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Exam Repts 50-272/92-21OL & 50-311/92-21OL on 921214-17.Exam Results:All Six SRO Candidates Passed All Portions of Exam & Issued Licenses
	ML20128C989
Person / Time
Site:	Salem
Issue date:	01/27/1993
From:	Meyer G, David Silk; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML18096B236	List: IR 05000272/1992021;
References
	50-272-92-21OL, 50-311-92-21OL, NUDOCS 9302100011
	Download: ML20128C989 (151)
	v • d • e

{{#Wiki_filter:I U.S. NUCLEAR REGULATORY COMMISSION

REGION I

, OPERATOR LICENSING EXAMINATION REPORT l ! l

REPORT NO.: 92-21(OL) I t 50-272 j DOCKET NOS.: i 50-311 1

i LICENSEE: Public Service Electric and Gas Company 244 Chestnut Street i Salem, N.J. 08079 FACILITY: Salem Generating Station. Units 1 and 2 LOCATION: Hancocks Bridge, New Jersey ' DATES: Decera'o er 14 - 17, 1992 EXAMINERS: T. Vehec, Pacific Northwest Laboratories (PNL) G. Benjamin, PNL CHIEF EXAMINER: dW < \ l N ! David M. Silk, Sr. Operations Engineer Date PWR Section, Operations Branch - ( APPROVED BY: ,4fg / vJof/ //J. 7/93 Glenn W. 516yer, ' Chief / / , _ Iate I P'WR Section, Operations Bral ch Division of Reactor Safety - '

i I 9302100011 930202 PDR ADOCK 05000272 y PDR

EXECUTIVE SUMMARY During this examination, six instant senior reactor operator (SRO) candidates were evaluated.

Written and operating examinations were administered during the week of December 14, 1992. All six candidates passed the examination and were issued licenses.

The training department staff was helpful and accommodating during the examination development, validation, and administration. The timely submittal of reference materials allowed the examination team to develop the written and operating examination despite time lost due to holidays. The availability of facility staff to review the written exammation prior to administration and the availability of the simulator to validate scenarios and job performance measures (JPMs) was helpful to the examination process.

Analysis of the examination results yielded several noteworthy observations regarding candidate or training program weaknesses. From the review of written examination results, several weak knowledge areas were noted in component cooling water system lineup, reactor protection system alarms, and confirmiag a steam generator tube rupture in a faulted steam generator. The operating portion of the examination revealed weaknesses in locating information contained in control room materials for administrative questions and performing containment ventilation operations.

During the operating examinatica, the simulator generally functioned well. However, there were two modei deficiencies noted. In one scenario, it was evident that there was minimal, if any, decay heat being modeled by the simulator. During a loss of all main and auxiliary t feedwater, core exit thermocouples (CETs) were either remaining constant or decreasing ,

slightly at times. The downward trend of the CETs could have caused an erroneous ! transition out of EOP-FRHS-1, Response to Loss of Secondary Heat Sink, without addressing ! the loss of feedwater condition. The other simulator deficiency was the unreliability of the availability of the P-250 cornputer ud to provide information to the candidates during , scenarios and JPMs.

i i l Overall, the evaluation of the candidates' performance and the observation of facility staff I and training facilities indicates that the training program is being implemented effectively. l

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DETAILS . 1.0 INTRODUCTION The NRC administered initial licensing examinations to six SRO instant candidates. The examinations were administered by the NRC chief examiner, with the assistance of two contractor examiners, in accordance with Revision 6 of NUREG 1021. All candidates passed the examination and were issued licenses. Overall, the candidates performed well in the operating portion of the examinatien, with several weaknesses noted after analyzing the examiratie results. The simulator performed well with only two noted deficiencies.

2.0 SUMMARY OF EXAMINATION RESULTS 2.1 Individual Examination Results _ _ _ SRO RO Written 6/0 N/A Simulator , 6/0 N/A Walk-through 6/0 ' N/A _ T Overall 6/0' N/A _- __.

_ 2,2 Generic Weaknesses A weakness is an operating exmnination item performed unsatisfactorily or with difficulty.

This item may not, by itself, result in an individual failure. A weakness may also be a written examination item when greater than 50% of the candidates incorrectly respond to the - item. A weakness could indicate an area where the training program should be assessed to determine if increased emphasis or instruction is warranted.

Operating Examinations The majority of the candidates demonstrated minor unfamiliarity with procedures examined-in 5,ection A of the operating examination by having difficulty locating the answers in the control room references. Candidates generdly obtained the correct responses but were not quick to do so.

Two candidates did not properly implement O'P-II-16.3.1, Containment Ventilation Operation, to reduce containment pressure. Both candidates skipped over step 5.6.1 and started at step 5.6.2. Step 5.6.1 directs operators to verify the status of containment

- ventilation radiation monitors prior to initiating a containment release.

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Written Examinations j I Ouestion # Area of Knowledge !

29 Knowledge of component cooling water (CCW) system valve alignment to prevent overpressurization of a CCW heat exchanger that has had CCW flow secured to it.

43 Knowledge of the conditions causing the actuation of the Reactor Protection System " GENERAL WARNING ALARM."

76 Knowledge of the method to confirm the diagnosis of a steam generator tube rupture on a steam generator that has had a doubled end steam line break inside of containment.

3.0 RELATED ISSUES During the operating examination, the simulator functioned well with the exception of a couple of problems. In one scenario, it was evident that there was minimal, if any, decay heat being modeled by the simulator because, during a loss of all main and auxiliary feedwater, the core exit thermoccuples (CETs) were remaining constant or even decreasing at times. The downward trend of the ClHs could have caused a transition out of EOP-FRHS-1, Response to Loss of Secondary Heat Sink, at step 14.2 back to EOP-TRIP-1, Reactor Trip or Safety injection, without addressing the loss of feedwater condition. In this scenario, the operators were at step 14.2 of FRHS-1 when the CETs were trending downward. The candidate in the SRO position correctly chose to remain in FRHS-1 because no action had been taken to restore feed flow to the steam generators. The other simulator problem was the unreliability of the P-250 computer during the conduct of scenarios and job performance measures.

4.0 CONCLUSIONS

Based upon observation of candidate performance and revievts of the grading of the written and operating portions of the examination, the NRC concludes that the facility training program satisfactorily trains its licensed operator candidates to operate its nuclear facility.

Several weaknesses were noted in candidate performance, but none were determined serious enough to deny issuance of licenses or to indicate inadequate training by the facility.

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ATTACIIMENT 1 SRO EXAMINATION AND ANSWER KEY U. S. NUCLEAR REGUIATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION 1 CANDIDATE'S NAME: 86 CI FACILITY: Salem 1 & 2 REACTOR TYPE: EMR-WEC4 DATE ADMINISTERED: 12_/12D1 If1STRUCTIONS TO CANDIDATE: Use the answer sheets provided to document your answers. Staple thic cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grede of at least 80%. Examination papers will be picked up four (4) hours after the examination starts.

CANDIDATE'S TEST V &QE SCORE % N~ % TOTALS c/ 5 c d FINAL GRADE All work done on thic examination is my own. I have neither given nor received aid.

Candidate's Signature

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5 ANSWER SHEET Page 2 l Multiple Choice (Circle or X your choice) j ' If you change your answer, writo your selection in the blank.

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002 a b c d 027 a b c d
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010 a b c d 035 a b c d

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A N S'W E R- SHEET Page- 3 Multiple Choice (circle or X your choice) If you change your answer, write your selcotion in.the blank.

051 a b c d 070 a b c d 052 a b c d 077 a b c d 053 a b c d 078 a b a d ___ 054 a b c d 079 a b c d 055 a b c d 080 a b c d ___ a b c d 081- a b c d 056 057 a b c d 082 a b- c .d-058 a b c d 083 a- b c d 059 a b c- d 084 .a b- c. d 060 a b c d 085 a b c d-061 a b c d 086- a-- b c d 062 a b c d 087 a- b c d-063 a b c d; 088 a b c d 064 a b c d 089 a b c d 065 a b c d 090 a b c d 066 a b c d 091- a !b c d 067 a b .c d- ___ 092 a b- c- d-068 a b c d 093- a b- c d 069 a b c d. 094 -a b- c d 070 a .b c d. 095 a' b c d-071 a b c .d- 096 a b c d , 072. a b c d- 097 a- -b c- d 073 a b c d- 098 a b c d 074- a b c d 099 'a -b c d 075- a b c d (********** END OF EXAMINATION **********)

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l Page 4 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS

. During the administration of this examination the following itles apply:

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

2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. This must be done after you complete the examination.

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

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

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

the examination cover sheet and each answer sheet.

6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

7. Before you turn in your examination, consecutively number each answer sheet, , including any-additional pages inserted when writing your answers on the examination question page.

8. Use abbreviations only if they are comronly used in facility literature.

Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it out.

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

10. Show all calculations, methods, or assumptiens used to abtain an answer to any short answer questions.

11. Partial credit may be given except on multiple choice questicls. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

12. Proportional grading will be applied. Any additional wrong information that is provided may court against you. 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 fivo responses, each of your responses will be wortn 0.20 points. 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 answers.

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

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Page 5 14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. 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 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 revoked.

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SENIOR REACTOR OPERATOR Page 6 QUESTION: 001 (1.00) WHICH ONE-(1) of the following is the MAXIMUM permissible circulating water thermal discharge Delta-T allowed by the New Jersey Pollutant Discharge Elimination System (NJPDES) for Salem? a. 30.5 degrees F.

b. 27.5 degrees F.

c. 24.5 degrees F.

d. 21.5 degrecs F.

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Page- 7 SENIOR REACTOR OPERATOR QUESTION: 002 (1.00) Given the following:

An Abnormal Plant condition exists.

" Troubleshooting Abnormal Plant

SC.0F-DD.ZZ-AD46(Q) Conditions" is in use.

WHICH ONE (1) of the following individuals is responsible for authorizing testing and/or troubles'. looting identified by the procedure? , a. Maintenance Supervisor.

b. The System Engineer.

' c. The Shift Fupervisor.

d. The. Operating Engineer.

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Page a SENIOR REACTOR OPERATOR . l QUESTION: 003 (1.00)

WHICH ONE (1) of the following MUST approve an EMERGENCY , Technical Specification Interpretation (TSI) prior to implementation in the control room? a. Station Operations Review Committee (SORC).

, b. Licensing Department.

c. 2 On-Shift licensed Senior Reactor Operators.

' d. Operations Manager lu!D an Operating Engineer.

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Unit 1 is in Moda 6.

- RCP #1 Motor is uncoupled from the pump.

- Loop i is full.

- Maintenance is working on the #1 RCP pump.

WHICH ONE (1) of the following prevents leakage of reactor coolant up the RCP shatt? a. Pump shaft mates with the top of the thermal barrier assembly, b. Seal Leakoff collects any RCS leakage up the shaft and.

directs it back to the VCT.

c. Seal injection is maintained during this condition.

d. Nozzle dam installation prevents RCS water from entering the RCP shaft area.

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Page 10 SEMIOR REACTOR OPERATOR QUESTION: 005 (1.00) <

Given the following:

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- A Site Blackout-has occurred.

- ESP buses are de-energized and no emergency diesel generator is running.

- The STA reports the status of the CSF's as follows: i l Heat Sink - RED Shutdown Margin - GREEN Containment - GREEN Inventory - YELLOW Core Cooling - RED Thermal Shock - GREEN WHICH ONE (1) of the following procedures should be used to mitigate this event following tra7sition from EOP-TRIP-1, REACTOR TRIP OR SAFETY INJECTION? a. EOP-FRCC-1 - ResponJe to Inadequate Core Cooling b. EOP-FRCI-2 - Response to Low RCS Inventory c. EOP-FRHS-1 - Loss of Secondary Heat Sink d. EOP-LOPA-1 - Loss of All AC Power

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Page II SENIOR REACTOR OPERATOR , - QUESTION: 006 (1.00) W11ICil ONE (1) of the following describes how to independently verify the position of a throttled valve during a system lineup in accordance with OD-7? ' a. Operator #1 positions the' valve and also performs the second verirication by measuring valve _ stem height.

b. Operator #2 observes Operator #1 positioning the valve and

does not operate the valve.

c. Operator #1 places the valve in the required position, Operator #2 independently places the valve.in the required position, d. Operator #1 places the valve in the required position _and observes Operator #2 placing the valve in the required position.

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Page 12 SENIOR REACTOR OPERATOR QUESTION: 007 (1.00) WHICH'ONE (1) of the following requires actions to be taken per OD-6," Circuit Breaker Reclosure Policy Following a-Trip"? a. Circuit breaker has opened due to a designed interlock. l b. Circuit breaker automatically opened as a consequence of a Reactor Trip /SI.

c. circuit breaker was inadvertently opened during breaker alignment, d. Circuit breaker trips open on overcurrent.

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l l Paga 13 l SENIOR REACTOR OPERATOR l QUESTION: 008 (1.00) WHICH ONE (1) of the following describes why the handwheel must be in place when performing alignments and verifications on Grinnel-Saunders diaphragm type valves? a. For accurate measurement of throttle valve positioning.

b. The valve diaphragm and stem are free to ride up and down without the handwheel installed, c. Independent verification is not possible without the handwheel in place.

d. Control Room position control or indication accuracy is not within acceptable tolerances.

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Page 14 SENIOR REACTOR OPERATOR QUESTION: 009 (1.00) WHICH ONE (1) of the following individuals, by job title, can request the cancellation of a " Temporary Release"? a. Job Supervisor. - l b. Nuclear Control Operator. j

c. Nuclear Shift Gupervisor. 1

h t d. Senior Nuclear Shift Supervisor.

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SENIOR REACTOR OPERATOR Page 15 QUESTION: 010 (1.00) WHICH ONE (1) of the following describes the process for-positioning and tagging of valves under the cognizance of the Chemistry Department? a. Valves are placed in the required position by operations personnel and the tags are hung by Operations personnel, b. Valves are placed in the required position by Chemistry personnel and the tags are hung by Chemistry personnel.

c. Valves are placed in the required position by Chemistry personnel and the tags are hung by Operations personnel.

d. Valves are placed in the required position by operations i personnel and the tags are hung by Chemistry personnel.

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___ Page 16 SENIOR REACTOR OPERATOR QUESTION: 011 (1.00) Given the following

- /21 Charging pump breaker has been tagged out for maintenance.

- Nuclear Shift supervisor has approved a YELLOW tag request-to all'ow a test of the pump breaker-by the Relay Technician, who is the person named on the YELLOW tag.

- The control Room operator prepared the tags and the Nuclear equipment operator hung the tags.

WHICH ONE (1) of the following named individuals, may grant permission to operate the charging pump breaker during normal operating conditions?: a. Relay Technician b. Nuclear Control Room Operator c. Nuclear Equipment Operator d. Nuclear Shift Supervisor

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Paga 17 SENIOR RE1.0 TOR OPERATOR

, QUESTION: 012 (1.00) k j tieen the following:

1 . A losu of the "C" 125 VDc Bus has occurred. * j WHICH ONE (1) of the following statomonts that describos the affect -

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! that this loss of power has on vital loads powered trom this busi <

t. Toad breakers may bo locally closed with the "close" * i puchbutton at tho breaker cubicle.

i l b. Load breakers may be operated from the control room aftor

opening and closing springs are manually cliarged, i -
c. In tho event of a SI, running SI loadn will strip and 10t'

i sequence back on-duo to a spring charging failure. , l d. The " Spring Charging Failuro" alarm will annunciato for- , i all 4KV load breakers.

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Page la SENIOR REACTOR OPERATOR QtrESTION: 013 (1.00)

Given the following

- The oncoming Nuclear control Operator informs the shift aupervisor that he witnessed a drum of 11ydrazine fall from the bed of an incoming truck into the marsh on the access road.

WilIC11 CNE (2) of the following HUST be notified t0* 'he Nucinar Shift Supervisor to perform IMMEDIATE actions fo.' this loss of control of chemicals outsido the protected area? a. llazardous Materiala Managnr b. Nuclear Site Protection (Fire Department) : c Radiation Protection Department d. Nuclear Regulatory Commission i U

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. Page 19 SENIOR REACTOR OPERATOR , d QUESTION: 014 (1.00) i i Given the following: Unit 1 is in Modo 5.

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Unit 2 is Definaled.

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 - Shift manning _is at minimum for the above conditions.

) WHICH ONE (1) of the following represents the MINIMUM number of l Equipment operators who must to on shift for BOTH units for the above conditions? (Sco attached Tech Spec tablo.) , !

j a. 2 b. 3 i c. 4 ) d. 5 .

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Pago 20

) SCt410R REACTOR OPERATOR
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QUESTION: 015 (1.00)

i Given the following: i I "he reactor has tripped due to a steam line break and partial loss of feedwater has occurred.

- EOP-LOSC-1, " Loas cf Secondary Coolant" is in use.

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  - The STA reports the critical safety function status as

, follows:

Shutdown Margin - GREEN

t Coro Cooling - YELLOW 1 lloat Sink - PURPLE I Thermal Shock - GREEN

, Containment - GREEN '

4 Coolant Inventory - YELLOW . ! Wi!ICH ONE (1) of the following' is the required MINIMUM frequency ,

' for monitoring the Critical Safety Functiono? l a. Monitor continuously.

I b. Monitor every 5_ minutes.

{; c. Monitor every 10 ninutes. l f d. Monitor overy 15 minutos.

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_. _ SENIOR REACTOR OPERATOR Pago 21 I QUESTION: 016 (1.00) WillCII ONE (1) of the following represents the MAXIMUM doses limit that may be authorized in 'sn emergency situation to save a life? a. 25 Rom o. 50 R2m i c. 75 Rom d. 100 Rom

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l Pago 22 SENIOR v2AfE'. OPERATOR i l

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c j wu:2 @:f ,.) of the following individuals is responsible for j yi.ay v.iscal accountability of the security keys on the Unit 1 l'

Primary key ring? . a. Primary Operator h b. Shift Clerk j s

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c. Nuclear control Operator

l d. Nuclear shift Supervisor

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Pago 23 SEllIOR REACTOR OPERATOR QUESTIOlis 018 (1.00) Given the following!

A Bank "C" Shutdown rod has fallen into the coro.

- A dropped rod recovery is in progross per S2.OP-AB. ROD-0002(Q), " Dropped Rod".

- Rod recovery has commenced.

-- An URGENT failure alarm is NOT present.

WilICil ONE (1) of the following explains why.the URGENT failure alarm is NOT onergized? a. There is no master cycler input for Shutdown Bank C.

b. Shutdown rods receive no input from the bank overlap unit.

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, c. Shutdown rods-have no multiplexing thyristors.

d. There is only one group of rods in Shut. lown Bank C.

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SENIOR REACTOR OPERATOR Pago 24 I l QUESTION: 019 (1.00) i WHICl! ONE (1) of the following is used na the reactor power input l

to the Rod Insertion Lituit (RIL) computor? I l a. Auctioneered liigh T-avg b. Auctionoorod High Power Range NI . c. Auctioneered High Delta T d. Auctioneered High Trof- :

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QUESTION: 020 (1.00) ' Given the following

 - Salem Unit 1 ir, operating at 30% steady stato renctor power.

- I&c technician receives permission to perform a calibration on PR N-41.

- The I&C.tochnician pu)is the tunes on N-42, then < reinserts.the fuses for N-42 and pulla the fuses for the correct chonnel, N-41, causing a reactor trip.

WHICH ONE (1) of the followdng describes the reason for the reactar trip? , a. PR neutron flux low setpoint trip b. OP Delta T trip d c. PR neutron flux high setpaint trip d. PR rate trip i

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Pago 26 SENIOR REACTOR OPERATOR ; QUESTIC:It 021 (1.00) Given the following H-41 is out of service.

- A shutdown is being performed IAW IOP-5, "Mininum Load to ! llot Standby".

- Reactor power is at 12%. WHICil OllE (1) of the following actions must be verified by the Nuclear Shift Supervisor to ensure that the low power trips are reactivated when power decreases below 10%? a. Verify that the " Power Mismatch Dypass" nwitch is in the "H-41" position.

b. Verify that the "Comparator channel Defeat" switch is in the "N-41" position. P c. Verify that the appropriato jumpers are installed in SSPS Train "A" and "B". r d. Vorify that the "Uppor" and " Lower Sect. ion Channol Defeat" switches are in tho'"N-43" position.

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Page 27 SEMIOR REACTOR OPERATOR QUEST.ON: 022 (1.00) ' Wi!ICli ONE (1) of the followinr3 is NOT an input to the Subcooling Margin Monitor? a. Core Exit Thermocouples b. Wide Range T-hot c. Containment Pressura d. RCS Pressure e _.__

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SENIOR REACTOR OPERATOR Pago 78 ' QUESTION: 023 (1.00) WHICl! ONE (1) of the following is removed PRIMARILY by the CATION bed domineralizer in the CVCS systom? a. Tritium b. Doron c. Cesiuta d. Sodium Hydroxido

A . E e. . - , x--x. a

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l } J Page 29 ' SENIOR REACTOR OPERATOR

, ' i ,

f QUESTION: 024 (1.00) i 1 WilICll OllE (1) t>f the following explains the requiremont to have all letdown orifice valves OPEN during solid plant operation? I a. Provides additional relief protection.

I i

b. Prevents loss of NPSil to the RilR pumps.

c. Provides maximum letdown purification flow.

l d. Prevents lifting of RHR suction relief.

l

i l .

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

Page 30 SENIOR REACTOR OPERATOR QUESTION: 025 (1.00) WHICH ONE (1) of the following describos the operation of 2SJ-67, SI Hini-flow isolation valve, from panoi RP47 a. The t' Power Lockout Switch" is overridden by a safety , injection signal.

b. The " Power Lockout Switch" locks out."oponing" pcwer to the valvo motor operator.

c. The "Recirc Overrido Switch" in the " NORMAL OPER" position can be used to control valvo position.- d. The "Recirc Overrido-Switch".in the "RECIRC OVERRIDE CLOSE" position directly operatos contacts in the MOV control circuit.

.

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-

I

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

       ,
       . _ . _ _ _ _ _ _ _

Pcga 31 SENIOR REACTOR OPERATOR QUESTION: 026 (1.00) Given the following!

 - Plant cooldown and depret.surization ic in progress.     -

At WilICil ONE (1) of the iollowing RCS ptcasures would the accumulators START to inject if the operator failed to close 11-14SJ54, Accumulator Isolation valves when required? , a. 900 psig.

b. 800 psig. , c. 700 psig. , d. 600 psig.

,

.
, ,  ;- - .2-,. _ . :. - , ; . . :- __ ., _,:...._.  - -

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

Page 32 SENIOR REACTOR OPERATOR - ;

0 1 i I 7d) i QUESTION: 027 (1.00) i

Given the following: l ) 5 - Unit 1 is at 100% power. . ' - PT-948A - Containment Pressure Ch el I is failed i high.

"

 - ALL Bistables for PT-948A havy  een placed in the    -

tripped condition. / WHICH ONE (1) of the following would occur if PT-948D-Containment - Pressure detector Channel IV fai16d high? a. Main Steam Isolation /and Containment Spray Actuation b. Containment Spray Actuation and Phase A - c. Main Steam Ipdlation, Safety' Injection ano Phase A d. Main SteamI / solation, safety Injection, containment Spray Act'uation, and Phase A

          ,
          -I

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

, SENIOR RfACTOR OPERATOR Page 33 I h QUESTION: 028 (1.00) l 7 9) i ' Given the following

Unit 1 is in 11ot Shutdow . d.

, " - RCS Cooldown has comme WilICil ONE (1)-of the followl is the MAXIMUM service water inlet temperature for perfofming an RCS cooldown at the normal rate? a. 80 degrees F / j

    /      l
         ;

b. 82 degrees F. / 'l c. 90 degr'oes F.

d. 92 Megrees F.

/

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/
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         <
       . _ _ . _ _ _ _

_ SENIOR REACTOR OPERATOR Pag 6 34 QUESTION: 029 (1.00) Given the followingt

Unit 1 is in Mode 4.

- CCW Heat Exchangers #12A/B are in service.

- CCW flow has been secured to the #11 CCW heat exchanger.

WHICH ONE (1) of the following actions must be taken to the CCW valves to prevent overpressurization of the #11 CCW heat exchanger? a. Close ONLY the heat exchanger inlet valve.

b. Close ONLY the heat exchanger outlet valve.

c. Open heat exchanger inlet AND outlet valves. i d. Close heat exchanger inlet AND outlet valves.

I l _ .

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  - - _ _ _ _ _ _ _ - _ _ _ _ _________________________j

_ _ _ _ . _ . _ _ _ _ _ . - . - _ _ _ . _ _ _ _ . . - . _ _ - - - _ - _ _ _ _ . - _ _ . - _ _ _ _ _ . - _ - _ . _ _ _ _ ._ _ _---_ ._....- l Pago 35 ! SENIOR REACTOR OPERATOR

            !
            :

QUESTION: 030 (1.00) !

WHICH ONE (1) of the following will cause an AUTOMATIC trip of the #11 Motor Driven Auxiliary reedwater Pump following an AUTO start signal? a. 1A Bus differential b. Phaso B overcurrent .I c. Modo 1 SEC Operations d. Auxiliary Foodwater Storage Tank low 1evel i s F I .I

            :
. . . - _ .
-

_ _ . . , . . _

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Pago 36 SENIOR REACTOR OPERATOR QUESTIONI 031 (1.00) WHICII ONE (1) of the following describes the setpoints and etiwr.id;nce_ required for AMSAC to actuato? U. 2/2 Turbino impulse pressures greater than 50%, 2/4 s/G's narrow rango levels less than 10%. b. 3/2 Turbinc impulso pressures greator than 50%, 2/4 S/G's narrow rango levels 100s than 10%. i cc 2/2 Turbino impulso pressures greatcr than 40%, 3/4 3/G's narrow range invols less than 5%. d. 1/2 Tarbino impelse procouros greator than 40%, 3/4 S/G'c rarrow rengt- levels less than 5%. Ib n _ _ _ . . _ _ _ _ _

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

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

SENIOR REACTOR OPERATOR Page 37 QUESTION: 032 (1.00) ,

Givon the followingt

 - Gas Decay Tank release is in progress por OP-II-13.3.1,
 "Gasoous Wasto Disposal System Normal operations"
 - Gas Decay Tank radiation monitors 2R42A and B are in alarm.

WilICll ONE (1) of the following is the reason for de-onorgizing the Auxiliary Building clovator as a consequence of the radiation alarms? a. To provent an unmonitored release through the olevator shaft.

b. To provent spreading of gaseous activity to other areas of the Auxiliary Huilding.

c. To provent the ignition of combustible gas, d. To provent inadvertent personnel access to the area.

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_ ___ _ _ _ _ __ - _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . l SENIOR REACTOR OPERATOR Pago 38 l

l QUESTION: 033 (1.00) WHICH ONE (1) of the following is the reason for the release rato limits specified in Technical Specification 3.11.2.1, "Gasoous

Effluents Doso Rate"? l a. Ensures that the doso at any time insido the oito boundary from radioactive effluents in within the dose limits of 10 CFR 20.

b. Provents exceeding 10 CFR 20 limits at-the site boundary during an accident condition.

c. Ensures that no member of the' general public at or beyond the Sito Boundary will receive a Bota/ Gamma skin dose of 1500 mrom above background.

d. Prevents exceeding 1500 mrom/ year thyroid doso rate above background to a child via the ingostion pathway.

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Page 39 SEllIOR REACTOR OPERATOR

QUESTION: 034 (1.00) ' 1 ! Given the following:

l

  -  Liquid radwaste release is in progress.

4 - NO high rad alarm is present, j

  -  WL-51, Liquid RadWaste isolation valve closes.

WilICl! OllE (1) of the following could have caused the closure of ' WL-51, Liquid Radwaste Discharge Valve?

a. Loss of control air to valvo. I i l b. Loss of 115VAC supply to valve. 1 I c. Control switch on Waste Disposal Panel CC-1 taken to

I

"CLOSE".

d. Local control switch at Alternhte S/D panel taken to

   "CLO3E".

, i

W

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

_ . _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ . _ . _ _ _ - . _ _ . SENIOR REACTOR OPERATOR pago 40 i l QUESTION: 035 (1.00) Given the following , ,

  - Tho unit is operating at 35% power.

- Reactor Coolant Pumn 12 trips.

' WilICl! O!!E (1) of the following describos the IMMEDIATE unit response to the RCP trip? Assume NO operator action.

a. Steam generator water level will shrink but a reactor trip will NOT occur.

b. Steam Generator water level will swell but a reactor trip , Will llOT occur.

c. Steam generator water icvol will shrink and a reactor- l trip WILL occur.

, d. Steam generator water lovel Will swoll and a reactor trip 1 WILL occur.

_ _ . . ;_ . ..,_ _ _ . . _ _ . . . . ., . _ _ _ . , _ . , _..;,.a...._.-.. ,._2-, ..- . . . _ . . . , _ - _ - - .

i Pago 41 l SENIOR REACTOR OPERATOR QUESTION: 036 (1.00) WHICH ONE (1) of the following Unit 1 Area Radiation Monitors has automatic actuations associated with a HIGH level alarm annunciating in the control room? a. 1R2 - Containment Radiation Monitor.

b. 1R7 - In-Core Seal Table Radiation Monitor.

c. 1R32A - Fuel llandling Crane Raditation Monitor, d. 1R5 - Fuel llandling Building Radiation Monitor.

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      !
. _ . . . - . - . - . . .. . . - _ - . . . . . _ . _ .- -
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SENIOR REACTOR OPERATOR Page 42 QUESTION 037 (1.00) ! Wi!ICl! ONE (1) of the following is the basis for maint.aining a minimum of 200 psid on the RCP seals during RCP startup and operation? a. Ensures that adequate seal coolincj flow from the RCS is available.

b. Prevents the #1 RCP seal from swapping from 4r face rubbing to a film riding seal.

c. Prevents the weight of the seal. ring from limiting cooling flow through the seal gap.

d. Ensures adequate cooling flow to the pump radial ' bearing.

, _ _

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_ _ _ _ . . - _ _ .. _ _ _ ___ _ __ _ .__. . _._ . _ . _ . . _ _ _ _ _ _ _ _ _ . . . . . _ . i SENIOR REACTOR OPERATOR Pago 43 !

b

} QUESTI0H 038 (1.00) , ,

,            ;

j Wi!ICil ONE (1) of the following describes how the proximity of the .

Rod Position Indication Search coil to the control rod shaft i t extension relates to the actual rod position?

i a. Indicates that the rod is fully inserted.

l l b. Indicates that the rod is fu)1y withdrawn.

. , '

c. Indicatos rod position within +/-lo steps of actual ;

j t position.

1 I i

d. Indicates rod position within +/-50 stops of actual -l
position. l'

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k Page 44 SENIOR REACTOR OPERATOR l QUESTION: 039 (1.00) . Given the following

           !
  - The Unit is in Hodo 6 with the reactor head installed.

- OP-II-1.3.4, " Filling and Venting the RCS" is in uso.

- A temporary tygon hose has boon connected to Loop 23 ; for use as RCS level indicaticn.

- The RCS is being drained to 97'6".

WHICH ONE (1) of the following explains why there is a concern about nitrogon pressure in the RCS7 > a. Causos gas binding of the RHR pumps.

b. causos a proportional increase in reactor vossol lovel indication on the tygon hoso.

c. causes invalid pressurizer lovel indication.

d. Causes diffictity establishing natural circulation flow if RHR flow woro lost.

,

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t

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

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

l ,

SENIOR REACTOR OPERATOR Page 45 QOESTION: 040 (1.00) i I Given the following:

- The blowdown phase of a large break LOCA is in progress.

- The pressurizer has gone einpty after- the reactor tripped.

- All RCPs are tripped.

' - The NCO reports that all the RVLIS channels-are displaying that there are no voids forming in the reactor vessel.

I WHICH ONE (1) of the following statements. describe the caase of the RVLIS indication?- i ' a. RVLIS indication can be erratic during the blowdown phase due to the rapid insurge of subcooled ECCS flow.

< b. RVLIS is not accurate during the blowdown phase of the LOCA.

' due to high flows through the reactor core.. , c. RVLIS indication was providing the correct. level indication i- during the blowdown transient.

$ d. RVLIS indications were normal for this event due to the , dynamic D/P cells being density compensated.

. ! ! t

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SENIOR REACTOR OPERATOR Page 46 o

-,

QUESTION: 041 (1.00) R WHICH ONE (1) of the following would result froia a break on the common REFERENCE LEG for Channel I pressurizer pressure instrument PT-455 and Channel I pressurizer level instrument LT-4597 a. PT-455 will-tail HIGH and LT-459 will fail HIGH.

b. PT-455 will fa31 LOW and LT-459 will fail HIGH.

c. PT-455 will fail HIGH and LT-459 will-fail-LOW. , d. PT-455 will feil LOW and LT-459 will fail LOW. ,, r @ l

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      .
       !

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SENIOR REACTOR OPERATOR Page 47

QUESTION: 042 (1.00)

, Given the following:

- Pressurizer Overpressure Protection System (POPS)

testing for 2PR1, PORV is ready _to be performed IAW OP-II-2.3.4, " Pressurizer Overpressure Protection Oper6bility Check and Arming of 2PR1 and-2PR2".- . UHICH ONE (1) of the following ntatements reflects a i requirement / interlock that must be met before 2PR1 can be successfully tested? a. Pressurizer pressure must be less than 375 psig, b. PT-403 must be OPERABLE.- c. 2PR6,-PORV block valve,Lmust be closed,

d. 2PR2, PORV, must be isolated.

.

4

--

,

     !
     .]
     .

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_ _ ... . _ . SENIOR REACTOR OPERATOR Pago 48 . QUESTION: 043 (1.00)

' WHICH ONE (1) of the following will NOT cause'the actuation of the Reactor Protection System " GENERAL WARNING ALARM"?. a. Loss of either 15 VDC power supply in RPS logic cabinet.

b. Reactor trip bypass breaker racked in.

. c. Loose or removed circuit card in RPS logic cabinet.

d. Loss of 115 VAC to or in SSPS Output cabinet.

,

9 l o

_ .

  . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

SENIOR REACTOR OPERATOR Page:49 QUESTION: 044 -(1.00) WHICH ONE (1) of the following is a " DEMI'.ND" signal _ that a Reactor Trip is required? I a. IRPI indicates all rods on bottom.

b. OHA F-40,." REACTOR TRIP" energized.

c. OHA F-36, " TURBINE TRIP AND P-9" energized and associated bistable coincidence on RP-4.

' d.' Reactor trip and associated bypass _ breakers OPEN. . p t.

.

#.

Y

' - - - - - _ _ _ _ _  _ _ _ _ _ _ _ _ _ _ _      __ _

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

d SENIOR REACTOR OPERATOR Paga 50 QUESTIOM: 045 (1.00) WHICH-ONE (1) of the following will cause an AUTOMATIC trip of a Containment. Iodine Removal System Fan? a. Phase A b. Safety Injection signal c. Roughing Filter HIGH differential pressure d. Charcoal Filter HIGH temperature

-

     \
     '

"

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    - -
,  _ ,
    .
     .

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

SENIOR REACTOR OPERATOR Page 51

      //dh'7;N
      !

QUESTION: 046 (1.00) W91C3 ONE (1) of the following could pro uce enough-hydrogen.

to exceed flammability limits inside ,the containment following a large break LOCA? (Assume all ECC Bystems function as designed.)

a. Radiolysis of water, b. Zirconium-Water r. ction. _ c. Corrosion of aluminum structural components.

d. Corrosion of/structural components painted with-zinc based pain't.

/

   /
  /
 /
 ,/
,/     \
-      - _ -

- -. - .- -
    , ,-

SENIOR REACTOR OPERATOR Page 52-i l

QUESTION: 047 (1.00) Given the following:

A_large break LOCA has occurred,>

- Prior to the LOCA, containment tempercture was 70 do3rees F.

- Following the'LOCA, containment pressure is~15.7 psia.- - .

-- The EOP's require that the Hydrogen'recombiner be placed in service.

WHICH ONE (1) . of the following values should be ' set' on the Hydrogen Recombiner potentiometer based-on the above conditions?

(Use II-15.3.1, " Hydrogen Recombiners- Normal Operation" Attached)

a; 46.2 Kw.

b. 48.4 KW.

c. 52.8 Kw.

d. 57.2 Kw.

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__ _ _ . . GENIOR REACTOR OPERATOR Page 53

    * lf -

QUESTION: 048 (1.00) 7 q)_

WHICH ONE (1) of the following is the basis - maintaining _ greater than 16% narrow range level in.theA cam generator with a tube leak? a. Reduces the release by "fil

/l/

ing the release" through a layer of water prior to it escaping to the atmosphere.

b. Reduces the_probabilit, of a greater tube rupture by avoiding thermal stresses associated with tube uncovery.

, ,/ c. Maintains the the[ mal stratification layer in the steam generator'to ass'ist in RCS/SG pressure equalization.

' d. Assures that'an adequate heat sink exists _for the reactor core by mafntaining a ninimum mass in the steam generatoys.

l l

   . - .

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

& SENIOR REACTOR OPERATOR Page 54 . i QUESTION: 049 (1.00) Given the following:

- Reactor power is at 25%.
- All plant controls are in automatic.

,

-
- Loop 21 MSIV insTvertently shuts.

WHICH ONE (1) of the following parameters would show an INITIAL INCREASE following this event? (Assume No operator actions are taken).

a. Steam generator #21 level.

j b. Steam Generator #22 pressure, c. Loop #24 Cold leg temperature, d. Steam Generator #23 level, i i i <

4 . d f n n e v , .an+

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n SENIOR REACTOR OPERATOR Page 55 i QUESTION: 050 (1.00) . Given the following:

- A large break LOCA has occurred on Unit 2.

- Radiation levels are increasing rapidly inside the Unit'2 Containment.

At WHICH ONE (1) of the following radiation. levels will 2R44A

insert adverse containment correction levels into the Subcooling-Monitor?

a. 1.0 E+2 R/HR b. 1.0 E+3 R/HR-c. 1.0 E+4'R/HR d. 1.0 E+5 R/HR i

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.
  .-.

SENIOP. REACTOR OPERATOR Page S6 , QUESTION: 051 (1.00) . Given the following: ,

- Loss of offsite power has occurred.

- Diesel Generator #1A has had an actuation of its fire protection system due to high operating temperatures.

- Alarms in the control room actuate indicating that there is a fire in the #1B Diesel Generator room.

- The "Firnt in with Lockout" signal is preventing CO2 deluge actuation in the #1B Diesel Room.

WHICH ONE (1) of the following describes what must occur to-actuate tha #3B Diesel Generator CO2 deluge system? a. Wait for the timed discharge sequence circuitry on the Fire Protection system to cycle and pick up the new signal on #1B Diesel room, b. Manually isolate the deluge system.for the #1A Diesel.

c. Energize the shunt trip coil for the #1B Diesel actuation system at the fire protection panel using the toggle switch provided, d. Bypass the electrical actuation circuitry by manually operating the pilot cabinet operating lever for the #1B Diesel actuation system.

. , _ _ . . . . ._ .

__ _ h SENIOR REACTOR OPERATOR Pago 57 QUESTION: 052 (1.00) Given the following:

 - Unit 2 is at 20% power during a startup.

- The Main Generator has just been synched to the grid -

 - The steam dumps are closed.

- Alarms are received in the control room indicating-that the 21A, 22B and 23A Circulating Water Pumps,have tripped.

WHICH ONE (1) of the following could cause the simultaneous trip of'the three circulating water pumps? a. Opening the 500KV Keeney Line (5015).

b. Opening the 500KV Hope Creek Tie Line-(5037).

c. Opening the 13KV supply from Hope Creek switchyard.

d. A Phase to Ground fault on the Salem 2E 4KVLbus.

, , , . .

SENIOR REACTOR OPERATOR Page 58 j QUESTION: 053 (1.00) Given the following:

- Transfer from Station Power Transformer (SPT), to the Auxiliary Power Transformer (APT), is in progress on Unit 1.

- Actions are being taken to transfer the first Group Bus.

~ WHICH ONE (1) of the following actions must be taken to ensure that the SPT and APT breakers do not trip and lock out when the first APT Infeed breaker to the Group Bus "CLOSE" pushbutton is depressed? , a. Verify. running and incoming voltages matched.

b. Ensure synchroscope is moving alowly in the fast direction.

c. Ensure that the Unit Isolation Trip Multi-Trip relay is ' reset.

d. Ensure that the SPT Infeed breaker Anti-Pump relay is reset.

t

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SENIOR REACTOR OPERATOR Page 59 (1.00) ' QUESTION: 054 ' Given the following:

- Diesel Generator 2A is in the process of being started
  ~

on Unit 2 to parallel it to the grid.

WHICil ONE (1) of the following describes the operation of the diesel generator voltage control switch during this evolution? ' a. Lowering the voltage control' switch manually has no

effect on the generator if. selected to the AUTO (Isochronous) mode of operation.

'

b. Raising the voltage control switch to a higher value, will cause the. generator to pick up a larger share of the reactive load after-breaker closure.

' c. Raising the voltage-control switch will correct a' synchroscope which is traveling slowly in the SLOW.

direction.

- ' d. Lowering the voltage control switch raises the speed of the generator when operating in the MANUAL node. prior to paralleling with offsite-source.

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,

SENIOR REACTOR OPERATOR Page 60 QUESTION: 055 (1.00)

WHICH ONE (1) of the following is an interlock that must be met > in order to close_a diesel generator output breaker from the

Control Room? a. Both Vital Bus infeed breakers must be OPEN.- b. Loading switch in MANUAL (Droop).

c. Voltage Regulator-in AUTO.

d. Diesel speed at 800 rpm.

.

   + ----e . , , - , -- , , ,

- .

SENIOR REACTOR OPERATOR Page 61 QUESTION: 056 (1.00) WHICH ONE (1) of the following is normally used to fill the - refueling cavity from the RWST during preparations for refueling? a. Gravity feed b.-Spont Fuel Pit Cooling pump c. RHR pump . d. Boric Acid transfer pumps

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   .

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SENIOR REACTOR OPERATOR Page 62 QUESTION: 057 (1.00)

Given the following:

- Main Turbine overspeed testing is in progress on Unit 2.

IAW S2-OP-ST.TRB-0003(Q), " Turbine Mechanical Overspeed Test".

- Generator output breakers are open.

- You are the SRO at the Turbine Front Standard.

WHICH ONE (1) of the following procedural / automatic trips provides overspeed. protection for this evolution? a. Mechanical overspeed trip at 1850 rpm.

b. Manual trip at control room frequency counter indication of 1870 rpm.

' c. Manual trip at turbine front standard-speed indication of 1980 rpm.

d. Electrical overspeed trip at 2025 rpm.

.

 -  .  . . ..

Pago 63 SENIOR REACTOR OPERATOR

j l QUESTION: 058 (1.00) l Given the followingt

Plant is in Mode 5.

RHR pump 21 is in service.

,

'

RCP 21 is in service.

- You are directed to start RHR pump 22.

' WHICH ONE (1) of the following describes why 2CV18, Letdown Heat Exchanger Outlet valve,-must be immediately adjusted'following the start of RHR pump 227 a. Prevent RHR pump cavitation.

b. Provent cooldown of the RCS.

, c. Provent lifting the lotdown relief, d. Provent RCS pressure from dropping below the minimum _ required for-RCP operation.

'

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        .. ..

I

' SENIOR REACTOR OPERATOR       Page 64 QUESTION: 059  (1.00)
         ,

Given the following:

 - Unit 1-has been shutdown for 3 days following a 6 month full power run.

- The RCS temperature is 120 degroes F.

. - The RCS is at midloop.

- A total loss of IUIR occurs.

- No core cooling is re-established.

WilICH ONE (1) of the following is the MINIMUM time required for the,RCS to reach saturation? a. 15 minutes-b. 45 minutes c. 120 minutes d. 180 minutes

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

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

SENIOR REACTOR OPERATOR Page 65 QUESTION: _060 (1.00) Given the following: - -- Reactor power-is at 85%

- A dropped rod recovery is in progress IAW S2.OP-

- AD. ROD-0002(Q), " Dropped Rod".

W!!ICH ONE-(1) of the following describes the reason for ensuring that T-ave is within 1.5 degrees F of program prior to operating the Rod Bank Selector Switch through the AUTO position? : a. Prevents misalignment of the Bank Overlap Unit.

~ b. Prevents misalignment of the Pulse to Analog converter, c. Prevents dropping rods with disconnected lift. coils, d._ Prevents misalignment of rod insertion limit computer.

SENIOR REACTOR OPERATOR Pago 66 QUESTION: 061 (1.00) . Given the following:

- All systems are in automatic.

- Reactor power is at.75% and increasing slowly.

- Pressurizer pressure is steady.

- Pressurizer level is increasing.

- T-ave is increasing.

- Containment parameters are normal.

- No operator actions have been taken.

WHICH ONE (1) of the following could cause the abov( symptoms to occur? a. Steam leak outside containment.

b. PORV leaking to PRT.

c. Loop T-ave drifting high.

d. Continuous rod withdrawal.

L

   .- - .

SENIOR REACTOR OPERATOR Page 67 QUESTION: 062 (1.00) ' Given the following:

- The reactor has tripped.

- 2 control rods have failed to insert on the trip.

WHICH ONE (1) of the following is the MINIMUM-volume that must be added to the RCS from the RWST as a result of the stuck rods? a. 1960 gallons b. 3920 gallons c. 5600 gallons d. 11200 gallons

em. + e a- , , .~r,

   .  -

SENIOR REACTOR OPERATOR-Page 68 i QUESTION: 063 (1.00) Given the following:

- A large break LOCA'has occurred.

-

- CET's are 1300 degrees F.

- Operators are performing actions of 2-EOP-FRCC-1,

 " Response-to Inadequate Core Cooling".

WHICH ONE (1) of the following is the reason for maintaining intact S/G-levels above the top of the U-tubes during' performance of this procedure? a. ToLensure_S/G inventory is adequate to avoid' thermal shocking the tubesheet.- b. To prevent depressurization of the S/G steam space.

c. To ensure-maximum heat transfer capability.

d. To maintain reflux boiling capability in the event ~ that SI cannot be restored.

'

t ,- r - 4 e v * y , 4 .- -

SENIOR REACTOR OPERATOR Page 69 QUESTION: 064- (1.00) WHICH ONE (1) of the following explains why it is preferable to leave the RCP's running during a small break LOCA if the RCS pressure RCP trip criteria;on the conditional action summary are met but there is no SI flow? a. To maintain two phase mixture level above the break-longer.

b. To provide heat removal through the. break and the S/G's.- c. To limit single phase inventory loss out of the break, d. To decrease loop transit time for' boron delivery to the core.

  '+ ~ ,

J l SENIOR REACTOR OPERATOR Page 70 (1.00) '

-QUESTION: 065 Given the following:
- Small break LOCA mitigation is in progress IAW 2-EOP-LOCA-1, " Loss of Reactor Ccolant".

- RHR pumps have been secured IAW Step 13 of that procedure.

- Containment pressure is 3 psig.

WHICH ONE.(1) of the following would require the manual start of the RHR pumps?

< a. RCS subcooling margin is 10 degrees F b. RCS pressure is 200 psig c. Charging pump at runout-flow conditions d. SI pump at runout flow conditions

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. - - .  . _ .

SENIOR REACTOR OPERATOR Page 71

4 , QUESTION: 066 (1.00) WHICH ONE (1) of the following indicates that the PRT rupture ) disk is ruptured following a pressurizer PORV failing OPEN? a. Pl.f temperature decreasing ' b. Relief line temperatures increasing c. PRT level low d. Pressurizer level decreasing

.. .. -. . . . - . - . .

SENIOR REACTOR OPERATOR Page 72 i

QUESTION: 067 (1.00) WHICH ONE (1) of the following would require that maintenance install jumpers on 2A East Valves and Misc. 230V Control Center terminals for 21SJ49, RHR discharge to cold leg when < S2.OP-AB.CR-0001(Q), " Control Room Evacuation," is in progress 7 , a. Automatic initiation of Safety Injection b. Manual Safety Injection alignment c. Boration to cold shutdown conditions d 2RP4 lockout switches not in valve operate position prior to evacuation.

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

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           !

Sell!!)R REACTOR OPFRATok Pago 73 + i ' QUESTION: 068 (1.00) : Wi!IcIl OllE (1) of the following is the reason that the RCP'c are ' tripped as a cubeoquent action of 82.OP-AB.cR-0002(Q), " control Room Evacuation due to Fire In the Control Room"? , a. Limits heat input by RCP's to the RCS during cooldown. i b. Prevent damage to the RCP's if a LOCA were to occur later in the ovent.

c. Prevent damage to the core if a partial loca of flow woro to occur later in the event.

. d. He local control of RCP auxiliarios in available. ,

d

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SENIOR Hih" TOR OPERATOR Ptige 74 e QUESTION: 069 (1.00) , eivon the following

  - Reactor power is 3 %
  - Reactor / Plant Stt.: tup in progress.

- NI-35 han. failed LOW

  - H1-36 in operable..        ,
  - I&C has been called to investigato.

WHICH ONE (1) of the following is the offect on the startup of this failuro? a. Verify / actuate reactor trip.

b. Power must be maintained loss than St.. c. Power must be reduced to lous than the POAH.

d. Rodo must bo driven in mars. ally to nhutdown the reactor.

.

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.  .- . .. . .
     . -. . - . ... -
          . --

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

SENIOR REACTOR OPERATOR Page 75

QUESTION: 070 (1.00) Given the followingt

- 2-EOP-FRSM-1, "Roopcnso to Nuclear Power Generation"  i has been entered,
- Charging flow is greater than-100 gpm.   '
- SJ1 and SJ2, RWST to Charging pump valves aro. HOT  >

open.

- Boric Acid pumps cannot be started.

WilICll 0140 (1) of the following actions must be taken to commenco rapid boration flow por 2-EOP-FRSM-1, " Response to Nuclear Power Generation"? a. Open cV-175,(Rapid boration valvo). > b. Start safety injection equipment manually. . c. Open SJ1 and SJ2 and establish boration from RWST through the BIT.

d. Initiate safety injection with actuation switchos, i.

l

   . .-  . .

SEllIOR REACTOR OPERATOR Pays 76 QUESTIONt 071 (1.00) WilICil ONE (1) of the following requires an immediato stop of the RCP's? a. Component cooling thermal barrior return valvo cc131-fails closed, b. Componont cooling lost to all RCPs.

c. Loss of soal injection flow to all ncPs.

d.-RCP motor bearing indicated-temperature of 180 degroon F for 2 minutos.

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i SE1410R PEACTOR OPERATOR Pago 77

           :
           ,

QUESTION: 072 (1.00) WilIC11 ONE (1) of the following is the basis for reducing T-ave to ' loss than 500 degrees F following a shutdown required by a Doso , Equivalent I-131 level greater than Technical Specification limit? a. Slows coolant / fuel roaction rato, immediately reducing the source term of the activity. ,

b. Prevents the release of activity following a steam generator tube rupture.- c. Minimizes the temperature related degradation of the CVCS domineralizers while the RCS clean-up is in progress. , d. Minimizes of the iodino spiking phenomena which occurs ,

due to the large change in TilERMAL POWER levol caused by the unit shutdown.

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SEllIOR REACTOR OPERATOR Page 78 , . 1 I

, QUESTION: 073   (1.00)        I

WIIICil ONE (1) of the following describes why the main turbino is tripped during all ATWS conditions? a. To initiato reactor shutdown from Doppler defect.

b. To mitigate the consequences of a loss of food ATWS.

c. To provc nt excessivo cooldown of the RCS. l

!
"

d. To provent exceeding the DNBR/LPD limits on the core.

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_ _ _ _ _ __ _ _ _ . .. SENIOR REACTott OPERATOR Pago 79 QUESTION: 074 (1.00) WilICl! ONE (1) of the following conditions 10 checked to verify thet the reactor is suberitical during performance of 2*EOP-FRSM-1, "Rosponse to Nuclear Power Generation"?

/#

a. IRPI all indicate zero.

b. Red bottom lights onorgized.

c. Negativo startup rato on Intermediate Range Indicators.

d. Power rango channels indicato poWor loss than 10%.

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

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, SENIOR REACTOR OPERATOR Page 80

4 A . i

QUESTION: 075 (1.00) ; i i Given the followings t i ! - S/G level is decreasing j i

 - Loop Delta-T is constant.

l

 - Containment pressure is increasing.         l
 - Steam Gonorator pressure is decreasing         j i

! WillCl! ONE (1) of the following transients is indicated by the - above conditions? f a. Steam Break insido containment b. Food Break inside containment c ' Steam Generator Tube Rupture a d. LOCA insido containment-

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

l

! Page 81 j SENIOP. REACTOR OPERATOR

L i ! QUESTION: 076 (1.00) , i Given the following:

A doublo ended main steam lino break insido containment has occurred.

j

 - Operators ar. responding to the event per 2-EOP-LOSC-1,
  " Loss of Secondary Coolant."

i WilICll ONE (1) of the following Will be available to the operator j to confirm the diagnosis of a SGTR'in the FAULTED S/G? , l , l

a. S/G sample'results. l l b. R-19 Blowdown radiation monitors.

I " c. R-15 Condensor Air Ejector radiation monitor. l

! l d. Uncontrolled S/G lovel increase. ! ! ,

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SENIOR REACTOR OPERATOR Page 82 QUESTION: 077 (1.00)

Given the followingt

- Vacuum in the main turbino condenser is decreasing.

- No cause has yet boon identified.

- A power reduction has commenced IAW directions in S2.OP-AB.COND-0001(Q), " Loss of Condenser Vacuum".

WilICil ONE (1) of the following is the reason that S2.OP-AB.COND- -'

0001(Q), "Losa of Condonsor Vacuum" limits the load decrease to loss than 5%/ minute? a. Minimizo turbino blado " flutter."

b. Prevent turbine' rupture disc failure, c. Maintain adequate vacuum for operation of condenser steam dumps.

' d. Prevent operation of the main condensor steam dumps.

, b

-,

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

Page 83 ; SE!1IOR REACTOR OPERATOR

        ;
        ,

QUESTION: 078 (1.00)

Given the following

 - Loss of All AC power has occurred.    ,.
 - 2-EOP-LOPA, " Loss of All AC Power and Recovery" is in-use.

' WilICl! 011E (1) of the following is the reason for isolating the VCT? a. CV-40 and 41, Charging Discharge valves, fail CLOSED on , loss of power.

b. CV-40 and 41, Charging Discharge valves, f all OPEli on loss of power.

c. Provent nitrogen and other non condensibics from interrupting natural circulation.

d. Auto makeup and RWST swapover are not available due to loss of power.

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        !

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

SENIOR REACTOR OPERATOR Pago 84 l QUESTION: 079 (1.00) ! Given the following:

- Loss of All AC power has occurred.

- 2-EOP-LOPA, " Lass of All AC Power and Recovery" is in use.

WHICH ONE (1) of the-following parameter trends verify that natural circulation has been established? a. RCS subcooling greater than 5 degrees.

b. RCS Narrow Range cold leg temperatures at saturation temperature for the pressurizer, c. RCS Wide Range Hot leg temperatures stable or decreasing.

d. Steam generator pressure slowly increasing.

._

_ _ _ .

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

. i Pago 85 SENIOR REACTOR OPERATOR i s QUESTION: 080 (1.00)

, WHICil ONE of the following sets of 115V Vital Instrument Duses ,

contains the TWO busos which would require an immediato i trip of the reactor if EITl!ER were de-energized? . ! a. 2A and 2B.

b. 2A and 2D.

t c. 2B and 20.

d. 2C and 2D

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SENIOR REACTOR OPERATOR Page 86

, !

' , 90ESTION: 081 (1.00)

;             ,

Given the following: )

  - Loss of 2A 115V Vital Instrument Bus has occurred.

,

  - All immediate action steps have been comple.ted.

- Pressurizer heaters are being operated in the " LOCAL- , MANUAL" mode for RCS pressure control.

i ' WilICl! ONE (1) of the following automatic pressurizer features is overridden by " LOCAL-MANUAL" operation of the pressurizer

heaters?

i

a. Safety Injection heater cutoff.

b. Pressurizer low lovel heater cutoff.

! c. Control group heators "ON/OFF" setpoint.

d. Pressurizer program / actual level deviation heater ! energization.

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l SENIOR REACTOR OPERATOR PGgo 07 QUESTION: 082 (1.00) Given the following:

     '
- Unit 1 is in Mode 6 with all circulators tagged out due to main condensor maintenance.

- Unit 2 in operating at 100% power.

- A release of radioactive liquid waste is necessary from tho /21 CVCS Monitor Tank.

WilICII ONE (1) of the following actions must occur in order.to-release the #21 CVCS monitor tank to Unit 1.

a. Component cooling water flow through the Unit 2 CCW !!X selected muct be greator than the required flow for the maximum relcaco rate on the permit.

b. Service water flow through the Unit' 2 CCW !!X selected must be greator than the required flow for the maximum release rato on the permit.

c. Total service water flow through both Unit 1-CCW HXs must be greater than the required flow for the maximum.

release rate on the permit, d. Discharge must go directly to.the Unit.2 Circulators, ' then through the sluice cross-connect to Unit l' discharge.

_ __ _ t w- sn -

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

SENIOR REACTOR OPERATOR Pago 88 QUESTION: 083 (1.00) Given the followingt !

 - 2-EOP-FRCC-1, " Response to Inadequate Core Cooling" is in use.

- RWST Low Level alarm has just energized. j

 - Red Paths exist on Core Cooling and lleat Sink. l WHICli ONE (1) of the following actions should be taken by the operating crew?

a. Transition to 2-EOP-LOCA-3, " Transfer to Cold Leg Recirculation" immediately, b. Transition to 2-EOP-LOCA-3, " Transfer to Cold Leg Recirculation" when Tshot and CET's show a definite decrease.

c. Transition immediately to 2-EOP-FRilS-1, " Response to Loss of Secondary llent Sink".

d. Remain in FRCC-1 until Core Cooling Purple Path established, than transition to 2-EOP-FRilS-1, " Response to Loss of Secondary lleat Sink".

_

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          !
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Page 89 SENIOR REACTOR OPERATOR

QUESTION: 084 (1.00) Given the followingt

A small break LOCA has occurred.

- Operators are responding-IAW 2-EOP-LOCA-2, " Post LOCA Cooldown and Deprosaurization".

WilICl! ONE (1) of tho following describes an appropriato action which would prevent an inadequato Coro cooling condition? , n. Maintain PZR levol greater than 20% and Pressuro g'ecater than 1000 psig to ensure subcooling can be maintained while steaming the S/G's.

b. Roduce ECCS flow and lower RCS presouro to increase heat removal through the break while maintaining a constant S/G pressuro.. , c. Increase heat removal via tho S/G's to increase cooldown and depressurization of the RCS allowing increased ECCS flow.

d. Stabilizo S/G pressure and levol, increano ECCS ! injection and RCS pressure to increase heat removal ' through the break.

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

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SrNIOR REACTOR OPERATOR Page 90 i i QUESTION 085 (1.00) Given the following

!

 - A less of RilR cooling has occurrod.

- Operator actions are being taken IAW S2.OP-AD.RllR-0001(Q), Loss of RilR.

! - The RIIR punips can not be rostered prior to the RCS reaching the projected boiling point.

- The following cooling methods are availablo: , 1. Cooling the RCS with S/G's.

2. Cooling the RCS with Spent Fuel Pool.

3. Ilot log injection.

4. Cold log injection.

WilICil OllE (1) of the following reflects the proforred order of implementation of the four (4) available cooling methods? a. 1, 2, 1, 4 b. 2, 3, 4, 1 c. 3, 4, 2, 1 d. 4, 1, 2, 3

.--  ,    . . _ _ . . _ . . . . _ . . _ . _ . . . , _ _ : .. - .- :2 ..~,.__~.,2-.

___.__.___._____._____.m._.______.__m.._..___._... _ _ . _ _ _ _ _ _ . _ . _ _ _ . . _

s i SEllIOR RFACTOR OPERATOR Pago 91

              .
              !

QUESTIOlis 086 (1.00) ;

              ;
              .

Given the followingt .

  - RilR is in servico at reduced inventory conditions.

- S 2. 0P- AD . R}iR-0001, " Loss of RilR" has boon implemented due to pump cavitation on the running pump. ,

WilICll 011E - (1) of the following could causo an observed increase in RCS loyal during performance of this procedure? a. Venting the RilR system.

b. opening GJ69, RilR suction from RWST.

c. Any. opening in the RCS boundary, d. liigh RilR flow rato when restarting pump. , r t

4-

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-- ._ _ _. . . .. ._ _ ._ __ _ _ _ _ __. _, . _ SENIOR REACTOR OPERATOR Pago 92

QUESTION: 087 (1.00) ,

Given the following

 - Unit 1 le at 100% power.   '
 - PT-474, Prossurizer pressura channel IV is failed LOW.

- All actions of OP.IV-10.3.1, " Removing, Returning to Service and Loss of a Protectivo System Channol" are

complete.

WHICH ONE (1) of the following describes the plant response to a failure of PT-455, Pressurizar pressura channel I, 1.0W? a. Reactor trip.

b. Reactor trip AHD safety' injection.

c. All heators energizo, pressuro increases to POT.V - setpoint, 1PR2 opens and IPR 1 does not, pressuro cycles around PR2 sotpoint.

d. All heators energize, pressure _ increases to PORV sotpoint, IPR 1 opons, 1PR2 does not, pressuro. cycles > around PR1 sotpoint.

,

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m.._ _ _ _ . . _ _ _ . _ _ _ _ _ . _ . . _ _ _ _ _ _ . _ . _ . . _.___._. _ _ ._.

Pago 93 SENXOR REACTOR OPERATOR QUESTION: 008 (1.00) WHICH ONE (1) of the following describes the reason that selection of-the same Auxiliary Building Ventilation exhaust unit on moro than one control bezel is not advisable? I a. Could prevent the vontilation system from maintaining negative prosauro on the Auxiliary Building, b. Could causo iuolation of one exhaust header.

c. Could prevent alignment of Post Accidunt ventilation for the Auxiliary Building.

d. Could overrido the thermostatic controls for the Coolers.

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SENIOR REACTOR OPERATOR Page 94 QUESTION: 089- (1.00) WilICll ONE (1) of the following determinos the temperature at which the RcS cooldown is TERMINATED whilo responding to a steam generator tube rupture per 2~EOP-SGTR-1, " Steam Generator Tube Rupture"? n. A temperaturo limit of 450 degrees F.

b. Temperature at which RilR can be placed-in service.

c. Ruptured steam generator pressure.

d. 50 degroos F subcooling in the RCS available.

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SENIOR REACTOR OPERATOR Pago 95 QUESTION 090 (1.00) Given the following

  - Reactor trip and SI have occurred.

- Operators aro using 2-EOP-SGTR-1, "Stoam Generator Tubo Rupture" to mitigate the event.

WHICH ONE (1) of the following describes tho applicablo RCP trip criteria when responding to the SGTR.

a. RCPs should be tripped anytimo during 2EOP-SGTR-1 when the RCP trip criteria are mot.

b. RCPs should be tripped during 2-EOP-SGTR-1 ONLY if the RCP trip critoria are mot at Step-I when the operator is specifically required to check the criteria, c. RCPs should be tripped during 2-EOP-EGTR-1 ONLY 11 the RCP trip criteria are not DEFORE beginning the cooldown and depressurization.

d. RCPc should be tripped during-2-EOP-SGTR-1 ONLY if the RCP trip critoria are mot JUPTER beginning the cooldown and depressurization.

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

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

      - . .  . . . .- .- .-
           !

SENIOR REACTOR OPERATOR Page 96 l l QUESTION: 091 (1.00) WilICH ONE (1) of the following Critical Safety Function conditions has the highest priority? a. ECCS is not in service, NO RCP's are running, and RVLIS full range is 96% b. RCS subcooling is 35 degroes F with one RCP running and ! RVLIS is 40% dynamic rango c. All S/G NR lovolo are less than 80 and total food flow ' l to the S/G's is 2E04 lb/hr.

d. Containment pressure is 25 psig i j l i.. I ( l l l l 5 l

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

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       . _ . - . - . - . - . - - . - . -
           !-

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pagg 97 SENIOR REACTOR OPERATOR : L i

QUESTION: 092 (1.00) j Wi!ICll ONE . (1) of the following occurs on a loss of Control Air? ;

a. CV18, Lotdown !! cat Exchanger outlet valvo fails OPEN.

b. #23 Charging pump speed control.er l falla'AS IS.

c. Feed regulating valves fall OPEN. i d. RCP ocal retut'n' valves fail'AS IS.

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_ _ - _ _ _ - _ _ _ _ _ _ _ _ - _ - _ - - - _ - - SENIOR REACTOR OPERATOR Pagn 98 QUESTION 093 (1.00) Givon the following

    - Unit 1 is at 100% power.

- S2.OP-AB.CA-0001(Q), " T<ssa of-Control Air" is in use by the operators.

Wi!ICll ONE (1) of the following requires a reactor. trip during performance of this procodure? a. Station Air heador pressure of 90 psig.

b. Control Air header pressuro'or'75 psig.-

c. Redundant-Air-panol fallo-to transfer to 2A header.

d. 2 or more Rod Drivo Vent Fan low flow alarms annunciato.

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_ _ --------- A u - - _ -- _-----_-.--_.---___--__-_-------.-_-.--_-__.-.a -___N.w-_a

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

i

Sr.HIOR REACTOR OPERATOR Pago 99 l I

i ' QUESTION 094 ( .1. 00) i j WilICil ONE (1) of the following is' HOT monitored by the critical

Safety Function Status Trees to determino loss of containment

integrity? j , a. Containmont hydrogon concentration. ! b. Containment temperature.

1

c. Conteinment uump lovel, i d. Containment pret.sure, s I h a

           :
           -
           ,
  , --,  . . _ . .  ,. ., , . . _ . . . . _ . _ ...~...,m..... _ . . , , . .

___ - . _ . . _ _ _ _ _ . _ . _ _ _ . . _ . . . _ _ ~ _ . . _ . _ . . - - . . . . _ _ . . . _ . _ . . . . _ _ _ - . _ . . . . _ _ _ . . _ . . . - _ _ _ _ . - . _ . SEllIon REACTOR OPERATOR Pago100 i

QUESTION: 095 (1.00)

WilICil ONE (1) of the following is an iminedtato action associated with the trip of a Steam Generator Feed Pu:ap? a. Bypasa Foodwater hoators.

D. Dypass condensato polishcro. f c. Start AFW nystem and take manual control of feed.

I d. Start additional condenanto pump.

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

SENIOR REACTOR OPERATOR Page101

 ..hSTION: 096 (1.00)

Given the following: A opent fuel assembly is being raised from its slot in the storage pool for return to the reactor.

Gas bubbles are'now coming to the surface of the pool.

- Level in the spent fuel pool is decreasing.

- Radiation in the Spent Puel area is increasing steadily.

WHICH ONE (1) of the following reflects the required response of personnel in the Puel handling building? a. Immediately evacuate all.personne.l.

b. Evacuate non-essential personnel when-radiation levels

exceed 250 nr/hr.

c. Evacuate all personnel when radiation levels exceed 1 R/hr d. Complete fuel transfe:r to reactor cavity prior to evacuation of all personnel.

l l

. .

I SENIOR REACTOR OPERATOR Page102

QUESTION: 097 (1.00) ;

. Given the following:

    .
- Reactor is at 100% steady state power BOL.

- All control systems are in automatic.

WHICH ONE (1) of the following will occur as a result of the controlling prosaurizer level instrument failing o (Assume no > , operator actions are taken.)

a. The reactor will trip on high-pressurizer. level.

~ b. The reactor will-trip on'high pressurizer pressure.

2. The level in the pressurizer will stahlf.ize at the letdown isolation setpoint.

' I d. The reactor will trip on low pressurizer pressure.

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   -. -

SENIOR REACTOR OPERATOR Page103 l QUESTION: 098 (1.00) WHICH ONE (1) of the following actions of 2-EOP-LOPA-1, " Loss of All AC Power" is performed in order to limit the potential for a radioactive release inside the Auxiliary Building? a. Isolation of the Seal Return line, b. Isolation of the RCP seal injection lines.

c. Isolation of the thermal barrier CCW return line, d. Isolation of the PORV'stop valves . s .I A

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

SENIOR REACTOR OPERATOR Pcgo104 QUESTION: 099 (1.00) Given the following:

A LOCA outside of containment has occurred on Unit 1.

- Operators'are using 1-EOP-LOCA-5, " Loss of Emergency Recirculation", due to the inability to isolate the leak. .

- RWST level is less than the low level alarm setpoint.

WHICH ONE (1) of the following is the reason for depressurizing the S/G's to less than 675 psig in 1-EOP-LOCA-5, " Loss of Emergency Recirculation"? a. To allow the S/G's to be fed from the condensate pumps.

b. To ensure adequate subcooling for restart-of the RCP's.

c. To set up conditions for controlled injection of the accumulators.

d. To ensure that the RHR pumps are injecting at maximum flow for as long as possible while attempting to makeup to the RWST.

 (**********'END OF EXAMINATION **********)'
  - , . -- ,

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

SENIOR REACTOR OPERATOR Page106 ANSWER: 001 (1.00)

* ANSWER b. [+1.0)

, REFERENCE: i * REFERENCE 1. SALEM: AD-17, Requirements of the New' Jersey Pollutant

Discharge Elimination System-(NJPDES) Permit. Directives 1 Pg. 1.

2. KA 194001K108 (3.5/3.4) 194001K108 ..(KA's) j ANSWER: 002 (1.00) d. [+1.0] < REFERENCE:

1. SALEM: SC.OP-DD.ZZ-AD46(Q) " Troubleshooting Abnormal Plant , Conditions", Section 3.1.

2. KA 19400A103 (2.5/3.4) 194001A103 ..(KA's) ANSWER: 003 (1.00) d. [+1.0) REFERENCE: 1. SALEM: AD-45, Technical Specification Interpretation Program Sec 3.1.13 2. KA 194001A102 ( 4 .1/ 3. 9 ) 194001A102 . .. (KA's) ANSWER: 004 (1.00) a. [+1.0] REFERENCE: 1. SALEM: LP No. 300S-000.00S-RCPUMP-04, ELO 2, Pg. 17 2. KA 003000K407_(3.2/3.4) 003000K407 ..(KA's)

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

SENIOR REACTOR OPERATOR Pege107 ANSWER: 005 (1.00) d.- [+1.0)

REFERENCE: 1. SALEM: AD-44, Emergency / Abnormal Procedures Program, Pg. 24 of 45 Step 6.3.3.D 2. SALEM L/P - 300S-EOPINT-02, Introduction to the Use of the EOP's, Obj. 10, Pg 13.

3 KA 194001A102 ( 4.1/ 3. 9 ) 194001A102 ..(KA's) ANSWER: 006 (1.00) b. [+1.0] REFERENCE: 1. SALEM: OD-7, System Alignment,.Pg 2 Sec 2.3.1 2. KA 194001K101 (3.6/3.7) 194001K101 ..(KA's) ANSWER: 007 (1.00)- d. [+1.0] REFERENCE: 1. SALEM: OD-6, " Circuit Breaker Realosure Policy Following a Trip", Pg 1.

2. KA 194001K107 (3.6/3.7) - 194001K107 ..(KA's) ANSWER: 008 (1.00) b. [+1.0] REFERENCE: 1. SALEM: OD-7, System Alignments, Pg 5, Sec. 3.1.5 2. L/P 300S-000,00S-ARIA 00-08, Administrative Requirements, Objf 4, Pg 26-27.

3 KA 194001K101 (3.6/3.7) 194001K101 ..(KA's)

      ,
     ,.

Page108 --SENIOR REACTOR OPERATOR ANSWER: 009 '(1.00)

-G. [+1.0)

REFERENCE: 1. SALEM: OD-8, TRIS Tagging Operations, Sec. 3.0, Pg. 9.

2. KA 194001K102_(3.7/4.1) 194001K102 ..(KA's) ANSWER: 010 (1.00) c. [+1.0) REFERENCE: 1. SALEM: NC.NA-AP.ZZ-0015, Safety Tagging Program,- Sec.S.I.26, pg 13-2 KA 194001K102 (3.7/4.1) 194001K102 ..(KA's) ANSWER:- 011 (1.00) a. [+1.0] REFERENCE: 1 SALEM: NC.NA-AP.ZZ-0015, Safety Tagging Program, Section 6.11.1 Pg. 52.- , 2. KA 194001K107 ( 3 . 7 / 4 .1) 194001K107 ..(KA's) ANSWER: 012 (1.00) a. [1.00]

-REFERENCE:

1. SD-42 " Electric Distribution" 2. KA 000058A203 (3.5/3.9): 000058A203 ..(KA's)

l

- ,

___ SENIOR REACTOR OP" ATOR Pago109 ANSWER: 013 (1.00) b.- [+1.0) REFERENCE: 1. SALEM: NC.NA-AP.ZZ-0938, Chemical Control Program, Sec.

5.5.1-5.5.2, Pg. 14.

2. KA 194001K111 (3.4/3.5) 194001K111 ..(KA's) ANSWER: 014 (1.00) b. (+1.0] REFERENCE: 1. -SALEM: . Technical Specifications, Section 6.2.20, Table 6.2 2. KA 194001K116 (3.5/4,2) 194001K116 ..(KA's) ANSWER: 015 (1.00) a. [+1.0] REFERENCE: 1. SALEM: AD-44 - Emergency / Abnormal Procedures Program, Sec 6.3.2, Pg 23.

2. KA 194001A111 (2.8/4.1) 194001A111 ..(KA's) l ANSWER: 016 (1.00) c. (1.0]. REFERENCE:. 1. SALEM: NC.NA-AP.ZZ-0024,-Radiation Protection Program,.Pg.8.

2. 10 CFR 50.20, Emergency. Dose Rate Allowances.

3. SALEM: L/P # 0215-000.00B-000002-01, EP training, i

SNSS,NSS,NSTA,NCO Duties / Responsibilities,'Obj. 44,_Pg.45; 4. KA- 194001K103 (2.8/3.4) 194001K103 ..(KA's) l l-l^ I l ! !' i e 2 4'

. -.  .. . . - - . . -- .. - - -

i ! SENIOR REACTOR OPERATOR- Page110 i . ANSWER: 017 (1.00)- c. [+1.0) , REFERENCE: ' 1. SALEM: AD-37 - Key Control, Soc. 3.4.2, Pg 3.

2 KA 194001A110 (2.9/3.9) + 194001A110 ..(KA's) ANSWER: 018 (1.00) d. [+1.0) REFERENCE: 1. SALEM: LP No. 300S-000.00S-ROC 000-01, ELO 2.1, Pg. 31.

2. SALEM: Proc. S2.OP-AB. ROD-0002(Q), NOTE Pg. 6 , 3 KA 001050A201 (3.7/3.9) 001050A201 ..(KA's) ANSWER: 019 (1.00) c. (+1.0) REFERENCE: 1. SALEM: LP No. 300-000.00S-ROC 000-01, ELO 3d, Pg. 48, 2. KA 014000A103 (3.6/3.8) 014000A103 ..(KA's) ANSWER: 020 (1.00) d. [+1.0]

-REFERENCE:

1. SALEM: 300-000.00S-EXCNIS-07, ELO 3.i, Pg 57, 2. KA 015000K405 (4.3/4.5) 015000K405 ..(KA's)

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

' SENIOR REACTOR' OPERATOR Page111 ANSWER: 021 (1.00) c. (+1.0) REFERENCE: 1. SALEM: Proc. No IV-10.3.1, Removing, Returning to Service and Loss of a Protective System Channel, Table 1 2. SALEM: LP No. 300S-000.00S-IOP500-01, Minimum Load to Hot Standby, ELO 4, Pg 6.

3. KA.015000K407 (3.7/3.8) 015000K407 ..(KA's) ANSWER: 022 (1.00) b. (+1.0) REFERENCE: 1. Salem: L/P 300S-000.00S-INCORE-04, ELO 2-3, Pg 44.

2. KA 017020A401 (3.8/4.1) 017020A401 ..(KA's) ANSWER: 023 (1.00) c. [+1.0] REFERENCE: 1. SALEM: 300S-000-00S-CVCS00-02, ELO 2a, Pg 29, 2. KA 004020K406 (2.3/2.7) 004000K406 ..(KA's) ANSWER: 024 (1.00) a. [+1.0)- REFERENCE: 1. Salem Unit-2 Procedure S2.OP-SO.CVC-0001(Q), Precaution-3.8, Pg 3.

2. KA 004010K403 (3.1/3.6) 004010K403 ..(KA's)

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

SENIOR REACTOR OPERATOR Pcgs112 ANSWER: 025 (1.00) ,

d. [+1.0) REFERENCE: 1. Salem L/P - 300S-000.00S-ECS00-07, ELO 2d, Pg. 32-33.

-2. Salem P&ID 205234 ' 3. KA 013000K201 (3.6/3.8) 013000K201 ..(KA's) ' ANSWER: 026 (1.00) d. [ +1. 0 ] -- ' REFERENCE: 1. Salem L/P - 300S-000.00S-ECS00-07, ELO 2f Pg. -79.

2. Salem Technical Specifications 3.5.1, Accumulators.

3. KA 006000A301 (4.0/3.9) 006000A301 ..(KA's) A5kSWERt 027 (1.00) x 3gl._.

a. [+1.0}s - /7 f) REFERENCE: '- 1. Salem L/P .00S-000.0 CSPRAY-06, ELO 4, _Pg. 15-16.

2.-Salem Logic' Diagram 221057 3. K3.ATI6000A301 (4.3/4.5)

-c26000K201 ..(KA's)

ANSW R: 028 (1.00)

"   && If7/1)

RErERE1cE: $ / g ' 1. S- 0.00S-CCWOOO-01, ELO 11, Pg. 14.

2.

SalemL/P,[- KA 00800pA 02 (2.9 1) 008000hf02 ..(KA's) . ew-

l l SENIOR REACTOR OPERATOR Page113

ANSWER:- 029 (1.00)

! b. .(+1.0]

REFERENCE: , 1. Salem L/P - 300S-000.00S-CCWOOO-01,-ELO 17a,-Pg. 46, 2. Salem Proc. 11-7.3.2, " Component Cooling Water System - Normal Operations" Precaution 3.3.

3. KA 008000G010 (3.1/3.2) 008000G010 ..(KA's) ANSWER: 030 (1.00) a. [+1.0) REFERENCE: 1. Salem L/P - 300S-000.00S-AFEEDW-10, ELO 7c,.Pg. 31.

2. KA 061000K406 (4.0/4.2) 061000K406 ..(KA's) ANSWER: 031 (1.00) c. [1.0) REFERENCE: 1. Salem L/P - 300S-000.00S-AFEEDW-10, ELO 7a/b, Pg. 32.

2. KA 061000K402 (4.5/4.o) 061000K402 ..(KA's) ANSWER: 032 (1.00) c. [+1.0] REFERENCE: 1. Salem L/P - 300S-000.00S-WASGAS-02, ELO 12a, Pg. 21.

.2. Salem Proc. OP-II-12.3.1, "Geseous Waste Disposal System Normal Operations" Precaution 3.4.

3. KA 071000A409 (3.3/3.5) 071000A409 ..(KA's)

 .   .
    .

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

        .. .
         .. .. . .. ..

SENIOR REACTOR OPERATOR Pagal14 ANSWER: 033 (1.00) d. [+1.0) or (_., REFERENCE: b f7fD 1. Salem L/P -300S-000.00S-WASGAS-02, ELO 10, Pg. 20.

2. Salem Tech Specs, 3.11.2.1, Bases, B 3/4.11-3 3. KA 071000G006 (1.9/2.9) 071000G006 . . (KA's) ANSWER: 034 (1.00) a. [+1.0) REFERENCE: 1. Salem L/P - 300S-000.00S-WASLIQ-02, ELO 6, Pg. 19, 2. Salem Proc. II-11.3.2b, Pg 6, 5.2.11.b.

3. KA 000059A101 (3.5/3.5) 000059K301 ..(KA's) ANSWER: 035 (1.00) a. [+1.0) REFERENCE: 1. Salem L/P, 3005-000.00S-EXCNIS-07, Obj. 3b, Pg 55.

2. Salem L/P, 300S-000.00S-SGHLCS-03, Obj. 1, Pg 13, 3. KA 000015K104 (2.9/3.1) 000015K104 ..(KA's) ANSWER: 036 (1.00) d. [+1.0] REFERENCE: 1. Salem L/P, 300S-000.00S-RMS000-00, Obj. Id/e, Pg 70-73.

2. Salem Proc. IV-11.3.1, Table 2 3. KA 072000K402 (3.2/3.4)- 072000K402 ..(KA's)

,
    - _ _ - _ _ _ _ _ _ - _ _ _ _ _ - _ _ -

 .  . . - . _

i SENIOR REACTOR OPERATOR Page115 . ANSWER: 037 (1.003-c. .(+1.0)

, REFERENCE:

$ 1. Salem S/D, 302S-SD/1:19, Pg 15.

l 2. Salem L/P 300S-000.00S-RCPUMP-04, Obj. 13,-Pg. 28.

' 3. KA 003000K407 (3.2/3.4)

.' 003000K407 ..(KA's)

d ANSWER: 038 (1.00) a. [+1.0] . REFERENCE: 1. Salem S/D, 302S-SD/1:25, Pg 2 2. Salem L/P 302/304-145.12-RPI-04, Obj. 1/2.1.3, Pg 8.

< 3. KA 014000A202 (3.1/3.6) 014000A202 ..(KA's) ' AMSWER: 039 (1.00) ' b. [+1.0)

REFERENCE: 1 1. Salem Proc. OP-II-1.3.4 " Filling and Venting.the RCS", Precaution 3.

, 2. Salem L/P 300S-000.00S-RCS000-05 Obj. 12, Pg'27 ,~ 3. KA 002000G010 (3.4/3.9) 4 002000G010 ..(KA's) . ! ANSWER: 040 (1.00)

b. (+1.0) . REFERENCE: 1. Salem S/D Chapter 24, Reactor Vessel Level Monitoring System 2. Salem L/P 300S-000.00S-RVLISO-05 Obj. 2 - 3. KA 002000K107 (3.5/3.7) - 002000K107 ..(KA's)

-SENIOR-REACTOR OPERATOR:   Pagn11C ANSWER: 041 (1. 00);
!

- b. [+1.0) REFERENCE: I 1. Salem ~P&ID 205201, D-2

2. Salem L/P 3005-0CO.00S-PZRP/L-00 Obj. 2 Pg 35.

3. KA 010000K601 (2.7/3.1)

010000K601 ..(KA's) . , ANSWER: 042 (1.00) c. [+1.0) REFERENCE: 1. Salem Tech Spec 3.10.3, Pg 3/4 4-31.

2.-. Salem L/P 300S-000.00S-PZRP/L-00 Obj. 12, Pg 37.

3. OP-II-2.~3.4, " Pressurizer Overpressure Protection Operability Check and Arming of 2PRI and 2PR2".

4. KA 010000K403 (3.8/4.1) 010000K403' ..(KA's) ANSWER: _043 (1.00)- d. [+1.0) REFERENCE: 1. Salem S2.OP-SO.RCP-0002(Q), Reactor Trip or Bypass Breaker Operation, Precaution 3.1.

2. Salem L/P 300S-000.00S-RXPROT-08 Obj. 2o1Pg_57-58.

3. KA 012000A202 (3.6/3.9) 012000A202 ..(KA's) ANSWER: 044 (1.00) c. [+1.0) REFERENCE: 1. Salem 2-SOP-TRIP-1,- Entry conditions.

2. Salem L/P 300S-000.00S-RXPROT-00 Obj. 15 Pg 62.-

'3. KA 012000A306 (3.7/3.7)

012000A306 ..(KA's)

  - . -  - . , . .,

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

-_

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

        ,      .

m SENIOR _ REACTOR OPERATOR: L -

           . Pago117:  ,

ANSWEh, 04 5 _ . (1. 00) _.

' b./(+1.0)- REFERENCE:- _1. Salem'L/P_300S-000._00S-CONTMT-02 Obj.112 lpg 60.

-

2. KA 027000A403'(3.3/3.2)- 027000A403 ..(KA's)

1 ,

      {.
   ~

SWER: 046 (1. 00)-: a.- (+ . g t/ ) REFERENC2: 1.' Sal d L/P 300S- 00S-CONTMT-023Obj.-;8 Pg 106_.

/ VKA 028000K503 (2.9/3.4)s

 .028000K503 ~  ..(KA's)

ANSWER: 047 (1.00) ._, c. (+1.0] REFERENCE:

1. Salem L/P 300S-000.00S-CONTMT-02 Obj. 15 Pg 111.j __

2. Salem Proc II-15.-3.1,." Hydrogen ~Recombiners --Normel-

Operation" Steps; 5.1. 6-5.1.' 8, ! Fig .- 1.

3. KA ~ _~ 028000A201 :(2. 9/ 3. 6) i:- 028000A201_ ...(KA's) 048 '(1.00)

i. NSWER:- - c. liO) q)-

           '
 . REFERENCE: .
 -

1..S2.OP-AB.'SG, ,ESteam GeneratorcTube-Leak,'AOP Basis D.f7.i 2.-.300S-000. dS-ABS -00 =. Steam; Generator . Tube . Leak, . Obj . < 3',1 p. ; 7-) ' 3. KA 000 307 (4.2/4.

- -

1000 7K307 -..(KA's

  -.
-

l:_ k J . , , f-

'

<
-        "

g f

    -,
+ v   e  sw., y y<- ) , & ,.e- m , . +,.+.#w 4- ,.e-,cwe. . r. .. - . c-r<w - em -w a

--. _ _
'

i SENIOR REACTOR OPERATOR Page118 i ANSWER: 049 (1.00) d. [+1.0) REFERENCE: 1. Salem S/D Chapter 25 - Stoam Generator Wa'-ar Level Control, Pg. 7.

2. KA 035010K601 (3.2/3.6) 035010K601 ..(KA's) ,

-ANSWER: 050 (1.00)

d. [+1.0) REFERENCE: l 1. Salem L/P 300S-000.00S- Obj. 2, Pg 13, 2. Salem S2.OP-AB. RAD-0001 (Q), Area Radiation Monitors, Att.3.

3. KA 000061A101 (3.6/3.6) 000061A101 ..(KA's) , ANSWER: 051 (1.00)

d. [+1.0)

REFERENCE: 1. Salem L/P 300S-000.00S-FIRPRO-03, Obj. 7b, Pg 71-72.

2. KA 086000A204 (3.3/3.9) 086000A204 ..(KA's) . . RNSWER: 052 (1.00)

c. [+1.0)

REFERENCE: 1. Solem L/P 300S-000.00S-CWATER-03, Obj. 2.

' 2. S2.OP-SO.CW-0001, " Circulating Water System Operation 3. KA 062000A201 (3.4/3.9) 062000A201 ..(KA's)

SENIOR REACTOR OPERATOR Page119 ANSWER: 053 (1.00) c. (+1.0) REFERENCE:

1. Salem L/P 300S-000.00S-4KV000-05, Obj. 12, Pg 56, 2. KA 062000K403 ( 2. 8 / 3.1) 062000K403 ..(KA's) . ANSWER: 054 (1.00) b. (+1.0] REFERENCE: 1. Salem L/P 300S-000.005-DIESEL-05, Obj. 4, Pg 38.

2. S2.OP-SO.DG-0001(Q), Pg. 15 3. KA 064000A401 (4.0/4.3) 064000A401 ..(KA's) ANSWER: 055 (1.00) a. (+1.0) REFERENCE: 1. Salem L/P 300S-000.00S-DIESEL-05, Obj. 7 Pg 31, 2. S2.OP-SO.DG-0001(Q), Pg. 14 3. KA 064000K401 ( 3 . 8 / 4 .1) 064000K401 ..(KA's) ANSWER: 056 (1.00) c. (+1.0) REFERENCE: 1. Salem L/P 300S-000.00S-RO-05, Refueling Operations, Obj 36, Pg 85 2. OP-II-8.3.3, Filling the Reactor Refueling Cavity, Pg 1 3. KA 034000A102 (2.9/3.7) 034000A102 ..(KA's)

SENIOR REACTOR OPERATOR PagO120 ANSWER: 057 (1.00) a-[+1.0) REFERENCE: 1. Salem I,/P 300S-000.00S-MNTURB-00, Obj. 12b, Pg 54, (Old reference, contradicts new procedure setpoint).

1. S2-OP-0T.TRB-0001(Q), " Turbine Auto Trip Mechanism Operational Test"., Precaution 3.8, Pg 3 3. KA 045000G014 (2.8/2.9) 045000G014 ..(KA's) ANSWER: 058 (1.00) - d. [+1.0) REFERENCE: 1. Salem L/P 300S-000.00S.RHR000-06, Obj. 16, Pg.36.

2. S2.OP-SO.RHR-0001(Q) - Initiating RHR, Precaution 3.9.2, Pg 4.

3. KA 005000A102 (3.3/3.4) 005000A102 ...(KA's) ANSWER: 059 (1.00) a. [+1.0) REFERENCE: 1. S2.OP-AB.RHR-001 (Q), " Loss of RHR", Attachment 1, p.15 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 same. 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 core.

The question doea not require detailed knowledge of the-saturation vs time-curve due to the very large time frame of.

the incorrect distractors.... 2. KA 000025K101 (3.9/4.3) 000025K101 ..(KA's)

-- ..- . - - . .  . .

SENIOR REACTOR OPERATOR Page121 ANSWER: 060- (1.00) a. [+1.0) REFERENCE: 1. Salem L/P 300S-000.00S-ABROD02-00 Obj 3 Pg 11 . 2. S2.OP-AB. ROD-0002(Q), Dropped Rod, Step 3.34, Caution.

3. KA 000003G007 (3.4/3.6) 000003G007 ..(KA's) ANSWER: 061 (1.00) d. [+1.0) REFERENCE: 1. Salem.L/P 300S-000.00S-ABROD03-00 Obj 3 Pg 8.

2. S2.OP-AB. ROD-0003(Q), Continuous Rod Motion Basos Document, Pg 5-6.

3. KA 000003A205 (4.4/4.6) , 000001A205 ..(KA's) ANSWER: 062 (1.00) d. [+1.0) REFERENCE: 1. Salem L/P 300S-000.00S-TRIP 02-03 Obj 2a Pg. 7.- 2. S2.EOP-ES-0.1, TRIP 2, Reactor Trip Response Bases Document, Pg 12.

3. KA 000005A203 (3.5/4.4) 000005A203 ..(KA's) i l ANSWER: 063 (1. 00 ) -

'c. .[+1.0]

REFERENCE: 1. Salem L/P 300S-000.00S-FRCC00-02 Obj 2.2, Pg 11.

2. 2-EOP-FRCC-1, " Response to Inadequate Core Cooling", Sheet.4,

Step 34.

3. KA 000074K308 .(4.1/4.2) 000074K308 ..(KA's) i

'

l

     . _ _ _ _ . -

c. _ . . . s SENIOR REACTOR OPERATOR - Page122 ANSWER: 064 (1.00) b. (+1.0] REFERENCE: 1. Salem L/P 3005-000.00S-RCPTAA-01 Obj 1, Pg 11.

2.. 2-EOP-LOCA-1, " Loss of Reactor Coolant",_ Sheet 1, CAS.

3. KA 000009K323 (4.2/4.3) 000009K323 ..(KA's) ANSWER: 065 (1.00) b. '[+1.0)

-REFERENCE:

1. Salem.L/P 300S-000.00S-LOCA02-02 Obj 2d, Pg 8, 2. 2-EOP-LOCA-1, "" Loss of Reactor Coolant", Step 13, Continuous * Action Step.

3. KA 000009K321 (4.2/4.A, 000009K321 ..(KA's) ANSWER: 066 (1.00) a. (+1.0] REFERENCE: 1. Salem L/P 300S-000.00S-SBLOCA-00 Obj 2, Pg 9.

2. 2-EOP-TRIP-1, " Reactor;-Trip or Safety Injection", Step 27-28 3. KA 000008A108- (3.8/3.8) 000008A108 ..(KA's) ANSWER: 067 (1.00) d. (+1.0) REFERENCE: 17 Saleu L/P 300S-000.00S-AOPEVA-00 Obj 2,~Pg.6 2.'S2.OP-AB.CR-0001(Q), " Control Roota-Evacuation", Step 3.18. , z 3. KA- 000068K318 - (4. 2/ 4. 5) - 000068K318 ..(KA's)

. _

_ _ .

  - _ - . - __-__ _ _   _

SENIOR REACTOR OPERATOR .Paga123 ANSWER: 068 (1.00) a. [+1.0) REFERENCE: 1. Salen L/P 300S-000.00S-AOPFIR-00 Obj 2, Pg 6-2. S2.OP-AB.CR-0002(Q), " Control Room Evacuation Due To Fire in-Control Room, Relay Room,.or Ceiling of the 460/230V-Switchgear Room ", Subsequent action I, Step 3.9.

3. S2.OP-AB.CR-0002(Q), "Centrol Room' Evacuation Duo To Fire in Control Room, Relay _ Room, er_ Ceiling of the 460/230V- s ' Switchgear' Room Technical. Basis Document", Pg 8.

4. KA 000068A112 (4.4/4.4) 000069A112 ..(KA's) ANSdER: 069 (1.00) b. [+1.0]

-REFERENCE:

1. Salem L/P 300S-000.00S-ABNIS1-00 Obj 3A, Pg. 8 2. Salem T.S. 3.3.1, Table 3.3-1, Item 5, Action 3 3. KA 015000K301 (3.9/4.2) 015000K301 . (KA's) ANSWER: 070 (1.00)_ c. [+1.0] REFERENCE: 1. Salem L/i' 300S-000.00S-FRSM01-01 Obj 6, Pg 9.- 2. Salem 2-EOP-FRSM-1, Step 2.

3. KA 000029G010 (4.5/4.5) 000029G010 ..(KA's) ANSWER: 071 (1.00) b. [+1.0) REFERENCE: 1. Salem'L/P 300S-000.00S-ABRCP1-00 Obj 1, Pg 12 . . 2. S;.OP-AB.CC-0001(Q), " Component Cooling Abnormality", RCP trip:

-criteria.   ~

3 KA- 000026G010 ~ ('3. 6/ 3. 5) 000026G010 ..(KA's) A

  ~
  ----  ---_--_.-_-_____..-__.____.-_.,_____LL___ _ __ _ _,_,
        ;__________ , _ _ _ _ __

 -~ - - -  . -

Page124 SENIOR REACTOR OPERATOR.

ANSWER: ~ 072 (1.00) b. (+1.0) REFERENCE: 1. Salem Technical Specifications Bases 4.4.9.

2. KA 000076G004 (2.1/3.7) 000076G004 ..(KA's) ANSWER: 073 (1.00) b. [+1.0) REFERENCE: 1. Salem L/P 3005-000.00S-FRSM01-01 Obj 2, Pg 8.

2. 2-EOP-FRSM-1, Response to Nuclear Power. Generation" Basis, Pg.

2.

3 KA 000029K312 (4.4/4.7) 000029K312 ..(KA's)- ANSWER: 074 (1.00) c. (+1.0) REFERENCE: 1. Salem L/P 300S-000.00S-FRSM01-01 Obj 2, Pg 8.

2. 2-EOP-FRSM-1, Response to Nuclear Power Generation" Step 11.

3. KA 000029A201 (4.4./4.7) 000029A201- ..(KA's) ANSWER: 075 (1.00) b. (+1.0) REFERENCE: 1. Salem L/P 300S-000.00S-LOSCTA-C2, Loss of Secondary Coolant Transients Description, Obj. 1.2, Pg 13.

2. KA 000040A201 -(4.2/4.7)- 000040A201 ..(KA's)' l l

:
'

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

.__-_  . _

SENIOR REACTOR OPERATOR. Page125 ANSWER: 076 (1.00)' a. (tl.0) REFERENCE: 1. Salen L/P 300S-000.00S-LOSC01-02, Loss of Secondary Coolant Obj. 5 Pg 10, 2. Salem 2-EOP-LOSC-1, " Loss of Secondary Coolant", Step 11.. 3. KA 000040K304 (4.5/4.7) 000040K304 ..(KA's) , ANSWER: 077 (1.00) d. [+1.0) REFERENCE: 1. Salem S2.OP-AB.COND-0001(Q), " Loss of Condenser Vacuum", Pg 1 Note.

2. Salen L/P - 3005-ABCOND-00,-Loss of Condenser Vacuum, Obj. 3, Pg. 5.

3. KA 000051G007 (2.3/2.5) 000051A202 ..(KA's) ANSWER: 078 (1.00) d. (+1.0) REFERENCE: 1. Salem 300S-000.00S-LOPA00-01, " Loss of ALL AC and Recovery", Obj. 3, Pg. 13.

2. Salem EOP-LOPA-1, " Loss of ALL AC and Recovery", Step 16 Bases.

3. KA 000055K302 (4.3/4.6) 000055K302 ..(KA's) ANSWER: 079 (1. 0 0 ) . l c. [+1.0) REFERENCE: 1. Salem 300S-000.00S-LOPA00-01, " Loss of ALL AC and Recovery", Obj. 5, Pg. 14.

2. Salem EOP-LOPA-1, " Loss of ALL AC and Recovery", Step 34.

3. KA 000055A202: (4.4/4.6)

000055A202- ..(KA's)

l l : l'

   . . . _ - ~ . - .

 -   _
     ,

SENIOR REACTOR OPERATOR Page126 ANSWER: 080 (1.00) d. _

[+1.0)

REFERENCE: 3. Salem 300S 000.00S-115VAC-05, "115 VAC Control Power System",0bj. 5, Pg. 24.

'2. Salen S2.OP-AB.115-0001-4, " Loss of 2A-2D Vital Instrument Buses, Immediate Actions.

3. KA 000057G010 (3.6/3.7) 000057G010 ..(KA's) ANSWER: 081 (1.00) b. (+1.0) REFERENCE: 1. Salem 300S-000.00S-PZRPRT-05, " Pressurizer and Prensurizer Relief Tank, Obj. 2f, Pg 14,24.

2. Salem S2.OP-AB.115-0001-4, " Loss of 2A-2D Vital Instrument Buses, Note prior,to Step 3.3.

3. KA 000057A106 (3.5/3.5) 000057A106 ..(KA's)

ANSWER: 082 (1.00) b. [+1.0) ' RSFERENCE: 1. Salem S2.OP-SO.CW-0001(Q), " Circulating Water ~ Pump-Operation", precautions and limitctions.

. 2 ., Salem II-11.3.2b, " Release of-~ Radioactive Liqui.d to the Circulating Water. System from 21 or 22; Monitor Tanks", Note, Pg 3.

3. KA' 000059A204 (3.2/3.5) , 000059A204 ..(KA's) . T

  . . . , . a y..,. n , , , ,-

- . .  -. . . - - .

SENIOR REACTOR OPERATOR Page127 ' ANSWER: 083 (1.00) u. (+1.0) REFERENCE: 1. Salem 300S-000.00S-FRCC00-02, "HCD/EOP FRCC1,2,3, Obj. 6 Pg 8.

3. Salem 2-EOP-FRCC-1, " Response to Inadequate Core Cooling", - Caution prior to Step 1.

3. KA 000074G010 (4.5/4.6) 000074G010 ..(KA's) j ANSWER: 084 (1.00) c. [+1.0) REFERENCE: 1. Salem 300S-000.00S-LOCA2C-05, Post LOCA Cooldown and Depressurization, Obj. 2d, Pg 8 2. Salem question bank #000284 3. KA 000011A113 (4.1/4.2) 000011A113 ..(KA's) . ANSWER: 085 (1.00) c. [+1.0] REFERENCE: 1. Salen 300S-000.00S-ABRHRO-00, Loss of RHR, Obj. 2, Pg E.

2. Salem S2.OP-AB.RHR-0001, " Loss of RHR, Step 3.31 Attachment ' lict.

3. KA 000025K301 (3.1/3.4) 000025K301 ..(KA's)

-ANSWER: 086 (1.00)    i b. [+1.0]

REFERENCE: 1. Salem 300S-000.00S-ABRHRO-00, Loss of RHR, Obj. 3a, Pg 6.

2. Salem S2.OP-AB.RHR-0001, " Loss of RHR, Step 3.14 Cautions.

3.-KA 000025A102 (3.8/3.9) 000025A102 ..(KA's) l I

. SENIOR REACTOR OPERATOR Pago128 ANSWER: 087 (1.00) n. {+1.0) REFERENCE:

1. Salem 300S-000.00S-ICF000-03, I&C FAILURE ANALYSIS, Obj. 20, Pg. 50.

2. Salem Logic Diagram 221055, Reactor Protection System

Pressurized Trip Signals. .

3. KA 000027A216 (3.6/3.9) 000027A216 ..(KA's) ! ANSWER: 088 (1.00) , b. [+1.0)

REFERENCE: 1. Salem 300S-000.00S-AOPVNT-00, Unit 1&2 Ventilation Malfunctions, Obj. 2, Pg. 6.

L 2. Salem II-17.3.1 " Auxiliary Building Ventilation Operation, Precautions.

3. KA 000060K202 ( 2. 7 / 3 .1) ~ 000060K202 ..(KA's)

ANSWER: 089 (1.00) ' c. [+1.0)

REFERENCE:

1. Salem 300S-000.00S-SGTR01-04,_ Steam Generator,--Obj. 6c, Pg. 9 2. Salem 2-EOP-SGTR-1, " Steam Generator Tube Rupture, Step 18.

3. KA 000038A136 (4.3/4.5) 000038A136 ..(KA's) , ANSWER: 090 (1.00) c. [+1.0) REFERENCE: 1. Salem 300G-000.00S-SGTR01-04, Steam Generator, Obj. 60,.Pg. 9

'2.' Salem 2-EOP-SGTR-1, " Steam Generator Tube Rupture, . Step 15.

3. KA 000038K308 ( 4 .1/ 4 '. 2 ) = 000038K308 ..(KA's) . A

 = _ . -  -  1 .

I SENIOR REACTOR OPERATOR Pago129-ANSWER: 091 (1.00) c. (+1.0) ': REFERENCE: 1. Salem-300S-000.00S-EOPINT-02, Introduction to the EOP's, Obj. . 9, Pg. 12.

2. Salem 2-EOP-CSPT-1, " Critical Safety Function Status-Tress, Fig's 1-6.

3. KA 000054G011 (3.4/3.3) 000054G011 ..(KA's) ANSWER: 092 (1.00) a. (+1.0) ' REFERENCE: 1. Salem 3003-000.00S-ABCA01-00, Loss of Control Air,.Obj. 3, Pg.

6.

2. Salem S2.OP-AB.CA-0001(Q), " Loss of Control Air", Att. 1, 3, KA 000065A*00 (2.9/3.3) 000065A208- ..(KA's) ANSWER: 093 (1.00) b. (+1.0) REFERENCE: 1, Salem 3 00S-000. 00S-ABCA01-00, Loss 01 Control' Air, Obj. 3, Pg.

6.

'2 . Salem S2.OP-AB.CA-0001(Q), " Loss of Control Air", Reactor trip-requirements.

3. KA 000065A206_(3.6/4.2) , 000065A206 ..(KA's) , n

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

SENIOR REACTOR OPERATOR Pago130 ANSWER: 094 (1.00) Q g(th b. (+1.0) o r- 4.

REFERENCE: 1. Salem 300S-000.008-EOPINT-02, Introduction to the EOP's, Obj.

9, Pg. 12.

2. Dalem 2-EOP-CSFT-1, " Critical Safety Function Statun Tross, Fic's 1-C 1. . 1 3. KA 1940d1A116 (3.1/4.4) j 194001A116 ..(KA's) I i F ANSWER: 095 .(1.00) b. (+1.0) ' REFERENCE: 1. Salem 300S-000.000-ADCN01-00, Main Foodwator/Condeanato System ' Abnormality, Obj. 1, P9 5.

l 2. Salem E2.OP-AD.cN-0001tQ), Main Foodwator/Condensato System

j Abnotwality", Immodlato Actions, i 3. KA 000054G010 (3.2/3.2) 000054G010 ..(KA's) .:

A.NSWE3: 096 (1.00)

i ' c. (+1.0) REFERENCE: 1. Salem 300S-000.00S-AFUEL2-00, Fuel handling Incident, Obj. 3, - Pg 8.

j 2. Salem S2 OP-AD.FUPL-0002 (Q). " Fuel Handling Incident", i Subsequent Actione.

1 3. KA 0000-6C011 (3.5/3.9) 00090CG011 ...(KA's) ! . i

4

'
.

        .__
'
  ,- _ _.L_. ,,m .. . . .a.. , __,...._.,.,.m - - , _ . . _ . . . . _ ;_.a,_;..,_.s..,.. .~._.....;...

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

SENIOR REACTOR OPERATOR Pago131 ANSWER: 097 (1.00) c. (+1.0) REFERENCE: 1. Imlom SD 20, Pt'osuurizer Pressure and Levol control.

2. flalom L/P 300S-000-00CoPZRP/L-00, Prosauritor Pressure Lavol Control, Obj. 11, Pg. 28-29.

3. KA 000028A10$. (3.8/3.9) 000020A101 ..(KA's) ANSWER: 098 (1.00) a. (+1.0) REFERENCE: 14 Salem 300S-000. 00S-LOPA00-01, Loss of AC power and Pocovery Obj. 3 Pg 15 2. Salem 2-EOP-LOPA-1, " Loss of All AC Power and Hocovery" Stop 40.

3. KA 000056K302 (4.4/4.7) 000056K302 ..(KA's) ANSWER 099 (1.00) o. (+1.0) REFERENCE 1. Salem L/P 300S-000.00S-LOCA05-03 Obj 2, Pg 9.- 2. 2-EOP-LOCA-5, ""Losc of Emergency Rocirculation", Shoot 4.

3. KA 000011K312 ~(4.4/4.6) 000011K312 ..(KA's)

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

__ _.

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

ANSWER KEY Pago 1 l ' 001 b 026 d j 002 d __ n n a e Odd / ] 003 d 2 8---- c =* N

/fi
004 a 029 b 005 d 030 a

) 006 b 031 c

007 d 032 c

h 008 b 033 (1 of G- 4/[*'tf7

009 a 034 a 010 c 035- a 011 a 036 d 012 a 037 c 013 b 038 a 014 b 039 b 015 a 040 b. , 016 c 041 b ! 017 c 042 c 018 d 043 d 019 c 044 c 020 d 045 b 021 c A a-c.l1$ih.

t{?h> 0M 022 b 047 c 023 c 4 _ nan __oA delcT/ Il'/f.)

024 a 049 d 026 -d 050 d _ _ _ _ . _ . . . .. _ . . . _ _ _ . . _ _ _ .

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

A 11 S W E R KEY Pago 2 051 d 076 a

             !
            *

052 c 077 d ,- 053 c 078 d 054 b 079 0 055 a 080 d ,

            ,
            '

056 c 081 b 057 e' o a Vrf91 002 b ,

 ,           t 058  d     083 a 059  a     084 c 060  a     085 c 061  d     086 b 062  d     087 a 063  c     -088 b 064  b     089 c     ,

065 b 090 c 066 a 091 c 067 d 092 a

068 a 093 b N hfq) $ 069 b 094 b or R 070 c 095 b 071 b 096 c 072 b 097 a 073 b 098 a 074 c 099 c 075 b ( * * * * * * * * * * END OF 2XAMIllATIO!1 * * * * * * * * * * ) _ _

..
+,m-.'e'qy q -. ..

9- i es.. m 9 % ..msy -. p... m .p , .yp.+y

        -
         ,~g ,y e 7 p%.

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

TEST CROSS REFERENCE Pago 1 Organizod by Quoation Huabcr OUESTION MALUE REFERENCE ; 001 1.00 0000001 002 1.00 8000002 003 1.00 8000003 004 1.00 8000004-005 1.00 8000005 006 1.00 0000006 007 1.00 8000007 008 1.00 8000008.

I 009 1.00 8000009 ' 010 1.00 8000010

011 1.00 0000011-012 1.00 8000012 013 1.00 8000014 014 1.00 8000015 015 1.00 8000016 016 1.00 8000017 017 1.00 0000018 018 1.00 8000019 019 1.00 8000020 ' 020 1.00 8000021 021 1.00 0000022 022 1.00 8000023 023 1.00 8000024 024 1.00 8000025 , 025 1.00 0000026 026 1.00 8000027-02-7 1 rOO - 8000028-*--g t/gd

    .03 8-- 1 r00 5000029 %

029 1.00 8000030 030 1.00 8000031 031 1.00 8000032 , 032 1.00 0000033 033 1.00 8000034 034 1.00 8000035 035 1.00 8000036 036' 1.00 80C0037 037- 1.00 8000038 038 1.00 8000039 039 1.00 8000040 040 1.00 8000041 041 1.00 8000042 042 1.00 8000043 043 -1.00 8000044 044 1.00 8000045 045 3.00 8000046-0 4 6 --- 1. u(T u000047 4 g

047- 1.00 8000048 13 048 1500 8000049"^ ; 049 1.00 8000050 l 050 1.00 8000051

l TEST CROSS REl'ERENCE Pago 2 l organ 1zod by Quootion Hunbor J OUESTION MALUE REFERENCE ' 051 1.00 8000052 052 1.00 8000053 053 1.00 8000054-054 1.00 8000055 055 1.00 8000056 056 1.00 8000057 057 1.00 8000058 058 1.00 8000059 059 1.00 8000060 060 1.00 -8000061 061 1.00 8000062 ; 062 1.00 8000063 t 063 1.00. 8000064 , 064 1.00 8000065 j 065 1.00 8000067 i 066 1.00 8000068 l 067 1.00 8000069 068 1 (4 8000070 069 1.0) 8000071 070 1. .'u 8000072 3

071 1.00 8000073 072 1.00 0000074 t 073 1.00 8000075 074 1.00 8000076 075 1.00 8000077 076 1.00 8000078 077 1.00 0000079 078 1.00 8000080 079' 1.00 0000001 - 000 1.00 '8000082 *

081 1.00 8000083 l 082 1.00 8000004 ' 083 1.00 8000005 084 1.00 8000086 '

    -005  1.00  8000087.-

' 086 1.00 8000088 .

087- 1.00- 8000089 i- 088 -1.00 8000090

, 089- 1.00 8000091 s 090- 1.00 ;8000092

,     091 _1. 00_  8000093    ,

092 1.00 8000094 093- 1.00 8000095 I;

    '094  1.00'  8000096'    '

i. 095 1.00 8000097 096 1.00 0000098 097 1.00 8000099 , 098 8000100

1.00 .

099- 1 00 8000066 49r00 g RbT6b ~ iOf?7

,
      
- - -  - -  _ _ _ _...:.
, i._ , . . . _ , . . . . . . _. , ,. ._ . . ~ . . . ,- ._.;,_.a_.. , . . _ _ . , . . . m_ . . , ,, , - m ... . . . . ., . , , , . .._m, . , , . . . . ,

_ _ . _ _ . ._ _ . . . _ , _ _ _ _ _ . . . . _ . __ _ . . . . TEST CROSS REFERENCE Pago 3

SRO Exam PWR Reactor ' Organizod by KA Group PLANT WIDE GENERICS QUESTION YAINE EA 005 1.00 194001A102 003 1.00 194001A102 002 1.00 194001A103 017 1.00 194001A110 015 1.00 194001A111 I 094 1.00 194001A116 , 006 1.00 194001K101 008 1.00 194001K101 010 1.00 194001K102 4 009 1.00 194001K102 " 016 1.00 194001K103 011 1.00 194001K107 007 1.00 194001K107 - 001 1.00 194001K108 013 1.00 194001K111 014 1.00 194001K116

   ---...

PWG Total 16.00 PLANT SYSTE-;0 Group I OUESTION VALUE Eh 018 1.00 001050A201 004 1.00 003000K407 037 1.00 003000K407 023 1.00 004000K406 024 1.00 004010K403 025 1.00 013000K201 019 1.00 014000A103 038 1.00 014000A202 069 1.00 015000K301 020 1.00 015000K405 021 1.00 015000K407-022 1.00 017020A401

  .-DM.- - 1.--G G G26000R20it d 'M#

031 1.00 061000K402 030 1.00 .061000K406-032 1.00 071000A409 l 033 1.00 071000G006 036 1.00 072000K402

   --- ..

PS-I Total

  -  Jer00'Ai 17, c 4 U'h5 l

l l

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

- -- .. .- .. .- - . . . . f TEST CROSS REFERENCE Page 4 .

SRO Exaa PWR Roaotor organizod by KA Group PLANT SYSTEMS Group II OUESTION VALUE Eh ' 039 1.00 002000G010 040 1.00 002000K107 026 1.00 006000A301 - 042 1.00 010000K403 041 1.00 010000K601 043 1.00 012000A202

044 1.00 012000A306 045 1.00 027000A403 047 1.00 028000A201 _0 4 6.- -1-r00-~128000K503 o % '/?/f> 056 1.00 034000A102 049 1.00 035010K601 052 1.00 062000A201 053 1.00. 062000K403 054 1.00 064 000A4 01-- r

055- 1.00 064000K401 051 1.00 086000A204 _____.

PS-II Total 4-ht0' rbl/7/t)

    (L #

Group III + OUESTION VALUE Eh i 058 1.00 005000A102 ,

     .
  .43 1 -^O- -000000A102-ci'd //y/V 029 1.00 ~008000G010   <

057 1.00 045000G014

   ---_..

PS-III Total 4rOO~ of t/?/4 *.

   -- LJW     ,

______ 4 PS Tota 1 Sv. vu N //</u 3(.09 ; i l I

        .
        !
        ,
     -
        --]
        '
..-n_ . . _ , . , - .  . . , . .

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

i i TEST CROSS REFERENCE Pago 5

SRO Exaa PWR Reactor Organicod by KA Group

EMERGENCY PLANT EVOLUTIONS <

I Group I l

OUESTIOH VALUE Kh l l 061 1.00 000001A205 l 060 1.00 000003G007

062 1.00 000005A203 084 1.00 000011A113 . f EMERGENCY PLANT EVOLUTIONS

l Group I.

' yALQH QUESTION Eh 099 1.00 000011K312 1 035 1.00 000015K104 071 1.00 000026G010 ;

074 1.00 000029A201 '

070 1.00 000029G010 I 073 1.00 000029K312

075 1.00 000040A201 i 076 1.00 000040K304 2 077 1.00 000051A202'

079 1.00 000055A202 078 1.00 000055K302 081 1.00 000057A106 080 1.00~ 000057G010.

012 1.00 000058A203

     '082 1.00 000059A204   -

034 1.00 000059K301-068 1.00 000068A132 067 1.00 000068K318 083 1.00 000074G010-063 1.00 000074K300 072 1.00 000076G004

     ----..

' EPE-I Total 25.00

          ,
          )

9_

._ - _ _ _ __ _ ..~._ _ _ _ _ .. _ _ .___. _._. _    . _._
       -
        . .._._._ _ _ _. . . _ .__
!

! I

f TEST CR038 REFERENCE pago 6 j

l 8RO Exam PWR Roaotor

organized by KA Group i i Group II , WESTION YMdIE M ' 066 1.00 000000A108 . 065 1.00 000009K321 i j 064 1.00 000009K323 086 1.00 000025A102

059 1.00 000025K101 ' 085 1.00 000025K301 087 000027A216

1.00 448 ---l i- 00 000037K9O W 000038A136 M kU , ! 089 1.00 ' 090 1.00 000038K308 I 095 1.00 000054G010 ' 091 1.00 000054G011

088 1.00 000060K202 050 1.00 000061A101

' 093 1.C0 000065A206 092 1.00 000065A208

    ----.

EPE-II Total , 1+r00"M/[7[Y)

     (Ts EMERGENCY PLANT EVOLUTIONS Group III WESTION VALUE M 097 1.00 000028A101 096 1.00 000036G011 098 1.00 000056K302

______ EPE-III Total 3.00

    --- --

EPE Total 4 ~ df-/f7[)

    .....

Test Total N /[*ff] ; 1s w

l l

      --
  ._

w4 y i.,.9 g.-,. ,-m- p : orc w. a p m , g y ,, -,. Sww-.;_.g -.& y y 99 .m. gm.

.. . . = _ - . - - - . ATTACllMENT 2 l'OST-EXAMINATION FACILITY COMMINIS l I l l Pubhc St'f vice ! riedne and Gan I Compny I l stanley L.aBruna Pu nc ser v w Ueuoc and Gaa company P o t u rah. w unes (md p to onors coa 339 vec ! v4er e 9.t.,* % , m ; December 17,1992 NTC-92-3220 Mr. T. Timothy Martin Regional Administrator U. S, Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 1415

Dear Mr. Martin:

WRITTEN EXAMINATION REVIEW COMMENTS - SALEM SENIOR REACTOR OPERATOR EXAMINATION On November 30,1992, and again on December iI,1992, our examination review team had an opportunity to review and comment on the proposed Salem SRO Written Examination. There were numerous comments and recommendations, most of which were incorporated into the examination We appreciate the efTorts of David Silk, Chief Examiner, who was both patient and open niinded concerning our comments and recommendations The pre-examination review continues to be an important pan of conducting these examinations in a manner fair to both the evaluators and candidates The post-examination review with the candidates and other instnictors raised issues that were not identified in the pre-examination review or were not solved by the changes that were incorporated. There are four such questions (Numbers 027,076,087,094) for which we have enclosed comments. The format for the comments is: e Question e Key Answer and Reference Data e Facility Comment e Facility Recommendation e Supporting Documentation (attached) ,M b iD o

- _ Mr. T. Timothy Martin -2- 12/17/92 There were four other questions (Numbers 029,057,082,096) that we are not formally challenging but, because they dealt with information in procedures that would be used in hand, were less than desirable and one question (003) has resulted in a review of the related reference to detennine if the step / action is appropriate or necessary.

If you have any questions, c.omments, or need additional information, please contact A. Orticelle (609) 339 3300, or O. Mecchi(609) 339 3857. They will provide the requested information or will see that you are contacted by the appropriate person.

Sincerely, n ,- ' Attachments (4) l C Mr. Lee 11. Bettenhausen Chief, Operator 1.icensing Branch U.S. NRC, Region i Mr. David Silk ChiefExaminer j U.S. NRC, Region 1

._

SENIOR REACT 0R OPERATOR Pega 32

QUESTION: 027 (1.00)

Given the followings

- Unit 1 is at 100% power.

- PT-948A - Containment Pressure Channel I is failed ! high.

- ALL Bistables for PT-948A have been placed in the tripped condition.

WHICH ONE (1) of the fo$ lowing would occur if PT-948D Containment Pressure detector Channel IV failed high? a. Main Steam Isolation and Containment Spray Actuation ' b. Containment spray Actuation and Phase A c. Main Steam . solation, Safety Injection and Phase A d. Main Steam Isolation, Safety Injection, Containment Spray Actuation, and Phase A ANSWER: 027 (1.00) a. (+1.0) REFERENCE:

1. Salem L/P - 300S-000.00S-CSPRAY-06, ELO 4, Pg. 15-16.

2. Salem Logic Diagram 221057 3. KA 026000A301 (4.3/4.5) 026000K201 ..(KA's) , , Facilitv Comment: There is no correct answer because the channel numbers and bistable numbers are inconsistent ' with the actual system logic. Both of the indicated bistables are fed from Containment Pressure,

Channel IV. In addition. the failure of a containment pressure signal would result in a bypass of ' that input to the Hi Hi Containment Pressure circuit so that the logic changes from 2/4 to 2/3.

The end result of the actions described in the question would be 1/3 signals necessary for an Si signal and Containment Hi Hi Press logic changed to 2/3.

In addition, this question is fundamentally the same as #87 in that they both deal with four channels of protection that utilize all four channels to generate one protection signal and three of those same channels to generate another through numerous bistables. Having two questions on an examination that test the same knowledg: presents a " double jeopardy situation" for the candidate who has not memorized the logic diagrams While our training does require a recall ur derstanding of the number of channels, coincidence, setpoint, etc., it is not intended that candidates memorize specific channel and/or bistable functions unless a failure of that channel would cause an immediate control response. Other ise, they are expected to utilize available references to answer the questions. [ Recommended Action: Delete this question based on there being no correct answer and an unnecessary required recall level of knowledge.

-

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

I

: SENI!R REACTOR CPERATOR      P'g3 81 I

QUESTION: 076 (1.00) Given the following

A double ended ' main steam line break inside containment. has occurred.

- Operators are responding to the event per 2-EOP-LOSC-1,

 " Loss of Secondary Coolant."

WHICil ONE (1) of the following will be available to the operator to confirm the diagnosis of a SGTR in the FAULTED S/G7 a. S/G sample results.

b. R-19 Blowdown radiation monitors, c. R-15 Condonsor Air Ejector radiation monitor, d. Uncontrolled S/G 1evel increase.

ANSWER: 076 (1.00) a. (+1.0) REFERENCE: 1. Salem L/P 300S-000.00S-LOSC01-02, Loss of Secondary Coolant

Obj. 5 Pg 10, 2. Salem 2-EOP-LOSC-1, " Loss of Secondary Coolant", Step 11, 3. KA 000040K304 (4.5/4.7) 000040K304 ..(KA's) Encilitycomment This question was the subject of much discussion during the examination review. The primary means utilized to identify a SGTR is an uncontrolled SG level increase or certain process radiation alarms. Given the coincident LOSC in the question, the process radiation alarms are not a factor because both Phase A and Steamline Isolation have occurred, if a SGTR exists and an uncontrolled SG levelincrease cannot be confirmed at Step 11 - LOSC, then at Step 14 - LOSC (LOCA Evaluation), or as a result of a Continuous Action Summary (CAS), the Containment Radiation Monitors might require a transition to LOCA-1. In both LOSC-1 and LOCA-1 there are steps requiring that a sample be drawn aAer certain related interloch aie cleared. However, that step in both procedures comes after the primary means (uncontrolled les el increase, process radiation) have been evaluated and the interlocks are cleared. Even though the uncontrolled level increase may not be a good indicator of a SGTR for this situation, the tenn "available" implies an on-line function. These factors create a problem for the candidate responding to this question on a written examination because the situation is much more complex than the question. In retrospect, a better question would have been the basis for requiring a SG sample, ktommended ActiOD: Given the technical complexity of the situation presented and the wording of the' question (available), either a. or d. should be considered to be correct.

-

i pago 92 6EllIOR REACTOR OPERATOR l QUESTION: 087 (1.00) Given the following!

- Unit 1 is at 100% power.

- PT-474, Pressurizar pressure channel IV is failed I4W.

- All actions of OP.IV-10.3.1, " Removing, Returning to Service and Loss of a Protective System Channel" are ! complete.

W11IcIl ONE (1) of the following describes the plant response to a failure of PT-455, Pressurizer pressure channel I, LOW 7 a. Reactor trip.

b. Reactor trip AND safety injectiot.. c. All heaters' energize, pressure increases to PORV netpoint, IPR 2 opens and 1PRI does not, pressure cycles around PR2 setpoint.

, d. All heaters energize, pressure increases to PORV setpoint, IPR 1 opens, 1PR2 does not, pressure cycles around PR1 setpoint.

ANSWERt 087 (1.00) a. [+1.0) REFERENCE: 1. Salem 3005-000.005-ICF000-03, I&C FAILURE ANALYSIS, Obj. 20, Pg. 50.

2. Salem Logic Diagram 221055, Reactor Protection System Pressurized Trip Signals.

3. KA 000027A216 (3.6/3.9) 000027A216 ..(KA's) Eacility commen_t: This question is ftmdamentally the same as #27 in that they both deal with four channels of protection that utilize all four channels to generate one protection signal and three of those same channels to generate another through numerous bistables; llaving two questions on an examination that test the same knowledge presents a "doublejeopardy situation" for the candidate who has not memorized the logic diagrams While our training does require a recall understanding of the number of channels, cobcidence, setpoint, etc., it is not intended that candidates memorize specific channel and/or bistable functions unless a falhwe of that channel would cause an immediate control responte. Otherwise they are expected to utilize available references to answer the questions, i.e., reavor protect.sn status panel indication.

, Eccommended Action: Delete this question based on an unnecessary required recall level of knowledge.

e e s> nee zr m m---4-- t v+

         ..-
(SENIOR l   REACTOR OPERATOR       Pcga 99 i

l QUESTION: 094 (1. 00) i WHICH ONE (1) of the following is NOT monitored- by the-Critical f Scfety Function Status Trees to determine loss of containment integrity? ! l a. Containment hydrogen concentration.

b. Containment temperature. j c. Containment sump level.

i d. Containment pressure.

' ANSWER: 094 (1.00)- , b. (+1.0) KEFERD'CE:

      .
        . . . -- -

1. Sales 300S-000.00S-EOPINT-02, Introduction to the'EOP's, Obj.-

9, Pg. 12.

) 2. Sales 2-EOP-{SFT-1, " Critical Safety Function Status Tress, Fig's 1-6.1. . 3. K.% 194001A116 (3.1/4.4) < 194001A116 ..(KA's) L u Facility Comment I A review of the applicable CSFST indicates that both a. and b.'are correct responses. Reference attached.

Recommended Action: Consider both a. and b. as correct responses.-

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

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

A'ITACllMENT 3 NRC POST-EXAMINATION RESOLUTIONS A. NRC Response to Facility Comments Ouesting_Nst Resolution 27 This question was reworded at the request of it.c facility. The most notable change to the question was the addition of channel numbers (1 IV) to the stem. The changes requested were added verbatim. The channel numbers provided by the facility for the bistables were incorrect with the result of there being no correct answer for the question. The question was deleted.

76 This question dealt with the diagnostic ability of the candidate and his famillarity with plant conditions during the response to a faulted-ruptured SG. The facility recommends that answers A and D be accep'ed r.s corr:ct, even though their argument supports the designatext answer. Comment not accepted. A will remain the only acceptable answer.

87 The question is supported by the facility learning objectives. No objections were raised by the facility during the geexamination review.

Comment not accepted. The question will remain in the examination.

94 Because hydrogen is not monitored by the CSFSTs, A will be accepted as a correct response also.

l B. Additional NRC Examination Changes l Ouestion No. Resolution , 28 This question was testing for knowledge of the service water temperature limit. The question was dSeted because of ambiguity caused by the wording "for performing an RCS cooldown at the normal rate." The possible emphasis on the cooldown rate could have been misleading.

. -- . -. . . _ -. Attachment 3 2 33 The NRC will accept C as an additional correct answer because this distractor was not totally incorrect. The wording did not specify the 1500 mrem / year as a maximum exposure limit. 500 mrem / year was the ( acct maximum limit. If the public is lin.ited to less than 500 mrem / year it will also not exceed the 1500 mrem / year as stated in distractor C.

46 This question was deleted, l'ased upon further review of the reference material. To exceed the hydrogen flammability limit several mechanisms of hydrogen production would be needed. The question implied that only one mechanism would produce the necessary hydream' therefore, the question was not valid.

48 This (, ., tion was deleted due to ambiguity in regard to the size of the SO tube leak, if one assumes that the leak is small and the AOP was being implemented then only one choice is correct, if one assumes a nipture occurred and the EOPs were being implemented then two of the distractors would also be correct according to the EOP bases.

Therefore, the question was deleted.

57 Due to a typographic errar in the answer key, the correct answer is A instead of C.

.

__

ATI'ACllMENT 4 SIMULATOR FIDELITY REPORT Facility Licensee: Public Service Electric and Gas Company 244 Chestnut Street Salem, N.J. 08079 Facility Docket Nos.: 50-272 and 50-311 Initial Examination Administered on: December 14 - 17, 1992 , This form is to be used only to report observations. 'I'hese observations do not constitute audit or inspection findings and are not, witho'ut further verincation 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 in response to these observations. During the examination the following items were observed: The NRC ran three scenarios on two different crews on the same day. Running the same scenarios permitted the exam team to observe the consistency of the simulator to replicate malfunctions. In all scenarios, the simulator was consistent and appeared to correctly model RCS and secondary n:sponses expect in one instance. The exception was the absence of core decay heat. _ lipecifically, on a loss of all main and auxiliary feedwater, the CETs were stable and in some instances were slightly decreasing.

Besides the decay heat :nodel ti:e only other deficiency was the unreliability of the plant computer (P-2Ni in the sin:ulator.

Overall, the simulator lxrformed well. During the three days that the .ilmulator was used for scenarios and JPMs, it modeled a variety of conditions and was useful in the licensing decision process.

' t _ _ _ _ ___ _____ _____5 }}

v t e Inspection Report - Salem - 1992021
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 9(0) 21(0) 81(0) Security Operator Exams Emergency Planning Other

IR 05000272/1992021

Text

REGION I

Dear Mr. Martin:

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