump-flow and amps are' oscillating.- ,

WHICH ONE- (1) 'of the following conditions: requires the running RHR-pump to be TRIPPED per-AOP-115.1, " RHR Pump Vortexing"? RHR flowiis-2500 gpm.

, KHR pressure'iss50-psi RCS Hot Ieg level is 12 inche .

d; RHR. temperature'is:212 degrees *

-

'

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_._. . _ . . . _ _ _ . - . . _ . - _ _ _ . _ . . _ _ _ . - . .- _ . _ . . _ .- ._ _ .

SENIOR REACTOR OPERATO ;PageE43-QUESTION: 071 (1.- 0 0 ) _

The following plant conditions exist:

-

- RHR is in service at reduced inventory condition AOP-115.1, "RHR Pump Vortexing", has been implemente WHICH ONE (1) of the following could cause an observed INCREASE-in RCS level during performance of AOP-115.1, " RHR Pump Vortexing"?

a. Venting the RHR syste b. Any opening in the'RCS boundary, c. Opening MVG 8809A, RWST to RHR PP_ d. High RHR flow rate when restarting' pum QUESTION: 072 (1.00)

The following plant conditions exist:

-

The reactor was operating at 100%= power when an ATWS occurre Pressurizer pressure is 2340 psi The reactor trip has NOT been verifie EOP-13.0, Response to Abnormal Nuclear Power Generation was entere Step 4, " INITIATE EMERGENCY BORATION OF RCS" is being performed and boric acid flow has been verifie WHICH'ONE (1) of the following is the NEXT required action? Start the turbine driven EFW pum Align valves to maintain EFW flow greater than 690 gp Locally-trip reactor trip and-bypass. breaker Open-pressurizer PORVs and block valves.

,

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-

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- SENIOR REACTOR OPERATOR Page.44 QUESTION: 073 (1.00)

The following plant conditions exist:

- The Unit has just received a Reactor Protection System input which requires a Reactor Tri The Reactor has NOT trippe The MANUAL Reactor Trip switches fail to actuate a1 Reactor Tri WHICH ONE (1) of_the following methods should be INITIALLY used to r SHUTDOWN the Reactor per EOP-13.0, " Response To Abnormal Nuclear Power. Generation"? Allow the control rods to insert AUTOMATICALLY at 72 steps per minut MANUALLY' insert control rods at 48 steps per minute, Trip the Motor Generator (MG) set output breaker Trip the Main Turbin ,

-

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' SENIOR REACTOR' OPERATOR Page . 45-QUESTION: 074 (1. 00 ):

The-following plant-conditions exist:

-

The Unit-is-in MODE One (1) Source Range Neutron Flux. Monitor is out of-

-servic The Source Range Audio _ Count Rate _ drawer is out of servic Core alterations are in progres WHICH ONE (1) of the following Technical Specification Action Statements should'be implemented? Suspend ALL operations involving positive' reactivity change l Emergency borate _the RCS until a boron concentration of 2150 ppm is establishe c .. Immediately evacuate the refueling area until the Audio Count Rate is returned to service, _ Replace the Reactor Ves'sel Head until both Neutron Flux Monitor and Neutron Flux Alarm are returned to service.

,

QUESTION: 075 (1.00)

WHICH ONE (1) of the following is an-IDENTIFYING characteristic of:

a RUPTURED Steam Generator per EOP-4.0,. " Steam' Generator Tube l Rupture"?- Level is DECREASING with MAXIMUM feedwater. flow, Level is STABLE with Blowdown isolated, Steam flow is greater than feedwater flow, Feedwater: flow-is greater than steam flo . - , , - . . - - -. -. :. - - . -

~ . , .~. . . - - - . . - . . -- ... . . . . . .- .. -

' SENIOR-REACTOR! OPERATOR Page 46

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' QUESTION: 076 .'(1. 0 0 )

The following plant conditions exist: -

-- The Unit is operating at 80%. .

_ _ _

-

A spurious low steam pressure SI signal is_-receive The cause of the spurious SI signal _has been corrected-and Safety-Injection Signal _has;been RESE During preparation to return the plant'to power-operation, the MSIVs FAIL to OPEN when their control '

switches are taken to OPEN.-

-

The Unit is currently in MODE 3 with steam pressure-downstream of the MSIVs at 1080 psi Tavg is 557 degrees WHICH ONE (1)-of the following is cause for the MSIV's failure to

-

OPEN? The MSIV-delta pressure is excessiv T1.s MSIV bypass valves are CLOSE The MSIV Motor Control Center breaker.is OPE The MSIV Isolation Signal has not-been RESE .m=- -,y + . . . , , ,w y 4 _ -- , . - -

. - - _ . . . . . . _ _ . _ _ . . - _ _ . _ _ , _ .

.

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iSENIOR REACTOR ~ OPERATOR - Page 47 _

'

! QUESTION: 077_ (1.00);

lWHICH ONE (1) of the following is the reason why EFW flowrate is-

~

-

procedurally restricted to less than.100 gpm when recovering a'

steam generator level if the level has fallen below 4% wide; range-indication? Ensure SG pressure transient condition does not occur which could result in an uncontrolled release through a Safety valv Ensure pressurizer _ level transient does not result in

pressure transient that-would actuate S Minimize thermal stress conditions on steam generator components.- Minimize RCS cooldown rate and prevent resultant thermal stress on RCS vesse QUESTION: 078 (1.00)

~

WHICH ONE (1) of the following is the reason for tripping ALL Reactor Coolant Pumps (RCPs) when Secondary Heat Sink is LOST?' To reduce heat input into the Reactor Coolant System, To reduce Reactor Coolant System pressur To reduce steam flow from steam generator ~

' To reduce decay heat input into the steam generator _

- . _ _ _ - - - _ _ _ - _ - - _

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SENIOR REACTOR OPERATOR' Page 48 QUESTION: 079 .(1.00)

zThe-following plant conditions _ exists

-

--

A HIGH radiation condition exiots on Reactor-Building-radiation monitor, RM-G17 _

-

AUTOMATIC actions have FAILED to occu WHICH ONE (1) of the following actions is required? Stop Train "A" AND "B" Purge exhaust fan Stop ONLY Train "A"-Purge exhaust fa ~ Close Reactor Building Purge discharge valves XVB-1A AND XVB-2 Close ONLY Reactor Building Purge discharge valve XVB-1A.

. -QUESTION: 080 (1.00)

WHICH ONE (1) of-the following ALTERNATE-methods of makeup are-available to the Spent Fuel Pool should level be lost and RHR trains are the source of the leak? Reactor Building Spray-pumps from the RWST' Spent Fuel Pool Cooling pumps from the-RWST

- Diesel' Fire pump from Monticello! Reservoir Filtered water from the fire service header-

_ ._ ._._ . - . , . ._ .._.._ -.. . . _ . _ _ _ - . . - - . - . _ - . _ ___ . . .. . _ . . -

l SENIOR. REACTOR OPERATOR Page'49 ,

- QUESTION: 081 (1.00)

!TheJfo11owing plant conditions' exist:

-:

-

' The Unit.has tripped from 100%. power due.to a-loss of- ,

off-site powe Diesels have~ STARTED and the sequencers have begun to LOA Fifteen (15) seconds after the sequencers start, Safety Injection is INADVERTENTLY actuate

WHICH ONE (1) of the following explains the response of the LEngineered Safety Feature load sequencers? The sequencers will-reset immediately and. start'to sequence the SI load The sequencers will reset in- twenty (20) seconds and start to' sequence the SI loads, The sequencers must be-reset MANUALLZ kr. fore the SI loads-will begin sequencing, The sequencers will start those: loads NOT started previously by the UV sequencer af ter twenty ;(:20) second . . .

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R SENIOR REACTOR' OPERATOR Page 50 QUESTION: 082 (1.00)

-The following plant conditions exist:

- .An undervoltage condition exists on 7.2KV ESF bus 1D Diesel- Generator "A" has FAILED to star WHICH ONE (1) of the following actions may be taken to mitigate the-undervoltage condition? START Diesel Generator'"B" to supply emergency power to 1D Position the Normal to Emergency control switch for.ESF bus 1DA to the EMERGENCY positio DE-ENERGIZE the power supply.to the ESF Loading-Sequence (ESFLS) and SHUT the alternate feeder breaker to 1DA from ,

XTF-3 Attempt to RE-CLOSE-the normal feeder breaker from the 115 KV Parr lin QUESTION: 083 (1. 00)

WHICH ONE (1) of.the following conditions REQUIRES initiation of a-manual reactor trip during~a loss of instrument air at 50% power?. Instrument air: pressure, drops to 80 psig.AND XVA02670-IA, TB Instrument Air-Header-Isolation Valve, is OPE Instrument air pressure drops to 70 psig AND IPV08324-IA, -

. Station Air. Supply Header Pressure Control Valve, is -

OPE Instrument air pressure drops to 60 psig AND Main-Feedwater Regulating valves are at MID-POSITIO ~ Instrument-air pressure drops to.50 psig AND LCV-459,

Letdown Line Isolation valve is at MID-POSITIO .

T d -

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SENIOR REACTOR OPERATOR - Page_51-

[ QUESTION: .084 -(1.00)

The L following plant conditions exist:

-

John and Harry have been chosen to make a containment entry to isolate a lea John has accumulated 700 mrem exposure for the prese'

quarter and_3000 mrem for the-year to dat Harry has accumulated 500 mrem exposure for the present quarter and 3900-mrem for the year to dat Dose rate in-the area they will be working in-is 1 Rem / hou Their stay-time will-be approximately fifteen 1(15)- -

minute WHICH ONE (1) of the following is the MINIMUM action that must be taken prior to allowing the containment entry? The Operations Shift' Supervisor must sign the approval for extension of the quarterly exposure limit for John, The Vice President, Nuclear Operations must sign the

' '

approval for extension of the quarterly exposure limit-for Joh The Operations Shift _ Supervisor must sign the; approval-for extension of the yearly exposure limit for Harr The Vice President, Nuclear Operations must sign the approval for extension of the yearly exposure limit for Harr _

E j

_ _ _ _ . . . - . . _ _ _ .

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-SENIORLREACTOR OPERATOR .Page 5 # QUESTION: 085 (l '. 0 0 )

.Thef following plant conditions exist:

-

The plant is operating at 100% powe Adjustments are to be made to the packing for the 1611,

"Feedwater Isolation valves".

- The valves will be BLOCKED OPEN while work is being performed on each valv WHICH ONE (1) of the following is the status of each valve while work is being performed? The valves are OPERABLE, but must be stroke timed following completion of work to verify operability, The valves _are INOPERABLE unless a dedicated _ operator is ,

standing by to remove the blocks should a feedwater isolation be require The valves are OPER'tBLE as long as the accumulators remain pressurize ' The valves are INOPERABLE until the blocks are REMOVED '

and the valve is stroke time QUESTION: 086 (1.00)

WHICH ONE (1) of the following individuals, by title,1nay be issued-a MASTER KEY to Radioactive Materials Storage Areas? _ HP Lab Specialist perf orming routine surveillanc Licensed-Control Room Operator performing normal plant tour Qualified' Auxiliary. Operator performing routine

'

' surveillance rounds, I & C Technicians performing a normal surveillance test on radiation monitoring equipmen .

r w-

= SENIOR REACTOR OPERATOR Page 53

. QUESTION: 087 (1. 00)

The following plant conditions exist:

-

A.normally throttled valve has been closed for-system -

isolation during maintenanc Maintenance is now complete and the system is being returned to norma WHICH ONE (1) of the following is the method for independently verifying the position of the THROTTLED valve? Observe the relative height of the valve ste Use local positice indicatio Observe the number of turns OPEN from the full CLOSED-position as the initial operator returns the system to servic Fully OPEN the valve and count the number of turns in the CLOSED direction until the desired position is obtaine QUESTION: 088 (1.00) q

WHICH ONE (1) - of the following is the MINIMUM required tagging -

necessary to assure a motor operated valve (MOV) is isolated for personnel safety per SAP-201, "Da'nger Tagging"? A red Danger Tag on thelvalve breaker and a red Hold Tag on the remote operating switc A red Danger Tag on the valve breaker and a red Danger-Tag on the local valve position handwheel, A red Danger Tag on the, local valve position handwheel .

and a yellow Caution Tag on the remote operating switc A red Danger Tag on the remote-operating switch.and a yellow-Caution tag on the local valve-position handwhee . . -. .. - . _ - - -. - . - . . . . . . - .-.- --- ~ ~.- .-.- - -.-.. ~. . . .

~ -SENIOR = REACTOR _ OPERATOR Pagei54

_

!

LQUESTION: 089 (1,'00)

WHICH ONE (1) of the'following is-a VIOLATION of-Administrative

' Procedures-when performing Emergency Maintenance?-

- _ Work is authorized by the Duty Shift Supervisor, Maintenance Work Request (MWR3)_ are-NOT complete Procedure deviations exist and are documented on the Off

-

Normal Occurrence, SAP-13 QC hold points are waive QUESTION: 090 (1.00)

Each of the following is a component / system-that is authorized to be OPERATED by maintenance personnel EXCEPT: Instrument air header isolation valves Bridge cranes , Demineralized water hose bib valves- Amertap System 4-p

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_ ,

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SENIORLREACTORf0PERATOR' ;Page)55

.9UESTION: 091 .(1. 0 0 )

.WHICH'ONE (1) of_the following is a' VIOLATION of Administrative

~ Procedures when troubleshooting on an OPERABLE piece'of equipment

~

that 1 could affect plant operating status? Major steps' involved in troubleshooting are documented on'

the MW The-shift Engineer is. notified of the troubleshootin '

. Leads are lifted as needed by maintenance personnel 1 without an approved Work Document to facilitate troubleshooting, Inspection covers are removed for visual inspection without an approved Work Documen QUESTION: 092 (1.00)

WHICH ONE (1) of'the following individuals, by' title,. may fill th position of Fire Brigade Leader per SAP-200, " Conduct Of-Operations"? Mechanical Maintenance Fire Brigade member I&C. Fire Brigade' member Fire' Protection' Officer- Turbine Building operator

. . - . - - . - . ., - - ..- - .- .- - -. .

. SENIOR REACTOR OPERATORL Page156 ,

-QUESTION: 093 (1.00)

'

WHICH ONE . (1) of the following requirements applyJ to_ the .

use/ application of a CAUTION statement that precedes Step 12 of an Emergency Operating Procedure? Applies ONLY to Step 12 " Action / expected response" items, Applies to Step 12 " Action / expected response" items AND

" Alternative Action" item Applies ONLY to ALL future " Action / expected response"-

items of the current Emergency-Operating Procedur Applies to ALL future " Action / expected response" items AND " Alternative Action" items of the current Emergency _

Operating Procedur QUESTION: 094 _ (1.00)

The following plant conditions exist:

-

Operators are performing Immediate Action Steps of EOP-1.0, " Reactor Trip or Safety Injection".

-

A RED path-condition exist on HEAT SIN WHICH ONE (1) of the following actions.should be taken if ALL power-is lost to the AC Emergency Buses?

- Complete Immediate Action steps _of EOP-1.0 and.-transition-to EOP-15, " Response To Loss Of SecondaryjHeat Sink". Transition to EOP-15, " Response To-Loss Of Secondary, Heat:

Sink", IMMEDIATEL .

!

' Complete Immediate Action steps!of EOP-1.0=and transition tofEOP-6.0, " Loss of All ESF AC Power", : Transition to EOP-6.0, " Loss of All ESF AC Power",

IMMEDIATEL '

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SENIOR-REACTOR OPERATOR .Page 57

QUESTION: 095' (1.00)

'WHICH-ONE (1) of the following statements defines the MINIMUM criteria for a '

HIGH RADIATION AREA? An area where a person could receive a dose between 10 mrem and 100 mrem in one hour, An area where a person could receive a dose between 100 mrem and 1000 mrem in one hour, An area where a person could receive a dose between 10 Rem and 100 Rem in one hour, An area where a person could receive a dose between 100 Rem and 1000 Rem in one hou ._

- -- . .

_ - ~ ~ . - . - ... - . - - - . . - . - . . - . ., - _ - _ . . . .

1SEN!OR R2 ACTOR-; OPERATOR .PageL58-

!

JOUESTION: 096 (1 ~. 0 0 )

The1following plant conditions exist:

- A Reactor Startup iscin progres . Control = Rods are being_ withdraw The Control Bank D rods have' reached desired estimated critical position (ECC), ,

-

The Reactor.is NOT critica =WHICH ONE (1) of the following actions should be taken next per GOP-APPENDIX A, " General Operating Precautions"? Trip the Reactor and remain in HOT STA DBY until the ~

Reactor Engineer can evaluate the proble ~ Insert the Control Rods'and lemain'in HOT STANDBY until the Reactor Engineer can evaluate the problem, Continue to withdraw the. Control ~ Rods until eitherfthe Reactor is. critical or until the MAXIMUM rod withdrawal limit is reache Maintain current Control Rod position and-dilute the RCS until either the Reactor is critical or until-a boron concentration equivalen't to +500 PCM greater.than thefEC position is reached.

,

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. -. . _ . . .__ .. . _ _ . - _ ~ .__

- __ _ ._. ._. _ _ .

,

J" SENIOR-REACTOR = OPERATOR- Page-59L l

.1 QUESTION: 097 _ (1. 0 0 )

'

The~following plant. conditions exist:-

--

The Reactor is in MODE The Shift Supervisor needs a control' room operator for ,; '

the upcoming evening shift on a Saturday night _to replace an.ill crew membe .

'

WHICH ONE (1) of.the following is the PREFERRED operator to fill'

this' vacancy per Administrative P'rocedure SAP-200, " Conduct Of-Operations"? An Operator who has worked his. normal dayshift, is willing to hold over on the evening shif An Operator who is-assigned to Administrative dayshif An Operator who is on his normal days off, d'. An_ Operator who is on vacation.

= QUESTION: 098 (1.00)

WHICH ONE'(1) of the following.is the MINIMUM-review / approval'

necessary for a-temporary procedure change?

' Discipline Supervisor Shift Supervisor c.- Qualified Reviewer AND Shift Supervisor Discipline Supervisor, Shift Superviscr, AND the Plant Manager h

n

.

..z- . u .- --- . . . . - - _ . _ . ._ - . . _ , _ , .-. - . . .. -- . . . - . .-. - . . . . - - . - _ . .-. .

' SENIOR REACTOR OPERATOR Page160 QUESTION: 099 (1.00)

WHICH ONE (1) oof the following is the approved method for pulling

fuses for a Tagout of'an electrical component?-

_,

The Duty Electrician shall pull'the fuses and install the Danger Tags, An operator shall pull the fuses and install the Danger Tag The Duty Electrician shall pull the fuses and an-operator shall install the Danger Tag An operator shall pull the_ fuses and the Duty Electrician shall install the Danger Tag QUESTION: 100 (1.00)

WHICH ONE (1) of the:following plant conditions _ defines requirements. for a Main Control Board component TAGGED with + n-ORANGE "R&R Entry" Tag? Assume 1NO other tags are installe ,

a .- The component is INOPERABLE and cannot be used for any reason, The component is-OPERABLE in.any plant MODE, bat should only be used in an emergency situatio _The component-is INOPERABLE, but can be operated if ~

neede The component is OPERABLE ONLY in MODES 5 an 6, but should only_be used in.an emergency. situation.

!

l (**********.END OF EXAMINATION.**********)

l l - - , . -,

SENIOR REACTOR OPERATOR 2 age 61 ANSWER: 001 (1.00) (41.0)

REFERENCE: VCS: IC-5, Rod Control, p. 13, Objective 6,7,8 KA 001000K602 (2.8/3.3) 001000A102 (3.1/3.4)

001000K602 ..(KA's, ANSWER: 002 (1.00) (+1.0)

REr: 1ENCE: VCS: IC-5, Rod Control, p. 26-30, Objective 6 KA 001000G008 (3.6/3.6)

L

_

001000G008 ..(KA's)

ANSWER: 003 (1.00) (+1.0)

- - . . _ _ _ _ _ _ _ _ _ _ _ . _

SENIOR REACTOR OPERATOR Page 62 REFERENCE: VCS: L13 - 4 , RCP, p. 19, Objective 6 and 7 KA 0030000007 (3.2/3.3)

L 003000G007 ..(KA's)

ANSWER: 004 (1.00) (+1.0)

REFERENCE: VCS: ARP-001, XCP-617, Pt. 2-1, p. 12 VCS: AB-4, Reactor Coolant Pump, Objectives 6, 7, 8 KA 003000A202 (3.7/3.9)

003000A202 ..(KA's) _

ANSWER: 005 (1.00) [+1.0)

. . . . .

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

SENIOR REACTOR OPERATOR Page 63 ,

i

'

- REFERENCE:

'

<

, AB-4, " Reactor Coolant' Pump", E.O. 6, page 2 ' KA 003000G010 (3.3/3.6)

i l

l

i 003000G010 . . (KA's) .

ANSWER (1.00)

' (+1.0)

i REFERENCE:

,

l'. AB-4, " Reactor Coolant Pump", E.O. 6, page 3 . AB 4, " Reactor Coolant Pump", E.O. 7,-page 3 . AB-3, " Chemical and Volume Control System", E.O. 6.

KA 003000G005 (3.4/3.8) i 003000G005 . . (KA's) .

~

,

ANSWER: 007 (1.00)

b. or ( + 1. 0 )

,

-REFERENCE:

. VCS: AB-3, CVCS, p. 16, Objective 5 and 7 KA 004010A211~ (3.1/3.1)

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.

004010A211 . . '( KA ' s ) .

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'

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SENIOR REACTOR OPERATOR Page 64 j l

i AMSWER: 008 (1.00) !

,

' [ 41. 0 )

REFERENCE:

1. . AB-3, "

Chemical And Volume Control System", E.O. 3, page 3 . IB 2, " Component Cooling Water", page 1 . KA 004000K110 (2.9/3.2)

>

004000K118 ..(KA's) i

!

'

ANSWER: 009 (1.00) . (+1.0) ,

REFERENCE: Technical Specifications Pg.,3.3-15a, 16, and 3.0-1 Note: This is an-obvious operability problem, and SI is -

-

'

required in MODES 1-4. With all trains of SI automatic actuation inop, the candidate should be aware without reference to the Tech Spec that 3.0.3 applie .. KA 013000G005 (3.6/4.2) "

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013000G005 ..(KA's)=

ANSWER: 010: (1.00)

c.- (+1.0]

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SENIOR REACTOR OPERATOR Page 65 REFERENCE: IC-8, " Nuclear Instrumentation", , page 4 .. KA 015000G007 (3.3/3.4)

015000G007 . . (KA's)

ANSWER: 011 (1.00) (+1.0] l

!

REFERENCE: IC-12, " Core Subcooling Monitor", E.O. 4, page . KA 017020K401 (3.4/3.7)

017020K401 . . (KA's)

- ANSWER: 012 (1.00) (+1.0)

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!

i SENIOR REACTOR OPERATOR . Page 66 i t

REFERENCE: l TB-7, "Feedwater System", , page 2 L i KA 059000K419 (3.2/3.4) ,

.

!

,

059000K419 . . (KA's) ,

,

'

ANSWER: 013 (1.00) (+1.0)

REFERENCE:

1.- IB-3, " Emergency Feedwater System", E.O. 5, pages 27-28, KA 061000K402 (4.5/4.6)

>

061000K402 . . (KA's)

' ANSWER: 014 (1.00)

d '. [+1.0)

REFERENCE:

1. - Instructor's Lesson Plan EOP-15, " Response.To Loss-Of Secondary Heat Sink", E.O. 8, page 1 > KA 061000K301 ( 4 . 4 /4 '. 6 )

061000K301 . .(KA's) ._

,

+,,.-e..,. +-,-n ,...n... , n . ~u,,+.,, ,, -,,+c-+,,.,

-

e.,e., ,

SENIOR REACTOR OPERATOR Page 67 ANSWER: 015 (1.00)

.

b, or [+1.0)

REFERENCE: VCS: AB 12, Waste Gas System, p. 21 & 22, Objectives 4, 5, 6 VCS: SOP 119, Waste Gas Processing, p. 31 & 32 KA 071000A407 (3.0/3.0) 071000A413 ( 3 . 0 / 3 .1 ) -

L-071000A413 071000A407 ..(KA's)

ANSWER: 016 (1.00) [+1.0)

REFERENCE: GS-9,-" Radiation Monitoring System", ., page 1 .- KA.068000A404 (3.8/3.7)

.

068000A404 ..(KA's)

ANSWER: 017' (1.00)

- (+1.0)

.

-.

-I

. . . . . ., ... ,....._...1,...,_.;, ...i....,_,.. . ....;., . , . . , - , . J,__,,.,_.,_._-,._,,_,_...._.,.,_,,,,,,, , . . _ , . , _ _ _ , , , _ . . ,

. __ _ _ _ _ .

_ _ . . _ _ - _ _ . . . ___ . . . ._ _ ._.. - - , _ _ _ _ . _ _ . . _ _ . _ _ . _ _ _ . _ _ _ . .

,

SENIOR REACTOR OPERATOR Page 68

REFERENCE: ,

, VCS: Technical Specifications pg. 3/4 11.2.5, Immediate -I

'

requirement KA 071000A429 (3.0/3.6)

.

b I

071000A429 . . (YA's)

ANSWER: 018 (1.00)

u (+1.0)

REFERENCE: Technical Specification LCO 3.1.1.4, page 3/4 1- , KA 002000G005 ( 3. 6 /4.1) '

002000G005 . . (KA's)

ANSWER: 01 (1.00) [+1.0)

.k b

r-,Aw -,- -- -- ,, m y r em w mr w - e- r= = v< v= r -w r ,r rn -,,-r---mwr,.r- ,w wrg .= r , - ~r %- m .

. # <+ -ra,s-+- = = t--o- s-

SENIOR REACTOR OPERATOR Page 69-REFERENCE: IC-13, " Reactor Vessel Level Indication System", ,

Figure IC1 .- KA 002000A403 (4.3/4.4)

002000A403 ..(KA's)

-ANSWER: 020 (1.00) [+1.0)

-REFERENCE: AB-10, Emergency Core Cooling System", E.O. 5, pages 12-1 . KA 006030K403 (3.4/3.6)

006030K403 ..(KA's)

.

ANSWER: 021 (1.00) -

.

b- . [+1.0)

!

,

!

,

vrb h-r-

'

~. g e n -.h, - ty e., r=r-**-*mvv-,- - - - , - --e--*~*- w= rwrtTb-

'

r- a=ein==~*** v --t + - + - - = = + - = ^--,r'='*

'

- = - * * - - - * ' rw *; re = -r-*

- . . _. . _ - _ ___ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ __

I

' Page 70

'

-- SEMIOR REACTOR OPERATOR l J

l REFERENCE: ! AB-10, " Emergency _ Cooling Water System", E.O. 10, page 2 . Technical Specification 3/4.5.1, " Accumulators" , page 3/4 5 I

-- 3 . KA 006000G005 (3,5/4.2) , 006000G011 (3.6/4.2)

t 006000G011 006000G005 . . (KA's) -

,

ANSWER: 022 (1.00) [+1.0) ,

-REFERENCE: AB-10, " Emergency Core Cooling System", , page 2 . KA 006020A107 (3.5/3.7)

006020A107 ..(KA'a)

t ANSWER: 023' (1.00) (+1.0) !

,

C-L s

1

--v' 1 T w' '=- ~ ew' ,e,.e+r-w Eee a-' - seme s -w e w- em --w-- w--w-----t-er+w-rae se r+ ce--+ tu -c' y -- e e's w- a '

rw- r- v' w'rwe - - ' , --

. .- --- . - . . . . . _ . - . - . - - .. . - - .. _ . - - . -- ~ - __ -. - -. . . - .

l

.I l

i SENIOR REACTOR OPERATOR Page.71 I i

l l

REPERENCE: '

'

l VCS:- IC-3, Pressurizer Level and Pressure Control, p. 14-16, i fig. IC 3.9, Objective 4 and 5 l KA 010000A107 (3.7/3.7) '

L-1 010000A107 ..(KA's) !

i

ANSWER: 024 (1.00) !+1.01 REFERENCE: IC-9, " Reactor Protection and Safeguards Actuation System", , page 2 . KA 012000A406 (4.3/4.3) -t

i 012000A406 ..(KA's)

ANSWER: 025 (1.00)

(+1.0]

.,

h

.

!

v i

m w- --

,,gr--r-- +m-r--nn e v, , n ><~-e- - - , ,,.ne, ,, -- c, ... , ~ -, ~er.w- - - , , -- -, -- --

.!

SENIOR REACTOR OPERATOR Page 72

.1 REFERENCE: IC-4, " Rod Position Indicati^on", E.O. 7, page . KA 014000A102 (3.2/3.6)

014000A102 ..(KA's)

'7."(SWER: 026 (1.00)

- (+1.0)

REFERENCE:

- FACILITY EXAM BANK QUESTION ft 774 GS-5, " Spent Fuel Pit Systema , E.O. 9, page . KA 033000A101 (2. 7/3. 3 ) , 033000A203 (3.1/3.5),

033000K401 (2.9/3.2)

033000A101 033000A203 033000K401 . . (KA's)

ANSWER: 027 (1.00)

. (+1.0)

. REFERENCE:

- FACILITY-EXAM BANK QUESTION-809-

' KA 033000G006 - (2.1/3.1) ,

,

h

>

033000G006 ..(KA's)-

,

t

,,,mi---www-,- -,,,,,w.y-,-yr.p+-y-

-

-<4 .-..m, ,e, . ..--rw-,,, ,- .,.e . .#-,,-,,, g e ,,.m, ,,y,.w.w.ym,.w,e,,7,+- -,-*s,+.,-yy.9..,

SENIOR REACTOR OPERATOR Page 73 ,

,

ANSWER:' 028 (1.00) [+1.0) ,

!

REFERENCE: :

TB-2, " Main Steam System", E.O. 7, page 1 . KA 039000K102 (3.3/3.3) ,

>

039000K102 ..(KA's)

ANSWER: 029 (1.00) (+1.0) ,

REFERENCE: VCS: 'AB-8, Reactor Building Spray System, p. 18-19,

,

Objectives 5 and 6 KA 064000K411 (3.5/4.0) :

L 064000K411 ..(KA's) .

.

i ANSWER:- 030 (1.00) , [ + 1. 0 ] -

o '.

t

-

'. , , . , - , , , , ,..,.,,, ...-,:-... - . .. - - ..,,,~ _ ,. .

-

., , . . . . . - . . . - . . . . . . - - -.

-SEMIOR_ REACTOR _ OPERATOR Page 74 ,

,

" REFERENCE: GS-9, "

Radiation Monitoring", E.O. 8, page 3 ; KA 073000K401 (4.0/4.3) ;

e

-

'

073000K401 . . (KA's) ,

ANSWER: 031 (1.00) -;

b.- [+1.0) l REFERENCE: TB-8, "

Circulating Water System", E.O. 7, pages 12~& 2 . -KA 075000K102 (2. 9 / 3.1)

.

075000K102 -. . . (KA's)

.

ANSWER: 032 (1.00) (+1.0) -

i

.

I y

- -

,

-

tr # hu* r--es s---e e- w tne-m -.P-. mum -,,i, - qa-+h w i- e ,yi-= . g- 4- pm'y 4 Tty-- ,grw ar, 'r-g-y--se--g& a--

r-'M--w m EP'Fg

_ . . .

SENIOR REACTOR OPERATOR Page 75-REFERENCE: GS-11, " Fire Protection System", E.O. 5, pas s . lUt 086000A401' (3.3/3.3), 086000G009 (2.9/3.3)

086000G009 086000A4L1 ..(KA's)

ANSWER: 033 (1.00) - (+1.0)

REFERENCE:

.1. - -VCS: IB-2, CCW, p. 22, Objective 3, 6, and 7 KA 000000K101 (3.1/3.1)

L 008000K101 ..(KA's)

ANSWER: 034- (1.00) [+1.0)

REFERENCE:, EOP-14, " Response To Inadequate Core Cooling", page . Instructor's Lesson Plan EOP-14, E.O. 7,8, page 24.- KA Numbers 028000K501 (3.4/3.9), 028000K502 (3.4/3.9).

028000K501 ..(KA's).

. _ ..

SENIOR REACTOR OPERATOR Page 76

!

ANSWER: 035 (1.00) ; (+1.0) !

REFERENCE: Instructor's Lesson Plan EOP-14.0, E.O. 7,8, page 2 . KA 028000G004 (3.3/3.5)

!

.

028000G004 ..(KA's)

,

ANSWER: 036 (1.00) (+1.0)

REFERENCE: GS-4, "" Fuel Handling System", , page 1 . KA 034000K403 (2.6/3.3) ;

034000K403 ..(KA's)

ANSWER: 037 (1.00) (+1.0)

l

.

.'

I

= . _ , - ,e,,,-- - , - - . . , - _ ~ , - . _ , . ~ . . - - - . . - . . . _ . . . - - . - - . - ~ . - - - - - - - . _ . . . - , - - - . . . . , . - . , , - - , , + , . . - - - - , - - , - - . ~ + . ,

I i

SENIOR REACTOR OPERATOR Page 77 l 1 REFERENCE:

1.- TB 9, " Main Turbine", E.O. 1, pages 4- l KA 045000A305 (2.6/2.9)

!

045000A305 ..(KA's)

,

P ANSWER: 038 (1.00) i+1.0)

i REFERENCE:

- IB-1, " Service Water System", E.O. 3, page 1 f 2. - KA 076000K109 (3.0/3.1) i c

t

076000K109 ..(KA's)

'!

-

ANSWER: 039 (1.00) [+1.0) ,

REFERENCE:

- 1. - _TB-12, " Instrument and Service. Air", E.O.- 6, page . KA 078000K402 (3.2/3.5)

.

&

078000K402 ..(KA's) .

P

'

l

>

" - - - - . - -

i

,

SENIOR REACTOR OPERATOR - Page 78 l

ANSWER: 040 (1.00) { [+1.0) :

,

REFERENCE:

,

" Instrument and Service Air", F.O. 8, page . TB 12, i KA 07 'K201 (2. 7/2. 9 )

t t

078000K20 . . (KA's)

ANSWER: 041 (1.00) (+1.0) .

.

REFERENCE: VCS: AOP-403.4, p. 1 . KA 000005G010 (3.4/3.6) ,

,

000005G010 . (KA's) .+

. ANSWER: 042 (1.00) , (+1.0)

,

r

~m u e w.m,-- - ..,.-,...,.L.n.--.

-

-g,.,,,, .,we-m -p--Ww ,-wwn,-

SEMIOR REACTOR OPERATOR Page 79 REFERENCE: VCS: Technical Specifications 3.1.3.1. . KA 000005G003 (3.1/3.6)

L 000005G003 ..(KA's)

ANSWER: 043 (1.00) [+1.0)

REFERENCE: AB-4, " Reactor Coolant Pump", E.O. 5, page 3 : KA 000026A204 (2.5/2.9)

000026A204 ..(KA's)

ANSWER: 044 (1.00) [ +1. 0) -

SENIOR REACTOR' OPERATOR - Page 80 REFERENCE: AOP-106.1, " Emergency Boration", paga . Instructor's Lesson Plan AOP+106.1,-Objective 2, page . KA 000024K301 ( 4 .1/4 . 4 )

000024K301 ..(KA's)

ANSWER: 045 (1.00) [+1.0)

...t REFERENCE: ,

' AOP-106.1, * Emergency Boration", pages 1- . Instructor's Lesson Plan - AOP-106.1, E.O. 6, pages 3- . KA 000024A202 (3.9/4.4)

!

'

.

000024A202 ..(KA's)

ANSWER: 046- (1.00)

- ( + 1. 0]

o F

't h

_ _

?

,

-- :w

SENIOR REACTOR OPERATOR Page S1 REFERENCE: AOP 4 0' . 5, " Pre 9surizer Pressure Control Channel Failure",

page 1 Instruc Jr's Lesson Plan AOP-401.5, , page . KA 000027A215 (3.7/4.0)

000027A215 ..(KA's)

ANSWER: 047 (1.00) [+1.0)

REFERENCE: Instructor's Lesson Plan EOP'-4.0, page 2 . KA 000040A201 (4. 2/4. 7) , 000040G007 (3.3/3.6)

'

000040G007 000040A201 . . ( KA ' s )-

ANSWER: 048 (1.00) [+1.0)

t lI i

I

1

'

t

-'

SENIOR REACTOR-OPERATOR Page 82 REFERENCE: BOP 1.0, " Reactor Trip / Safety Injection Actuation", Reference Pag l KA 000040A204 (4.5/4.7) j i

I

,

000040A204 ..(KA's)

ANSWER: 049 (1.00) (+1.0)

REFERENCE: AOP-206.1, " Loss of Condenser Vacuum", page . Instructor's Lesson Plan AOP-206.1, E.O. 4, page '

- KA 000051A202 (3.9/4.1)

000051A202 ..(KA's)

'

ANSWER: 050 (1.00) [+1.0)

__ F l

__

,

-.

, . - _ _ _ . - _ . _ _ . _ _ . . . - _ - . _ _ . _ _ . _ _ .. ..__ _ _. -. - ...__._ _, _.__. _ _ _ .

SENIOR REACTOR OPERATOR Page 83 t

l REFERENCE: !

FACILITY EXAM BANK QUESTION 1059 e 2.= IB 5, " Diesel Generator System", E.O. 8, page 1 . KA 000055A102 (4.3/4.4) ;

000055A102 ..(KA's)

- ANSWER: 051 (1.00)

, [+1.0)

,

~ REFERENCE: , EOP-6.0, " Loss of=All ESF AC Power", page l.- Instructor's Lesson Plan EOP-6.0, E.O. 3, page 1 . KA 000055G010 (4.1/4.3)

,

000055G010 ..(KA's)

ANSWERt 052- (1.00)

= [+1.0)

,

t

-

~.

m- .._;._,,..,._.-....__,,....-. - . - . . .- . . _ , ., . . _ . - . . _ . , _ - . . , _ . _ _ , , . _ , - - . , _ . . . _ . .

- . .-. - - -.=-- .. . . . .-.. -- _ .--_ -. ....-.- .-_ - . . -.-.- .. . . - . .-. - - _ - .-.

t-SENIOR REACTOR OPERATOR Page 84 REFERENCE: i i

'

- FACILITY EXAM BANY, QUESTION 1065 . GS-2, " Safeguards Power", E.O. 3, Figure GS ; KA 000057A214 (3.2/3.6)

!

l 000057A214 . . (KA's)

-!

ANSWER: 053 (1.00) (+1.0) i REFERENCE: EPP 001, " Emergency Action Levels", page 1 . KA 000067G002 (3.2/4.1)

,

t 000067G002 . . (KA's) .

-ANSWER: 054 (1.00)

, [+1.0)

,

t

.-_

g ,p g wbi g --aggym-pqr+sy+ .wg--g -ge eye y gym 9 y-q-a,qe-- ,v.w. ,. y, gg. s -+ wty*w= g*t-yy- =,W<*-- y cw y y- y * t-W?'

_ _ . . . ,

SENIOR REACTOR OPERATOR Page 85 L

REFERENCE:

1.- Instructor's Lesson Plan AOP-600.1, E.O. 7,-page . KA 000068K318 (4.2/4.5)

000068K318 ..(KA's)

_

ANSWER: .055 (1.00) (41.0)

-REFERENCE: AOP-600,1, " Control Room Eva'cuation", page 7, Instructor's Lesson Plan AOP-600.1, , page . KA 000068A21.'. (4.3/4.4)

000068A211 ..(KA's) -

- - - _ , - - -

ANSWER: 056 (1.00) (+1.0)

.i

.

i

.

l SENIOR REACTOR OPERATOR Page 86 f

i-REFERENCE:

VCS: EOP-17.0, Response to High Reactor Building Pressure, ;

Rev. 4 ,

' KA 000069A202 (3.9/4.4)

i i

!

e i

000069A202 . . (KA's) !

!

I ANSWER: 057 (1.00) '

b. or~ [+1.01 ,

REFERENCE: Technical Specification 3.6.1, " Containment Integrity", page .

3/4 6- . KA 000069A201 (3.7/4.3) ;

000069A201 . . (KA's) '

=-

IMSWER: 058 (1.00) , (+1.0)

.

k F

-rv , r ,w- v y ~: .. ,4 ._,n,., e..-g-n s ,, , .- .,..e - + , . ~ , , - - s,n-- , , . , - . . , , - , u-+-- .,- , + , ,~, -

SENIOR REACTOR OPERATOR

'

Page 87 i

!

REFERENCE: EOP-12, " Monitoring Of Critical Safety Functions", Attachments i 1- ! KA 000069G011 (4.0/4.2)

,

000069G011 . .- ( KA ' s ) :

,

ANSWER: 059 (1.00) (1.0)

REFERENCE: I

- Westinghouse ERG Background document FR- . Instructor's Lesson Plan EOP-15, E.O. 5,_page 4 . KA 000074K311 (4.0/4.4)

000074K311 ..(KA's)

.

ANSWER: 060 (1.00) [+1.0]

.

>

l l.

,

l ,

. . . _ . , - _

-

SENIOR REACTOR-OPERATOR Page 88 REFERENCE: VCS: Emergency Operating Procedures Lesson _ Plan, EOP-12.0, Monitoring Critical Safety Functions, p. 2 . KA 000074A207 (4.4/4.7) 017020K503 (3.7/4.1)

000074A207 ..(KA's)

_

ANSWER: 061 (1.00) ( + 1. 0 )

REFERENCE: VCS: GS-6, Primary Chemistry and Sampling, Objectives 6, 9, >

10, p.22, 29, 30 VCS: CR-2, Plant Chemistry Control, Objectives, 7, 12, p. 17 .

- VCS: Technical Specifications 3. . KA 000076K305 (2.9/3.6)

_

_ _

000076K305 . . ( KA ' s ) -

ANSWER: 062 (1.00)

- {+1.0)

n

"

. . . . . . . .

_- _ - _ - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ . _ _ - .-_

SENIOR REACTOR OPERATOR Page 89 REFERENCE: Instructor's Lesson Plan AOP-403.3, "

Continuous Ro Withdrawal", E.O. 4, page . KA 000001G010 (3.9/4.0)

000001G010 ..(KA's)

__

ANSWE (1.00) ( + 1. 0)

REFERENCE: Instructor's Lesson Plan AOP-403.6, , page . KA 000003G0102 (3.9 /3. 8)

000003G010 ..(KA's)

_

ANSWER: 064 (1.00) (+1.0)

l

_ - - - - - --_ - -.m.- - . - - . _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ . - - - _ . _ _ ___..____._.m__,___._m_ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ . _

..

SENIOR REACTOR OPERATOR Page 90 REFERENCE: VCS: TB-5, Turbine Control and Protection System, p. 52, Objective 1 and 7 KA 000007K301 (4.0/4.6)

L 000007K301 ..(KA's)

_

ANSWER: 065 (1.00) (+1.0)

REFERENCE: FACILITY EKAM BANK QUESTION 1028 TB-2, " Main Steam System", E.O. 7, page 2 . KA 000007K103 (3.7/4.0)

_

000007K103 ..(KA's)

ANSWER: 066 (1.00) [+1.0)

- , - . _ . . ..- .-... - . _ . . . . - _ _ _ - . . . - - . - .

. ..

SENIOR--REACTOR-OPERATOR -Page 91

REFERENCE: -'

FACILITY EXAM BANK QUESTION 2115-2: Instructor's Lesson-Plan EOP'-12, " Monitoring of Critical Safety Function",.E.O.' 2, page . -KA 000007G012 ;(3.8/3.9)

!

--000007G012 ..(KA's) >

.

ANSWEP. : 067 (1.00) [ + 1'. 0 ]

^t REFERENCE: FACILITY EXAM BANK-QUESTION 2176 . Instructor's Lesson-Plan EOP-2.0, E.O. 5, page 2 . KA 000008G011 '(4.0/4.1) G011 ..(KA'a)

LANSWER: 068 (1.00)

! (+1.0)

l

!

':

,

y + -..-,--y 7 - - ,,

.. - - - -- _ . . - - . - . - - - _ _ . - - - - . . .

.

SENIOR REACTOR _ OPERATOR Page;92 REFERENCE:

t Ic.- Instructor'a Lesson Plan EOP-2.0, page 1 _

s KA_000011A103 (4.4/4.4), 000009A215 _ (3. 3/3. 4 )

000011A103 000009A215 ..(KA's)

ANSWER: 069 (1.00) (+1.0)

REFERENCE: VCS: GOP-9, Mid Loop Operation, Attachment IV Generic Letter 88-17 VCS: AB-7, Residual _ Heat Removal System, Objectives 6,7,8' , _

24 Generic-Letter 88-17 Note: This question reveals if the candidate'is--sufficiently-sensitive to the issue of loss of-RHR,'and<the very short time frame-available to-respond to sam Industry events have occurred where RHR has been lost at reduced. inventory, and one key-issue is the operators were often-not aware how little -

time was available until saturation'was_ reached in the cor The question _does not require detailed knowledge of the saturation vs. time curve due to the very.large time frame of the incorrect distractors.... KA'000025K101-(3.9/4.3)

000025K101 ..(KA's)

.

ANSWER: 070 (1;00) [+1.0]-

. .

. - . . _ -

SENIOR REACTOR OPERATOR Page 93;

-! REFERENCE:-

1.- Instructor's-Lesson Plan-AOP-115.1, E.O._4,: page . . lAOP-115.1,."RHR Pump Vortexing", pag . KA 000025G010-(3.9/3.9).

000025G010 ..(KA's)

ANSWER: 071 (1.00)

c.- i+1.0)

REFERENCE:

1. . AOP-115.1, "RHR Pump Vortexing", p. . KA 000025A102 (2.9/2.8)

000025A102 ..(KA's)

ANSWER: 072 (1.00)

(+1.0)

>:

.A;

[:

l I-

.

'T , ~-s- , . , , .&

. . , . _ _ _ _ __ __ . . _ ._ .. _ _. , , . , . . _ . , . _ _ _ .. _

SENIOR-REACTOR OPERATOR , Page'94

REFERENCE: EOP-13.0, " Response to Abnormal Nuclear Power Generati'on", i

KA 000029G010 (4.5/4.5);

.y-1

,

-1

!

000029G010 ..(KA's) ,

..,

ANSWER: 073 (1.00) [+1.0]

REFERENCE: EOP-13, " Response To Abnormal Nuclear Power Generation", NOTE, page .- Instructor's Lesson Plan EOP-13, E.O. 6, page 1 .- KA 000029A114 (4.2/3.9)

.

000029A114 ..(KA's)

f

..

ANSWER: 074 (1.00)

a ._ - [ + 1. 0 ) . ,

n

.

1 e I

>

~

cSENIOR-REACTOR OPERATOR- Page 95 REFERENCE:

_

"

1. = -Technical Specification 3/4.9.2, Refueling Operations Instrumentation", page 3/4 9-2, KA 000032G008 (2.8/3.3).

000032G008 ..(KA's)-

ANSWER: 075 (1.00) [+1.0]

l REFERENCE:

1 Instructor's Lesson Plan'EOP-4.0, E.O. 5, page 6, KA 000037A203 (4.4/4.6)

000037A203 ..(KA's)

l ANSWER: 076 (1.00) - [+1.0)

,

c-T'W , , - .

. -.y.-. my

SENIOR REACTOR OPERATOR- Page!96 REFERENCE:

' li- FACILITY' EXAM BANK QUESTION.90 .c KA 000038A212: (3.8/4.2)

000038A212 . . (KA's) -

-r ANSWER: 077 (1.00)

e [+1.0)

REFERENCE: . Instructor's Lesson Plan E-15.4, "

Response to Steam Generator Low Level", E.O. 5, page 11, KA 000054K102 -(3.6/4.2)

000054K102 . . (KA's)

'

ANSWER: 078 (1.00) '

a .- [ + 1. 0)

'l

_

j p 9M F s er m e 5mM2- ty y- m- T- +O~s- + - w - - --

,.. . - _ . . . - .

.--.m~--~ .. . -. . . - -

t

TSENIOR: REACTOR OPERATOR - Page 97

..

4 REFERENCE:

-

,

cInstructor's-Lesson Plan EOP-15, E.O. 5, page'.8.-

'

-- 1 '. -

2.- KA 000054K3042- (4.'4/4.6)

-

,

+

000054K304 . . '( KA' s )

i :

LANSWER: 079 (1.00) i

,

c.- [+1.0)

.

REFERENCE: GS-9, " Radiation Monitoring System", E.O. 8, page 1 . KA 000061A101' (3.6/3.6)

.

000061A101 . . (KA's)

' ANSWER: 080 (1.00) (+1.0]

,

-REFERENCE:

'

1. - Instructor's Lesson Plan AOP123.1, E.O. 6,.page-5.

.

' KA 000036G010 -(3.7/3.8)

, ,

000036G010 . . (KA's)

g y -

pg --yg p

.. - - - - - . - - - _ . _ ~ - ~ . . - . . . . - -... . .

. SENIOR REACTOR OPERATOR Page 98'

ANSWER: 081 (1.00) ( + 1 '. 0 ]

' REFERENCE:

1. . GS-2, " Safeguards Power", E.O. 6, page 44, .KA 000056A247 (3. 8/3. 9 )-

000056A247 ..(KA's)

ANSWER: 082 (1.00) (+1.0]

jREFERENCE:

4 FACILITY EXAM BANK QUESTION 1097 GS-2, " Safeguards Power", E.O. 3, Figure GS . K7 000056A102 (4.0/3.9)

000056A102 ..(KA's)

.

E ANSWER: 083 (1.00) '

' - [+1.0)

l

L .

W

-

r m - e= ---r-a v ,- 9--,&- w v ., - . J -

. . _ . _ . . . _ . _ . _ . - . _ _ - _ -. . _ _ . . _

'

-SENIOR REACTOR' OPERATOR Page 99

REFERENCE: : AOP-220.1,'" Loss-Of Instrument Air", page 2, Instructor's Lesson Plan AOP-220.1,-E.O. 6, page 4.-

3.- KA:000065A206 (3.6/4.2_) 000065G010 ( 3 . 2 /3 .-3 ) . A206 000065G010 . . (KA's)

ANSWER: 084 (1.00) [+1.0)

' REFERENCE: ; HPP-153, " Administrative Exposura Limits", pages 1- . KA 194001K103 (2.8/3.4)

194001K103 . . - ( KA ' s )

(1.00)

-ANSWER: 085 Jd.- [+1.0)

s

>

f m ~a+_> ,

. . .. . . .. . ..~ , . - ~ -..-=.--- - - . . . - . . - . - - . . . . . . _ _ .

'

$ SENIOR REACTOR OPERATOR Page100:

'

-REFERENCE: SI 92-02, " Specific ~ Operations-Guidelines",,page l

'

_

- ,KA.194001K101 ( 3 . 6 / 3 .' 7 ) ,

.

194001K101 ..(KA's)

-

. . ANSWER : 086 (1.00) .[+1.0]

REFERENCE: ,

-- l . SAP-140, " Plant' Key-Control", page . KA 194001K105 - (3.1/3.4)

.

._

194001K105 ..(KA's)

s

ANSWER: -087 (1.00)

' [+1.0)

l'- ,

( ,

I

!

, .. ... . _ . . .. -

,

<

(SENIOR-REACTORIOPERATOR; ~-Page101=

-REFERENCE:

1.- : SAP 2153, " Independent Verification", page 6.- -

-3.-

KA 194001K101 (3.6/3.7)

i 194001K101 ..(KA's)

ANSWER: 088 (1.00)

' [+1.0)

REFERENCE:. .

a

1L SAP-201, " Danger Tagging", page . KA 194001K102 (3.7/4.1)

,

394001K102' . . - ( KA ' s )

ANSWER: 089= (1.00)

.b.-

[+1.0)

-

a

, SENIOR REACTOR' OPERATOR- Page102L

.

'

.:- REFERENCE:

1.: - SAPa300, " Conduct of Maintenance", page . > KA 194001A110 (2 ~ 9/3.9) .

194001A110 . . (KA's)

~ ANSWER:

_ 090 (1.00)

J [+1.0)

REFERENCE: SAP-300, " Conduct of Maintenance", pagesl17-1 . KA 194001A110 (2.9/3.9)

,

194001A110 . . (KA's)

..

- . ANSWER: 091' (1.00) [+1.0)

.

d

._ - ,

If7 -

T7 e" K- m'Pr-- *TM-+ -- +--I+ --e- W'- - * * e - --- -- - ^ - - - - - - - - -

_ SENIOR REACTOR OPERATOR Page103 REFERENCE: SAP-300, " Conduct of Maintenance", pages 10-11, KA 194001A110 (2.9/3.9)

194001A110 ..(KA's)

ANSWER: 092 (1.00) (+1.0)

REFERENCE: SAP-200, " Conduct Of Operations", Attachment 1, page . KA 194001K116 (3.5/4.2)

194001K116 ..(KA's)

ANSWER: 093 (1.00) [+1.0]

_ _ _ _ _ _ _ _ _ _ _ _ _ - _

.

i fSENZOR REACTOR._ OPERATOR Page104 I:

-REFERENCE::

1. - . EP-2, " Usage of Emergency Operating' Procedures", page 1 . KA -194001A102 ' (4.1/3.9)

194001A102 ..(KA's)

ANSWER: 094' (1.00) [+1.0)

. . . _ .

-REFERENCE:

.. EP-2, " Usage of Emergency. Operating Procedures", page 22 2.- KA 194001A102 ( 4 .1/ 3 . 9 ) -

-

l194001A102 ..(KA's)

.;

!

ANSWER: 095 (1.00)

, [+1.0]

.

,

'

.

'

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--

m SENIOR' REACTOR OPERATOR . Page105'

REFERENCE:

.1 '. CR-6, " Protection Against Radiation",= Enabling Objective:14, page-3 .- KA 194 001K103__ . (2. 8 /3. 4 )

194001K103 ..(KA's)

ANSWER: 096 (1.00) [+1.0] :

' REFERENCE: GOP-APPENDIX A, " General Operating Precautions", page 6.

, KA 194001A102 (4.1/3.9)

.

194001A102 ..(KA's)

L ANSWER: '097 (1400)

l ( + 1. 0 ] -

4 -

'l

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SENIOR REACTOR. OPERATOR' Page10.6_ ,

REFERENCE

' l '. --

-

SAP-200, " Conduct Of' Operations", page 11'.

.2.- KA 194001A103 - (2.5/3.4 ) .

~194001A103 ..(KA's)

ANSWER: 098 (1.00) * [+1.0]

REFERENCE: SAP-139, " Procedure Development, Review, Approval, and'

Control",.page 2 '. 2 . SO 92-01, page . KA 194001A101 (3.3/3.4)

"

194001A101 ..(KA's)

ANSWER: 099- (1.00)

. [+1.0)-

REFERENCE:

,1.: SI 92-02, page ! KA-194001K107 (3.6/3.7)

,,

194001K107 ..(KA's)

. .2,. . . - . . _ . - . _ . - - . - . . _

. . . . .. . _ .- ..

--

' SENIOR REACTOR OPERATOR Page107:

- ANSWER: 100 (1,00)

-;c - (+1,0]

. REFERENCE: SI 92-02, page 6-2,- .KA 194001K102 (3,7/4.1)

-194001K102 . (KA's)

.

T

,

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

V

- - -

-i SENIOR REACTOR OPERATOR Page 1 AN'SWER K'E Y-

,

i MULTIPLE CHOICE 023 -c

.001 d 024 d ,

l 002 c 025 c-1 003 b 026 b '

.. !

004 b 027 b-005 c 028 c 006 d -029 b 007- b or a 030 c 008 d 031 b 009 c 032 c 010 c 033- c

011 d 034 d 012 c 035- b'

013 a 036 a

.014 d 037 d 015 b or d 038 d 016 a 039 d

~

..017 c 040 c 018 d 041 d 019 b 042 d 020 a 043 c-021 tr 044 c 022 a 045 a-

.

_ _

'

g -

m -

. . . . . . . . . . .. ., . .. .,.- -...-...---- .-. -

.

, , --. , - .

SENIOR REACTOR OPERATOR Page- 2 J

A N s WLE RL KEY ,

046 a 069 a LO47- c 070 c-048 d 071- c ,

.

049 a 072 c 073 a ,

051 a 074 a-052- d 075 c 053 b 076- d-054 b 077 c >

-055 c 078 a 056 c 079 c t

. -057 b or a 080 b L

h 058 c 081- a 059 a- -082 c-

-060 b 083- d

'

-061 c 084 .d 062 b 085 d

'063 a

064- b- 087 c 0651 b- -088 b 066- .a- 089 b

'067' d- 090 a'

068' b 091 c L 3

,

.- . . , . . . .. . . . ..

.. . . . .- . - . . -- . . . ~ .. .~ ..

(SENIOR REACTOR OPERATORi Page- :3'

.;

ANSWER KEY

,

092 d

.093 b 094 d 095 b 096 c 097 c 098 c 099 c 100 c

i a

.

--

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

-

. - __ . _ _ .

-

. . . . _--__. _ _ - - ___ . .. . . . .

TEST CROSSIREFERENCE- .- Page - l'

SRO Exam P W-R R Le a c t- o -r-O r.g a n i.z e d b. y Question N u m b e-r QUESTION VALUE REFERENCE 001 1.00 8000013 002 1.00 8000027 003 1.00 8000017 004 1.00 8000021 005 1.00 8000083 006 1.00 8000137 007 1.00 8000025 008 1.00' 8000070 009 1.00 8000014 010 1.00 8000088 011 1.00 8000080-012- 1.00 8000085 013 1.00 8000082 014 1.00 8000132 015 1.00 8000001_

016 1.00 8000086 017- 1.00 800001 .00 8000047 019 1.00 8000079 020 1.00 8000073 021 1.00 8000074 022 1.00- 8000075 023 1.00- 8000028:

024 1.00 8000078 025 1.00 8000077 026 1.00 8000064 027 1.00 8000108-028 1.00 8000053 029 1.0 .00 8000066-031 1.00 8000056 032 1.00 8000060 03 .0 .00 8000129 036 1.00 8000062-037 1.00 8000057 038 l '. 0 0 ' 8000067!

039 1.00 8000058'

. 040 1.00- !800005 e041 1.00 8000009 042 1.00- 8000010 043 1.00 8000110

044- 1.00 8000089-1 045 1.00 8000090-

-046 1.00- 8000113

<

047 1.00 8000106-048' 1.00 8000122-

,. .. . - .

-. .. . . . ~ . .. ~..-. . . - ., - . .. ..- . .. ,, . . . . . . . . . . . . . . - . . - ~ . . . . . . . - .

,

049 - 1,00 =8000127'_-

,

, ..

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k

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y 3 g _

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w ., ,

.

TEST CROSSjREFERENCE= Page 2'

S R O- Exam PWR - R e a c t: o r O r g a n i=zie d by Q u e sLt i o n- N u m.b-e-r-QUESTION VALUE REFERENCE 050 1.00 8000118 051 1.00 8000126 052 1.00 8000119 053 1.00 ~8000116 054 1.00 800009 .00 8000115 056 1.00 8000012 057 1.00 8000121 058 1.00 8000123 059 1.00 8000006 060 1.00 8000096 061 1.00 8000008 062 1.00 8000092 063 1.00 8000093 064 1 00 8000016 065 1.00 8000117-066 1.00 8000124 ;

067 1.00 8000125-068 1.00 8000107 069 1.00- 8000003 070 1.00 8000097 071- 1.00 8000098 072 1.00 8000005 073 1.00 8000104 074 1.00 .8000105 075 1.00 -8000091-076 1.00 8000111-077 1.00 8000094 078 1.00 -8000095~

079 1.00 8000065 080 '

' 1.~ 0 0 '

-

8000101

, 081 1.00 -8000103-082 1.00- 8000120-083 1.0 :

084 1.00 '8000031

>

085 1.00 8000032:

086 '1.00- 8000034_ .

'

-087 1.00 :8000035- "

088- - 1 .10 0 8000036 089 1.00 8000037 0901 1.00 8000038 09 '

1.00

'092 1.00 8000040 093 1.00 8000041 094_ 1.00 8000042 095, 1.00 8000044 096 1.00 8000046 097: 1.00 8000048 ,

, . . . . . ,

1. *#-A 3-- M,.m Jd,eN1J *i6.2,,MiJ iMwM<* 2 4Em.4_a ,W_pm.. a P

- . 'h-

.

098- 1.00 - 8000049

.

I NI

,

t J

t b

b I

f L

,

i

,k

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-

4 3 4 m - a- 2. _ , _h !_'-___

. .. . .. . . . . . - . . . . ..- . -. . - .. . - . - .-.. -_ ..

,

.

TEST CROSS REFERENCE Page 3

S R-O-

-

E x a-m_ P W R' R:e a c t o r-

.

l

'

~

0 r g a;n'i z e'd by- Quest-ion N u'm b e r r

QUESTION- VALU REFERENCE-

.__

099 - 1,00 8000050 100 - 1.00 8000051

......

'

,

100.00

. .....

& M . .e e 100.00

.

h h

!

-

a r

..

m-P

'

9 v -, m4 c' , --,4 . *s- -, m --

TEST CROSS REFERENCE Page 4 SRO Exam PWR React or Organized by KA Group PLANT WIDE GENERICS QUESTION VALUE KA 098 1.00 194001A101 094 1.00 194001A102 096 1.00 194001A102 093 1.00 194001A102 097 1.00 194001A103 091 1.00 194001A110 090 1.00 194001A110 089 1.00 194001A110 085 1.00 194001K101 087 1.00 194001K101 100 1.00 194001K102 088 1.00 194001K102 095 1.00 194001K103 084 1.00 194001K103 086 1.00 194001K105 099 1.00 194001K107 092 1.00 194001K116

......

PWG Total 17.00 PLANT SYSTEMS Group I QUESTION VALUE KA 002 1.00 001000G008 001 1.00 001000K602 004 1.00 003000A202 006 1.00 003000G005 003 1.00 003000G007 005 1.00 003000G010 008 1.00 004000K118 007 1.00 004010A211 009 1.00 013000G005 025 1.00 014000A102 010 1.00 015000G007

011 1.00 017020K401 012 1.00 059000K419 014 1.00 061000K301 013 1.00 061000K402 016 1.00 068000A404 015 1.00 071000A413 017 1.00 071000A429 ,

--_ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

......

PS-I Total 18.00

. . . . , , . . . . . . _ _ . . . . . .

TEST CROSS REFERENCE Page 5 S R O- Exam PWR Reactor-Organized by KA Group PLANT SYSTEMS Group II

!

QUESTION VALUE KA j j'

019 1.00 002000A403 018 1.00 002000Gi O5 021 1.00 006000G019 022 1.00 006020A107 '

020 1.00 006030K403 ;

'

023 1.00 010000A107 024 1.00 012000A406 035 1.00 -028000G004 034 1.00 028000K501

"

076 1.00 033000A101 027 1.00 033000G006 036 1.00 034000K403 028 1.00 039000K102 029 1.00 064000K411 *

030 1.00 073000K401 031 1.00 075000K102 I

'

032 1.00 086000G009

......

PS-II Total 17.00 Group III QUESTION VALUE KA

.-

033 1.00 008000K101 037 1.00 045000A305 .

038 1.00 076000KiO9 !

040 1.00 078000K201-039 1.00 -078000K402

......

PS-III Total 5.00*

......

PS Total- 40.00-l_ EMERGENCY PLANT EVOLUTIONS

'

Group I QUESTION VALUE KA

-- .

?

--. + , .....,,m~ . ,~, ,- ,-#.

_ - _ _ _ _ _ _ _ _ _ _ _ _ -

062 1.00 000001G010-

-063 1.00 000003G010 042 1.00 000005G003

.

_ _ .-. _

E

_

' . .... ~,___ L _._..w__ _- - --- 2

_ . . . . .

TEST CROSS REFERENCE Page 6 SRO Exam PWR Reactor Organized by KA Group EMERGENC' JT EVOLUTIONS Group I QUESTION VALUE KA 041 1.00 000005G010 068 1.00 000011A103 -

045 1.00 000024A202 044 1.00 000024K301 043 1.00 000026A204 073 1.00 000029A114 072 1.00 000029G010 048 1.00 000040A204 047 1.00 000040G007 049 1.00 000051A202 050 1.00 000055A102 051 1.00 000055G010 052 1.00 000057A214 053 1.00 000067G002 055 1.00 000068A211 054 1.00 -000068K318 057 1.00 000069A201 056 1.00 000069A202 058 1.00 000069G011 060 1.00 000074A207 059 1.00 000374K311 061 1.00- 000076K305-

.. . . . . .

EPE-I Total 25.00 Group II QUESTION- VALUE KA 066 1.00 000007G012 065 1.00 000007K103 064 1.00 000007K301-067 E1.0 G011-071 1.00? 000025A102 070 1.00- 000025G010 069 1.00- 000025K101-046 1.00 000027A215 074 1.00 000032G008 075 1.00 000037A203 076 1.00- 000038A212 077 1.00- 000054K102-078 1.00 000054K304 079 1.00 000061A101-l

_ --__ _ - _ - _ - - . _ - _ _ _ _ . _ _ _ _ _ _ _ . . _ _ _ _ . _ . _ _ . _ _ _ . _ _ _ _ _ _

083 1.00 000065A206

......

%

hm .. .. . . . .... ... . _ . . . . . ___ ____ . _ _ __ a

i

'

TEST CROSS REFERENCE Page 7 ,

'

,

!

SRO Exam PWR Reactor

.- Organized by KA Group i

!

.

EMERGENCY PLANT EVOLUTIONS Group II i

QUESTION VALUE KA

EPE-II Total 15.00 .

-;

Group III ;

QUESTION VALUE KA 000 1.00 000036G010 t 082 1.00 000056A102 081 1.00 000056A247 EPE-III Total 3bb ,

......

,

......

EPE Total 43.00

......

...... -

Test Total 100.00 #

'

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-

NRC Official-Use Only y, 0 Sun <mt"92-301 M a s {e s-. ACek~

,

Nuclear Regulatory Commission Operator Licensing Examination-l

-. ,

i

,

This document is removed from l' Official-Use'Only category'on date of examination, i

l NRC Official Use-Only

.. i i

- I (-

- .. _

U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION REACTOR OPERATOR LICENSE REGION 2 CANDIDATE'S NAME:

FACILITY: V. C. Summer 1 REACTOR TYPE: PWR-WEC3 ,

i DATE ADMINISTERED: 92/12/07 INSTRUCTIONS TO CANDIDATE:

Use the answer sheets provided to document your answers.- Staple thisicover-sheet cx1 top of the answer sheets. Points for each question are indicated i parentheses after the question. The passing grade requires a final grade of at-least 80%.- Examination papers will be picked up four _(4) hours af ter the examination start CANDIDATE'S-TEST VALUE SCORE %'

_

100.00- % TOTALS-FINAL GRADE

.

All_ work done on'this examination is my ow I have neither given nor

received ai _

Candidate's. Signature

,

?

i ..

I O

. -- - - - - y REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice)

If you= change your answer, write your selection in the blan 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 003 a b c d 026 a b c d

_

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 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 014 a b c d 037 a b c d 015 a b c d ___

038 a b c d 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 c d

'

020 a b c d 043 a b .c d 021' a b c d 044 a b d-022 a b c .d 045' a b c d L

,_ _ .

. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

REACTOR OPERATOR Page 3 ANSWER S 11 E E T Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blan a b c d 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 059 a b c d 082 a b c d 060 a b c d 083 a b c d 061 a b c d 084 a b c d 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 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 068 a b c d 091 a b c d L ;

~. .

REACTOR OPERATOR Page 4 ANSWER SHEET Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blan a b c d 093 a b c d 094 a b c d 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 100 a b c d (********** END OF EXAMINATION **********)

l

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4 .. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ _a

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Page 5 NRC RULES AND GUIDELINES FOR LICENSE EXN4INATIONS During the administration of this examination the following rules apply:

1. Cheating on the examination means an automatic denial of your application and could ret' tit in more seve*e penaltie . After the examination has been coupleted, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. This must be done after you complete the examinatio . Restroom trips are to be limited and only one applicant at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provide USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG . Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question pag . Use abbreviations only if they are commonly used in facility literatur Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answe Write it ou . The point value for each question is indicated in parentheses after the questio .

10. Show all calculations, methods, or assumptions used to obtain an answer to any short answer question . Partial credit may be given except on multiple choice questions. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN , Proportional grading will be applied. Any additional wrong information that is provided may count against yo For example, if a question is worth one point and asks for four responses, each of which is worth 0.25 points, and you give five responses, each of your responses will be worth 0.20 point If one of your five responses is incorrect, 0.20 will ba deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answer . If the intent of a question is unclear, ask questions of the examiner only.

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I Page 6 14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheet In addition, turn in all scrap pape . Ensure all information you wish to have evaluated as part of your answer is on your answer. shee Scrap paper will be disposed of immediately following the examinatio . To pass the examination, you must achieve a grade of 80% or greate . There is a time limit of four (4) hours for completion of the examinatio . When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoke 'l l

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REACTOR OPERATOR Page 7 i

QUESTION: 001 (1.00)

WHICH ONE (1) of the following is the source of the Tref signal? It is calculated from auctioneered high T-av It is calculated from NI normalized power level, It is calculated from averaged T-ho It is calculated from turbine first stage pressur QUESTION: 002 (1.00)

WHICH ONE (1) of the following conditions will result in an URGENT- ,

FAILURE alarm for the Rod Control System? Loss of a 24VDC power supply Loss of an auxiliary power supply A blown fuse Failure of the slave cycler timer QUESTION: 003 (1.00)-

WHICH ONE (1):of the following is the reason-that all the RCP i breakers open on underfrequency? To protect the RCP anti-rotation device from damage due to abnormal coastdown, To preserve the RCP flywheel kinetic energ To avoid water harmer-transients in the RCS induced-by rapid RCP speed chang , To reduce the-probability of a' stress induced RCP sheared'

shaft acciden ;

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REACTOR-OPERATOR Page 8 QUESTION: 004 (1.00)

The following plant conditions exist:

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The Unit is in MODE Reactor Coolant pump A was started at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months and secured at 0920 hours0.0106 days 0.256 hours 0.00152 weeks 3.5006e-4 months to repair an oil lea At 0930 hours0.0108 days 0.258 hours 0.00154 weeks 3.53865e-4 months Maintenance personnel report that the repairs.are complet WHICH ONE (1) of the following actions should be taken following a request from maintenance personnel to start RCP A?

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a, .The RCP can be started'immediately with the present conditions in order to perform a leak check of the oil syste .. The RCP can be started with the present conditions at --

0950 hour0.011 days 0.264 hours 0.00157 weeks 3.61475e-4 months The RCP can be started immediately provided PCV-145, LO PRESS LTDN is placed in MA The ""P can be started at 0940 provided RCS temperature-is 2?d degrees F. more than seal injection _ temperatur _

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QUESTION: 005- (1.00)

The following plant conditions exist: ;

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The Unit is in MODE .;

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RCS temperature is 210 degrees '

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A bubble exists in-the Pressurize RCP "A" is RUNNIN '

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The Component Cooling Water System supply to the Reactor Coolant Pumps (RCPs) has been isolated for repair of a lea WHICH ONE (1) of the following is the reason Seal Water injection .:

is REQUIRED to be in service to the Reactor Coolant pumps (RCPs) l per SOP-101, " Reactor Coolant System a ? ! The RCS is FUL An RCP is RUNNIN ' A bubble is established in the Pressurize f-i RCS temperature is 210 degrees r i

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QUESTION: 006 (1.00)

The following plant conditions exist ;

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The Volume Control Tank (VCT) level is at 39%.-

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Automatic make-up is in progres A leak develops-in the reference' leg associated withithe i automatic level controller _(LT-112).

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Assume no operator action.- ;

LWHICH ONE (1) of the following_ describes the FIRST system response?; .LCV-115A begins to diver Automatic makeup to.the VCT will STOP.- Indicated VCT-(LT-112)-level will DECREAS ' Charging pumps will-automatically shift suction to the RWS t

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_ REACTOR-OPERATOR lage 10 QUESTION: 007 (1. 00)

WHICH ONE (1) of the following leaks could result in a DILUTION of the-RCS? Regenerative Heat Exchanger leak Letdown Heat Exchanger leak Reactor Coolant Drain Tank Heat Exchanger leak Seal Water Heat Exchanger leak QUESTION: 000 (1.00)

WHICH of.the following Charging pumps is assumed to be OPERABLE during a control room fire?

a- "A" Charging pump ONLY b.- "B" Charging pump ONLY "A" and "C" Charging pumps "B" and "C" Charging pumps

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-REACTOR OPERATOR '/ age 11 QUESTION: 009 (1.00)

The following plant conditions exist

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The reactor is in MODE It has been determined that BOTH SI trains automatic actuation logic and actuation relays did NOT meet the 1 surveillance acceptance criteria for the test previously ;

conducted in MODE 5, but the manual initiations passed :

the surveillance WHICH ONE (1) of the following actions should be taken? ! Stay in MODE 4 and do NOT allow transition to MODE Continue with the plant startup, no restraints appl Enter and comply with Technical Specification 3. j Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months verify the actuation relays are in the required state for the existing plant conditio t QUESTION: 010 (1.00)

WHICH ONE (1) of the following is the design interlock or operating practice that is used to prevent'another automatic ESP actuation :

following an automatic ESF actuation and subsequent reset? '

Manually blocking SI from the control boar Placing-the ESF actuation-trains in tes F

! The P-4 interlock, actuated by_the opening of the reactor trip _ breakers, The seal-in feature of the reset circuitry, which disarms-auto actuatio .s t

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QUESTION: 011 (1.00)

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The following plant conditions exist i

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The reactor is operating at 9% power, conducting a plant startu ,

- Control systens are in their normal system alignmen ,

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The output of Power Range NI N 44 gradually fails HIGH.'

- Main Feedwater Regulating valves are in MANUA Main Feedwater Regulating Bypass valves are in AUTOMATI Assume no operator actio WHICH ONE (1) of the following is the INITIAL plant response to this failure? Control rods automatically insert, Steam generator levels increas CMPTR DELTA FLUX LMT EXCEEDS annunciator-illuminate ; Auto rod stop at 20% auctioneered NI powe QUESTION: 012 (1.00)

WHICH ONE (1)-of the following systems receives an input from the Power Range Nuclear Instrumentation? Rod Position Indication System Audio Count Rate-System Steam Generator Water Level-Control System , High Flux At Shutdown Alarm System-

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REACTOR OPERATOR Page 13 ;

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QUESTION: 013 (1.00) +

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Each of the following is an input to the Core Subcooling Monitor EXCEPT: Wide range hot leg RTD Wide range cold leg RTD j Narrow range Pressurizer pressure Average Core Exit Thermocouple temperature ;

QUESTION: 014 (1.00)

WHICH ONE (1) of the following supplies cooling to the Reactor Building Cooling System during a Loss Of Coolant Accident (LOCA)? Component Cooling Water System Service Water System Industrial Cooling Water System d.- Chilled hater System

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. QUESTION: 015 (1.00)

WHICH ONE (1) of the-following conditions.will trip ALL Feedwater-

-pumps?

a.- HIGH feedwater-pump discharge pressure as indicated on-PS - 3242 ( A)- Steam Generator levelHof 80%-on "A" Steam Generator Safety Injection Signal:on Train "A" Deaerator Storage "ank . level-- of two - (2) feet

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REACTOR OPERATOR , Page 14 j

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i QUESTION: 016 (1.00)

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WHICH ONE (1) of the following conditions is indicated by the AMBER

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l light located above the Feedwater Booster pump control switch being i LIT? +

-f The pump is OPERATIN ; LOW oil pressure condition exist , Overcurrent condition exist d. . The pump is TRIPPED due to HIGH Steam Generator leve '

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QUESTION: 017 (1.00) .

WHICH ONE (1) of the following conditions will result in CLOSING

.all Feedwater Isolation valves?

, Feedwater temperature is 220 degrees All Steam Generators indicate 13% leve ! Feedwater Forward Flushing valve (XVG-1689A)-is 50% OPE ; HIGH-HIGH Reactor Building sump leve .

QUESTION:-018 (1.00)

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WHICH ONE (1) of the'following serves as a BACKUP Water source to the Emergency Feedwater Supply? Service Water System Demineralized Water Storage System i Circulating Water System Fire Protection Water System-

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QUESTION: 019 (1.00)

WHICH ONE (1) of the following conditions will AUTOMATICALLY start ,

a Motor Driven Emergency Feedwater pump with the reactor at 50% '

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power?

, "A" Steam Generator level is 15%. l "A", "B", AND "C" Steam Generator levels are 18%. Zero (0) Volts on Bus 1EA Trip of "A", "B", OR "C" Main Feedwater pumps QUESTION: 020 (1.00)

The following plant conditions exist:

1 - The Unit has tripped from 100% powe .

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Emergency Feedwater flow (EFW) CANNOT be establishe Steam Generator levels. indicate 5% wide range-

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leve .

- . Steam Generator pressures indicate 500 psi ,

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All Reactor Coolant pumps have been TRIPPE WHICH ONE (1) of the following methods of RCS cooling is ,

IMMEDIATELY available? Natural circulation. cooling established _by low pressure steaming through the Steam Generator Power-Operated Relief valves-(PORVs). Natural circulation cooling established by Condensate System makeup to the Steam Generators, Natural circulation cooling established by Main Feedwater_ ,

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makeup to the Steam Generator Bleed and feed cooling of the RCS_usingfthe CVCS-System and two (2) opened Pressurizer Power-Operated Relief valves (PORVs). _ _ _ _ __ b

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REACTOR OPERATOR Page 16 QUESTION: 021 (1.00)

The following plant conditions exist:

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A waste gas release was in progres Waste Gas Discharge Radiation Monitor, RM-A10, is in servic HCV00014, Vent Stack Hand Control Valve has tripped shu A new analysis and release permit have been requeste WHICH ONE (1) of the following is the MINIMUM action necessary for the waste gas release to continue? HCV-00014, Vent Stack Hand Control Valve, must be reset locally at the valv HCV-014, Waste Gas Discharge Control Valve controller must be taken to ZERO then the valve re-opene HCV-014, Waste Gas Discharge Control Valve selector switch must be cycled to closed then re-opene HCV-014, Waste Gas Discharge Control Valve selector switch must be cycled to CLOSED then re-opened AND HCV-0014, Waste Gas Discharge Control Valve controller must be taken to ZER )

QUESTION: 022 (1.00)

WHICH ONE (1) of the following AUTOMATIC actions result from a HIGH radiation condition on Liquid Waste Effluent Monitor, RM-L5? Liquid radioactive waste discharge control valve (RCV-018) CLOSE Waste Monitor Tank pumps TRI Liquid waste flow is DIVERTED to the Nuclear Blowdown Monitor tan Liquid Effluent To Fairfield Penstocks valve (PVD-6910)

CLOSE REACTOR OPERATOR Page 17 QUESTION: 023 (1.00)

WHICH ONE (1) of the following AUTOMATIC actions results from a HIGH radiation condition on Reactor Building radiation monitor, R G17A? Train "A" AND "B" Purge exhaust fans STO ONLY Train "A" Purge exhaust fan STOP Reactor Building Purge discharge valves XVB-1A AND XVB-2A CLOS ONLY Reactor Building Purge discharge valve XVB-1A CLOSE QUESTION: 024 (1.00)

The following plant conditions exist:

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The reactor is in MODE 2 at 3% power and slowly-increasing.-

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Tavg indication is 555' degrees Tavg/ Tref deviation alarm is NOT rese WHICH ONE (1) of the following actions is the MINIMUM necessary to comply with Technical Specification 3.1.1.4, " Minimum Temperature For Criticality"? Restore Tavg to greater than 557 degrees within 15 minutes.

> i Restore Tavg to greater than 557 degrees within 30 ;

minute ' Determine RCS-Tavg is. greater than minimum required a least once per 45 minute Determine RCS Tavg is greater than minimumLrequired at least once per 30 minutes, t

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REACTOR OPERATOR Page 18 QUESTION: 025 (1.00)

The following plant conditions exist:

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The Unit has trippe Offsite power has been los Upper range Reactor Vessel Level Indication System (RVLIS) indicates G5%.

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Narrow range RVLIS indicates 50%.

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Wide range RVLIS indicates 15%.

WilICll ONE (1) of the following is the current level / condition in the Reactor Vessel? % %- t (full) but voids are present, Four (4) inches above the Hot Leg _

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REACTOR OPERATOR Page 19 i

QUESTION: 026 (1.00) ,

The following plant conditions exist: ,

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The Unit has tripped from 100% powe Safety Injection has actuate .

- Train "A" and "B" Safety Injection Reset switches have been placed in the RESET positio RWST level is 20%. '

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Reactor Building sump level indicates two (2) fee AUTOMATIC swap over of RHR suction to the Reactor Building sump has NOT occurred.-

WHICH ONE (1) of the following is.the reason AUTOMATIC swap over of RHR suction to the Reactor Building sump has NOT occurred? RWST level must be less than eighteen (18) percen , Train "A" and "B" Safety Injection has been RESE NORMAL / RESET switches are in the RESET position, Reactor Building sump level must be five (5) feet, f QUESTION: 027 (1.00)

WHICH ONE (1) of the following-"A"-Accumulator parameters needs to be corrected prior to declaring the "A" Accumulator OPERABLE while the reactor is operating at 100% power? Volume is 7500 gallon Boric acid concentration is 2015 pp Pressure is 650 psi ~ Isolation valve (MVG-8808)' is OPEN- and the RED indicating ,

light-is ON.

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QUESTION: 028 (1.00) i To WHICH ONE (1) of the following systems / locations is an ECCS Accumulator vented when lowering pressure? : Reactor Building atmosphere ] Reactor Coolant Drain Tank (RCDT) Pressurizer Relief Tank (PRT) Reactor Building Purge System

, QUESTION: 029 (1.00)

The following plant conditions exists

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The reactor is at 100% powe Pressurizer heatera are OF Normal pressurizer spray valves are MODULATIN Pressurizer PORVs are CLOSE i

.WHICH ONE (1) ' of the following RCS pressures is appropriate for the above conditions? psig psig psig psig -

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m-=-+-peea=w'---myesq g p e

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REACTOR OPERATOR Page 21 QUESTION: 030 (1.00)

The following plant condition exists:

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A MANUAL reactor trip has been initiated by use of the spring loaded control switche The red indicating lights below the control switches are DE-ENERGIZE The green indicating lights below the control switches are ENERGIZE WHICH ONE (1) of the following conditions exists? The shunt trip coil is DE-ENERGIZED and all Reactor Trip breakers are CLOSE The undervoltage coil is ENERGIZED and all Reactor Trip breakers are OPE The shunt trip coil is ENERGIZED and all Reactor Trip breakers are CLOSE The undervoltage trip coil is DE-ENERGIZED and all Reactor Trip breakers are OPE .

,; ION: 031 (1.00)

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WHICH ONE (1) of tne follcwing is the system accuracy for the Rod Position Indication System when the COIL B output is lost? +10, -10 steps , -4 steps +4, -10 steps , +10 steps

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REACTOR OPERATOR Page 22 QUESTION: 032 (1.00)

WHICH ONE (1) of the following will provide an OPEN signal to the Reactor Building Spray Header Isolation valves, MVG-3003A & B? Main Steam Isolation Signal Safety Injection Signal Containment spray Actuation Signal Reactor Trip Signal 1 QUESTION: 033 (1.00)

The following plant conditions exist: '{

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The Unit has tripped from 100% pows The Emergency Core Cooling System has ACTUATE Reactor building pressure is 4 psig and INCREASING rapidl Operators have attempted to initiate Reactor Building Spray by actuating ONE (1) switch from each set of Reactor Building Spray Actuating switches on the MC The Reactor Building Spray System has NOT actuate .

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WHICH ONE (1) of the following is the reason Reactor Building Spray

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has NOT actuated? Spray System Isolation valves are BLOCKED from OPENING until Reactor Building pressure reaches 12.05 psig, RWST suction valves to the Reactor Building Spray pumps do not OPEN until Reactor Building pressure reaches 12.05 psi Reactor Building Spray pumps are BLGCKH3 from STARTING until Reactor Building pressure reaches 12.05 psi BOTH switches from ONE (1) set of Actuating switches have to be ACTUATED simultaneously to initiate Reactor Building Spray flow.

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l REACTOR OPERATOR Page 23-QUESTICM: 034 (1.00)

WHICH ONE (1) of the following provides temperature control of the Spent Puel Cooling and Transfer Loop? AUTOMATIC control of Component Cooling Water MANUAL control of Component Cooling Water AUTOMATIC control of Service Water MANUAL control of Service Water QUESTION: 035 (1.00)

Each one (1) of the following is a concern when level is reduced BELOW minimum in the Spent Fuel Cooling System EXCEPT: Minimize radiation _ levels at the pool surface,

_ Minimize the heat load on the Spent Fuel Pool Heat Exchanger (s) .

' Minimize velocity of.a dropped spent fuel shipping cask, Minimize the amount -of ' iodine released from a ruptured

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fuel assembly.

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REACTOR OPERATOR Page 24 l

l QUESTION: 036 (1.00)

WHICH ONE (1) of the following AUTOMATIC actions occur when HIGH activity is sensed by Radiation Monitor, RM-L3, in the Steam Generator Blowdown line? Reactor Building Isolation valves, PVG-503A, B, and C, CLOSE to isolate blowdcwn flow, Blowdown Flow Control valves, IFV-4701A, B, and C, CLOSE to isolate blowdown flo Discharge Isolation valve, XVT-524, CLOSES and Isolation -

valves, XVT-525 and XVT-541, OPEN to discharge blowdown to the Nuclear Blowdown Holdup tan Discharge Isolation valve, XVT-524, OPENS and Isolation valves, XVT-525 and XVT-541, CLOSE to divert blovdown to the Nuclear Blowdown Monitor tan QUESTION: 037 (1.00)

The following plant condition exists:

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RCS Pressure is 2050 psi WHICH ONE (1) of the following will cause the Main Steam Isolation

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Valves (MSIV) to close automatically? Steamline pressure in "A" mainsteam line - 725 psig, Containment Pressure - 6.5 psi Low-low water level on one steam generator AND P-1 P-1 AND Tave less than 540 degrees REACTOR OPERATOR Page 25 QUESTION: 038 (1.00)

WHICH ONE (1) of the following describes the FAILURE mode of the Steam Generator Power-Operated Relief valves, PORV-2000, 2010, and 2020, on a loss of the selector control signal from the Steamline Power Relief Mode (PWR RLF/ AUTO)

switch? OPEN CLOSED Steam Dump Mode - Overpressure Mode QUESTION: 039 (1.00)

WHICH ONE (1) of the following describes the normal power supply to the 7.2 KV electrical buses 1A, 1B, and 1C? Directly from the unit auxiliary transformer (XTF-2)

which is supplied from the Main Generator, Directly from the main transformer (XTF-1) which is supplied from the switchyard,

~ Directly from the unit auxiliary transformer (XTF-2)

which is supplied from the switchyar Directly from the main transformer (XTF-1) which is supplied from the Main Generato _ ._-

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REACTOR OPERATOR Page 26 QUESTION: 040 (1.00)

WHICH ONE (1) of the following is the MAXIMUM length of time that the 125 VDC Batteries are designed to supply Class 1E loads during a station blackout? hours hours hours hours -

QUESTION: 041 (1.00) ,

WHICH ONE (1) of the following will generate a shutdown signal when in the TEST mode for the ESF Diesel Generator? Jacket water temperature - 192 degrees F High lube oil temperature - 172 degrees F High crankcase pressure - 0.6 inches water vacuum Lube oil pressure - 65 psig

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QUESTION: 042 (1.00)

WHICH ONE (1) of the following is the type of fire protection provided in the Diesel Fire Pump' room? Dry pipe reaction system Wet pipe sprinkler system Halon 1301 system CO2 extinguishing system i

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REACTOR OPERATOR Page 27 QUESTION: 043' (1.00)

The following plant conditions exist:

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A fire has occurred requiring flow from either the electric fire pump or the diesel-fire pum The electric fire pump has failed to star IPS-4911, Diesel Fire Pump Discharge Pressure Switch, is INOPERABLE ..

WHICH ONE (1) of the following actions will start the diesel fire pump from the local control panel? Place the mode selector switch in the TEST position to open the discharge drain valv , Place the mode selector switch in OFF/ RESET and depress the START pushbutton, Place the mode selector switch in MAN 1 and depress the START pushbutto Ensure the mode selector switch is in OFF/ RESET, then cycle the feeder breaker to XSW-1C ,

QUESTION: 044 (1.00)

WHICH ONE (1) of the following describes how the RCS cooldown rate-is controlled when using RHR to perform a normal cooldown of the RCS? Both CCW flowrate and RHR flowrate passing:through the RHR heat exchange . CCW flowrate through the RHR heat exchanger, while maintaining RHR flowrate constant at 3750;gpm.

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c.- RHR flowrate passing through the RHR heat exchanger,'

while. maintaining total RHR system flowrate constant at-3750 gp Total RHR system flowrate is controlled at a value between 500 gpm and 4000 gp >

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-QUESTION: 04 (1.00).

The following plant conditions exist:

---The Unit is in MODE '

- RHR is in service 1 utilizing SOLID' pressure contro WHICH ONE (1) of the following component' failures will result in a

- HIGH pressure condition iri the - RCS? . Letdown Flow Control valve HCV-142 fails CLOSED and RHR ^

suction valves fail-to CLOS __ Letdown Flow Control. valve HCV-142 fails OPEN and RHR suction relief valves fail to OPE Pressure Control valve PCV-145 fails CLOSED and RHR suction relief valves fail to OPE Pressure Control valve PCV-145 fails OPEN and RHR suction valves fail to CLOS f

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REACTOR. OPERATOR- -Page 29 QUESTION: 046 (1~. 0 0 )

The following plant conditions exist:

-- The Unit is operating at 100%-powe PORV PCV-445A is leaking steam into the Pressurizer Relief Tank (PRT) .

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PRT temperature is 130 degrees WHICH ONE (1) of the following actions are required per SOP-101,

" Reactor Coolant System"? Monitor PRT temperature until it reaches 180 degrees then aligS the Reactor Makeup Water System-to makeup to the PRT chrough the spray header to reduce the temperatur Align the Reactor Makeup Water-System to-makeup _to the-PRT through the spray header to reduce the temperature, Monitor PRT temperature-until it reaches 180 degrees then align the~ Reactor Coolant _ Drain Tank 1(RCDT) pumpsfto the PRT to circulate water through the heat exchangers and back to the spray header to reduce temperature, Align the Reactor Coolant Drain Tank ~(RCDT) pumps to the PRT to circulate water through the heat exchangers and back to the. spray header to reduce temperatur QUESTION: 047 (1.00)

?WHICH ONE _(1) of the following containment hydrogen'co'ncentrations Lis the MAXIMUM allowable prior to placing Hydrogen-Recombiners in.-

service while performing EOP-14.0, " Response-To Inadequate Core-Cooling"? .5% .5%

. .5% .5%

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REACTOR OPERATOR Page 30 QUESTION: 048 (1.00)

The following plant conditions exist:

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A Loss of Coolant Accident (LOCA) has occu: re The Hydrogen Recombiners are INOPERABL WHICH ONE (1) of the following is the preferred method to REDUCE hydrogen concentration in the Reactor Building per EOP-14.0,

" Response To Inadequate Core Cooling"? Add Hydrazine to the Reactor Building Spra START the Reactor Building Alternate Purge Syste Vent the hydrogen to the Atmosphere, Purge the Reactor Building with nitrogen ga QUESTION: 049 (1.00)

WHICH ONE (1) of the following conditions PREVENTS arming of ALL the Condenser Steam Dumps? Circulating Water pump breakers NOT closed Loss of Stator Cooling water HIGH condenser vacuum LOW Tavg ,

_ _ _ _ - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ - _ . . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ - _ _ _ _

REACTOR OPERATOR Page 31 QUESTION: 050 (1.00)

WHICH ONE (1) of the following describes Main Turbine operation when the Main Generator output breaker is CLOSED? Steam flow is controlled by the Stop valves to control turbine spee Steam flow is controlled by the stop valves to control generator loa Steam flow is controlled by the Control valves to control turbine spee Steam flow is controlled by the Control valves to control generator load.

QUESTION: 051 (1.00)

WHICH ONE (1) of the following is a load supplied by the Service Water System during normal plant operations?

' Diesel Generator jacket water coolers Reactor Building air dryers

, Residual Heat Removal (RHR) heat exchangers CRDM switchgear coolers

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- REACTOR-OPERATOR Page,32-QUESTION: 052

-

(1.' 0 0 )

The'following plant-conditions exist: -

-

The reactor,is at 80F powe Control rods are in AUTOMATI A Tave-Tref-deviation of 7 degrees F exist Control rods are-NOT movin "RHICH ONE (1) of the following actions should be taken? Adjust Tave to within 5 degrees of Tref by adjusting turbine load and-do-NOT attempt-to move rods until -

troubleshooting is complete, Place the rod control in manual and verify control ro operability by moving rods in-5 steps, then out 5 step Trip the reactor and enter EOP-1.0, " Reactor Trip / Safety Injection-Actuation". Place-rod control in manual-'and adjust rods'to match.Tave to within 1.5 degree F of Tre >

-QUESTION: 053 (1.00)

WHICH ONE (1) of the following conditions would require the IMMEDIATE trip of a Reactor Coolant Pump (RCP) ? } . Delta Pressure across the No -1 seallis 250 psi . Motor bearing temperature is 200 degrees RCP shaft peak to peak composite vibration is 15 mils and stabl .VCT-pressure is 16 psi ._

!

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REACTOR OPERATOR Page 33 QUESTION: 054 (1.00)

WHICH ONE (1) of the following conditions requires EMERGENCY BORATION per AOP-106.1, " Emergency Boration"? Shutdown Margin of 2.00% delta k/k while in MODE 1 One (1) shutdown bank rod remaining out of the core following a reactor tri A 100% Load Rejection from the Main Turbine Generator ppm boron concentration in the RCS while in MODE 6 -

QUESTION: 055 (1.00)

WHICH ONE (1) of the following is the MAXIMUM time the Reactor Coolant Pumps (RCPs) can operate following the LOSS of Component Cooling Water (CCW)? minutes minutes minutes minutes

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. . _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . . _ - . _ _ _ _ _ _ _ _ _ _

REACTOR-OPERATOR- Page 34

~ QUESTION: 056 (1.00)

The following plant conditions exist:

-- The Unit is operating at 100% powe .

_

-

Pressurizer pressure control transmitter PT-444 has-failed HIG WHICH ONE (1) of the following IMMEDIATE actions should be taken per AOP-401.5, " Pressurizer Pressure Control Channel Failure"? Close PORV 444 Close PORV 444B Block valv Close PORV 445 Close PORV 445B Block valv QUESTION: 057 (1.00)

WHICH ONE (1) of the following Critical Safety Functions-has the

HIGHEST order of priority? Containment Heat Sink Integrity Inventory

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. ~ . _ . . . . . . -. . ..- -- - . .. . .- - . . . . . .. - -.

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= REACTOR OPERATOR- . Page 35

,

QUESTION: 058 (1.00)

LWHICH ONE (1) of the following is an example of a Steam Generator depressurizing in an UNCONTROLLED manner? (Assume all plant systems operate as designed.) , A steam break exists on the inlet to the HP Turbine '

causing a rapid decrease in Steam Generator pressur An Atmospheric Steam Dump has failed OPEN causing a rapi decrease in Steam Generator pressur , A steam break exists on-the steam supply line to the-EFW -

turbine (between the Main Steamline and-XVG-2802A)

causing a rapid decrease in Steam Generator pressur A feedwater flow controller failure has'resulted in over feeding the "A"~ Steam Generator causing-a rapid decrease in Steam Generator pressur QUESTION: 059 (1.00)

WHICH ONE (1) of the following satisfies criteria for MANUALLY-actuating Safety Injection per EOP-1.0, " Reactor. Trip / Safety Injection Actuation"? Pressurizer pressure is 1900 psig and DECREASIN Reactor Building pressure is-3.3-psig'and-STABL Steam Generator "B" pressurelis 760 psig and^ Steam.-

Generator'"C" pressure-is 850 psig, Pressurizer level is 4% and DECREASING',

,

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REACTOR OPERATOR Page 36 QUESTION: 060 (1.00)

The following plant conditions exist:

-

Main Condenser vacuum is INCREASIN Main Turbine load is DECREASIN WHICH ONE (1) of the following as the condenser vacuum pressure that will INITIATE a Main Turbine TRIP? inches Hg increasin inches Hg increasin inches Hg increasing, inches Hg increasing.

QUESTION: 061 (1.00)

The following plant conditions exist:

-

Operators are attempting a MANUAL start of Diesel Generator during a Station Blackou The speed signal generator output failed HIGH when the DG reached ten (10) RP The Low Speed Relay (115 RPM) has now LOCKED OUT power 1 from both solenoid operated start valve _

WHICH ONE (1) of the following methods can be used to START D/G A? PRESS and HOLD the EMERGENCY START pushbutto DEPRESS the GEN RELAY RESET pushbutto MANUALLY override the main air start valv DISCONNECT the leads to the shutdown solenoid t

- - - - - - ________- - -_-___--____-_-_________-________________ -__ - - ___ - __-__ _ _ _ _ _

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REACTOR OPERATOR Page 37 QUESTION: 062 (1.00)

The following plant conditions exist:

-

A normal electrical alignment exists when AC power is LOST to Vital Bus 1D >WHICH ONE (1) of the following will provide power to Safeguards Bus APN-5903? V Vital AC Bus APN-1FB

" Inverter XIT-5901 Inverter XIT-5908

- "B" Battery Bus DPN-1HB QUESTION: 063 (1.00)

The following plant conditions exit:

-

The Unit Control _ room has been evacuate Reactor Coolant Pumps-have been TRIPPED and operators are attempting to verify natural circulation coolin RCS pressure is 2000 psi RCS Thot (Th)-is.604 degrees F.'and STABL RCS Tcold (Tc) is-540 degrees'F. and STABL Steam Generator pressure is 1000 psig and-STABL Each of the following parameters is an indication that natural-circulation cooling has been1 established per AOP-600.1, " Control-Room-Evacuation", EXCEPT: _ RCS subcooling RCS Thot (Th) RCS Tcold (Tc)

- Steam Generator pressare

'

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" REACTOR OPERATOR- -Page 38- -

!

- QUESTIOth - 064 - (1. 0 0 ) -

Each-of the following is a Major Action Category ~of EOP-17.0,

-

" Response to_High Reactor Building Pressure" EXCEPT:- Verify containment isolatio Check for excessive hydrogen concentration and determine appropriate actio Reduce heat _ input to containment by shutting down RCPs and closing primary PORV >

d.- Verify containment heat remova QUESTION: 065 (1. 00)

WHICH ONE (1) of the following constitutes a LOSS of containment integrity per Technical Specifications 3/4.6, " Containment Systems"? While at 100% power, an electrician opens the outer Reactor Building airlock door toiperform maintenance activities on the CLOSED INOPERABLE inner Reactor Building door without prior approval, While performing an operability test of two normally-open, redundant Reactor Building isolation valves at 100%--

. power, one of:the valves fails'to clos f While in MODE 6, the equipment door is closed but held in; place by _ only. 'f our _ (4) bolts.

' The Integrated Containment Leakage Rate Test results

~

, indicate a leakage rate of 0.4 of the maximum allowable: ,

l- leakage _(La). ~!

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. REACTOR OPERATO Page 39

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QUESTION: 066 (l'. 0 0)

The following plant conditions exist:

-

Following a LOCA the core exit thermocouples indicate'710 I degrees RCS pressure is 350 psi RVLIS-indicates 30%. ;

NHICH ONE (1) of the following conclusions can be drawn from the above information? Limited core melting has occurre Core uncovery is occurrin Core cooling mode is reflu Core exit thermocouples~have faile QUESTION: 067 (1.00)

The following plant conditions exist:

-

The reactor has been at 100% power for 30 day _

-

Chemistry' reports that RCS activity is.l.5 microCuries per gram DOSE' EQUIVALENT I-13 WHICH ONE (1)' of the following actions will reduce RCS activity? Vent the Volume Control Tank (VCT) to the Waste Ga System, Place cation demineralizertin service"and maximize

= letdow " Maximizciletdown.through-the mixed bed demineralize Reduce letdown to minimum and establish an RCS ph between:

6.5 and 7.5 by chemical injectio ,

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. _ . - _ . . _ - _ . . . _ . . _ . _ . _ _ _ , - _ . . . _ . _ . _ _ , _ _ __. _ _ _ . . - . _ _ _ - , .

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' REACTOR OPERATOR- Pagef40 QUESTION: 068- (1.00)

The_following plant conditions exist:

-

React'or is 30% increasing 3%/ hou .

-

Control rods are in AUTOMATI Bank "D" rods are at 140 oteps and stepping out with NO demand signa WHICH ONE (1) of the following actions are required to be performed FIRST per AOP-403.3, " Continuous Rod Withdrawal"? Place the ROD CNTRL BANK SEL switch to the CBD positio Place the ROD CNTRL BANK SEL switch to the MAN position, TRIP the reactor and enter EOP-1.0,_" Reactor Trip / Safety Injection Actuation". Enter EOP-13.0, " Response To Abnormal Nuclear Power Generation", to ensure the reactor is TRIPPE ,

'

QUESTION: 069 (1.00) r The following plant conditions exist: -

'

-

The Unit is operating at 100%.

-

Two (2) Bank "A" rods have dropped into'the cor WHICH ONE (1) of the following actions.is' required to beLtaken IMMEDIATELY per AOP-403.6, " Dropped Control Rod"?- Place the ROD CNTRL BANK SEL switch.toLthe. CBA position, Place the ROD CNTRL BANK SEL switch-to the MAN positio TRIP the reactor and enter EOP-1.0, "

Reactor-Trip / Safety Injection Actuation", REDUCE Turbine load and stabilize Tavg with Tref.

, .

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= REACTOR OPERATOR- _

Page'41 g QUESTION: 070 (1.00)

WHICH ONE (1) of.the following is-a reason for the time delay between a turbine trip and a generator trip? RCP overspeed can occur during a major LOCA1resulting in a high enough RCS flowrate to damage the reactor internals, A loss of RCS flow due'to a failure of auto bus transfer would not be as serious because the reactor would have

-

been shutdown for the duration of the time dela The thermal stresses imposed on the RCP shafts are far greater if the RCP's are stopped immediately (due to high RCS delta t) after turbine trip, The time delay allows circulating transformer currents

~

induced by the trip transient to stabilize following the plant trip, reducing the probability of damage to the RCP motor QUESTION: 071 (1.00)

-The following plant conditions exist:

-

The Reactor tripped from 100% powe EOP-1.0 " Reactor Trip / Safety Injection Actuation" has-been entere '

-

The-turbina did NOT-trip as expecte WHICH ONE (1) of the following actions should be immediately performed per EOP-1.0, to' trip /stop the-' turbine? Manually close MSIV ,

. ManuallyLclose turbine stop/ intercept valves..

+ Manually trip the generator exciter breake ' Locally trip the Turbine from the front standard.

.

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. . . - . . . . - . . . . --. - - .. . . - - ~ _ . . - . . . . .- . - -, -. .- -

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REACTOR ~OPERATORr Page142

_ QUESTION: 072 (1.00)

- WHICH~ONE~(1) of the following conditions may result from'the -

FAILURE ~of one (1) Combined Intermediate valve to CLOSE on a

'

. turbine trip? Moisture Separator Reheater (MSR) level would increase, causing water induction into the LP turbin * Expansion of steam in the MSR resulting in an overspeed of the turbin Depressurization of the shell side of the MSR by blowdown to the condenser, causing a MSR tube ruptur Unbalanced torque between the "A" and "B" LP turbines resulting in turbine shaft failur ,

QUESTION: 073 (1.00)

The following plant conditions exist:

,

-

The Unit has tripped from 100% power at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months , i- -

Safety Injection Actuation occurred at 0805. hour RCS pressure is 1300 psig and DECREASING slowl ,-

-

Pressurizer level is 100 %.

L -

RCS Thot (Th) is 557 degrees F.

' -

Containment pressure is 1 psig and INCREASING. slowl .WHICH ONE (1) of the following events has occurred?

' Main. Steam line break inside the Reactor Building Inadvertent Safety Injection RCS cold leg break

, Pressurizer vapor space break- '

-

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l REACTOR OPERATOR *

Page 43 QUESTION: 074 (1. 00 )

WHICH ONE (1) of the following conditions requires ALL-Reactor Coolant Pumps-(RCPs) to be TRIPPED? ~1 RCS pressure is 500 psig and SI flow has NOT been-verifie RCS pressure is 1200 psig and SI flow HAS been verifie ,

' RCS pressure is 500 psig and Reactor Building pressure is 8 psi RCS pressure is 1200 psig and Reactor Building pressure is 11 poi ~ QUESTION: 075 (1.00)

The following plant conditions _ exist: ;

-

7te reactor has been shutdown for 120 hour0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months The RCS temperature is 140 degrees The'RCS is at mid-loo A total loss of RHR occur No core cooling is re-establishe WHICH ONE-(1) of the following is the MINIMUM time required for the RCS to reach saturation? minutes minutes minutes minutes

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REACTOR OPERATOR Page 44 QUESTION: 076 (1.00)

The following plant conditions exist:

-

The plant is at mid-loo RHR pump amps, discharge flow and pressure are fluctuatin RCS Hot Leg level is 11.3 inches on RCS Mid-Loop Monitorin WHICH ONE (1) of the following actions should be taken FIRST per AOP-115.5, " Loss of Residual Heat Removal System While at Mid-Loop Operations"? - Trip the operating RHR pump, Place FCV-605A(B), A(B) Bypass (RHR Heat Exchanger Bypass valves), in Manual and throttle close ' Close LCV-459 and 460 (Letdown Line Isolation valves), Attempt to start the alternate RHR pum QUESTION: 077 (1.00)

The following plant conditions exist:

- RHR is in service at reduced inventory condition ~

-

- AOP-115.1, "RHR Pump Vortexing", has been implemente WHICH ONE (1) of the following could cause an observed INCREASE in RCS level during performance of AOP-115.1, "RHR Pump Vortexing"?

a. Vanting the RHR system, b. Any opening in the RCS boundar c. Opening MVG 8809A, RWST to RHR PP d. High RHR flow rate when restarting pum _--_ -_ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ .

_ - - _ _ _ _ _ _ _ _ __

REACTOR OPERATOR Page 45

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l I

QUESTION: 078 (1.00)

The following plant monditions exist:

-

The reactor was operating at 100% power when an ATWS occurre Pressurizer pressure is 2340 psi The reactor trip has NOT been verifie EOP-13.0, Response to Abnormal Nuclear Power Generation was entere Step 4, " INITIATE EMERGENCY BORATION OF RCS" is being performed and boric acid flow has been verifie _

WHICH ONE (1) of the following is the NEXT required action? Start the turbine driven EFW pum Align valves to maintain EFW flow greater than 690 gp Locally trip reactor trip and bypass breaker ;, Open pressurizer PORVs and block valve .)

QUESTION: 079 (1.00)

The following plant conditions exist:

,

- The Unit has just received a Reactor Protection System input which requires a Reactor Tri The Reactor has NOT trippe The MANUAL Reactor Trip switches fail to actuate a Reactor Tri WHICH ONE (1) of the following methods should be INITIALLY used to SHUTDOWN the Reactor per EOP-13.0, " Response To Abnormal Nuclear Power Generation"? Allow the control rods to insert AUTOMATICALLY at 72 steps per minut MANUALLY insert control rods at 48 steps per minut Trip the Motor Generator (MG) set output breakers, Trip the Main Turbin )

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- _ _ _ _ _ _ - - - - _-

..

REACTOR OPERATOR Page 46 QUESTION: 080 (1.00)

The following plant conditions exist:

-

The Unit is in MODE One (1) Source Range Neutron Flux Monitor is out of servic The Source Range Audio Count Rate drawer is out of servic Core alterations are in progres WHICH ONE (1) of the following Technical Specification Action Statements should be implemented? -- Suspend ALL operations involving positive reactivity change Emergency borate the RCS until a boron concentration of 2150 ppm is established, Immediately evacuate the refueling area until the Audio Count Rate is returned to service, Replace the Reactor Vessel Head until both Neutron Flux Monitor and Neutron Flux Alarm are returned to servic _

QUESTION: 081 (1.00) -

WHICH ONE (1) of the following is an IDENTIFYING characteristic of a RUPTURED Steam Generator per EOP-4.0, " Steam Generator Tube Rupture"? Level is DECREASING with MAXIMUM feedwater flo Level is STABLE with Blowdown isolate Steam flow is greater than feedwater flo Feedwater flow is greater than steam flo _ _ -

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.l LREACTOR; OPERATOR' Page 47 i

I QUESTION: 082 (1.00) ;

The following plant conditions exist:

- The-Unit is operating at-80%.

-

A spurious low steam pressure SI signal is receive The cause of the spurious SI signal has been corrected and Safety-Injection Signal has been RESE '

-

During preparation to return the plant to power _

operation, the.MSIVs FAIL to OPEN when:their-control switches are taken to OPE The Unit is currently in MODE 3 with steam pressure downstream-of_the MSIVs at 1080 psi Tavg is 557 degrees WHICH ONE (1) of the following is cause for the Mb1V's failure to OPEN? The MSIV delta pressure is excessiv The MSIV bypass valves are CLOSE The MSIV Motor Control Center breaker is OPE The MSIV Isolation Signal has not been RESE .

'

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REACTOR, OPERATOR' Page'48 ,

i

QUESTION: ~ 083 (1.00) ,

-~ WHICH ONE T(1)~ of the following is the reason why EFW flowrate--is procedurally restricted to less than 100 gpm when-recovering a:

steam _ generator leval if the level has fallen-below 4% wide range Lindication?- Ensure SG pressure transient condition does_not-occur which could result in an uncontrolled; release through a '

Safety valve, Ensure pressurizer level transient does not result in-pressure transient that would actuate S!. Minimize thermal stress conditions on steam generator component Minimize RCS cooldown rate and prevent resultant thermal stress on RCS vesse QUESTION: 084 (1.00) -

WHICH ONE (1).of the following is the reason for tripping ALL Reactor Coolant Pumps (RCPs) when Secondary Heat Sink is LOST? To reduce heat input into the Reactor Coolant System, To reduce' Reactor Coolant System pressur '

_ To_ reduce steam flow from steam generator To reduce decay heat input into the steam generators.

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REACTOR OPERATOR Page 49 QUESTION: 085 (1.00)

WHICH ONE (1) of the following results from the Controlling Pressurizer Level channel failing LOW? Charging Flow Control valve (FCV-122) CLOSE ALL Pressurizer heaters TRI Letdown isolation valve (PVT-8152) CLOSE Actual pressurizer level DECREASE __

QUESTION: 086 (1.00)

The following plant conditions exist:

-

An undervoltage condition exists on 7.2KV ESF bus 1D Diesel Generator "A" has FAILED to star WHICH ONE (1) of the following actions may be taken to mitigate the undervoltage condition? START Diesel Generator "B" to supply emergency power to 1D Position the Nornal to Emergency control switch for ESF bus 1DA to the EMERGENCY position, DE-ENERGIZE the power supply to the ESF Loading Sequencer (ESFLS) and SHUT the alternate feeder breaker to 1DA from XTF-3 Attempt to RE-CLOSE the normal feeder breaker from the 115 KV Parr lin _._ . _ _ _ . . . .. .._ _ _ .. _ _ ___ _ _ _ . - ____ m . .____ _ _ _ _ .

REACTOR OPERATOR Page 50 ,

l QUESTION: 087 (1.00)

'

WHICH ONE (1) of the following conditions-REQUIRES initiation of a manual reactor trip during a loss of instrument air at 50% power? j Instrument air pressure drops to 80 psig ANDLXVA02670-IA, !

TB Instrument Air Header Isolation Valve, is OPE ' Instrument air pressure drops to 70 psig AND IPV08324-IA, Station Air Supply Header Pressure Control Valve, is OPE I Instrument air pressure drops to 60 psig AND Main Feedwater Regulating valves are at MID-POSITIO >

'

,' Instrument air pressure drops to 50 psig AND LCV 459, Letdown Line Isolation valve is at MID-POSITIO QUESTION: 088 (1.00) !

The following plant conditions exist:

- The plant is operating at 100% powe !

- Adjustments are to be made to the packing for the 1611, .

"Feedwater Isolation valves".

- The valves will be BLOCKED OPEN.while work is being performed on each valv WHICH ONE (1) of-the following is the status-of each valve while work is being performed? The valves are OPERABLE, but must be stroke timed following completion of work to verify operabilit The valves are INOPERABLE unless a dedicated operator is standing by to remove the blocks should affeedwater isolation be require The valves are OPERABLE as long as the accumulators ,

.

remain pressurize ; The valves are INOPERABLE until the blocks are REMOVED-and the valve is stroke time .

.

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' REACTOR OPERATOR Page 51

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- QUESTION: 089 (1.00)

WHICH ONE (1) of the following is added to the RCS for oxygen

'I control while in MODE 5? Lithium Hydroxide e Ammonium Hydroxide Hydrazine Hydrogen ;

!

QUESTION:-090 (1.00) ;

f The following plant conditions exist

-

A normally throttled valve has been closed for system i isolation during maintenance, -

Maintenance is now complete and the system-is being '

returned to norma WHICH ONE (1) of the following is the method'for independently ;

~

verifying the position of.the: THROTTLED valve? Observe the relative height of the valve ste . Use local position indication, Observe the number of turns:OPEN from the full CLOSED position as the initial operator returns-the system to servic , Fully OPEN the valve and count'the number of. turns in the CLOSED direction until the desired position is obtaine >

.

i . - .- . - . - . - . . . . . . , , . -. .. . - . - . . - . -

. . . . - - . -

_ _ . - _ _ _ _ _ . . . _ - - . . . . _ _ _ _ . _ _ . . _ . _ _ . . _ _ . _. _ _ _ . . . . _ _ _ _ . . _ . - - .

l REACTOR OPERATOR Page 52 l

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-:

QUESTION: 091 (1.00) l

,

WHICH ONE (1) of the following is_the MINIMUM required tagging !

necessary to assure a motor operated valve (MOV) is isolated for personnel safety per SAP 201, " Danger Tagging"? ,

, A red Danger Tag on the valve breaker and a red Hold Tag !

on the remote operating switch, A red Danger Tag on the valve breaker and a red Danger ,

Tag on the local valve position handwhee ! A red Danger Tag on the local valve position handwheel and a yellow Caution Tag on the remote operating switc A red Danger Tag on the remote operating switch and a

,

yellow Caution tag on the local valve position handwhee ;

,

QUESTION: 092 (1.00)

Each of the following is a component / system that is authorized to -

be OPERATED by maintenance personnel EXCEPT: Instrument air header isolation valves !

, Bridge cranes Demineralized water hose bib valves t Amertap Syste _ _

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REACTOR OPERATOR Page 53 i

. QUESTION: 093 (1.00)

WHICil ONE (1) of the following individuals, by title, may fill-the position of' Fire Brigade Leader per SAP-200, " Conduct Of i Operations"? ! Mechanical Maintenance Fire Brigade member I&C Fire Brigade member j Fire Protection Officer , Turbine Building operator i

!

r QUESTION: 094 (1.00)

- .;

WHICll ONE (1) of the following designates which steps in Emergency Operating Procedures must be performed in SEQUENCE? Steps that are designated with bullets, Steps that are designated with asterisk Steps that are designated by~1etters, Steps that are designated by check off boxe ,

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-

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REACTOR OPERATOR Page 54

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QUESTION: 095 (1.00)

WHICH ONE (1) of the following statements defines the MINIMUM criteria for a !

HIGH RADIATION AREA? ! An area where a person could receive a dose between 10 mrem and 100 mrem in one hour, An area where a person could receive a dose between 100 mrem and 1000 mrem in one hou , An area where a person could receive a dose between 10 Rem and 100 Rem in one hou An area where a person could receive a dose between 100 Rem and 1000 Rem in one hou ,

,

QUESTION: 096 (1.00)

-The following plant conditions exist:

-

An operator has been assigned to operate the Refueling Machine f rom 0000 hours0 days 0 hours 0 weeks 0 months until 1300 hours0.015 days 0.361 hours 0.00215 weeks 4.9465e-4 months on Saturday,

32-05-199 Dose rate for the area is 25 mR/ hou The operator's exposure for-the week is as follows:

Monday, 11-30-1992........... 25 mR

Tuesday, 12-01-1992.......... 30 mR

Wednesday, 12-02-1992........ 50 mR-

Thursday, 12-03-1992........ . 0 mR

Friday,. 12 04-1992............ 25 mR

~

WHICH ONE (1) of the following is.the expected accumulated WEEKLY--

dose for the operator following his present work assignment?

mR b. . 130 mR ; mR mR

,

W y, ,vy--r- ,mm-- ,,p. r,,m- =t,-=- e' *

REACTOR OPERATOR Page 55 QUESTION: 097 (1.00)

WHICH ONE (1) of the following is the MINIMUM review / approval necessary for a temporary procedure change? Discipline Supervisor Shift Supervisor Qualified Reviewer AND Shift Supervisor Discipline Supervisor, Shift Supervisor, AND the Plant Manager QUESTION: 098 (1.00)

WHICH ONE (1) of the following is the approved method for pulling fuses for a Tagout of an electrical component? The Duty Electrician shall pull the fuses and install the Danger Tag An operator shall pull the fuses and install the Danger Tags, The Duty Electrician shall pull the fuses and an operator shall install the Danger Tag ~ An operator shall pull the fuses and the Duty Electrician shall install the Danger Tag : uu= um u $

REACTOR OPERATOR Page 56 QUESTION: 099 (1.00)

WHICH ONE (1) of the following plant conditions defines requirements for a Main Control Board component TAGGED with an ORANGE "R&R Entry" Tag? Assume NO other tags are installed, The component is INOPERABLE and cannot be used for any reason, The component is OPERABLE in any plant MODE, but should

'

only be used in an emergency situatio The component is INOPERABLE, but can be operated if neede The component is OPERABLE ONLY in MODES 5 an 6, but should only be used in an emergency situatio QUESTION: 100 (1.00)

WHICH ONE (1) of the following is required to be reviewed by the oncoming operator at the controls (OATC)? Plant Status and Alarm Log Removal and Restoration (R&R) Log Station Orders and Special Instructions Log Technical Specifications and Control Board Log (**********- END OF EXAMINATION **********)

i i

REACTOR OPERATOR Page 57 '

!

,

ANSWER: 001 (1.00)

A (+1.0) I

!

REFERENCE: ,

. VCS: IC-5, Rod Control, p. 13, Objective 6,7,8 KA 001000K602 (2.8/3.3) 001000A102 ( 3.1/ 3 . 4 )

!

,

001000K602 ..(KA's)

.

ANSWER: 002 (1.00) i 4 i (+1.01

{

REFERENCE: IC-5, " Rod Control", E.O. 6, pages 29-30.- KA 001000G008 (3.6/3.6)

'

001000G008 ..(KA's)

,

ANSWER: 003 (1.00) (+1.0)

.

.-

l l

p._

1 -

- . - . .-...- .- - , , - -.

_. _. ._ ._.._ _ _ _ _ _ _._ - .. _ . . _ _ . _ _ . . _ . . . . _ . _ . _ _ . _ _ _ _ _ _ . _ . . _ - .m

_

REACTOR OPERATOR Page 58 -

.

REFERENCE:

i VCS: AB-4, RCP, p. 19, Objective 6 and 7 KA 003000 GOO'/ (3.2/3.3) -,

L 4 i

i t

t 003000G007 ..(KA's)

!

ANSWER: 004 (1.00) [+1.0)

REFERENCE:- SOP-101, " Reactor Coolant System", pages 2- . KA 003000G005 (3.4/3. 8) , 003000G010 '(3.3/3.6) 4

003000G005 003000G010 ..(KA's)

,

'I

' ANSWER: 005 (1.00) [+1.0)

__-

REFERENCE:

. .

1.- AB-4, " Reactor Coolant Pump", E.O. 6, page-3 . AB-4, " Reactor Coolant Pump",_E.O.,7, page.3 . AB-3, " Chemical and Volume Control. System", E.O. , KA 003000G005 (3.4/3.8)

003000G005 ..(KA's)

!

ANSWER: 006 (l '. 0 0 )

'

b. or (+?. 0)

.

&

_ ,

p

'

,. s ,,----..,v.., . , . < , - ~ ...,,a. ,.s,.-,.-,w,,,,,,,w,-,m-e,,m,- , -.,ew...,n-m---_,m., ,,. am,... , . , . - - gw-, n.7n,,,, , - - , .

_ _ . _ _ . _ _ . . _ . _ _ _ _ . REACTOR OPERATOR

'

Page 59 ;

REFERENCE: i

' VCS: AB-3, CVCS, p. 16, Objective 5 and 7 ;

2.- KA 004010A211 (3.1/3.1)

L i

!

004010A211 ..(KA's) .j ANSWER: 007 (1.00) [+1.0) l REFERENCE:- AB-3, " Chemical And Volume Control System", E.O. 3,-page 3 _

. IB-2, " Component Cooling Water", page-1 . -KA 004000K118 (2.9/3.2)

.

004000K118 ..(KA's)

,

ANSWER: 000 (1.00) (+1.0)

REFERENCE: AB-3, " Chemical and. Volume Control System", E.O.;8, page-3 . KA 004000K202 (3.3/3.5)

004000K202 ..(KA's)

l l

,

h v

..

. - - - . - , - , . _ . . _ _ . _ . . - , _ . . - _ - . _ . . . _ _ . . _ _ _ . _ . _ . . . _ _ . . _ _ - - , . . _ _ _ _ _ _ . _ . . _ . . _ ,

-

REACTOR OPERATOR Page 60 ,

,

!

ANSWER:- 009 (1. 00)

' (+1.0)

REFERENCE: Technical Specifications Pg. 3.3-15a, 16, and 3.0-1 Note: This is an obvious operability problem, and-Ent is !

required in MODES-1-4. With all trains of SI automatic j

t actuation inop, the candidate should be aware without reference to the Tech Spec that 3.0.3 applie . KA 013000G005 (3.6/4.2) .

'

L

!

013000G005 ..(KA's)

ANSWER: 010 (1.00) , (+1.0) .

REFERENCE: VCS: IC-9, p. 40,-Obj. 5 2.- KA 013000K412 (3.7/3.9)

-L i

.

l-L 013000K412 ..(KA's)

L ANSWER: 011 (1.00)

b.= (+1.0)

.. .. . . . - - - . . - . , ,_ - , , - - -

_ _ _ _ . _ _ _ . - . - . _ . . - -__ _ . _ _ . . . _ _ . . _ . . _ _ . _. - . ____ _ _ . __. ..___. _ _ _ -. REACTOR OPERATOR Page 61 i

-REFERENCE: j l

1.- VCS: IC-2, Steam Generator Water Level Control, p. 9, Ob , ->

6, 7, and 8 ! A change in NIS power causes an anticipation of higher reactor -i power by the S/G feed bypass valve control system. This ,

causes the bypass valve to open more and raise S/G leve ' KA 015000K105 (3.9/3.9) '

L i

.

i

.

015000K105 . . (KA's)

ANSWER: 012 (1.00) + ( + 1. 0)

,

REFERENCE:

I IC-8, " Nuclear Instrumentation", , page 4 . KA 015000G007 (3.3/3.4)

.-

015000G007 . . (KA's)

E (1.00)

'

ANSWER: 013 [+1.0)

,

P

.

w , ,. b , w , ww.,,,-, , , ,n-..-..,e-r -,,,w..,.,-_,--=v3.-. rw, --,e---

'

e"t *-'*" '***-Vrer

'

wvv~-- '

-. _. _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _

l

REACTOR OPERATOR Page 62

l REFERENCE: IC-12, " Core Subcooling Monitor", E.O. 4, page .-

-

YA 017020K401 -(3.4/3.7)

,

!

l i

~

017020K401 ..(KA's)

r ANSWER: 014 (1.00) i [+1.0) .

REFERENCE: AB-9, " Introduction To Engin'eered Safety Features System",

E.O. 3, page 1 + .KA 022000K101 (3.5/3.7)

!

022000K101 ..(KA's)

<

ANSWER: 015 (1.00)

' (+1.0)

,

_ _

f i

l

! s W

h

. . . _ . . . . . . . . _ , . . . . _ , , . . _ . , _ . . . , . . . . - , _ . _ - , . . ., - - , - .,

. . ~ . . . - - . . -. . -.-.. . - . . - . - - . . . . . . - . - . -

. . - . - . .. . _ ~ .

REACTOR OPERATOR Page 63

REFERENCE:

- TD-7, " Feedwater System", E.O. 5, pages 14-1 .- KA 056000K103 (2.6/2.6)

,

i 056000K103 . . (KA's)

TANSWER: 016 (1.00)

c .- [+1.0)

REFERENCE: TB 7, "

Feedwater System", E.O. 5, page 5, KA 059000G008 (3.0/3.1)

.

t 059000G008 . . (KA's)

ANSWER: 017 (1.00) . [+1.0)

i i

A

-p P

r b

u , ,e --w ,av ,em-.a r----..w-,,-,n-,, , ,,-- - , , , w, n , , , - . , , , , - - , . , , , , , , , -r,- uw- r- e --,N,,-.w--, e,, y--

. . _ _ _ _ _ _ _ . . . _ _ . . . _ _ _ _ _ . . _ _ . _ _ . _ - _ . . . - ._. _ - - . . _ . . . _ . .. - _ . . _ . . . .

. REACTOR OPERATOR Page 64 '

REFERENCE: B-7, "

Feedwater System", , page 2 ! KA 059000K419 (3.2/3.4)

i

!

059000K419 ..(KA's)

ANSWER: 018 (1.00) [+1.0)

REFERENCE: IB-3, " Emergency Feedwater System", E.O. 3, page . KA 061000K107 (3.6/3.8) 3

^

061000K107 ..(KA's)

. ANSWER: . 019 (1.00) - [+1.0].

,

l

. . . e s'. ~y. , ,

'

.-E, m- . . . ~ --~-_-~,-~.-r .m.' ,,,._,.-wy-v..w, , -- 2.y.-,--,.,,-,Em. . ~ . , - . - - , -

REACTOR OPERATOR Page 65 i

l REFERENCE: IB-3, " Emergency Feedwater System", E.O. 5, pages 27-2 . KA 061000K402 (4.5/4.6)

061000K402 ..(KA's)

-_

ANSWER: 020 (1.00) (+1.0]

REFERENCE: Instructor's Lesson Plan EOP-15, " Response To Loss Of Secondary Heat Sink", E.O. 8, page 1 . KA 061000K301 (4.4/4.6)

061000K301 ..(KA's)

ANSWER: 021 (1.00)

b, or d, [+1.0)

REFERENCE: VCS: AB-12, Wasto Gas System, p. 21 & 22, Objectives 4, 5, 6 VCS: SOP 119, Waste Gas Processing, p. 31 & 32 KA 071000A407 (3.0/3.0) 071000A413 (3.0/3.1)

L 071000A413 071000A407 ..(KA's) _ ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . ___________- _ _ _ _ _ _ ___

_ _ _ _ __

_ _ . , , .

REACTOR OPERATOR Page 66 ANSWER: 022 (1.00) [+1.0]

REFERENCE: GS-9, " Radiation Monitoring System", E.O. 5., page 14, KA 068000A404 (3.8/3.7)

OG8000A404 ..(KA's)

ANSWER: 023 (1.00) [+1.0)

REFERENCE: GS-9, " Radiation Monitoring System", E.O. 8, page 1 . KA 072000K401 (3.3/3.6)

072000K401 ..(KA's)

ANSWER: 024 (1.00) (+1.0)

...

i-t

.'

R'JACTOR OPERATOR Page 67 {!

!

REFERENCE: l

! VCS: Technical Specification 3.1.1.4, p. 3.1-6 l KA 002000G005 (3.6/4.1) l L !

!

-j 002000G005 ..(KA's) i

i ANSWER: 025 (1.00) l [ + 1. 0)

'

t-REFERENCE:

IC-13, " Reactor Vessel Level Indication System", ,

Figure IC1 + KA 002000A403 (4.3/4.4) -

!

,

002000A403 ..(KA's)

+

_ .-

ANSWER: 026 (1.00) (+1.0) .

I r

i

' I

._

!

, - . - . .m a._.,m -. ,,s., __ .,,,._,.,...,.,,m, , , . . , . . . . . . . . . , . . . . ,___,,2, . ,__ _,-. - .. , ,, , ,,,__.m., ,_

RI? ACTOR OPERATOR Page 68 REFERENCE: AB-10, " Emergency Core Cooling System", E.O. 5, pages 12 1 . KA 006030K403 (3.4/3.6)

006030K403 ..(KA's)

ANSWER: 027 (1.00) [ 41. 0)

REFERENCE: AB-10, " Emergency Cooling Water System", E.O. 10, page 21, Technical Specification 3/4.5.1, " Accumulators", page 3/4 5- . KA 006000G005 (3.5/4.2), 006000G011 (3.6/4.2)

006000G011 006000G005 ..(KA's)

ANSWER: 028 (1.00) [+1.0)

. .

_

REACTOR OPERATOR Page 69 REFERENCE: AD-10, " Emergency Core Cooling System", E.O. 3, page 2 . KA 006020A107 (3.5/3.7)

006020A107 ..(KA's)

ANSWER: 029 (1.00) (+1.0)

REFERENCE: VCS: IC-3, Pressurizer Level and Pressure Control, p. 14-16, fig. IC 3.S*, Objective 4 and 5 KA 010000A107 (3.7/3.7)

L 010000A107 ..(KA's)

-

ANSWER: 030 (1.00) !+1.0)

. . . . . .

. _ _ . - _ . _ _ . . . _ . . _ . . . _ . _ . . . _ _ _ -

.

_.-_....___._.m__.._.. _ _ . _ _ _ . . _ . ,

i l

REACTOR OPERATOR Page 70 i REFERENCE: _ IC-9, " Reactor Protection and Safeguards Actuation System",

E.O. 4, page 2 l KA 012000A406 (4.3/4.3)

!

i I

,

012000A406 ..(KA's)

ANSWER: 031 (1.00) I

! [+1.0)

.

REFERENCE: IC-4, " Rod Position Indication", E.O. 7, page . FA 014000A102 (3.2/3.6)

014000A102 ..(YA's)

ANSWER: 032 (1.00) (+1.0) REFERENCE:

l l VCS: . AB-8, Reactor Building' Spray System, p. 13-16,

~ Objectives 5 & 6 KA 026000K101 (4.2/4.2)-

'

- ?

026000K101 ..(KA's)

.

t

..-,m.-<..,,n -,- , , , . . . - , , , , , , , , , , , , - . , + - .-- , , , , , , , . - - - - , . . . , ,

- . . - - - - _ - - _ - . . . _ - . _ . . . . _ - . . ~ . - . . ~ . . . . - . - - . . . - . . . - . - - . . . . . - _

. _

!

.

-REACTOR OPERATOR Page 71 l

!

l ANSWER: 033 (1.00) {

t (41.0) ,

REFERENCE: I

!' AB-8, " Reactor Building Spray System", E.O. 5, page . KA 026000A401 (4.5/4.3) ,

.

t

.

026000A401 ..(KA'n)

. ANSWER: 034 (1.00) (+1.0)

REFERENCE: GS*5, " Spent Fuel Cooling System", E.O.,6, page . KA 033000A202 (2.7/3.0)

-033000A202 ..(KA's)

ANSWER: 035- (1.00) (+1.0]

. . -

!

' REACTOR OPERATOR Page 72 ?

t

' REFERENCE: FACILITY EXAM BANK QUESTION # 774 i GS-5, " Spent Fuel Pit System", E.O. 9, page !' KA 033000A101 (2.7/3.3), 033000A203 (3.1/3.5),

033000K401 (2.9/3.2)

i 033000A101 033000A203 033000K401 ..(KA's)

ANSWER: 036 (1.00) .

, [+1.0)

REFERENCE: TB-1, " Steam Generator and Blowdown", E.O. 5, pages 12 1 . KA 035010K403 (2.6/2.8)

i

,

f 035010K403 ..(KA's)

.

ANSWER: 037 (1.00) (+1.0)

,

REFERENCE:

l '. VCS: IC-9, p. 40-42, Objective _5 ,

L - KA 039000K405 (3. 7,3. 7)

i- L 1

,

039000K405: ..(KA's)

l l

l

- . . - _ - _ - . _ _ . _ - - - - . _ _ _ _ _ _ _ _ - . _ . - - . - _ _ _ _ _ . . . . _ . . _ _ . _ ._ . . , . _ . - . - _ . . . . . . . _ _ . ~ ; _ 2. _ .

,. _- .. -_ - _ -. .

!

REACTOR OPERATOR Page 73 ANSWER: 038 (1.00) : [+1.0)

REFERENCE: !

. TB-2, " Main Steam System", E.O. 7, page 1 ! YA 039000K102 (3.3/3.3)

,

039000K102 ..(KA's)

ANSWER: 039 (1.00) ' [ + 1. 0 )

REFERENCE: GS-1, " Service Power System", E.O. 3, page . KA 062000K201 (3.3/3.4)

062000K201 ..(KA's)

_ _

ANSWER: 040 (1.00) [+1.0)

_ . ._____.-___.._.____--____.m_.____ __._.. . ________._ ..._ . . __._.__ . _ .-_

,

!

REACTOR OPERATOR Page 74 {

t'

REFERENCE: GS 3, "

DC Power", E.O. 8, page . KA 063000A101 (2.5/3.3) l

-!

,

i L

.

,

!

063000A101 ..(KA's) ;

,

t ANSWER: 041 (1.00) (+1.0)

REFERENCE: VCS: IB-5, Diesel Generator Syctom, p. 34, Objective 5 KA 064000K402 (3.9/4.2) ,

L i

..)

064000K402 ..(KA's)

.

ANSWER: 042 (1.00) (+1.0)-

,

t P

i P

-

? -(77gwyy mg g-e u r e.ey es" yi9 M gy,r--&ad' -d--pw evy vvv y -- y

_ _ _ . _ _ _ _ - - _ - _ _ - _ - _ _ _ _ _ _ _ _ __ __- _ _ _ _ _ __ _ _ - _ _ - _ - _ _ _ _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - - _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _

l REACTOR OPERATOR Page 75 REFERENCE: VCS: GS-11 Fire Protection System, Objective 4, p. 17 KA 086000G007 (3.0/3.2)

086000G007 ..(KA's)

I

- ANSWER: 043 (1,00) [+1.0) l

l-REFERENCE: 1 1.- GS-11, " Fire Protection System", E.O. 5, page ; KA 086000A401 (3,3/3.3), 086000G009 (2.9/3.3) !

l 086000G009 086000A401 . . (KA's)

a ANSWER: 044 (1.00) ' (F1.0) ,

.i

.

,

w . v, ., m - + w E w.w...r-e,-%w., -4- ,w# -#, -u c ,w - s a.-,. s.,w,, r-. E e ,w.m.- - %- , . . , . . ~ , + , . -

- -

- . , - =<re.fe wumm .v_er==, *r',_-,w ew -se ,-vr.e-%,4,-p,,-v-w -

- _____ _ _ _ _ _ _ - ____ _____-_____-_____-__ _

REACTOR OPERATOR Page 76 REFERENCE: VCS: SOP 115, p.14 KA 005000K410 (3.1/3.1)

005000K410 ..(KA's)

.

!

ANSWER: 045 (1.00) [+1.0)

REFERENCE: AB-7, " Residual Heat Removal System", E.O. 8, page 2 . KA 005000A202 (3.5/3.7)

,

P 005000A202 ..(KA's)

a ANSWER: 046 (1.00)

l'- (+1.0) ;

I

'

,

l L

p ;

_ . - . _ , _ _ . . . ._ _ _ . _ . . . . _ _ . . . . _ . _ _ _ ._ . ... - _

. _ . _ - _-- ___ _ _ _ ---__ -_.. -__- - -.-------- - - - - -_-

i REACTOR OPERATOR Page 77 4- REFERENCE: AB-2, "

Reactor Coolant System", E.O. 3, page 37, KA 007000G008 (3. 0/2. 8)

007000G008 ..(KA's)

N4SWER: 047 (1.00) [+1.0)

.

'

REFERENCE: EOP-14, " Response To Inadequate Core Cooling", page 6, Instructor's Lesson Plan EOP-14, E.O. 7,8, page 2 . KA Numbers 028000K501 (3.4/3.9), 028000K502 (3.4/3.9).

.

-

,

028000K501 ..(KA's) [

ANSWER: 048 (1.00)

- [+1.0)

REFERENCE: Instructor's Lesson Plan.EOP-14.0, E.O. 7,8, page 24.- KA 028000G004-: (3.3/3.5)

028000G004 ..(KA'a)-

,

k f

k-ye c- ,y--+,3.-wyw,w+-,wm,--,,y--ww,,,--wLyv-- w.ve, w-,,=-e,se-w----- r --w---%-ar -w,, .--4 ..-e -r- - . . . , , '- = --, -e'e-+-+ -a+- w a wi -m-<4-+-

.. . . - . . - - - . . . . . . . . - .. -. -.... - - - . . - . ... - . .

. REACTOR OPERATOR Page 78 ;

,

ANSWER: .049 (1.00) [+1.0]

REFEP.ENCE: IC-1, " Steam Dump-System", E, , Figure 1C1.1 . KA 041020A408 (3.0/3.1)

.

041020A408 ..(KA's)

ANSWER: 050 (1.00)

(+1.0)

REFERENCE: TB-9, " Main Turbine" , , pages 4- . KA 045000A305 (2.6/2.9)

045000A305 ..(KA's)

ANSWER: 051 ( l'. 0 0 ) [+1.0]

REFERELNE:

'1'. IB-1, " Service Water System", E. O. 3,.page 3.

KA 076000K105 (3.8/4.0)

076000K105 ..(KA's)

,

vi--- -,1w., + +-, . . - ,

. . , ,_ - . . _ - . _ - - _ _ _ _ . _ . . _ . , _ - . _ . _ . _ _ . . - . _ . _ . . . _ , - _ . _ . .

.

REACTOR OPERATOR Page 79 i

.

- ANSWER: ' 052 (1;00)

- d . -- (+1.0) -

,

- REFERENCE:-

1.- VCS: AOP-403.4, p. 1 KA 000005G010 (3.4/3.6) >

+

000005G010 ..(KA's)

ANSWER: 053 (1.00) [+1.0) .

REFERENCE: AB-4,~" Reactor Coolant Pump", E.O. 5, page 37, KA.000015A208 (3.4/3.5)

.

'000015A208 ..(KA's) -

ANSWER: 054 (1.00)-

i [+1.0)

J w

W

yf g . -. - - - - - - , -

e

_ _ _ _ _ _ _ _ __ - - _ _ _ _ _ _ _ _ _

REACTOR OPERATOR Page 80 REFERENCE: AOP-106.1, " Emergency Boration", page . Instructor's Lesson Plan AOP-106.1, Objective 2, page . KA 000024K301 (4.1/4.4)

000024K301 ..(KA's)

-_

ANSWER: 055 (1.00) {+1.0]

REFERENCE: AB-4, " Reactor Coolant Pump", E.O. 5, page 3 . KA 000026A206 (2.8/3.1)

000026A206 ..(KA's)

ANSWER: 056 (1.00) [+1.0]

--. . . _ ._ _ . _ _ _ _ . _ . _ . . _ _ , . _ . _ . _ . . _ - - . __ . - , .__

^

~'

REACTOR OPERATOR

-

PageL81-

'

REFERENCE: -

1- -AOP-401.5, " Pressurizer Pressure Control Channel Failure",-

page- .

' 2 . -- Instructor's Lesson-Plan AOP-401.5, E.O.-4, pag .- KA 000027A215 (3.7/4.0)

000027A215 ...(KA's)

ANSWER: 057 (1.00)

- (+1.0)

REFERENCE: Instructor's Lesson Plan EOP-12.0, " Monitoring of Critical Safety Functions", E.O. 6, page 1 . KA 000040G012 (3.8/4.1)

.

000040G012 . .- ( KA' s )

-

ANSWER: 058 (1.00) [+1.0)

.

._-

.

y 4s -,, c- -- _

.,7, -m<- 6 -c' -*

. .. . . - - - . . _ _ .. . . . . . .. .. - _.. -- -. __ __.-_

,

~

.' REACTOR OPERATOR . Page:82

'

' REFERENCE:

. _

Instructor!s Lesson Plan EOP-4.'0, page 2 . KA 000040A201 (4.2/4.7), 000040G007 (3.3/3.6)

,

000040G007 000040A201 . . ( KA' s )-

i

.?

ANSWER: 059 (1.00) [+1.0]

REFERENCE: EOP-1.0, " Reactor Trip / Safety-Injection Actuation", Reference-Pag . KA 000040A204 (4.5/4.7)

o

, 000040A204 ..(KA's)

ANSWER: 060 (1.00)- [+1.0)

,p

,

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v -

y + 4 w- e wa --- war -s----

REACTOR OPERATOR Page 83 REFERENCE: AOP-206.3, ' Loss of Condenser Vacuum", page . KA 000051A202 (3.9/4.1)

000051A202 ..(KA's)

-

ANSWER: 061 (1.00) [+1.0)

REFERENCE: FACILITY EXAM BANK QUESTION 1059 IB-5, " Diesel Generator Syrtem", , page 1 . KA 000055A102 (4.3/4.4)

000055A102 ..(KA's)

_

ANSWER: 062 (1.00) [+1.0)

_ _ _ _ _ - _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

____ ___ _ _____ - _- ___ __ . _ _ - . _ _ . _- _.

REACTOR OPERATOR Page 84 REFERENCE: FACILITY EXAM BANK QUESTION 1065 GS-2, " Safeguards Power", E.O. 3, Figure GS . KA 000057A214 (3.2/3.6)

000057A214 ..(KA's)

ANSWER: 063 (1.00) [+1.0]

REFERENCE: AOP-600.1, " Control Room Evacuation", page . Instructor's Lesson Plan AOP-600.1, E.O. 7, page . KA 000068A211 (4.3/4.4)

000068A211 ..(KA's)

ANSWER: 064 (1.00) (+1.0)

.. .

. . .

- . . . . . . , _. . _ _ _ . . . _ _ _ _ . _ _ _ . .. . - . _ . _ . . ,. . .

-v l REACTOR OPERATOR Page-05 ,

REFERENCE;

'la -VCS: EOP-17.0, Response to High Reactor Building Pressure, ,

Rev 4 KA 000069A202 (3.9/4.4)

,

000069A202 . . (KA's)

.

ANSWER: 065 (1.00) ,

b. or [+1.0)

. REFERENCE: Technical Specification 3.6.1, " Containment Integrity",-page 3/4 6- ~

2.- KA 000069A201 (3 ~. 7/4. 3 )

s 000069A201 . . (KA's)

ANSWER: 066 (1.00)

~ [ +1. 0) :

.

-r - , r- 'e - - , . , . , , , , ,

. _ . - . . .. , ~ . .- -

_ - . = ~ - - . - = - . . . . . _

REACTOR-OPERATOR Page 86-REFERENCE:

' 1. - VCS: _ Emergency _ Operating Procedures-Lesson Plan, EOP-12.0,-

Monitoring Critical Safety Functions, p. 2 = _ KA 000074A207 -(4.4/4.7) 017020K503 -(3. 7/4.1)-

s 000074A207 ..(KA's) ,

ANSWER: 067 (1.00) (+1.0)

- REFERENCE: _VCS: GS-6, Primary Chemistry and Sampling, Objectives 6,-9,-

10, p.22, 29, 30 _ _

-- VCS: CR-2, Plant Chemistry Control, Objectives, 7,_12, p. 17 VCS: Technical Specifications 3. . KA 000076K305 (2.9/3.6)

s 000076K305 ..(KA's)

,

- ANSWER: 068 (1.00)

1 (+1.0)

j -.

'

. , , , - , . , , - , , . - , . , . , - . ~ . . , .,. .....J: . - ... ;-. ,, _ - a

REACTdR OPERATOR ~Page:87

RsFERENCE: Instructor's Lesson Plan AOP-403. 3, " Continuous Rod Withdrawal", E.O. 4, page .- KA 000001G010. (3.9/4.0)

000001G010 ..(KA's)

ANSWER: 069 (1.00)

c. . . [+1. 0)

REFERENCE: Instructor's Lesson Plan AOP-403.6, E.O. 4,-page . KA 000003G0102 (3.9/3.8)

!

l 000003G010 ..(KA's)

ANSWER: 070 (1.00)

b .- (+1.0)

^l-I

. :.!

t f

l'

-l

!

'

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-

'l L

'! . . . _ . .- ... . _ . ._. _ .- ._-._._ _ _ _ _ . . . . . . _ _ . . _ .

. REACTOR OPERATOR Page 88 ii

- REFERENCE: l 1. - VCS: TB-5, Turbine Control and Protection System, p. 52, Objective'1 and 7 -

' '- - - KA 000007K301' (4.0/4. 6) -

-

+

'

-L ,

-

000007K301 ..(KA's) I

' ANSWER: 071 (1.00)

~ (+1.0)

'

REFERENCE: VCS: EOP-1.0, p. 2 KA 000007K301 (4.0/4.6)

L 4

,.

000007K301 ..(KA's)

- ANSWER: -072 (1.00)

b.- - [+1. 0]

I

J

'f p

-

'e y y -- w w- ,w q y.---.-s-- -e -.-.g g r,,y.,

.1

-

REACTOR-OPERATOR _

Page-_89

.

REFERENCE:.

' FACILITY EXAM BANK QUESTION 1028 TB-2, " Main Steam System", , page 2 .:3 . KA'000007K103 (3. 7/4. 0)

.

000007K103 ..(KA's)

ANSWER: 073 (1.00) [+1.0]

' REFERENCE:

- FACILITY EXAM BANK QUESTION 2176 Instructor's Lesson Plan EOP-2.0, , page 2 . KA 000008G011 (4.0/4.1)

000008G011 ..(KT's)

' ANSWER: 074- (1.00)

. [+1.0)

..

<

k a , . . - -,.. . .

- _ . . - ._ __ .. ___ _ _ _ _ _ . _ _ .. . -- _

-REACTOR OPERATOR': Page:90

REFERENCE:' - Instructor's Lesson' Plan-EOP-2.0, page 1 .- KA 000011A103 (4.4/4.4), 000009A215 (3.3/3.4).

000011A103 0000094215 ..(KA's)

-ANSWER: 075 (1.00)

. (+1.0)

REFERENCE: VCS: GOP-9, Mid Loop Operation, Attachment-IV

' Generic Letter 88-17

- VCS: AB-7, Residual Heat. Removal System, Objectives 6,7,8, . Generic Letter 88-17 Note: This question reveals if the candidate is sufficientl sensitive to the issue of loss of RHR, and the-very short-time-frame available to respond to sam Industry events-have occurred where RHR has been lost at reduced inventory, and one

~

key issue is t'. operators were often not aware how little

'

time,was available until saturation was reached in-the cor The question does not' require detailed knowledge of the

-

saturation vs. time curve due to the very large time! frame of the incorrect distractors.... KA 000025K101 (3.9/4.3)

000025K101 ..(KA's)

ANSWER: 076 (1.00) ( +1. 0 ) - .

y - - ww- -

.- -. . . ~ . . . -. . ~ _ - - . . ~ -- - - . . . . - . . ~ . - - . . .- ,,

REACTOR OPERATO Page 91 LREFERENCE: - : VCS:- AOP-115.5, " Loss of Residual Heat Removal System While at.Mid-Loop Operations", .~- .KA 000025G010 (3,9/3.9)

L

.

000025G010 ..(KA's)

ANSWER: 077 (1.00) ( +1. 0 )

REFERENCE:

, AOP-115.1, "RHR Pump Vortexing", p. . KA 000025A102 (2.9/2.8)

.

000025A102 ..(KA's)

ANSWER: 078 (1.00)

. (+1.0)

REFERENCE:

l' . -EOP-13.0, " Response to Abnormal Nuclear Power Generation", . .KA 000029G010 ( 4 . 5 / 4 .-5 )

L

-

l 000029G010 ..(KA's).

!.

l-l-

_ _

. .. . - _ . _ _ _ ._ .--. _ _

- . . _ _ _ __ ..m . _ .. _ .__ -- _ _ _ . - _ _

REACTOR OPERATOR- Page 92-ANSWER: _ 079 _1.00)

( ( +1. 01 REFERENCE: EOP-13, " Response To Abnormal Nuclear Power Generation", NOTE,- -

page . Instructor's Lesson Plan EOP-13, E.O. 6, page 1 . KA-000029A114 (4.2/3.9)

000029A114 ..(KA's)

ANSWER: 080 (1.00) (+1.0]

'

REFERENCE: ' Technical-Specification 3/4.9.2, " Refueling Operations

-

Instrumentation", page 3/4_9-2, KA 000032G008 '(2.8/3.3)

000032G008 ..(KA's)

L

'

ANSWER: 081 (1.00) [+1.0].

.

'

_ -_ . . . , , _ . , _ _ . . , _ . . - - , _ _ ___,._a . -, . . . _ __ _ . . . _ . _ _ _ .. . . . . _ _.__. ___ _.-. _ - . .. .

-REACTOR-OPERATOR- Page:93'

.

!REFERENCEi

= Instructor's Lesson Plan EOP-4.0, , page E 2 . . KA 000037A203 (4. 4 /4. 6) -

.

000037A203 ..(KA's)

ANSWER: 082 (1.00)

' [ + 1. 0 )

REFERENCE:. . FACILITY EXAM BANK QUESTION 90 . KA 000038A212 (3.8/4.2)

.

000038A212 ..(KA's)

.

ANSWER: 083 (1.00) [+1.0]

.t 2 '

'

't>

, , , ~ - , , . . - - , , = . . . , , , .,-c, , :,. . . . , ~ , , , , , - . . ., ,. , .,

REACTOR OPERATOR Page 94 REFERENCE: Instructor's Lesson Plan E-15.4, " Response to Steam Generator Low Level", E.O. 5, page 1 . KA 000054K102 (3.6/4.2)

000054K102 ..(KA's)

-_

ANSWER: 084 (1.00) [+1.0]

REFERENCE: Instructor's Lesson Plan EOP-15, , page 8, KA 000054K3042 (4.4/4.6)

000054K304 ..(KA's)

_

ANSWER: 085 (1.00) [+1.0)

_ _ _ _ _ - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. . . -

- . . . . . . . ~.. - - . - --. - . - . - . - ~ . -. , . - ,

. REACTOR:. OPERATOR Pagec95 ,

- REFERENCE: Instructor's Lesson Plan AOP'-401.6,. , page .

- KA 000028A212 .(3.1/3. 5)

,

?

000028A212 ..(KA's)

,

' ANSWER: 086 (1.00) [+1.0)

REFERENCE:

- FACILITY EXAM BANK QUESTION 1097

, GS-2, " Safeguards Power", , Figure GS .- KA 000056A102 (4. 0/3. 9 )

000056A102 ..(KA's)

- ANSWER:- 087 (1.00) -

- d '. [+1.0]

,

e v

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v,- - - - ~ . - , - - - , . 4

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_.... . .. .

.

I REACTOR OPERATOR Page 96 REFERENCE: AOP-220.1, " Loss Of Instrument Air", page . Instructor's Lesson Plan AOP-220.1, E.O. 6, page . KA 000065A206 (3.6/4.2) 000065G010 (3.2/3.3)

000065A206 000065G010 ..(KA's)

ANSWER: 088 (1.00) [+1.0)

REFERENCE: SI 92-02, " Specific Operations Guidelines", page . KA 194001K101 (3.6/3.7)

194001K101 ..(KA'b)

ANSWER: 089 (1.00) [+1.0)

______ - - . - - . . . .

REACTOR OPERATOR Page 97 REFERENCE: CR-2, " Plant Chemistry Control", page 1 . KA 194001A114 (2.5/2.9)

194001A114 . .(KA's) -

ANSWER: 090 (1.00) [+1.0)

REFERENCE: SAP-153, " Independent Verification", page 6, KA 194001K101 (3.6/3.7)

-

194001K101 . .(KA's)

ANSWER: 091 (1.00) [+1.0)

-_ ---__._-_-_-___.____.L_____-________-________--_____.__--.__-____ -_--__.___-_a

. - , . . . . .- . .- -. . - _ . .. .. . . . . .- - - . .- -. . . , . . ,

! REACTOR OPERATOR' .Page 9 REFERENCE: SAP-201, " Danger Tagging"i page ~ KA 19.4001K102 (3 ; 7/4.1) _ ,

,

.

-

194001K102 ..(KA's)

ANSWER: 092 (1.00) [+1'0)

.

-

REFERENCE: SAP-300, " Conduct of Maintenance", pages 17-18.

, KA 194001A110 (2.9/3.9)-

194001A110 ..(KA's)

. ANSWER: 093- (1.00)

i d =. [ +1. 0].

i f

i ; 11 _ . , . , _ , . - - . . . . , , . ..i... ,

REACTOR _ _ OPERATOR'

-

Page 99-REFERENCE: -

1.- -SAP-200, " Conduct Of_ Operations", Attachment.1,:page ':. - -KA 194001K116 (3.5/4.2)

._

194001K116 ..(KA's)

'-ANSWER: 094 (1.00) (+1.0)

REFERENCE: EP-2, " Usage of Emergency _ Operating Procedures",fpage-1 . KA 194001A102- (4.1/3.9)

194001A102 ..(KA's)

(1.00)

~

ANSWER: 095 b.- (+1.0)

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REACTOR OPERATO _ Page100

. REFERENCE:

l '. - -;CR-6, " Protection Against Radiation", Enabling Objective 14, page'3 . YA-194001K103 (2.8/3.4)

a

194001K103 ..(KA'a) 4

.

(1.00)

'

-ANSWER: 096

. [+1.0)

REFERENCE: CR-6, " Protection Against Radiation", Enabling-Objective 2 . KA 194001K103 (2.8/3.4)

'

194001K103 ..(KA's)

ANSWER: 097 (1.00)

c '. [+1.0]

.REFERENC E: SAP-139, " Procedure Development, Review,' Approval, and'

'-

.Controla, -page 2 . SO 92-01, page _

-s ji '- 3 . KA 194 001A101 -_( 3. 3 /3. 4 )

l-194001A101 ..(KA's)

l:-

-

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w

. REACTOR-OPERATOR- Page101'

+

ANSWER: 098 (1.00)

-Lc . ( + 1'. 0 )

- REFERENCE:

. SI 92-02, page .' 2 . KA 194001K107 (3.6/3.7) ,

,

194001K107 ..(KA's)

ANSWER: 099 (1.00) [+1.0)

. REFERENCE: SI 92-02, page 6 2, KA.194001K102 (3.7/4.1)

.

194001K102- ..(KA's)

- .

'

- ANSWER: 100 (1.00)

.. (+1.0)

.

I-y - E N A ---

. . . ~~. . . - - - .. . . . . . -.

. REACTOR OPERATOR- Page102:

REFERENCE:-

J 1' . . SAP-200, " Conduct.of Operations", Attachment ' KA 194001A103 (2.5/3.4). .

194001A103 ..(KA's)

i, t

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1.

.

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

I'

.; -

.. .- . . . . - . . - - . . - . - . - . - . - , - . . . . . . . . . - - ...- -. .

.

-RBACTOR OPERATOR- Page: 1

,

ANSWER KEY

,

MULTIPLE CHOICE 023 c 001- d 024 d

'

002 d 025 b 003- b 026 a

004 b 027 b 005 d 028- a ,

-006 b or a 029 c -

007 d 030 d

'

'008 d 031 c 009 c 032 b 010 c 033 d

,

_011 .b 034 b

.012 c 035 'b 013 d 036 c-014 b- 037 b 015 d ~ 038 c 016 c 039' a

>

017 c 040 b-018: a- =041 c 019 ~a 042- b

? -. -

_ 020 d 043 c

,

021- b or d 044 c .

022- a 045 .c

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r , . . . . , ~ , - -- ,- ,, . -

.. - . . . , . . . ~ . . ._ . - - . - _ ~ . - - - .- - - . . , .-

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,

LREACTOR OPERATOR- - Page L2 ANSWER KEY

-

.

i 046 d 069 c '

O'47 d 070 b -. !

048 b 071 d 049- a 072 b

'

050 d 073 d 051 a 074 b 052 d 075 a 053 b 076 a 054 c 077 c-055 c 078 d

'056 a 079 a 057 'b 080 a-058 c 081 c-

'059 d 082 .d

-- 0 6 0 ' c 083 c 061: c 084 a 1062 d- 085 b

063 c- 086 -c

. 064 c 08 d ,

f065' b or l 0 8 8 -. d ,

'066~ b 089 c-067 c- 090 b- 091 b--

.

, < ., -- _ .,,,.-~ - . - - - .

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REACTOR-OPERATOR -Page

-

ANSWER KEY

,

%

092 a i093 d ,

.094 c 095 b 096 c 097 c 098 c 099 c 100 b

,

l l

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!

l (********** END.OF EXAMINATION **********)

I

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p

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. . . . . . - . . . - - - - - -

. . . - - . - . - . . - . . - . - . - - .

_ -

'

~ TEST CROSS REFERENCE Page. 1 RO Exam PWR Reaetor f

Organized by -Oueation N u m b e r-

l QUESTION VALUE REFERENCE e

.i

!

001 1.00 8000013 !

002 1.00 8000076 -l 003 1.00 0000017 004 1.00 8000133 005 1.00 8000137 006 1.00 8000025

-

007 1.00 8000070 008 1.00 8000071 009 1.00 8000014 010 1.00 8000029 011 1.00 8000019 i 012 1.00 8000088 013 1.00 8000080 014 1.00 8000087 015 1.00 8000054 016 1.00 8000055 017 1.00 8000085 018 1.00 8000081 019 8000082 '

1.00 020 1.00 8000132 ,

021 1.00 8000001-022 1.00 8000086 023 1.00' 8000131 024- 1.00 8000022 [

025 1.00 8000079 026 1.00 8000073 027- 1.00- 8000074 028 1.00 8000075 029 1.00 8000028 030 1.00 8000078 031 1.00 8000077 032 1.00 8000002 033 1.00 8000072 034 1.00- 8000063 035 1.00 8000064 036 1.00 -8000052 .

037 1.00 8000024- .l 038 1.00 8000053 0391 . 00 1 8000134 040 :1.00 8000061 041 1.00 8000026 042 1.00 8000007 043' .

1.00 -8000060- '

044 .1.00 '8000030-045- 1.00 8000068 1.00

'

046 B000069 047 1.00 '8000128 -

048 1.00 8000129

_. -.

434 _ , . . ,---- - - - -- -, -._- - _. -_--- . . - - - -- - . - - - - -

. - - _ _

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.mm g --- -_-- -- - - - - -

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049 1.00 8000130

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l TEST CROSS REFERENCE Page 2 1

!

RO Exam PWR- Reactor ;

Or9anized by 0uestion number

!

QUESTION VALUE REFERENCE 050 1.00 0000057 ;

051 1.00 8000135 052 1.00 8000009 .

053 1.00 8000109 4 954 1.00 8000089 055 1.00 8000114 056 1.00 8000113 l 057 1.00 8000023-058 1.00 8000106 !

059 1.00 8000122 !

060 1.00 8000112 061 1.00 8000118 i 062 1.00 8000119 063 '

.00 8000115 '

064 1 00 8000012 065 1. 'O 8000121 !

066 1.01 8000096 067 1.0L 8000008 -;

068 1.00 8000092- >

069 1.00 8000093 i-070 1.00 8000016 071 1.00 0000018 072 1,00 0000117 073 1.00 8000125 074- 1.00 8000107 075 ~1.00 8000003 076 1.00 8000004 077 1.00 800009 .00 :8000005 079 1.00 8000104 080 1.00 8000105 081 1.00 8000091 1082 1.00 8000111 y 083 1.00 8000094'

084' 1.00 8000095 085 l' 00 8000102'

086 - 1.00 8000120 L -087 1.00 8000100

088 1.00 8000032 L

'

089: 1.00 8000033 -

090 1.00 8000035 091- 1.00 0000036 092 1.00 8000038 093 l'. 0 0 - 8000040 094 1~. 0 0 8000043 095 1.00 8000044 096 .1.00- 8000045 097 1.00 0000049

-

-. _ . - - _ - ._ __- ..-,,._.,. , . _ .. _ -_ _ _ _ - - _ - _ . . , . _ _ . . _ -

- - - . , - - - , - - - - - - _ _ _ _ _ __

098 1.00 8000050

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TEST CROSS REFERENCE Page 3 RO Exam PWR Reacto i O r g a n-i z e d- by Question Number [

.

QUESTION VALUE REFERENCE i

A

~

099 1.00 8000051 100 1.00 8000136 -

......

100.00

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..*...

100.00

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. . _ ,

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TEST CROSS REFERENCE Page 4 !

.

RO Exam PWR Reaetor

0 r g a n i z e'd by KA Group PLANT WIDE GENERICS -

QUESTION VALUE KA 097 1.00 194001A101 094 1.00 194001A102 100 1.00 194001A103 092 1.00 194001A110 089 1.00 194001A114 088 1.00 194001K101 090 1.00 194001K101 ,

099 1.00 194001K102 ,

091 1.00 194001K102 096- 1.00 194001K103 095 1.00 194001K103 098 1.00 194001K107 093 1.00 194001K116

......

PWG Total

'

13.00 PIJWT SYSTEMS Group I QUESTION VALUE KA 002 1.00 001000G008 001 1.00 001000K602

,

'

004 1.00' 003000G005 005 1.00 003000G005 003 1.00 003000G007 '

007 1.00 004000K118 008- 1.00 004000K202 006 1.00 004010A211 009 1.00 013000G005 "

l- 010 1.00 013000K412 012 1._00 015000G007 011 1.00 015000K105 013 1.00 017020K401 014 1.00- 022000K101-

-015 1.00 056000K103 ,

016 1.00 059000G008 l 017 1.00 059000K41 .00 061000K107 -

020 1.00 061000K301 019 1.00 .061000K402 .

022 1.00' :068000A404:

021- 1.00~ 071000A413 ,

. _ . - . _ , _ _

-

. . . _ . . . - _ _ . _ . . . , . , _ , . _ _ - . . _ . _ . _ _ , _ . _ _ . _ _ - - - _ . _ _ . _ . _ . _ . _ - _ . . . _ .

a,h-.A4 1.-LM

1 023 1.00 - 072000K401 - l

. . ....

-

N

.

'

w=

/s >- r-+'4 -h--Ps%?6 m e ee"-4 d'*r- W- "D'de- M w Y vw -PM'l-N ey9 *+r-*- d- F*1TT 't-W"' s-

. , _ _ _ _ _ _ ...___ _ . _ _ . _ ._..-_ _ _ _ -_ _- ___. _ - _ . _ . _ . _ _ _ _ _ _ _ _ _ . .

TEST CROSS REFERENCE Page 5 RO Exam PWR Reactor [

Organized by KA Group PLANT SYSTEMS !

Group I ;

'

QUESTION VALUE KA l

PS-I Total 23.00 ,

Group II QUESTION VALUE KA

'

025 1.00 002000A403 92 /. 1.00 002000G005 32 7 1.00 006000G011 -

028 1.00 006020A107 026 1.00 006030K403 029 1.00 010000A107 030 1.00 012000A406 031 1.00 014000A102 033 1.00 026000A401 032 1.00- 026000K101 035 1.00 033000A101 034 1.00 033000A202 036 1.00 035010K403 038 1.00 039000K102 037 1.00 039000K405 039 1.00 062000K201-

- 040 1.00 063000A101 041 1.00 064000K402 042 1.00 086000G007 043 1.00 086000G009

- - .....

PS-II Total 20.00 Group III QUESTION- VALUE . KA 045 -1.00 005000A202 044 1.00 005000K410

.

046- 1.00 007000G008 048 1.00 028000G004-047 1.00 028000K501 049 1.00 041020A408 "

.

050 1.00 045000A305

- 051 1.00 076000K105

. .....

w e e-ee w+ *- --e-wsy+ss-w+ .w<,r-=,e ,=-.sde--.- -.w,, -y -eww---+=.- e w e - , --ve -- e- w------w w -3 e e-wm e*-ev-?w-Tw -+:---w-+ ~+wa-y e e w--- w-r--y-=

. _. _. _ _ _ _ _ _ _ _ - _ . _ _ _ _ . _ . _ . . _ _ _ . . . _ _ _ _ - -.. _ .-=. _ __..__ _ .. E a PS-III Total 8.00 1

......

-!

.

I

)

I,

,

'

C i

i N

Y E

P t

t i

i

..

s t

. ' f

' '

- , ..... , m,-, , ..- . . - ,. . . .; .

._ . __ .. . _ , ..... , _ . . .. . - - - ~ . ,, _, . , . . . _ ,

. . - . - . . . - - -

_

i TEST CROSS REFERENCE Page 6 .

l

!

RO Exam PWR Reactor Organ 1 zed by KA Group

PLANT SYSTEMS QUESTION VALUE KA

_ __

PS Total 51.00 t

EMERGENCY PLANT EVOLUTIONS

Group I i QUESTION VALUE KA 052 1.00 000005G010 053 1.00 000015A208 054 1.00 000024K301 055 1.00 000026A206

<

056 1.00 000027A215 059 1.00 000040A204 -,

058 1.00 000040G007 057 1.00 000040G012-060 1.00 000051A202 061 1.00 000055A102 062 1.00 -000057A214 063 1.00 000068A211 065 1.00 000069A201 064 1.00 000069A202 066 1.00 000074A207 067 1.00 000076K305

......

EPE-I Total 16.00 i i Group II QUESTION VALUE KA l

'

068 1.00 000001G010 L-- 069- 1.00-- -

_000003G010 072 1.00 000007K103 07 .00 000007K301 >

070 1.00 000007K301 073 1.00 000008G011 074 11 .10 0 000011A103 .

077 1.00 000025A102-076 1.00 000025G010 075 1.00 000025K101 ,

079- 1.00 000029A114 078 1.00 -000029G010

_ _ - . . . . __ _ _ _ . . . . _ _ . - _ . _ . _ . _ _ _ _ . . _ . . _ . _ _ _ _ . . . _ _ . , , , . . _ _

- , . _ . _ . . . - . - . , - . - ,

__ _

-

,

000 1.00 000032G008 081 1.00 000037A203 082 1.00 000038A212

4

_

mer

_ _ _ _

-

_

_ , , . _ _ _ . - _ _ . _ - - - - - - - - - - - - - - - " - ^ -

, .

.

TEST CROSS REFERENCE Page 7 RO Exam PWR Reactor Or9anized by KA Group EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA 083 1.00 000054K102 084 1.00 000054K304

......

EPE-II Total 17.00 Group III-QUESTION VALUE KA 085 1.00 000028A212 086- 1.00 000056A102 087

'

1.00 000065A206

......

EPE-III Total 3.00

......

EPE Total 36.00

......

.

......

,

......

Test Total 100.00

,,.,J

v t e Inspection Report - Summer - 1992301 (Operator Exams)
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1992 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 17(0) 19(0) 21(0) 22(0) 23(0) Security Operator Exams 301(0) Emergency Planning Other

IR 05000395/1992301

Text

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