ML20207J493
| ML20207J493 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/01/1999 |
| From: | Michael B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-348-98-300, 50-364-98-300, NUDOCS 9903160299 | |
| Download: ML20207J493 (1) | |
See also: IR 05000348/1998300
Text
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March 1,1999
NOTE TO: NRC Document Control Desk
Mail Stop 0-5-D-24
FROM:
8uJ WN
Beverly Michael, LicensingAssistant, Operator Licensing and Human
Performance Branch, Division of Reactor Safety, Region 11
SUBJECT: OPERATOR LICENSING RETAKE EXAMINATIONS ADMINISTERED
ON MARCH 13,1998 AT THE FARLEY NUCLEAR PLANT
DOCKET NOS. 50-348 AND 50-364
On March 13,1998, Operator Licensing Examinations were administered at the referenced
facility. Attached, you will find the following information for processing through NUDOCS
and districation to the NRC staff, including the NRC PDR:
Item #1 - a) Facility submitted outline and initial exam submittal, i
designated for distribution under RIDS Code A070. t
b) ' As given operating examination, designated for distribution under
RIDS Code A070.
Item #2 - Examination Report with the as given written examination attached,
. designated for distribution under RIDS Code IE42.
' Attachrnents: As stated
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9903160299 990301
PDR ADOCK 05000348
V PM
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April 6. 1998
Southern Nuclear Operating Company, Inc. !
ATTN: Mr. D. N. Morey
Vice President
P. O. Box 1295 i
Birmingham. AL 35201 l
SUBJECT: EXAMINATION REPORT 50-348.364/98-300
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Dear Mr. Morey:
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On March 12. 1998, the facility administered a written re-take examination to
one Senior Reactor Operator (SRO) candidate. On March 13. the NRC l
administered a Job Performance Measure (JPM) re-take examination to another j
SRO candidate. both of whom had re-applied for licenses to operate the Farley
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Nuclear Flant (FNP) Units. At the conclusion of the examination, the chief !
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examiner discussed the examination and preliminary findings with those members
of your staff identified in the enclosed report (Enclosure 1).
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A copy of the written examination questions and answer key, as noted in i
Enclosure 2. was provided to members of your training staff at the conclusion
of the examination.
The facility comments regarding the written examination are included in this
report as Enclosure 3. The NRC's Resolutions to these facility comments are
included as Enclosure 4.
The chief examiner concluded that the candidates' performance on the written
and the JPM re-take examinations were satisfactory.
In accordance with 10 CFR 2.790 of the Commission's regulations. a copy of
this letter and the enclosure will be placed in the NRC Public Document Room.
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DISTRIBUTION CODE
1E42
AB C QR D260 Sffj7, ,
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. SNC 2
Should you have any questions concerning this examination, please contact me
at (404) 562-4638'.
Sincerely.
L
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Thomas . Peebl . Chief l
Operato Licensi g and Human l
Perfo mance Branch l
Division of Reactor Safety
Docket Nos.. 50-348, 364
License Nos.. NPF-2. 8
Enclosures: 1. Examination Report
2. Written Examination !
3. Facility Comments
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4. NRC's Resolution to Facility Comments ;
cc w/encls:
M. J. Ajluni. Licensing j
Services Manager. B-031
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Southern Nuclear Operating l
Company. Inc.
42 Inverness Center Parkway
Birmingham AL 35201-1295
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R. D. Hill, Jr.
General Manager. Farley Plan; '
Southern Nuclear Operating
Company. Inc.
P. O. Box 1295
Birmingham AL 35201-1295
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l (cc w/encls cont'd - See page 3)
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(cc w/encls cont'.O ~
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J. D. Woodard
l Executive Vice President l
Southern Nuclear Operating i
Company Inc. i
'
e P. O. Box 1295
Birmingham, AL. 35201-1295
State Health Officer
Alabama Department. of Public Health !
434 Monroe Street !
Montgomery AL 36130-1701 i
M. Stanford Blantcn '
Balch and Bingham Law Firm )
P. O. Box 306 )
1710 Sixth Avenue North j
Birmingham, AL 35201 i
Chairman l
Houston County Commission
P. O. Box 6406
Dothan.'AL 36302 -
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Distribution w/encls: ;
P. Skinner. RII l
R. Carrion. RII ;
J. Zimmerman. NRR !
PUBLIC l
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NRC Resident Inspector )
U. S. Nuclear Regulatory Commission
7388 N. State Highway 95
Columbia. AL 36319
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OFFICIAL RECORD COPY DOCLHENTNAME: I:\0Sil\LA\far\ fig 3300.RfA !
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION II
Docket Nos.: 50-348, 50-364 '
Report Nos.: 50-348.364/98-300
Licensee: Southern Nuclear Power Company
Facility: Farley Nuclear Plant
Location: Columbia. AL
Dates: March 12 -13. 1998
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Examiners: #m _ . .
RonaT6'F. Aiello. Chief License Examiner
Approved by: Thomas A. Peebles. Chief
Operator Licensing and Human Performance Branch
Division of Reactor Safety ,
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. EXECUTIVE SUMMARY i
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Farley Nuclear Plant I
bRCExaminationReportNo. 50-348, 364/98-300 !
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On March 12 and 13. 1998 The NRC conducted an announced operator licensing
written and JPM re-take examination in accordance with the guidance of
Examiner Standards. NUREG-1021. Interim Revision 8. These examinations I
implemented the operator licensing requirements of 10 CFR S55.43 and 55.45. ,
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Ooerations
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One SRO candidate received a written re-take examination. This i
examination was administered by the facility on March 12. 1998.
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One SRO candidate received a JPM re-take examination. The NRC
administered this examination on March 13, 1998.
. Candidate Pass / Fail
SRO R0 Total Percent
Pass 2 0 2 100%
Fail 0 0 0 0%
. The examiner concluded that the candidates' performance on the written
and JPM examinations were satisfactory. (Section 05.3).
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Report Details
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l Summarv of Plant Status l
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During the period of the examination, both units were in Mode 1. '
I. Doerations
05 Operator Training and Qualifications j
05.1 General Comments
The Licensee developed operator licensing initial written and JPM re- 1
take examinations, under the guidance of the NRC, to be administered by '
the faulity and the NRC respectively under the requirements of an NRC j
security agreement, in accordance with the guidelines of the Examiner !
Standards (ES). NUREG-1021. Interim Revision 8. One SRO upgrade re-take i
applicant received and passed the written examination. One SRO upgrade
re-take applicant received and passed the JPM operating examination.
05.2 Pre and Post-Examination Activities
a. Scooe
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The NRC reviewed the licensee's examination submittal using the l
criteria specified for examination development contained in NUREG
1021 Interim Rev 8.
b. Observations and Findinas
The licensee developed the SRO written and JPM retake examinations.
All materials were submitted to the NRC on time. The Chief Examiner
reviewed. modified and approved the examination prior to
administration. The NRC conducted in-office and onsite preparation
prior to examination administration. The examination met the
criteria set forth in NUREG 1021 Interim Rev, 8.
The written examination was reviewed and approved in the regional
office. Four of the written examination questions contained
signif nant technical errors that resulted in either question
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deletici or answer modification (See enclosure 3 for details), These
types of errors should have been identified during the facility's
technical and managerial reviews.
The NRC conducted the preparation visit for the operating exam on
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March 12. 1998. One JPM set was validated. There were no direct !
look-up JPM follow-up questions. Most of the JPM follow-up questions ;
were either comprehensive or analytical. ;
c. Conclusion
The NRC concluded that the facility had placed emphasis on ensuring
that the examination was technically accurate with a few exceptions
(see Enclosure 3), and discriminating.
05.3 Examination Results and Related Findinas. Observations. and Conclusions
a. General
The chief examiner reviewed the results of the written and JPM
examinations. The overall performance of the candidates was
satisfactory. The chief examiner identified no discrepancies.
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V. Manacement Meetinos !
XI. Exit Meeting Summary
On March 13. 1998, the chief examiner discussed the examination results
with the Operations Training Supervisor. Dissenting cciaments were not i
received from the licensee. No proprietary information was identified. ;
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PARTIAL LIST OF PERSONS CONTACTED ,
Licensee- .
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- B. Badham. Supervisor Safety Audit Engineering Review '
W. Coggins Performance Modification and Maintenance
Support Supervisor ;
- P. Crone. Engineering Support Performance Review Supervisor
- J. Deavers. Senior Plant Instructor
- S. Fulmer. Training and Emergency Preparedness Manager
D. Grissette. Operations Manager
- D. Hall. Operations Instructor
- R. Hill. FNP Plant Manager !
C. Nisbitt. Assistant Plant manager. Support ;
- W. Oldfield. Nuclear Operations Training Supervisor
- J. Powell. Senior Plant Instructor
- G. Waymire. Technical Manager -
NRC
- B. Caldwell. Resident Inspector
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ENCLOSURE 3
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FACILITY RECOMMENDATIONS FOR CHANGES TO EXAMINATION !
OUESTIONS i
Question Number 1:
Change the correct answer to "a". The question stem gives the condition that
Pu, fails high and asked for the initial response of the rod control system. ;
During the time frame that impulse pressure is failing, the power mismatch
circuit (see attached ) will be causing a maximum rod speed signal based on !
the large difference in the rate of change of NI-44 and Pug. This rate of !
change signal will be brief and then rod speed will:be determined by the i
difference between median T;y and 1;g (generated from Pu,,). The question
developer failed to take into account the momentary difference in the rate of ;
change and based the original answer on the 1;y-1;g difference that would i
exist.
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Given the following plant conditions:
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Loop B Tivg channelis 576 degrees F.
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Loop C Tavg channelis 572 degrees F. .
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. Rod Control System is in auto with control bank D at 215 steps !
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Which one of the following explains how the Rod Control System will initially respond if the ;
selected Pimp pressure failed high7 '
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a. Rods will step out at 72 steps / minute.
b. Rods will step out at 48 steps / minute. !
c. Rods will step'out at 8 steps / minute. l
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d. Rods will not move. l
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ANSWER: c. )
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KA: 001A1013.8/4.2 LEVEL: ANALYSIS l
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REFERENCE: OPS-52201E, pg.10-13 i
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LEARNING OBJECTIVE: 052201E13
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HISTORY: 052201E05006 was used on one the candidates audit exams The stem and I
- distracters a, b, and c, have been modified.
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JUSTIFICATION:
a ' Rods would move out at this speed if the Tref were not high limited,
b. Rods would move out at this speed if average value of Tavg was used instead of the median
value.
c. Rods will move out at this speed because the out put of the Rod-Speed Programmer is above
1* but not greater than 3*.
. d. Rods would not move if the Temperature Mismatch Channel out put was from "B" loop Tavg
instead of the median Tavg.
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DC Hold Cabinet (Ficure 4)
A failure in a power cabinet may require replacement of a printed circuit card, fuse, or other
component. To avoid the possibility of dropping rods during maintenance and to avoid the need for
an extemal power source, each power cabinet contains three switches 'ised to energize any one of
the three groups of stationary gripper coils from a separate 125n0-volt DC power source. Placing
more than one group in entire system on hold bus may result in overloading of supply. !
This power source is the DC hold cabinet. The 125V DC supply is used to assure latching
of the stationary grippers. The 70V DC supply is used to hold tlie grippers without overheating the
coils.
Control Rod Drive Mechanism I
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The CRDM is z.three-coil, electromagneticjack that raises and lowers a 144-inch drive rod, !
which attar'2es to the control rod assemblies. Tlie three coils, mounted outside the pressure
housing, actuate armatures contained within the housing. The movable and stationary gripper
armatures operate latches that grip a grooved drive rod. The stationary gripper latches are used to
hold the drive rod in position. The movable gripper latches, which are raised and lowered by the
lift coil armature, are used to raise and lower the drive rod. Each step of the mechanism moves the
drive rod 5/8 inch. Refer to the Reactor Vessel and Core Components lesson for the design and
construction details of the CRDM.
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OPERA TIONS
Instrumentation and Controls
Reactor Control Unit (Fieure 5)
The reactor control unit consists of two channels: (1) the power mismatch channe.' and (2)
the temperature mismatch channel. The power mismatch channel provides an error signal
whenever there is a rate of change between turbine power and reactor power. (During constant
power operation, the error signal will be zero even if turbine power and reactor power are not
equal.) The temperature mismatch channel produces an error signal proportional to the deviation
between median T ., and Pimp generated Tnr. (The error signal will be zero only if the difference
between T.,, and Turis zero.) This is the normal control channel.
70 OPS-402041/52201E
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The error signals produced by these channels are summed and touted to a bistable and a
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function generator. hhe bistable determines the direction of rod motion, and the function i
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determines the rod speed. The resultant output will be sent to the logic cabinets.
Temoerature Mismateh Channel
The temperature mismatch channel receives inputs of median T.,, and turbine first stage i
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impulse pressure (Pmp). Prior to entering a differential amplifier, the T. g signal passes through a
lead / lag card for dynamic conditioning. The lead / lag card provides dynainic compensation by
producing an output that anticipates the actual plant T,,g when T.,g is changing.
On a ramp up in T.,g, the output of the lead / lag card will be the value of actual T.,g at som:
future point in time. This compensates for the delay between the time when temperature begins to
increase in the reactor and the time when the increase will actually be sensed by the resistance
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temperature detectors (RTDs) in the loops.
The Pim, signal, a measure of turbine load, feeds into a function generator. The function
generator creates a Tnr signal programmed to vary as a function of plant load. Again, for purposes l
of discussion, the program is 547 F at zero-percent power to 575 F at 100-percent power. The Tnt
signal passes through a lag circuit for dynamic compensation prior to entering the different ial
amplifier.
The T,v, and T,erinputs connect to the differential amplifier, which performs the following
function:
Tem >r = (T,er- T .g)
The Tem >r signal will be summed with the Pom>, signal from the power mismatch channel.
Power Mismatch Channel
The power mismatch channel receives inputs from nuclear power (N-44) and Pimp. Pimp is
conditioned to produce a tubine power signal that may be compared with the nuclear power signal
from N-44. When compared in a differential amplifier, nuclear power and turbine power produce
an error output signal equivalent to the difference between turbine and reactor power multiplied by
a gain.
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The output of the differential amplifier supplies a signal to a derivative card. This card
produces an output c * when the turbine-reactor power deviation is changing. When the deviation
is constant, the output of the derivative card, and, therefore, the output of the power mismatch
circuit, equals zero.
Any output obtained from the derivative card enters a function generator, which serves as a
non-linear gain unit. A small input to the unit will be amplified with a gain of only 0.24, resulting
in little rod motion. However, if the input is greater in magnitude, the gain becomes 1.2, lending
greater weight to the error signal and resulting in increased rod motion.
He output of the non-linear gain unit enters a variable gain unit. The variable gain unit
varies the gain applied to the error signal inversely with turbine power. The variable gain unit
compensates for the fact that a step of rod motion produces a greater change in power at high power
levels than at low power levels. Therefore, the power error (Pm) signal must be reduced as power
increases to reduce the rod motion at higher power levels. The variable gain is accomplished by
dividing the error signal by the output from the power compensation unit. The power compensation
unit generates a function that varies inversely with Pimp.
The output of the variable gain unit, a P.. signal, inputs to a summing unit to be summed
v;ith the Tmsignal from the temperature mismatch channel. He output of the variable gain unit is
provided with a defeat switch, which is located in control cabinet eight of the 7300 cabinets, along
with the rest of the reactor control unit. This switch, operated by the I&C deputment using
procedures under their control, allows the power mismatch channel to oe isolated trom the rest of
the reactor control unit for calibration or maintenance purposes. I&C procedures and the PLS l
document require the rod control system to be in manual control any time this switch is open. If the
rod control system were operated in automatic with the mismatch channel defeat switch open, the
rod control system would be without the benefit of the anticipatory response provided by this
channel, causing a possible improper response of the system.
The output of the summing unit, which can either be positive or negative, provides an input to
the rods in/out bistable and a function generator. The rods in/out bistable provides the signal to !
direct rod motion (in or out). The polarity of the input signal to the bistable will dictate the
direction the rods are to move. If the input signal exceeds the output setpoint in the positive
direction, a rods-out command will be genemted. The output serpoints equate to 1.5 F
12 OPS-402041/52201E
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temperature error. The rod motion command will reset at * 1"F. This 0.5 F lockup will prevent -
unnecessary rod motion near the bistable output setpoint.
The function generator determines the rod speed based on the magnitude of the enor signal.
The rod speed varies from 8 steps per minute (0 to * 3"F error) to 72 steps per minute (* 5 F
error). The rod speed varies linearly from eight steps per minute to 72 steps per minute ( 3 to
5"F enor).
Rod Control System (Ficure 6)
Bank Selector Switch (BSS)
The BSS has eight positions designated as follows:
1. SBA
2. SBB
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3. MAN
4. AUTO
5. CBA
6. CBB
7. 'CBC
8. CBD
The position of the BSS is sensed by several components in the logic cabinet. The BSS
position determines the speed input to the pulser, selects the direction input to the master cycler,
and provides the bank selection input to the bank overlap unit (BOU). This all takes place in the
logic cabinet.
A rod speed meter on the MCB indicates calculated rod speed from the reactor control unit.
l Since speed signals are always being calculated, even with no rod motion, the meter always
indicates some speed. The indicated rod speed depends on the control mode selected by the BSS.
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r - - - POWER-MISMA~CH
- - - - - -- RATE CHANNEL ,
TEMPERATURE-MISMATCH CHANNEL
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NUCLEAR POWER ' P,yp I
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446 WT H 447 T,yg A IG A * T,yc C
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SELECTOR
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UNEAR
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ROD-SPEED PROGRAMER
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CONTROL-ROD SPEED AND DIRECTION SIGNAL
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REACTOR CONTROL UNIT
FIGURE 3
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Question Number 5:
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The answer to this question should be "a". The stem of the question states
that the Digital Rod Positioning Indication (DRPI) experiences a loss of power
to the Data A cabinet. The following is an excerpt from DRPI lesson material
(attached):
Half Accuracy: The system will still function with either data bank
inoperable but with reduced or half accuracy. Table 4 of OPS-52201F. reflects
! the accuracy available with Data A out of service. The central control cards
will not receive any information from Data A coils. At three steps even
though a Data A coil has been penetrated. data from the detector encoder card i
is inhibited, so no knowledge of this is received by the central control card. l
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It assumes zero coils have been penetrated until a Data B coil is penetrated. l
At nine steps, the first Data B coil will be penetrated. The central control
cards now have information of one coil being penetrated. When either data
- bank is inoperative, the information from the operating data bank is doubled.
The central control cards now assume that two total coils have been
penetrated, and the indication will display 12 steps. The worst case
indication occurs at nine steps where the rod may be plus nine steps or minus
three steps. Plus or minus one (+/-1) must be added to this for manufacturing
tolerances and temperature changes, providing an accuracy of plus 10 minus
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four (+10- 4) accuracy when using Data B only. The accuracy for Data (A)
failed ir / *10 -4) not +4 -10 as the question indicates (see justification for
distracte ). This accuracy would make answer "a" correct. Answer "b" is
incorrect cause 156 is outside the -4 accuracy for group 1. "c" is incorrect
because 150 is beyond the -4 accuracy for group 1. and "d" is incorrect
because 150 is beyond the -4 accuracy for group 2. This error occurred due to i
the exam developer writing the answer based on data B being failed, when the
stem actually specifies data A as the failed channel.
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remain at six steps until a second coil has been penetrated at nine steps. Now, with the rod
somewhere between nine steps and 15 steps, the indication will show 12 steps. This means that
actual rod position can be as much as plus or minus three ( 3) steps from indicated position.
(Table 3 shows this relationship.) '
As can be seen from Table 3, when the rod is at three steps, the coil at three steps may or
may not have been penetrated enough to make it change state. In either case, the indication will
be off by three steps. In addition to the three steps inaccuracy, one additional step must be added
to the inaccuracy to account for manufacturing tolerance of th5 coils and tube, the placement of
!
the coils on the tube, and the expansion or contraction of the tube with temperature changes. The
final full accuracy of the system then becomes plus or minus four (14) steps.
HalfAccuracy
The system will still function with either diita bank inoperable but with reduced or half
accuracy. Table 4 reflects the accuracy available with Data A out of sersice. The central control
i
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cards will not receive any information from Data A coils. At three steps, even though a Data A
coil has been penetrated, data from the detector encoder card is inhibited, so no knowledge of
.
this is received by the central control card. It assumes zero coils have been penetrated until a
Data B coil is penetrated. At nine steps, the first Data B coil will be penetrated. The central
control cards now have information of one coil being penetrated. When either data bank is
inoperative, the infonnation from the operating data bank is doubled. The central control cards
now assume that two total coils have been penetrated, and the indication will display 12 steps.
The worst case indication occurs at nine steps where the rod may be plus nine steps or minus
three steps. Plus or minus one ( 1) must be added to this for manufacturing tolerances and J
temperature changes, providing an accuracy of plus 10 minus four (+10 -4) accuracy when using
Data B only.
Table 5 illustrates the accuracy received if Data B has had a failure. When the first Data
A coil is penetrated, the central control cards double this information. This means that with as I
~
low as three steps, the indication can read 12 steps or still read zero steps. After adding the plus
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8 OPS-52201F
s
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QUESTION No. 005: .
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Given the following plant conditions:
-
Unit 2 is at 50% power.
l -
Control Bhik D rods are :
- Grciup 1 at 161
. Grcup 2 at 160
If the Digital Rod Position Indication System (DRPI) experiences a loss of power to the Data "A"
cabinet, which one of the following Control Bank D DRPI indications are within the limitations
ofDRPI:
Group 1 Group 2
a. 168 162
b. 156 168
c. ~_50 162
d. 162 150
ANSWER: d. -
KA: 014A202 3.1/3.6 LEVEL: ANALYSIS
REFERENCE: OPS-52201F, pg. 8
_
LEARNING OBJECTIVE: 052201F09
HISTORY: New
JUSTIFICATION:
a. With Data "A" failure accuracy will be +4 and -10,168 on group 1 is outside the accuracy
range .
b. Both group 1 and group 2 are outside the accuracy range.
c. Group 1 is outside the accuracy range of DRPI but is plausible if candidate only remembers
12 steps of tech specs as the accuracy.
d. 162 is within the +4 limit and 150 is within the -10 limit.
-
, . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ __ _ . . _ _ . _
-
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Question Number 13:
i
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Delete the question due to no correct answer. Answer "a" is incorrect
'
because the trip of both main feedwater pumps signal is an auto start signal
for the motor driven auxiliary feedwater (MDAFW) pumps only, it does not start
the turbine driven auxiliary feedwater (TDAFW) pump. Answer "b" a safety
injection signal is also an auto start signal for the MDAFW pumps only, it
does not start the TDAFW pump. Answer "c" steam generator low level is an
alarm signal only, the actual automatic start signal is steam generator
low-low level. Answer "d" the AMSAC signal is not active in this case
because power has been below 40% for longer than 240 seconds (see attached).
The validity of the examination outline is not affected by this deletion
because there was another question regarding the auxiliary feedwater system
and there were 18 other questions in this group to evaluate required knowledge
and ability.
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, . QU. ESTION No. 013: .
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l Unit I has been holding at 33% power for the last 2'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, which one of the following signals
l
[ will result in the Auto start of the Turbine Driven AFW pump? !
- i
a. Trip of both main feedwater pumps. I
s
b. SafetyInjection. .
i
c. steam generator low level. l
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d. AMSAC signal.
ANSWER: d.
KA 061K402' 4.5/4.6 LEVEL: MEMORY
l REFERENCE: OPS-52102H, pg. 9
LEARNING OBJECTIVE: 052102H13
HISTORY: New
JUSTIFICATION: ?
a. Trip of both MFPs will auto start MDAFW Pumps but not the TDAFW Pump.
b. SI signal will auto start MDAFW Pumps but not the TDAFW Pump. .
': c.~ This is tiie only valid signal for these conditions
,
d. The AMSAC signal is not active due to being less than 40% power.
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.. - - - . -- - . _ .
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Operation
i
The MDAFW pumps may be controlled from either the MCB or the HSP. The pumps will
automatically start on any one of the following:
, 1. A steam generator 10-10 level of 25% on Unit 1 (25% on Unit 2) (2/3 level
l
1 instruments in 1/3 steam generators) and no LOSP
2. Both main feed pumps tripped and no LOSP .
l 3. An engineered safety feature (ESF) sequencer signal
4. An LOSP sequencer signal
j 5. AMSAC (2/3 steam generators < 10% level on Unit 1 [< 10% level on Unit 2];
blocked below C-20) {< 40%}
Turbine-Driven Auxiliary Feedwater Pum2
, One TDAFW pun 2p provides emergency feedwater flow to the steam generators if off-site
l
power is unavailable. The seven-stage pump is rated at 700 gpm at 1227 psig. Main steam directly
l
from the steam generator provides the power for the turbine. The pump is located on the non-rad
t
side,100 foot elevation.
The condensate storage tank supplies the TDAFW pump through two locked open isolation
'
, valves and check valve. An altemate supply may be drawn from the service water system through
l
l two motor-operated isolation valves (MOV-3216 and either MOV-3209A or B) and a locked open
manual isolation valve located by the TDAFW pump room. The TDAFW pump, like the MDAFW
,
pumps, has a miniflow linc containing a locked open isolation valve, check valve, and a flow
1
orifice. A bypass line around the miniflow line provides for system performance and pump flow
testing. The bypass line isolation valve is normally locked closed. The miniflow and bypass lines
'
retum flow to the condensate storage tank.
EumoInstrum ntation l
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Flow instmment FISL-3218 provides a low flow alarm on the MCB at 80 gpm. A pump
suction pressure instrument (PT-3217) provides both local and MCB indication as well as a low I
suction pressure alarm on the MCB at 22.5 psig. Pressure instnunent PT-3222 provides both local i
and MCB indication of pump discharge pressure. Pump bearing temperatures alarm on the
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Omniguard panel in the main control room.
8 OPS-40201D/52102H
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. Turbine Operation (Fiewe 3 and 3A) *
Connections on the main steam lines from steam generators B and C supply steam to the
TDAFW pump. Steam flows through two parallel lines into a common line, which feeds the
TDAFW pump. An air-operated isolation valve (3235A and B) located in each line will admit
steam to the TDAFW pump upon receiving a start signal. Each of the valves has an air reservoir
associated with it. These valves are in the main steam valve room.
The air reservoir ensures that on a loss-of-instmment air the respective isolation valve can
be opened. The reservoir may be supplied from either instrument air or the emergency air
compressor. Ifinstmment air pressure falls below 80 psig, the solenoid-operated supply valve to
the air reservoir will automatically close. The valve will automatically reopen when pressure
retums to 80 psig. A low pressure alarm for instmment air will sound on the MCB at 60 psig.
HV-3235A and B are normally closed. However, a warming line keeps the supply piping at
main steam temperature to prevent or minunize the thermal shock during pump starts. The
warming line isolation valves (HV-3234A and B) close on a T-signal and can be controlled
remotely from the BOP panel. This supply of wamung steam condenses in the steam header and as
the level of condensate increases, LCV-3608 opens, draining the condensate to the auxiliary steam
condensate tank.
_
During TDAFW pump operation, the steam passes through steam admission valve
HV-3226, the trip throttle valve MOV-3406, the govemor valve, and the TDAFW pump turbine.
The steam exhausts to the atmosphere.
The TDAFW pump may be controlled from either the MCB or the HSP. The pump
automatically starts on the folining:
1. Steam generator lo-lo level of 25 percent (2/3 level instruments in 2/3 steam
generators)
2. Undervoltage signal of 64.4% on RCP buses (blackout) (1/2 UV relays on 2/3
l buses)
! 3. AMSAC (2/3 steam generators < 10 .:1; blocked below C-20 after 260 secs)
'
Upon receiving a start signal, the steam , ply valves (3235A and B) and the steam
admission valve (3226) will open.
The trip throttle valve and governor valve, integral with the turbine, control the steam flow
9 OPS-40201D/52102H
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Ouestion Number 87:
l
l
Delete the question. The stem of the question ::tates that the date 1/10/98 is
the issue date of TCN 3C. FNP-0-AP-1 paragraph 7.1.1.1 (attached) requires
that the dates for which the change is to be effective be listed in the lower
l right hand corner and a one time change shall be valid for the indicated dates
only and this period shall not exceed 90 days. The information provided in
the stem regarding the effective dates was incomplete in that it only provided
the date issued. Additionally. the candidate requested clarification
l (Facility recommendations enclosure 3) about counting the issue date and wa. l
l
- told yes it counts by the proctor. The incomplete stem information and the l
! answer provided prevented the candidate from having to evaluate when the time I
requirement actually began and could have also misled him in the correct
counting of the 90 day period. The deletion of this question does not affect
the validity of the examination outline because there were four questions
l remaining in this category to sample the required knowledge and ability.
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QUESTION No. 087: .
. You are about to use a System Operating Procedure'that has the following markings:
.
l - The applicable portion of the procedure has been changed by TCN 3C. l
) - In the lows right-hand comer of the page is the statement "One Time Only."
E
- In the lower right-hand comer of the page is written " Issued on 1/10/98."
l Which one of the following is the latest date this TCN could be valid?
a. 1/30/98
. b. 3/10/98
c. 3/30/98
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d. 4/10/98 '
ANSWER: d.
l
KA: G1213.1/3.2 LEVEL: MEMORY
l
REFERENCE: FNP-0-AP-1, pg.15 i
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LEARNING OBJECTIVE: 052303A01 i
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' HISTORY: New l
JUSTIFICATION:
-
a.' When a TCN is within 20 days of the end duration the responsible individual is notified.
I
b. A TCN shall be approved or denied within 60 days ofimplementation.
!
c. An outstanding TCN over 80 days old will be referred to an Assistant General Manager for
disposition or extension.
d. A one time only TCN will not exceed 90 days.
, .. .
-. -_ . _ _ _ . . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ _ . _ _ . . _ _ _ _ . . _ _ _ _ _ .
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06/19/97 13:28:50
ENP-0-AP-1
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Temporary changes shall be documente'd using the Procedure Request Form
(Figum 1). The individual assigned to prepare the temporary change will fill out
Items 1 through 3 of the Procedure Request Form and verify that the procedure or
manual has been screened for 10 CFR 50.59 applicability per paragraph 5.1.
7.1.1 One time only changes
In addition to the TCN, the lower right-hand comer of the replacement
and/or additional page(s) shall show: l
7.1.1.1 The dates for which the change is to be effective.
7.1.1.1.1 The one time only temporary change shall be valid
for the indicated date(s) only and this period shall
!
not exceed 90 days.
7.1.1.2 That this is a one time only change.
l
7.1.2 Temporary changes Required By Plant Conditions
In addition to the TCN, the lower right-hand comer of the replacement
and/or additional page(s) shall show (1) the plant condition for which the l
change is to be effective, and (2) that this is a one time only change.
7.2 Review ofTemporary Changes *
The temporary change will be reviewed by a qualified reviewer as stated in
Section 4 ofthis procedure. The qualified reviewer will complete Item 4 of the
Procedure Request Form and designate:
7.2.1 Any required cross-disciplinary review or PORC review in item 5 of the
Procedure Request Fonn.
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'7.2.2 The temporary chcnge approval authority in Item 6 of the Procedure
Request Form.
7.2.3 The Final appmval authority in Item 7 of the Procedure Request Form.
7.3 - Approval Requirements for Temporary Changes .
!
Temporary changes shall be approved as specified in this paragraph prior to
implementation.
!
7.3.1 The approval authority shall ensure that:
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-15- Revision 35 !
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.
ENCLOSURE 4
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.+
NRC RESOLUTION OF COMMENTS ,
i
1. SR0'Oue ion # 1
l
Comment accepted. The answer key was changed to accept choice "a" as
the correct answer.
I
2. SRO Question # 5
Comment accepted. The answer key was changed to accept choice "a" as l
the correct answer.
3. SR0 Question # 13
Comment accepted. The question was deleted.
l
4. SRO Question # 87
Comment accepted. The question was deleted.
.
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. - . _ _ . _ _ _ _ . .
T
QUESTION No.1: !
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[
l Giv;n the following plant conditions:
! -
Loop A Tavg channelis 575 degrees F.
- # #' *
Loop B Tavg channelis 576 degrees F.
-
l Loop C Tavg channelis 572 degrees F.
-
Rod Control System is in auto with control bank D at 215 steps
Which one of the following explains how the Rod Control System will initially respond if t
select:d Pimp pressure failed high? i
-
a. Rods will step out at 72 steps / minute. ,
I
b. Rods will step out at 48 steps / minute. i
I
c. Rods will step out at 8 steps / minute.
d. Rods will not move.
4
ANSWER: /
t
i
KA: 001 A1013.8/4.2 LEVEL:
'
ANALYSIS
REFERENCE: OPS-52201E, pg.10.-13
\
LEARNING OBJECTIVE: 052201E13 ;
HISTORY:
distracters OS2201E05006
a, b, and c, have been modified. was used on one the candidates audit exams The ste
JUSTIFICATION:
i 1
a. Rods would move out at this speed if the Tref were not high limited.
! b. Rods
value.
would move out at this speed if average value of Tavg was used instead of the m
c.
Rods will move out at this speed because the out put of the Rod-Speed Programmer is above
1 but not greater than 3 i
d.
Rods would not move if the Temperature Mismatch Channel out put was from "B"
instead of the median Tavg.
I loop Ta
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DISTRIBUTION CODE
IE42
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QUESTION No. 003:
l
Given the following plar:t conditions:
-
Unit 1 is at 100% power, with CVCS aligned for normal operation
-
One orifice is olation valve is in service.
-
VCT levelis 32%.
-
All controls nre in automatic.
- LT-112, VCT level transmitter, fails high.
Which one of the follow:ng describes the final actual VCT level? (Assume no operator action.)
a. Increases to 100% (full).
l- b. Increases to 71% and stabilizes.
l
! c. Cycles between 20% and 40% due to auto-makeup.
.
l
l d. Decreaces to 0% (empty).
l ANSWER: c.
l
j KA: 004K605 2.5/2.5 LEVEL: MEMORY
l
' REFERENCE: OPS-52101F, Table 2, pg. T-2
L LEARNING OBJECTIVE:. 052101F09
l
HISTORY: 052101F09031 with 3 of 4 distracters changed..
JUSTIFICATION: i
l !
! a. Level will initially decrease as LCV-115 diverts. Level will decrease until auto Makeup starts !
L when VCT level reaches 20% but will stop increasing when auto makeup stops at 40%.
1
'
b. Level will not reach 71% because auto makeup will stop at 40% and with LCV-115 open f
level will begin decreasing.
I
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c. Level will cycle between 20% and 40% because makeup flow is grea:er than flow through
LCV-115.
d. Level will decrease because LVC-115 is open until 20% level when auto makeup begins and
. causes level to increase,
t
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QUESTION No. 004:
Given the following plant conditions:
-
Two minutes ago an MSIV inadvedently closed causing secondary safeties to lift and
a reactor trip and safety injection due to low pressurizer pressure signal to be i
generated.
-
The reactor trip breakers failed to open.
-
The operators tripped the reactor by opening the CRDM MG set supply breakers. )
-
It is now desired to reset SI and secure SI equipment. RCS pressure is now 1800 psia.
Which one of the following will prevent resetting SI from the Main Control Board under these
conditions?
a. RCS pressure is less than SI setpoint.
b. The SI timing relays. l
!
c. RCS pressure is below the P-11 setpomt. l
d. Permissive P-4 has not actuated.
ANSWER: d.
KA: 013A402 4.3/4.4 LEVEL: MEMORY )
.
REFERENCE: OPS-5220ll, pg. 32
LEARNING OBJECTIVE: 05220lI32.a.
HISTORY: New
JUSTIFICATION:
a. Pressurizer pressure below the SI setpoint willinitiate an SI signal but will not prevent
resetting SI.
b. SI will not reset if the 60 second timer is active, but the timer timed out 1 minute ago.
c. P-11 is the set point above which a blocked SI signal will auto unblock. Being below P-11 will
not prevent resett;ng SI.
d. With the reactor trip breakers closed the required P-4 signal will not be generated and SI
cannot be reset from the MCB.
_ _. _ ._. _ . _ . . _ . _ _ _ _ _ _ . _ . . _ . _ . . ___ _ .. . _ - . . . _
,
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>
QUESTION No. 005,
Given the following plant conditions: 9
-
Unit 2 is at 50% power. !
-
Control bank D rods are :
. Group 1 at 161
l
e Group 2 at 160
If the Digital Rod Position Indication System (DRPI) experiences a loss of power to the Data "A"
l cabinet, which one of the following Control Bank D DRPI indications are within the limitations of
l
DRPI:
l
Group 1 Group 2
'
a. 168 162
b. 156 168
c. 150 162
d. 162 150
l%
ANSWER: /.
KA: 014A202 3.1/3.6 LEVEL: ANALYSIS
REFERENCE: OPS-52201F, pg. 8
LEARNING OBJECTIVE: 052201F09
HISTORY: New
JUSTIFICATION:
a. With Data "A" failure accuracy will be +4 and -10,168 on group 1 is outside the accuracy
range.
b. Both group 1 and group 2 are outside the accuracy range.
c. Group 1 is outside the accuracy range of DRPI but is plausible if candidate only remembers
12 steps of tech specs as the accuracy.
d. 162 is within the +4 limit and 150 is within the -10 limit.
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. . .- - - .- __. __ -
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QUESTION No. 006:
Given the following plant conditions:
I - Reactor startup is in progress.
l -
Source range Channel N31 indicates 9E4 eps.
-
Source range Channel N32 indicates 8.5E4 cps.
-
Intermediate range channel N35 indicates 3E-10 amps.
-
Intermediate range channel N36 indicates 2.lE-11 amps.
Which one of the following statements describes the condition of the nuclear instmments?
(Ranges ofIndication for the Excore Instrumentation System, is attached).
a. N35 is overcompensated.
b. N35 is undercompensated.
l c. N36 is overcompensated.
l
d. N36 is undercompensated.
ANSWER: c.
I KA: 015A303 3.9/3.9 LEVEL: COMPREHENSION
REFERENCE: OPS-52201D, pg. 20
LEARNING OBJECTIVE: 052520R03 ,
HISTORY: 052520R04019
l
l JUSTIFICATION:
a. N35 reads higher than N36 but corresponds to SR indication so is not overcompensated.
b. N35 reads higher than N36 and could be caused by undercompensation but N35 corresponds
to SR indication so is not undercompensated.
l
c. N36 reads lower than N35 which corresponds to SR indication therefore N36 is
overcompensated.
l d. N36 reads lower than N35. Undercompensation would result in N36 reading higher than N35.
l
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.. -- . . . _ - . - . . - . . . . - . . . - . . . . . . - . _ - _ - . - . . . - - _ . . . . . . -
{
' 0
10 -10
- 90 f '
.4 LINEAR
10 SCALE R
A .
-20 N
. g
5 g -10 G
10 -0 C
T
E g
- -8 g
'O
. E
D
10 A
r .
E
,
_ -8 R
10 A
N , ,
- S ----
9E
~ $ S -
10
30
R
- 4 C
10
I 11
. _ _ ..
R 10
AMPS
A
_ 3
10 N
G
E
_ 2
to
- 1
10
- 0 -
to CPS
(-
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DETECTOR RANGES
FIGURE 4
1 -.__ . _ _ . _ . . . _ . ~ . _ . ~ . . _ _ _ - - . _ . . . _ _ _ _ _ . . . _ . . _ _ _ _ _ . . _ _ . _ . . _ . _ . . . _ .
. , .
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- QUESTION No. 007:
Given the following plant conditions:
-
The subcooled margin monitors (SMM) are selected to the "CETC" mode.
l Hottest RTD = 605 F
-
Hottest core exit T/C = 612*F
L -
Hottest upper head T/C = 619 F
- PT-402 = 2245 psig . i
_ PT-403 = 2255 psig
-
PT-455 = 2235 psig
-
PT-457 = 2200 psig -
Which one of the following is the subcooling margin displayed on channel A7
a. 34 F l
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b 38 F
c. 41 F
d. 48 F
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- ANSWER: c. l
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KA: 017K502 3.7/4.0 LEVEL: ANALYSIS
REFERENCE: OPS-52202E, pg. 8 ,
l
- LEARNING OBJECTIVE: 052202E13 <
l
HISTORY: OS2202E24008 with stem and all four distracters modified. Similar questien used on
- HTL-24 course final.
1
JUSTIFICATION:
a. 34' F determined by using corrected pressure on PT-455 of 2250 psia and incorrect hottest !
head T/C reading of 619 F.
-
b. ' 38 F dete1 mined by using corrected pressure on PT- 457 of 2215 psia and correct T/C
reading of 612 F. PT-457 is not an input to channel A. j
c. 41*F determined by using corrected pressure on PT-455 of 2250 psia and correct T/C reading i
of612*F;
, .
d. 48'F determined by using corrected pressure on PT-455 of 2250 psia and incorrect RTD
.
reading of 605 F.
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l QUESTION No. 008:
Given the following plant conditions:
l -Unit 1 is at 100% power.
- All containment (CTMT) fan coolers are operating in fast speed.
l - A loss of offsite power occurs.
l
- Emergency diesels have started and are supplying the ESF buses. l
- A Safety Injection (SI) signal is received.
l
l Which one of the following describes the response of the containment fan cooling units? The
nmning CTMT fan coolers will stop and:
a. all must be manually restaned. i
b. all will auto stan in slow speed.
l
c. two of the fan coolers will be staned in fast speed by the sequencer.
d. two of the fan coolers will be started in slow speed by the sequencer.
ANSWER: d. ,
4
KA: 022A301 4.1/4.3 LEVEL: MEMORY I
REFERENCE: OPS-52102C, pg. 2 & 3
LEARNING OBJECTIVE: 052102C10
HISTORY: New
JUSTIFICATION:
a. The fan coolers will stop on the shed signal but 2 units will be started in slow speed by the
sequencer.
b. This would be correct if only the SI had occurred.
c. The fan coolers will stop on the shed signal but only 2 fan coolers will be started in slow speed
not fast speed by the sequencer.
d. The units will stop on the shed signal but 2 fan coolers will be started in slow speed not fast
speed by the sequencer.
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_ . - _ . _ _. . _ . _ - _ _ . _ - . ~ _ _ . _ . ._ __. _ __. _ _ _ . . . _ . _ _
QUESTION No. 009:
Which one of the following will result in the loss of BOTH Containment Spray Pumps? l
!
a. Loss of 4160V Buses F and G. l;
!
b. Loss of 4160V Buses F and J.
c. Loss of 4160V Buses H and G.
d. Loss of 4160V Buses H and J.
ANSWER: a.
KA: 026K201 3.4/3.6 LEVEL: MEMORY
REFERENCE: OPS-52102C, Table 3, pg. T-3
i
LEARNING OBJECTIVE: 052102CO3 l
l
HISTORY: New
JUSTIFICATION:
a. Both buses are ESF buses in different trains, bus F supplies Spray Pump A and bus G supplies
Spray Pump B.
b. Both buses are ESF buses in different trains, but only Bus F supplies a Spray Pump.
l
c. Both buses are ESF buses in different trains, but only Bus G supplies a Spray Pump. '
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d. Both buses are ESF buses in different trains, but neither supplies a Spray Pump.
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QUESTION No. 010:
Given the following plant conditions:
- Unit 1 is at 60% power and increasing during a startup. ;
- All systems are in automatic. 1
- Two Condensate pumps are mening.' !
- Condenser hotwell level is near the low end of the operating band. j
Which one of the following conditions will result if one of the operating Condensate Pumps trip? I
a. One ofthe main feedwater pumps will trip.
b. The Mc cz Driven Auxiliary Feedwater pumps will start.
.
c. The standby Condensate Pump will start.
j
d H~otwell fill valves V901 and V902 will open. 1
1
ANSWER: c.
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l
KA: - 056A204 2.6/2.8 LEVEL: COMPREHENSION ,
l
REFERENCE: OPS-52104C, pg. 20 ;
LEARNING OBJECTIVE: 052104C09.d.
1
HISTORY: New !
JUSTIFICATION: i
e
a. A feed pump could trip on low suction pressure within 30 seconds if the standby Condensate .
Pump did not start. I
l
b. The Motor Driven AFW pumps will start ifone SG level reaches 25%.
c. The standby Condensate Pump will auto start.
d. The hotwell fill valves will open to makeup if hotwell level drops which should not happen
under these conditions.
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_ _- -
.- - _ - - _ _ - . - . - . _ . - - . . - . . - . . .. . . - . - . .-
1-
,
.
'
QUESTION No. 011:
Given the following plant conditions: !
t
-
Reactor poweris 33%. ;
l -
S/G 1 A levelis 79%.
l -
S/G 1B leve1is 80%.
l
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S/G IC level is 79.5%.
Which ene of the following automatic action sequences will result from the above situation? -
t
a. Turbine trip, Reactor trip, Feed Pump trip.
[
i
b. Turbine trip, Feedwater Isolation, Feed Pump trip. ;
c. Turbine trip, Reactor trip, Feedwater Isolation, Feed Pump trip. (
.
>
d. Turbine trip, Reactor Trip, Feedwater Isolation. l
ANSWER: b.
i
KA: 059K104 3.4/3.4 LEVEL: COMPREHENSION !
REFERENCE: OPS-5220ll, pg. 31
,
LEARNING OBJECTIVE: 05220lI20.e.
HISTORY: New
JUSTIFICATION: ,
a. The turbine will trip, the reactor will not trip below P-9 (35%) when the turbine trips, and the l
feed pumps will trip.
l
b. The turbine will trip, Feedwater isolation will occur and the feed pumps will trip.
c. The reactor will not trip below P-9 (35%) when the turbine trips, Feedwater isolation will
occur and the feed pumps will trip.
d. The turbine will trip, the reactor will not trip below P-9 (35%) when the turbine trips,
Feedwater isolation will occur.
_. _ . _ _ . _ _ . _ _ . . _ _ _ _ _ _ _ _ . _ . . _ . _ _ . _ . _ _ _ _ . _ _ _ _ _ _ . . - _ _
r
!
QUESTION No. 012: l
!
FRP-H.1, " Response to Loss of Secondary Heat Sink", directs the operator to stop all RCPs at I
Step 6 of the procedure. Which one of the following is the reason why the RCPs are stopped so j
earlyin the procedure?
l
a. To allow the operator more time to take corrective action before feed and
bleed is required. )
b. To minimize the potential for RCP damage due to loss of RCS pressure.
c. To allow the development of natural circulation. i
d. To minimize the potential for RCP damage due to thermal shock from cold
HHSI flow.
l
ANSWER: a. l
l
KA: 061K301 4.4/4.6 LEVEL: COMPREHENSION
REFERENCE: OPS-52533F, pg.16
LEARNING OBJECTIVE: 052533F07
HISTORY: 052533F07008 with stem and three distracters changed.
JUSTIFICATION:
a. Operation of RCPs will short,en the time before the initiation of feed and bleed is required.
b. RCP could be damaged ifoperated at reduced RCS pressure but in this procedure pressure
remains at or above the PORV setpoint.
c. Stopping RCPs would eventually result in establishing natural circulation if there were a heat
sink. Which is not the case in this procedure.
d. When SI is initiated it shifts the source of water for seal injection from the VCT to the RWST.
l
The actual temperature difference is minimal and should not affect the RCP's.
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.. . . . -- . - - - - - - - . - -. . - . - - - . . - - . - - . - .
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QUESTION No. 013:
!
Unit I has been holding at 33% power for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, which one of the following signals
will result in the Auto start of the Turbine Driven AFW pump? ;
I a. Trip of both main feedwater pumps.
!
l b. . Safety Injection. ;
c. steam generator low level.
i
d. AMSAC signal.
j
ANSWER: . _
[
KA: 061K402 4.5/4.6 LEVEL: ORY
REFERENCE: OPS-52102H, pg. 9 1
LEARNING OBJECTIVE: 052102H13 v
h(
HISTORY: New
JUSTIFICATION: -
!
a. Trip of both MFPs will auto start MDAFW Pumps but not the TDAFW Pump,
b. SI signal will auto start MDAFW Pumps but not the TDAFW Pump.
c. This is the only valid signal for these conditions
d. The AMSAC signalis not active due to being less than 40% power.
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QUESTION No ' 014:
i
Which one of the following describes the normal power source for an Auxiliary Building 125V
DC bus?
L a. 120V AC Distribution Panel through a battery charger, through a Transfer !
! Safety Switch to the DC bus.
b. 208V AC MCC through a battery charger, through a Transfer Switch to the
l bus. '
'
l
c. 600V AC Load Center through a battery charger to the DC bus. j
d. - 600V AC MCC through a battery charger to the DC bus.
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ANSWER.c. '
1
KA: 063K201 2.9/3.1 LEVEL: MEMORY
REFERENCE: OPS-52103C, pg. 3 & 4
LEARNING OBJECTIVE: 052103C03.a.
HISTORY: New
JUSTIFICATION:
a. This is the power supply for the Cooling Tower 125 VDC distribution,
b. This is the power supply for the Senice Water Intake Stmeture 125 VDC distribution.
c. This is the correct power supply for the Auxiliary Building 125 VDC distribution.
d. This is the power supply for the Turbine Building 125 VDC distribution.
.
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l' QUESTION No. 015:
!- i
^
Given the following plant conditions:
l' - Unit 1 is at 100% power. !
l - Excess letdown is in service due to a problem with normal letdown >
l - All other systems are operating normally.
l
l .Which one of the following would be the approximate Reactor Coolant Drain Tank (RCDT) level
,
i
after two hours if the initial RCDT level was at 50%7 (Tank Capacity Curve attached.) ;
a. 50 %
i
b. 55%
c. 63%
!
d. 100% !
ANSWER: b. '
RCP #2 sealleakoff: 3 gph x 3 = 9 gph, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> x 9 gph = 18 gallons
!
50% = 176.48 gallons, 176.48 + 18 = 194.48 gallons = ~55% '
KA: 068K104 2.4/2.5 LEVEL: ANALYSIS
REFERENCE: OPS 52106A, pg. 4,5, & 6
LEARNING OBJECTIVE: 052106A06
i
HISTORY: 052106A06001 with stem, answer and 3 distracters modified.
JUdTIFICATION:
a. The RCDT level would remain at 50% if the level control valve was in auto and one RCDT
pump was running. The training material states that the pumps are not kept running but are
. started to pump down the tank as necessary.
b. This number is correct since the only liquid entering the RCDT would be the RCP #2 seal leak
off.
c. This number would obtained if3he applicant incorrectly used a value of 15 GPH for excess
letdown flow and combined this number with the seal leak off flow. However, the LCV fails
to the VCT position.
d. The RCDT would fill if excess letdown were aligned to the RCDT , but the conditions in the
stem would cause excess letdown to be aligned to the VCT.
p
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_ . . _ _ - - . . . _ - _ . _ _ . . - - . - . - . - - - . _ . _ _ . - - . - . _ - - - - _ - - . . _ _ _ _
- - -
'
volu=e II Cur-te 2S
Re ac:c t C olant Drun Tank Capacity
N1G21T001
Capaci:y (Gallons) vs :1 1.evel
'
Rev. 2. Dececse r 14, 1981, C.A.F.
App rove d:
i
4
[D I ~ ~L2- 2%
l Tecnnicai ;:foerintencient. Date
l ' LEVEL
.
CALLONS '
. LEVEL CALLOHs
1
1
0.0 19.28 51.0 180.09
! 1.0 20.45 52.0
'
2.0 22.69 183.71
53.0 187.32
i
'
( 3.0 25.00 54.0 190.94
4.0 27.37 55.0 194.54
, 5.0 29.81 56.0 198.15
1 6.0 57.0
. u.
3.2.31
e 4 . 6, <
. 201.74 ,
r
58.0 205.33
- 8.0 37.49 59.0 208.92
l
S.0 40.16 60.0 212.49
! 10.0 42.58 61.0 216.05
! 11.0 45*.65 62.0 219.60
j 12.0 48'.46 63.0 222.14
l ee . u- e 1 . e...
v
.
e 4. u. .n. .
cac.e<
14.0 54.23 .65.0 230.19
> 15.0 57.18 66.0 232.68
j
16.0 60.17 67.0 237.17
l , 17.0 63.20 68.0 240.62
j 18.0 66.26 69.0 244.08
!
19.0 69.36 70.0 247.50
40.0 72.50 71.0 250.91
- 21.0 75.67 72.0
,
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24.0 35.36 75.0 264.21
, 25.0 58.64 "' 6 . 0 267.60
26.0 91.96 77.0 270.86
27.0 95.29 78.0 274.09
,
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-,.. c.
) 29.0 102.04 80.0 280.45
30.0, 105.45 81.0 283.59
e1... 10ec.co ^^
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,ee- . n. 1 1.e. . .e..e ce
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37.0 129.31 88.0 304.49
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?9.0 126.90 90.0 310.08
40.0 140.46 91.0 312.80
41.0 144.04 F2.0 115.46
a e . n.
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43.0 151.21 94.0 320.64
44.0 154.81 95.0 323.14
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I CAPACITY L) VS LWEL (F
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2- _ APPROVED:
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- -- - - - -- -- -
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{. 0. W'YSV _ _ ff -f ; I'
i- 5 ) - TECHNICAL MANAGER DATE - s I_
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0.75 8.6 69.0
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$ l
~ QUESTION No. 016: l
Which one of the following sets of Waste Gas Tank parameters is within Technical Specification -i
= Limiting Conditions for Operation?
Curie content Hydronen Oxvnen
a. 71,500 curies Iodine >4% 2%
b. 71,500 curies Noble Gases >4% 4% -l
l
c. 71,500 curies Noble gases >4% 2% }
d. 71,500 wries Iodine >4% 4%
ANSWER: a.
KA: 071K504 2.5/3.1 LEVEL: COMPREHENSION
REFERENCE: FNP-1-SOP-51.0, pg. 4, Precaution 3.12 and Technical Specification 3.11.2.5
and 3.11.2.6.
LEARNING OBJECTIVE: 052302G06 I
I
HISTORY: New
JUSTIFICATION:
a. Curie content is above limits but it is not Noble gas , Hydrogen is above the limit but oxygen
is within limits.
b. Curie content, hydrogen and oxygen are above the limits !
1
c. Noble gases are above the limit , hydrogen concentration is above the 4% referenced by the !
TS but oxygen is within limit. l
d. Oxygen concentration is above the 2% referenced by the TS but hydrogen is below 4% by
volume.
.
.-r - *
_ _ - . -
.
QUESTION No. 017:
Using the attached figure, determine which one of the following statements is correct regarding
the proper alignment / operation of the High Range Containment Area Monitor R-27A/B7
a. The Electronic Check Source (E.C.S.) push-button, when depressed, will
insert a radioactive check source to verify proper detector operation.
b. The Channel Test push-button, when depressed, will cause the meter to read in
the midrange of the value selected on the Off/ Test function switch.
c. The scale of the meter is changed to 3 decades 10'- 10' when the function l
switch is placed in the 1 - 10 position.
d. The ALERT and ALARM push-buttons only light indicating an alert or alarm
condition.
ANSWER: c.
KA: 073A402 3.7/3.7 LEVEL: MEMORY
REFERENCE: OPS-52106D, pg. 22
LEARNING OBJECTIVE: 052106D05
HISTORY: 052106D05015 with distracter b. modified.
l
JUSTIFICATION:
a. The ECS push-button applies a DC test voltage to the detector and electronics. A source is
not involved.
b. The channel test push-button will only illuminate the ALERT, HIGH and CHANNEL TEST
lights.
c. The scale of the meter is changed to read only the 3 decades selected by the function switch.
d. The ALERT and HIGH push-buttons, in conjunction with an internal adjustment, are used to
adjust the alert and high alarm set points.
l
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- - . . . . _ ~ . . . - - . .. - .. -.= .. .___ _ _.= -_.__.. _ - - . - .....
DETECTOR ASSEMBLY ,
l
-- = m n
- . . -.
7
@ o - ; i
@ O f '
C
< ,
3
2 10 104
10 ig 5
s
20
\ / 10
1 I 103 106 g 210
10
10 10
$y 10 10
4
g i itill
4 88 3
j,5 2 , 1.5 2 ALL 1-10
4
8
3 8 0FF'
TLST
ALERT . NIGN
U
SAFE
CONTAINMENT MONITOL
VICTDREEN
ggggy
CgEL
V J
HIGH RANGE CONTAINMENT AREA MONITOR
'R-27A&B'
FIGURE 4
'
QUESTION No. 018:
Given the following plant conditions:
- Unit 1 is at 75% power.
- The rod control system is in automatic.
An increase in which one of the following parameters would result in a smaller ga_in output of the
power mismatch channel in the Reactor Control Unit?
a. Reference temperature (Tar.r).
b. Output of the Temperature Mismatch Channel (Tragon).
c. Power range channel N-44.
d. First stage impulse pressure (Pam).
ANSWER: d. I
KA: 001G2.1.28 3.2/3.3 LEVEL: COMPREHENSION
REFERENCE: OPS-52201E, pg.12
LEARNING OBJECTIVE: 052201E03 ,
HISTORY: New
JUSTIFICATION:
a. Tatr is an input used in the Temperature Mismatch Channel of the Reactor Control Unit not
an input to the Power Mismatch Channel.
b. Tranon is the output of the Temperature Mismatch Channel of the Reactor Control Unit not
an input to the Power Mismatch Channel. j
c. is an input to the Power Mismatch Channel but is compared with Pawto provide an error
signal that is input to the non-linear gain amp:ifier.
d. PIMP is used as the power compensation input to the variable gain unit.
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._.
QUESTION No. 019:
Which one of the following is the reason for maintaining a minimum pressure of 18 psig in the
VCT during normal power operation?
a. To ensure proper coolant flow across RCP seal #2.
b. To ensure adequate hydrogen concentration in the RCS coolant.
c. To prevent adequate suction pressure during a multiple charging pump starts.
d. To provide adequate charging pump recirculation backpressure during normal
operations.
ANSWER: a.
KA: 004K104 3.4/3.8 LEVEL: MEMORY
REFERENCE: OPS-52101F, pg.19 and Appendix 1 of OPS-52101D
LEARNING OBJECTIVE: 052101D11
HISTORY: New
JUSTIFICATION:
a. The 18 psig provides adequate backpressure on the #1 seal leakoff to ensure proper flow
across the #2 seal.
b. The gas used in the VCT, hydrogen, ensures adequate hydrogen concentration in the RCS not
the 18 psig overpressure.
c. The minimum volume in the VCT provides adequate suction pressure for the charging pumps
the pressure requirement is for seal flow.
d. The charging pump miniflow recirculation lines which return to the VCT contain orifices to
provide back pressure.
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1
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! l
, QUESTION No. 020: 1
l
Given the following plant conditions:
-
Unit 1 is in operation at 100% power.
-
Loop 1 and 3 Tavg meters indicate 575 degrees F.
- Loop 2 Tavg meter indicates off scale HIGH.
-
Loop 1 and 3 Delta T meters indicate 100 percent.
-
Loop 2 Delta T meter indicates 0%.
Which one of the following is the cause of these indications? ;
a. Loop 2 Thot failed low.
b. Loop 2 Thot failed high. j
c. Loop 2 Tcold failed low.
i
d. Loop 2 Tcold failed high. l
ANSWER: a.
KA: 002K512 3.7/3.9 LEVEL: ANALYSIS i
,
REFERENCE: OPS 52201J, pg.1 l
!
LEARNING OBJECTIVE: 052201J01
HISTORY: New
JUSTIFICATION:
a. That failing low would cause the delta-T meter to indicate 0% but the Tavg meter would also 1
indicate low.
b. Thot failing high would result in Tavg indicating high but delta-T would also indicate high. l
c. Teold failing low would result in Tavg indicating low and delta-T indicating high.
d. Tcold failing high would result in the indications provided in the stem.
i
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. - - .. . . - . - - . - . - ~ _ . - - . . . - - - . . - . . . - - . -
'
s
QUESTION No. 021:
l
l Given the following plant conditions:
1
-
Unit 1 is in Mode 4 with the following RWST conditions:
-
RWST volume at 472,000 gal.
. RWST boron concentration is at 2305 PPM.
-
RWST temperature is at 34 degrees F.
Which one of the follow'mg is the operability status of the RWST?
l
l a. Inoperable because of the potential for boron crystallization in the RWST.
b. Inoperable because there is insufficient borated water inventory to meet DBA
conditions.
c. Inoperable because there is insufficient volume to ensure adequate NPSH
l
during a DBA.
d. Inoperable because in the event of an accident long term shutdown margin
cannot be ensured.
ANSWER: a.
KA: 006A115 3.3/3.9 LEVEL: MEMORY
REFERENCE: OPS-52102B, pg. 7 and Technical Specification 3.1.2.(. l
1
LEARNING OBJECTIVE: 052102B15
HISTORY: New
JUSTIFICATION:
J
a. The temperature given is below the minimum given in TS for operability. l
l
'
b. The RWST volume and concentration given in the stem meet the TS operability volume and
concentration.
c. The RWST volume is above the minimum TS requirement,
d. The boron concentration given is above the minimum required by TS.
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.
QUESTION No. 022:
Gi 4 the following plant conditions: '
-
Unit 2 is at 100% power. !
-
All systems are in automatic.
-
A transient with the CVCS has resulted in:
l * Pressurizer level increasing to 61%, and
. - Pressurizer pressure increasing to 2270 psig.
.
Which one of the following describes the Pressurizer heaters and spray status for these
conditions?
a. Backup heaters on, Variable heaters at maximum voltage, Spray valves closed.
b. Backup heaters off, Variable heaters at minimum voltage, Spray valves closed.
c. Backup heaters off, Variable heaters at maximum voltage, Spray valves
throttled open. .
d. Backup heaters on, Variable heaters at minimum voltage, Spray valves
throttled open.
ANSWER: d.
!
KA: 010K108 3.2/3.5 LEVEL: ANALYSIS / COMPREHENSION
' REFERENCE: OPS 5220lH, pg. 5,6 and 17
. LEARNING OBJECTIVE: 05220lH07.f.
HISTORY: New .
JUSTIFICATION: ,
a. The backup heaters would be on due to + 5% level increase but the variable heaters would be
at minimum voltage and the spray valves would be partially open.
b. The variaWe heaters would be at minimum voltage but the backup heaters would be on and
the spray valves would be partially open.
c. The spray valves would be partially open but the backup heaters would be on and the variable
heaters woald be at minimum voltage.
d.: The backup heaers would be on due to the + 5% level increase, the variable heaters would
nave gone to minin um voltage at 2250 psig and the spray valves would have staned opening
at 2260 psig.
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, - -
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QUESTION No. 023:
A pla .; startup is in progress and operators are collecting critical data with all systems except rod I
control in automatic when the comrolling pressurizer level channel fails high. Which of the will be I
the initial result (assun ing no opentor action)? l
a. Pres:;urizer power operated relief valve cycling.
b. Pres::urizer level controlling at 54%.
i c. Reactor trip on high pressurizer pressure.
d. Reactor trip on high pressurizer level.
ANSWER: a.
KA: 011 A210 3.4/3.6 LEVEL: ANALYSIS
REFERENCE: OPS 5220lH, Figure 7 and Appendix 2
LEARNING OBJECTIVE: 05220lH14.b. !
HISTORY: New
JUSTIFICATION:
1
a. Controlling level failing high will result in charging flow reducmg to muumum, pressunzer 1
level decreasing until letdown isolates then increasing PZR level until the pressurizer goes
solid, raising pressure until the PORV opens.
b. Plausible since level program is high limited.
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c. The reactor trip on high pressure is 2385 psig, which is above the set pressure of the PORV.
d. This would be the result if power were above 10%
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QUESTION No. 024:
Given the following plant conditions: 1
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A Unit I shutdown is in progress.
-
The intermediate range high flux trip bistables havejust been verified to be reset.
Which one of the following will cause an automatic reactor trip:
a. A source range channel energizes.
b. IB stanup transfonner becomes deenergized.
c. . Turbine trip.
' d. l A RCP breaker trips.
ANSWER: b. l
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KA: 012K406 3.2/3.5 LEVEL: COMPREHENSION
REFERENCE: OPS 52201I, pg. 33
LEARNING OBJECTIVE: 052201114.c.
HISTORY: New
JUSTIFICATION:
a. A source channel energizing would initiate the required coincidence for a reactor trip, but due
to plant conditions (intermediate range trip must verified reset prior to going below 13%) the
source range trip would be blocked.
b. The RCP buses are transferred to the start up transformers prior to the step to verify the
intermediate range trip. The IB start up transformer supplies two the RCP buses to meet the
coincidence for the RCP undervoltage/underfrequency trip which is not blocked until power is
< 10%.
c. The turbine trip reactor trip is blocked below P-9 (35%). ;
d. The RCP breaker position trip changes to a 2/3 coincidence when power is <35%. The plant l
conditions above ensure that power is <35%, therefore a single RCP breaker opening will not i
result in a reactor trip.
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QUESTION No. 025: ;
,
Given the following plant conditions:
- - Unit 1 is at 60% power.
- All control systems are in automatic.
Which gns of the following events will cause Loop B ovenemperature (OT) delta-T trip setpoint
- to increase? Assume each event occurs separately.
a. Channel 456 pressurizer pressure detector fails high.
b. Loop B Tavg unit output fails high.
c. N42 power range lower detector fails high.
j d. Power is increased to 103%.
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l ANSWER: a.
KA: 012K611 2.9/2.9 LEVEL: ANALYSIS
REFERENCE: OPS 52201J, pg. 4,13,14 and Figure 11.
LEARNING OBJECTIVE: 052201I12
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' HISTORY: New-
JUSTIFICATION:
a. An increase in pressure will result in an increase in the OTDT setpoint.
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j b. An increase in Tavg will result in OTDT decreasing.
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l c. The delta I input to OTDT cannot increase the setpoint.
d. As power is increased the OTDT setpoint would decrease.
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QUESTION No. 026:
Given the following Unit 1 plant conditions:
-
Steam generator level decreases to the level shown. !
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Turbine load stays at the load indicated. .
Which one of the following plant conditions will result in an AMSAC actuation?
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Steady State
2/3 Steam Generator Turbine Load ;
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a. 7% for 18 sec 39% 40%-
b. 9% for 26 sec 40% 41 %
c. 25 % for 60 sec 42% 44 %
d. 79% for 260 sec 44 % 47%
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ANSWER: b. ,
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KA: 016Kil2 3.5/3.5 LEVEL: MEMORY r
. REFERENCE: OPS 5220lK, pg. 7
LEARNING OBJECTIVE: 05220lK05
,
HISTORY: New
JUSTIFICATION:
f
,
a. AMSAC will not be actuated because it is less than the minimum time for AMSAC actuation
of 25 sec.
b. AMSAC will actuate because 2/3 SG levels are below 10% for > 25 seconds with turbine
power above 40%. -
c. AMSAC will not be actuated because SG levels are above 10%. 25% is the SG Lo Lo Level
Reactor Trip setpoint.
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~d. AMSAC will not actuate because SG levels are above 10%. 79% is the SG Hi Level Turbine
Trip and Feedwater Isolation and 260 seconds is the time delay to enable AMSAC based on i
turbine load.
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i ' QUESTION No. 027:
E :
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! Given the following pla11 conditions: 1
- A large break LOCA has occurred on Unit 2 thiny minutes ago.
- Hydrogen corcentration inside containment is 4.5%. )
Which one of the following actions should be taken to reduce hydrogen concentration?
a. Place only one electric hydrogen recombiner in service within the next 30
minutes at a power setting of 100 kilowatts.
b.' Place both electric hydrogen recombiners in service within the next 30 minutes
at a power setting of 50 kilowatts.
c. Place the post accident containment venting system in service within the next
30 minutes.
d. Place the post LOCA containment air mixing system in service within the next
30 minutes.
ANSWER: c.
KA: 028K502 3.4/3.9 LEVEL: MEMORY
REFERENCE: OPS 52102D, pg. 5
LEARNING OBJECTIVE: 052102D05.
HISTORY: 052102D05008 with stem and three'distracters modified. Similar question used on
the HLT-24 course written final.
JUSTIFICATION:
a. Recombiner operation should commence within one hour of the event, however, the maximum
power setting is initially set at approximately 68 kilowatts and recombiners should not be put
into service if hydrogen concentration is above 4%.
b. Recombiner operation should commence within one hour of the event, however, recombiners
should not be put into service if hydrogen concentration is above 4%.
1
c. The post accident containment venting system should be used when hydrogen concentiation is i
above 4%.
d. . The post LOCA containment air mixing system is used to prevent the formation of hydrogen
pockets and not to reduce the concentration of hydrogen.
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QUESTION No. 028: l
Given the following spent fuel pool conditions.
-
Pool water level is 20 feet above the fuel. 1
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Pool water Boron concentration is 2300 PPM. l
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Pool reactivity is Keff= .96. s
Which one of the following describes whether the above conditions are in compliance with limits !
imposed on the spent fuel pool? i
Ecol Level Pool Boron Conc. Pool Reactivity
a. compliance compliance compliance ;
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b. compliance non-compliance compliance j
c. non-compliance compliance non-compliance
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d. non-compliance non-compliance non-compliance
ANSWER: c. j
KA: 033G2222 3.4/4.1 LEVEL: MEMORY ;
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REFERENCE: OPS 52108D, pg. 36 - 38 and Tech Specs 3.9.1 and 3.9.11 i
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LEARNING OBJECTIVE: 052302M03, and 052302M14 I
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'HISTORYi New-
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JUSTIFICATION-
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a. One of the choices is correct.
b. None of the choices is correct. i
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c. All three of the choices are correct. i
d. Two of the three choices are correct.
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QUESTION No. 029:
Which one of the following statements describes the fuel transfer system FRAME INTERLOCK
BYPASS?
a. The switch is actuated from the Pit Control Panel to allow operation of the !
conveyer car when the upenders are not in the horizontal position.
b. The switch is actuated from the Reactor Control Panel to allow operation of
the conveyer car when the upenders are not in the horizontal position.
c. The switch is actuated from the Pit Control Panel to allow operation of the
manipulator crane if the gripper tube is not up or up disengaged.
d. The switch is actuated from the Reactor Control Panel to allow operation of
the manipulator crane if the gripper tube is not up or up disengaged.
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ANSWER: a.
KA: 034G2130 3.9/3.4 LEVEL: MEMORY
REFERENCE: OPS 52108D, pg.16 & 17
4
LEARNING OBJECTIVE: 052108D10
HISTORY: New
JUSTIFICATION:
a. The named switch is located on the Pit Control Panel and performs the function descdbed.
,
b. The named switch is not located on the Reactor Control Panel but does perform the function
described.
c. The named switch is located on the Pit Control Panel but does not perform the function
described.
d. The named switch is not located on the Reactor Control Panel nor does it perform the
function described.
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QUESTION No. 030:
Given the following plant conditions:
!
- Unit 1 is operating at 75% power. ,
- The 1 A SG selected steam pressure channel fails low.
{
l Which gns of the following describes the SGFP speed change and the reason for the speed i
change?
l )
a. SGFP speed increases due to actual feed flow signal increasing. )
b. SGFP speed decreases due to actual delta-P decreasing.
c. SGFP speed increases due to program feed flow increasing. '
d. SGFP speed decreases due to program delta-P decreasing.
ANSWER: d.
KA: 035K101 '4.2/4.5 LEVEL: ANALYSIS
REFERENCE: OPS 52201B, pg. 7 & 8
LEARNING OBJECTIVE: 052201B14 and 15
HISTORY: 052201B16024 with stem and 2 distracters modified.
JUSTIFICATION:
a. The failure of the named steam pressure channel which is in the control circuit will not result
in a changein feed flow,
b. The actual Delta-P is determined using steam line pressure not SG pressure.
c. Feed pressure not flow is the input to the feed pump control circuit.
d. Failure of the selected .SG pressure channel will result in decreasing the program delta-P
which results in the SGFPs slowing down.
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L ' QUESTION No. 031: I
Given the following plant conditions. ;
-
The Unit 2 reactor has been tripped due to a loss ofinstrument air.
-
No station air compressors are mnning.
-
SG atmospheric relief valves are being controlled locally in the lower equipment room. i
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1 A emergency air compressor is mnning.
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Tavg is stable at 555 F and being controlled by the lowest set pressure SG safety.
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Which one of the following conditions will result if air pressure to the atmospheric reliefvalves is -
adjusted to 47 psig?
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a. Tavg will lower rapidly, the atmospheric reliefs are fully open.
!
b. Tavg will lower slowly, the atmospheric reliefs are partially open.
c. Tavg will remain constant, the atmospheries reliefs are closed.
d. Tavg will raise slowly, the atmospheric reliefs are closed but decay heat
prodt.ction results in RCS heating.
ANSWER: a.
KA: 039K305 3.6/3.7 039A105 3.2/3.3 LEVEL: COMPREHENSION
REFERENCE: OPS 52104A, pg. 8
LEARNING OBJECTIVE: 052104A15
HISTORY: 052104 A15026 with stem and one distracter modified. Used on class quiz.
JUSTIFICATION:
a. At 45 psig the valve is fully open.
b. At 26.5 psig the valves start to open, therefore the valve would be full open at a pressure of
45 psig not partially open.
c. Tavg would not be constant but would be decreasing rapidly due to the full open valve.
d. Tavg would likewise not be rising but would be decreasing rapidly due to the full open valve. l
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QUESTION No. 032:
Which one of the following describes the NORMAL, EhERGENCY and BACKUP power j
supplies to Emergency 4160V AC Bus IF7
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NORMAL EMERGENCY BACKUP l
a. S/U IB l-2A DG S/U 1A
b. S/U 1A 1-2A DG S/U IB
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c. S/U 1A IBDG S/U IB
d. S/U IB IB DG S/U 1A
ANSWER: b.
KA: 062K201 3.3/3.4 LEVEL: hEMORY
REFERENCE: OPS 52103B, pg. 2,4 & 5
LEARNING OBJECTIVE: 052520E01
HISTORY: Similar question used on the HLT-24 remedial exam but stem and three distracters
changed. i
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JUSTIFICATION:
a. The Normal and Backup supplies are reversed. l
- b. This is the correct listing for Bus F.
c. The Normal and backup are correct, but the emergency is incorrect.
,
d. This would be correct ifit was thought that IF was B train.
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QUESTION No. 033:
- Given the following plant conditions:
-
The "A" air receiver for DG IC is out of service.
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Neither of the air compressors is capable of running.
!
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- Which one of the following describes the design maximum possible number of DG stan attempts
. that are possible given the above conditions?
a. 1
l b. . 3
c. 5
d. ' 7
ANSWER: c.
KA: 064K607 2.7/2.9 LEVEL: MEMORY
REFERENCE: OPS 52102I, pg. 7 ,
LEARNING OBJECTIVE: 052102I04
HISTORY: New
JUSTIFICATION:
1
a. Staning air to only one of the air headers will start the diesel.
b. Three starts are possible but a total of five starts is the design maximum.
l' c. Five stans are possible.
d. Seven starts is not possible.
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QUESTION No. 034: ,
Which one of the following switch position combinations are required for the 1-2A Diesel ;
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Generator to start when the remote start push button on the EPB is depressed?
EPB Mode Selector Switch DLCP Mode 4 Selector Switch
a. MODE 2 OFF
b. MODE 2 M-4 ,
c. MODE 3 OFF ,
d. MODE 3 M-4
ANSWER: a.
KA: 064A401 4.0/4.3 LEVEL: MEMORY
REFERENCE: OPS 52102I, pg. 26 & 27
LEARNING OBJECTIVE: 052102I19
HISTORY: New
JUSTIFICATION:
a. In Mode 2 the remote push button will start the diesel. j
b. In M-4 the Control Room has no control.
c. In Mode 3 the io al start /stop push buttons are operational the remote are not. I
d. In M 4 het Control Room has no control.
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QUESTION No. 035:
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Given the following plant conditions:
- Unit 1 is Mode 5. !
- A Containment purge is in progress. !
L - STP-50, Radiation Monitor Monthly Source Check, is being performed on Spent Fuel
Ventilation Monitor R-25A.
- The red alarm light is illuminated.
Which one of the following describes the system response to these conditions?
a. The automatic actions of R-25A will be blocked while performing this STP.
b. The 1 A fuel handling area supply and exhaust fans trip and the 1 A fuel handling
area supply snd exhaust dampers close. The IB fuel handling area supply and
exhaust fans start.
c. The fuel handling area supply and exhaust fans trip, the fuel handling area
supply and exhaust dampers close and the penetration room 1 A and IB ,
filtration units start,
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d. The fuel handling area supply and exhaust fans trip, the fuel handling area
supply and exhaust dampers close and the containment purge supply and
exhaust valves close.
ANSWER: c.
KA: 073K101 3.6/3.9 LEVEL: MEMORY
REFERENCE: OPS 52106D, pg.15 and Table 1, pg. T-lb
LEARNING OBJECTIVE: 052106D08
HISTORY: 052106D08006 with stem and one distracter modified.
JUSTIFICATION:
'
a. If the alarm circuit actuates during the STP the automatic functions provided by the
instmment will occur.
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b. The A fuel handling area fans will trip, the fuel handling area dampers will close, but
penetration room filtration units will start, not the other train of SFP HVAC.
c. These are the correct automatic actions that will occur when R-25A alarms.
d. The fuel handling area fans will trip, the fuel handling area dampers will close, but the
containment purge valves will not close.
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QUESTION No. 036:
Which one of the following conditions is correct for holding at 65% power:
a. All fans on B and C cooling towers not running.
b. A cooling tower isolated for maintenance.
c. The cooling tower bypass system in service.
d. A single waterbox isolated to repair a tube leak..
ANSWER: d.
KA:. 075K401 2.5/2.8 LEVEL: ANALYSIS
REFERENCE: FNP-1-SOP-26.0, pg. 3; FNP-1-UOP-3.1, pg.11; OPS 5220lG, pg.17
LEARNING OBJECTIVE: OS2104D01
HISTORY: New
JUSTIFICATION:
a. A least one group of fans must be started per UOP-3.1 when ramping up and power level is
between 30-60%.
b. SOP-26.0 prohibits running two circulating water pumps through two Cooling Towers and
both pumps are required to be in service at 65%. !
c. SOP-26.0 limits the operation to only one circulating water p imp during operation of the
cooling tower bypass system and both pumps are required to be in service at 65%.
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d. AOP-25.0 allows isolating a waterbox for leak repair at 600 megawatts (860 x 65% = 559
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MW). !
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QUESTION No. 037: i
l Given the following Unit 2 plant conditions: i
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RHR is in service. !
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RCS pressure is 320 psig and INCREASING.
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RCS temperature is 340 degrees F and INCREASING.
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ALL system lineups are in a normal shutdown configuration.
Which one ofthe following will act f!_rit to prevent overpressurizing the Reactor Coolant System?
t
a. Pressurizer PORV will open.
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b. RHR loop suction valves will auto close. '
c. RHR pump suction reliefvalves will open. !
d. RHR pump discharge reliefvalves will open.
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ANSWER: c.
KA: 005K109 3.6/3.9 LEVEL: ANALYSIS
REFERENCE: OPS 52101K, pg. 6 & 7
LEARNING OBJECTIVE: 052101K12
HISTORY: New ;
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JUSTIFICATION:
a. The PORVs normally provide RCS overpressure protection except when in a cold, solid
condition.
b. The RHR inlet iso'.ition valves will auto close if RCS pressure exceeds 700 psir, but the
suction reliefvalves will lift first.
c. The RHR suction relief valves will lift at 450 psig to protect the RCS from overpressure.
d. The RHR discharge relief valves will lift at 600 psig to protect RHR system components.
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QUESTION No. 038: !
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Which one of the following statements identifies systems and/or comisonents that alldischarge to l
the pressurizer relief tank (PRT)? !
a. Pressurizer PORVs, Reactor Vessel Head Vents, and VCT relief valve.
b. Pressurizer safety valves, Accumulator relief valves, and RHR relief valves. !
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c. Seal return heat exchanger reliefvalve, pressurizer PORVs, and the nitrogen
system.-
d. CVCS letdown piping relief valve, charging pump suction relief valve, and ;
RHR relief valves. ;
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ANSWER: d.
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KA: 007A301 2.7/2.9 LEVEL: hEMORY j
REFERENCE: OPS 52101E, pg. 9
LEARNING OBJECTIVE: 052101E24
HISTORY: 0052101E24007 with all four distracters modified.
JUSTIFICATION: 6
a. PORVs and Reactor Vessel Head Vents discharge to the PRT, the VCT relief does not.
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b. Safety valves and RHR discharge reliefvalves discharge to PRT, the Accumulator relief.
valves do not. 4
c. The PORVs and the Nitrogen system discharge to the PRT, the seal return heat exchanger
relief valve does not.
d' . The letdown piping relief valve, charging pump suction relief valves and the RHR relief valves
all discharge to the PRT.
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, QUESTION No. 039:
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Given the following Unit 2 plant conditions: !
-- Operating at 100% power, for the last month.
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All systems are lined up for normal operation.
-
While in AUTO rod control, Control Bank "D" starts stepping in slowly, but at a
noticeable rate.
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Which one of the following events will cause this response?
a. A tube leak in a RCP Thermal Barrier Heat Exchanger.
b. A tube leak in the Seal Water Heat Exchanger.
c. A tub: leak in the Letdown Heat Exchanger,
d. A tub: leak in the on service CCW heat exchanger.
ANSWER: b.
KA: 008K301 3.4/3.5 LEVEL: ANALYSIS
REFERENCE: OPSS2102G, pg. 5
LEARNING OBJECTIVE: 052520103
HISTORY: New question although a CCW system leak question was used on the SNC Final
remedial examination, end on one of the audit exams.
JUSTIFICATION:
I
a. CCW cools the RCP Thermal Barrier heat exchanger but a leak would not result in control !
rods stepping in.
b. CCW cools the Seal Water heat exchanger and a leak would result in slowly diluting the RCS
causing the control rods to move in.
c. CCW cools the letdown heat exchange but a leak would not result in control rods stepping in.
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l d. CCW is cooled by service water but a leak in the heat exchanger will not cause inward rod ;
motion. l
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QUESTION No. 040:
Given the following Unit 2 plant conditions:
-
The plant hasjust had a 40% load rejection from 100%.
-
All systems are operable and in automatic.
-
Actual Tavg is 12 degrees F higher than Tref.
Which one of the following describes the response of the steam dump valves?
V501 A & V501C & V501B & V501D & '
V501E V501G V501F V50lH
a. Full Open Modulating Closed Closed
b. Full Open Full Open Closed Closed
c. Full Open Full Open Modulating Closed
d. Full Open Full Open Full Open Modulating
ANSWER: c.
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KA: 041 A408 3.0/3.1 LEVEL: ANALYSIS l
REFERENCE: OPS 5220lG, pg. 8,9, and 10.
LEARNING OBJECTIVE: 052201G16
HISTORY: Similar type of question used on SNC Final remedial exam to test Tavg-Turbine Trip
response. This question, with a different stem and distracter format, is testing the Tavg-Loss-of-
Load controller setpoint.
JUSTIFICATION:
a. Under the given plant conditions the first two banks of steam dump valves would be fully
open, the third bank would be partially open and the four+.h bank would be closed.
b. Under the given plant conditions the first two banks of steam dump valves would be fully
open, the third bank would be partially open and the fourth bank would be closed,
c. Under the given plant conditions the first two banks of steam dump valves would be fully
open, the third bank would be partially open and the fourth bank would be closed.
d. Under the given plant conditions the first two banks of steam dump valves would be fully
open, the third bank would be partially open and the fourth bank would be closed.
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, QUESTION No. 041: !
l !
Given the following plant conditions. ;
_
- Unit 1 is holding at 33% power for Reactor Engineering. j
- Trefis 557 F.
- Loop 1 Tavg is 560 F.
- Loop 2 Tavg is 559 F. ~
- Loop 3 Tavg is 560 *F. !
- The RO inserted control rods in MANUAL one step.
-. Rod motion did not stop, when the IN-HOLD-OUT switch was released.
'
j - Rods were placed in AUTO but continue to slowly step in. j
Which one of the following is the required operator response? .
1
a. Manually trip the reactor because rod motion should have stopped. l
b. Place rod control in MANUAL and restore rods to program.
I
c. Adjust turbine load to restore Tavg to program.
d. Allow rod control to restore Tavg to program.
ANSWER: d.
KA: 000001 AA101 3.5/3.2 LEVEL: ANALYSIS
REFERENCE: FNP-1-AOP-19.0, pg. 2
LEARNING OBJECTIVE: 052520S03
HISTORY: 052520S01020 with stem modified.
JUSTIFICATION: I
)
a. The procedure requires a reactor trip if unexplained rod motion does not stop. Rod motion in
AUTO is due to the difTerence between median Tavg and Tref being 3 F. I
b. Rod control was in MANUAL initially. The procedure requires rod control to be placed in
AUTO. If rod motion is still unexplained then a reactor trip is required. I
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c. The rod control system is restoring Tavg to Tref therefore adjusting turbine load, a step in the
procedure is not required.
d. The procedure allows AUTO rod motion until Tavg is within 1 degree of Trc..
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QUESTION No. 042:
i Given the following plant conditions: i
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- Unit 1 is at 90% power, ramping to 100%.
l
- Rod controlis in MANUAL.
- COMP ALARM ROD SEQ /DEV OR PR FLUX TILT alarm is illuminated.
! - Power has decreased.
- Tavg has decreased. ;
- Rod F6 is indicating 1I steps below its group step counter. I
Which one of the following operator actions is required?
a. Open the disconnect switch for rod F6.
b. Restore RCS Tavg to program.
c. Place rod controlin AUTO.
d. Trip the reactor.
ANSWER: b. - l
KA: 000003G2411 3.4/3.6 LEVEL: MEMORY )
i
I
REFERENCE: FNP-1-AOP-19.0, pg. 2 & 3
LEARNING OBJECTIVE: 052520S01
i
HISTORY: 052520S1004 with stem and three distracters changed.
JUSTIFICATION:
a. Opening the disconnect switches for all but the affected rod is a requirement to recover the
misaligned rod.
i
b. This is the procedural requirement if a rod has " fallen" less than 12 steps from its demanded )
- position.
c. This is the procedural requirement if unexplained rod motion is occurring and rod controlis in !
MANUAL.
d. This is the procedural requirement if a rod drops more than 12 steps.
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. - . . . _ . - - - - _ - - ._ _-._- - .- .- . . - - _ . - . - . -
t
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L - QUESTION No. 043:
{
!
With one control rod misaligned from its group by more than 12 steps and determined to be l
l
'
inoperable, Technical Specification ACTION requires limiting power to 75% of RATED
THERMAL POWER. l
!
Which one of the following provides the basis for this power limit. ;
a. Provides assurance of fuel rod integrity during continued operation. !
b. Ensures the affected rod remains above the Rod Insertion Limit. :
c. Prevents power range NI channels from exceeding thermal power limits due to i
rod shadow'mg effects. '
d. Ensures the limits of QUADRANT POWER TILT RATIO are not exceeded.
ANSWER: a. I
1
KA: 000005AK106 2.9/3.8 LEVEL: MEMORY
REFERENCE: Technical Specifications Bases 3/4.1.3
LEARNING OBJECTIVE: 052302E09.
HISTORY: New
JUSTIFICATION:
a. This is the reason for the power limit.
b. Other operating procedures maintain rods above the Rod Insertion Limits.
i
i'
c. The reduction of the high neutron flux trip setpoint to 85% prevents exceeding thermal power
limits.
d. When QPTR limits are determined to be exceeded then power is required to be reduced to
less than 50%.
!
.. . _ ,
_ . ~ _ _ - _ _ _ _ . _ _ _ _ . _._._.__.__m_ . _ _ _ _ . _ . _ _ _ _
QUESTION No. 044:
Given the following plant conditions:
i - A large break LOCA has occurred on Unit 1.
- The actions required by EEP-0, REACTOR TRIP OR SAFETY IN)ECTION, have
been completed.
- Transition to EEP-1, LOSS OF REACTOR OR SECONDARY COOLANT, has been
made.
'
- EEP-1 has been completed up to Step 18," Check when to transfer to cold leg
recirculation".
- All ECCS systems are operating normally. l
- RCS pressure is stable at 200 psig.
Which one of the following mechanisms describes the PRIMARY method of decay heat removal.
a. Condensation of reflux boiling in the Steam Generators.
b. Natural circulation cooling between the RCS and Steam Generators.
c. ~ Forced circulation cooling between the RCS and Steam Generators.
d. Forced circulation cooling from the RWST through the cold legs and out the
break.
ANSWER: d.
'
l
KA: 0000llEA21-0 4.5/4.7 LEVEL: ANALYSIS / COMPREHENSION
REFERENCE: OPS 52530B, pg. 4
LEARNING OBJECTIVE: 052102B02.
HISTORY: New
JUSTIFICATION:
a. Reflux boiling would not have been established as long as there is injection flow.
I
b; Natural circulation flow would not have been established as long there is injection flow. !
i
c. Forced circulation would not be established because the RCPs would have been stopped. )
' d. This is the method of decay heat removal.
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, - - .. _.- . - . - . .-- ~ - - - - - . . - . - . - - .- . _ . . - .. . ~ .. - -- .
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QUESTION No. 045.
Given the following plant conditions:
- Unit 2 has had a loss of off-site power and an inadvertent Safety Injection.
- The actions of EEP-0, REACTOR TRIP OR SAFETY INJECTION, have been
completed.
- The crew has transitioned to ESP-1.1, SI TERMINATION.
- Off-site power has been restored at Step 6. Of ESP-1,1.
Which one of the following is the reason that the Reactor Coolant Pumps (R cps) are not started
until Step 22. of ESP-1.17
!
a. To prevent damage to the Reactor Coolant Pumps.
b. To allow time SI flow to be terminated to prevent excessive cooldown of the
reactor vessel.
I
c. Provide time for an engineering evaluation prior to staning the RCPs.
d. To allow time for steam generator level to be restored to that required to cover
the U +ubes.
ANSWER: a
KA: EO2EK22 3.5/3.9 LEVEL: MEMORY
!
REFERENCE: OPS 52531E, pg. 25 l
LEARNING OBJECTIVE: 052531E07 or 052531ElI
HISTORY: New
JUSTIFICATION:
a. - The RCPs are not started until all support co"ditions have been established and verified in l
order to prevent damage to the pumps.
b. SI flow during a LOSP could cause a large cooldown of the reactor vessel, but starting the l
RCPs will not cause a further reduction of temperature. !
c. An engineering evaluation is' required only if both sealinjection and CCW have been lost.
d. This is a concern in FRP-C.1 but it does not apply for these plant conditions. .
)
,
t -,. ,a .r
. ... . . ._ __ .. _ . . _ .- . .-.._ - _ _ _ _ _ _ _ _ _ . _ _ - - - .
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L QUESTION No. 046:
A reactor coolant pump #2 seal failing open would be indicated by:
a. #2 seal leakoff high flow alarm, #1 seal leakoff high flow alarm, rising RCDT
l level
b. #2 seal leakoff high flow alarm, #1 seal leakofflow flow alann, rising RCDT i
, level i
1
,
l c. #2 seal leakoff high flow alarm, low standpipe level alarm not clearing while
l filling
d. #1 seal leakoff high flow alarm, #1 seal low delta p alarm, seal injection flow I
l
mismatch between .RCPs
ANSWER: b
KA: 000015AAl22 4.0/4.2 LEVEL: ANALYSIS
<
REFERENCE: FNP-1-ARP-1.4, DAS and DCl
l
LEARNING OBJECTIVE: OS2101D11
HISTORY: 052101D17005. !
1
JUSTIFICATION. ,
a. A number 2 seal failure would cause number 1 seal leakoff to be high not low.
- b. This is the correct answer. I
i c. A number 2 seal failing open would not cause RCP standpipe level to fall.
d. A number 2 seal failing open would not cause a low RCP seal delta p.
1
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... . .. . . - . ..
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QUESTION No. 047: ,
l
l
Given the following plarit conditions:
'l
- I A, IB, and IC RCPs are off and cannot be renarted.
L
)
j -
RCS core exit thermocouples indicate 490 F. !
RCS pressure is 1400 psig. !
-
RCS cooldown rate is 90 F per hour.
-
SG atmospheric wilefvalves are being used to control cooldown.
3
. ~ Auxiliary spray being used to depressurize the RCS. !
-
PRZR levelis 90% and increasing. l
-
RVLIS is 0% upper head, > 100% upper plenum.
- . RCS depress.arization has begun in ESP-0.3, Natural Circulation Cooldown with
i
Allowance fcr Reactor Vessel Head Steam Voiding. ]
' Which one of the following actions is required?
a. Reduce charging flow.
b~. Isolate SI accumulator.
c. Repn ssurize RCS to collapse the void in the reactor vessel head.
d. Perform actions of FRP-I.3, RESPONSE TO VOIDS IN REACTOR VESSEL
- to collapse void.
ANSWER:- a.
KA: - E10EK33 3.4/3.6 LEVEL: COMPREHENSION
REFERENCE: OPS 52531L, pg. 26 and FNP-1. ESP-0.3, pg.10
. LEARNING OBJECTIVE: 052531LO6 ,
HISTORY: 052531LO6003 with stem and one distracter modified.
JUSTIFICATION:
'
. a. Reduction of charging flow is required to prevent the pressurizer from going solid.
b. Isolation of the SI accumulators is not required until RCS pressure is less than 1000 psig.
c. '.With RVLIS indication greater than 44% in the upper plenum repressurization is not required, j
'd. _ This procedure should not be performed if a controlled natural circulation cooldown is in -
progress.
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,
QUESTION No. 048:
l
l Given the following plant conditions:
- AOP 27.0 has been entered.
- Initial RCS boron concentration was 900 PPM.
- RCS Tavg is 521 F fuMowing an uncontrolled cooldown.
- Normal emergency boration has been in progress for 2 minutes.
- FI-110, BORIC ACID EMERG BORATE, indicates 40 gpm.
- Charging flow is 60 gpm.
Which one of the following is the minimum amount of time emergency boration must continue?
(Step 7.3 ofFNP-1-AOP-27.0 attached).
a. Two minutes.
b. Four minutes.
c. Six minutes.
d. Eight minutes.
ANSWER: b.
KA: 000024AK101 3.4/3.8 LEVEL: ANALYSIS / COMPREHENSION
REFERENCE: FNP-1-AOP-27.0, pg. 4
LEARNING OBJECTIVE: OPS 52521 A04 i
HISTORY: New
JUSTIFICATION:
a. If 60 gpm charging flow instead of 40 gpm emergency boration flow is used the incorrect
answer of 2 minutes could be chosen as the answer.
b. This is correct based on no interpolation and the most conservative values.1200 PPM
requires 60 gallons per F below 525 F = 60 X 4 = 240 gallons ofboron required. Emergency
boration flow rate is 40 gpm which would require 6 minutes of emergency boration. Boration
has been in progress for 2 minutes, therefore a emergency boration must continue for a
minimum of 4 more minutes.
c. If the 2 minutes of emergency boration that has already taken place is not taken into account
then 6 minutes could be chosen as the answer.
d. If the 2 minutes of emergency boration that has already taken p. ace is mcorrectly added to the
total time of 6 minutes required to add 240 gallons of boron then 8 minutes could be chosen
as the answer.
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FNP-1-AOP-27.0 EMERGENCY BORATION Revision 7
!
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.
Step Action / Expected Response Response NOT Obtained
i I I I I I
NOTE: In response to an uncontrolled cooldown below 525'F, the cold
shutdown boron concentration is the maximum boron concentration
required regardless of the extent of the cooldown.
7.2 Check RCS TAVG - LESS THAN 7.2 Perfonn the following.
525*F.
7.2.1 Verify shutdown margin
greater than Technical
Specification requirement
using FNP-1-STP-29.1,
SHUTDOWN MARGIN CALCUIATION
(TAVG 547'F) or
FNP-1-STP-29.2, SHUTDOWN
MARGIN CALCULATION
(TAVG < 547'F OR BEFORE THE
INITIAL CRITICALITY
FOLLOWING REFUELING).
7.2.2 WHEN shutdown margin
greater than Technical
Specification requirement,
THEN proceed to step 8. l
7.2.3 Continue emergency baration
and return to step 5.
7.3 Continue emergency boration
based on initial boren
concentration and RCS TAVG.
Aporoximatq_Boration
Initial RCS Each *F TAVG
Boron Is Less Than
Concentration 525*F
0 ppm 50 gal
300 ppm 52 gal
600 ppm 55 gal
1200 ppm 60 gal
1500 ppm 64 gal ,
1800 ppm 68 gal _
.
Page Completed
Page 4 of 6
. . . . -_ - - - - . - . - . - - . . - . - . . -.. . . . -
QUESTION No. 049: :
Given the following plant conditions:
- Unit 2 is at 33% power.
- A loss of Component Cooling Water has occurred.
- The crew is taking the actions of FNP-2-AOP-9.0, LOSS OF COMPONENT
COOLING WATER. i
Which one of the following actions specified by AOP-9.0 will require main generator load to be
reduced? (Consider each action separately.)
a. Stopping 1 A RCP due to increasing temperatures.
b. Aligning the charging pump suction to the RWST.
c. Securing loads on the miscellaneous header.
d. Aligning the IB Charging Pump to the opposite train.
ANSWER: b.
KA: 000026AK303 4.0/4.2 LEVEL: ANALYSIS
REFERENCE: FNP-1-AOP-9.0, pg. 6
LEARNING OBJECTIVE: 052520I02
HISTORY: New
!
JUSTIFICATION:
a. Reactor power is below the single loop loss of flow setpoint.
b. This alignment will borate the RCS and require turbine load to be reduced to maintain Tavg. j
c. This is an action required by AOP-9.0 but does not require load reduction. !
l
d. This is an action required by AOP-9.0 but, although this action is considered urgent, does not !
require load reduction. ;
i
-_- -- , _
.- -- .-. - - - - - . . - - . - - - - - . . . -.- -
QUESTION No. 050:
Given the following plant conditions:
- Unit 1 failed to trip following the receipt of a valid trip signal. .
- FNP-1-FRP-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATWT, has
been entered.
- All attempts to trip the reactor have failed.
- Manual rod insertion and emergency boration is in progress.
- Both intermediate range start up rates are NEGATIVE.
- All power range instruments indicate below 5% power. ;
- Control banks are partially insened.
- Shutdown banks are fully withdrawn.
Which one of the following states the action required under these conditions?
a. Stop driving rods, stop emergency borating and go to EEP-0.
b. Continue driving rods, continue emergency borating and go to EEP-0.
c. Continue driving rods, stop emergency borating and go to EEP-0.
d. Stop driving rods, continue emergency borating and stay in FRP-S.I.
ANSWER: b.
KA: 000029EK311.4.2/4.3 LEVEL: COMPREHENSION
REFERENCE: OPS 52533 A, pg. 36 & 37
LEARNING OBJECTIVE: 052533A16
HISTORY: 052533Al2014 with two distracters modified.
JUSTIFICATION:
I
a. The conditions allow returning to EEP-0 but emergency boration and rod insenion should
continue.
b. Rod insertion and emergency boration should continue until adequate shutdown margin is
established and EEP-0 can be reentered. l
c. The conditions allow returning to EEP-0 but emergency boration should continue until
adequate shutdown margin is established.
d. Emergency boration and rod insertion should continue, but transition to EEP-0 should occur i
i
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<
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QUESTION No. 051:
Oiven the following plant conditions:
l:
-
Unit I was operating at 100% power.
L -
A Reactor Trip and Safety Injection hasjust occurred due to high Containment
l- pressure.
-
Pressurizer level and prersure are falling rapidly
RCS temperature is 520-/ and falling
Which one of the following parameters is the cause of the above indications:
a. Steam line break.
b. Small break LOCA.
c. Pressurizer vapor space break. I
d. Steem generator tube rupture.
ANSWER: a.
KA: 000040AA203 4.6/4.7 LEVEL: ANALYSIS / COMPREHENSION
REFERENCE: OPS 525210, pg. 3
LEARNINO OBJECTIVE: 053002J15.
HISTORY: .New
. JUSTIFICATION:
-- a. A steam line break will produce all these symptoms
. b. A small break LOCA will produce all these symptoms but RCS temperature falling.
c. A Pressurizer vapor space break will produce all these symptoms but pressurizer level and
RCS temperature falling.
d. ~ A steam generator tube nipture will not cause containment pressure to rise.
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QUESTION No. 052:
Given the following plant conditions:
.
-
Steam generator IB is faulted due to a steam break outside of containment.
! -
The crew is performing actions of EEP-2 " FAULTED STEAM GENERATOR
l ISOLATION."
-
Only the TDAFW pump is mnning.
-
Steam generator narrow levels are as follows: ;
- A = 19%
. B = 7%
- C = 15%.
Which one of the following actions concerning the AFW pumps is required?
a. Isolate the TD AFW pump from IB steam generator by isolating air to the
main steam valve room.
b. Isolate the TDAFW pump from IB steam generator by closing the manual
isolation valve in the main steam valve room.
c. The TDAFW pump should n_o_t be isolated from IB steam generator for these l
conditions.
d. Isolate the TDAFW pump from IB steam generator at the hot shutdown panel .
'
after ensuring it will continue to run from IC steam generator supply.
ANSWER: d.
KA: 000040AA110 4.1/4.1 LEVEL: COMPREHENSION .
I
REFERENCE: OPS 52530C, pg.13
LEARNING OBJECTIVE: 052530C05
HISTORY: New
JUSTIFICATION:
a. This would close the steam supply from the IB steam generator to the TDAFW pump, but it
would also close the supply from the IC steam generator.
b. Since the fault is outside containment,1B steam generator would be blowing down to the
main steam valve room and the manual isolation valve would be inaccessible.
c. This would be correct if there were no way to isolate the TDAFW pump without affecting the
IC steam generator supply.
d. The procedure directs that HV-3226 be failed open in the lower equipment room and then the
TDAFW pump can be isolated from 1B steam generator at the hot shutdown panel.
. _. .. -. _ _ . ._ .- _ _ _ . _ . _ . _ . . . . _ _ _ . _ _ . _ . . . . _ _ . _ - . _ _ _ _ _ . _ _ _ _ _ . . _ . . _ .
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' QUESTION No. 053: j
!
Given the following plant conditions:
i
- Unit 1 experienced a main steam line break at 1000. ,
- At 1200 the following conditions existed: I
- l
l RCS pressure was 700 psig and decreasing.
'
- RCS temperature was 350'F and decreasing. ~
- At 1300 the following conditions exist:
- RCS pressure is 500 psig and decreasing.
- RCS temperature is 260 F and decreasing.
Which one ofthe following is the applicable INTEGRITY Critical Safety Function path at 13007 ,
(FNP-1-CSF-0.4, Page 1 should be attached if required.)
j
a. Red'
b. Orange -
c. Yellow
d. Green
ANSWER: b.
KA: E08EA21 3.4/4.2 LEVEL: ANALYSIS /COMPREIENSION
REFERENCE: FNP-1-CSF-0.4, pg.1
LEARNING OBJECTIVE: 052533K13
HISTORY: New
JUSTIFICATION:
a.. A red path is an option in the Integrity CSF but since cold leg temperature decrease was less
than 100 F in the last 60 minutes this is not the correct path.
b. An orange path is the correct path based on the conditions given.
c. ' A yellow path is an option but since the cold leg temperatures are not greater than 270 F this. I
is not the correct path.
i
d. A green path is an option but since the cold leg temperatures are not greater than 310 F and
RCS pressure is_not less than 450 psig this is not the correct path. l
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FNP-1-CSF-0.4 INTEGRITY REVISION 7
G
(J m ==
,
g2_ _
l
m o y
$
US I 4
E o '
210 245 270 300
COLD LEG TEMPERATURE --*-
U O
--
_ _ _ _ _
. a ~s fgpfp 3
-
ALL RCS
I'RESS - NO GO TO
CL TEMP (IN F;
n a :.;3 ta m II
" FRP-P.1
-
L AST 60 MIN)
F OlNTS TO U
FIGHT OF YES
LIMIT A ALL RCS
=
CL TEMPS NO GO TO
k IN LAST
60 MIN
FR P-P.2
GRTR THAN YES I
270* F
ALL RCS
TEMP DECR CL TEMPS
NO
IN ALL CL NO L iNtAST
&+ IN LAST
60 MIN
LESS THAN
1000 F
YES
60 MIN
GRTR THAN
VES
-
._
_ . .,, / h-h GO TO
~~i' "'gtyj) FRP-P.1
un
ALL RCS NO
CL TEMPS
F GRTR THAN
270* F
YES
RCS PRESS NO
LESS THAN
j
I 450 PSIG GO TO
FRP-P.2
YES
ALL RCS
"
-
TEMPS
GRTR THAN
l ' CSF
3100F YES
!
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l' s CSF
NI
. . _ _ _ _ . _ . _ . - _ . .. _ _ _ . _ _. . .- _ _ - _ . _ __
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FN P-1 -CS F-0.4 INTEGRITY REVISION 7
A z.
(
N
l 4
1
INTEGRITY
'
RCS PRESSURE - TEMPERATURE CRITERIA
- - . . . ..
. _
.
,
_.
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.. . _ . ;
'
-- .
. . . . . ., .
.
. . .
. . . . .
' i
,
2500-2550---------------------------------------.
.
- - -
-
. - . . !
. ..
.
. .
p
~
.
-2050----------------
2000 ,
m
m
D
so :T
M A-
,)
^\ m
L a
$
$0
z E 1500
'
..
<L ~
E
W .
9
M
O
E
1000 -
)
. . . . . .. . ..
l
INTEGRITY INTEGRITY INTEGRITY INTEGRITY. . . . . .
RED PATH ' ORANGE PATH YELLOW PATH ' GREEN PATH
REGION .
REGION REGION REGION j
- .
- ..
500 - - - -- -
- ,
. .
ll
.. \
0- - - - -.. - - - - -
.r -.
.
210 245 270
175 200 225 250 275 300 325
l
i
E
4 RCS COLD LEG WIDE RANGE TEMPERATURE
(*F)
!
e
i
-. _ -._ - -_ .-.- - -_- - - - - . . ..-. - . .. -.-
,
,
QUESTION No. 054:
Given the following plant conditions:
- Unit 2 has been holding at 33% power.
l
- Condenser vacuum is slowly degrading. )
- Which one ;of the following alarms / indications will be the first to actuate?
a. LO VAC TURB TRIP will alarm.
b. TURB'COND VAC LO-LO will alarm.
c. CONDENSER AVAILABLE C-9 will go out.
d. l A or IB SGFP TRIPPED will alarm.
ANSWER: b.
KA: 000051 AA202 3.9/4.1 LEVEL: MEMORY
REFERENCE: FNP-1-ARP-1.7 GJ2, pg.1
LEARNING OBJECTIVE: 052520H01
HISTORY: New
I
-JUSTIFICATION:
a. This alarm causes a Turbine Trip on decreasing vacuum at 4.41 PSIA.
l
b. This alarm actuates on decreasing vacuum at 2.7 PSIA, when greater than or equal to 30% i
power
i
c. : Will go out on a decreasing vacuum at approximately 10.8 PSIA.
d, This alarm will actuate as a result of decreasing vacuum at 5.9 PSIA. I
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_ _ .
QUESTION ho. 055:
,
Which one of the following concerns would an operator most likely be confronted with during a
totalloss of AC power over an extended period of time?
a. Inadequate core cooling condition.
i
b. An unmonitored release ofradioactivity. i
c. Loss of containment integrity. !
!
d. Pressurized thermal shock.
l
ANSWER: a.
KA: 000055EK302 4.3/4.6 LEVEL: COMPREHENSION
REFERENCE: OPS 52532A, pg. 6 & 7
LEARNING OBJECTIVE: 052532A03 i
HISTORY: New
l
'
JUSTIFICATION:
a. Significant voiding in the SG U-tubes will prevent natural circulation and loss ofinventory will ,
preclude reflux boiling.
b. RCP seal leakage which is occurring will not lead to an unmonitored release.
c. RCP seal leakage will not result in a challenge to containment integrity.
d. Although RCS pressure will remain high and a RCS cooldown is in progress, it is
procedurally stopped prior to PTS becoming a concern.
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k
(. 4
l QUESTION No. 056: '
, Given the following plant conditions:
! .
.
l -_ A loss of all AC power has occurred on Unit 2.
- The actions required by FNP-2-ECP-0.0, LOSS OF ALL AC POWER, are in progress.
- SG atmospheric reliefvalves are being controlled locally to reduce SG pressure to less !
l than 200 psig.
! - A low steam line pressure SI signal has been received.
!
- Steam line pressure is 350 psig and RCS cold leg temperatures are at 325 F.
l - Source range startup rate is +0.2 dpm and steady. ;
-- The STA reports there is a Yellow path on subcriticality.
Which one of the following actions should be taken?
1
a. 'Begin an emergency boration.
b. Stop dumping steam.
c. Continue to lower SG pressure to < 200 psig.
d. Proceed immediately to FRP-S.2
ANSWER: b.
')
KA: 000055G2416 3.0/4.0 LEVEL: ANALYSIS
REFERENCE: FNP-1-ECP-0.0, pg. 29
I
LEARNING OBJECTIVE: 052532A08
HISTORY: 052532A08015
!
,
JUSTIFICATION: l
a. Emergency boration would be an action to mitigate the positive SUR, but cannot be done
without AC power.
b. If SUR is above zero the ECP-0.0 RNO requires securing dumping steam to heat up the RCS
and establish subcriticality.
I
c. If SUR is above zero the ECP-0.0 RNO requires local control of atmospheric relief valves to
raise SG pressure. 1
- d. While in ECP-0.0, CSFs are monitored for infonnation only.
.
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QUESTION No. 057:
l
Given the following plant conditions:
- Unit 1 is holding at 85% power due to problems with the IC Condensate Pump. I
- Rod control is in MANUAL with Bank D rods at 218 steps.
- VCT LT-112 failed low 30 minutes ago.
!
- I & C is troubleshooting Power Range Nuclear Instrument N-42 because of a blown
fuse.
Which one of the following conditions will occur if power is lost to the IC 120V AC Vital Bus?
a. A reactor trip will occur. '
b. A boration of the RCS will begin.
c. Control rods will begin stepping in. !
d. Only automatic rod withdrawalis blocked.
ANSWER: a.
KA: 000057AA219 4.0/4.3 LEVEL: ANALYSIS / COMPREHENSION I
REFERENCE: OPS 52103D, pg. T-4b
LEARNING OBJECTIVE: 052103D20
HISTORY: 052103D20014 with stem and two distracters modified.
JUSTIFICATION:
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a. A reactor trip will occur since another PRNI channel has already been placed in a tripped
condition.
b. A boration of the RCS would have occurred if power was lost to 1 A 120V AC Vital Bus.
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c. Although loss of power will occur to another PRNI control rods will not step in.
d. A rod block would occur if this fault was the first PRNI to lose power.
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! QUESTION No. 058:
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Given the following plant conditions:
- Unit 2 has had a steam generator tube rupture.
- 380,000 gallons ofcontaminated secondary condensate has resulted.
Which one of the fo!!owing methods would be used to store this volume to allow decay of short
lived radioactivity and prevent an inadvertent release?
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a. Condensate storage tank.
i b. Condenser.
c. Turbine building sump.
d. Waste monitoring tanks.
ANSWER: a.
KA: 000059G2311 2.7/3.2 LEVEL: MEMORY
REFERENCE: FNP-0-AOP-2.1, pg. 5
LEARNING OBJECTIVE: 052520B01.
HISTORY: New
JUSTIFICATION:
a. The CST can hold up to 500,000 gallons and is one of the options in AOP-2.1.
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b. The condenser is one of the options in AOP-2.1 but can hold only 280,000 gallons.
c. The turbine building sump is one of the options in AOP-2.1 but each sump can only hold
30,000 gallons.
d. Although the waste monitoring tanks are used in contaminated liquid waste processing they
can only hold 10,000 gallons each.
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QLESTION No. 059:
Which one of the following is the length of time it will take to receive a Reactor Coolant Pump
(RCP) high bearing temperature following loss of Component Cooling Water flow to the RCP
motor oil cooler?
r
a. One minute.
f
b. Two minutes.
c. Five minutes.
d. Thiny minutes.
ANSWER: b.
KA: 000062AA206 2.8/3.1 LEVEL: MEMORY
REFERENCE: FNP-1-AOP-9.0, pg. 3
LEARNING OBJECTIVE: 052520102
HISTORY: New
JUSTIFICATION:
a. One minute is the approximate length of time an RCP must be running before the oil lift pump
can be secured.
b. Approximately two minutes after loss of CCW flow to the RCP motor oil coolers high bearing
temperatures will occur on any running RCP.
c. Time limit for starting an RCP if all RCP's have been idle with seal water flow established
during solid plant operations. UOP 1.1 appendix 5.
d. Minimum idle time prior to restarting an RCP after an attempted stan.
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QUESTION No. 060: l
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l Which one of the following conditions will require the use of reactor head vents to assist in plant l
recovery when operating from the hot shutdown panels due to a cable spreading room fire? ,
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a. Loss ofReactor Coolant Pumps. !
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b. Pressurizer level decreasing below 15% level. !
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c. Steam Generator levels decreasing below 25% level. j
d. Pressurizer pressure dropping in an uncontrolled manner. ,
ANSWER: b.
KA: 000067AK304 3.3/4.1 LEVEL: MEMORY l
R2FERENCE: FNP-1 AOP-28.1, pg.13
LEARNING OBJECTIVE: 052521C04.
HISTORY: New
JUSTIFICATION:
a. Natural circulation can be used following loss of RCPs
b. Pressurizer level decreasing below 15% will result in letdown isolation with the inability to
reopen LCV 459 and 460 requiring the use of the head vents for removing mass from the
RCS.
c. Control of steam generator levels is available at the hot shut down panels.
d. The PORV block valves can be shut to stop pressure decrease and allow recovery.
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QUESTION No. 061:
Given the following plant conditions:
- Ui.it 1 Control Room has been evacuated due to inaccessibility.
- The actions of FNP-1-AOP-28.0, CONTROL ROOM INACCESSIBILITY, are in
progress.
Which one of the following components is electrically isolated from the Control Room by direct
actine transfer?
a. Main Steam Isolation Valves,
b. Pressurizer PORVs and their block valves.
c. Charging Pump Suction Valves LCV-115B & D.
d. Turbine Driven AFW Pump and Flow Contrcl Valves. l
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ANSWER: d.
KA: 000068AK303 3.7/4.3 LEVEL: MEMORY
REFERENCE: OPS 52202D, pg. 6
LEARNING OBJECTIVE: 052202D06
HISTORY: This is a new question. A similar question,052202D06007, used on a previous quiz
tested the same learning objective in a different manner.
JUSTIFICATION:
)
1
a. MSIVs are isolated by indirect acting transfer.
b. Pressurizer PORVs and block valves are isolated by indirect acting transfer.
c. 115B & D are isolated by indirect acting transfer.
d. Turbine Driven AFW Pump and Flow Control Valves are isolated by direct acting transfer.
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QUESTION No. 062:
Which one of the following conditions concerning the Personnel Airlock would exceed a Limiting
Condition for Operation and require entering an Action Statement of Technical Specifications?
a. The auxiliary airlock fails its LLRT while control rod unlatching is in progress.
L b. Welding cables are laid through both airlock doors during RCS fill and vent at
the end of an outage.
c. The outer and inner doors are opened simultaneously during a normal ,
cooldown prior to aligning RHR to the RCS. ;
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d. The door interlocks are cMated while reactor vessel head is being tensioned.
,
ANSWER: c.
KA: 000069G222 3.4/4.1 LEVEL: ANALYSIS
REFERENCE: Technical Specificatiori 3.6.1.3
LEARNING OBJECTIVE: 052302J04.
HISTORY: New
JUSTIFICATION:
a. The requirement for the airlocks is one door being closed during core alterations, the air lock
does not have to be OPERABLE.
b. RCS fill and vent is only done during MODE 5 and the technical specification LCO is not
applicable in MODE 5.
c. A normal cooldown requires RHR to be aligned to the RCS prior to 310 F (still in MODE 4).
The technical specification LCO is applicable in MODE 4 and requires at least one air lock
door be closed.
d. The plant condition described is MODE 6 and the interlocks are not required operable in
MODE 6
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l QUESTION No. 063. >
l. Given the following plant conditions: ,
- A reactor trip and safety injection has occurred on Unit 2. :'
- FNP-2-FRP-C.1, RESPONSE TO INADEQUATE CORE COOLING, has been
entered.
- At Step 4, Establish RCP suppon conditions, it is determined RCP support conditions
are NOT available.
- RWST level has decreased to 12 feet. l
Which one of ti.e following actions must be performed IMMEDIATELY? i
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a. Continue in FRP-C 1 at step 5, Check SI accumulator discharge valve status.
b. Stop all operating Reactor Coolant Pumps. I
c. Restore Reactor Coolant Pump support conditions,
d. Transfer to cold leg recirculation per ESP-1,3.
ANSWER: d.
KA: 000074G244 4.0/4.3 LEVEL; MEMORY
REFERENCE: OPS 52533C, pg.14 & 15
LEARNING OBJECTIVE: 052533C04
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HISTORY: Similar question used on one of the audit exams given to the candidate but stem, i
correct answer and one distracter are modified. j
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JUSTIFICATION: .
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a. Since there is no RNO for Step 4 continuing with Step 5 would be correct if RWST had not
decreased below 12.5 feet. )
b. RCPs are not stopped until Step 15 ofFRP-C.1.
c. There is no requirement to restore RCP support conditions.
d. Step 1 of FRP-C.1, Monitor RWST level, is a continuous action step which requires going to
ESP-1.3 anytime RWST level decreases below 12.5 feet.
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. QUESTION No. 064:
!
Which one of the following is the reason that Technical Specifications requires Tavg to be
immediately decreased below 500*F after the reactor is shut down for excessive Reactor Coolant
Activity.
a. Minimize temperature related degradation of the CVCS demineralizers during
subsequent R.CS clean-up.
b. Prevent the uncontrolled release of radioactivity if a steam generator tube
rupture occurs.
c. Mini:nize potential for containment contamination from inadvertent pressurizer
PORV operation.
d. Allow rapid depressurization of the RCS to reduce iodine spiking from the .
fuel.
ANSWER: b.
KA: 000076AK305 2.9/3.6 LEVEL: MEMORY
REFERENCE: Technical Specification Bases 3/4.4.9, pg. B 3/4 4-6
LEARNING OBJECTIVE: 052302H08.
HISTORY: New
JUSTIFICATION:
a. Letdown volume and flow rate determine the temperature of the liquid flowing through the
demineralizers.
b.' A steam generator tube rupture with high RCS activity could result in an uncontrolled release
-
through the atmospheric steam relief valves.
c. Activity in the containment can be monitored and release controlled.
d. Iodine spiking is not a concern following plant shutdown.
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QUESTION No. 065:
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Given the following plant conditions:
i
- An inadvertent Safety Injection Signal has generated a reactor trip on Unit 1. !
- Reactor power is decreasing. ;
- All rod bottom lights are lit. l
- Reactor trip breaker A and reactor trip bypass breaker A are open.
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- Reactor trip breaker B is closed and reactor trip bypass breaker B is open. i
- BYP & PERMISSIVE SAFETY INJECTION ACUTATED status light is lit. ;
- TSLB214-1,2,3, and 4 are lit. ;
- 160V ESF Bus A Train power available light is lit. ,
'
.J0V ESF Bus B Train power available light is NOT lit.
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of the following describes the operator actions required for the first four steps of FNP- !
, rGACTOR TRIP OR SAFETY INJECTION, for the above conditions?
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a. Verify the expected actions of Steps 1,2,3 and 4. .
b. Perform RNO actions of Step 1, then verify expected actions of Steps 2,3 and
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c. Verify expected actions of Steps 1 and 2, perform RNO actions of Step 3, then i
verify expected actions of Step 4.
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d. Perform RNO actions of Step 1, verify expected actions of Step 2 perform l
RNO actions of Step 3, then verify expected actions of Step 4.
ANSWER: a. l
KA: 000007G2.4.49 4.0/4.0 LEVEL: MEMORY
REFERENCE: FNP-1-EEP-0, pg. 5-7
LEARNING OBJEr'TIVE: 052530A03
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HISTORY: New
JUSTIFICATION:
a. The conditions presented in the stem require the first 4 Steps be verified only,
b. Step 1 requires " Check any reactor trip and its associated reactor trip bypass breaker open."
The "A" trip and bypass breakers meet these conditions, therefore step 1 RNO actions not
required.
c. . Step 3 requires " Verify at least one train of 4160 V ESF busses energized." Train A power
available light lit meets these conditions, therefore step 3 RNO actions not required. l
d. Seejustifications b. and c. above.
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QUESTION No. 066: !
Given the following plant conditions: I
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1 - Unit 2 is at 18% power rolling the main turbine
- Tavg is on program
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Which one of the following describes the immediate resocyxof the main feedwater regulating
..
bypass valves and the motor driven auxiliary feedwater pumps (MDAFW) if a reactor trip occurs ;
from these conditions?
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i a. The main feedwater regulating bypass valves close and the MDAFW pumps
remain secured. '
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b. The main feedwater regulating bypass valves remain open and the MDAFW
,
pumps auto start.
c. The main feedwater regulating bypass valves close and the MDAFW pumps
auto stan.
d. The main feedwater regulating bypass valves remain open and the MDAFW
pumps remain secured.
ANSWER: a.
KA: 000007EA102 3.8/3.7 LEVEL: COMPREHENSION
REFERENCE: OPS 5220ll, pg. 35 '
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l LEARNING OBJECTIVE: 052530Al2 !
HISTORY: New
JUSTIFICATION:
1
a. The main feedwater regulating bypass valves close due to Tavg < 554 F and the reactor trip
- breakers open but the SGFP does not trip therefore the MDAFW pumps will not start until i
steam generator level goes below 25%
l b. The main feedwater regulating bypass valves should close and the MDAFW pumps should not
auto stan.
c. The MDAFW pumps should not auto stad i
d. The main feedwater regulating bypass valves should close.
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QUESTION No. 067:
!
Given the following plant conditions: i
- Unit 2 was shutdown due to a pressurizer safety valve seat leakage. !
- Pressurizer pressure is 2235 psig.
- PRT pressure is 15 psig.
- PRT temperature is 150*F
- Ambient heat losses are negligible.
- Steam quality in the pressurizer bubble is 100%. I
Which one of the following would be the approximate temperature on the downstream tailpipe
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temperature indicator, on the MCB, caused by the leaking pressurizer safety valve? i
a. 200 F 5
b. 220"F
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c. 250 F
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d. 280 F '
ANSWER: b.
KA: 000008AK202 2.7/2.7 LEVEL: COMPREHENSION
REFERENCE: Steam Tables
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LEARNING OBJECTIVE: 052530A17.
HISTORY: A calculation of this type was used on one of the candidates audit exams but the i
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stem, correct answer and 2 distracters were changed
JUSTIFICATION:
)
a. 200 F is a value the applicant might choose if saturation conditions of 0 psia were used.
b. Since the leakage throttling action is isenthalpic, the leakage at 2250 psia will reach saturation
conditions at 30 psia, which corresponds to a temperature of 220 F. l
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c. 250*F is a value the applicant might choose if saturation conditions of 15 psia were used. I
d. 280 F is a value the applicant might choose if the isentropic line was used instead of the
constant pressure line. !
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QUESTION No. 068:
l Which one of the following is the reason the Reactor Coolant Pumps (RCPs) are required to be l
'
, tripped during a small break LOCA?
1
a. To eliminate RCP heat input into the Reactor Coolant System (RCS).
b. To avoid RCP impeller damage as dissolved gases come out of solution at .
lowe. pressures. !
c. To avoid significant clad heatup if RCP trip is delayed and occurs later in the j
accident. l
d. To prevent RCP seal damage during subsequent RCS depressurization.
ANSWER: c.
KA: 000009EK323 4.2/4.3
REFERENCE: OPS 52530A, pg. 30
. LEARNING OBJECTIVE: 052530A07 LEVEL: MEMORY
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HISTORY: New :
>
JUSTIFICATION: i
a. Although the RCPs would be adding heat under these conditions this is not the reason the l
pumps are tripped. !
b.~ This is a concern with RCPs but the pumps do not have to be tripped as long as RCS
subcooling is maintained at the core exit.
c. The RCPs are tripped in a timely manner to prevent significant core uncovery and clad heatup
if RCP trip is delayed and the RCPs are subsequently stopped.
d. Seal damage will not occur if seal injection flow is maintained as required by a caution in
EEP-0 and EEP-1.
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QUESTION No. 069;
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Which one of the following procedures can be entered based solely on operatorjudgment?
a. FNP-1-ESP-1.3, TRANSFER TO COLD LEG RECIRCULATION.
b. FNP-1-ESP-3.1, POST-SGTR COOL DOWN USING BACKFILL.
c. FNP-1-ECP-0.1, LOSS OF ALL AC POWER RECOVERY WITHOUT SI
j REQUIRED.
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d. FNP-1-ECP-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION.
ANSWER: D.
l KA: W/E11EA2.1 3.4/4.2 LEVEL: MEMORY
REFERENCE: FNP-1-ECP-1.1, pg. 1
LEARNING OBJECTIVE: 052532D01.
HISTORY: New
JUSTIFICATION:
a. ESP-1.3 is entered when RWST level is less than 12.5 ft from steps in 8 procedures and a
foldout page but not at the operator's discretion.
b. ESP-3.1 is entered from two other procedures; EEP-3 and ESP-3.2 but not at the operator's
discretion.
c. ECP-0.1 is entered from ECP-0.0 but not at the operator's discretion.
d. ECP-1.1 can be entered from four other procedures and based on operatorjudgment, when
emergency coolant recirculation capability or containment spray recirculation capability is lost
due to the clogging of sump suction screens.
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QUESTION No. 070: ,
AOP-28.2 FIRE IN THE MAIN CONTROL ROOM has the operator open LCV 1 ISB !
and LCV115D CHARGING PUMP SUCTIONS FROM THE RWST prior to exiting the l
control room. What is the purpose of this action?
l
a. To prevent a loss RCS inventory due to charging pump cavitation. '
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b. To ensure adequate shutdown margin by borating the RCS.
c. To maintain VCT level above the auto makeup setpoint.
d. Because a VCT level transmitter failure could prevent valve operation.
ANSWER: a.
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KA: 000022AA108 3.4/3.3 LEVET : MEMORY
REFERENCE: FNP-1-AOP-28.2 latest change
LEARNING OBJECTIVE: OS2101F05, and 052521C04
HISTORY: New
JUSTIFICATION:
a. A fire in the control room could cause one of the VCT OUTLET valves to close thus a loss of
suction source to the charging pumps.
b. Opening the valves would cause a RCS boration.
c. Opening the valves would prevent a loss of VCT level.
d. VCT level failing high would prevent auto roll over but not manual operation
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QUESTION No. 071:
Given the following plant conditions:
f
- Unit 2 is shutdown and on Residual Heat Removal (RHR) cooling down per UOP-2.2 l
- RHR Train A is in service.- !
- A non-recoverable loss ofinstrument air occurs.
Which one of the following describes how the RCS temperature is affected by the loss of l
instrument air? ;
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a. Decreases because the RHR heat exchanger discharge valve, RHR-HV-603A, j
fails open.
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b. Increases because the RHR heat exchanger bypass valve, RHR-FCV-605A, i
fails closed. l
c. Decreases because the RHR to letdown line valve, CVC-HCV-142, fails open.
d. Increases because the RHR miniflow valve, RHR-FCV-602A, fails closed.
ANSWER: a.
KA: 000025AK101 3.9/4.3 LEVEL: ANALYSIS l
REFERENCE: FNP-1-AOP-6.0, pg. 9
LEARNING OBJECTIVE: 05210lK10
HISTORY: New i
JUSTIFICATION:
!
a. Los's of IA will cause HV-603A to fail open maximizing flow and cooldown through the heat
exchanger decreasing RCS temperature. ,
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b. Loss of IA will cause FCV-605A to fail closed increasing flow and cooldown through the
heat exchanger decreasing RCS temperature.
c. Loss oflA will cause HCV-142 to actually fail closed and should have no affect on RCS
temperature.
d. Loss ofIA will have no effect on FCV-602A because this valve is motor operated.
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QUESTION No. 072:
A plant startup is in progress at the point of rolling the main feed pump when
PT-444 fails higi. Assuming no operator action what would be the plant response: I
a. RCS pressure would cycle around 2000 psig.
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b. RCS pressure would rise to 2335 psig.
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c. Reactor trip would occur at 1865 psig.
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d. Reactor trip would occur at 1850 psig.
ANSWER: d.
KA: 000027AA202 3.8/3.9 LEVEL: COMPREHENSION
REFERENCE: OPS 5220lH, pg.12 & 13
LEARNING OBJECTIVE: 05220lH25
HISTORY: New
JUSTIFICATION:
a. This could be correct for a PT 445 failure. j
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b. This could be correct if PT 444 were to fail low. !
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c. This would be correct if reactor power were above 10%, but that does not agree with plant ;
conditions i
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d. RCS pressure will fall rapidly until 2000 psig when P-11 closes the PORV, but the spray
valves being open would continue to reduce pressure until the reactor tripped due to low
pressurizer pressure safety injection.
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QUESTION No. 073: .
Given the following plant conditions:
{
- Unit 2 reactor is shutdown and the reactor trip breakers are open.
- RCS pressure is 2200 psig.
-. RCS temperature is 540*F and slowly decreasing. ,
- Source range channel N-31 is out of service for repairs. ;
- Source range channel N-32 fails low. !
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Which one of the following actions is required?
'a. ' Stop the cooldown and commence a heatup.
b. Verify shutdown margin within one hour.
c. Place channel N-32 in the tripped condition within six hours.
d. Borate to cold shutdown conditions. j
ANSWER: b. l
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KA: 000032G2.1.33 3A/4.0 LEVEL: MEMORY I
REFERENCE: Technical Specifications, Table 3.3-1, pg. 3/4 3-2 and Action 5
LEARNING OBJECTIVE: 052302G01
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HISTORY: New i
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JUSTIFICATION ,
a. Although this would be prudent there is no requirement.
b. This is the TS requirement for the number of operable channels less than the minimum
required.
c. Some NI channels are ter ired to be placed in the tripped condition within six hours; the SR
channels are not,
d. If the shutdown margin ; determined to be less than required by TS then emergency boration
is required.
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QUESTION No. 074:
Given the following plant conditionsi
- Critical data has been recorded and power escalation to 2% is in progress.
- NI-35 hasjust lost compensating voltage.
Which one of the following is the required operator action?
a. Enter EEP-0, REACTOR TRIP OR SAFETY INJECTION.
b, Stabilize reactor power prior to increasing power above 5%.
c. Go to BLOCK on both Intermediate Range Block Switches.
d. Decrease power below 1E-10 amps and manually energize the Source Ranges.
ANSWER: b.
- KA: 000033G2.4.4 4.0/4.3 LEVEL: COMPREHENSION
REFERENCE: OPS 52201D, Appendix 1
LEARNING OBJECTIVE: 052520R08
HISTORY:-New
JUSTIFICATION:
a. The given plant conditions should not result in a reactor trip.
b. This action would is required because power level is > P-6 and <P-10.
c. This action would be required if the failure had resulted in the IR HI FLUX AUTO / MAN
ROD STOP alarm but puwer level is too low for that to occur.
d. This action would be required if the reactor was required to be shut down for this failure, but
reactor shutdown is not required since power level is > P-6.
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QUESTION No. 075:
Given the following plant conditions:
- Unit 1 is operating at 100% power.
- SG TUBE LEAK ABOVE SETPT is in alarm. I
- R-70A, N-16 primary to secondary leak detection system, shows an increase from 10
gpd to 65 gpd in I hour and now has remained stable for the last hour.
- R-15, condenser air ejector gas monitor, shows a slight upward trend. l
- The pressurizer level recorder shows no change in level trend.
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Which one of the following actions is required?
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a. Confirm leak rate using FNP-0-CCP-31, LEAK RATE DETERMINATION l
prior to initiating any other action.
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b. Be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c. Raise the R-70A alert setpoint to 95 gpd.
d. Trip the reacter and go to EEP-0. I
ANSWER: c.
KA: 000037AK307 4.2/4.4 LEVEL: ANALYSIS l
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REFERENCE: FNP-1-ARP-1.6 FG1 and AOP-2.0.
LEARNING OBJECTIVE: 052520B01
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HISTORY: 052520B01003 with stem and 3 distracters modified.
JUSTIFICATION:
a. This is one of the actions within the procedures but it is not required to be completed prior to
any other action,
b This is one of the actions within the procedures but it is not required until leakage reaches 140
SPd.
c. This is an action if the leakage rate has not exceeded 60 gpd/hr and is above 30 gpd.
d. This action is one of the options required for a leak rate increase of > 60 gpd in < than I hour.
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QUESTION NO. 076
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A rapid cooldown is in progress due to a SG tube rupture. Which of the following is l
correct per EEF-3 ? l
a. RCPs must be tripped if subcooling falls to 16 *F during the cooldown.
1
b. AFW should be admitted to the ruptured SG narrow range level falls to 30%. l
c. Coo!down rate should be maintained less than 100 F/ hour.
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d. Cooldown must be stopped if pressurizer level goes below 7%. l
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ANSWER: b.
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~ KA: 000038EA136 4 3/4.5 LEVEL: MEMORY ;
REFERENCE: OPS 52530D, pg. 50
LEARNING OBJECTIVE: 052530D09
HISTORY: 052530D03007 with stem and 2 distracters modified.
JUSTIFICATION:
a. RCP trip criteria is discussed in the first step of EEP-3 but is not applicable once the
cooldown is started.
b. Ruptured SG level must be maintained above 30% to prevent the ruptured SG from
depressurizing during the cooldown.
c. The 100 F/ hour limit does not apply to the initial cooldown following a SG tube rupture
d. Pressurizer level is a criteria for terminating RCS pressure reduction..
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QUESTION No. 077:
j
Which one of the following statements explains why Auxiliary Feedwater (AFW) Flow as .
'
procedurally restricted to 100 gpm when recovering steam generator (SG) level if the level has !
fallen below 8% on the wide range indication? (Assume that heat sink requirements are met.) l
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a. To minimize water hammer in the SG feed ring. l
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b. To prevent disruption of RCS natural circulation.
c. To minimize thermal stresses to the SG components. l
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d. To prevent runout conditions of the AFW pumps.
ANSWER: c.
KA: 000054AK102 3.6/4.2 LEVEL: MEMORY I
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REFERENCE: FNP-1-FRP-H.5, pg. 3
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LEARNING OBJECTIVE: 052102H24.
]
HISTORY: New i
JUSTIFICATION:
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a. Although water hammer is a concern in starting flow in closed systems it is not a concern in
!
b. Adding AFW to a SG with a low level will enhance natural circulation, not disrupt it. !
c. Thermal stress concern is the reason AFW flow is restricted.
,
d. AFW pump runout is always a concern but the AFW system / components prevent AFW pump
runout.
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QUESTION No. 078: ,
t
- _ Given the following plant conditions
- Unit 2 is at 80% power ramping to 100%.
- Both Main Feedwater Pumps are operating. !
- All systems are aligned for automatic operation. ;
- SGFP SUCT PRESS LO alarm has actuated.
- SGFP pressure is 290 psig and decreasing.
)
Which one of the following is an action required by the operator? i
a. Manually trip one SGFP.
I
b. Start the standby condensate pump. :
c. Begin a rapid load reduction to 60%. j
d. Manually trip the reactor and enter EEP-0.
ANSWER: b.
KA: 000054G449 4.0/4.0 LEVEL: ANALYSIS / COMPREHENSION
REFERENCE: FNP-1-1 ARP-1.10 KB4. Pg.1
l LEARNING OBJECTIVE: 052201B19.
HISTORY: New
JUSTIFICATION:
l a. There is no requirement to trip one of the SGFPs because there is an automatic trip on low
! suction pressure.
b. The standby condensate pump will auto start at 275 psig but the ARP requires it to be
manually started prior to reaching 275 psig.
c. A rapid load reduction is required if the standby condensate pump is not available.
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d. This action would be appropriate if suction pressure decreases below 275 psig for 30 seconds.
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QUESTION No. 079:
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l Given the following plant conditions:
- Unit I reactor has tripped from full power.
- Safety injection has occurred.
- AFW flow cannot be established.
- All Steam Generator (SG) narrow range levels are at 15% and decreasing slowly.
- FNP-1-FRP-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, has been
entered.
- The following indications are observed:
- Pressurizer pressure is 1900 psig.
- SG wide range levels 48%,42% and 41%
e Contaimnent prenure is 10 psig.
. FifM hottest core exit temperature is 705 F.
Which one of the following actions should be performed as quickly as possible?
a. Stop all RCPs and initiate bleed and feed.
b. Establish main feedwater flow to an intact SG with one SGFP,
c. Establish condensate flow to an intact SG.
d. Cooldown the RCS and place RHR in service.
ANSWER: a.
KA: EOSEA21 3.4/4.4 LEVEL: COMPREHENSION
REFERENCE: OPS 52533F, pg. 24 & 25 and FNP-1-FRP-H.1, PG. 3 i
LEARNING OBJECTIVE: 052533F05
HISTORY: 052533F05006 with 3 distracters modified.
JUSTIFICATION: '
a. With the given conditions bleed and feed must be immediately initiated.
b. Establishing main feedwater flow is one of the major actions of FRP-H.1 if AFW flow cannot
be established but the bleed and feed criteria have precedence.
c. Establishing condensate flow is one of the major actions of FRP-H.1 if AFW flow and main
feedwater flow cannot be established but the bleed and feed siteria have precedence.
d. Placing RHR in service is one of the first steps in FRP-H.1 if hot leg temperatures are less
than 350 F.
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QUESTION No. 030: I
Given the following plant conditions:
- Unit 2 is operating at 100% power. !
- A leak has developed in the Instrument Air (IA) system and IA pressure is decreasing ;
slowly.
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- IA PRESS LO alarm hasjust actuated. i
Which one of the following IA System AUTOMATIC response has occurred? -
a. Instrument air to the turbine building has isolated. f
b. Instrument air to the service building has isolated. I
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c. All air compressors have started.
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d. Service air has been isolated.
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ANSWER: d.
KA: 000065AA208 2.9/3.3 LEVEL: MEMORY
IEFERENCE: FNP-1. AOP-6.0, pg. 2
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LEARNING OBJECTIVE: OS2520F01
HISTORY: Similar question used on the HLT-24 audit exam with the stem and two distracters
modified.
,
JUSTIFICATION:
a, This auto isolation will not occur until 45 psig. The IA PRESS LO alarm actuates at 75 psig.
b. This auto isolation will not occur until 55 psig. The IA PRESS LO alarm actuates at 75 psig.
c. This auto start will not occer until 70 psig. The IA PRESS LO alarm actuates at 75 psig.
d. This auto isolation will occur at 80 psig. j
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QUESTION No. 081:
'
!
Given the following plant conditions:
!
- Unit 1 is operating at 75% power.
- All control systems are in automatic.
- Tavg input to the Master Level Controller has failed at 554.6 F.
- Which one of the following would be the resultant pressurizer level?
a. 21.4 % ,
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b. 29.8 %
c. 46.5 % l
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d. 54.9 %
ANSWER: b.' i
KA: 000028AK203 2.6/2.9 LEVEL: ANALYSIS
l
REFERENCE: OPS 52201H pg.17,18,19 & 20 and Figure 8. I
LEARNING OBJECTIVE: 0552201H15
f
HISTORY: 05220lH15010 with stem, answer and 3 distracters modified.
JUSTIFICATION: i
i
a. This is the no load pressurizer level setpoint. l
h. This is the level that would be maintained if the Tavg input to the controller was at 554.6 F.
!
c. This is the level that would be maintained if the controller was functioning properly at 75%
power. ;
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d. This is the 100% power pressurizer level setpoint.
)
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QUESTION No. 082: !
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Which one of the following conditions, occurring during refueling operations, would require
emergency boration per Technical Specification 3/4.9.1, BORON CONCENTRATION?
!
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a. Keffis 0.95.
b. Keff decreases by 1% from 0.95 in a period of I hour.
l . c. Boron concentration of the RCS is 1950 PPM. !
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d. Boron concentration decreases by 50 PPM from 2050 in a period of I hour.
ANSWER: c. '
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l KA: 000036G2.1.33 3.4/4.0 LEVEL: MEMORY !
1
l REFERENCE: Technical Specification 3.9.1
LEARNING OBJECTIVE: OS2302M03 I
,
HISTORY: New
l
JUSTIFICATION:
a. Keffs0.95 requires no action.
b. I % delta k/k is discussed in the TS LCO as a conservative allowance for uncertainties but
Keffis 5 0.95 and decreasing so no action is required.
c. Emergency boration is required if boron concentration decreases below 2000 PPM.
l d. The TS includes a 50 PPM conservative allowance for uncertainties but if boron concentration
is > 2000 PPM no action is required.
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QUESTION No. 083:
Given the following plant conditions: I
.
'
- Unit 2 was operating at 95% power with all systems aligned for normal operation.
- The electrical grid was experiencing slowly decreasing frequency and voltage causing the
startup tranformer breakers to 2F and 20 4160 VAC buses to open
'
Which one of the following will cause the reactor to trip?
a. NIS negative rate
b. Temporary loss of power to NIS power ranges j
c. RCP Undervoltage trip
d. RCP breaker position
ANSWER: a.
KA: 000056AA242 4.1/4.1 LEVEL: MEMORY
REFERENCE: OPS 5220ll, pg. 24 & 25
LEARNING OBJECTIVE: 052201I12
HISTORY: New
JUSTIFICATION:
a. A loss of power to 2F and 2G buses deenergizes the CRDM mg sets resulting negative rate.
b. The NIS will remain energized due to the 120VAC inverters
c. The RCP ous undervoltage trip is well below the degraded grid actuation.
d. The RCP breaker position relays would deenergize if they were aligned to the SOLA
transformer but this no longer the normal alignment.
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QUESTION No. 084;
Given the following plant conditions: )
- An ALERT has been declared on Unit 2.
- The Plant General Manager has assumed the duties of Emergency Director. -
.
- The TSC, EOF and EOC are being manned. ;
,. Which one of the following personnel can grant permission to enter the MCB horseshoe area.
a. Emergency Director- j
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b. Shin Supervisor
)
c. Unit Operator !
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d. Shia Technical Adviser
ANSWER: c.
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KA: ' G113 2.0/2.9 - LEVEL: MEMORY
REFERENCE: FNP-0-ACP-2.0.
- LEARNING OBJECTIVE: 05230316 4-
HISTORY: New.
JUSTIFICATION:
a. Although the ED is responsible for directing emergency actions the Unit Operator (UO) or
operator at the controls (OATC) are responsible for controlling MCB horseshoe area.
b. The Shin Supervisor is in the At the Controls Area personnel and does not require permission
to enter without request, but he can not grant others access.
.
c. The UO can give permission to enter the MCB horseshoe Area..
.
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d. During these conditions,is considered At the Controls Area personnel and does not require I
permission to enter without request, but he can not grant others access.
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QUESTION No. 085:
l An oncoming Reactor Operator has worked the following schedule:
Day 1 2 3 4 :5 6 7 8 9 10. 11 12 13 14
Hours 12 8 12~- 8- 12 8 12 8 10 8 12 8 12 7 1
!
Which one of the following is the maxim _um number of hours the individual may work on day 14 i
without obtaining special approval assuming the operator has a minimum of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> off between
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each shift?
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a. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
b. 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
c. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
d. 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />
ANSWER: c.
KA: GIS 2.3/3.4 LEVEL: ANALYSIS
REFERENCE: Technical Specification 6.2.2, pg. 6-3
LEARNING OBJECTIVE: 052302R01
HISTORY: Similar question used on NRC exam given on 3/24/97 but this question has been
- modified to test a different knowledge.-
JUSTIFICATION:
a'. The individual can work 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> according to TS but this is not the maximum he/she can
work on day 14.
b. The individual can work 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> according to TS but this is not the maximum he/she can
work on day 14. !
c. This is the maximum the individual can work on day 14 without exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48
hour period.
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d. . This is the maximum an individual can work in any 72-hour period but working 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> on !
day 14 would exceed working 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.
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QUESTION No. 086: ,
Which one of the following positions can not be filled by an individual with a Unit 1 only license?
a. Operations Manager
b. Unit 2 Superintendent
c. On-Call Operations Manager
d. Shift Supervisor In Charge.
ANSWER: d.
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KA: G2.1.4 2.3/3.4 LEVEL: MEMORY
REFERENCE: FNP-0-AP-16, pg. 2 through 7.
LEARNING OBJECTIVE: 052303H01.
HISTORY: New
JUSTIFICATION:
a. The operations manager is required to hold an SRO license but it is not required to be a dual
unit license.
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b. The Unit 2 superintendent is required to hold an SRO license but it is not required to be a dual
unit license,
c. The On-Call Operations Manager is required to hold an SRO license but it is not required to
be a dual unit license.
d. - The Shift Supervisor In Charge must hold an SRO license on both units.
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. QUESTION No. 087:
You are about to use a System Operating Procedure that has the following markings:
- The applicable portion of the procedure has been changed by TCN 3C.
- In the lower right-hand corner of the page is the statement "One Time Only."
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- In the lower right.'tand corner of the page is written " Issued on 1/10/98."
Which one of the following is the latest date this TCN could be valid?
a. 1/30/98
b. 3/10/98
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c. 3/30/98
d. 4/10/98 '
ANSWER:
KA: G1213.1/3.2 LEVEL: 4EMORY
REFERENCE: FNP-0-AP-1, pg.15
LEARNING OBJECTIVE: 052303A01
. HISTORY: New
l JUSTIFICATION:
l a.' ' When a TCN is within 20 days of the end duration the responsible individual is notified.
b. A TCN shall be approved or denied within 60 days ofimplementation.
c. An outstanding TCN over 80 days old will be referred to an Assistant General Manager for
l
disposition or extension.
d. A one time only TCN will not exceed 90 days.
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a
d
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QUESTION No. 088:
I
Which one of the following conditions would allow personnel to be used instead of hold tags?
a. The use of personnelinstead is not permitted at Farley Nuclear Plant.
b. For a component inside a high radiation area if the personnel are stationed at
every possible entrance to the area.
i c. In a aituation specified by FNP-0-AP-6, PROCEDURE ADHERENCE.
d. For a component that is " inoperable" but left in service as per FNP-0-AP-13,
CONTROL OF TEMPORARY ALTERATIONS. 1
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ANSWER: c.
KA: G213 3.6/3.8 LEVEL: MEMORY
REFERENCE: FNP-0-AP-14, pg. 2
LEARNING OBJECTIVE: 052303G13
HISTORY: New
JUSTIFICATION:
a. Personnel can be used in lieu of hold tags in emergency situations in accordance with AP-14.
b. Components inside a high radiation area require tagging (but may not require verification.)
c. AP-14 states " Personnel shall not be used in lieu of hold tags except in emergency situations
in accordance with FNP-0-AP-6, PROCEDURE ADHERENCE."
d. This component status is discussed in AP-13, but is not exempt from hold tags.
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, QUESTION No. 089:
^
Which one of the following is the preferred method to reduce turbine load to less than 50% during
a rapid load reduction?
a. Operator Automatic.
b. Turbine Manual using Throttle Valve Close.
l c. Valve Position Limiter.
d.' Turbine Manual using governor valve Fast Action.
ANSWER: a.
KA: G19 2.5/4.0 LEVEL: COMPREHENSIVE
REFERENCE: FNP-1-AOP-17.0, pg. 8
LEARNING OBJECTIVE: 052520Z06
HISTORY: New
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JUSTIFICATION:
a. AOP-17,0, Action / Expected Response for a rapid load reduction requires " Reduce turbine
load at desired rate in OPERATOR AUTO."
b. Closing the turbine throttle valves in manual will reduce turbine load but this method is not
discussed in the AOP.
c. Reducing the VPL in Operator Auto will result in the go vemor valves going closed but this
method is not discussed in the AOP.
d. This is the method discussed in AOP-17.0 as a respnse not obtained if Operator Auto does
not work.
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QUESTION No. 090:
,
l Which one ofthe following describes the process, used during trouble shooting an
instrumentation channel's behavior, that visually compares the affected instrument channel
indication to other instrument channels measuring the same parameter?
a. Channel Calibration
b. ' Channel Functional Test
c. Channel Check
'
d. Channel Verification
ANSWER: c.
KA: G220 2.2/3.3 LEVEL: MEMORY
. REFERENCE: Technical Specification Definition 1.4 !
LEARNING OBJECTIVE: 052302805
HISTORY: New :
JUSTIFICATION:
a. Channel calibration is discussed in TS but refers to adjustment of channel output rather than a i
comparison.
b. Channel Functional Test is discussed in TS but refers to injecting simulated signals into a
channel.
c. The method discussed in the stem is described in TS as a Channel Check.
d. The channel functional test discusses verification of channel operability but there is no channel i
verification .
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l QUESTION No. 091:
]
With the plant at 100% power which one of the following positions must recommend, by
sienature. releasing a Work Authorization for an RCS hot leg sample containment isolation valve
before the Work Authorization may be released?
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a. Dispatcher
b. Shin Supervisor i
c. Shia Chemist i
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d. Shin Foreman Inspectmg
l ANSWER: d. !
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KA: -G2.2.19 2.1/3.1 LEVEL: MEMORY i
REFERENCE: FNP-0-AP-52, pg.14 !
LEARNING OBJECTIVE: 052303N14
!
HISTORY: New
JUSTIFICATION: j
a. Dispatcher may release work authorizations that are non operatior.al impact, this being a
CTMT isolation valve then the work has operational impact.
b. The Shin Supervisor releases work authorizations only upon the Shia Foreman Inspecting's
recommendation.
c. The Shin Chemist can release work authorizations for systems for chemical sampling or
analysis equipment that is not power plant equipment Since this valve is a CTMT isolation !
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l valve, thea it affects nuclear safety and is therefore power plant equipment.
d. Work authorizations for power plant equipment must be recommended for release by a
signature of the Shin Foreman Inspecting before the Shin Supervisor can release it.
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QUESTION No. 092: l
l' Given the following plant conditions:
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- Unit 1 is at 100% power. '
- Reactor Head Vent Valve SV 2213 A solenoid has been determined to be burned out.
- A deficiency repon (DR) has been written.
l Which oce of the following would be the Work Order Priority Code assigned to this DR7 l
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a. Priority E j
b. - Prio. ity 1 l
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I c. Prio:ity 2 l
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l d. Priority O
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ANSWER: d.
KA: G217 2.3/3.5 LEVEL: MEMORY .
REFERENCE: FNP-0-ACP-52.1, Table 2, pg.1
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LEARNING OBJECTIVE: 052303N09
HISTORY: New.
JUSTIFICATION: i
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a. Priority E is for emergency work (LCO 5 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).
b. Priority 1 is for LCO requiring power reduction in s 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
c. Priority 2 is for all other LCOs.
. d. Priority O are for items that can only be ~ked during a Unit outage.
.
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QUESTION No. 093:
Given the following plant conditions.
- Unit 1 is in a refueling shutdown.
- The Spent Fuel Crane Bridge is in operation. !
- There is a spent fuel assembly on the hoist. !
- The assembly is moving through the transfer canal. ;
Which one of the following lists the operating speeds required for the bridge and the hoist?
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a. The bridge must be moved at slow speed and the hoist may be moved at fast !
l speed. :
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b. The bridge may be moved at fast speed but the hoist must be moved at slow l
speed. I
y c. The bridge and hoist may be moved at fast speed.
d. The bridge and hoist must be moved at slow speed.
ANSWER: a.
KA: G226 2.5/3.7 LEVEL: MEMORY
REFERENCE: FNP-1-FHP-5.18, Table 1, pg.1
LEARNING OBJECTIVE: 052511B01
HISTORY: Similar to 05251IB01001 with stem and two distracters modified.
,
JUSTIFICATION:
a. The bridge is in a " zone" that requires slow speed operation. The hoist is outside " zones" that
require slow speed operation. l
b. The bridge must be moved at slow speed. The hoist can be operated at fast speed.
. c. The bridge must be moved at slow speed. There is no requirement in this " zone" to move the .
hoist at slow speed.
d. The bridge must be moved at slow speed. There is no require. ment in this " zone" to move the
hoist at slow speed.
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. QUESTION No. 094:
'
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. Given the following conditions
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- You are a fully documented Radiation Worker. )
-- Your TEDE quarterly exposure is 1.0 rem. i
- Your TEDE annual exposure is 2.5 rem. ;
'
- The dose rate from a small valve is 6 R/hr at 6 inches away.
- You are working at a distance of 4 feet from the valve. l
Which one of the following times describes the maximum time you may work at your present l
location BEFORE you must obtain an extension to your total dose equivalent exposure (TEDE)? l
\
a. ' S hours 20 minutes
b. 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 40 minutes
c. 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 00 minutes
,
. d. ' 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> 40 minutes
' ANSWER *c.
IiD 2=12D' 2
(6 R/hr)(0.5 R)(0.5 R) = I (4 R)(4 R)
I = (6 R/hr)(0.25) / (16) = 93.7 mr/hr @ 4 A
(4 R - 2.5 R) / 93.7 mr/hr = 1500 mr / 93.7 mr/hr = 16.008 hrs = 16 hrs 00 minutes
KA: G2.3.1 2.6/3.0 LEVEL: ANALYSIS
REFERENCE: FNP-0-M-001, pg. 5 i
LEARNING OBJECTIVE: G40102A03 and G40102A04
' HISTORY: A calculation of this type was used required on one of the candidates audit exams,'
but this question uses the admin. limit in the new revision to M001 not the 10 CFR 20 limits i
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JUSTIFICATION: l
a. The first approval required for a dose extension above 2000 mrem is at 3000 mrem, 500 mrem
. above where the worker is at now and correct for the old admin. limits.. i
. b. This value could be calculated if a 1000 mrem increase were used instead'of 1500 mrem. l
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c. This is the correct answer. I
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d. This value could be calculated if the 10 CFR 20 limit of 5000 mrem was used.
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QUESTION No. 095: j
Which one of the following exposure limits shall NOT BE EXCEEDED for a Planned Special -
Exposure (PSE)? !
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a. I r em total effective dose equivalent (TEDE) in any one calendar year and 5
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rem TEDE during the worker's lifetime.
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l b. 5 rem total effective dose equivalent (TEDE) in any one calendar year and 25 l
l rem TEDE during the worker's lifetime.
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c. 10 rem total effective dose equivalent (TEDE) in any one calendar year and 50 I
rem TEDE during the worker's lifetime. l
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d. 25 rem total effective dose equivalent (TEDE)in any one calendar year and
rem TEDE during the worker's lifetime.
ANSWER: b.
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KA: G2.3.2 2.5/2.9 LEVEL: ANALYSIS
REFERENCE: FNP-0-M-001, pg. 6
LEARNING OBJECTIVE: G40102A07
HISTORY: New
JUSTIFICATION: .
a. ' One rem is the Farley annual exposure guideline and 5 rem is the 10 CFR 20 annual limit.
b. These are the limits imposed for a planned special exposure.
c. . Ten rem is a limit for emergency exposure to prevent a substantial loss of property.
d. One hundred rem is a limit for emergency exposure for lifesaving operations.
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' QUESTION No. 096:
Which one of the following activities requires a Special Radiation Work Permit?
a. An entry into an Airborne Radioactivity Area.
b. An entry into a Radiological Restricted Area.
c. An entry for a Planned Special Exposure.
d. Entry into a Radiological Exclusion Area.
ANSWER: d.
KA: G2.3.7 2.0/3.3 LEVEL: MEMORY
REFERENCE: FNP-0-M-001.
LEARNING OBJECTIVE: G40103C02
HISTORY: New
JUSTIFICATION:
a. This requires tracking ofinternal exposure but does not require a special RWP.
b. This requires the area to be surveyed before entry but it does not require a special RWP.
c. A Planned Special Exposure requires the approval in writing from the executive vice
president, but it does not require a special RWP.
d. Entry into an Exclusion Area requires a special RWP.
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QUESTION No; 097: I
Given the following conditions:
L - Unit 1 is in MODE 5.
- A non-refueling outage is in progress.
l - You have entered containment and are approaching an area with the following indications:
. The area is roped offwith three horizontal radiation ropes.
. The area is conspicuously posted with 8-sided signs.
. . . There is a flashing red light activated.
i Which one of the following describes the type of area you have encountered? 1
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a. . High Radiation Area -
b. Radiological Exclusion Area
c. Airborne Reactivity Area
d. Radiography Area
- ANSWER: b.
1 KA: G2.3.10 2.9/3.3 LEVEL: MEMORY
REFERENCE: FNP-0-M-001, pg.11
LEARNING OBJECTIVE: G40103D01
HISTORY:' A similar question was used on the SNC final remedial exam to test the same KA.
The stem of this question and all distracters have been significantly modified.
JUSTIFICATION:
a. A high radiation area requires barricades and posting with 3-sided sign.
- b. A radiological exclusion area requires the posting described in the stem of the question.
c. An airborne radioactivity area does not require the posting described in the stem.
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d. A radiography area is posted with a 4 sided sign.
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/ QUESTION No. 098: I
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Given the following plant conditions.
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- An event occurred on Unit I at 0800 that resulted in a reactor trip and safety l
injection.
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- The event was initially classified at 0815 as an Alert.
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- The event classification was upgraded at 0830 to a Site Area Emergency.
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- The event classification was upgraded at 0845 to a General Emergency.- !
Which one of the following is the maximum time by which accountability must be completed?
a. 0830
b. 0845
c. 0900
d. 0915
ANSWER: c.
KA: G438 2.2/4.0 LEVEL: ANALYSIS
REFERENCE: FNP-0-EIP-10.0, pg. 7
LEARNING OBJECTIVE: 053002K05
HISTORY: New-
JUSTIFICATION:
a. This time is 30 minut,, figm the time of the event initiation. Accountability must be
completed within 30 c.anutes of.dinouncing a general evacuation which should' occur at the
declaration of the Site Area Emergency.
b. This time 30 minutes from the declaration of the Alert. Accountability must be completed I
within 30 minutes of announcing a general evacuation which should occur at the declaration
of the Site Area Emergency. l
c. This time is 30 minutes from the declaration of the Site Area Emergency and is the time i
accountability should be completed.
d. This time is 30 minutes from the declaration of the General Emergency. Accountability must
be completed within 30 minutes of announcing a general evacuation which should occur at
the declaration of the Site Area Emergency. j
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/ QUESTION No. 099: l
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l Given the following plant conditions: I
- An event has occurred on Unit 2 that resulted in a reactor trip and safety injection.
~ - Containment pressure is 30 psig. 1
- Radiation monitors R-2, R-11, and R-12 have exceeded their alarm values.
- Containment humidity is high.
- A General Emergency has been declared. 1
- Which one of the following is a protective action recommendation that should be made?
a. Recommend immediate sheltering of the general population and controlling
access in the 10 mile downwind zones.
b.' - Recommend locating and evacuating hot spots.
c. Recommend implementing control of food and water supplies pending
sampling and analysis and possible confiscation in certain areas.
d. Recommend monitoring of environmental radiation levels.
ANSWER: a.
KA: G444 2.1/4.0 LEVEL: COMPREHENSION
- REFERENCE: FNP-0-EIP-9.0, pg. 8 & 9
LEARNING OBJECTIVE: 053002J10
HISTORY: New
JUSTIFICATION:
a. This is one of the recommendations that must be made if the Emergency Classification is
based solely on plant conditions and not on dose projections. j
b. This is one of the recommendations that must be used if the Emergency Classification is based j
on dose projections.
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c. This is one of the recommendations that must be used if the Emergency Classification is based
on dose projections.
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d. This is one of the recommendations that must be used if the Emergency Classification is based
on dose projections.
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a QUESTION No.100:
l- Given the following plant conditions:
- Both Units are at 100% power.
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- A fire has been reported inside the Unit 1 RCA.
- The Fire Brigade is on the scene fighting the fire.
Which one of the following personnel is responsible for directing the Fire Brigade Chiefin fighting
the fire?
-a. Shift Supervisor Unit 1.
I b. Shift Supervisor Unit 2.
c. Shift Foreman Operating Unit 1.
d. Shift Foreman Operating Unit 2..
ANSWER: a.
- KA: G427 3.0/3.5 LEVEL: MEMORY
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REFERENCE: FNP-0-ElP-13, pg. 3
LEARNINGOBJECTIVE: 053002N02
HISTORY: New .
JUSTIFICATION:'
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a. The Unit 1 SS directs the Fire Brigade Chiefin fighting the fire.
b.- The Unit 2 SS is normally the SS in Charge, becomes the ED and does not direct specific
plant activities.
c. The Unit 1 Shift Foreman Operating is responsible for Unit 1 System Operators supervision.
Id. The Unit 2 Shift Foreman Operating is responsible for Unit 2 System Operators supervision.
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