IR 05000387/1994015
| ML17158A448 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 08/17/1994 |
| From: | Meyer G, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17158A447 | List: |
| References | |
| 50-387-94-15OL, 50-388-94-16OL, NUDOCS 9408290149 | |
| Download: ML17158A448 (93) | |
Text
U. S. NUCLEAR REGULATOR COMMISSION REGION 1 DOCKET/REPORT NOS:
50-387/94-15 50-388/94-16 LICENSEE:
FACILITY:
DATES:
Pennsylvania Power and Light Company Susquehanna Steam Electric Station, Units 1 & 2 Berwick, Pennsylvania August 8-10, 1994 Julian H. Williams, Sr. Operations Engineer CHIEF EXAMINER:
S'~
J an H. Williams, Sr. Operations Engineer Da Section, Division of Reactor Safety APPROVED BY'lenn W. Meyer, Chief BWR & PWR Sections Division of Reactor Safety (h Date 9'408290l49 940818 PDR ADOCK 05000387 V
SUSQUF<3IANNA STEAM 5<LZCTRIC STATION, UNITS 1 &2 l<2AAMINATIONREPORT NOS. 50-337/94-15 AND 50-388/94-16 One examiner administered an initial examination at Susquehanna Steam Electric Station to one senior reactor operator (SR'nstant applicant during the week of August 9, 1994.
Quca>fiick The applicant passed all portions of the examination.
Generic strengths and weaknesses were not identified due to the single applicant.
No problems were experienced in the preparation and administration of the examination, which indicated effective facility management oversight and contro DETAILS 1.0 INTRODUCTION The NRC ad'ministered an initial examination to one senior reactor operator (SROI) instant applicant.
The examination was administered in accordance with NUREG-1021, "Operator Licensing Examiner Standards," Revision 7.
2.0 SUMMARYOF E3DLMINATIONRESULTS AND RELATED FINDINGS 2.1 Examination Results The results of the examination are summarized below:
Written 1 passed and 0 failed Operating 1 passed and 0 failed Overall 1 passed and 0 failed In a letter dated August 11, 1994 (Attachment 2), the facility provided comments and proposed resolutions on five written exam questions.
The NRC resolution (Attachment 3) of these comments accepted the five proposed resolutions, which included the deletion of two questions and answer changes to three questions.
2.2 Facility Generic Strengths and Weaknesses A summary of strengths and weaknesses was not identified due to the single applicant.
No conclusive training program feedback could be derived with the limited amount of observations and examination data.
2.3 Reference Material The reference material met the guidelines of ES-201, Attachment 2, of NUTMEG-1021.
2.4 Preexamination Activities The facility reviewed the written examination on July 27, 1994.
The simulator scenarios and job performance measures (JPMs) were validated on August 8, 1994, on the facility's simulator and in the'plant.
The facility staff, who were involved with these reviews, signed security agreements to ensure that the initial examination was not compromise.5 Management Oversight and Controls The facility reference materials provided for the examination preparation were in an easy to use format.
No problems were experienced in the validation and administration of the examination.
The applicant performed well during the examination.
Based upon these observations, the examiner concluded that management oversight and control was effective.
3.0 EXITMEETING An exit meeting was conducted on August 9, 1994, with Mr. Art Fitch, Operations Training Supervisor.
No problems or concerns were identified.
Attachments:
1.
SRO Examination and Answer Key 2. Facility Comments 3.
NRC Resolution of Facility Comments 4.
Simulator Fidelity Report
ATTACH1VXEPFT 1 SRO EXAlVHNATIONAND ANSWER KEY
HRC Official Use Only Nuclear Regulatory Commission Operator Licensing Examination This document is removed from Official Use Only category on date of examination.
NRC Official Use Only
U. S.
NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE CANDIDATE'S NAME:
FACILITY:
SUS UEHANNA STEAM ELECTRIC STATION REACTOR TYPE:
BWR-QE4 DATE ADMINISTERED: Au st
1 94 INSTRUCTIONS TO CANDIDATE:
Use the answer sheets provided to document your answers.
Staple this cover sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires a final grade of at least 80:.
Examination papers will be picked up four (4) hours after the examination starts.
TEST VALUE CANDIDATE'S SCORE FINAL GRADE o
All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature
'ENIOR REACTOR OPERATOR A N S W E R SHEET Page
Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
MULTIPLE CHOICE 001 a
b c
d 002 a
b c
d 003 a
b c
d 004 a
b c
d 005 a
b c
d 009 a
b c
a b
c d
d 006 a
b c
d f
007 a
b c
d 008 a
b c
d 031 a
b c
032 a
b c
d d
033 a
b c
d 023 a
b c
d 024 a
b c
d 025 a
b c
d 026 a
b c
d 027 a
b c
d 028 a
b c
d 029 a
b c
d 030 a
b c
d 012 a
b c
013 a
b c
d d
011 a
b c
d 034 a
b c
035 a
b c
036 a
b c
d d
d 014 a
b c
d 015 a
b c
d 016 a
b c
017 a
b c
018 a
b c
d 019 a
b c
d 022 a
b c
d 020 a
b c
d 021 a
b c
d 037 a
b c
d 038 a
b c
039 a
b c
d d
040 a
b c
d 041 a
b c
d 042 a
b c
d 043 a
b c
d 044 a
b c
d 045 a
b c
d
SENIOR REACTOR OPERATOR ANSWER SHEET Page
Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
046 a
b c
d 047 a
b c
d 048 a
b c
d 049 a
b c
d 050.
a b
c d
051 i
052 a
b c
d a
b c
d 053 a
b c
d 054 a
b c
d 055 a
b a
b 057 a
b c
d c
d c
d 058 a
b c
d 059 a
b c
d 060 a
b c
d 061 a
b c
d 062 a."
b c
d 063 a
b c
d 064 a
b c
d 065 a
b c
d 066 a
b c
d 067 a
b c
d
a b
c d
069 a
b c
d 070 a
b c
d 071 a
b c
d 072 a
b c
d 073 a
b c
d 074 a
b c
d 075 a
b c
d 076 a
b c
d 077 a
b c
d 078 a
b c
d 079 a
b c
d 080 a
b c
d 081 a
b c
d 082 a
b c
d 083 a
b c
d 084 a
b c
d 085 a
b c
d 086 a
b c
d 087 a
b c
d 088 a
b c
d 089 a
b c
d 090 a
b c
d 091 a
b c
d
'ENIOR REACTOR OPERATOR A N S W E R S
H E E T Page Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
092 a
b c
d 093 a
b c
d 094 a
b c
d 095 a
b c
d 096 a
b c
d 097 a
b c
d 098 a
b c
d 099 a
b c
d 100 a
b c
d (**********
END OF EXAMINATION**********)
Page
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS ing the administration of this examination the following rules apply:
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1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.
This must be done after you complete the examination.
3.
Restroom trips are to be limited and only one applicant at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
4.
.Use black ink or dark pencil ONLY to facilitate legible reproductions.
5.
Print your name in the blank provided in the upper right-hand corner of the. examination cover sheet and each answer sheet.
6.
Mark your answers on the answer sheet provided.
USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
7.
The point value for each question is indicated in parentheses after the question.
If the intent of a question is unclear, ask questions of the examiner
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only.
9.
When turning in your examination, assemble the completed examination wit:h examination questions, examination aids and answer sheets.
In addition, turn in all scrap paper.
10.
Ensure all information you wish to have evaluated as part of your answer is on your answer sheet.
Scrap paper will be disposed of immediately following the examination.
11.
To pass the examination, you must achieve a grade of 80~ or greater.
12.
There is a time limit of four (4) hours for completion of the examination 13.
When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may b~
denied or revoke l
'ENIOR REACTOR OPERATOR Page
STION:
001 (1.00)
The RCIC system was operating for a surveillance test when the
"RCIC STEAM LINE LOGIC A HI DIFF PRESS" and the
"RCIC STEAM LINE LOGIC B HI DIFF PRESS" annunciators came in.
The RCIC system did not isolate as expected.
The PCO was unable to shut the inboard and outboard isolation valves (F007 and F008)
so he shuts the RCIC Steam Supply Valve (F045).
SELECT the proper emergency classification, if any.
a. Unusual Event b. Alert c. Site Emergency d.
None Applicable QUESTION:
002 (1. 00)
Unit 1 was operating at 100. power when a large break LOCA occurred.
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eactor water level decreased to approximately-170 inches and has not t been restored five minutes after the event occurred.
The operator performing ON-159-002,
"Containment Isolation," reports that the drywell nitrogen makeup inboard and outboard isolation valves are stuck open.
SELECT the proper emergency classification.
a.
Unusual Event b. Alert c. Site Emergency d. General Emergency
SENIOR REACTOR OPERATOR Page
STION: 003 (1.00)
If the APRM Gain Adjustment Factor (AGAF) is 0.99...
(Choose One)
a.
a non-conservative condition. exists b. the APRM gain should be adjusted c. the indicated thermal power is greater than the actual thermal power d. the indicated thermal power should be compared with the power determined by a heat balance QUESTION:
004 (1.00)
The
"HPCI OUT OF SERVICE" annunciator is lit.
WHICH ONE (1) of the following conditions would NOT activate this alarm?
a.
HPCI Inverter Power Failure b.
HPCI Auxiliary Oil Pump Motor Overload c. Valve F066,
"Steam Exhaust to Suppression Pool," not full open d. Valve F004,
"CST Suction," control switch in CLOSE position
i SENIOR REACTOR OPERATOR Page
ESTION:
005 (1.00)
Unit 1 is cooling down following a reactor scram.
Reactor pressure is 350 psig and decreasing.
Both RHR systems are in the Suppression Pool Cooling Mode providing maximum cooling.
A spurious Division 1 LPCI initiation signal is received.
Which ONE (1) of the following correctly describes the lineup of the RHR systems following the spurious signal?
a.
F015A (RHR Injection)
open F028A (Suppression Chamber Spray Test Shutoff)
open b.
F015A (RHR Injection)
open F024A (Test Return)
closed F028A (Suppression Chamber Spray Test Shutoff) closed c.
F015B (RHR Injection)
open F024B (Test Return)
closed d.
FO15B (RHR Injection)
open FO24A (Test Return)
closed F048B (RHR Heat Exchanger Shell Bypass)
closed STION:
006 (1.00)
Which ONE (1) of the following conditions will cause all SRM ROD BLOCKS to be bypassed?
a. All IRMs on range 3 or above b.
SRMs read greater than 100 counts c. All IRMs are on range 8 or above d.
SRMs read greater than 2 E+5 counts
'SENIOR REACTOR OPERATOR Page
STION:
007 (1.00)
The operator has selected a "Ten Minute History" for water level on the Safety Parameter Display System.
The plot is indicated in white.
The status of the water level information is:
a. indeterminate b. in the safe range c. in the caution range d. in the danger range QUESTION:
008 (1.00)
The reactor is at 5: power and the Mode Selector, Switch is in STARTUP.
Nuclear instrumentation readings are as follows:
APRM Channel A
PRM Channel E
RM Channel F
RM Channel A
IRM Channel G
IRM Channel H
5'o 4~o 105'o 106o 125'o Which ONE (1) of the following correctly describes the automatic actions that should occur?
a.
Rod block but no half scram b. Half scram but no rod block c.
Rod block and half scram d.
No rod block and no half scram
'ENIOR REACTOR OPERATOR Page
ESTION:
009 (1.00)
During normal power operations, the temperature of the downstream piping of an SRV should be approximately In the event of an SRV leaking past its closed seat, the temperature will increase.
In this case, an alarm will actuate at a downstream piping temperature of Which of the following sets of temperatures correctly completes these statements?
a.
85 degrees F,
225 degrees F
b.
135 degrees F,
225 degrees F
c.
85 degrees F,
250 degrees F
d.
135 degrees F,
250 degrees F
QUESTION:
010 (1.00)
actor pressure has been reduced to 470 psig and has been held at this pressure for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Determine the lowest reactor pressure that the reactor can be reduced to over the next 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without exceeding the maximum technical specification allowable cooldown rate.
CHOOSE ONE a,
136 psig b.
146 psig c.
156 psig d.
166 psig
SENIOR REACTOR OPERATOR Page
ESTION:
011 (1. 00)
A common mode failure causes a loss of all 125 VDC busses.
Which of the following functions is NOT disabled due to the loss of 125 VDC?
a. Reactor Feed Pump trip b. Safety Relief Valve ADS function c. Safety Parameter Display System d.
HPCI automatic initiation QUESTION:
012 (1. 00)
Unit 1 is operating at 85: power when an unexpected decrease in reactor water level occurs with the feedwater level control system in automatic three element control mode.
Which of the following failures will cause this event?
a.
one feedwater flow indicator failed downscale b. total steam flow summer failed upscale c.
one feedwater flow indicator failed upscale d. total feedwater flow summer failed downscale
SENATOR REACTOR OPERATOR Page
ESTION:
013 (1. 00)
While operating Unit 2 at 50% power (Steady state),
an operator increases the controlling EHC pressure regulator setpoint 1 psi.
Which of the following is the expected behavior of the reactor power during this event?
a. power increases and stabilizes at a slightly higher value b. power increases and then returns to its original value c. power decreases and stabilizes at a slightly lower value d. power decreases and then returns to its original value QUESTION:
014 (1.00)
E0-100-113, Level/Power Control, is in progress under ATWS conditions with boron injecting into the reactor.
Which of the following is the primary reason for inhibiting ADS under these conditions?
a. Prevents injection systems from injecting and diluting the boron concentration b. Prevents thermal stresses on the reactor vessel by an uncontrolled cooldown c. Prevents unnecessary suppression pool water temperature increase d. Prevents injecting large volumes of cold water which may result in substantial positive reactivity addition
'ENIOR REACTOR OPERATOR Page
ESTION:
015 (1. 00)
Which of the following describes the conditions which will allow opening the MSIVs with condenser vacuum at 2.5 inches Hg if the Low Vacuum Bypass Switches are in BYPASS?
a.
Mode switch in RUN Main turbine stop valves closed Main turbine control valves open b.
Mode switch in RUN Main turbine stop valves closed Main turbine control valves closed c.
Mode switch in Startup Main turbine control valves closed Main turbine stop valves open d.
Mode switch in Startup Main turbine stop valves closed Main turbine control valves open ESTION:
016 (1. 00)
A manual scram has been inserted on both RPS channels and the rods have failed to insert.
The full core display has transferred to full in/full out display mode.
Which of the following is the cause for the failure to scram?
a. hydraulic lock in the SDV b. failure of the scram valves to open
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c. blockage in the scram air header
.d; faiiur4 of one RPS scram trip system
SENIOR REACTOR OPERATOR Page
ESTION:
017 (1.00)
Which of the following is the Unit 1 Technical Specification Bases for maintaining the minimum water level in the spent fuel storage pool?
a.
To keep the spent fuel cooled b.
To provide adequate shielding for personnel working in the area c.
To remove iodine gap activity following a fuel rupture accident d.
To maintain the pool at a low operating temperature QUESTION:
018 (1. 00)
Which of the following conditions requires entry into "Secondary Containment Control" ?
a.
Zone 1 HVAC fails to start for a surveillance test e
b.
Zone
HVAC exhaust radiation is 2.8 mr/hr c.
Zone 1 differential pressure indicates
+0.23 inches water d.
The HPCI equipment area radiation is two times its setpoint
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'ENIOR REACTOR OPERATOR Page
STION:
019 (1. 00)
Unit 1 is shutting down with reactor temperature at 250 degrees F and decreasing.
Both loops of RHR have been placed in shutdown cooling.
An inadvertent closure of the shutdown cooling suction valves has occurred.
IDENTIFY the minimum level in the RPV that ON-149-001, Loss of Shutdown Cooling, requires to be maintained in order to establish natural circulation.
a.
13 inches b.
27 inches c.
45 inches d.
60 inches QUESTION:
020 (1. 00)
Which of the following must be completed by a licensed operator to
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aintain his license in an "active status" per the requirements of 10CFR 5.53,
"Conditions of Licenses" ?
The operator shall actively perform the functions of a licensed operator for a minimum of:
a.
two 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts or one 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift per month b.
seven 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts or five 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts per calendar quarter c.
one 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift or one 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift per month d. five 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts or four 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts per calendar quarter
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'ENIOR REACTOR OPERATOR Page
ESTION:
021 (1.00)
On-118-001,"Loss of Instrument Air", requires that the reactor be scrammed at a specific air pressure.
The reactor is scrammed at this pressure because (CHOOSE ONE)
a.
a loss of reactor instrumentation is imminent b. scram valves could be opening prior to the scram discharge volume vent and drain valves closing, resulting in a release to secondary containment c. control rod drive mechanisms could be damaged due to the outlet scram valve opening prior to the inlet scram valve d. inleakage to the scram discharge volume from drifting open scram valves could preclude a successful scram QUESTION:
022 (1.00)
valve equipped with remote indication is required to be verified in e closed position.
The valve is currently showing dual indication.
DENTIFY how the valve is to be verified in accordance with OP-AD-001,
"Operations Shift Policies and Work Practice".
a. Locally attempt to further close the valve manually.
If it will not close any further, then the valve is considered verified closed.
b. Locally operate the valve in the open direction and then the closed direction.
If the valve opened and. then closed then it is verified closed.
c. If the remote dual indication cannot be changed.to'losed indication the valve CANNOT be verified closed.
d. Remotely attempt to close the valve.. If dual indication remains, consider the valve closed and have the electrical department correct the indication proble 'ENIOR REACTOR OPERATOR Page
ESTION:
023 (1.00)
A licensed reactor operator has worked the following schedule during a refueling outage:
Thursday-scheduled day off Friday-7am to 7pm Saturday-7am to 7pm Sunday-7am to 3pm Monday-7am to 3pm Tuesday-7am to 3pm Wednesday-7am to 9pm Which of the following work schedules is the maximum allowed for the following Thursday without additional authorization in accordance with OP-AD-001, Operations Shift Policies and Work Practices?
a.
4am to Sam b.
4am to 12 (noon)
c.
7am to 3pm d.
7am to 5pm QUESTION:
024 (1.00)
allows "reasonable action that departs from a license condition or a technical specification in an emergency when this action is immediately needed to protect the public health and safety..."
These actions must be approved prior to being taken.
SELECT the minimum approval level required per OP-AD-001, Operations Shift Policies and Work practices.
a. VP-Nuclear Operations b. Manager-Nuclear Operations
.
c. Shift Supervisor d. Licensed Senior Reactor Operator
'ENIOR REACTOR OPERATOR Page
ESTION:
025 (1.00)
Which of the following will result in the lowest ALARA exposure?
a.
one individual performing a job in a 60 mr/hr gamma field for 60 minutes b.
one individual installing temporary shielding in a 60 mr/hr gamma field for 30 minutes and then performing the job in a
mr/hr gamma field for 60 minutes c.
two individuals performing a job in a 60 mr/hr field for 35 minutes d. two individuals installing temporary shielding in a 60 mr/hr gamma field for 15 minutes and then both individuals performing the job in a 6 mr/hr gamma field for 40 minutes QUESTION:
026 (1. 00)
n accordance with OP-AD-001, Operations Shift Policies and Work ractices, if an electrical breaker trips, at tempt (s) at eclosing the breaker is/(are)
allowed before electrical maintenance is called to investigate.
CHOOSE ONE (1):
a.
b.
C.
d.
'SENIOR REACTOR OPERATOR Page
STION:
027 (1. 00)
Determine the correct value for total core flow indication from the following data:
Loop 'A'otal jet pump flow Loop 'B'otal jet pump flow
'A'ecirculation pump speed
'B'ecirculation pump speed Loop 'A'ischarge valve Loop 'B'ischarge valve CHOOSE ONE a.
35 E+6 ibm/hr b.
40 E+6 ibm/hr c.
45 E+6 ibm/hr d.
50 E+6 ibm/hr 45 E+6 ibm/hr 5 E+6 ibm/hr 35%
O~o open open
/
ESTION:
028 (1.00)
OP-AD-001, Operations Shift Policies and Work Practices, indicates that some procedure types require step by step (in sequence)
compliance whereas other types do not.
Which of the following does NOT require step by step (in sequence)
compliance?
a. Surveillance Operations (SO)
b. Operations Procedures (OP)
c.
Emergency Support (ES)
d. Alarm Response (AR)
~g
'ENIOR REACTOR OPERATOR Page
ESTION:
029 (1.00)
Which statement below describes the response of the scram discharge volume valves to a half scramP a. Both dump valves do not change position.
Both sets of vent and drain valves remain open.
b.
One dump valve regositions, but both sets of vent and drain valves remain open.
c.
One dump valve repositions causing one set of vent and drain valves to close.
d. Both dump valves reposition causing both sets of vent and drain valves to close.
QUESTION:
030 (1.00)
A Rod Worth Minimizer rod group has insert and withdrawal limits of otch 12 and notch 24.
The alternate limits for this group are:
a.
14 and 26 b.
14 and 22 c.
10 and 26 d.
10 and
SENIOR REACTOR OPERATOR Page
ESTION:
031 (1.00)
The Rod Worth Minimizer Low Power Alarm Point is sensed by:
(CHOOSE ONE)
a.
The reference APRM b. Total steam flow c. Total feed flow d. Turbine first stage pressure QUESTION:
032 (1. 00)
During the blowdown phase of a Design Basis LOCA, one pair of suppression chamber to drywell vacuum breakers fail open.
Which of the following describes the effect this would have on the peak primary containment pressures?
a. increase suppression chamber and drywell pressure b. increase suppression chamber pressure with no effect on drywell pressure c. increase drywell pressure with no effect on suppression chamber pressure d. increase the time drywell pressure remains at its peak pressure with no effect on suppression chamber pressure
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'ENIOR REACTOR OPERATOR Page
STION:
033 (1.00)
The Reactor Core Isolation Cooling (RCIC) system initiated at -30 inches and raised level to +54 inches.
IDENTIFY the response of the RCIC to the high level and subsequent level decrease to -30 inches.
a.
RCIC turbine trips at level 8 and must be manually reset to allow the turbine to restart at -30 inches to raise level.
b.
Steam supply valve (F045) to the turbine will close at Level
and the high level seal-,in must be manually reset to allow the F045 valve to reopen at -30 inches.
c.
The RCIC turbine governor valve closes at Level 8 and the governor valve will automatically reset restarting the turbine at -30 inches.
d.
Steam supply valve (F045) to the turbine closes at Level 8 and the high level seal-in is automatically reset to allow the F045 valve to reopen to restart the turbine at -30 inches.
(1. 00)
Which of the following would be a direct result of excessive
"steam carryunder" during 100% power operation?
a. recirculation pump NPSH decreases b. main turbine efficiency decreases I
c.
steam quality exiting the reactor decreases d. steam line radiation levels increase
L SENIOR REACTOR OPERATOR Page
STION:
035 (1. 00)
Unit 1 is crating at 100% power.
Diesel generator 'A's out of service for xtensive repair and has been replaced with diesel generator
'E'.
The fol wing equipment is out of service:
480 VAC loa center 1B280 480 VAC load enter 1B210 480 VAC motor ontrol center OB516 480 VAC motor c trol center 1B227 SELECT the statement th t is in accordance with technical specifications from those listed below:
a. Re-energize the equ'ent within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least Hot Shutdown within the xt 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Cold. Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b. Re-energize the equipment within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least Hot Shutdown within the next 1) hours and Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c. Re-energize the equipment with 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least Hot Shutdown within the next 12 hou and Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d. Re-energize the equipment or be in t least Hot Shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdo within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SENIOR REACTOR OPERATOR Page
ESTION:
036 (1.00)
Loop 'B'f RHR is operating in Shutdown Cooling mode.
Water level decreases from +45 to -130 inches.
The RHR injection valve (F015B)
closes due to the shutdown cooling isolation signal.
Which of the following describes the response of the RHR injection valve (F015B) to a LOCA signal at -129 inches?
a.
F015B opens when LOCA signal is received b.
F015B opens after the RHR pump suction valves realigned to the suppression pool c.
F015B opens only by manually opening the valve d.
F015B opens after the operator manually resets the F015B shutdown cooling reset QUESTION:
037 (1.00)
Unit 1 was operating at 75% power and 90: rod line when both
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~
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~
ecirculation pumps trip.
Power reduces to 46:.
The operator is equired to
(CHOOSE ONE)
a. place mode switch in Shutdown b. drive rods until power is below the 80: rod line c.
commence a normal shutdown d. restart a recirculation pump
SENIOR REACTOR OPERATOR Page
ESTION:
038 (1. 00)
The recirculation pumps are operating in master manual control.
Pump speed has been increased using the master controller.
No adjustments have been made to the individual M/A transfer stations.
The operator inadvertently depresses the manual pushbutton for the 'B'ecirculation pump.
The 'B'ecirculation pump will: (CHOOSE ONE)
a. return to the speed it was prior to being transferred to the master controller at the maximum rate of scoop tube movement b. return to the speed it was prior to being transferred to the master controller at a rate limited by the M/A transfer station c. remain at the speed set by the master controller d. return to the speed it was prior to being transferred to the master controller at the maximum rate allowed by the Error Limiting Network QUESTION:
039 (1. 00)
e general caution in the Emergency Operating Procedures list a drywell temperature and level indication for the upset, shutdown, and extended wide range level instruments.
If drywell temperature is greater than the value stated, then level indication is not valid below the value stated for that instrument.
The level is not valid because:
(CHOOSE ONE)
a. at low reactor pressures the variable leg will flash at temperatures greater than the stated drywell temperature b. at high drywell temperatures the variable leg density will decrease causing invalid readings c. at low indicated levels reference leg density causes on scale indications with level below the instrument's monitoring range d. at low drywell temperatures the reference leg will cause erronously high indicated levels
SENIOR REACTOR OPERATOR Page
ESTION:
040 (1.00)
Unit 1 is operating at 100% power and the "Off Gas Recombiner Panel OC673 System Trouble" alarm occurs.
Investigation reveals an "Offgas Unit 1 isolated" alarm.
Which of the following statements describes the plant response if no operator, action is taken?
a.
There will be no major effect on plant operations at 100% power.
b. Main condenser vacuum will decrease but not to the point where a
main turbine trip will occur.
c. Main condenser vacuum will slowly decrease, taking about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to reach the main turbine trip setpoint.
d. Main condenser vacuum will decrease at a fairly rapid rate, reaching the main turbine trip setpoint in less than 10 minutes.
QUESTION:
041 (1.00)
Unit 1 was at 100 'ower when the 'C'eactor feed pump trips due to losure of the turbine exhaust valve.
Which of the following is the expected plant response if no operator action is taken?
a.
Feed flow decreases, reactor water level decreases, A and B feed pumps return the reactor water level to normal, reactor power returns to 100%
b.
Feed flow decreases, reactor water level decreases, reactor recirculation is unchanged, reactor scrams on low water level, turbine trips, reactor water level returns to normal.
c.
Feed flow remains constant as A and B feed pumps pick up the load, reactor power remains at 100%.
d.
Feed flow decreases, reactor water level decreases, recirculation runs back, reactor power decreases, reactor water level returns to nornal, reactor power stabilizes at about 70%
'ENIOR REACTOR OPERATOR Page
ESTION:
042 (1.00)
Which of the following loads is NOT supplied by the Main Steam system?
a.
steam jet air ejector b. turbine building heating coils c. reactor feed pump turbine d. main condenser heating coils QUESTION:
043 (1.00)
Which of the following signals will NOT isolate the MSIVs?
a.
MSL high flow b.
MSL high radiation c.
MSL high pressure d. condenser vacuum low
SENIOR REACTOR OPERATOR Page
ESTION:
044 (1.00)
A controller failure on the Unit 1 GRRCCW heat exchanger has caused the associated temperature control valve (TV-10938) to fail closed.
Which of the following is the expected system response'P a.
An 'Isolate Offgas'ignal will be processed due to low condenser flows.
b. A 'Recombiner Shutdown'ignal will be processed due to low condenser flows.
c.
An 'Isolate Offgas'ignal will be processed on a high recombiner temperature.
d. A 'Recombiner Shutdown'ignal will be processed on a high recombiner condenser temperature.
QUESTION:
045 (1.00)
he "auto transfer permissive switch" must be 'on'o permit auto losure of:
I.
Tie Bus Bkr OA10S02 II.
Aux Bus Bkr 1A10104 III. Aux Bus Bkr 1A10204 IV. Alt Fdr Bkr on 4 KV ESS Bus CHOOSE ONE a. I and II b. I and IV c. II and III d. IV
'ENIOR REACTOR OPERATOR Page
ESTION:
046 (1.00)
An RHR Division I LOCA signal is present concurrently with a main generator primary and backup lockout relay trip.
Identify which equipment from the following list which will be load shed from associated bus if previously in service.
I.
Circulating water pumps II.
Turbine Building Chiller A III. ESW pump C
IV.
Service water pumps CHOOSE ONE a. I and IV b. II only c. II and III d. III only ESTION:
047 (1. 00)
During plant operator rounds it is noticed that both bulbs in the ground detection circuit on 1D652 are much dimmer than normal but of the same brightness.
Which statement below describes system condition?
a.
There is no ground indicated by this condition b.
A ground exists on both the positive and negative sides c.
A ground exists only on the positive side d.
A ground exists only on the negative side
SENIOR REACTOR OPERATOR Page
ESTION:
048 (1. 00)
While Unit 1 is at 20: power, a leak develops on the EHC piping.
EHC header pressure is decreasing and cannot be recovered.
Which of the following describes the plant response?
Assume no operator action.
a. Turbine trip; reactor does not scram; reactor pressure is controlled with bypass valves b. Turbine control and stop valves drift shut; reactor scrams on turbine stop valve closure or turbine control valve fast closure c. Turbine trip and reactor scram on high pressure d. Turbine control valves and stop valves drift shut; reactor scrams on high pressure QUESTION:
049 (1.00)
With the reactor operating at 20 'ower, which one of the following describes the result if the output of the low value gate from the EHC ressure regulator fails high?
a. backup regulator controls reactor pressure at higher value b. load limit maintains total steam flow at 100-.
c. reactor scram on MSIV closure d. turbine trip caused by turbine control valve closure
SENIOR REACTOR OPERATOR Page 3r ESTION:
050 (1. 00)
A LOCA has occurred, but no low pressure ECCS pumps are running.
The ADS timer has timed out. If you now start an RHR pump the ADS will...
(CHOOSE ONE)
a. actuate b. actuate when the 102 second timer times out again c. not actuate unless the 102 second timer is reset d. not actuate until a full loop of RHR is running QUESTION:
051 (1.00)
Core spray pumps 'A'nd 'C'n Unit 1 were started using the manual initiation pushbutton for testing purposes.
If Unit 2 experiences a
LOCA, which of the following describes the automatic actions that occur in the Unit 2 Core Spray system?
a.
Core Spray pumps '2A', '2B', '2C'nd '2D'tart after a 15 second time delay b.
Core Spray pumps '2A'nd '2C'tart after a 25 second time delay and pumps '2B'nd '2D'tart 15 seconds later c.
Core Spray pumps '2C'nd '2D'tart after a 15 second time delay and pumps '2A'nd '2B'o not start d.
Core Spray pumps '2A'nd '2C'o not start and pumps '2B'nd
'2D'tart after a 15 second time delay
'ENIOR REACTOR OPERATOR Page
I STION:
052 (1.00)
The RHRSW crosstie valves HV-112-F075 and HV-112-F073 are used in emergency conditions to:
I.II.
III.
IV.
spray containment cool the alternate unit's heat exchanger flood the reactor vessel crosstie RHRSW B loop to the A loop of RHRSW CHOOSE ONE a. I and III only b. I, II and III c. II and IV d. III and IV UESTION:
053 (1.00)
oncerning the RHR inboard and out board injection valves F015 and F017, which statement below is correct?
a. Both F015 and F017 are powered from the swing buses.
b. Both valves cannot be opened simultaneously with RPV pressure at 400 psig.
c. If both F015A and F017B valves are closed, an initiation signal in Div 1 caused by drywell pressure above 1.72 psig and RPV pressure less than 436 psig, both valves will open.
d.
F015 will remain interlocked open with a LOCA initiation signal presen ~ ~
'ENIOR REACTOR OPERATOR Page
ESTION:
054 (1.00)
In the RHR fuel pool cooling assist mode, the RHR pump minimum flow valve F007 is closed and the appropriate leads are lifted in order to (CHOOSE ONE)
a. protect the RHR pump motor from over heating b. prevent draining the RPV to the suppression pool c. prevent pumping fuel pool water to the suppression pool d. prevent pumping suppression pool water to the fuel pool QUESTION:
055 (1.00)
If the HPCI system is in standby readiness lineup, which of the following describes the correct lineup?
a.
F002 and F003 steam supply valves open, F001 steam admission valve open, the stop valve and control valve closed b.
F002 and F003 steam supply valves open, F001 steam admission valve closed, the stop valve and control valve closed c.
F002 and F003 steam supply valves open, F001 steam admission valve closed, the stop valve and control valve open d.
F002 and F003 steam supply valves closed, F001 steam admission valve closed, the stop valve and control valve closed
SENIOR REACTOR OPERATOR Page
ESTION:
056 (1.00)
Unit 1 is at 100% power when a loss of AC power occurs (station blackout).
HPCI receives a low reactor water level initiation signal.
The HPCI Auxiliary Oil Pump breaker trips during the auto start attempt.
Which statemnet below describes how the HPCI system responds to this condition?
a.
HPCI will not start since the stop and control valves require oil pressure to open.
b.
HPCI will start and run at low speed due to the governor requiring aux oil pump oil pressure to control HPCI turbine speed.
c.
HPCI will start and will probably experience bearing damage if operator action is not taken within 2 minutes.
d.
HPCI will start since the shaft driven oil pump will supply the lubrication and hydraulic needs.
ESTION:
057 (1.00)
Which of the following statements describes the response of the RCIC system when the RCIC Topaz inverter is deenergized when an automatic initiation signal is received?
a.
RCIC will run because all AC powered valves are in their proper position for injection.
b.
RCIC will"trip on overspeed during startup due to a loss of speed control.
c.
RCIC will not run because the governor valve fails closed without power.
d.
RCIC will start and remain at low rpm until the operator takes manual contro 'ENIOR REACTOR OPERATOR Page
I ESTION:
058 (1.00)
If an ARM has an auxiliary unit with a local audible alarm, where should this alarm be silenced?
(CHOOSE ONE)
a. At the auxiliary unit associated with the alarming ARM b. At the ARM Indicator and Trip Unit in the upper relay room c.
From the main control room d. At the doorway of the affected area QUESTION:
059 (1.00)
One method used to identify excess drywell leakage is the stroking and/or backseating of valves in the containment.
Which statement concerning this practice is NOT correct?
a. Stroking a valve could change the packing configuration so that the leakage increases significantly.
b. Backseating a valve can result in the calculation of the unidentified leak rate being lower even though the total leak rate remains the same.
c. Primary Containment Isolation valves are operable as long as they are not backseated moxe than 30 minutes.
d. Stroking a valve may isolate the packing from the process pressure, therby reducing leakage rat ~
SENIOR REACTOR OPERATOR Page
STION:
060 (1. 00)
The PCO doing his rounds reports that the 'B'ontainment Radiation Monitor Iodine channel failed downscale three hours ago.
The 'A'RM system is INOP due to a sample line moisture problem.
The plant is in Condition 1.
Which of the following statements is true?
a.
An LCO should be entered because the Iodine and Noble Gas channels are required to be operable.
b. An LCO should be entered because the Iodine and Particulate channels are required to be operable.
c. An LCO should be entered because no Iodine channel is operable.
d.
No LCO should be entered because neither Iodine channel is required to be operable.
QUESTION:
061 (1. 00)
An MSIV isolation has occurred and all conditions for reset are met.
pon pushing the N4S Reset pushbuttons on 1C601, which of the following ill occur'?
a.
N4S logic A, B, C and D white lights will energize b.
SBGT Fan(s) will stop c.
ADS gas supply swaps back to CIG d.
CREOASS fan stops
'ENIOR REACTOR OPERATOR Page
STION:
062 (1.00)
Which statement below describes how a high temperature (265 degrees F)
on the SBGT filter train 'A'ffects the 'A'rain when both trains are in "auto lead" but neither is running?
a.
The fire protection deluge system auto actuates on the 'A'rain to extinguish a possible fire.
b. Dampers open to allow cooling via natural convection.
c.
The 'A'rain fan auto starts and the makeup outside air dampers close to ensure a cooling flowpath.
d.
The 'A'rain fan auto starts and the crosstie damper and the cooling outside air inlet dampers auto close.
QUESTION:
063 (1.00)
A Zone 1 isolation has just occurred with the unit at 100: power.
The main steam tunnel coolers will: (CHOOSE ONE)
a.
remain running b. trip and stay tripped c. start immediately d. trip and auto re-start after 60 seconds
~ SENIOR REACTOR OPERATOR Page
STION:
064 (1.00)
A reactor startup is in progress following a refueling outage.
'A'ode switch is taken out of operate inadvertently.
A full reactor scram occurs.
Which of the statements below could be the cause of the scram?
a.
IRM 'A's also inop b.
APRM 'B's failed downscale c. Shorting links are not installed d.
SRM 'D's reading 2 E+5 cps QUESTION:
065 (1.00)
Which of the following is an accurate statement concerning the scan mode of the Reactor Manual Control System?
a.
Samples the activity control cards for information concerning the status of all control rods.
b. Controls the directional control valves and returns information concerning the present state of the HCU.
c.
Samples all HCUs for information concerning the status of scram valves and accumulators.
d. Works in conjunction with various control rod blocks initiating systems to determine when a control rod motion block is require " SENIOR REACTOR OPERATOR Page
ESTION:
066 (1.00)
Regarding ARI initiation logic, which of the following system responses is expected if a reactor high pressure trip unit relay remains de-energized during an ATWS?
a. ARI will be able to be actuated by the 4 RPV level 2 trip inputs b. ARI cannot auto actuate c. ARI will be able to be actuated by the 3 remaining high pressure trip inputs d. ARI cannot be manually actuated QUESTION:
067 (1.00)
The letdown mode of RWCU can be lined up to LRW or the main condenser.'hy is it prohibited from lining up to both simultaneously?
(CHOOSE ONE)
a. isolation valves are interlocked preventing simultaneous opening b. excess flow could damage the resin c. potential to siphon LRW to the condenser d. potential to lose condenser vacuum QUESTION:
068 (1.00)
While trending RPV pressure on SPDS, the secondary CRT fails.
What actions permit viewing the RPV trend data on SPDS?
(CHOOSE ONE)
a. Depress the CRT TRANSF pushbutton b. Depress the CRT FAIL pushbutton c. Depress the SYS TROUBLE pushbutton d. Depress the PANEL RESET pushbutton
" SENIOR REACTOR OPERATOR Page
STION:
069 (1.00)
Who may grant permission for the Refueling Platform to be left unattended when items are tied off over the side?
(CHOOSE ONE)
a. Unit Supervisor b. Refuel Floor SRO c. Refuel Floor Manager d.
Ops Outage Supervisor QUESTION:
070 (1.00)
The plant is operating at 100% power.
The RADWASTE EFFLUENT MON HI RADIATION annunciator alarms.
Abnormally high radiation trends are observed on the Effluent Radiation Recorder RR-06433.
The Chem Tech reports that there has been an offsite release of:
8.60 E+6 micro Ci/min of Noble Gas 1.23 E+3 micro Ci/min of Iodine-131 7.70 E=3 micro Ci/min of particulates EPB dose rate projected to exceed 500 mR/hr in one hour Which of the following is the proper emergency classification?
a.
Unusual Event b. Alert c. Site Emergency d. General Emergency
SENIOR REACTOR OPERATOR Page
STION:
071 (1. 00)
An emergency D/G start signal is developed when all breakers to a 4 KV bus open simultaneously.
Choose the condition(s) that open all three feeder breakers.
a. Associated ESS transformer lockout b.
Bus lockout c.
65'. bus UV for more than 10 seconds d. Division I core spray LOCA signal QUESTION:
072 (1.00)
The station has experienced a Station Blackout and no AC power has been available for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
After a D/G is started, which of the following need NOT be performed in a timely manner to prevent overheating of the D/G?
a. close the D/G breaker b. ensure diesel room vent fans start after 2 minutes time delay c. ensure ESW is aligned d. ensure UPS aligned-breaker closed
SENIOR REACTOR OPERATOR Page
STION:
073 (1.00)
When the Transfer Switches at the Unit 1 Remote Shutdown Panel are in the EMERG position which of the following statements is NOT correct?
a.
The auto initiation capability for the 'A'oop of RHR is inoperable.
b.
The NSSSS isolation signals to RHR valves 1F008 and 1F009 are defeated.
c.
The high RPV pressure signals to RHR valve 1F008 and 1F009 are operable.
d.
The relief mode of SRVs 'A', 'B'nd 'C's inoperable.
QUESTION:
074 (1.00)
Which of the following statements describes the bases for opening HPCI TEST LINE ISO VLV HV-155-F011 prior to leaving the control room per ON-100-009, Control Room Evacuation?
a.
The valve is opened to provide a flow path for HPCI.
b.
The valve is opened to ensure that keepfill protection is available.
c.
The valve is opened to ensure that the RCIC piping remains filled and vented.
d.
The valve is opened so that RCIC can be cycled between vessel injection and full flow test for increased vessel heat remova 'ENIOR REACTOR OPERATOR Page
ESTION:
075 (1. 00)
Unit 1 experienced a reactor scram on low RPV level due to loss of all reactor feed pumps.
ON-100-101,"Scram",
and EO-100-102,
"RPV Control",
were entered.
The 'A'eed pump was restarted and level decrease was stopped at -50 inches.
When level was at 20 inches and increasing, the
'A'eed pump tripped.
Level decreased and is now at 0 inches.
SELECT the statement regarding use of the EOPs in this situation.
a. continue through E0-100-102, and use other systems to restore level b. exit EO-100-102 and continue in ON-100-101 c. exit EO-100-102 and enter EO-100-114, RPV Flooding d. re-enter EO-100-102 at the beginning QUESTION:
076 (1.00)
When an SRV is cycling the pressure is reduced to 935 psig.
Why was this ressure selected?
a.
When MSIVs are open, it allows time to place RHR in suppression pool cooling.
b. If MSIVs are open, maximum steam flow through bypass valves is achieved.
c. This ensures cooldown rate will not be violated.
d. If MSIVs are open, it saves pneumatic gas supplie 'ENIOR REACTOR OPERATOR Page
STION:
077 (1.00)
A scram has occurred and the following conditions exist:
No systems can be aligned for injection MSIVs are closed Suppression pool pressure is 5 psig Drywell pressure is 8 psig In combination with the above, WHICH of the following conditions is providing adequate core cooling'
a. reactor level is unknown,
SRVs are open and RPV pressure is 65 psig b. reactor level is -190 inches, no SRVs are open and RPV pressure is 800 psig c. reactor level is unknown, 6 SRVs are open and RPV pressure is 50 psig d. reactor level is -170 inches, no SRVs are open and RPV pressure is 100 psig
SENIOR REACTOR OPERATOR Page
ESTION:
078 (
00)
A transient on Uni 1 resulted in a loss of coolant.
plant conditions exi t:
The following Reactor scrammed All rods at
Div 1 low pressure E
S has initiation signal Div 2 low pressure EC has no initiation signal HPCI and RCIC are injec ing RPV level is -128 inches n wide range RPV pressure is 6SO psig Suppression pool temperatur is 109 degrees F
Drywell temperature is 180 d grees F
Drywell pressure is 6.2 psig Suppression pool level is 23 f et No secondary containment proble exist Which of the following statements is co rect?
a. Div 2 low pressure ECCS failed to tart on high drywell pressure, the operator should star core spray and LPCI b. Wide range instrument may be used to etermine RPV level c.
RPV level cannot be determined and EO-1 0-114, RPV Flooding, should be entered d. Adequate core cooling is assured, and one oop of LPCI should be placed in drywell spray
" SENIOR REACTOR OPERATOR Page
ESTION:
079 (1.00)
An unidentified leak has occurred which has caused containment pressure and suppression pool level to inciease.
The following conditions exist:
power is 20%
reactor pressure is 1000 psig suppression pool level is 38 feet and increasing drywell pressure is 1.6 psig and increasing feedwater is maintaining reactor water level steady As Unit Supervisor, which action should be directed first?
a.
Shutdown the reactor and isolate feedwater.
b. Minimize RPV injection from sources external to primary containment and decrease suppression pool level.
c. Perform a rapid RPV depressurization per E0-112, Rapid Depressurization.
d. Manually scram the reactor and reduce RPV pressure QUESTION:
080 (1.00)
Which of the following is the purpose of maintaining the plant on the safe side of the Heat Capacity Temperature Limit Curve?
a.
To prevent exceeding the pressure suppression limit during a design basis LOCA b.
To prevent excessive dynamic loads on the suppression chamber structure during a design basis LOCA c.
To prevent exceeding the primary containment pressure limit during an emergency blowdown d.
To prevent excessive dynamic loads on the submerged suppression chamber components during an emergency blowdown
SENIOR REACTOR OPERATOR Page
ESTION:
081 (1.00)
When spraying the drywell to reduce drywell temperature per EO-100-103, PC Control, which statement below is the basis for spraying BELOW the Drywell Spray Initiation Limit Curve'
a. Spray the drywell before the temperature reaches the point at which an emergency blowdown is required.
b. Spray the drywell prior to the point where the spray will flash to steam and damage critical electrical components.
c. Spray the drywell prior to the point where boiling of the RPV level reference legs will cause unreliable measurement.
d. Spray the drywell before the point where spraying will result in negative differential pressures in the containment.
QUESTION:
082 (1. 00)
EO-100-103, Primary Containment Control, contains the Heat Capacity
~
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vel Limit Curve.
Per this curve, the safe region ends any time ppression pool level drops below 12 feet.
WHY is a level less than
feet considered unsafe by this curve?
a.
Chugging will occur in the downcomers during a LOCA.
b. Any energy from the drywell will be released directly to the suppression pool air space.
c.
The HPCI exhaust line becomes uncovered and directly pressurizes the suppression pool airspace.
d. Vortex limits for low pressure ECCS pumps are challenge SENIOR REACTOR OPERATOR Page
ESTION:
083 (1. 00)
Which of the following conditions places primary containment integrity over adequate core cooling?
a. Suppression pool level and pressure cannot be maintained below the "Pressure Suppression Limit" b. Drywell temperature cannot be maintained below 340 degrees F
c. Suppression pool level and RPV pressure cannot be maintained below the
"SRV Tail Pipe Level Limit" d. Drywell hydrogen/oxygen concentrations cannot be restored and maintained below combustible limits QUESTION:
084 (1.00)
Regarding the
"CS and RHR Vortex Limits" curve, which statement describes the consequences of operating the ECCS pumps below these limits?
a.
Low ECCS flow can result in overheating of the ECCS pump motor windings.
b. Flow induced vibration can result in damage to ECCS pump motor bearings.
c. High differential pressure can result in suction strainer damage.
d. Stopping and starting pumps without venting the system high point can result in fluid hammer and component damag 'ENIOR REACTOR OPERATOR Page
STION:
085 (1.00)
A LOCA is in progress.
The reactor is shutdown and all rods are at position 00.
RCIC area temperature is 155 degrees F and RCIC area differential temperature is 95 degrees F.
HPCI area temperature is 90 degrees F and HPCI area differential temperature is 25 degrees F.
MSIVs are closed.
HPCI started and was stopped by the PCO.
RCIC is maintaining reactor water level constant.
Which one of the following systems should be isolated?
a. control rod drive hydraulics b.
RCIC c.
HPCI d. Fire suppression QUESTION:
086 (1. 00)
The plant is operating at full power, E0-100-105, Radiation Release
~
~
ntrol, has been entered.
The EPB dose rate is projected to exceed 500
/hr in one hour.
Which of the following actions must the operator perform?
I.
Perform scram imminent actions II.
Scram the reactor III. Cooldown RPV in accordance with RPV Control IV.
Rapidly depressurize the reactor CHOOSE ONE a. I only b. I and II only c. I, II. and III only d. I, II, III and IV
SENIOR REACTOR OPERATOR Page
ESTION:
087 (1.00)
Due to high winds a loss of offsite power on both units has occurred and all diesel Generators failed to start.
How would, you monitor drywell temperature?
(CHOOSE ONE)
a.
use a digital voltmeter and table at panel 1TC621D b. use TR-15790A1(B1)
c. use a digital voltmeter and table at panel 1C614 d. use a RTD reader at panel 1C614 QUESTION:
088 (1.00)
Unit 1 was operating at 100% power with all equipment available.
Successive failures have resulted in loss of both RB chilled water loop circ pumps.
Assuming the proper automatic actions occur and no operator actions take place, which of the following will have cooling water flow to it?
I.
Reactor recirc pump motor II.
Drywell unit coolers III. Zone I and III supply unit cooling coils IV.
RWCU non-regenerative heat exchanger CHOOSE ONE a. I and XI b. II and III only
~ ~,
c.
1 an d. ZI, II1 and, X ~ SENIOR REACTOR OPERATOR Page
STION:
089 (1. 00)
A tube leak is identified on a condenser water box.
Which of the following actions should be initiated to allow repairs to the water box?
a.
reduce power to about 90~ for ALMA concerns b. reduce power to about 75< to maintain condenser vacuum c. reduce power to about 60-: for ALMQ. concerns d. reduce power as necessary only to maintain condenser vacuum QUESTION:
090 (1.00)
The unit has experienced a reactor recirc runback.
All APRMs are fluctuating between 45% and 58: power.
What action is required?
a.
no action required b.
scram c. insert CRAM array rods d. increase recirc flow to stabliize APRMs QUESTION:
091 (1.00)
The reactor automatically scrams due to high level in the scram discharge volume.
How soon is the NRC Operations Center required to be notified?
a. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> b. within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> c. within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> d. within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
'ENIOR REACTOR OPERATOR Page
ESTION:
092 (1. 00)
In accordance with 10 CFR20.1004,
"the the quantities expressed as dose equivalent".
is the SI unit of any of (CHOOSE ONE)
a.
Rad b. Gray c. Sievert d.
Rem QUESTION:
093 (1.00)
The Day Shift Supervisor must respond to the "Operator Concerns Form" within:
a.
7 days of receipt b.
5 days of receipt c.
10 days from submission date d.
30 days from submission date QUESTION:
094 (1.00)
When returning a licensed operator to on-shift duties, who is responsible for determining if the operator requires a physician examination?
a. the operator himself or herself b. Shift Supervisor c. Manager-Nuclear Operations d. Site nurse
SENIOR REACTOR OPERATOR Page
ESTION:
095 (1. 00)
Which of the following parameters is NOT used to determine jet pump operability?
a. indicated diffuser to lower plenum dP of individual jet pumps versus established patterns b. indicated core plate differential pressure versus established recirc pump speed-loop flow characteristics c. indicated total core flow versus established total core flow derived from recirc flow measurements d. indicated recirc loop flow versus established pump speed-loop flow characteristics QUESTION:
096 (1.00)
Which of the following is a limitation on the use of portable two-way radios in the Reactor Building?
a.
Radios will not be used in the Reactor Building on the 749'evel.
b. Radios will not be used within 15'f the SLC explosive valves.
c. Radios will not be used within 30'f the Remote Shutdown Panel.
d. Radios will have a maximum power output of.5 k '" SENIOR REACTOR OPERATOR Page
ESTION:
097 (1.00)
A component cannot be returned to operable status until which of the following is(are)
documented on the System Status Record Report?
I.II.
III.
IV.
CHOOSE ONE Permit Cleared Operability Testing Complete Work Complete Date WA Review Complete a. II and III only b. I and III only c. I, II and IV only d. I, II, III and IV QUESTION:
098 (1.00)
~
~
ile at 11% power a request is made to enter the drywell to identify the source of leakage.
Select the proper course of action.
a.
Cannot enter b.
Can enter-Shift Supervisor approval c.
Can enter-Supervisor Industrial Safety approval d.
Can enter-Shift Supervisor and Supervisor of Industral Safety approval
SENIOR REACTOR OPERATOR Page
STION:
099 (1.00)
Which of the following procedure changes can be authorized and implemented prior to obtaining Plant Superintendent approval?
a.
PCAFs which involve a 50.59 Action b.
PCAFs to Check-Off Lists c.
PCAFs which involve an unreviewed safety question d.
PCAFs which constitute an intent change QUESTION:
100 (1.00)
A shift of drywell cooling from reactor building chilled water is caused by:
I.
high reactor building chilled water temperature of 80 degrees F
II.
low flow in the reactor building chilled water loop after a
'minute time delay III. loss of 480 VAC power, 1B253042(1B263042)
to reactor building chilled water loop circ pump 1P214A(B).
(Loss of both)
IV.
loss of 120 VAC power 1Y21914 CHOOSE ONE a. I, III and IV only b. I, II and IV only c. II, III and IV only d. I, II, III and IV (**********
END OF EZA+XNATION **********)
'ENATOR REACTOR OPERATOR Page
SWER:
001 (1. 00)
REFERENCE:
EP-PS-100, TAB 6, Emergency Classification
[4.7/4.7]
294001A116
.. (KA's)
ANSWER:
002 (1. 00)
c
~e J.
REFERENCE:
EP-PS-100, TAB 6, Emergency Classification
[4.7/4.7]
294001A116
.. (KA's)
ANSWER:
003 (1.00)
REFERENCE:
SY017 I-4, APRN, LO
[3.4/3.4]
215005A107
...(KA s)
ANSWER:
004 (1.00)
~ SENIOR REACTOR OPERATOR Page
REFERENCE:
[4.3/4.1]
206000G008
.. (KA's)
'ANSWER:
005 (1.00)
REFERENCE:
Ec 9
[4.1/4.3]
219000A214
.. (KA's)
ANSWER:
006 (1. 00)
REFERENCE:
[3.7/3.7]
215004K401
.. (KA's)
ANSWER:
007 (1. 00)
" SENIOR REACTOR OPERATOR Page
REFERENCE:
[4.2/4.2]
295009A201
..(KA s)
ANSWER:
008 (1.00)
REFERENCE:
[4.0/4.0]
215003K402
.. (KA's)
WER:
009 (1. 00)
REFERENCE:
SY017 C-4, ADS System,
[3.3/3.4]
218000A101
.. (KA's)
ANSWER:
010 (1.00)
'SENIOR REACTOR OPERATOR Page
REFERENCE:
earn Tables
[3. 8/4. 1]
295006G007
.. (KA's)
ANSWER:
011 (1.00)
REFERENCE:
SY017 G-3, DC Distribution 125 VDC, LO 5
[3.3/3.3]
295004K203
.. (KA's)
WER:
012 (1.00)
REFERENCE:
SY017 D.-3, Reactor Feedwater System, LO RSb ON-145-001, RPV Level Control System Malfunction
[4.0/4.0]
295009A102
..(KA's)
ANSWER:
013 (1.00)
SENXOR REACTOR OPERATOR Page
REFERENCE:
Y017 A-S, EHC Pressure Control and Logic, LO 10
[3.9/4.0]
295025K101
..(KA's)
ANSWER:
014 (1.00)
REFERENCE:
E0-100-113, Level/Power Control PP002A, LO R7
[3.7/3.8]
295007K302
..(KA's)
SWER:
015 (1.00)
X
REFERENCE:
SY017 H-2, Main Steam System
[3.0/3.1]
LO 6 295002A106
..(KA's)
ANSWER:
016 (1.00)
< SENIOR REACTOR OPERATOR Page
REFERENCE:
0-100-113, Level/Power Control P002A, LO R7
[3.8/3.9]
295015K201
..(KA's)
ANSWER:
017 (1.00)
REFERENCE:
Unit 1 Technical Specification Bases 3/4.9.8 and 3/4.9.9
[2.9/3.2]
295023A202
.. (KA's)
ANSWER:
018 (1. 00)
REFERENCE:
EO-100-104, Secondary Containment Control PP002A, LO R1
[4.2/4.3]
~
2950346011
..(KA's)
ANSWER:
019 (1.00)
"'SENIOR REACTOR OPERATOR Page
REFERENCE:
-149-001, Loss of Shutdown Cooling
[3.5/3.5]
295021A203
..(KA's)
ANSWER:
020 (1. 00)
REFERENCE:
OI-AD-044 section 4.1.4
[2.7/3.7]
294001A103
..(KA's)
ANSWER:
021 (1.00)
REFERENCE:
ON-118-001, Loss of Instrument Air
[3.0/3.1]
295019G007
.. (KA's)
ANSWER:
022 (1.00)
REFERENCE:
OP-AD-001, Operations Shift Policies and Work Practices, section 6.9.3
[3.7/3.7]
294001K101
.. (KA s)
" SENATOR REACTOR OPERATOR Page
SWER:
023 (1. 00)
d REFERENCE:
OP-AD-001, Operations Shift Policies and Work Practices, section 6.3
[2.7/3.7]
294001A103
..(KA's)
ANSWER:
024 (1.00)
REFERENCE:
OP-AD-001, Operations Shift Policies and Work Practices, section 6.19
[4.2/4.2]
294001A102
..(KA's)
ANSWER:
025 (1. 00)
REFERENCE:
CFR20
[3.3/3.6]
294001K104
..(KA's)
ANSWER:
026 (1.00)
'ENIOR REACTOR OPERATOR Page
REFERENCE:
P-AD-001, Operations Shift Policies and Work Practices, section 6.18
[3.6/3.6]
294001K107
.. (KA's)
ANSWER:
027 (1.00)
REFERENCE:
SY017 L-8, Reactor Recirculation System and Motor-Generator Set LO 9
[3.2/3.2]
202001A303
.. (KA's)
ANSWER:
028 (1.00)
FERENCE:
OP-AD-001, Operations Shift Policies and Work Practices
[4.2/4.4]
294001A102
..(KA's)
ANSWER:
029 (1.00)
REFERENCE:
SY017 K-2, Control Rod Drive Hydraulic System LO 3m
[3.5/3']
201001A105
..(KA's)
h
'ENIOR REACTOR OPERATOR Page
SWER:
030 (1. 00)
d REFERENCE:
SY017 K-6, Rod Worth Minimizer LO 8 and
[3.0/3.0]
201006K514
.. (KA's)
ANSWER:
031 (1.00)
REFERENCE:
SY017 K-6, Rod Worth Minimizer LO 11
[3.1/3.2]
201006K104
..(KA's)
ANSWER:
032 (1.00)
REFERENCE:
SX017 E-1, Primary Containment Structure LO 3
[3.4/3.4]
295024A116
..(KA's)
ANSWER:
033 (1.00)
'
'SENIOR REACTOR OPERATOR Page
REFERENCE:
[3.9/3.9]
217000A305
.. (KA's)
ANSWER:
034 (1.00)
REFERENCE:
SY017 J-1, Reactor Vessel and Internals LO 2c
[3.2/3.2]
290002K102
..(KA's)
ANSWER:
035 (1 0)
d ERENCE:
Unit 1 Te SY017 M-Specs sections 3.8.1, 3.8.3.1 and 3.0.3 ch Specs LO 6
[.9/3.9 2 2001G005
.. (KA's)
ANSWER:
036 (1. 00)
REFERENCE:
[3.6/3.5]
205000A403
.. (KA's)
SENIOR REACTOR OPERATOR Page
SWER:
037 (1. 00)
REFERENCE:
ON-164-002, Loss of Recirculation Flow
[3.3/3.3]
295001A102
..(KA's)
ANSWER:
038 (1. 00)
REFERENCE:
SY017 L-9, Reactor Recirc Control L04
[3.0/3.0]
202002K402
..(KA's)
ANSWER:
039 (1.00)
REFERENCE:
E0-100-100, Cautions PP002, Emergency Operating Procedures Training LO 2
[3.5/3.7]
295028K101
..(KA's)
ANSWER:
040 (1.00)
~ SENIOR REACTOR OPERATOR Page
REFERENCE:
N-172-001, Off Gas System Isolation
[3.1/3.1]
295002K207
..(KA's)
ANSWER:
041 (1.00)
REFERENCE:
ON-'145-001, RPV Level Control System Malfunction
[3.8/3.8]
259001K109
..(KA's)
ANSWER:
042 (1.00)
ERENCE:
P&ID M-141
[3.4/3.4]
239001K101
..(KA's)
ANSWER:
043 (1.00)
REFERENCE:
l ON-159-002, Containment Isolation
[3.8/3.8]
239 001K401
..(KA's)
i SENIOR REACTOR OPERATOR Page
SWER:
044 (1. 00)
REFERENCE:
SY017 F-2, Offgas Recombiner
[3.0/3.3]
271000A301
..(KA's)
ANSWER:
045 (1.00)
REFERENCE:
AR-015-001 C02, C03, D02
[3.1/3.4]
262001K403
.. (KA's)
ANSWER:
046 (1.00)
REFERENCE:
ON-134-001, Loss of Reactor Building Chilled Water
[3.4/3.5]
262001A303
~.(KA's)
ANSWER:
047 (1.00)
SENATOR REACTOR OPERATOR Page
REFERENCE:
-188-001, 250 VDC System
~
~
[3.2/3. 33 2 63 0 0OA3 01
.. (KA's)
ANSWER:
048 (1.00)
REFERENCE:
ON-'193-002, Main Turbine Trip
[3.8/3.8i 241000K311
.. (KA's)
ANSWER:
049 (1. 00)
ERENCE:
ON-193-001, Turbine EHC System Malfunction
[3.4/3.4)
241000A114
.. (KA's)
ANSWER:
050 (1. 00)
REFERENCE:
OP-183-001, ADS and SRVs section 3.2.3
[4.0/4.1]
218000K101
.. (KA s)
SENIOR REACTOR OPERATOR Page
SWER:
051 (1.00)
REFERENCE:
SY017 C-2, Core Spray
[3.2/3.6l 295001A205
.. (KA's)
ANSWER:
052 (1.00)
REFERENCE:
EO-100-103, Primary Containment Control
[3.8/3.8)
203000K118
.. (KA's)
ANSWER:
053 (1.00)
REFERENCE:
OP-149-001, RHR System
[3.5/3.5]
203000K108
.. (KA's)
ANSWER:
054 (1.00)
I SENIOR REACTOR OPERATOR Page
REFERENCE:
P-149-003, RHR Operation in the Fuel Pool Cooling Mode
~
~
[3.2/3. 3]
233000G007
..(KA's)
ANSWER:
055 (1.00)
REFERENCE:
OP-152-001, HPCI System
[3.9/3.8]
206000A303
.. (KA's)
ANSWER:
056 (1.00)
FERENCE:
OP-152-001, HPCI System
[4.4/4.4]
295003A103
..(KA's)
ANSWER:
057 (1.00)
REFERENCE:
OP-150-001, RCIC System
[3.4/3.5]
217000K601
.. (KA's)
SENXOR REACTOR OPERATOR Page
SWER:
058 (1.00)
b REFERENCE:
OP-179-001, Area Radiation Monitoring System
[3.3/3.2]
272000A310
..(KA's)
ANSWER:
059 (1.00)
REFERENCE:
ON-100-005, Excess Drywell Leakage Xdentification
[3.8/3.6]
223002G013
..(KA's)
ANSWER:
060 (1.00)
REFERENCE:
[3.4/3.6]
223001A110
..(KA's)
ANSWER:
061 (1.00)
'ENIOR REACTOR OPERATOR Page
REFERENCE:
N-159-002, Containment Isolation
[3.4/3.5]
223002K406
..(KA's)
ANSWER:
062 (1.00)
REFERENCE:
OP'-070-001, Standby Gas Treatment System
[3.0/3.1]
261000A304
..(KA's)
ANSWER:
063 (1.00)
ON-159-002, Containment Isolation
[3.9/4.0]
290001A301
..(KA's)
ANSWER:
064 (1.00)
REFERENCE:
Tech Spec 3.3.1-1(d)
AR-103-001-A01/9
[3.3/3.6]
295001A101
..(KA s)
'ENXOR REACTOR OPERATOR Page
SWER:
065 (1.00)
REFERENCE:
SY017 K-7, Reactor Manual Control System
[3.6/3.7]
201002G007
..(KA's)
ANSWER:
066 (1.00)
, REFERENCE:
AR-103-001 AR-104-001
[3.6/3.9]
295037K308
.. (KA's)
ANSWER:
067 (1.00)
REFERENCE:
OP-161-001, RWCU System, section 3.3
[3.2/3.2]
204000G010
.. (KA's)
ANSWER:
068 (1.00)
SENIOR REACTOR OPERATOR Page
REFERENCE:
Y017 M-5, SDPS
[4.0/3.8]
295025G006
..(KA's)
ANSWER:
069 (1. 00)
REFERENCE:
OP'-181-001, Refuel Platform Operation, Section 3.2.2.(i)(3)
[2.9/3.5]
234000G010
..(KA's)
ANSWER:
070 (1.00)
FERENCE:
EP-PS-100, TAB 6, Emergency Classifications
[4.2/4.5]
295038G011
.. (KA's)
ANSWER:
071 (1.00)
REFERENCE:
[4.2/4.3]
295003A102
.. (KA's)
'ENXOR REACTOR OPERATOR Page
WER:
072 (1.00)
d REFERENCE:
E0-000-031, Station Power Restoration
[3.8/4.0]
2950036006
.. (KA's)
ANSWER:
073 (1. 00)
REFERENCE:
E0-100-009, section 4.8
[3.5/3.7]
295016K303
.. (KA's)
ANSWER:
074 (1.00)
REFERENCE:
ON-100-009, Control Room Evacuation
[4.0/4.0]
295016A108
..(KA's)
ANSWER:
075 (1.00)
'ENIOR REACTOR OPERATOR Page
REFERENCE:
P002, EOP and ES Training LO 3
[4.2/4.2]
295031A102
..(KA s)
ANSWER:
076 (1.00)
REFERENCE:
EO'-100-102, RPV Control PP002A, EOP training
[4.0/4.1]
LO 7 295031K207
..(KA's)
ANSWER:
077 (1.00)
REFERENCE:
E0-100-102, RPV Control Bases
[4.0/4.3]
295031K304
~
.. (KA')
ANS R:
078 (1. 00)
REFERENCE:
EO-100-102, PV Control
[4.4/4.4]
295031A201
..(KA's)
),(j~
'ENXOR REACTOR OPERATOR Page
SWER:
079 (1.00)
d REFERENCE:
E0-100-102, RPV Control
[3.9/3.9]
295029A201
.. (KA s)
ANSWER:
080 (1.00)
REFERENCE:
EO-100-103, Primary Containment Bases
[3.2/3.6]
295026K203
..(KA's)
ANSWER:
081 (1.00)
REFERENCE:
E0-100-103, PC Control Bases
[3.7/4.1]
'95028K201
..(KA s)
ANSWER:
082 (1.00)
" SENIOR REACTOR OPERATOR Page
'EFERENCE:
0-100-103 PC Control Bases
[3.5/3.8]
295030K207
.. (KA's)
ANSWER:
083 (1.00)
REFERENCE:
EO-'100-103, PC Control Bases
[4.1/4.2]
295024K101
..(KA's)
ANSWER:
084 (1.00)
d ERENCE:
E0-100-103, PC Control
, [3.5/3.8]
295030K307
.. (KA's)
ANSWER:
085 (1.00)
REFERENCE:
E0-100-104, Secondary Containment Control
[3.7/3.9]
295032A105
.. (KA's)
'ENIOR REACTOR OPERATOR Page
SWER:
086 (1.00)
REFERENCE:
E0-100-105, Radiation Release Control
[3.3/4.3]
295038A201
..(KA's)
ANSWER:
087 (1.00)
REFERENCE:
EO-100-030, Att A
[3.4/4.2]
295003G012
.. (KA's)
ANSWER:
088 (1.00)
REFERENCE:
ON-134-001, Loss of RB Chilled Water
[3.3/3.4]
295018A101
..(KA's)
ANSWER:
089 (1.00)
'ENIOR REACTOR OPERATOR Page
REFERENCE:
-142-002, Circulating Water Condenser Leak
~
~
~
[3.1/2.9]
295002A107
.. (KA's)
ANSWER:
090 (1.00)
REFERENCE:
ON-'178-002, Core Flux Oscillations
[3.3/3.4]
295001A106
..(KA's)
ANSWER:
091 (1.00)
X ~p.
ERENCE:
NDAP-QA-724, Significant Operating Occurrence Reports
[3.7/4.3]
295006G008
.. (KA s)
ANSWER:
092 (1. 00)
REFERENCE:
CFR 20.10004
[3.8/3.8]
294001K103
.. (KA's)
'ENIOR REACTOR OPERATOR Page
SWER:
093 (1. 00)
REFERENCE:
OI-AD-060, Operations Concerns, section 4.6
[3.3/4.2]
294001A109
.. (KA s)
ANSWER:
094 (1.00)
REFERENCE:
OI-AD-044, Return to Shift Duty/Job Promotion
[3.7/3.7]
294001A103
.. (KA's)
ANSWER:
095 (1.00)
REFERENCE:
[3.4/4.2]
202001G005
..(KA's)
ANSWER:
096 (1.00)
'EN1OR REACTOR OPERATOR Page
REFERENCE:
AP-00-0316, section 6.3.5
[3.1/3.2]
294001A104
..(KA's)
ANSWER:
097 (1.00)
REFERENCE:
NDAP-QA-0302, System Status and Equipment Control, section 6.1.l.f
[3.2/3.4]
294001A115
.. (KA's)
ANSWER:
098 (1. 00)
a ERENCE:
NDAP-QA-0309, Primary Containment Access and Control
[3.2/3.4]
294001K114
.. (KA's)
ANSWER:
099 (1. 00)
REFERENCE:
NDAP-QA-003, Procedure Change Process, section 6.3.3
[2.9/3.4]
294001A101
.. (KA's)
SENIOR REACTOR OPERATOR Page
SWER:
100 (1. 00)
d REFERENCE:
ON-134-001, Loss of RB Chilled Water t3.1/3.2]
295018K307
..(KA's)
(**********
END OF EXAMINATION**********)
ATTACHhG<WT2 FACILI'IYCOMMENTS