IR 05000400/1992030
| ML18022A887 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/12/1992 |
| From: | Casto C, Ernstes M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18022A886 | List: |
| References | |
| 50-400-92-30, NUDOCS 9203030165 | |
| Download: ML18022A887 (100) | |
Text
~R REGII('Ip.0 A0O
~O
+**++
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 ENCLOSURE
EXAMINATION REPORT - 50-400/92-300 Facility Licensee:
Carolina Power and Light Company
'acility Name:
Shearon Harris Nuclear Plant Facility Docket No.:
50-400 Examinations were administered at Shearon Harris Nuclear Plant near New Hill, North Carolina.
Chief Examiner:
Approved By:
Michael.- E., Ernstes
.c.
II h 'r es A. Casto Chief Operator Licen ng Sec ion
Division of Reactor Safety Z -/5-f~
Date Signed a <2.
Date S'gned SUMMARY Scope:
During the week of January 27, 1992.
Written and operating examinations were administered to eight Senior Reactor Operators (SRO) applicants.
Results:
All eight candidates passed the examinations.
A weakness was noted on the written examination performance.
Although all candidates passed the examination, half of those taking the examination had
.scores of 84 or less.
Specific knowledge areas which were deficient are detailed in paragraph 3c.
9203030165 F2021"~
PDR ADOCK 05000400 V
REPORT DETAILS 1.
Facility Employees Attending Exit L. Hartin, Manager Nuclear Training W. Powell, Training Manager J. Collins, Manager Operations J.
Bryan, Manager Harris Training Unit Simulator C. Briney, Operations Specialist, Nuclear Assessment Dept.
J.
Boska, Manager License Training A. Barbee, Senior Specialist, Operator Training 2.
NRC Personnel Attending Exit
- H. Ernstes, Examiner, DRS S. Cahill, Examiner, DRS G. Harris, Examiner, DRS C. Casto, Chief Operator Licensing Section 2,
DRS J. Tedrow, Senior Resident Inspector D. Roberts, Resident Inspector
- Chief Examiner 3.
Discussion a
~
Written Exam b.
Although all candidates passed the examination, four of the eight had scores of 84 percent or less.
There were 12 questions which were missed by half of the candidates or more.
The following are some of the knowledge areas which were tested by these 12 questions.
~ Fire pre-plans
~
CFR 20 exposure limits
~ basis for Axial Flux Difference Tech.
Spec.
. ~ range of operation for in-core thermocouples
~ basis for actions in Functional Recovery Procedures (FRPs)
~ required response for dropped control rods
~ emergency boration flow paths
~ expected flux levels following a reactor trip Operator Performance A weakness was observed in candidates'amiliarity with industry events.
Albeit a small sample size of events was discussed, it is essential for operators to learn from the experience gained through significant events at other facilities.
Discussion with the licensee staff indicates that there may not be an efficient process for disseminating industry events to licensee candidates and
h possibly licensed operators.
There was a strong evidence of the operators studying the JPH exam bank.
Although familiarity with the exam process is, highly encouraged, the operators should be learning the procedures vice concentrating on the JPHs.
This was apparent during the administration of JPHs which were different from those found in the facility exam bank.
When the operator received a cue that was contrary to the way the facility JPH was written, he became confused and would often back track in the procedure to check his actions.
C.
Examination Administration There was an excessive number of observers in the simulator booth.
This created distractions for the simulator operators with a
resulting impact on the
- scenarios.
Observers were at times conducting unrelated business in the booth.
This included outside phone calls being received and a high noise level.
Unsolicited assistance from the observers further distracted the simulator operators while entering or removing Malfunctions.
The number of problems associated with the administration of the simulator scenarios had a deleterious effect.
There were several malfunctions which were not entered at the desired time.
In two instances, malfunctions were improperly entered which required additional scenarios to be run.
These deficiencies impair the examiners'bility to effectively evaluate the candidates.
Additionally, delays and extra scenarios compound the stress experienced by the candidates during the examination process.
The card reader to the diesel building was inoperable.
This required security's assistance in entering the building.
This card reader is continuously exposed 'to the outside environment often rendering it inoperable.
The assistance and cooperation of the control room personnel during the administration of the walkthrough examinations was particularly noteworthy.
Exit Meeting At the conclusion of the site visit, the examiners met with those representatives of the plant staff indicated in paragraph I, to discuss the results of the examinations and inspection finding ENCLOSURE
U. S.
NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION
CANDIDATE'S NAME:
FACILITY:
REACTOR TYPE:
Shearon Harris
PWR-WEC3 INSTRUCTIONS TO CANDIDATE:
DATE ADMINISTERED:
01 27 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 804.
Examination papers will be pxcked up four (4) hours after the examination starts.
TEST VALUE CANDIDATE'S SCORE
~O0 100.00 TOTALS All work done on this examination is my own.
I have neither given nor
.received aid.
an x a e s xgna ure DRAFT COPY
SENIOR l'EACTOR OPERATOR Page
QUESTION:
001 (1.00)
Which one o'f the following is a symptom of a continuous withdrawal of a control bank?
a.
Lowering pressurizer level b.
Actuation of PZR backup heaters.
c.
"Over Power Delta-T Block Rod C-4 ALERT" alarm.
d.
"RCS Loop Tavg Low-Low Alert" alarm.
ANSWER:
001 (1. 00)
c ~
REFERENCE:
AOP 001~
p
000001A205 (4.4/4.6)
000001A205
.. (KA's)
QUESTION:
002 (1.00)
Which one of the following actions does NOT automatically occur upon receipt of a "Fuel Handling Building (FHB) High Radiation" alarm?
a.
FHB Emergency exhaust fans start.
b.
FHB Fuel Pool Purification System starts.
c.
FHB Operating Floor supply fans stop.
d.
FHB loading area isolation dampers shut.
ANSWER:
002 (l. 00) SENIOR l'EACTOR OPERATOR Page
REFERENCE:
5g p 55 KA:
000061K302 (3.4/3 ')
000061K302
.. (KA's)
QUESTION:
003 (1.00)
"SGTR With Loss of Reactor Coolant:
Subcooled Recovery",
contains a CAUTION which reads:
"With RCP's secured, steps to depressurize the RCS and terminate SI should be performed as quickly as possible after the cooldown has been initiated...".
Which one of the following is the purpose of this CAUTION?
a.
Minimize possible pressurized thermal shock of the reactor vessel.
b.
Minimize possible pressurized.thermal shock of the S/G tubes.
c.
Ensure the RCP minimum seal delta-p is maintained.
d.
Minimize the potential for S/G overfill.
ANSWER:
003 (1. 00)
a ~
REFERENCE:
EOP-LP-3.5, LO A.8 EOP-EPP-020, p.17 KA:
000038A136 (4.3/4.5)
000038A136 (KA's)
SENIOR R'EACTOR OPERATOR Page
QUESTION:
004 (1.00)
According to AOP-016,
"Excessive Primary Plant Leakage",
which one of the following is the criteria'or initiating Safety Injection?
a. If PZR level cannot be maintained greater than 15<.
b. If VCT level cannot be maintained with letdown isolated.
c. If the leakage exceeds the capacity of one charging pump.
d. If the leakage exceeds normal auto VCT makeup capability.
ANSWER:
004 (l. 00)
d.
REFERENCE:
AOP-016, p.5 KA:
000009K321 (4.2/4.5)
000009K321 (KA's)
QUESTION:
005 (1.00)
According to the immediate actions of AOP-017,
" Loss. of Instrument Air", which one of the following plant conditions should be present before the operator initiates AFW and trips the turbine?
a.
Letdown valves going closed.
b.
c A significant decrease in S/G level.
Feedwater reheater bypass valves close.
d.
Feedwater cascading valves close.
ANSWER:
005 (1.00) SENIOR REACTOR OPERATOR
\\
REFERENCE AOP-LP-3.10, LO 1.1.3, p.4 AOP-017, p.3 KA:
000065G010 (3.2/3.3)
Page
000065G010
.. (KA's)
QUESTION:
006 (1.00)
Which one of the following will initiate an automatic start signal for the Turbine Driven Auxiliary Feedwater Pump?
a. SI signal.
b. Loss of power to either emergency bus.
c. Trip of last running main feedwater pump.
d. Low-low level in one S/G (1/3 detectors).
ANSWER:
006 (1. 00)
b.
REFERENCE:
9g p 26 KA:
061000K406 (4.0)
[4.0/4.2]
061000K406
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
007 (1.00)
Which one of the following is a function of the Incore Thermocouple System (ITS)?
a.
Used to determine subcooling margin.
b. Used to calibrate excore nuclear instruments c. Used for calorimetric calculations.
d.
Used as density compensation to RVLIS.
ANSWER:
007 ( 1. 00)
a ~
REFERENCE:
SD-106 1. 0 KA:
017020K401
[3.4/3.7]
017020K401
.. (KA's)
QUESTION:
008 (1.00)
Which one of the following main turbine generator trips has a 30 second time delay before the generator output breakers will open?
a.
An electrical fault.
b.
Low condenser vacuum.
c.
Low turbine bearing oil pressure.
d. Thrust bearing failure.
ANSWER:
008 (1. 00) SENIOR REACTOR OPERATOR Page
REFERENCE:
GEN LP
~ Oi LO 1 ~ 1 ~ 6g p 18 MT LP 3 ~ Oi LO 1 ~ 1 ~ 27'
KA:
045000A304 (3.4/3.6)
045050K101
..(KA's)
QUESTION:
009 (1.00)
Concerning the loss of an Emergency DC Bus, which one of the following statements is correct?
a.
Loss of emergency dc bus DP-1A-SA will prevent the AFW Terry Turbine from operating.
b.
Loss of either emergency dc bus will cause the main feed regulating valves to fatal as-is.
c.
Loss of either emergency dc bus will cause the feedwater isolation valve to fail open.
d.
Loss of emergency dc bus DP-1B-SB will prevent Emergency Diesel Generator 1B from starting.
ANSWER:
009 (l. 00)
d.
REFERENCE:
LO 1
3 AOP-025, p.8 SD-155.01, p.16 KA:
063000K301
[3.7/4.1]
063000K301
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
010 (1.00)
Assume both CCW pumps are lost.
Which one of the following actions is required immediately according to AOP-014,
"Loss of Component Cooling Wateri> o a. Isolate letdown.
b. Line-up SW to standby ESW header.
c.
Connect
"C" CCW. pump to emergency bus.
d. Stop Spent Fuel Cooling pumps.
ANSWER:
010 (1. 00)
REFERENCE:
AOP-LP-3.0, LO A.2 AOP-014, p.15 KA:
000026K303 (4.0/4.2)
000026K303
.. (KA's)
QUESTION:
011 (1.00)
When work is performed on systems containing fluids or gases, a
Clearance must be issued if the system pressure and temperature limits are exceeded.
Which ONE of the following correctly list those limits?
a.
140 psi, 100 degrees F.
b.
200 psi, 140 degrees F.
c.
100 psi, 140 degrees F.
d.
140 psi, 200 degrees C SENIOR REACTOR OPERATOR Page
ANSWER:
011 (1. 00)
c ~
REFERENCE:
SHEARON HARRIS:
ADOP-LP.2.2 L.O. 1.1.4 AP-020
[/3 4]
194001K109
.. (KA's)
QUESTION:
012 (1.00)
The plant is operating at full power and all systems are functioning within their normal operating bands.
Which ONE of the following conditions/malfunctions would require an IMMEDIATE trip of the affected Reactor Coolant Pump, per AOP-018,
"Reactor Coolant Pump Abnormal Conditions?"
a. Thrust bearing temperature increases to 180 degrees F.
b. Seal inlet temperature increase to 210 degrees F.
c. Motor winding temperature increase to 310 degrees F.
d.
Pump shaft vibration level increases to 5 mils.
ANSWER:
012 (1. 00)
c ~
SENIOR PEACTOR OPERATOR Page
REFERENCE:
SHEARON HARRIS:
AOP-018, page
AOP-LP-3.11, LO A.2
[4.0/3.9]
003000G014
..(KA's)
QUESTION:
013 (1.00)
WHICH one of the following Technical Specifications leakage definitions correctly classifies Pressurizer PORV leakage?
a.
IDENTIFIED LEAKAGE b.
CONTROLLED LEAKAGE d.
RCS PRESSURE ISOLATION VALVE LEAKAGE ANSWER:
013 (F 00)
a ~
REFERENCE:
SHEARON HARRIS:
OST-1026 Reactor Coolant System Leakage, Tech.
Spec. Definitions 1.17 K/A 002000G011 (3.3%4.0)
002000G011
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
014 (1.00)
WHICH one of the the column below manual SI?
1 ~
~
3 ~
4 ~5.6.
,7.
following correctly identifies those valves from that receive an actuation signal on receipt of a I
CSIP suction valves from RWST.
CSIP suction valves from VCT.'harging header isolation valves to the RCS.
RCP seal injection isolation valves.
CSIP miniflow common isolation valve.
RHR suction valves from RWST Accumulator discharge isolation valves.
a.
2, 3, 4, 6,
and
b.
1, 2, 4, 5, and
c.
1, 3,
4, 5,
and
d.
1, 2, 3,
5, and
ANSWER:
014 (1.00)
d.
REFERENCE:
SHEARON HARRIS:
SIS-LP-3.0, Section 2.4, L.O. 1.1.6 (4.2%4.4)
006030K101
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
015 (1.00)
Given the following:
The plant is operating normally at 50< powe All systems are in automatic.
WHICH one of the following correctly describes the effect on the listed parameters resulting from a sudden closure of the "B" loop main steam line isolation valve?
a. Total steam flow remains the same, loops "A" and "C" Tavg decreases, loop "B" Tavg increases.
b. Total steam flow and-loops "A" and "C" Tavg remain the same, loop "B" Tavg increases.
c. Total steam flow and loops "A" and "C" Tavg increases, loop "B" Tavg decreases.
d. Total steam flow remains the same, loops "A" and "C" Tavg increases, loop "B" Tavg decreases.
ANSWER:
015 (1.00)
a ~
REFERENCE:
SHEARON HARRIS:
T&AA-LP-2.25, pages 19 and 20, L.O. 1.1.3 (3.2/3.4)
035010A205
.. (KA's)
SENIOR PEACTOR OPERATOR Page
QUESTION:
016 (1.00)
WHICH one of the following is a possible cause of 'a continuous rod withdrawal?
a. Tref input to the rod control system failing low.
b. Tavg input to the rod control system failing high.
c. Valve 1CS-291, RWST to Charging pumps fails open.
d. Loss of control air to 1CS-156, Boric acid to CSIP suction.
ANSWER:
016 (1. 00)
c ~
REFERENCE:
SHEARON HARRIS:
AOP-001, page
AOP-LP-3.1 (4 '/4.6)
000001A205
.. (KA's)
SENIOR PEACTOR OPERATOR Page
QUESTION:
017 (1.00)
Step 15 of EOP-EPP-008, SI Termination, directs the operator to establish secondary pressure control using the steam dumps.
WHICH one of the following describes the action the operator should take upon subsequently determining secondary pressure is decreasing UNCONTROLLABLY?
a. Immediately shut all MSIV's, SG PORV's and isolate Main Feedwater, Auxiliary Feedwater, SG sampling and blowdown lines.
b. Initiate the Secondary Integrity Criteria of EPP-008 fold-out page and transition to EPP-014, Faulted Steam Generator Isolation.
c. Initiate the Secondary Integrity Criteria of EPP-008 'fold-out page and transition to EPP-015, Uncontrolled Depressurization of All Steam Generators.
d. Verify all Steam Generator PORV's are shut and transfer steam dump control to the STM PRESS mode then verify that all steam dumps are shut.
ANSWER:
017 (1 ~ 00)
b.
REFERENCE:
SHEARON HARRIS:
EOP-EPP-008, SI Termination, Fold-out page, EOP-LP-3.1, L.O. I.A.7 (4.1/4.2)
000040G010
.. (KA's)
SENIOR PEACTOR OPERATOR Page
QUESTION:
018 (1.00)
The following alarms annunciate over a period of about one minute.
FAN E-12 E-13 EXH TEMP HI FAN E-12 E-13 EXH TEMP HI-HI EXH FAN E-12 CHAR FLTR TRBL EXH FAN E-13 CHAR.FLTR TRBL WHICH one of the following is the most likely cause of these alarm annunciations?
a.
Loss of cooling water to the Fuel Handling Building ventilation coolers.
b. Loss of power to the Fuel Handling Building fans.
c. Fire in the Fuel Handling Building or charcoal filter.
d. Fuel Handling Building ventilation damper failure.
ANSWER:
018 (1. 00)
c ~
REFERENCE:
SHEARON HARRIS:
APP-ALB-023 pages 139, 171, 173 and 179 (3.3/3.5)
000067A203
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
019 (1.00)
You have been directed to maintain a constant actual SG water level of 66% narrow range at. the ACP. Given a copy of AOP-004, Attachment 6,
Which ONE of the following indicates how SG level indication should change at the ACP as SG pressure decreases from 1000 psig to 200 psig in order to maintain a constant mass in the SG?
a.
SG level indication at the ACP remains constant b.
SG level indication at the ACP increases c.
SG level indication at the ACP decreases d.
SG level cannot be determined at the ACP ANSWER:
019 (1. 00)
b.
REFERENCE:
AOP-004,SEC 3.3, step
AOP-LP-3.0; L.O. 1.1.2
[]
000068A204
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
020 (1.00)
Upon declaration of an emergency, the Shift Foreman is designated to act as the initial Site Emergency Coordinator until the Technical Support Center is activated.
If the Shift Foreman is Unavailable, which ONE of the following is the primary alternate to act as the Site Emergency Coordinator prior to activating the Technical Support Center?
a.
Roving Senior Control Operator b.
Control Room Senior Control Operator c.
Manager Shift Operations ANSWER:
020 (1.00)
a ~
REFERENCE:
PEP-102 Rev.
9 page
ADOP-LP-2.0 L.O. 1.1.3 l/4')
194001A116
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
021 (1.00)
During shift turnover, you, the oncoming Shift Foreman, are informed that one of your licensed Reactor Operators (RO) has called and will be one hour late.'ne licensed RO, a Senior Reactor Operator (SRO),
and an unlicensed Shift Technical Advisor are present for the oncoming crew'.
The operators in the offg'oing crew have all worked the last 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.
Unit 1 is in Mode 3.
Which ONE of the following is the necessary course of action for staffing the oncoming shift in accordance with Technical Specifications?
a.
No action required because only one RO is required for Mode 3.
b.
No action required because two RO's are required for Mode 3 but Technical Specifications allow the crew to be short one operator for up to two hours to accommodate unexpected absences.
c.
Authorize, with the concurrence of the offgoing Shift Foreman, holding one RO over from the offgoing crew because two RO's are required in Mode 3.
d.
Obtain Plant General Manager authorization to hold one RO oyer from the offgoing crew because two RO's are required in Mode 3.
ANSWER:
021 (1. 00)
d.
REFERENCE:
Tech.
Spec. 6.2.2.f OMM-001 3.3.1 ADOP-LP-2.0 L.O. 1.1.4 f/3 43 194001A103 (KA')
SENIOR PEACTOR OPERATOR Page
QUESTION:
022 (1.00)
Which ONE of the following correctly fills in the blanks for a Site Emergency level event?
The SHNPP Emergency Plan requires the State and Counties to be notified within and all individuals on site to be HK a.
15 minutes, 15 minutes b.
15 minutes, 30 minutes c.
30 minutes, 15 minutes d.
30 minutes, 30 minutes ANSWER:
022 (1.00)
b.
REFERENCE:
PLP-201 (Emergency Plan)
4 ~ 2 bg 4 ~ 6
~ 2 (p
2 g 4 7)
EPT-LP-2.1 L.O. I.A.3,6
[/4')
194001A116
..(KA's)
SENIOR PEACTOR OPERATOR Page
QUESTION:
023 (1.00)
Which ONE of the following statements is correct concerning Fire Pre-plans?
a.
Fire pre-plans are meant. to be followed as a procedure by the Fire Brigade Team Leader b.
Fire pre-plans are for the unrestricted use of all SHNPP personnel c.
Fire pre-plans contain lists for Control Room Operators of all plant equipment and cabling in an affected area and actions to be taken if they are damaged d.
Fire pre-plans contain guidance on restoring fire protection and ventilation systems to normal after a fire ANSWER:
023 (1. 00)
d.
REFERENCE:
FPP-012 ADOP-LP-2.3 L.O. I.A.3
[/F 5]
194001K116
..(KA's)
P SENIOR PEACTOR OPERATOR Page
QUESTION:
024 ( 1. 00)
Which ONE of the following individuals holds primary responsibility for the safe movement of fuel and core components insa.de the containment and Fuel Handling Building (FHB) and is also responsible for directly supervising fuel handlinq in the containment or, the FHB if fuel z.s being handled only zn the FHB?
a.
Refueling Coordinator b.
Shift Foreman c.
SRO Fuel Handling I
d.
FHB Operator ANSWER:
024 (l. 00)
c ~
REFERENCE:
PLP-616 5.3.2, p 5,10,11 LP-PP-2.8 L.O. I.A.2
[/3 4j 194001A103
.. (KA's)
I SENIOR PEACTOR OPERATOR Page
QUESTION:
025 (1.00)
A clearance requires a piece of equipment to be isolated for maintenance; Which ONE of the following valve types would be an acceptable barrier BUT-would also require a qaggzng or blocking device in order to meet the equipment isolation requirements of the clearance per OMM-014?
a.
a fails-open motor-operated valve with the local operation switch tagged off b.
a fails-closed air operated valve with the air supply isolated c.
a fails-open air operated valve with the air supply isolated d.
a fails-closed solenoid valve with the power supply isolated ANSWER:
025 (1. 00)
c ~
REFERENCE:
OMM 001 5 ~ 1 ~ 23
~ 6 g OMM 014 5 ~ 1 ~ 1 ~ 9 g 10 ADOP-LP-2.2 L.O. 1.1.4 l/4'3 194001K102
..(KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
026
.(1.00)
On September 25, 1991 a
22 year old radiation worker with no previous exposure received
Rem during a life saving effort.
Which ONE of the following statements is correct concerning the Maximum Legal exposure per 10CFR20 that this individual can
'eceive on October 1st, 1991?
( Assume an NRC Form 4 is on file )
a.
No additional exposure is legally allowed since the individual has received an acute dose.
b.
The individual could legally receive up to
REM exposure.
c ~
d.
The individual could legally receive an additional dose not to exceed 2 REM.
The individual could legally receive an additional dose not to exceed 1.25 REM.
ANSWER:
026 (1. 00)
c ~
REFERENCE:
10CFR20.101 PP-LP-2.7 L.O. 1.1.2 JTA 343*029*H1*02 l/3')
194001K103
..(KA's)
SENIOR PEACTOR OPERATOR Page
QUESTION:
027 (1.00)
In an Emergency situation where it is necessary to perform life saving actions, which one of the following is the maximum allowable whole body dose?
a.
25 rem b.
50 rem c.
75 rem d.
100 rem ANSWER::
027 '1. 00)
c ~
REFERENCE:
PLP-201, Em. Plan 4.6.3.5 PP-LP-2.7 L.O. 1.1.2 L/3'1 194001K103
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
028 (1.00)
A point source in the auxiliary building is reading 500 mrem per hour at a distance of two,feet.
Background radiation level is negligible.
Two options are possible to perform a mandatory task near this radiation" source OPTION 1:
Operator X can perform the task in 30 minutes working at a distance of 4 feet from the point source OPTION 2:
Operators Y and Z can perform the task in 75 minutes at a distance of 6 feet from the point source Which ONE of the following is preferable and consistent with the ALARA concept?
a.
Option 1 should be used because of the lower total exposure.
b.
Option 2 should be used because of the lower individual exposure.
c.. Option 1'hould be used because fewer individuals and minutes are used.
d.
Option 2 should be used because of the lower dose rate per individual.
ANSWER:
028 (1. 00)
a ~
REFERENCE:
HPP-020-10.1.1.10
[/3 5]
194001K104
..(KA's)
SENIOR PEACTOR OPERATOR Page
QUESTION:
029 (1.00)
When directed by an EOP, a foldout page is applicable.
These foldouts contain action steps that must be performed.
With the foldout in use, at what point should these actions be taken?
a.
When the foldout page is first referenced.
b.
When the symptoms associated with that foldout are noticed.
c.
When a stable plant condition is reached.
d.
When a change in plant status would have no adverse effects on recovery efforts.
ANSWER:
029 (1.00)
b.
REFERENCE:
EOP User Guide 5.3.6 EOP-LP-3.19 L.O. I.A.3
[/3 ~ 9 j.
194001A102
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
030 (1.00)
A clearance is required to be canceled but the clearance holder is not available on site or by phone.
Which of the following constitutes sufficient authorization to cancel the clearance?
Any Senior Control Operator after physically verifying the work is complete and restored b.
The Manager Shift Operations after verifying the work is complete by either inspection or direct report from an operator c.
A co-worker of the clearance holder who was also involved with the task which required the clearance d.
The clearance holder's immediate supervisor ANSWER:
030 (1. 00)
d.
REFERENCE:
AP-020 5.1.4 ADOP-LP-2.2 L.O. 1.1.4 (/4 ~ 1j 194001K102'. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
031 (1.00)
Under which ONE of the following situations would operation of equipment with an active clearance on it be permitted per AP-020?
a.
Valve position verification per OMM-1 by operations personnel b.
During the performance of a surveillance if the equipment wall be restored in less than 5 minutes c.
Post-maintenance testing to confirm equipment availability when authorized by the Shift Foreman d.
During an emergency when authorized by any on-duty licensed Control Room Operator ANSWER:
031 (1. 00)
a ~
REFERENCE:
AP-020 5.1.2 ADOP-LP-2.2 L.O. 1.1.4 194001K102
.. (KA's)
'
SENIOR REACTOR OPERATOR Page
QUESTION:
032
- (1.00)
A 36 year old welder with a current lifetime exposure of 80 rem is the onl'y individual qualified to perform a critical repair which is expected to result in an exposure of 800 mrem.
His current received dose is 1000 mrem this quarter-and 5000 mrem for this year.
Which ONE of the following is the only correct course of action?
Assume the welder has'a valid NRC Form 4 on file.
a.
The welder will NOT be able to perform the repair because
CFR 20 lifetime exposure limits would be exceeded b.
The welder will NOT be able to perform the repair because
CFR-20 yearly exposure limits would be exceeded c.
The welder can, perform the repair if a yearly dose extension is approved by the Vice President SHNPP d.
The welder can. perform the repair with Manager RC and Manager E&RC approval as long as the dose he receives this quarter does not exceed 3000 mrem.
'ANSWER:
032 (1. 00)
c ~
REFERENCE:
AP-503 5.1.4 PP CP 2 '
L 0
1 'g6
[/3 4)
194001K103
.. (KA's)
SENIOR REACTOR OPERATOR
Page
QUESTION:
033 (1.00)
Which ONE of the following is the preferred method used for verifying the position of a locked throttle valve that was positioned last shift?
There is no reason to, doubt that the position of the valve is in other than normal.
a.
Close the valve, counting the number of turns until fully closed then restore the valve to it s throttled position by opening the same number of turns b.
c ~
Administratively determine that the valve was positioned to the correct throttle point and has not been operated since by reviewing valve lineup records Observe another operator close and reposition the valve to it's throttled position as part of a valve lineup d.
Visually verify locking device is in place and that the valve appears to be in the correct position ANSWER:
033 (1. 00)
d.
REFERENCE:
OMM-001 5.1.23.4 PP-LP-3.11 L.O. 1.1.2 (/3 7]
194001K101
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
034 (1.00)
A periodic audit of Category II locked valves is being conducted.
Which ONE of the following zs the preferred method in accordance with the ALARA program for verifying the position of a valve that is located in a High Radiation Area?
a.
Check the valve by operating in the shut direction for 1/4 turn then return to full open b.
Visually verify stem travel and position indicators are in the correct position and that the locking device is installed c.
Administrative'ly verify the valve position by reviewing on-file valve lineups or Independent Verification forms d.
No verification is requiied.
The valve can be deleted from the checklist and will be checked the next time that High Radiation Area is entered.
ANSWER:
034 (1.00)
c ~
REFERENCE:
OMM-011 5.5 PP-LP-3.11-2.2.2.C.1.c; L.O. 1.1.2 f/3')
194001K101
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
035 (1.00)
Which ONE of the following is an allowable method of performinq the second check of a valve position for Independent Verification purposes after a valve is initially positioned shut?
a.
Verify the position by exerting force to operate the valve in the shut direction without damaging the seat or stem b.
Verify the position 'by exerting force to operate the valve in the open direction to dislodge the valve from it's seat, then restore in the shut direction c.
Visually verify stem height position d.
Observe solenoid energized-to-open lights are extinguished ANSWER:
035 (1.00)
a ~
REFERENCE:
PP-LP-3.11'.O.
1.1.8 PLP-702 5.3 (/3')
194001K101
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
036 (1.00)
From which ONE of the followinq parameters does the RIL computer comparator determine the setpoznts for the Rod Bank Lo Alarm and Rod
.Bank Lo-Lo Alarm?
a. auctioneered high Tave b. auctioneered high turbine impulse pressure c. auctioneered high nuclear instrument power level d. auctioneered high delta-T ANSWER:
036 (l. 00)
d.
REFERENCE'ODCS-LP-3.1; L.O. 1.1.5
[3.6/3.8]
014000A103
.. (KA's)
QUESTION:
037 (1.00)
A Rod Control Non-Urqent Failure Alarm is received from the Logic Cabinet while operating at steady-state 984 power.
Which ONE of the following describes the effect of this alarm on rod movement?
a.
No effect, all rods capable of moving b. Automatic rod motion blocked, manual and individual bank movement permitted c. Automatic and manual rod motion blocked, individual bank movement permitted d. All rod motion is blocked
SENIOR REACTOR OPERATOR Page
ANSWER:
037 (1.00)
a ~
REFERENCE RODCS-LP-3.0; L.O. 1.1.12 SD-104 4.F 1
[3.7/3.9]
001050A201
.. (KA's)
QUESTION:
038 (1.00)
During the recovery of dropped rod in Control Bank A, an Urgent Failure Alarm is received once the withdrawal of the rod commences.
Which ONE of the following cabinet alarms is the cause for the Control Board alarm?
a. Logic Cabinet:
Slave Cycler Failure-incorrect input to slave cycler b. Logic Cabinet:
Card Interlock-removed card signal from continuity monitors c.
Power Cabinet:
Logic Error Detector-current command signals to stationary and movable gripper coils lost simultaneously d.
Power Cabinet: Regulation Failure Detector-output current to coils does not match demand ANSWER:
038 (1. 00) SENIOR REACTOR OPERATOR Page 35.
REFERENCE:
SD-104 4.1.4.3 RODCS-LP-3.0; L.O. 1.1.11
[3'/4.1]
001000K409
..(KA's)
QUESTION:
039 (1.00)
An inadvertent SI actuation has occurred due to improper testing of the system.
Which ONE of the follow'ing describes all the conditions necessary to manually reset the inadvertent SI actuation?
a.
No conditions necessary, can reset immediately b. Must wait 60 seconds and remove test'ignal that caused the actuation c. Must wait 60 seconds and have P-4 Reactor Trip signal present d. Must remove test signal that caused the trip and have P-4 Reactor Trip signal present ANSWER:
039 (1. 00)
c ~
REFERENCE:
SD-103 Fig. 7.27 ESFAS-LP-3.0; L.O. 1.1.5
[3.9/4.3]
013000K401
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
040 (1.00)
Technical Specifications state that with Axial Flux Difference (AFD)
outside the target'band limit, the operator must,.within 15 minutes restore AFD to within the target band or, within 30 minutes reduce thermal power.
The basis for these limits is to:
a. allow time for identification and correction of a dropped rod b. allow time for identification and correction of a failed excore detector c. prevent axi'al power shifts from causing a redistribution of xenon d. prevent exceeding the Heat Flux Hot Channel Factor Fq(z) in any one channel ANSWER:
040 (1. 00)
c ~
REFERENCE:
Tech.
Spec.
B3.2.1 NIS-LP-3.0; L.O. 1.1.18
[2.6/3.7]
015020G006
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
041 (1.00)
Which ONE of the following statements correctly describes a purpose and operation of the Chemical and Volume Control System's cation resin bed demineralizer?'.
Is normally in use to provide full-time mechanical filtration of corrosion products b. Is normally in use to control Reactor Coolant System (RCS)
pH by releasing lithium hydroxide II c. Is normally off-service but is used to control RCS oxygen by releasing dissolved hydrogen when needed d. Is normally off-service but is used to remove cesium fission products ANSWER:
041 (1. 00)
d.
REFERENCE:
-CVCS-LP-3.0 1.1.3 SD-107 2.3.3
[2.5/3.1]
004010K612
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
042 (1.00)
The Chemical and Volume Control System is cross tied to the Residual Heat Removal System (RHRS) at two points.
Which ONE of the following is NOT a function of these connections?
a. Provides additional letdown during heatup of the RCS and establishment of the pressurizer bubble b. Allows for purification of the RCS during low temperature and pressure shutdown conditions c. Used for safety injection recirculation phase d. Allows the CSIP's to pump flow through the RHRS Heat Exchangers to augment cooldown ANSWER:
042 (,1. 00)
d.
REFERENCE:
SD-107 5'.2 CVCS-LP-3.0 2.7.B.2; L.O. 1.1.7
[3.4/3.9]
004010K101
.. (KA's)
QUESTION:
043 (1.00)
Which ONE of the following is the maximum temperature the incore thermocouples are capable of indicating?
a.
1800 deg F
b.
2000 deg F
c.
2300 deg F
d.
2500 deg F
SENIOR REACTOR OPERATOR Page
ANSWER:
043 (l. 00)
c ~
REFERENCE:
SD-106 3.F 1 Reg.
Guide 1.97 Rev.
[3.1/3.3]
017020K403
.. (KA's)
QUESTION:
044
'(1.00)
The CVCS system has an interlock between the letdown isolation valves (1CS-1 and 1CS-2)
and the orifice isolation valves (1CS-7, 1CS-S, 1CS-9).
Which ONE of the following is the purpose of his interlock?
a. Prevents over-pressure excursions which could damage the letdown heat exchanger b. Prevents depressurization of the letdown line resulting in steam flashing in the regenerative heat exchanger c. Prevents excessive initial flow which would occur 'by opening of the letdown isolations if all the orifice isolations were open d. Prevents cooling of the water in the line from developing large differential pressures across the isolation valves ANSWER:
044 (1.00)
b.
REFERENCE:
SD-107 4.6.2 CVCS-LP-3.0; L.O. 1.1.5
[3.0/3.4]
004020K403
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
045 (1.00)
Which ONE of the following correctly explains how the. Power Range Nuclear Instruments compensates for gamma flux?
a.
No compensation is necessary since gamma flux is much smaller than neutron flux and proportional to power in the Power Range b.
A pulse height discriminator is utilized to filter out the gamma'ignal current pulses which are smaller than the neutron signal pulses c.
A compensated ion chamber. is. utilized which produces ecpal, off--
setting currents proportional to gamma flux to cancel xt out d.
An electronic multiplier applies a power level dependent fraction to factor out the portion of the output signal due to
~
gamma flux ANSWER:
045 (1. 00)
a ~
REFERENCE:
SD-105 NIS LP 3 '
L 0 1 ' 'g12g13
[2.9/3.2]
015000K501
..(KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
046 (1.00)
Which ONE of the following is a function of the sodium hydroxide (NaOH)
additive to the Containment Spray System (CSS)?
a.
Removes hydrogen from the post-accident. containment atmosphere b. Enables the spray to condense steam more effectively, aiding in more rapid depressurization c. Maintain a final recirculation sump pH of at least 8.5 d. Prevent clogging of the* CCS nozzles ANSWER:
046 (l. 00)
c ~
REFERENCE:
CSS-LP-3.0; L.O. 1.1.4
[2.8/3.2]
026020K401
.. (KA's)
I
SENIOR REACTOR OPERATOR Page
QUESTION:
047 (1. 00)
Which ONE of the following is NOT a design feature of the Containment Recirculation Sumps?
a.
The velocity of water in the immediate sump screen area is limited to allow high density particles to settle out on. the floor b.
The RHR and CSS pump intakes are completely separated by a solid wall which does not allow sump liquid to flow between them, preventing cross-clogging of both intakes c.
A curb on the containment floor is located in front of the sump to prevent heavy debris from washing into the sump d. Two-stage stainless steel screens limit the size of. any passed particles to less than the spray nozzle diameter ANSWER:
047 (1. 00)
b.
REFERENCE:
SD-112 3.5 CSS-LP-3.0; L.O. 1.1.12
[2.8/3.3]
026000K405 (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
048 (1.00)
With reactor power in the intermediate range, if an Intermediate Range Nuclear Instrument (IRNI) is OVERCOMPENSATED then the indicated IRNI power level will be (1)
'han the actual power level; The automatically energize when the highest reading IRNI reaches z s se point during a reactor shutd'own.
Which ONE of the following correctly fills in the statement blanks?
a.
(1) lower, (2) will b.
(1) lower, (2) will not c.
(1) higher, (2) will d.
(1) higher, (2) will not ANSWER:
048 (1.00)
a ~
REFERENCE:
SD-105 NIS-LP-3.0; L.O. 1.1.12,14
[3.7/3.8]
[3.1/3.5]
015000K407
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
049 (1.00)
What is the reason for limiting the number of successive Reactor Coolant Pump starts in a given period?
a. Minimize pitting of the starter contacts b. Prevents overheating of the motor windings c. Prevents excessive torsional stresses on the motor shaft d. Minimizes axial wear on the seal package ANSWER:
049 (1. 00)
b.
REFERENCE:
OP-100 4.0 RCS-LP-3.0; L.0.1.1.22
[3.3/3.6]
003000G010
.. (KA's)
QUESTION:
050 (1.00)
Which ONE of the following is an input to the Auxiliary Feed Pump Turbine Governor Valve controller?
a.
AFW Flow Control Valve differential pressure b.
TDAFW Pump discharge flow c. Main steam header pressure d.
AFW Pump turbine speed ANSWER:
050 (1. 00) SENIOR REACTOR OPERATOR Page
REFERENCE:
AFS-LP-3.0 2.4.6B, LO 1.1.6 KA:
061000K601 (2.5/2.8)
061000K601
.. (KA's)
QUESTION:
051 (1.00)
During a review of the 0900 BOP log readings it s noticed that the A-SA EDG Day Tank level was recorded at 84% which is below the required 85%
IAW with Tech.
Spec.
3.8.1.1.
Assuming that indicated level is still reading 84% which ONE of the actions below is the required action?
a.
Declare A-SA EDG inoperable from the last level recorded at
> or = to 85:
b.
Declare A-SA EDG inoperable from the last level recorded at
>85<o c.
Instruct the BOP to manually initiate makeup until indicated level is restored to > or = 854 d.
Verify fuel amount is sufficient by referencing oil specific gravity vs.
MCB% indication curve ANSWER:
051 (1.00)
d.
REFERENCE:
T.S. 3.8.1.1 T.S.I.88-002 DE-LP-3.0; L.O. 1.1.10
[3.4/3.9]
064000G011
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
052 (1.00)
Assuming the air compressors do not operate, how many consecutive times can one of the two air receivers roll and start the emergency diesel?
a.
4 start sequences b.
5 start sequences c.
9 start sequences d.
10 start sequences ANSWER:
052 (1. 00)
b.
REFERENCE:
DE-LP-3.0 2.4-D.8 SD-155.01 2.4
[3.1/3.5]
064000A304
.. (KA's)
QUESTION:
053 (1.00)
Pressurizer level transmitter LT-459 is selected for control when the reference leg for LT-459 develops a slow leak.
Which ONE of the following describes the instrument and plant response?
a ~
b.
c ~
d.
LI-459 PZR LVL INDICATION INCREASES DECREASES INCREASES DECREASES LI-460 PZR LVL INDICATION DECREASES INCREASES DECREASES INCREASES VCT LEVEL INCREASES INCREASES DECREASES DECREASES
SENIOR REACTOR OPERATOR Page
ANSWER:
053 (1.00)
a
~
REFERENCE:
PZRLC LP
~ Oi L 0
1 ~ 4~7 SD-100.03 4.7. 1
[3.4/3.6j 011000A210
.. (KA's)
QUESTION:
054 (1.00)
The lA-SA Emergency Diesel Generator (EDG) started automatically on an undervoltage condition on 6.9 kV bus 1A-SA.
The diesel is up to rated speed and voltage and the A-SA EDG output breaker (106) is shut.
Which one of the following signals could trip the Emergency Diesel Generator after this start a.
High crankcase pressure.
b.
Low lube oil pressure.
c.
Engine overspeed.
d.
Overcurrent ANSWER:
054 (1.00)
c ~
REFERENCE:
DE-LP-3.0; L.O. 1.1.5 SD-155.01 4.2 OP-155
[3.9/4.2]
064000K402 (KA's)
< ~
SENIOR REACTOR OPERATOR Page
QUESTION:
055 (1".00)
Given the following:
Steam Generator tube leakage has been reported to be:
SG A 4 gph SG B 10 gph SG C 22 gph Which ONE of the following Technical Specification being exceeded?
a.
.Individual SG primary to secondary leakage ONE steam generator b.
Individual SG primary to secondary leakage TWO steam generators c.
Individual SG primary to secondary leakage THREE steam generators
'imits is limit in limit in limit in d.
Total primary to secondary leakage through ALL steam generators daily limit ANSWER:
055 (1. 00)
a ~
REFERENCE:
1 '
'
[2.9/3.73 035000G011
.. (KA's)
C SENIOR REACTOR OPERATOR Page
QUESTION:
056 (1.00)
Given the following:
Spent Fuel Pool Cooling 'A'rain suction line has a large leak all systems are in a normal configuration for Mode
operation Which ONE of the following will occur assuming no operator action?
a.
The 'A'ooling Pump will lose suction when pool level falls to 6 feet above the top of the fuel b.
The 'A'ooling Pump will lose suction when pool level falls to 18 feet above the top of the fuel c
d.
Automatic makeup from the demineralized water system will start at 22 feet above the fuel which will maintain level at that height At the Lo-Lo Level setpoint of 282.0 feet, LS 5110A will send an isolation signal to the suction line valves, isolating the leak ANSWER:
056 (1 ~ 00)
b.
REFERENCE:
SD-116 2.2
[2.9/3.2]
033000K401
.. (KA's)
SENIOR REAOTOR OPERATOR.
Page
QUESTION: 057'1.00)
Which ONE of the following water source and path combinations-is NOT a possible method of manually maintaining the required level in the Spent Fuel Pool in the event of a leak in the Spent Fuel'ool Cooling System?
a.
RWST via the Containment Spray system b.
ESW via emergency connections c.
CCW Surge Tank via CCW emergency connections d.
Demineralized Water via the Fuel Pool Purification Pumps
'NSWER:
057 (1. 00)
c ~
REFERENCE:
SD-116 5.2
[3.1/3.5]
033000A203
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
058 (1.00)
An operator-initiated Automatic Bus Transfer from the Unit Auxiliary Transformer (UAT) to the Start-up Transformer (SUT) is in progress.
All conditions to shut the SUT supply breaker 107 have been met.
Which ONE of the following would be the result if the 107 breaker switch was taken to close and held in the close position?
a.
No action would occur until the switch was released b.
Breaker 107 would close and UAT Supply Breaker'108 would open c.
Breaker 107 would close and an alarm would be received after 3 seconds due to parallel UAT/SUT operation d.
Breaker 107 would close and then open after 1 second to prevent extended parallel UAT/SUT operation ANSWER:
058 ( 1. 00)
c ~
REFERENCE:
SD-156 6.9kV-LP-3.0 2.5.3; L.O. 1.1.5
[2.8/3.1]
062000K403
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
059 (1.00)
A LOCA that resulted in automatic Safety Injection is in progress when a Lo-Lo RWST signal is received on 4 of 4 level channels.
Which ONE of the following sets of valve operations occurs automatically due to this signal?
a.
1SI-322 and 1SI-323, RWST to RHR Pumps Isolation Valves CLOSE b.
1CS-291 and 1CS-292, RWST.to Charging/SI Pumps Isolation Valves CLOSE c.
1RH-25 and 1RH-63, RHR Pump Discharge to Charging/SI Pump Suction Isolation Valves CLOSE d.
1SI-300, 1SI-301, 1SI-310, and 1'SI-311, Containment Sump to RHR Pumps Isolation Valves OPEN ANSWER:
059 (1. 00)
d.
REFERENCE:
SIS-LP-3.0; L.O. 1.1.6 SD-110
[4.2/4.3]
006020A304
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
060 (1.00)
Which ONE of the following is the basis for the Reactor Coolant Pump ratchet and pawl Ants-Reverse Rotation Device?
a.
Prevents a reverse-rotating pump which would cause excessive motor starting current if started b.
Prevents'everse RCS loop flow which would bypass the reactor, reducing core flow c.
Prevents reverse pump rotation which would damage the pump seals due to the low reverse 'rotation speed d.
Prevents rapid reversal of pump on loss of power which transmits excessive shock to the pump bearings ANSWER:
060 (1. 00)
a ~
REFERENCE:
RCS-LP-3.9, 2.5.B, 5.a.5
[3.6/3.8]
002000K602
..(KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
061 (1.00)
On-line testing of the Train A Reactor Trip Breaker is in progress and the Train A Bypass Breaker is shut.
Which ONE of the following is the system response if the Train B Bypass Breaker is shut?
a.
Train B General Warning Alarm, breakers remain shut b.
Reactor Partial Trip Annunciator Alarm, breakers remain shut c.
Train B Reactor Trip Breaker opens d.
All Reactor Trip and Bypass Breakers open ANSWER:
061 (1. 00)
d.
REFERENCE:
SD-103 4.0 p. 13, RPS-LP-3.0g L.O. A.9
[3.3/3.6)
012000K604
.. (KA's)
r SENIOR REACTOR OPERATOR Page
QUESTION:
062 (1.00)
A depressurization transient has-been initiated by the failure of PT-444 and associated PORV actuation.
Given the following current RCS pressure instrument indications:
PT-444 PT-445 PT-455 PT-456 PT-457 2500 psig 2050 psig 2050-psig 1950 psig 2050 psig Which ONE of the following describes the expected current status of the Pressurizer PORV's assuming the pressure transient started at 2235 psig?
a.
All Pressurizer PORV's open b.
PCV-445B and PCV-445A open, PCV-444B closed c.
PCV-445B and PCV-445A closed, PCV-444B open d.
All Pressurizer PORV's closed ANSWER:
062 (,1. 00)
C ~
REFERENCE SD 100.03, 4.6 PZRPCS-LP-3.0 2.5.3; L.O. 1.1.4
[3.8/4.1]
010000K403 (KA')
SENIOR REACTOR OPERATOR Page
QUESTION:
063 (1.00)
The plant is operating at 1004 reactor power with the pressurizer pressure master controller in automatic.
If the setpoznt is raised from 69% to 71> then controller output -will and RCS pressure will a. increase, decrease b. increase, increase c. decrease, increase d. decrease, decrease ANSWER:
063 (1. 00)
c REFERENCE:
SD 100.03, sec 4.1
[3.7/3.7]
010000A107
.. (EA's)
SENIOR REACTOR OPERATOR Page
QUESTION; 064 (1.00)
. Given the. following and a copy of T.S. 3.3.3.1:
-Control Room Normal Outside Air Intake Monitors are alarminq and reading 1.0 E -5.
Automatic actions have been initiated.
-Emergency Filtration Fan is in operation-RM-23 for NORTH Side Emergency Outside Air Intake is NOT alarming and is reading 2.0 E -6-RM-23 for SOUTH Side Emergency Outside Air Intake alarming at 1.5 E -5-Operators have selected the NORTH Side Outside Air Intake-plant is in Mode
An I
& C Technician reports that one of the two NORTH Emergency Intake RM-23 channels was calibrated 13 months ago and the, Digital Channel Operational Test was last performed 44 days ago.
Which ONE of the following is the correct action required by Technical Specifications?
a.
Minimum channel operability requirements are met so NO action is required b.
Close the NORTH Side Intake within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and place Control Room Ventilation in Recirculation mode c.
NORTH Side Intake may be left open up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after which it must be shut and Control Room Ventilation placed in Recirculation mode d.
Close the NORTH Side Intake immediately and open the SOUTH Side Intake after increasing it's alarm setpoint to 10x the original value of 4.9 E -6 ANSWER:
064 (1. 00)
b.
REFERENCE:
RMS-LP-3.0 T.S. 3.3.3.1 SD-118 4.2.26,27
[3.1/3.6]
073000G005
.. (KA's)
SENIOR REACTOROPERATOR'age
QUESTION:
065 (1.00)
The Component Cooling Water (CCW) System has developed a leak with CCW Pump "A" running.
The automatic action that occurs when the CCW Pump
"A" discharge pressure decreases to 80 psig is:
a. Automatic makeup is initiated to the CCW surge tank b.
The line to the Gross Failed Fuel Detector isolates c.
1CC-252, CCW return from RCP thermal barriers. closes d.
The standby CCW pump starts ANSWER:
065 (1. 00)
d.
REFERENCE:
CCWS-LP-3.0 2.6.3; L.O. 1.1.6
[3.1/3.3)
008000K401
.. (KA's)
QUESTION:
066 (1.00)
Which ONE of the following inputs would cause a steam dump arming signal to be generated?
a.
3 of 4 turbine throttle valves closed b.
Load rejection of 5% within 2 minutes c.
4 degree F deviation between Tave and Tref d. Input to 2 of 3 auto stop trip header pressure instruments is
<1000 psig
SENIOR REACTOR OPERATOR Page
ANSWER:
066 (1.00)
d.
REFERENCE:
SDCS-LP-3.0 1.1.4 SD-126.01
[2.7/2.9]
041020K603
.. (KA's)
QUESTION:
067 (1.00)
Which ONE of the following choices would be the resulting lineup of the Emergency Service Water (ESW) equipment listed'below after the automatic actions generated by a Safety Injection Signal occur?
Booster Pumps (2)
to ESW isolations 1SH
& 40 (3)
return to Auxiliary Reservoir 1SW 270 6 271 (4)
a.
(1)
ON-(2)
OFF (3 )
SHUT (4)
OPEN b.
(1)
ON (2)
ON (3)
SHUT (4)
OPEN c.
(1)
OFF (2)
OFF (3)
OPEN (4)
OPEN d.
(1)
ON (2)
ON (3)
OPEN (4)
OPEN
SENIOR REACTOR OPERATOR Page
ANSWER:
067 (1.00)
b.
REFERENCE:
ESW-LP-3.0; L.O. A.15 SD-139
[3.6/3.8]
076000K116
.. (KA's)
QUESTION:
068 (1.00)
Which ONE of the following is NOT a basis for the 45 psig entry condition to FR-J.1,
"Response to High Containment Pressure" ?
a. Deformation of the containment structure is likely if design pressure is exceeded b. High pressure indicates significant energy release to the containment c. Design basis containment leakage could be surpassed if design pressure is exceeded d. Hydrogen ignition could cause a higher peak pressure that would deform the containment structure ANSWER:
068 (1. 00)
a ~
REFERENCE:
EOP-LP-3.13; L.O. I.A.5
[3.5/3.6]
000069G007
.. (KA's)
SENIOR REACTOR OPERATOR Page,61 QUESTION:
069 (1.00)
The plant is at 85% power when a loss of both Main Feed pumps occurs.
The RO attempts to manually trip the reactor but the reactor fails to trip.
The RO then attempts to trip the rod drive M/G but the breakers will not open.
Which ONE of the following is the reason why it is necessary to trip the turbine within 30 seconds in this situation?
a. Provides another trip signal to the Reactor Protection System in
'n attempt to shutdown the reactor b. Prevents an uncontrolled cooldown of the RCS c. Forces the RCS to heat up due to the loss of load which adds negative reactivity d. Maintains S/G water inventory ANSWER:
069 ( 1. 00)
d.
REFERENCE:
EOP-FRP-S.1 EOP-LP-3.15 L.O. A.6
[4.2/4.3]
000029K306
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
070 (1.00)
Given the following alarms and indications in the Control Room:
-"TWO OR MORE RODS AT BOTTOM" annunciator ALB 13-7-3-2 Rod Bottom Lights lit POWER RANGE LOWER DETECTION HIGH FLUX DEVIATION OR AUTO DEFEAT annunciator LB 13-5-4
-"ROD CONTROL URGENT ALARM" annunciator ALB13-7-'1-slowly lowering Tavg-Reactor Power 74, slowly decreasing Which ONE of the following describes the required operator response to this event in accordance with AOP-001, Malfunction of Rod Control System?
a. Manually withdraw control rods to match Tavg and Tref b. Monitor nuclear power and maintain steady-state conditions c.
Open the liftcoil disconnects for the dropped rods d. Proceed with an orderly shutdown to Hot Shutdown conditions ANSWER:
070 (1. 00)
d.
REFERENCE:
AOP-LP-3.1; L.O. 1.1.3 AOP-001 3.2
[3.4/3.6]
000003G009
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
071 (1.00)
Which ONE of the following is an acceptable emergency boration path for-the charging pumps to deliver boric acid to the RCS per AOP-002,
"Emergency Boratzon"?
a.
Seal Water Supply Lines b. Alternate Charging to Loop C cold leg c. Auxiliary Spray d. High Head Safety Injection charging lines via the BIT ANSWER:
071 (1. 00)
a
~
REFERENCE CVCS-LP-3.0 AOP-002-4.0.2 AOP-LP 3.2g L.O. 1.1.6 (3.6/3.7]
000024G007
..(KA's)
QUESTION:
072 (1.00)
Which ONE of the following is NOT a major step performed in FRP-J.1,
"Response to High Containment Pressure?
a. Identify and isolate unexpected sources of water in containment b.
Check for and isolate faulted Steam Generator c. Monitor containment hydrogen concentration and remove excess d. Confirm containment heat removal and containment isolation
SENIOR REACTOR OPERATOR Page
ANSWER:
072 (1. 00)
a ~
REFERENCE:
EOP-LP=3.3; L.O. I.A.1
[3.8/4.2]
000069K301
.. (KA's)
QUESTION:
073 (1.00)
Which ONE of the following is a correct set of conditions for terminating Safety Injects.on when an Adverse Containment condition exists?
a. Pressurizer level indicates 30%
b. Narrow range S/G levels indicate 25%
c.
RCS subcooling indicates 28 deg F
d. Total feed flow to intact S/G's is 228 kpph ANSWER:
073 (1.00)
d.
REFERENCE:
Path 1 Step
EOP-LP-3.1; L.O. I.A.6
[3.4/3.9]
000011A208
..(KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
074 (1.00)
The basis for depressurizing all intact steam generators to 90 psig in step 11 of FRP-C.1,
"Response to Inadequate Core Cooling," is to:
a.
ensure core exit thermocouple temperatures are reduced to less than 1200 degrees F..
b. reduce secondary pressure to,aid in recovery of steam generator water level.
c. reduce'RCS pressure to initiate core reflood.
d.
enhance natural circulation cooling of the reactor core.
ANSWER:
074 (1. 00)
c ~
REFERENCE:
EOP-LP-3.10; L.O. I.A.6 EPP-C.1
[4.0/4.4]
000074K3 11
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
075 (1.00)
Given the following situation:
ATWS in progress; reactor cannot be manually tripped turbine xs tripped emergency boratxon is. in progress AFW pumps are running and feeding S/G's RCS pressure is 2395 psig, PORV's open, block valves shut These conditions require the RCS to be depressurized by opening the PORV block valves per FRP-S.l.
Which ONE of the following zs the basis for reducing pressure in this situation?
a. Prevent rapid overpressurization event that would actuate the pressurizer safety valves b. Increase the flow of borated water into the RCS c. Begin a slow, controlled cooldown and depresurrization which minimizes positive reactivity feedback from a negative MTC d. Minimize primary to secondary leakage in the most limiting ATWS event, a
SGTR, until other recovery actions can be taken ANSWER:
075 ( 1. 00)
b.
REFERENCE:
EOP-FRP-S.1 EOP-LP-3.15 L.O. A.6
[4.2/4.3]
000029K311
.. (KA's)
SENZOR REACTOR OPERATOR Page
QUESTION:
076 (1.00)
Which one of the following would be an indication of a control bank failure to move during a power increase from 504 to 100:?
Assume a
constant boron concentration.
a.
"ROD CONT SYS URGENT ALARM" annunciator b.
"ROD CONT SYS NON-URGENT ALARM" annunciator c. High pressurizer level indication and possible alarm.
d. Increasing Tavg and possible
"RCS Tref/Tavg HIGH-LOW" annunciator ANSWER:
076 (1. 00)
a ~
REFERENCE:
AOP-001 AOP-LP-3.1; L.O. 1.1.4
[3.3/4.1]
000005A201
..(KA's)
QUESTION:
077 (1.00)
Which one of the following electrical loads would lose power if off-site and on-site (all AC power)
were lost?
a.
Rad monitor system RM-23 trains b.
RCP oil lift pump c. Fire detection local control panels d. Generator air side seal oil 'backup pump
SENIOR REACTOR OPERATOR Page
ANSWER:
077 (1. 00)
b.
REFERENCE:
RCS-LP-3.0; L.O. 1.1.5 120v UPS-LP-3.0
[3.7/4.1]
000055A204
.. (KA's)
QUESTION:
078 (1.00)
Which one of the following describes the basis for establishing hot leg recirculation following a large break LOCA?
a.
To terminate boiling in the hot legs and to ensure balanced cooling of the core.
b.
To terminate boiling in the.hot legs'and to prevent boron precipitation.
c.
To terminate boiling in the core and to ensure balanced cooling of the core.
d.
To terminate boiling in the core and to prevent boron precipitation.
ANSWER:
078, (1. 00)
d.
REFERENCE:
EOP-LP-3.3; L.O. I.A.5 EPP-011
[3.8/4.2]
000011K313
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
079 (1. 00)
Which one of the following Functional Restoration Procedures should have first priority if they are all in a RED path condition?
a. FRP-P.l,
"Response to Imminent Pressurized Thermal Shock" b.
FRP-C.1,
"Response to Inadequate Core Cooling" c.
FRP-H.1,
"Response to loss of Secondary Heat Sink" d. FRP-J.1,
"Response to High Containment Pressure" ANSWER:
079 (1.00)
b.
REFERENCE:
EOP Users Guide 5.2.2
[4'/4.4]
000074G012
.. (KA's)
QUESTION:
080 (1.00)
Which one of the sets of conditions from the list below require emergency boration in accordance with AOP-002,
"Emergency Boration"?
1. Abnormal control rod insertion.
2. Unexplained increasing RCS temperature or nuclear power.
3.
Shutdown margin less than 1850 pcm in Mode 2.
4.
Two or more control rods not fully inserted after a reactor trip.
a ~
1 /
2 /
b.
1, 2,
C.
1(
3(
d.
2, 3,
SENIOR REACTOR OPERATOR Page
ANSWER:
080 (1. 00)
b.
REFERENCE:
AOP-LP-3.2; L.O. 1.F 1
[4.1/4.4]
000024K301
.. (.KA's)
QUESTION:
081 (1.00)
If a loss of power to a 6.9 KV emergency bus 1A-SA occurs with the train B MDAFW pump running, S/G feed flow must be maintained above 25 kpph in order to:
a. Maintain an adequate heat sink for the reactor b. Ensure minimum flow through the MDAFW Pump while recirc flow is secured c. Prevent S/G Low-Low Level from causing an auto-start signal to be processed for the secured pump d.
Be able to secure the TDAFW Pump that started on the power loss ANSWER:
081 (1. 00)
b.
REFERENCE:
[4.1/4.4]
000057K301
..(KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
082 (1.00)
The Control Room Operators are responding to an ATWS condition.
They take manual actions to reduce core power but when the turbine stop valves are checked they are found open.
An operator attempts a manual trip of the turbine but the turbine will not trip.
Which ONE of the following is the next required action per EOP-FRP-S.1?
a. Attempt to manually runback turbine b. Initiate Safety Injection c. Shut the MSIV's and bypasses d. Ensure all AFW Pumps running at full capacity ANSWER:
082 (1.00)
c ~
REFERENCE:
EOP-LP-3.15; L.O. I.A.2 FRP-S.1
[4.4/4.5]
000029A209
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
083 (1.00)
EOP-001 has been directly entered due to a loss of power on all AC emergency busses.
Which ONE of the following is true regarding transition out of EPP-001 if no form of power has been restored to the AC emergency busses?
a. Function Restoration Procedures (FRP's)
should be implemented as any RED terminus is encountered.
b. FRP's should be implemented as time permits.
c. Only ATWS and Loss of Heat Sink FRP's should be implemented.
d. Transition to FRP's is NOT permitted.
ANSWER:
083 (l. 00)
d.
REFERENCE:
EOP User's Guide Rev.
2, Section 5.1.6, p.
[4.1/4.1]
000055G011
.. (KA's)
t SENIOR REACTOR OPERATOR Page
QUESTION:
084 (1. 00)
The reactor is at 100% power when several annunciators alarm that would be indicative of a reactor trip but the rods have not dropped.
The RO attempts a manual reactor trip and manual insertion of the control rods.
None of these actions 'is successful.
Which ONE of the following should-be performed in accordance with FRP-S.1?
a. Deenergize 1D2, lE2 (rod drive M/G Set power supplies)
b. Emergency borate c. Actuate SI d. Allow the 'RCS to heatup thereby inserting negative reactivity ANSWER:
084 (1. 00)
b.
REFERENCE:
FRP-S.1 EOP-LP-3.15 L.O A.2
[4.5/4.5]
000029G010
.. (KA's)
'
SENIOR REACTOR OPERATOR Page
QUESTION:
085 (1.00)
A large break LOCA has resulted in automatic actuation of containment spray.
At this point Path-1 directs the Reactor Coolant Pumps (RCP's)
to be tripped.
Which ONE of the following is the basis for tripping the RCP's at this point?
a.
Delays the onset of two phase flow b. Preempt the RCP's tripping on low pressure because it is assumed that zf containment spray actuates an RCS depressurization is in progress c. Reduces the containment high pressure transient by lowering the energy release rate to containment from forced flow d. Prevents RCP motor bearings from overheating on loss of Component Cooling Water ANSWER:
085 (1.00)
d.
REFERENCE'OG ERG EOP-LP-3.1 L.O. A.8
[4.1/4.2]
000011K314
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
086 (1.00)
You have entered FRP-H.1, Loss of Secondary Heat Sink.
Assuminq normal containment conditions, which ONE of the followinq sets of conditions requires tripping-of all RCP's and immediate initiation of bleed and feed?
Core Exit Thermocouples a. Steady b. Steady c. Increasing d. Lowering Pzr P
(psig)
2225 2300 2225 2000 S/G-A WR Level
20
S/G-B WR Level
20
S/G-C WR Level
- 30
ANSWER:
086 (1. 00)
c ~
REFERENCE:
EOP-LP-3.11 FRP-H.1 Foldout
[4.4/4.5]
000054A104
.. (KA's)
SENIOR PEACTOR OPERATOR Page
QUESTION:
087 (1.00)
All S/G levels decrease to less than 54 due to a loss of feedwater event and the required reactor trip does NOT automatically occur.
The operating crew enters FRP-S.1 after step 1 of PATH-1 and the reactor is locally tripped 'successfully.
Upon completion of FRP-S.1, the STA informs the SRO that a
RED path exists on the Heat Sink status tree.
The operating crew should perform which ONE of the following next?
a. Return to the start of Path-1 and complete immediate action steps of Path-1 b. Return to Path-1 at step 2 and complete immediate action steps of Path-1 I
c. Implement FRP-H.1 after Path-1 immediate actions have been completed d. Implement FRP-H.1 and verify actuation of safeguards equipment ANSWER: '87 (1.00)
d.
REFERENCE:
EOP Users Guide-Rev.
2, Section 5.2.3, p.
(3.4/3.3]
000054G011
..(KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
088 (1.00)
Which ONE of the following is NOT one of the bases for securing the.
RCP's in FRP-H.1, Loss of Secondary Heat Sink?
a.
RCP heat addition causes S/G's to dry out sooner b.
RCP operation causes SI pumps to see a higher discharge pressure, lowering SI flow c.
RCP heat addition holds up RCS pressure, hindering bleed and feed d.
RCP operation during bleed and feed could result in pump cavitation ANSWER:
088 (1. 00)
d.
REFERENCE:
EOP-LP-'3.11 p.
12; L.O. A.6.b
[4.4/4.6]
'00054K304
..(KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
089 (1.00)
Which ONE of the following correctly describes the behavior of reactor power following a trip from 100% power?
a.
Prompt drop to middle of Intermediate Range, SUR decreases to constant -1/3 dpm, levels off in Source Range b. Prompt drop to approximately 5% in Power Range, SUR decreases to
'onstant-1/3 dpm, levels off in Source Range c. Prompt drop to approximately 54 in Power Range, SUR decreases slowly throughout Intermediate Range, SUR reaches constant -1/3 dpm in Source Range d. Prompt drop to approximately 50% in Power Range, SUR decreases to constant -1/2 dpm, levels off in Source Range ANSWER:
089 (1.00)
b.
REFERENCE:
EOP-LP-3.15 no L.O.
[3.6/3.9]
000007K104
..(KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
090 (1.00)
Which ONE of the following combinations of the list below is sufficient to verify a reactor trip has occurred per EOP Path 1?
1.2.3.4.5.6.7.
Trip Breaker RTA open Trip Breaker BYA open Trip Breaker RTB open Trip Breaker BYB open Rod bottom lights lit Decreasing reactor flux Power Range channels less than
a. 1,2,5,7 b
1i2g3i6 c. 3,4,5,6 d. 2,3,5,6 ANSWER:
090 (1.00)
c ~
REFERENCE:
[4.3/4.5]
000007A206
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
091 (1.00)
Which ONE of the followinq is the reason for stopping both RHR pumps when a large leak occurs zn the RCS while on RHR cooling?
a.
To prevent pump damage
'from operation with no flow b.
To minimize RCS inventory loss c.
To prepare for swapping RHR Pump suction from the RCS to the RWST d.
To prepare for SI initiation with the CSIP s which will provide core cooling ANSWER:
091 (1. 00)
a ~
REFERENCE:
AOP-LP-3.8; L.O. A.5 AOP-20 1.4.6
[3.9/4.3]
000025K101
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
092 (1.00)
Following a S/G Tube Rupture Path-2 directs the operators to maintain the ruptured S/G water level between 104 and 154 vice 10% to 50ro for a nonruptured S/G.
Which ONE of the following is the reason for maintaining a lower band for a ruptured S/G?
a.
To minimize the generation of contaminated secondary water which must be disposed of b.
To prevent overfill of a ruptured S/G, limiting the radiological release potential c.
To limit RCS cooldown d.
To prevent condensing the=S/G steam space which would raise S/G pressure ANSWER:
092 (1. 00)
b.
REFERENCE:
EOP-LP-3.2 p.
6, 13 Rev. 4; L.O. A.10.g EOP Path-2
[4.2/4.5]
000038K306
..(KA's)
\\
SENIOR REACTOR OPERATOR Page
QUESTION:
093 (1.00)
Which ONE of the following is the most preferable manual method 'of determining temperature for subcooling margin if the plant process computer is unavailable?
a.
The average of all core exit thermocouple readings on the Inadequate Core Cooling Monitor b.
The single highest reading core exit thermocouple reading on the Inadequate Core Cooling Monitor c.
The average of all active loop wide range hot leg temperature indicators d.
The single highest reading active loop wide range hot leg temperature indication ANSWER:
093 (1.00)
b.,
REFERENCE:
EOP Users Guide 6.2
[4.2/4.2]
000009A116
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
094 (1.00)
You are in the process of recovering from a reactor trip that resulted in all Reactor Coolant Pumps (RCP's) tripping.
All prerequisites per
,OP-100,
Section 5.1 for starting a pump have been met.
Which pump should be started first and why?
a.
"A" RCP because normal letdown comes from Loop 1 and a spray line taps into Loop 1 ensuring optimal pressure control b.
"B" RCP because normal CVCS charging taps into Loop 2 and flow is desired through this loop to ensure adequate thermal mixing of the colder makeup water c. "B" RCP because the pressurizer surge and spray lines tap off Loop 2 ensuring optimal pressure control d.
"C" RCP because testing has determined this pump develops more flow than RCP's
"A" or "B" ANSWER:
094 (1. 00),
c ~
REFERENCE:
EOP-EPP-005, p.
RCS-LP-3.0; L.O. 1.1.20
[3.6/3.7]
000007A104
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
095 (1.00)
Given the following:
-SGTR has occurred-all RCP's are operating-RCS depressurization is in progress per PATH-2-both spray valves are full open for maximum spray flow Your RO reports that the effectiveness of pressurizer spray in reducing RCS pressure is severely diminished and that depressurizatzon termination criteria will not be met for a considerable period of time due to what appears to be noncondensible gases in the pressurizer.
Which ONE of the following courses of action should be performed in accordance with the EOP's?
a. Continue with the depressurization using pressurizer spray until the termination criteria are met since Path-2 still considers spray "available" b. Secure spray and cooldown the RCS further with the steam dumps while limiting injection flow to lower pressurizer level and pressure c. Secure spray and initiate Auxiliary Spray in conjunction with further RCS cooldown with the steam dumps d. Declare pressurizer spray unavailable and open a pressurizer PORV to continue the depressurization ANSWER:
095 (1. 00)
d.
REFERENCE:
EOP-Path-2 step
EOP Users Guide 6.15
[4'/4.1]
000038A104
.. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
096 (1.00)
Which ONE of the following describes the status of the CVCS system after a complete loss of Instrument Air pressure with no operator action?
a.
Letdown orifices and isolations SHUT, letdown directed to VCT (1CS-20),
CSIP taking suction from VCT, charging flow path normal (1CS-231 OPEN),
RCP seal injection flow (1CS-243 OPEN)
b. Letdown orifices and isolations OPEN, letdown directed to RHT f1CS-20),
CSIP taking suction from RWST,'charging flow path isolated (1CS-231 SHUT),
RCP seal injection flow (1CS-243 OPEN)
c. Letdown orifices and isolations SHUT, letdown directed to RHT
[1CS-20),
CSIP taking suction from RWST, charging flow path isolated (1CS-231 SHUT),
RCP seal injection flow (1CS-,243 OPEN)
d. Letdown orifices and isolations SHUT, letdown directed to VCT (1CS-20),
CSIP taking suction from VCT, charging flow path normal (1CS-231 OPEN),
RCP seal injection SECURED (1CS-243 SHUT)
ANSWER:
096 (1. 00)
REFERENCE:
AOP-LP-3.10; L.O. I.A.3 AOP-017 attach.
[2.9/3.3]
000065A208
.. (KA's)
SENZOR REACTOR OPERATOR Page
QUESTION:
097 (1.00)
Which ONE of the following groups contains events that will EACH independently cause RCS pressure to INCREASE during solid plant operations?
a.
Letdown isolates; RCS temperature increases; charging flow increases; pressurizer heaters energized b.
Letdown isolates; RCS temperature decreases; charging flow increases; pressurizer heaters energized c. Letdown is established; RCS temperature decreases; charging flow decreases; pressurizer heaters energized d. Letdown isolates; RCS temperature increases; charging flow increases; pressurizer spray is actuated ANSWER:
097 (1. 00)
a
~
REFERENCE:
AOP-LP-3.11, II.J.5; L.O. A.7
[4.0/4.1]
000027A211
".. (KA's)
SENIOR REACTOR OPERATOR Page
QUESTION:
098 (1.00)
Which ONE of-the following load sets receives a start signal from the EDG Sequencer on a Program A, "Loss of Off-Site Power",
loading?
a.
RHR Pumps b.
RAB Exhaust Fans (E.6)
c. Emergency Lighting d. Reactor Shield and Support Cooling Fans (S-2 and S-4)
ANSWER:
098 (1. 00)
d.
REFERENCE:
SD 155
~ 02'
~ 1 ~ 1~
~ 1
~ 4 I
[3.8/3.9]
000056A247
.. (KA's)
QUESTION:
099 (1.00)
Which ONE of the following is a function of the FHB Auxiliary Crane?
a. 'Removal and installation of the spent fuel cask removable barrier b. Movement of fuel assemblies from the transfer system to the reactor c.
Movement of the spent fuel cask from the decontamination enclosure to the flooded cask loading pool d. Changing out fuel assembly thimble plugs and RCCA's
SENIOR REACTOR OPERATOR Page
ANSWER:
099 (1.00)
a ~
REFERENCE:
SD 115 4 ~'2g4
~ 3g4
~ 4
[2.9/3.5]
000036K201
..(KA's)
QUESTION:
100 (1.00)
The plant is at steady-state 354 power with the Unit Auxiliary Transformers (UAT's) supplying AC electrical power when local indications show unstable off-site voltage and frequency. It is then reported that loss of the grid is imminent.
Which ONE of the following actions is required to be taken?
a. Manually start one EDG, parallel onto the Emergency AC bus and secure that UAT b. Manually start both EDG's, then sequentially parallel each one onto it's Emergency AC bus and secure the respective UAT c. Energize the Emergency AC buses by alternately opening their supply breakers and observing that each EDG and sequencer starts d. Trip the unit output breakers and verify the EDG's and sequencers start ANSWER:
100 (1. 00) SENIOR REACTOR OPERATOR Page
REFERENCE:
AOP-28 AOP-LP-3.14 L.O. A.2
[3.7/3.9]
000056G010
.. (KA's)
.(*****+****
END OF EXAMINATION**********)
TEST CROSS REFERENCE Page
SRO Exam PWR Reac tor Or an i z ed b
Quest ion Numb e r QUESTION VALUE REFERENCE 001 002 003 004 005 006 007 008 009 010 011 012 013 014 015 016 017 018 019 020 021 022 023
024 025 026 027 028 029 030 031 032 033 034 035 036 037 038 039 040 041 042 043 044 045 046 047 048 049 1.00 1.00 1.00 1.00 1.'0 1. 00 1. 00 1. 00 1. 00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 20656 20657 20658 20663 20664 20671 20673 20684 20688 20694 26029 26034 26051 26053 26060 26072 26083 26086 26088 9000540 9000541 9000542 9000543 9000544 9000545 9000546 9000547 9000548 9000549 9000550 9000551 9000552 9000553 9000554 9000555 9000556 9000557 9000558 9000559 9000560 9000561 9000562 9000563 9000564 9000565 9000566 9000567 9000568 9000569
TEST CROSS REFERENCE Page
SRO Exam PWR Reac tor Or anized b
Quest ion Number QUESTION VALUE REFERENCE 050 051 052 053 054 055 056 057 058 059 060 061 062 063 064 065 066 067 068 069 070 071 072 073 074 075 076 077 078 079 080 081 082 083 084 085 086 087 088 089 090 091 092 093 094 095 096 097 098 1. 00, 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00
~
1.00 1.00 F 00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00
'.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 9000570 9000571
, 9000572 9000573 9000574 9000575 9000576 9000577 9000578 9000579 9000580 9000581 9000582 9000583 9000584 9000585 9000586 9000587 9000588 9000589 9000590 9000591 9000592 9000593 9000594 9000595 9000596 9000597 9000598 9000599 9000600 9000601 9000602 9000603 9000604 9000605 9000606 9000607 9000608 9000609 9000610 9000611 9000612 9000613 9000614 9000615 9000616 9000617 9000618
TEST CROSS REFERENCE Page
SRO Exam Or anized b
PWR Reactor Question Number QUESTION VALUE REFERENCE 099 1.00 9000619 100 1.00 9000620 100.00 100.00
I~
TEST CROSS REFERENCE Page
S R
0 r
E a n xam ized PWR R
b K A e a c t o r Grou PLANT WIDE GENERICS QUESTION VALUE PWG Total 029 021 024 022 020 034 033 035 030 025 031 026 032 027 028 011 023 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00'.00 1.00 17.00 194001A102 194001A103 194001A103 194001A116 194001A116 194001K101 194001K101 194001K101 194001K102 194001K102 194001K102 194001K103 194001K103 194001K103 194001K104 194001K109 194001K116 PLANT SYSTEMS Group I QUESTION VALUE 038 037 049 012 042 041 044 039 036 048 045 040 007 043 047 046 006 050 009 1.00 1.00 1.00 1.00 1.00 1.00 1.00 F 00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 001000K409 001050A201 003000G010 003000G014 004010K101 004010K612 004020K403 013000K401 014000A103 015000K407 015000K501 015020G006 017020K401 017020K403 026000K405 026020K401 061000K406 061000K601 063000K301
TEST CROSS REFERENCE Page
SRO Exam PWR Reactor Or an i z ed b
.
KA Grou PLANT SYSTEMS Group'I QUESTION VALUE PS-I Total Group II 19. 00 QUESTION VALUE 013 060 059 014 063 062 053 061 057 056 055 015 058 052 051 054 064 PS-II Total Group III 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1. 00 17.00 002000G011 002000K602 006020A304 006030K101 010000A107 010000K403 011000A210 012000K604 033000A203 033000K401 035000G011 035010A205 062000K403 064000A304 064000G011 064000K402 073000G005 QUESTION VALUE 065 066 008 067 PS-III Total 1. 00 1.00 1.00 1.00 4.00 008000K401 041020K603 045050K101 076000K116 PS Total 40. 00 EMERGENCY PLANT EVOLUTIONS Group I
TEST CROSS REFERENCE Page
S
R 0 E
r a n xam ized P WR'
K Reactor A
G r o u EMERGENCY PLANT EVOLUTIONS Group I QUESTION
.VALUE 016 001 070 076 073 078 085 071 080 010 082 084 069 075 017 077 083 081 018 019 068 072 079 074 EPE-I Total Group II 1. 00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 000001A205 000001A205 000003G009
'00005A201 000011A208 000011K313 000011K314 000024G007 000024K301 000026K303 000029A209 1.00 1.00 1.00 1.00
,1. 00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 24.00 000029K306 000029K311 000040G010 000055A204 000055G011
= 000057K301 000067A203 000068A204
000069G007 000069K301 000074G012 000074K311 1.00 000029G010 QUESTION VALUE 094 090 089 093 004 091 097 095 003 092 086 087 088 002 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00
.1. 00 1. 00 1.00 000007A104 000007A206 000007K104 000009A116 000009K321 000025K101 000027A211 000038A104 000038A136 000038K306 000054A104 000054G011 000054K304 000061K302
TEST CROSS REFERENCE Page
S R 0 Exam P
W R Reactor Or an i z ed b
KA. Grou EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE 096 005 EPE-II Total Group III 1.00 1.00 16.00 000065A208 000065G010 QUESTION VALUE 099 098 100 EPE-III Total 1.00 1.00 1.00 3.00 000036K201 000056A247 000056G010 EPE Total 43.00 Test Total 100.00
ENCLOSURE 3 SIMULATOR FIDELITY REPORT Facility Licensee:
Facility Name:
Carolina Power and Light Harris Nuclear Plant Facility Docket No.:
50-400 Operating
'Tests Administered on:
January 28 through 30, 1992 This form is to be used only to report observations.
These observations do not constitute, in and of themselves, audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b).
These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations.
No licensee action is required solely in response to these observations.
During the conduct of the simulator portion of the operating tests,'he following items were observed
ITEM DESCRIPTION DEH CCW Load swings of 40-80 Mw were observed during downpower, this is much higher than what would be observed in the plant.
Failure of CCW pumps to automatically start could not be simulated.
Pumps could only be failed off or on.