IR 05000247/1997002
| ML20217E764 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 10/02/1997 |
| From: | Curley V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9710070190 | |
| Download: ML20217E764 (1) | |
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_ October 2, 1997 * *
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' NOTE T0: NRC Document Control Desk-
, Hail Stop 0-5-D 24 FRON: EcII boalen . Lic nsing Assistant
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ing Licensi{ Branch, R SUBJECT: OPERATOR-LICENSING EXAMINATION ADMINISTERED ON March 17-20. 1997 , AT Indian Point 2 ,
DOCKET #50- 247 (Written Retake Exam)
On March 17-20. 1997 Operator Licensing Examinations were administered at the referenced facility. Attached. you will find the following information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
s Item #1 - a) Facility submitted outline and initial exam submittal, designated for distribution under RIDS Code A070.
b) As given operating examination. designated for distribution under RIDS Code A070.
Item #2 - Examination Report with the as given written examination attached, designated for distribution under RIDS Code IE42.
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April 15, 1997 p Mr. Stephen Vice President - Nuclear Power.
Consolidated Edison Con.peny of g New York, I,nc. .
Indian Point 2 Station Broadway and Bleakley Avenues -
Buchanan, NY 10511
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SUBJECT: INDIAN POINT TWO RETAKE EXAMINATION REPORT NO. 50-247/97 02 (OL)
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Dear Mr. Quinn:
I During the week of March 17,1997, the NRC and your training dr.partment administered ;
retake examinations to two employees of your company who hao applied for licenses to-operate the Indian Point Unit 2 station, but had failed portions <>f their first examination administered in October 1996. These retake examinations were developed by the Indian Point
. Unit 2 training department, utilizing the guidance of NUREG '4021, " Operator Licensing Examiner Standards,". Rev. 7 and Generic Letter 95-06.
I Thes results of the examinations were that both the Senior Reactor Operator (SRO) and the Reactor Operator (RO) candidate passed their exam and were issued licenses.
. Should you have any questions concerning this examination, please contact me at 610-337-5211. :
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Sincerely, t' l ,
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Glenn W. Meyer, Chief Operator Licensing and Human Performance Branch f
Division of Reactor Safety Docket No. 50 247 ,,..
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Enclosure:
Examination Report No. 50-247/97-01 (OL) and w/ Attachments 1 and 2 J N)/
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9704180198 970415 PDR ADOCK 05000247
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Mr. Stephen
REGION I '
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o 475 ALLENDALE ROAD e*4 KING oF PRUsstA, PENNSYLVANIA 19406 1415
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April 15, 1997
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Mr. Stephen Vice President - fluclear Power Consolidated Edison Company of i New York, Inc, lJ 8 Indian Point 2 Station Broadway and Bleakley Avenues
- - Buchanan, NY 10511 SUBJECT: INDIAN POINT TWO RETAKE EXAMINATION REPORT NO. 50 247/97-07 (OL)
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Dear Mr. Quinn:
. During the week of March 17,1997, the NRC and your training department administered
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retake examinations to two employees of your company who had applied for licenses to operate-the Indian Point Unit 2 station, but had failed portions of their first examination administered in October 1996. These retake examinations were developed by the Indian Point
, Unit 2 training department, utilizing the guidance of NUREG 1021, " Operator Licensing Examiner Standards," Rev. 7 and Generic Letter 95 06.
The results of the examinations were that both the Senior Reactor Operator (SRO) and the
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Reactor Operator (RO) candidate passed their exam and were issued licenses.
Should you have any questions concerning this examination, please contact me at 610 337-5211.
Sincerely,
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Glenn W. Meyer, Chief Operator Licensing and Human Performance Branch Division of Reactor Safety Docket No. 50-247
Enclosure:
Examination Report No. 50-247/97-01 (OL) and
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. Mr.-Stephen E Quinn - ' 2~
REGION I==
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Docket No. 50 247 Iteport No: 97 02 (OL)
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Licensee: Consolidated Edison Facility: Indian Point Unit 2 Dates: March 17 20,1997 Examiner: Joseph D' Antonio, Operations Engineer Approved i Glenn Meyer, Chief Operator Licensing and Human Performance Branch Division of Reactor Safety
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{DR ADOCK 05000247 """
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EXECUTIVE SUMMARY l Examination Report 50 247/97 02(OL) g On March 18,1997, one examiner administered an initial retake simulator examination to one Senior Reactor Operator (SRO) candidate. On March 20,1997, the training department administered an Initial retake written examination to one r, actor operator
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p candidate. These examinations had been dt.veloped by the facility e.alning department in j accordance with NHC guidance and approved, e ! ninistered, and graded by the NRC. ;
9attatteA Both candidates passed their examinations. - :
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The facility did a good job in developing the examination.
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In the simulator, the SRO candidate was generally soft spoken to the point where the other crew members would have to walk tsway from their boards to hear what she was saying.
She did better once scenario events led to EOP entry and had no difficulty directing crew actions.
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lhoort Details l
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06 Operator Training and Qualification 05.1 Examination Develooment a.- Scope 8' '
The fa'cility staff developed the written and operating examinations and submitted these proposed examinations for NRC review and approval. The NRC reviewed ths
. proposed examinations, provided comments, and approved the examination for administration. The examinations, as administered, reflected incorporation of the NRC comments, b. Observations and Findinos The NRC had minor comments on approximately twenty percent of the written exam. Half of the comments concerned the substantial number of negatively
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worded questions. This was an !ndication of a generally well prepared exam with some time pressure to meet schedule diadlines. This was understandable since the decision to perform this retake as a facility generated exam and the scheduled date lef t the f acility with approximately 1 month less than the normally allotted time for these exams. Nonetheless, the NRC comments were satisfactorily resolved for the administered exam.
The content of the simulator scenarios was good. The only necersary modification was to start one scenario at a different power level to provide an a:p power rnaneuver as a normal event. However, the f acility will need to considerably expand the level of detail provided in the " operator actions" writeup for the evaluator in future examinations,
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c. Conclusions The facility did a generally acceptable job of developing the examinations.
05.2 Exam Results a. Scope-Initial rctake licensing examinations .were administered to one reactor operator and -
one senior reactor operator applicant. ,,..
. b. Observations and Findinos -
Both candidates passed their examinations.
The items missed by the RO applicant on the written examination did not indicate any one particular area of weakness.
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The SRO caV.:dato in the simulator was generally soft spoken to the point where the other crew members would have to walk away from their boards to hear What she was saying. This trait renders crew briefs an absolute necessity to ensure all crew members are aware of all relevant information. The candidato dia a good job both with crew briefs and in involving crew members in decision making. Once scenario events resulted in EOP entry, she raised her volume somewhat and had no l
, difficulty, directing crew actions. l c. Conclusions Summary Results
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Pass / Fall Pass / Fail Written 1/0 walved Simulator waived 1/0 Walk through waived waived Overall 1/0 1/0
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X1 Exit Meeting The NRC expressed its appreciation for the facility examination development and validation efforts and facility accommodation of the needs of the examination process. Strengths and weaknesses observed in the operating examinations and exam development were discussed.
PARTIAL LIST OF PERSONS CONTACTED J. Ferrick, Manager Operations Training Attachments:
1. Written Examination and Answer Key 2. Simulation Facility Report
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ATTACHMENT 1 WRITTEN EXAMINATIONS AND ANSWER KEYS
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Indian l'olnt Unit 2 Consolidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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Consider the following initial conditions when selecting your answer:
- Reactor Power 100 %
RCS T. 559'F
,p * Control. Rods Automatic / Hank D @ 215 Steps
e Chemistry has advised the control room that #21 CVCS Mixed Bed Demineralizer resin is exhausted. The Reactor Operator is directed to coordinate with the Nuclear
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.. NPO to place #22 CVCS Mixed lied Demineralizer in service.
- #22 CVCS Mixed lied Deminerallrer resin was replaced last week. The demineralizer has semained isolated since resin replacement. No operations have been performed on this demineralizer since that time.
Which of the following statements describes the effect,if any, that placing #22 CVCS Mixed lied Demineralizer in service without saturating at existing RCS boron concentration will have on the following RCS parameters:
- RCS T. ,
- Control Rod Position e RCS Iloton Concentration
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A.- RCS T.,, DECREASE Control Rod l'osition INCRiiASli RCS lloren Concentration INCREASE 11. RCS T,,, NO CllANGli Contiol Rod Position NO Cil ANGE
, , RCS lloron Concentration NO CilANGE C. RCS T.,, INCREASE Control Rod Position DECREASE RCS lloron Concentration -DECREASE .."
D. RCS T _ _ _ DECREASE Control Rod Position DECREASE RCS lloron Concentration INCREASE Page 1 of 100
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i Indian Point Unit 2 Consolidated Fdison Company ofNY
, REACTOR OPERATOR EXAMINATION :
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Which of the following statements conectly describes the status of the Main and Low i Flow Feedwater Regulating Valves after a reactor trip from 100% power and subsequent 7
! five (5) minute cooldown to 530'F due to a stuck open atmospheric steam dump valve? l
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A. The Main Feed Regulating Valves are CLOSED, and the Low Flow Feed
., . Regulating Valves are 75% OPEN (100% power position). !
) i 4 IL The Main Feed Regulating Valves AND the Low Flow Feed Regulating Valves !
are CLOSED. ;
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C. The Main Feed Regulating Valves are full OPEN due to large level enor signal i j caused by " shrink", The Low Flow Feed Regulating Valves are 75% OPEN ,
! (100% power position) ;
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D. The Main 1 ced Regulating Valves aue full OPEN due to large level enor signal
caused by " shrink". The Low Flow Feed Regulating Valves ?te CLOSIID.
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Indian l'olnt Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION
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Consider the following initial conditions when selecting your answer:
e Reactor Power 100 %
- RCS T.., 559'F
,p e Control, Rods Automatic / Bank D @ 215 Steps The narrow range hot leg RTD for Loop 23 fails instantaneously high due to an open circuit. Which of the following statements conectly describes the effect of this failure on the Rod Control System?
A. Rods automatically insert to restore indicated T.,, to T,s. Rod withdrawal is blocked in AUTOMATIC and MANUAL
\ No rod motion occurs. Rod insertion and withdrawal are blocked in AUTO only. M ANUAL od insenion and withdrawal are available.
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No sod motion occurs _ AUTOM ATIC rod insertion and withdrawal are blocked.
MANUAL rod insettlon is available. MANUAL rod withdrawal is blocked.
D. Rods automatically insert to restore T... to T,n. AUTOM ATIC and MANUAL rod insertion and withdrawal are blocked.
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indian Point Unit 2 Consolidated lidison Cornpany of Nl'
REACTOR OPERATOlt EXAMINATION
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Consider the following event when selecting your answer:
- A sinall break LOCA has occuned
- The reactor was tripped froin 100% power
. A Safety, Injection was initiated due to low Pressurizer level i,,
e Emergency Operating Procedure ES 1.2, Post LOCA Cooldown and Depressuritation is being used to reduce safety injection flow.
i Which of the following staternents correctly describes the anticipated response of Pressuriter level and pressure irnmediately after the first Safety injection Pump is stopped?
A. Pressuriier level AND pressure will INCREASE due to voiding in the reactor head.
11. Pressurizer level will DECitEASli, Pressurirer pressuie will INCitEASE as water in the Pressuri/er flashes to steam.
C. Pressmirer level AND pressure will DECREASE due to reduced injection flow.
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D. Pressuriier pressure will DECRi!ASE and Pressurizer level will INCREASE due to voiding in the reactor head.
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Indian Point Unit 3 Consolidated Edison Cornpany ofNY REACTOR OPERATOR EXAMINATION
If a Main Stearn it.olation Valve (MSIV) began to drift closed, which of the following will
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, A. Safety 1 1jection due to liigh Steamline AP
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t 13. Safety injection due to liigh Steam flow from the unaffected Steam Generators C. Turbine Trip from the MSIV 86 relay D. Reactor Trip due to level shrink on the affected Steam Generator
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Indian Point linit 2 Consolidated Edison Compnny of NY Ri? ACTOR OPERATOR EXAMINATION
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Consider the following event when selecting your answer:
With the plant at 100% Reactor Power, control rod 118 (center of core), drops approximately 100 steps (62 inches) from the bank position of 215 steps. Which of the following statements correctly describes the effect, if any, that this event will have on axial l
'# core power ilistribution (61)7 i Assume that a turbine sunback D_QliSElT occur.
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A. A i will become more negative due to reduced power generation in the top of the i core and increased power generation in the bottom of the core.
A i will become less negative due to increased power generation at the top of the and reduced power generation at the bottom of the core.
C. A 1 will not ch:.nge since rod ll-8 is at the center of the coie and will affect all quadrants equally.
D. A 1 will initially become more negative then return to its original value when positive reactivity from the power coefficient returns reactor power to its original value.
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i indian l'olnt Unit 2 Consolidated lidison Cornpany ofNl'
j REACTOR OPERATOR EXAMINATION
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During operation at 100% reactor power,6.9 KV Bus I nonnal supply breaker (UT-1)
trips open due to a relay failure and a reactor trip occurs. From the choices below, select the protection signal which initiated the reactor trip?
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% Twolo6plossof flow B 6.9 KV bus undervoltage
. Reactor Coolant Pump under frequency.
K. Single hx)p loss of flow
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Indian Point Unit 2
. Consolidated lidison Company ofNY MEACTOR OPERATOR EXAMINATION
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Consider the following event when selecting your answer:
. Following a reactor trip from 100% power due to a loss of both Main Boller Feed Pumps, l the Motor Driven Auxiliary Feedwater Pumps are being used to control Steam Generator
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(SG) Levels., Shortly after the AFW system is placed in service the piping downstream of the Auxiliary I' ecd Regulating valve to #24 SG suptures. Which of the following statements correctly describes the automatic response of the AIAV system to this failure?
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The #24 SG auxiliary feed tegulating valve will automatically close. Feed flow to
- 21, #22 and #23 SG will not be affected.
IL The #22 SG and #24 SG auxiliary feed regulating valves will automatically modulate to prevent pump runout and maintain sufficient AIAV discharge pressure to maintain auxiliary feed flow to #22 SG. Feed fiow to #21 and #23 SG will not be affected, hThe #23 SG and #24 SG auxiliary feed regulating valves will automatically modulate to prevent pump runout and maintain sufficient AIAV discharge pressure to maintain auxiliary feed flow to #23 SG. Feed flow to #21 and #22 SG will not be affected.
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Since #23 and #24 SGs are supplied by the same AISV pump, feed flow to lx>th SGs will be completely lost. Feed flow to #21 and #22 SG will not be affected.
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inclian l'oint Unit 2 Consolidated Edison Campany of NY REACTOR OPERATOR EXAMINATION
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Due to failure of a pre amplifier in the circuit for NIS Power Range Channel N42, the lower detector output has failed to zero. Total indicated power from NIS Power Range Channel N42 indicates 51% with the Reactor at 100% power. In order to repair the pre-amplifier the channel must be ternoved from service by removing the instrument power fuses.
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I Which of the following statements conectly states the effect that this operation will have on the Power Range Nuclear instrumentation System and Reactor Protection System?
The high Dux bistables for N42 will be inhibited from tripping. A minimum of two of the remaining three power ranpc channels must sense a high flux condition to trip the reactor.
h The high Hux bistables associated with N42 will tiip. A minimum of one of the three remaining power range channels must sense a high flux condition to trip the reactor.
C. The high flux bistables associated with N42 will trip. A minimum of two of the three remaining power range channels must sense a high Dux condition to trip $e reactor.
. The high Hux bistables for N42 will be inhibited from tripping, A minimum of one of the three remaining power range channels must sense a high Hux condition to trip the reactor.
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indian l'oint Unit 2 Consolidated ihlison Cornpany of Nl' l REACTOR OPEllATOlt EXAMINATION Which of the following conectly identifies the instruments that provide the pressure signal for the detennination of the sulcooling value displayed on Flight Panel (FD)? l
,, k. Pressuri f er Pressure (Channels I and II) 1 k. RCS OPS System Pressure Instruments (PT-413/433/443)
Pressurizer Pressure (Channels til and IV)
D. RCS Wide Range Pressure Instmments (PT-402/403)
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ludian l'olnt Unit 3
- Consolidated Edison Company ofNY RFACTOR OPERATOR EXAMINATION
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Consider the following indications when selecting your answer:
- Reactor Power 100 %
e Tm 559'F e Control llank D 215 steps / automatic lN I
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Contain0ient Area Radiation Monitor (R2) Ahtnning l * Containment Atmosphere Radiation Monitors (R4 f42) Normal ,
l * Containment Area Radiation Monitor (R7) Alanning
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- Charging Pump Cell Area Radiation Monitor (174) Alarming-
- Sampic Cell Area Radiation Monitor (R6) Alarming
Plant Vent Radiation Monitors (R43/R44) Nonnal From the list below select the event which would explain the indications described above?
Reactor core fuel element failure 1. RCS leak in containment
. RCS/CVCS leak in charging cell 1. RCS/CVCS leak in sample cell
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indian l'oint Unit 2 Consolidated Edison Cornpany ofNl'
ItEACTOlt OPEllATOlt EXAMINATION
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While discharging #1_3 Waste Distillate Storage Tank (WDST) to the river using #14 Waste DistillMcJunsff.LPhnilfiWDTP)ian aTaniT0ccurs on Panel SAF-1,"1154 LIQUID WASTE DISTILLATL411 RAD /rROUBLE" due to a high radiation signal, Which of the following correctly identifies the anticipated response of the following Liquid 8I Itadwaste S stem components:
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. #13 and 14 WDTPs
. #13 and 14 WDTP Discharge Valves (SOV CT 965 MCV and SOV CT 982 MCV)
. Conunon WDTP Discharge Valve (SOV CF-971 FCV)
A #13 WDTP ltUNNING
- 14 WDTP STOPPED
- 13 WDTP Discharge Valve (SOV CT-965-MCV) CLOSED
- 14 WDTP Discharge Valve (SOV CT 982-MCV) CLOSED Conunon WDTP Discharge Valve (SOV CT-971 FCV) OPEN i
11. #13 WDTP STOPPED
- 14 WDTP STOPPED
- 13 WDTP Discharge Valve (SOV CT-965-MCV) CLOSED
- 14 WDTP Discharge Valve (SOV CT-982-MCV) CLOSED Conunon WDTP Discharge Valve (SOV CF-971-FCV) OPEN C. #13 WDTP ltUNNING
- 14 WDTP STOPPED
- 13 WDTP Discharge valve (SOY Cr-965-MCV) OPEN
- 14 WDTP Dischaige Valve (SOV CT-9S2-MCV) CLOSED Conunon WDTP Discharge Valve (SOV CT-971-FCV) CLOSED
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D. #13 WDTP STOPPED
- 14 WDTP STOPPED
- 13 WDTP Discharge Valve (SOY CT-965-MCV) CLOSED
- 14 WDTP Discharge Valve (SOV CT-982-MCV) CLOSED Conunon WDTP Discharge Valve (SOV CT-971-FCV) CLOSED "
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Indian l'olns Unit 2 Consolidated Edison Company ofNY.
REACTOR OPERATOR EXAMINATION
'the Senior Reactor Operator has directed you to init' ate a Containment Building Pressure Relief. Approximately 15 minutes after the release has been initiated the S.1338, Meteorological Data Display on the Accident Assessment Panel stops functioning. Which of the following actions should you take to compensate for this failure?
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Immediately stop the release and have no further releases until the display is repaired.
Record the Plant Vent Radiation Monitor (R44) reading every hour until d.c release is terminated.
Verify that meteorological data is available and record meteorological data every hour throughout the remainder of the release, k Stop the release and prepare a new release permit using the most adverse meteorological conditions.
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i Indian Point Unit 2
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Consolidated Edison Company ofNY
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REACTOR OPERATOR EXAMINATION
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During power operation the CVCS Automatic VCT Makeup System initiates blended makeup due to the VCr level reaching the low level setpoint. An air kak on the supply
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line to Boric Acid Flow Control Valve, FCV 110A, causes air pressure to the valve diaphragm to decrease to O PSIO and prevents the valve from responding to the flight
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panel controller (FIC-Il0A). The air leak is small enough that it does not have any
8 significant effect on instrument air header pressure.
Which of the following statements conectly describes the efrect,if any, that this failure
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will have on the boric acid concentration of the Reactor Coolant System?
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. RCS lloric Acid Concentration will DECREASE because FCV 110A fails CLOSED on a loss of air pressure, causing blended makeup to have a lower than desired boric acid concentiation.
I RCS lloric Acid Concentration will INCREASE because FCV 110A fails OPEN on a loss cf air pressure, causing blended makeup to have a higher than deslied boric acid concentration.
1 RCS lloric Acid Concentration will NOT CilANGE because FCV 110A is only used when the Makeup Selector Switch is in the llORATE position.
. RCS I)oric Acid Concentration will NOT CII ANGE because FCV Il0A is only used ;
when the Makeup Selector Switch is in the M ANUAL position.
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indian l'olns Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION
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Consider tim following conditions when selecting your answer:
- A small break LOCA has occurred
. RCS pressure is at 1900 PSIG p * The Safety injection System has automatically actuated due to liigh Containment
Pressure'
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. ALL ESF equipment has responded as designed
- RCS break flow is approximately 800 GPM e No Charging Pumps are operating Select the response which correctly describes the status of the following ESF equipment, and RCS parameters which will exist whcDJJutbluqndiliQns are reached in the Reactor Coolant System? Assume that a cooldown ll AS NOT been initiated.
. S1 Pump Status
. RilR Pump Status
- L RCS Pressure
. - RCS Ilicak Flow 4
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h. St Pump Status- Running - Recirc to RWST RilR Pump Status Running - Flow to RCS RCS Presstne LESS Til AN 1500 PSIG and GRENFER Til AN 150 PSIG RCS lireak Flow LESS TilAN 800 GPM h St PumpStatus Running - Flow to RCS RilR Pump Status Running - On theirculation RCS Pressure LESS TilAN 1500 PSIG and GREATER Til AN 150 PSIG RCS litcak Flow LESS TilAN 800 GPM K S1 Pump Status Running Recirc to RWST RilR Pump Status Running Flow to RCS RCS Pressure LESS TilAN 150 PSIG RCS lireak Flow GREATER TilAN 800 GPM
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y. S1 Pump Status Running - Flow to RCS RilR Pump Status Running - Flow to RCS RCS Pressure LESS TilAN 150 PSIG -
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indian l'oint thuit 2 Consolidated ihlison Company ofNl'
REACTOR OPl?RATOR EXAMINATION
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Consider the following initial conditions when selecting yout answer:
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- Reactor Power i X 10IR amps l
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- Tm 547'F
, * Plant strutup in progress During a plant startup, af ter the Main Steam isolation Valves have been opened and Main Condenser Vacuum has been established, a fault is detected on the Station Auxiliary Transformer causing a loss of 138 KV electrical power to the station. Assuming no operator action, which of the following describes the inunediate impact that this failure will have on the operation of the following:
- Condensate Pumps
- Condenser vacuum
. Reactor
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Condensate pumps Stopped Condenser vacuum Stable Reactor Critical Condensate pumps Running Condenser vacuum Decreasing Reactor Tripped C. Condensate pumps Stopped Condenser vacuum Deetcasing Reactor Tripped 1. Condensate pumps Running Condenser vacuum Stable Reactor Critical
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indian l'oint their 2 Consolidated lidison Company of Nl'
ItEACTOlt OFFitATOlt EXAMINATION
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During power operation a catastrophic failure of the service water supply line to the #23 Containment Fan Cooler Unit results in Hooding of the containment building including the Hlling of the reactor cavity sump to the 46 ft. elevation of the containment building.110w and why could this failure isnpact the ability of the Emergency Safeguards System (liSF)
, to perform its design function if a design bases large brcak LOCA were to occur?
Assume all other ESF equipment operates as designed when the accident occurs.
.. Containment pressure may exceed design limits duc 'o loss of cooling water to the Containment Fan Cooler Units which are required to meet the minimum safeguards equipment Icquiten.ents for containinent pressure control.
Accumulator isolation valves will be submerged and may fail to open when signaled sesulting in failure of the accumulators to inject into the itCS.
' Cold leg recisculation will be impossible since the recirculation pumps will be submerged preventing the establishment of any recirculation path.
D. Containment pressure and water level may exceed design limits due to reduction in contalmnent free volume before the accident occuned.
Page 17 of 100
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indian Point Unit 2 Conwlidated Ediwn Company ofNY REACTOR OPERATOR EXAMINATION
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Consider the following initial conditions when selecting your answer:
- Reactor Power 100 %
- T. .,
. 559'F
- Control tods Bank D @ 215 Steps /Autornatic
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During power operation with the above conditions Power Range Channel N41 fails to inaxiinutn detector output (120%). Which of the foll~"ing staternents correctly describes the effect that this failure will have en the Rod Control System?
l A. Contiol rods will autornatically insert to reduce indicated reactor power to equal tuihine power.
l B. Control rods will automatically insert until the rate of change of indicated reactor power veisus turbine power decays to 7ero.
C. Contiol rod inotion will be inhibited due to the overpower rod stop.
Control tods will automatically insert due to a turbine sunback initiated by the change in indicated reactor power exceeding 5% in 5 seconds.
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Page 18 of 1(X)
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Indian l'alnt Unit 2 Consolidated Rdison Company ofNY REACTOR OPERATOR EXAMINATION While releasing #24 Large Oas Decay Tank (LGDT) the Nuclear NPO inadvertently open's RCV 014 (LGDT Release Hand Control Valve) to the full open position causing a the R-43/R 44 PLANT VENT HI RAD / TROUBLE alann to actuate on panel SAF 1 in the ,
control room, due to R-44, Plant Vent Gas Monitor, exceeding the High Radiation Gas J alarm setpoint, A,
Which of the following statements de=cribes the impact,if any, that this event will have on l the following equipment: .
- RCV-014
- Pall Supply Fan ;
- Pall Exhaust Fan
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h RCV 014 _ Closes No change Pall Supply Fan !
i PAB Exhaust Fan No Change B.RCV-Ol4 Closes
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Pall Supply Fan Trips Pall Exhaust Fan No Change
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C. RCV-014 No change l'All Supply Fan Trips PAB lixhaust I an Trips D. RCV-014 Closes Pall Supply 1 an No Change Pall lixhaust Fan Trips
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Indian Point Unit 2 Consolidated Edison Company o,fNl'
l REACTOR OPERATOR EXAMINATION I During power operation the APPROACillNG ROD INSERTION UMIT 12.5" and the ROD INSERTION LIMIT 0" alarms are actuated on Panel SAF in the control room. l
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Investigation reveals that all control and shutdown rods are fully withdrawn and that the rod insertion limit is not being violated, indicating that the alanns have been actuated due
, to a malfunction.
Which of the following components,if malfunctioning, could cause the alarms to actuate?
A. Individual Rod Position Indicator voltage decreases from 3.45 VDC to 0 VDC.
13. Rod 1101 tom Rod Stop flistable C Pulse to Analog Converter (P/A Converter)
Rod llottom Rod Stop liypass 131 stable
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Page 20 oi 100
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indian l'olnt Unit 2 Consolidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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Consider the following initial conditions when selecting your answer:
. Reactor Power 60 %
- T. 554*F
, * Control rods Bank D @ 200 Steps /Autornatic During surveillance testing it is discovered that the #22 Reactor Coolant Pump is not performing at design conditions. The flow rate in loop #_22 is 7% less than design due to degradation of the diffuser ring hi the pump cming. Flow in the remaining loops are at or above design and total core flow is reduced by 5%. Which of the following statements describes how this condition will affect the following:
. Core A T
- Departure From Nucleate lloiting Ratio (DNilR)
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A. Core 6 - T lower than design DNilR Lower than design 11. Core A - T lower than design DNilR liigher than design C. Core A - T liigher than design DNDR liigher than design D. Coic A - T liigher than design DNilR 1.ower than design r
1 0 AL ^
~2 'J-008 R c. .-
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Page 21 of 100
indian Point Unit 2 Consolidated Ediscn Company ofNY REACTOR OPERATOR EXAMINATION Which of the following conectly identifies the signals that are used to pmvide input to tiie MBFP Speed Control System for determining the following parameters:
- Actual Feed Regulating Valve Differential Pressure (AP)
- Power 8' * % Stattup Signal
. AP Main Feed 11eader Pressure - Main Steam lleader Pressure
. Power Jurbine First Stage Pressure (PT-412A)
% Startup Signal MBFP Governor Control Switch 13. A P Main Feed lleader Pressure - Main Steam 11eader Pressure
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Power Total Stm Flow (sum of all controlling steam flow channels)
% Stanup Signal MilFP Governor Control Switch
% AP Main Feed licader Pressure - Main Steam lleader Pressure Power Total Stm Flow (sum of all controlling steam flow channels)
% Startup Signal MilFP Foxboro Speed Controller Manual Setting
. AP Main Feed licader Pressure - Steam Generator Picssure
- Power Total Stm Flow (sum ot all controlling steam flow channels)
% Stanup Signal MilFP Governor Control Switch
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Page 22 er 100
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i indian l'olnt Unit 2 Consolidated Edison Company ofN}'
REACTOR OPERATOR EXAMINATION
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While retrieving a dropped control in Control Bank D Group 1 with the unit at 80% i power, a ROD CONTROL URGENT FAILURE alana is received on panel S11F 1 in the l control room. Which of the following statements conectly describes the cause of this alarm?
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h. Remaining rods in Control Bank D Group 1 are not moving when demanded. .
l k.' Alann is caused by the affected rod position deviating by more than 12 steps from bank demand position. - ,
. Alann is caused by control bank rods moving with no bank overlap.
h All rods in Control llank D Group 2 are not moving when demanded
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Page 23 of 100
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Indian l'oint ihoit 2 Consolidattvl Edison Cornpany ofNl'
REACTOR OPERATOR EXAMINATION
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Consider the following event when selecting your answer:
. Reactor Power 0%
+ T.., 530*F
- Total Steam Flow 0.75 X 10 1bm/hr 0, * Pressurizer Level 32 %
- Pressurizer Pressure 2235 PSIG The plant is at hot shutdown due to a tube leak in #21 Steam Generator (SG). The Main Steam Isolation Valve for #21 SG is closed and the Reactor Operator is perfonning a cooldown to Cold Shutdown using the condenser steam dumps.
While positioning the condenser steam dump valves, a controller failure causes ALL of the steam dumps to OPEN fully. A Safety injection Signal is received. Directly afterwards the Reactor Operator notes the following parameters:
. Reactor Pawer 0%
= T.,, 523 F j . Total Steam Flow 0 lbm/hr
- Pressurizer Level 25 %
- Pressurizer Pressure 2110 PSIG Which cf the following liSF actuation signals initiated the Safety injection signal?
%. Iligh Containment Pressure IL liigh Steam Flow ceincident with Low Tm K Low Pressurizer Pressute D. Main Steam Lme AP (21 SG GREATEP. Til AN 22,23,24 SG)
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indian l'oint Unit 2 Consolidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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Select the answer which correctly identifies the automatic reactor trip signals that are blocked when the POWER DELOW P-7 permissive is enabled?
8, \ A. Pressurirtr Low' Pressure Reactor Trip
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Picssurizer liigh Level Reactor Trip Two Loop Loss of Flow 6.9 KV Bus Unden'oltage
. Pressurizer Low Pressure Reactor Trip Pressurizer liigh level Reactor Trip Steam Generator lo : o Level Trip 6.9 KV llus Undervoltage (\ Pressurizer Low Pressure Reactor Trip Pressurizer Low Level Reactor Trip T3vo Loop Loss of Flow 6.9 KV 13us Undervoltage l
Pressurizer liigh Pressure Reactor Trip Pressurizer liigh Level Reactor Trip
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Steam Generator Lo-Lo Level Trip I
6.9 KV llus Undervoltage Page 25 of 100
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indian l'oint Unit 2 Conwlidated Edison Company of Nl'
RFACTOR OPERATOR EXAMINATION e
While performing a natural circulation cooldown using ES-0.2, Natural Circulation Cooldown, you are directed to depressurize the Reactor Coolant System to 1890 PSIG after verifying that the RCS Ilot leg temperatures are less than 550'F, Which of the following statements correctly describes the reason for the maximum limit on hot leg ,
$
, temperature Ixfore depressurization can commence?
. Ensua that the AT limit between auxiliary spray fluid temperature and the RCS is not violated.
Ensure that wide range hot leg temperatures are aforoximately saturation temperature for SG pressure.
Ensure that RCS subcooling is almve the RCP Termination Criteria for the E-0 series of procedures.
D. Ensure that adequate subcooling exists to prevent void formation in the reactor head when pressure is reduced to 1890 PSIG.
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Page 26 of 100
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Irullan l'olnt Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION
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During power operation an increase in RCS leakage is noted during a routine RCS leak rate surveillance test. The subsequent RCS leakage safety evaluation detemunes that Reactor Coolant Drain Tank in leakage has increased by the same amount that RCS leakage has increased. Which of the following leakage sources is the most probable cause?
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K. CVC'S Letdown Line Relief valve leakage IL Reactor Vessel Flange 0 Ring (. Pressurizer Power Operated Relief Valve (PORV) Leakage k. Reactor Coolant Pump #3 Seal Leakage l
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Page 27 of 100 j
indian l'oint ilnit 2 Consolidated Edison Company of Nl'
REACTOR OPERATOR EXAMINATION
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During power operation a Pressurizer insurge results from a step decrease in power from'
100% to 90%. As Reactor Operator you observe both Pressurizer spray valves throttling open, and all Backup heaters energized simultaneously. As Reactor Operator you should:
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8' A. Immedistely TRIP the Backup heaters B. Immediately CLOSE the Spray valves C. Ensure lleaters and Spray continue to operate as designed D. Take the appropriate action for the controlling Pressurizer Pressure channel failing
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Indian Point Unit 2 Consolidated lidison Company ofNY
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REACTOR OPERATOR EXAMINATION
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Which one of the following signals are used to automatically ARM the RCS .
Overpressurization System (OPS) during a plant cooldown to Cold Shutdown?
A. Two out,of three (2/3) RCS Wide Range Cold leg Temperature RTDs LESS TilAN 8, setpoint II. Two out of three (2/3) RCS Wide Range llot Leg Temperature RTDs LESS THAN ,
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setpoint
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Two out of three (2/3) RCS Wide Cold Leg Pressures GREATER THAN setpoint D. Two out of three (2/3) RCS Wide Cold Leg Pressures Li!SS Tit AN setpoint
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Page 29 of 100
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Indian Point Unit 2 Consolidated Edison Cornpany of NY ,
REACTOR OPERATOR EXAMINATION While performing Emergency Operating Procedure FR 11 1 Loss of Secondary lleat Sink',
following a Loss of Coolant Accident (LOCA), you are directed to discontinue use of the piocedure if RCS Pressure is LESS TilAN non faulted Steam Generator (SG) pressure (s).
Which of the following statements conectly describes why it is not necessary to continue with heat sink restoration if RCS Pressure is LESS T11AN non-faulted SG pressure (s)?
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FR 11 1, will initiate RCS Bleed and Feed. A bleed path is not necessary if RCS pressure is low.
. Since RCS pressure is less than non faulted SG pressure, RiiR flow may be established to remove core decay heet.
C. Secondary heat sink is not icquired in this condition since core decay heat is removed by break flow.
9, Since RCS pressure is less than non-faulted SG pressure, an inadequate Core Cooling condition probably exists, which is a higher priority procedure.
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Page 30 of 100
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- indian l'oint Unit 2 l Consolidated Edison Company ofNY f
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REACTOR OPERATOR EXAMINATION
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During power operation a Main Steam Line Hreak occurs inside containment on the steani line from #24 Steam Generator (SG). A Safety injection signal is received from the Main Steam Line AP circuit. During recovery operations the control room operators are unable to close the Main Steam Line Isolation Valve for #24 SG.
Which of the following statements is conect regarding the ability of the operators to control Reactor Coolant System (RCS) temperature during this event?
\. RCS temperature will continue to decrease until ALL SGs have dried out. Subsequent temperature control will be perfonned by limiting Auxiliary Feedwater flow.
11. RCS temperature will continue to decrease until 24 SG has dried out. Subsequent temperature control will be perfonned using the remaining SGs, and auxiliary feed water flow.
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if RCS temperature will continue to decrease until 24 SG has dried out. Subsequent I
temperature control will be performed using the Atmospheric Steam Dump Valve (s)
on the intact SGs since an automatic Main Steam Line Isolation was actuated by the Main Steam Line AP Safety injection signal.
RCS temperature will continue to decrease until ALL SGs have dried out. Subsequent temperature control will'tx performed by using RCS Bleed and Feed.
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Page 31 of 100
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Indian Point l] nit 2 Consolidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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Which of the following statements correctly describes the purpose of the POWER ABOVE P-10 pennissive?
\. Automati,cally blocks the Source Range liigh Flux and Intermediate Range liigh Flux 8, reactor trips. Prevents re-instatement of the Source Range instruments.
. Allow operator to manually block the Power Range liigh Flux Low Setpoint and Source Range liigh Flux reactor trips. Prevents re instatement of the Source Range instruments.
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Automatically blocks the Power Range liigh Flux Low Setpoint and Intermediate
' Range liigh Flux reactor trips. Prevents re-instatemera of the Source Range instruments.
D. Allow operator to manually block the Power Range liigh Flux Low Setpoint and Intermediate Range liigh Flux reactor trips. Prevents re-instatement of the Source Range instruments.
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Page 32 of 100
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Indian Point Unit 2 Consolidated li'dison Company ofNY-REACTOR OPERATOR EXAMINATION During power operation Channel 1 Pressurizer Pressure Instrument fails hich to maximuih'
output (100%). While performing the subsequent actions of the Abnormal Operating
- Instruction (AOI) the Reactor Operator is directed a trip the following bistables:
- -* Pressurizerliigh Pressure ReactorTrip E.T d * Pressuriict Low Pressure Reactor Trip :Itt
- Pressurizer Low Pressure Safety injection IL L -
- Pressurizer law Pressure SI Unblock Which of the following correctly states the expected status (illuminated / extinguished) of the associated bistable proving lamp as each bistable trip switch is placed in the trip l-position?
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l A. Pressurizer liigh Pressure Reactor Trip Extingui,shed l
Pressurizer Low Pressure Reactor Trip Extingtiished Pressurizer Low Pressure Safety Injection Extifiguished Pressurizer Low Pressure SI Unblock ,11,luminated
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h Pressurizer liigh Pressure Reactor Trip Extinguished Pressurizer Low Pressure Reactor Trip Illuminated Pressurizer Low Pressure Safety Injection Illuminated Pressurizer Low Pressure SI Unbh>ck Extinguished
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C. Pressurieer liigh Pressure Reactor Trip , Illuminated Pressurizer Low Pressure Reactor Trip ".Extinguisheh Pressurizer Low Pressure Safety injection ! Extinguishcll Pressurizer Low Pressure SI Unblock illuminateff D. Pressurizer liigh Pressure Reactor Trip Extinguished Pressurizer Low Pressure Reactor Trip llluminated Pressurizer Low Pressure Safety injection Extinguished Pressurizer Low Pressure SI Unblock Extinguished
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Page 33 of 100
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Indian l'oint linit 2
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Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION
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Consider the following event when selecting your answer:
While the plant was at 100% Reactor Power a loss of the 138 KV offsite power source occurred, followed by a Turbine Trip on low vacuum and subsequent reactor trip. The Reactor Operator is performing Step 1 of E-0, Reactor Trip or Safety Injection.
Which of the following statements correctly lists the indications that the Reactor Operator will use to verify that the Reactor trip has occurred?
Rod Ilottom Lights -
Reactor Trip Breaker Position B. Neutron Flux Individual Rod Position Indication C. Reactor Trip Breaker Position Neutron Flux Reactor Trip Breaker Position
- Bank Step Counters l
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Page 34 of 100
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Indian Point Unit 2 Consolidated Edison Company of NY REACTOR OPERATOR EXAMINATION
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While performing actions directed by the Emergency Operating Procedures, the Reactor Operator resets the Containment Spray signal. After depressing the reset push-buttons, the Reactor notes that the white indicating lights above the buttons illuminate and remain illuminated after the buttons are released.
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Which of tiie following statements correctly describes the reason the lights illuminated when the reset push-buttons were depressed and remained illuminated when the buttons were released?
' A. The lights illuminated when the buttons were depressed to indicate that the spray signal was reset. The lights remyned illuminated indicating that an automatic Containment Spray actuation signd was pre.sent.
\. The lights illuminated when the buttons were depressed to indicate that the spray signal was reset. The lights remained illuminated indicating that both containment spray pumps were running.
C. The lights illuminated when the buttons were depressed to indicate that the spray signal could NOT be reset. The lights remained illuminated indicating that an automatic Containment Spray actuation signal was present.
. The lights illuminated when the buttons were depressed to indicate that the spray signal was reset. The lights remained illuminated indicating that both containment spray pumps were NOT running.
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l indian l'oint Unit 2 Consolidated Edison Cornpany ofNY REACTOR OPERATOR EXAMINATION
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Which of the following correctly identifies the Process Radiation Monitors which are capable of AUTOMATICALLY terminating a Containment Purge if they sense a high radiation condition?
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'8! A. R-44, Pfarit Vent Gas Monitor
' R-43, Plant Vent Particulate Monitor R-42, Containment Gas Monitor -
B. R-43 Plant Vent Particulate Monitor R-41, Containment Particulate Monitor R-42, Containment Gas Monitor C. R-43, Plant Vent Particulate Monitor R-44, Plant Vent Gas Monitor R-41, Containment Particulate Monitor h R-44, Plant Vent Gas Monitor R-41, Containment Particulate Monitor R-42, Containment Gas Monitor
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Indian Point Unit 2 Consolidated I?dison Company of Nl'
REACTOR OPERATOR EXAMINATION
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While performing a plant cooldown using the Condenser Steam Dumps, a 6.9 KV MOTOR TRIP (COMMON) alarm is actuated. The Reactor Operator notes a step decrease in steam flow at the same time the alarm occurs. No other annunciators are actuated.
d - Which one of the following 6.9 KV motors is the most likely cause of the alarm?
h Reactor Coolant Pump B. Circulating Water Pump-
, ' lleater Drain Pump
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t. Condensate Pump i
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Page 37 of 100
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Indian l'oint Unit 2 Consolidated lidison Company ofNY REACTOR OPERATOR EXAMINATION Which of the following lists correctly identifies the plant equipment protected by concentrated foam fire suppression systems?
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\. Ilydroy il Seal Oil Unit 8,
Main an'd Auxiliary Transformer Oils Systems Clean and Dirty Oil Storage Tanks Turbine Oil Reservoir h liydrogen Seal Oil Unit Boiler Feed Pump Console Clean and Dirty Oil Storage Tanks-Turbine Oil Reservoir (, llydrogen Seal Oil Unit iloiler Feed Pump Console
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Support Facility Igmtion Oil Tanks
! Turbine Oil Reservoir L Main Generator lloiter Feed Pump Console Clean and Dirty Oil Storage Tanks Turbine Oil Reservoir Page 38 of 100
indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION
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Due a fire in the transformer yard, fire main pre [suNis decreasing because of heavy demand. Which of e.: fe!! swing senectiens correctly identifies the sequence in which the standby fire ramps will automatically stan to maintain fire header pressure?
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- 8' A. Fire Main Booster Pumps Fire Diesel Pump Standby Pressure Maintenance Pump B. Fire Diesel Pump Standby Pressure Maintenance Pump Fire Main Booster Pumps C. Fire Main Booster Pumps
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Standby Pressure Maintenance Pump
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Fire Diesel Pump Standby Pressure Maintenance Pump Fire Main Booster Pump Fire Diesel Pump
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Page 39 of 100 J
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Indian Point Unit 2 Consolidated Edison Cornpany ofNY
. REACTOR OPERATOR EXAMINATION Following a unit trip from 50% power, a station blyackout .cu due to a fault on the 138'
KV offsite power supply. Which one of the following selections correctly identifies the E480 VAC equip,mment that will automatically start WHEN the 480 VAC safeguards busses
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breakers are re-energized? ~
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d NOTE:NO S1 SIGNA 14s) EXIST l
. Two Non Essential Service Water Pumps Three Essential Service Water Pumps
, 21 and 23 Motor Driven Auxiliary Feed Water Pumps
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. Two Essential Serv;ce Water Pumps -
21 AND 23 Auxiliary Feed Water Pumps 21,22,- AND 23 Component Cooling Water Pumps
[C Three Essential Service Water Pumps 21 AND 23 Auxiliary Feed Water Pumps 21,22, AND 23 Component Cooling Water Pumps D. Three Essential Service _ Water Ptmps 21 AND 23 Auxiliary Feed Water Pumps 21 AND 23 Component Cooling Water Pumps
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indian l'olnt Unit 2 Consolidated Edison Company ofNY
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REACTOR OPERATOR EXAMINATION.
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When loading electrical equipment on the 480 VAC busses while recovering from a unit trip and station blackout you are directed to limit the load on Transformers 5,2,3 and 6, to less than 200 Amps. Which of the following selections correctly identifies the indication -
(parameter) that will be used to verify compliance with this direction?
l;9 .
- h. Station Service Transformer High Side (6.4 KV) Anuneter(s)
l . - 480 VAC Ilus Ammeters
' 6.9 KV Station Auxiliary Anuneters
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Sum of individual equipment ammeters I
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Page 41 of 100
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indian l'olut Unit 2 Consolidatedndison Company ofNY REACTOR OPERATOR EXAMINATION. '
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Following a Safety injection you receive indication that 125 VDC Control Power has becii lost to all equipment powered from 125 VDC bus #21, Which of the following selections describe liOW and WliY this failure WILL or WILL NOT impact your ability to satisfy the following potential EOP requirements:
!O e - RCP T' rip Criteria e Si Reduction
- Close Accumulator Isolation Valves A. RCP Trip Criteria - No impact, RCPs can always be tripped from CCR
- Si Reduction - Si pumps must be tripped locally at 480 VAC Switchgear Accumulator Isolation valves - No impact, control power supplied from individual breaker AC feed 11. RCP Trip Criteria - RCPs that have lost control power must be tripped locally S1 Reduction - No impact, control power automatically transferred to 23 DC llus Accumulator Isolation Valves - No impact, control power supplied from individual breaker AC feed C RCP Trip Criteria - RCPs that have lost control power must be tripped locally SI Reduction - Si pumps must be tripped locally at 480 VAC Switchgear Accuniulator Isolation Valves - No impact, control power supplied from individual breaker AC feed D. RCP Trip Criteria - RCPs that liave lost control power must be tripped locally Si Reduction - No impact, control power automatically transferred to 23 DC 13us Accumulator Isolation Valves - Valves that receive control power from 21 DC bus cannot be closed RI
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Page 42 ot' 100
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Indian l'oint finit 2 Consolidated Ihlison Company of bT REACTOR OPEllATOR EXAMINATION With the unit at ilot Shutdowa (llSD) a full load test is being performed on #21 Emergency Diesel Generator (EDG). #22 and #23 EDGs are operable and in AUTOMATIC. After the EDG is fully loaded, a fire in the Station Auxiliary Transformer causes a loss of 6.9 KV power. Which of the following statements describe the expected response of the #21 EDG and 480 VAC bus SA?
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I i #21 EDG tiips and restarts when returned to AUTO Bus 5A Normal Feeder llreaker opens All loads strip lllackout loads sequence start
\ #21 EDG trips and restarts when returned to AUTO Bus 5A Normal Feeder Breaker opens All loads except running blackout loads strip Non-running blackout loads sequence start when bus is re-energized h #21 EDG continues to run Bus 5A Normal AND Emergency Feeder Breakers open Bus SA Emergency Feed Breaker closes 111ackout loads sequence start D. #21 EDG continues to run Bus 5 A Normal Feeder Breaker opens All loads except running blackout loads strip Non-running blackout loads sequence start Page 43 of 100 l
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Indian l'oint Unit 2 Consolidated inliwn Cornpany ofNl'
ItEACTOR OPERATOlt EXAMINATION
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Consider the following initial conaitions when selecting your answer:
. Reactor Power 100 %
- #21 RiiR Pump OOS for last six hours (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO action time)
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8' While oper'ating with the above conditions the control room receives a report that a lubricating oil leak on the oil cooler for #21 EDG has been discovered making #21 EDG INOPERABLE (7 day LCO action time). It is estimated that it will take 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to repair the leak. Which of the following statements correctly describes the actions that must be taken to ensure compliance with technical speci0 cations?
g Complete repairs on both #21 EDG and #21 RliR within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> or place the phnt in llot Shutdown I. Complete repairs on #21 EDG within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> or place the plant in llot Shutdown f Verify operability of remaining safeguards equipment and continue operatior.
observing Technical Specification limits and action times for equipment out of service.
D. Complete repairs on #21 RiiR Pump or be in liot Shutdown within seven hours.
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Page 44 of 100 l
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Indian Point Unit 2
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REACTOR OPERATOR EXAMINATION
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You have been directed to conduct a plant cooldown from 110t Shutdown (llSD) to 350 F at 50 F/hr using the condenser steam dumps. Approximately one hour after positioning the steam dump valves in the MANUAL mode on the Steam Pressure controller to establish the desired cooldown rate you notice that the cooldown rate has -
, - decreased fro,m 50*F/hr to 20*F/hr. Steam dump valve position has not been changed.
- s-Which of the following statements correctly describes the reason that the cooldown rate has decreased and the actions necessary to maintain a constant cooldown rate?
l A. Steam flow has decreased due to reduced steam pressure, Valves must be gradually opened as the cooldown progresses.
. Steam dump pressure controller will not allow Main Steam Pressure to decrease below the dial setting in M ANUAL or AUTO. Setpoint must be gradually reduced as the cooldown progresses.
AT between the steam temperature and the condenser cooling water (circulating water) has decreased. Cooldown rate cannot be increased unless circulating water.
flow is increased.
D T between RCS temperature and feedwater temperature has decreased. Feed flow must be increased to increase heat removal.
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Page 45 of 100
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Indian l'oint Unit 2 Consolidated Edison Company of bT REACTOR OPERATOR EXAMINATION
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- - Consider the following initial conditions when selecting your answer:
- Reactor Power 10 %
- - - RCS T.,, 548'F e Control Rods Bank D 125 steps / MANUAL 8, e Main T0rbine Stanup approaching synchronous speed
- llP Steam Dumps AUTOM ATIC/ Steam Pressure Mode While performing a plant startup, the turbine trips due to an overspeed condition. Which of the following selections is co Tect regarding the effect that this malfunction will have on the following RCS parameters:
L . RCS T.,e
. Average Loop AT
- Pressurizer Pressure Pressurizer Level
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A. RCS T.,. INCREASE a
Average Loop AT INCREASE RCS Pressure INCREASE
' Pressurizer Level INCREASE y nxI
'll RCS T.,, DECREASE Average Loop AT INCREASE RCS Pressure DECREASE ,.
Pressurizer level DECREASE
.4 C, RCS T. , INCREASE Average Loop AT INCREASE RCS Pressure DECREASE
. Pressurizer Level DECREASE '-
D. RCS T.,, INCREASE -
Average Loop AT DECREASE ,,-
RCS Pressure INCREASE Pressurizer Level ' INCREASE
' Page 46 of 100 l
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Indian l'olnt Unit 2 Consolidated Edison Company ofNY -
REACTOR OPERATOR EXAMINATION
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in Accordance with System Operating Procedures, WilEN RCS temperature is GREATER TliAN or EQUAL TO 350*F, the CCW Pump Auto Stan key switch must be in the NORMAL position, and WilEN RCS temperature is LESS TilAN 350*F, the CCW Pump Auto Start key switch is placed in the BYPASS Position.
8 Which ONE f)f the following statements co Tectly describes the reason for placing the CCW Pump Auto Stan key switch in the BYPASS position WiiEN RCS temperature is -
LESS TilAN 350 P!
. Block the CCW standby pump auto stan feature to prevent water hammer when the RiiR system is in service.
Allow operation of the CCW system with three pumps mnning to meet heavy demand l imposed by RilR heat load.
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% Technical Specifications allow defeating the Auto start feature below 350*F if three CCW pumps me OPERABLE, D Permit operation of the CCW system with less than three pumps running when CCW
- is Bowing through RilR lleat Exchangers. !
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Page 47 of 100
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indian l'oint Unit 2
_ Consolidated Ediso*: Company of Nl'
REACTOR OPERATOR EXAMINATION
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During power operation, the control rocm coordinates with the NPOs to shift the Essential Service Water lleader from the 1-2-3 lieader to the 4-5-6 licader. When the-necessary valving is completed the SWP Mode Control Switch on the SBF-1 panel is inadvertently left in the 1-2,3 position. The service water system is operating in the Three lleader Configuration.
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Which ONE of the following components will be supplied with service water if a Safety injection signalis initiated?
A. Instrument AirCompressorlleat Exchangers l- !!. Emergency Diesel Generators C. CCW lleat Exchangers D.- Containment Fan Cooler Units
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i Page 48 of 100
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Indian l'oint Unit 2 Consolidated Edison Company of Nl'
ItEACTOR OPERATOR EXAMINATION
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Which ONE of the following selections identifies ALL of the conditions which will initiate a Containment Ventilation isolation signal?
h. liigh Radiation (containment, R-41, R-42)
0, V Containtnent Isolation Phase B
\ Containment Phase A lsolation Signal Containment lii-lii Pressure Signal Manual Containment Spray Signal B. liign Radiation (containment, R-41, R-42)
liigh radiation (plant vent, R-44)
Containment Phase A isolation Signal Containment lii-Ili Pressure Signal Manual Containment Spray Signal liigh Radiation (containment, R-41, R-42)
liigh radiation (plant vent, R-44)
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Containment Phase A isolation Signal Containment Ili-Ili Pressure Signal Station Blackout 1. liigh Radiation (containment, R-41, R-42)
liigh radiation (plant vent, R-44)
Containment Phase A isolation Signal liigh Radiation (containment R-2, R-7)
Manual Containment Spray Signal Page 49 of 100 ,
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Indian l'olnt Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION
Which ONE of the following conditions would be considered a loss of containment integrity during normal operation?
A. An autornatic containment isolation valve is found to be inoperable in the CLOSED 8,
position'
11. Weld channel zone pressure indicates 50 psig, C. Personnel hatch inner door indicates OPEN, outer door indicates CLOSED.
D. Weld channel seal to the equipment hatch is lost.
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Page 50 of 100
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Indian Point IJnit 2
- Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION l. -
In accordance with procedure Residual licat Removal (RilR) System flow rate should ti AIJE6SI 2500 GPM per RilR Pump /RilR lieat Exchanger.
Which ONE of the following statements is correct regarding the reason for this limitation?
g* .
hMinimize turbulence downstream of the associated R11R liest Exchanger and -
liCV-638 and llCV-640.
k Prevent the RllR pumps from reaching runout conditions.
G. Prevent vortexing at the RCS llot Leg loop connection.
.: Ensure RilR Gow is evenly distributed via the Si manifold to all RCS cold legs.
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Page 51 of 100 P
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Indian l'oint IInit 2 Conwlidated Ihliwn Cornpany of Nl'
l 11isACTOlt OPIsilATOlt EXAMINATION
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Which ONE of the following indications is used by the Erne gency Operating Procedures to verify that Natural Circulation Cooling flow has t>een established?
5 < SG Predure - at saturation pressure for RCS llot Leg temperature h RCS Cold Leg Ternperatures - at saturation temperature for SG pressure C. Core Exit Ternperatures - at satt ration ternperature for current RCS pressure D. RCS Pressure - stable or increasing
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Page 52 of 100
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indian l'oint l/ nit 2 Consolidated lidison Cornpany ofNY REACTOR OPERATOR EXAMINATION When using Einergency Operating Pmcedure(EOP) E 1,less of Reactor or Secondary '
Coolant, you rue directed to operate Reactor Coolant Purnps (RCP) in accordance with
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EOP FR-C.2, Response to Degraded Core Cooling, rather than tripping the RCPs as directed by EOP E-1,if the RVLIS Dynamic Range indication indicates less than the value obtained fiorn the following table:
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44 % -
4 RCPs 1/S 309b -
3 RCPs 1/S 20 % -
2 RCPs 1/S 13 % -
1 RCP I/S Which of the following statements is cortect regarding the reason for the guidance described above?
A. RCPs operating under these conditions inay seiz.c when tripped preventing their testart in future recovery actions, h Tripping RCPs under these conditions could lead to core uncovering and an inadequate Core Cooling Condition.
( Tripping RCPs under these conditions would result in an increase of mass flow from the break due to phase separation of the Guid.
Tiipping RCPs under these conditions will result in a loss of RCS pressure control due to the loss of Pressuiizer spray capability.
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Page 53 of 100
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Indian l'ulnt Unit 2 Conwlidated Ihlison Cornpany of Nl'
REACTOR Ol'ERATOR EXAMINATION
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Which ONE of the following conc . ions is a condition which would required a control rod to be declared INOPERAllLE7 k. Control Rod 118 in Control Bank D is mis aligned froin the bank by 9 steps for
- 8, 12 hourfwith the bank at 218 steps.
11. Control Rod 11 14 in Shutdown Bank C is inechanically bound at 223 steps and will not trip, k. Control Rod D-4 in Shutdown llank A has a rod drop time of 1.6 seconds.
\. Control Rod I,-3 in Control Bank A, Rod llottom llistable fails to initiate a turbine runtrack.
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l' age 54 of 100
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Irulian l'oint Unit 2 l Consolidated Ihlison Cornpany ofNl'
ItEACTolt Ol'EllATOlt EXAMINATION Consider the following indications when selecting your answel:
- Reactor Power 50 %
- RCS Tm 553 F e #23 RCP,#2 Seal Standpipe LOW Level Alarm actuated 8, e #23 RCP Sealinjection flow 8.0 GPM e #23 RCP #1 Seal leakoff flow 3.0 GPM While operating with the above conditions and indications the Reactor Operator reports that the RCS leakage calculation is normal. Which ONE of the following failures or inalfunctions would support ALL of the above indications?
A. Failure of the #23 RCP #2 Seal 11. Failure of the #23 RCP #3 Seal C. Failure of the #23 RCP #1 Seal D. Failure of the #23 RCP Seal Package (#1,#2, and #3 Scal)
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i Page 55 of 100 l
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indian l'oint Unit 2 Conwlidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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While the reactor is in the Refueling Condition the Reactor Operator identines an unexplained increase in Source Range count rate and a steady positive 0.15 DPM Source Range Startup Rate indication. The Senior Reactor Operator directs the Reactor Operator to initiate boration of the RCS per A3.4, Uncontrolled Reactivity Addition.
l:8 Which one of the following boration now paths and methods is the prefened inethod for l completing this task in accordance with A3.4, Uncontrolled Reactivity Addition?
l l .
A. RWST via I.CV 11211. Emergency RWST Makeup Stop 11. MOV-333, Emergency lloration Stop to charging pump suction C. Normal boration nowpath at maximum rate to charging pump suction 1). Normal bosation nowpath at maximum iate to volume Control Tank
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Page 56 of 100
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Indian l'oint Unit 2 Consolidated Edison Cornpany of N}'
Iti ACTolt OPEllATOlt EXAMINATION
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Which ONE of the following will NOT result in the automatic stan of a Conv ;nent Cooling Water Pump?
A. A loss of offsite electrical power is followed by Unit Trip AND Safety injection.
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\. CCW lleader Pressure decreases to 60 psig with two CCW pumps mnning.
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An inadvertent Safety injection Signal is actuated due to an instrument failure. The l 480 VAC busses are ALL energized from offsite electrical power.
A Station 111ackout Signal is actuated due to a loss of offsite power following a Unit Trip. No Safety injection signal is present.
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Page 57 of 100
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ludian Point Unit 2 Consolidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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Consider the following event when selecting your answer:
- The controlling Pressurizer pressure channel (QlANNELj) has failed high with the unit at 100'7c power
,p * RCS pressure is stabilized at 2115Tsig by manually closing the Pressurizer spray ;
valves After the plam is stabilized the Senior Reactor Operator reviews the following Technical SpeciGcation Limit:
Rext9tSitt anLSAtc.!!1fttnurslefnpriattttc;13sl l lilow itaig The following DNH related parameters penain to four loop steady state operation at inwer lesels greater than 98% of full rated power:
. Itcaelor Coolant System T,y 6 587.2*F e Pressuruer Pressure b 2190 psia
. Itcactor Coolant System Total Flow Itate 2 331.840 GPM licm (b), Pressuriter pressure, is not applicable during either a thermal power change in execss of $% of rated thermal power per minute, or a thermal power step change of 10% of rated thermal power.
Under the applicable ope:ating conditions, should reactor coolant T,y. or Pressuriter pressure exceed the values given in items (a) and (b) the parameter shall be restored to its applicable range within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Which of the following statements correctly describes the appropriate response,if any, the Semur Reactor Operator should take to ensure compliance with this technical speciGcation?
A. Specification for Pressuriter Plessure is not applicable because the transient was induced by an instmment failure.
IL Specification for Pressurizer Pressure is not applicable because there was no change in reactor power.
C. Specification is not applicable because it only applies when power is 2 98% of ,,
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IMI[illEULLElWEllwhich is equivalent to 98% of the maximum attainable power level of 108% (iiigh Flux Trip Setpoint)
D. Restole RCS pressure to GREATER TilAN 2190 psia within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce power to 6 98%.
Page 58 of 100
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Indian l'olnt Unit 2
>
Consolidatedlidison Company ofNY REACTOR OPERATOR EXAMINA'IlON A main steam line avpture has occurred with the plant at hot shutdown resulting in a Main Steam Line AP Safety injection Signal. After analyzing the following indications detennine which ONE of the following selections correctly identifies the ponion of the Main Steam System that has suptured?
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!0 8 e 21 SG Pressurc 780 psig(decreasing slowly)
e 22 SG Pressure 782 psig (decreasing slowly)
e 23 SO Pressure 340 psig (decreasing rapidly)
e 24 SG pressure 775 psig(decreasing slowly)
- ALL Main Steam Line Flow Indications 0 lbm/hr
- ALL MSIVs indicate OPEN
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A. 23 SG Main Steam Line Upstream of MSIVs outside containment
!L 23 SG Main Steam Line between SG and Flow Element C. 23 SG Main Steam Line Downstream of MSIVs D. 23 SG Main Steam 1.ine between Flow Element and Containment Penetration
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Page 59 of 100
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Indian Point Unit 2
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Consolidated Edison Company ofNY
! REACTOR OPERATOR EXAMINATION
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) Consider the following event when selecting your answer:
! * Reactor Power 15 % l
} * T. 547'F :
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j,a * Turbine.Startup in progress Turbine at synchronous speed ,
i + 11P Steam Dumps Pressure Mode - AUTOM ATIC l
,. During plant startup with the above conditions, a failure of the exhaust boot (expansion joint) between #21 Low Pressure Turbine and the condenser results in a rapid loss of condenser vacuum to atmospheric pressure over a period of 5 minutes.
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! - Assuming that all plant protection AND control systems function as designed, which ONE j- .of the following statements correctly describes the response
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(INCREASE / DECREASE /NO CliANGE) of the following parameters pta result of this
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. Reactor Power
- * Turbine RPM e Steam Flow
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T. .
- Reactor Power DECREASE Turbine RPM DECREASE
$ cam Flow NO CilANGE T.,, NO CilANGE
Reactor Power DECREASE I '1:urbine RPM DECREASE
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Jteam Flow INCREASE
T.,, INCREASE C. Reactor Power DECREASE Turbine RPM DECREASE.
' Steam Flow -DECREASE ,,..
. T.,, INCREASE
! - Reactor Power NO CilANGE
} -- 'urbine RPM DECREASE
- Steam Flow - NO CilANGE i - T.,. NO CllANGE
Page 60 of 100
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Indian Point Unit 2
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Consolidated Edison Company ofNY l
A REACTOR OPERATOR EXAMINATION
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During power operation at 100% reactor power a loss of electrical power to #21
- Instrument Bus occurs causing the loss of Irr-412A. Which ONE of the following statements describes the response of the High Pressure Steam Dump System?
i ly Assume tha,t no equipment or instrumentation was out of service before the failu~re, ii
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. Loss of load interlock will trip arming the steam dumps.
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- T,,, will fail to 547'F causing the liigh Pressure Steam Dumps to OPEN.
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.. Steam dump actuation will be inhibited due to loss of power to loss of load interlock.
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l D. T,,,will fail to 547'F. liigh Pressure Steam Dumps will OPEN jf loss of load i interlock trips.
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Indian l'oint linit 2 Consolidated lalison Company of Nl'
11EACTOlt OPEltATOlt EXAMINATION
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The Conventional NPO has reponed a sinali electrical fire on the 5' elevation of the Turbine Building. The following fire fighting equipinent is available to the Fire Biigade.
Select the equipment which is most suitable for extinguishing a fire of this type?
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A. Portable bO2 Gre extinguisher 11. liigh pressure water hose C. Water stream portable fire extinguishe D. Dry chemical Dre extinguisher
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Page 62 of 100
Indian Point linit .?
Conwlldated Ediwn Company of Nl'
REACTOR OPEllATOR EXAN11 NATION
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A fire in the control building requires that the Cential Control Room be evacuated. You have been designated as the l'irst RO and the SRO has directed you to trip the Reactor locally. Which of the following selections list in ORDER the locations and equipment froin which you would accoinplish this task, assuming that you are unsuccessful after each
. , atteinpt. ,
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Cable Spreading Room Reactor Trip lireakers Cable Spreading Room . Rod Drive htG Set lireakers MOV Switchgear lloom - Rod Drive h1G Set lireakers l 4b W Switchgear Room - Ilus 3 A and 5A Supply lltcakers l
6.9 KV Switchgear - Station Service Transfonner 3 and 5 Supply lireakers 11. 480V Switchgear Room Rod Drive h10 Set Ilreakers 480V Switchgear Room lius 2A and 6A Supply lireakers Cable Spreading Room - Reactor Trip lireakers Cable Spreading Rooin - Rod Drive htG Set Ilreakers 6.9 KV Switchgear - Station Service Transformer 2 and 6 Supply llicakers C. Cable Spicading Room Reactor Trip lireakers Cable Spreading Room Rod Drive h10 Set lireakers 480V Switchgear Room - Rod Drive h1G Set llicakers 480V Switchgear Room - llus 2A and 6A Supply 11reakers
'i.9 KV Switchgear - Station Service Transformer 2 and 6 Supply llreakers 1. Cable Spicading Room - Reactor Tiip 11reakers Cable Spicading Room Rod Contial System Pow er Cabinets SUV Switeligear Room - Rod Drive h1G Set lireakers 4hlV Switchgear Room - Ilus 2 A and 6A Supply llicakers 6.9 'Y Switchgear - Station Service Tiansformer 2 and 6 Supply lireakers Page 63 of 100
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indian l'oint finit 2 Consolidated Ihlison Company of Nl'
REACTOR OPERATOR EXAMINATION
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While verifying Containment isolation valves are in the correct position following a Safety injection actuation due to Hi Ili Containment Pressure, you notice the following "Two is Tnie" indication for MOV 222, RCP Seal 1.cakof f Containment isolation Valve:
e Left side oflight illuminated - AMilER light p
- Right side oflight Extinguished Which ONE of the following selections is conect regarding the expected position of MOV-222, AND the indicated position of MOV-222 with respect to the "Two is Tme" indicating lights?
l A lixpected Position Ol'EN Indicated Posit en OPl!N l
11. Expected Position Cl.OSED Indicated Position CLOSED I C. Expected Position OPEN Indicated Position Cl.OSED f
D. Expected Position Cl.OSED Indicated Position OPEN
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Page 64 of 100
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Ind'an l'olnt linit 2 Conwlidated Ihliwn Company of N)'
REACTOR OPERATOR I?XAMINATION
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Following a reactor trip you are directed by ES 0.1 Reactor Trip Response, to verify that all control rods are fully inserted. Which ONii of the following rod position indications would meet the criteria for a control rod NOT being fully inserted?
' A. Individ'ual IRPI teading 5.5 inches 1( Group Step Counters not indicating zero (0)
C. Proteus Computer rod position 13 steps
}{. 0.05 Volts on Digital Volt Meter (DVM)
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Page 66 of 100 l
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Indian Point finit 2 Conwlidated Ihliwn Company of Nl'
REACTOR OPEltATOlt EXAMINATION
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While conducting a natural circulation cooldown using ES 0.2, Natural Circulation Cooldown. you are directed to control SG levels. Which ONE of the following actions if perforined could ternporarily irniede or reduce natural circulation flow?
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8' A. Steam p'enerator #21 level is allowed to slowly increase to 55% as seen on the narrow range level indicator.
11. Auxiliary feed flow to #21 SG is sapidly increased froin 50 Gl'hi to 200 Gl'M
l Cy Stearn pencrator #21 level is allowed to slowly decrease to 35% as seen on the narrow I
iang 'evelindicator Auxiliary feed flow to #21 SG is :apidly decicased from 200 GPM to 50 GPM
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Page 67 of 100
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Indian l'ai :t linit 2 Conwlidated Ediwn Company of A'l'
REACTOR OPERATOR EXAh11 NATION s
While perforrning the inninediate actions for a Reactor Trip using EOP E-0, Reactor Trip or Safety injection, you are directed to de-encrpire 480 VAC busses 2A and 6A to trip the reactor since a reactor trip cannot be verified by available indication. After 480 VAC bus 2A and 6A are re-energized you are directed to depress the Blackout Relay Reset 480V push button on the SC panel jf the Main Generator Output Breakers are CLOSED.
Which ONE of the following staternents is conect regarding the reason that the Blackout Relay inust be reset at this time?
. Resetting the lilackout Relay at this tiene reinoves the 480 VAC Bus 2A Unde > oltage Signal froin the blackout logic and .lE the Main Generator Output Hrcakers are CLOSliD (86P and 86BU selays reset) a station blackout signal will be avoided.
11. Resetting the lilackout Relay at this tirne seinoves the 480 VAC lius 2A ANI) 480 VAC lius 6A Undervoitage Signals from the blackout logic and if the hiain Generator Output Breakers are CLOSED (86P and 86BU relays reset) a station blackout signal will be avoided.
C Resetting the lilackout Relay at this time reinoves the 480 VAC llus 6A Undervoltage Signal froin the blackout logic and IE the hiain Generator Output llicakets are CLOSED (86P and 86BU relays reset) a station blackout signal will be avoided, hiain Generator Output liscalets should aheady be OPEN. Depressing the Illackout Relay Reset 480V push button will tiip the 86P and 86BU relays causing the hiain Generator Ouiput lireakers to OPEN.
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Page 68 of 100
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loulian l'oint thall 2 Consolidated iklison Company of Nl'
iti?ACTOlt Ol'EllATOlt EXAMINATION
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Consider the following event when selecting your answer:
- All Si egyipment has operated as designed, i8, * E01' ES 1.2, Post LOCA Cooldown and ikpressurization has been implemented.
While performing the actions required by EOP ES 1.2, you are directed to establish maximum charging flow to the itCS. Which ONE of the following statements is correct
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regarding the reason for this action when performing a Post LOCA Cooldown and l Depressurization?
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A. Maximum charging flow is established in order to provide maximum auxiliary spray flow capability in the event that Ilea .or Coolant Pumps ate not running and normal spray is unavailable.
11. Maximum charging Ilow is established to ensure that maximum boration capability exists.
C. Maximum charging flow is established in an attempt to achieve St Termination Criteria, thus avoiding the tedious task of Si iteduction.
D. Maximum charging flow is established in order to provide sufficient makeup so that Si pumps can be more icadily reduced during the Si iteduction sequence.
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Page 69 of 100
indian l'oint linit 2 Consolidated lidison Company of NY REACTOR OPEllATOR EXAMINATION
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Consider the following conditions when selecting your answer:
. Large lireak LOCA has occurred )
- NIS Power Range indication 0%
. Intermed, late Range Startup Rate 1/3 DPM
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. Core Eiit Thermocouple Temperatures 1210'F l
. Reactor Coolant Pumps TRIPPED l
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. RVLIS Natural Circulation Range 30 %
. RCS Subcooling <0F e Steam Generator Narrow Range Levels All < 10%
. TotM Auxiliary feedwater flow 740 gpm
. I'.essurizer level 0%
The Watch Engineer reports that the above conditions require entry into the Functional Restoration Procedures.
Which ONE of the following Functional Restoration Procedures must be implemented?
A. FR-C.1, Response to inadequate Core Cooling 11. FR-S.1, Response to Nuclear Power Generation / ATWS C FR il.l. Response to Loss of Secondary licat Sink D. I R-i.3, Response to Voids in the Reactor Vessel Page 70 of 100
Indian l'oint Unit 2 Consolidated Iklison Cornpany of N)*
REACTOR OPERATOR EXAMINATION
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Following a Manual Itcactor Trip you are unable to verify that the Reactor is, tripped in accordance with E-0, Reactor Trip or Safety injection, and FR S.1, Response to Nuclear Power Generation /ATWS, is implemented.
Which ONE qf the following describes the sequence of actions directed to be used in FR-
'8, S.1, Response to Nuclear Power Generation /ATWS, for tripping the Main Turbine if the turbine trip cannot be verified? (Assume each action is unsuccessfulin tripping the
, turbine)
A. Manually trip the turbine from the CCR, CLOSE the MSIVs from the CCR and trip the turbine locally at the " Front Standard."
B. Manually trip the turbine from the CCR, manually runback the turbine in the CCR and trip the turbine locally at the "Fioni Standard."
C. Manually trip the turbine from the CCR, CLOSE the MSIVs from the CCR, manually runback the turbine in the CCR and locally CLOSE the MSIVs D. Manually trip the turbine from the CCR, trip the turbine locally from the " Front Standard," manually mnback the turbine in the CCR and locally CLOSE the MSlVs
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Page 71 of 100
indian Point Unit 2 Consolidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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Consider the following events when selecting your answer:
- A tube rupture has occurred in #22 Steam Generator (SG)
- A Safety injection (SI) signal has been actuated
,,, e EOP E 3 Steam Generator Tube Rupture has been implemented
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The SRO has directed you to dump steam from the intact SGs at the maximum rate to establish a Core Exit Temperature of 488'F AND then stop the cooldown.
I Which ONE of the following statcments correctly describes the reason for reducing RCS temperature to this value? -
. Reduce RCS pressure by causing an outsurge from the Pressurizer to minimize leakage into the #22 SG.
h Establish sufficient subcooling ii the RCS so that the RCS will remain subcooled after pressure is decreased to #22 SO pressure.
Establish sufficient subcooling in the RCS so that the Reactor Coolant Pumps will not have to be tripped when the RCS pressure is decreased to #22 SG pressure.
Reduce temperature of RCS fluid leaking into #22 SG to reduce #22 SG pressure to minimize potential of radioactive release through the atmospheric steam dump valve.
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Page 72 of 100
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indian Paint Unit 2 Consolidated thfison Cornpany ofNY l
REACTOR OPERATOR EXAMINA*' 40N Consider the following event when selecting your answe e Reactor Power 80 %
- RCS T.., 557'F
, e Control, Rods Control Bank D 195 steps / Manual ;
e RCS Boron Concentration 980 ppm Beginning Of Life (BOL)
After withdrawing control rods to adjust T., you note that when you release the in-llold-Out switch Control 13ank D room continue to withdraw for an additional 20 steps. As a result T.., increases to 5*F above program.
Which GNE of the following statements is TRUE regarding this event?
A. IF the same event occurred at EOL the INCREASliin RCS T., would have been GRiiATER.
' . IF the same event occurred at EOL the INCRiioSii in RCS T.., would have been the SAME.
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IF the same event occurred at EOL there would have been NO INCRiioSliin RCS T.s e.
h. IF the same event occurred at EOL the INQlliAEliin RCS T,w would have been LOWER.
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Page 73 of 100
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Inclian 1% int Unit 2 Consolidated Ihlison Cornpany of Nl' j l
REACTOR OPERATOR EXAMINATION l
l Which ONE of the following AUTOMATIC actions will NOT occur if a single control ;
tod drops from the fully withdrawn position to the fully inserted position while the reactor is at 100% power?
si '
\. NIS Dropped Rod Stop 11. APPROAC111NG ROD INSERTION LIMIT 12.5" and the ROD INSERTION LIMIT 0" alarms actuate on Panel S A.
\
% Turbine Runback
\
l{. Rod llott im Rod Stop
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Page 74 of 100
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ludian Point Ihtir 2 Consolidated iklison Company of NY REACTOR OPERATOR EXAMINATION
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bnsider the following event when selecting your answer:
. A Pressurizer Power Operated Relief Valve (PORV) has OPENED and failed to close.
,
, . A Safety injection Signal has been actuated,
- RCS Pressure stable 1480 psig IN D l
. Pressurizer Relief Tank pressure 30 psig 45
. Pressurizer relief tank ternperature 140*F l . Pressuriter Level 90%
l e Pressurizer Relief Tank Level 82 %
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Using the indications and conditions provided determine which ONE of the following
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temperatures would be indicated on the PORV Downstream Temperature Indicator?
A. 593"F B 275'F
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C. 250"F D. 140oF
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Page 75 of 100 L - -
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Indian l'aint Iinit 2 Conwlidarnilidiwn Company of Nl'
ItEACTOlt OPEllATOlt EXAMINATION
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Abnorrnal Operating Instruction ( AOI) 3.1, Chernical Voluine Control System (CVCS)
Malfunctions, states that when preparing to start a Charging Pump, the Charging Putnp controller must be placed in manual AND set for approximately 20% before the purnp is started?
Which ONh'of statements is correct regarding the reason for setting the controller to 20%
before starting the Charging Pump?
Ensure that a low bearing oil pressure trip does not occur during the charging pump start.
h Minimize staning cunent on the changing pump inotor.
Balance MANUAL. signal with AUTO signal before the pump is started.
{. Provided a minimum of 8 GPM seal injection flow to each itCP as soon as the pump is started.
Page 76 of 100
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Indian l'oint Unit 2 Consolidated Ihlison Cornpany of NY REACTOR OPERATOR EXAMINATION
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When operating the Residual lleat Removal System (RllR) with the Reactor Coolant System (RCS) at reduced inventony care inust be taken to control RCS water level such that the RilR pumps do not cavitate or tv:come airbound. SOP 1.2, Draining Reactor Coolant System, and AOI 4.2.1, Loss of Residual lleat Removal System, impose g restrictions op minimum RCS water level based on certain conditions / parameters.
Which ONE of the following parameters / conditions is a factor in determining the minimum allowable itCS water level?
h Reactor Coolant Sy:. tem Temperature 11. RilR System i10w Rate Which RilR lleat Exchanger is in service
. Which RilR Pumpis Running
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Page 77 of 100
indian l'olnt thsit 2 Consolidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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Consider the following indications when selecting your answer:
- Reactor Startup in progress gdp
+ Source Range N31 Count Rate 2X10' cps .,,goh
, . SourceJtange N32 Count Rate 4X10' cps
- Intermediate Range N35 Cunent <lX 10 " amps
- Intermediate Range N36 Cunent <lX 10'" amps e Source Range N31 Startup Rate 0.5 dpm
. Source Range N32 Startup Rate 0.1 dpm l
- Control Rods Control llank D - 100 steps /M ANUAL l Using ONLY the infonnation provided detennine which ONE of the following actions ai,d l associated reason is approp iate regarding the continuation of the reactor startup?
A. Source Range N32 is icading high due to a failure in the Pulse lleight Discrimination circuitry and should be considered inoperable. The startup may continue without further action.
(. Source Range N31 is reading low due to a failure in the Pulse licight Discrimination circuitry and should be considered inoperable. The startup may continue without further action.
Intermediate Range N35 AND N36 are not responding. Startup may continue as long as neutron Oux remains in the source range.
h. Nuclear instnnnentation is NOT indicating as t.nticipated. The approach to criticality SilALL be stopped AND no actions SilALL be taken which could add positive icactivity until the disciepancy is resolved.
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Page 78 of 100
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ludian l'aint thoit 2 Conwlidated Ihliwn Company of N)*
IllsACTOR Ol'ICitATOlt I?X AMINATION
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Technical Specifications requises that itCS activity (f or nuclides other than uitiurn with half lives of rnore than 30 ininutes) tw 1.liSS Til AN 60/li-bar pCi/cc when itCS ternperature is GRIIATliR TilAN 500'F. Which of the following correctly describes the reason for the 500'Flimit?
- I A. The probability for a I arge lineak LOCA has increased h The saturation pressure for 500*1
- is less than the S/G Safety valves' lift setpoint for Steam Generator Tube Rupture release concerns.
I C. Tb: volatility of the radioactivity increases above 5(XFl!.
11 Nuclear Instnnnentation suay be indicating incorrectly due to the increased radiation levels resulting from the Ingh activity.
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Page 65 of 100
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indian l'oln tinit 2 Conwlidated Ihliwn Company ofNl'
ItEACTOR OPERATOR EXAh11 NATION
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A scactor trip has occuned. Apptoximately 30 minutes after the reactor has tripped the Reactor Operator is performing actions directed by ES-0.I, Reactor Trip Response, when he notes the following indications:
Intermedjate Range N35 IX10* amps (stable)
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. Interme'diate Range N36 1X 10'" amps (stable)
. Source Range N31 0 cps (stable)
. Source Range N32 0 cps (stable)
. SOURCE RANGE LOSS OF DETECTOR VOLTAGE annunciator illuminated Which ONE of the following statements correctly describes the response,if any, that the Reactor Operator should take regarding these indications?
k. Manually te-enerpi/c the Source Range NIS by depressing the Train A and Train 11 Intennediate Range l'ennissive Override push buttons.
ig Manually re-eneigize the Source Range NIS by depressing the Train A and Train 11 Powet Range Pennissive Overside push buttons.
Initiate rapid boric acid injection in accordance with, A 3.4, Uncontrolled Reactivity Addition.
. Monitor Intermediate Range N35, and verify reinstatement of the Source Range NIS when both Intennediate Range instiuments are less than IX 10* amps.
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i Page 79 of 100
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Indian l'oint Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION
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Consider the following indications when selecting your answer:
. Reactor Power 100 %
Tm 559 F
)p
. Total RGS Leakage 0.8 ppm (includes SG tabe leakage)
. SG Tube Leakage 0.25 gpm l
The Senior Ltch Supervisor has directed the control room operatc,rs to perform a reactor shutdown due to increasing secondary side activity caused by SG tube leakage.
l Which ONE of the following statements is correct regarding the anticipated change,if any, l
in total RCS leakage as a result of the plant shutdown?
l A. Total RCS leakage will increase, dae to the increase in SG tube leakage.
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( B. Total RCS leakage will remain the same, increase in SG leakage will be offset by decrease in other RCS leakage.
i C. Total RCS leakage will decrease, due to the decrease in SG tube leakage, D. Total RCS leakage will remain the same, decrease in SG leakage will be offset by increase in other RCS leakage.
Page 80 of 100
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indian l'oint I) nit 2 Consolidated Ihliwn Company ofNl'
REACTOR OPERATOR EXAMINATION
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While recovering from a Reactor Trip due to a loss of both main feedwater pumps, EOP FR li.5, Response to Steam Generator Low Levelis entered due to indication that water level in #23 SG has decreased to 0% WIDE RANGE level. In accordance with FR-il.5, the SRO directs you to feed #23 SG at LESS THAN 100 GPM UNTIL water level is GIGATER TilAN 10% as indica.ed on #23 SG WIDE RANGE level indication.
.;9 .
Which ONE of the following statements is TRUE regarding the reason for limiting feed water flow to #23 SG to LESS TH AN 100 GPM until level is GREATER TilAN 10%?
A. Feed now is limited to prevent a rapid RCS cooldown which could result in challenge to the INTEGRITY critical safety function.
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l B. Feed flow is limited to prevent unnecessary thermal shock to a "liot Dry SG" which
! could result in SG tube failure.
C. Feed flow is limited to prevent Gashing in the SG which could result in lifting the SG Safety Valves.
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D. Feed flow is limited to prevent runout conditions on the #23 Auxiliary Feed Water PumI'.
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Page 81 of 100 o
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-indian l'oint Unit 2 -
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Consolidated lidison Company of Nl'
REACTOR OPERATOR EXAMINATION .
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1Which ONE of the following indications i' used by the Emergency Operating Procedure E-3, Steam Generator Tube Rupture, to IDENTIFY the RUPTURED Steam Generator'l b. Safety Inlection flow deviation for the offected loop, p, y- -
H. Afrected Steam Generator's pressure increase, liigh Radiation from the affected SG blowdown line.
y q. liigh Radiation from the Steam Jet Air Ejector Vent Radiation Monitor R45.
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Page 82 of 100
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Indian l'oint Unit 2 Consolidated Edison Cornpany of Nl'
REACTOR OPERATOR EXAMINATION
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Following a manual Reactor Trip from 100% power you are performing Step 3 of E-0, Reacter Trip and Safety injection," Check if SI Actuated", and note the fo'!owing indications:
. SI Annupciator Not Lit
!8, . SI Pum'ps None Running
. Pressurizer Pressure 2090 psig (stable)
. Steamline AP All < 50 psid (stable)
. Steam Line flow All < 100,000 lbm/hr (stable)
. Containment Pressure 0.75 psig (stable)
. Pressurizer Level 5% (stabic)
. RCS Subcooling Margin 87 F(stable)
Which ONE of the following actions should you take in response to these indications, and >
W11Y is the action required?
A. Manually initiate Safety injection due to Pressurizer Low Level. -
l 11. Manually Initiate Safety injection due to failure of the liigh Steam Flow Si to actuate.
C. Check RCS subcooling table. If LESS Til AN required, manually initiate Safety injection due to Low Subcooling.
D. Transition to ES-0.I, Reactor Trip Response, Safety injection is NOT required.
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Page 83 of 100 o
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Indsan l'aint Unit 2 Consolidated Edison Company of N)*
REACTOR OPEllATOR EXAMINATION
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A Large 13reak Loss of Coolans Accident has occurred. Approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after aligning for Cold Leg Recirculation, the Senior Reactor Operator implements EOP ES-1.4, Transfer to liot Leg Recirculation. Which ONE of the following statements is correct regarding the reason for placing liot Leg Recirculation in service at this time?
.s* -
A. Hot leg recirculation is implemented to sweep non-condensable gasses from the reactor head region.
it liot leg recirculation is implemented to cool the reactor head to enable RCS depressurization without additional void formation.
C. Ilot leg iecirculation is implemented to ielill the n actor vessel and preclude fuel rod damage at the top of the core.
l D. Ilot leg recirculation is impleinented to pievent boron precip'. ation in the core.
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Page 84 of 100
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y m /W hkndiarr vi
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-^{h Conwlidawd Edison Cornpany ofNY REACTOR OPERATOR EXAMINATION Following a Steam Generator Tube Rupture, EOP E-3, Steam Generator Tube Rupture, has been implemented.
l The Reactor Operator is controlling RCS temperature using the Condenser Steam Dumps, The Senior Reactor
. Operator directs the Reactor Operator to depressurize the RCS to LESS THAN RUPTURED SG pressure using
the Pressurizer PORVs since the RCPs have been tripped.
2 8#
j The Reactor Operator dLQES the steam dump valves and prepares for depressurization of the RCS. Before i commencing depressurization the Reactor Operator notes the following indications:
j- e RCS Wide Range Cold Leg Temperatures 480'F (increasing slowly)
- Core Exit Temperature 488*F(increasing slowly)
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- Pressurizer Level (hot calibrated) 09 i
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-e RCS Pressure 1370 psig (increasing)
- Ruptured SG Level 889 (Narrow Range - increating)
- e Ruptured SG Pressute 1015 psig (increasing slowly)
Just prior to closing the Pressurizer PORVs the Reactor Operator notes the following indications:
i .* RCS Wide Range Cold Leg Temperatures 506*F(increasing slowly)
- Core Exit Temperature 510 F(stable)
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- Pressurizer Level thot calibrated) 80% (increasing rapidly)
j- * RCS Pressure 760 psig (increasing)
. Ruptured SG Level 859 (Narrow Range - decreasing slowly)
i e Ruptured SG Pressure 1005 psig (decreasing slowly)
,
Which ONE of the following statements could explain Al1 of the changes in the above indications that have
[
occurred since the Reactor Operator commenced depressurization?
N.
. N RCS cold leg temperature has INCREASED due to Hashing of RCS Quid in the hot leg.
@R level has INCREASED due to increased makeup Dow AND voiding in the Reactor llead.
7 1[ tured SG level is DECREASING due to SG Guid back0lling the RCS.
. RCS cold leg temperature has INCREASED due to REDUCED natural circulation now rate.
Pressurizer level has INCREASED due to increased makeup Gow AND voiding in the Reactor Head.
x Ruptured SG level is DECREASING due to SG Guid back0lling the RCS.
RCS cold leg temp rature has INCREASED due to REDUCED natural circulation now rate. ,;-
dicated Pressurizer level has INCREASED to due Dashing of the fluid in the level instrument teference leg.
Rt tured SG level is DECREASING due to SG Guid back0lling the RCS.
.
. RCS cold leg temperature has INCREASED due to STOPPAGE of natural circulation Dow.
Pressmizer level has INCREASED due to increased makeup Dow AND voiding in the Reactor Head.
Ruptured SG level is DECREASING due to steaming through the Atmospheric Steam Dump valve.
Page 85 of 100
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-Indian l'oint Unit 2 -
Consolidated Edison Company of Nl'
REACTOR OPERATOR EXAMINATION
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A refueling operator has just placed an irradiated fuel assembly in the containment side upender when the Refueling SRO notes that refueling cavity water level is decreasing
- rapidly.
g Which ONE of the following actions should be directed by the Refueling SRO?
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A. Disengage the manipulator from the fuel assembly and lower the upender to the fully lowered position, TilEN evacuate containment.
B. Withdraw the fuel assembly from the upender and move it to the reactor, THEN evacuate containment.
l. C. Withdraw the fuel assembly from the upender and store it in the manipulator mast, TilEN evacuate containment.
D. Disengage the manipulator from the fuel assembly and leave upright, TliEN evacuate containment.
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Page 86 of 100
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Indian l'oint Unit 2 Consolidated Edison Company of Nl'
REACTOR OPEltATOlt EXAMINATION
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Following a Reactor Trip and Safety injection you are verifying 480 VAC Busses energized by offsite power using EOP E-0, Reactor Trip or Safety injection. You note the following conditions associated with the 480 VAC Busses:
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- All 480 VAC Bus Normal Feeder Breakers OPEN
. All Eme'rgency Diesel Generators in Service )
. All 480 VAC Bus Emergency Feed Breakers CLOSED Which ONE of the following actions should NOT be performed when completing this IMMEDIATE ACTION step?
\ A. Start ONE charg;ng pump in MANUAL at maximum speed.
l 11. Ensure the following MCCs - ENERGlZED
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- MCC 26A
= MCC 26B
+ MCC 26C I
l C. Reset Lighting l
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D. Ensure the following MCCs - ENERGlZED
. MCC 24 A
= MCC 27A
- MCC 29A
= MCC 211
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Page 87 of 100
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Indian 1% int Unit 2 Consolidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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You have been directed to coordinate the completion of a System Check Off List (COL)
for the Safety injection System. You note that the COL requires Independent Verification.
Which ONE,of the following statements is TRUE regarding acceptable practice when
' 8, conducting INDEPENDENT VERIFICATION of items contained in a COL that requires independent verification?
. Operator Performing the INDEPENDENT VERIFICATION (Second Checker) may perform the verification at the same time as the First Checker and should record the AS FOUND position of each component.
B. Operator Performing the INDEPENDENT VERIFICATION (Second Checker) must perform the verification independently of the First Cliecker, and should reposition the component IF it is not in the required position.
k. Operator Performing the INDEPENDENT VERIFICATION (Second Checker) may-perform the verification at the same time as the First Checker and should record the
. AS LEFT position of each component. ,
b Operator Performing the INDEPENDENT VERIFICATION (Second Checker) must perform the verification independently of the First Checker, and should record the AS FOUND position of each component.
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Page 88 of 100
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indian l'oint Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION
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A maintenance mechanic calls the control room and reports that a valve he has just removed from the service water system has a black and white STOP TAG attached to the .
handwheel that requires the valve to be in the OPEN position.
i l
Which ONE of the following actions should you take regarding this report?
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h No action is necessary since the valve was tagged OPEN. Work protected by this tagout is not affected.
B. Obtain the tagout number and report the finding to the work control center. Work protected by this tagout should cease until the discrepancy is corrected. !
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Direct the maintenance mechanic to place the tag on an adjacent valve which is in the OPEN position. Work protected by this tagout is not affected.
. Obtain the tagout number and record the finding in the SRO log. Work protected by this is not affected since the valve was tagged in the OPEN position.
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Page 89 of 100
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Indian l'oint Unit 2 Consolidated Edison C<nnpany of N)*
REACTOR OPERA' LOR EXAMINATION
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While investigating an increase in RCS leakage, preparations are made to make a containment entry using SAO 219. Containment Entry and Egress. Which of the following actions is required if entry is to be made within the crane wall on the 46'
elevation?
38 -
Reactor Shutdown is required for all entries in the vicinity of RCS loop piping.
It llealth Physics and Operations will determine the need to reduce radiation levels by reducing power.
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p. If entry is planned for less than 10 minutes within crane wall, no power reduction is
\ required.
. If total job exposure will exceed 5 person-rem, a power ieduction to 75% is required.
l Page 90 of 100 1
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.' indian l'aint Unit 2 '.
Consolidated Ediwn Cornpany ofNl'
REACTOR OPERATOR EXAMINATION
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. During power operation the 1)lESEL BLDG. FIRE PROT OPERATION annunciator is
. activated. Which ONE of the following statements is correct regarding the response of
. the Emergency Diesel Generator (EDG) Building Fire Protection System?
A. EDG Btiilding Fire Protection System is a dry line system and water spray is activated to the entire building by a deluge valve. The above alarm could mean that all of the
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spray nonles have actuated.
B. EDG Build'ing Fire Protection System is a concentrated foam system and foam is activated locally by thermostats. The above alarm could mean that the foam system has actuated.
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C. EDG Building Fire Protection System is a liALON system and H ALON is activated locally by thermostats. The abm e alarm could mean that II ALON system has actuated.
D. EDG Building Fire Protection System is a wet line system and water spray is activated locally at each spray noule. The above alarm could mean that one of the spray nozzles has actuated.
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Page 91 of 100
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indian Point Unit 2 Consolidated Edison Cornpany ofNl'
REACTOR OPERATOR EXAMINATION
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Station policy requires that each individual follow practices that ensure that their personal radiation exposure is kept As Low As Reasonably Achievable (ALARA). Consider the followit g situation:
. Maintenance must be perfonned in an area where the general area radiation levels are 8, '
5 mr/hriand 100 mr/hr '
. The maintenance activity is estimated to take I person I hour
. Most of the radiation is due to contamination of the floor in the area.
Select the ALARA practice that would result in the lowest achievable radiation exposure (total Person-Rem) for this job?
s A. Equip worker with a plastic face shield and protective clothing against contamination.
N B. Decontaminate the area (requires 2 people and 45 minutes) hefore commencing work, l \
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. Cover the floor in plastic sheeting (requires 2 peopic 30 minutes) before commencing l work.
Cover floor with lead blankets ( requires 3 people 30 minutes) hefore commencing work.
Page 92 of 100 l
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Indian Point Unit 2 Conwlidated Ediwn Company of NY ItEACTOR OPEllATOR EXAMINATION
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You are performing a Reactor Startup. Your watch relief from the on coming shift has arrived just as you are ready to withdraw the conttol banks to approach criticality.
Which ONE of the following statements describes an acceptable practice when conducting shift turnover during a Reactor Startup?
y .-
e Conduct shift turnover at the flight panel as you continue the reactor startup ensuring that you are not distracted from monitoring neutron flux.
B. Do not begin control bank withdrawal. Conduct shift turnover after neutron flux has stabilized.
C. Allow your relief to continue the reactor startup as you relay pertinent watch turnover infonnation to him.
D. Continue the startup while the rest of the crew conducts watch turnover.
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Page 93 of 100
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frulian l'oint linit 2 Consolidated Edi.wn Cornpany of N)*
REACTOR OPERATOR EXAMINATION While perfonning a plant cooldown to 350 F using POP 3.3, Plant Cooldown, you notice a handwritten notation in the right margin of the page that contains the following information TPC 96-153.
Which ONE of the following statements is wrrect regarding the significance of this 88 notation? * *
A. The notation refers to a Temporary Procedure Change. I must refer to the Temporary Prceedure Change Log Book located in the control room to perform the associated step.
B. The notation refers to a Temporary Procedure Change. The step that is affected will be lined out and the correct information entered in the body of the procedure.
C. llandwritten notations are not permitted in plant procedures. I should notify the SRO and obtain a clean copy of the procedure.
D. The notation refers to a Temporary Procedure Change. I must refer to the Temporary Procedure Change Request Form located at the beginning of the procedure to perform the associated step.
. i-t Page 94 of 100
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indian Point Unit 2 Consolidated Edison Cornpany of Nl'
REACTOR OPERATOR EXAMINATION An event has occurred requiring notification of the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Which ONE of '
the following communications systems should be used to perform this notification?
A. Emergen,cy Notification System (ENS) Phone B. Radiological Emergency Conununication System (RECS)
C. Microwave Phone Line D. State Emergency Management Office (SEMO) Radio
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Page 95 of 100 A
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Indian l'oint Unit 2 Consolidated tidison Company of Nl'
REACTOlt OPEllATOlt EXAMINATION Which ONE of the following is the minimum Emergency Plan Classification which would .
require " level 2" staffing?
A. Site Area Emergency
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General Emergency h Alert 3 Notification of Unusual Event l
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Page 96 of 100
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indian 1*oint Unit 2 Consolidated Edison Company ofbT REACTOR OPERATOR EXAMINATION
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Which ONE of the following statements corTectly identifies the tagout protection that must be provided if a worker is going to work in a tank that is connected to any system? -
Each source of energy must be isolated by TWO CLOSED, and TAGGED isolation i8 valves or by ONE CLOSED, LOCKED and TAGGED isolation valve.
B. Each source of energy must be isolated ( if available ) by TWO CLOSED, LOCKED and TAGGED isolation valves.
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\p. Each source of energy must be isolated ( if available ) by TWO CLOSED, LOCKED
\ and TAGGED isolation valves or a safety person must be stationed at the tank opening to assist in emergency egress. -
. Each source of energy must be isolated by at least ONE CLOSED, LOCKED and
= TAGGED isolation valves, or a safety person must be stationed at the tank opening to assist in emergency egress.
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Indian l'oint Unit 2 Consolidated Edison Cornpany ofNY ltEACTOlt OPEllATOlt EXAMINATION
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Which ONE of the following gas samples would indicate that the gas space in the associated equipment contains a potentially fiammable mixture?
A. Volume Control Tank 97% liydrogen,1% Oxygen,2% Nitrogen
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B. 23 CVCS lioldup Tank 26% liydrogen, 0% Oxygen,74% Nitrogen C. 22 Large Gas Decay Tank 7% liydrogen, !?% Oxygen,76% Nitrogen D. Main Generator 94% llydrogen,1% Oxygen,5% Nitrogen
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Page 98 of 100
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. Indian l'oint Unit 2 Consolidated Edison Cornpany ofNY REACTOR OPERATOR EXAMINATION
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Choose the selection that correctly completes the following statement:
The Condensate Storage Tank area is a(n) , the Simulator Building
- is in the and the Main Turbine is in the .
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A. Isolation Zone, Owner Controlled Area, Protected Area
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B. Vital Atea, Isolation Zone, Protected Area
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- C. Exclusion Zone, Owner Con / trolled Area, Protected Area D. Vital Aien, Owner Controlled Area, Protected Area
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Indian l'olnt Unit 2 Consolidated Edison Company ofNl'
REACTOR OPERATOR EXAMINATION
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Which ONE of the following Central Control Room log sheet entry examples would require that the entry be RED circled and explained in the remarks section?
A NON-Technical Specification reading exceeds NORMAL limits specified on log
8# sheet.
. An error is made entering a reading and a correct ion to the entry is required.
C. Reading exceeds MIN / MAX limit specified on log sheet, i Reading is taken one hour late due to startup activities.
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Page 100 of 100
Reactor Operator Exam Answer Key - 3/8/97
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1. C 26. D 51. A 76. A -
2. B 27. B 52. B 77. B 3. C 28. C 53. B 78. D
.,) 4 C .. 29. A 54 B 79. A 5. C 30. C 55. B 80. C 6. A 31. B 56. C 81. B 7. D 32. D 57. A 82. C 8. C 33. B 58. D 83. A
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9. B 34. C 59. B 84. D l
10 D 35. A , 60 C 85. B 11. A 36. D 61. D 86. A 12, D 37. B 62. A 87. C 13. C 38. B 63. C 88. D 14 B 39. D M. D 89. B 15. B 40. C 65. B 90. B 16. C 41. A 66. C 91. D I 7. D 42. B 67. B 92. A 18. B 43. D ex C 93. B 19. A 44. D 69. D 94. D 20, C 45. A 70. A 95. A 21. D 46. D 71. C 96. C 22. B 47. D 72 B 97. B
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23. D 48 C 73. D 98. C 24. B 49. B 74. B 99. D 25. A 50. D 75. B 100. C
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l - ATTACHMENT 2 SIMULATION FACILITY REPORT- ,
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FACILITY LICENSEE: Consolidated Edison FACILITY DOCKET NO: 50 247 l )9 Operating Tests administered: 3/18 1997 This form is to be used only to report observations. These observations do not constitute audit or inspection findings, and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC -
certification or approval of the simulation facility other than to provide information, which may be used in future evaluations. No licensee action is required in response to these observations.
No discrepancies in simulator performance were noted.
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