ML20199J402
| ML20199J402 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 11/22/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20199J392 | List: |
| References | |
| 50-271-97-09, 50-271-97-9, NUDOCS 9711280138 | |
| Download: ML20199J402 (92) | |
See also: IR 05000271/1997009
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..U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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Docket No. 50 271
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- License No. . DPR 28
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Peport No.
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Licensee:' Vermont Yankee Nuclear Power Corporation ~
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Facility: Vermont Yankee Nuclear Power Station
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Loc 6 tion: Vernon, Vermont &
Exerninkticat %ri6d:. September 2 4,1997
Examiners: D. Florek, Senior Operations Engineer ,
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' S. Dennis, Examiner in Training ,
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-Approved by: G. Meyer, Chief, Operations and Human i
Performance Branch
Division of Reactor Safety
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EXAMINATION SUMMARY
Examination Report 50 271/97 009 (OL)
initial exams were administered to three senior reactor operator (SRO) upgrade applicants
and one SRO instant applicant during the period of September 2-4,1997, at the Vermont
Yankea Nuclear Power Station.
.QMBATIONS
Two of four applicants passed the exam. One SRO upgrade applicant failed the written
and operating portion of the eram and one SRO upgrade applicant failed the operating
portion of the exam. Some weak areas of understanding were identified during the written
exam. Directing shift operations related to execution and use of emergency procedures
was a significant item of weakness noted in the operatita examination.
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Reoort Details
05.1 Operator initial Exams
a. Scope
The examiners administered initial exams tn three upgrade SRO applicants and one
SRO instant applicant in accordance with NUREG 1021, " Examiner Standards,"
Interim Revision 8.
b. Observations and Findinas
The results of the initial examinations are summarized below:
PASS / Fall
Written 3/1
Operating 2/2
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Overall 2/2
The Vermont Yankee (VY) staff reviewed the written exam and assisted in the
validation of the operating exam during the week of August 18,1997. The VY
staff provided comments on the examination that significantly improved the
examination. The VY staff, who were involved with the examination review, signed
security agreements to ensure that the initial examinations were not compromised.
In a letter, dated September 10,1997 (see Attachment 2), VY provided four
comments on the written examination. The NRC partially accepted one of the four
comments. As a result, one question was deleted from the examination. The NRC
resolution of facility comments is summarized in Attachment 3.
The following summarizes the written examination questions that were missed by at
least three applicants, indicating a weakness in the understanding of the subject.
Ques 8 Knowledge of IRM response to a loss of 24vde power.
Ques 21 Knowledge of the ECCS pump and ADS logic response to a
variable line break.
Ques 50 Ability to determine individuals required to receive a whole
body count based on actual or predicted exposures.
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Ques 66 Knowledge of the purpose of the heat capacity temperature ,
limit curve ,
Ques 83 Knowledge of the impact on RCIC and EDG LNP start logic due ,
to energization of the alternate / local control panels.
During the operating test, at least two applicants performed poorly in each of the
following areas:
Refueling operations. :
Providing effective simulator crew briefings to include plans and future
expectations.
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SRO leadership and command in directing the crew when executing the
emergency operating procedures.
The above test items represent areas of weak understanding or performance and are
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provided to enable improvement of the training program.
During the development and administration of the examination, the examiners noted
several human f actors items that may have influenced the weak performance in the
simulator The work table for the EOP flowcharts was located at the thigh lovel
rather than at the waist level requiring considerable bending. The EOP flowcharts
were stored flat on a pile which required sorting through a pile of flowcharts to
select the correct flowchart. There was insufficient laydown space for the applicant
to concurrently execute the flowcharts, and some applicants had to stand a >
flowchart on the floor. Several flowcharts were labeled as "OE 3102 Alternate
Level Control Page #/Pages" which required another sort once the correct
flowchart number was identified. Operations management acknowledged the
observations and indicated that an assessment of the control room layout was
planned along with the current in-progress revision of the EOPs and that the above
observations would be considered and evaluated,
c. Conclusions
Two of the applicants passed the examination. Two SRO upgrade applicants failed
the examination. Some weak areas of understanding were identified during the
written exam. Directing shift operations related to execution and use of emergency
procedures was a weakness noted in the simulator scenarios.
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E8 Review of UFSAR Commitments
A recent discovery of a licensee operating their facility in a manner contrary to the
updated final safety analysis report (UFSAR) description highlighted the need for a
special focused review that compares plant practices, procedures, and/or
parameters to the UFSAR descriptions. While performing the examination activities
discussed in this report, the examiners reviewed portions of the UFSAR that related
to the selected examination activities, questions or topic areas.
The examiners identified a difference between the technical specification and
UFSAR regarding the reason for the main steam isolation valve closure
VY UFSAR Rev 12 Section 7.2.3.6.6 indicated that the MSIV scram
" anticipates a reactor vessellow water level scram."
VY Technical Specification Bases 2,1.G indicated that the MSIV scram
" anticipates the pressure and flux transients which occur during normal or
inadvertent isolation valve closure."
Whereas the differences noted above do not affect plant practices, procedures or
parameters, the differences warrant assessment and clarification in future revisions.
V. Manaaemtat Meetinas
X1 Exit Meeting Summary
At the conclusion of the examination, the examiners discussed their observations of the
examination proceas with members of VY management. VY acknowledged the
observations. The VY personnel present at the exit included the following:
Vermont Yankan
M. Baldm., Assistant Operations Manager
K. Bronson, Operations Manager
S. Brown, Operations Training Supervisor
D. Dalgler, Operations Instructor
L. Doan, Assistant Operations Manager
B. Finn, Training Manager
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S. Dennis, Operations Engineer ,
D. Florek, Sr. Operations Engineer; .
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- . G. Meyer, Chief Operator Licensing and Human Performance Branch '
Attachments:
1.. SRO Examination and Answer Key
2.- . Facility Comments on Written Examinations i
3. NRC Resolution of Facility Comments i
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4. Simulation Facility Report -
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, ATTACHMENT 1
SRO Examination and Answer Key
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U. S. NUCLEAR REGULATORY COMMISSION
SITE SPECIFIC EXAMINATION
SENIOR OPERATOR LICENSE
REGION 1
APPLICANTS Nt AE:
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FACILITY: Vermont Yankee
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- REACTOR TYPE
- BWR-GE3
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DATE ADMINISTERED: September 2.1997
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INSTRUCTIONS TO APPLICANT:
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Use the answer sheets provided to document your answers. Staple this cover sheet on top of
the answer sheets. Points for each question are indicated in parentheses after the question.
The passing grade requires a final grade of at least 80.00%. Examination papers will be picked
up four (4) hours after the examination starts.
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TEST VALUE APPLICANTS SCORE FINAL GRADE %
100.00
All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature
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SENIOR REACTOR OPERATOR Page 2
ANSWER SHEET I
Multiple Choice - (Circle or X your choice). If you change your answer, write your selection in !
the blank,
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MULTIPLE CHOICE. 023 a b c d ;
' 001 a b c - d _ 024 'a b c d
002 a b c d 025 a b c d
026 a b c d
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011 a b c d 034 a b c d
012 a b c d - 035 a b c d
013 a - b c d 036 a b c d
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014 a - b c - d 037 a b c d ,
015 a b c d ___ 038 a - b c d
016 a b c - d ___ 039 a b c d
017 a b c d 040 a b c d
018 a b c- d 041 abcd
019 a b c d 042 - a b c d
020 a b c d 043 a b c d
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SENIOR REACTOR OPERATOR ' rage 3
' A N S W E R- S H li E T
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- Multiple Choice (Circle or X your choice)/ if you change your answer, write your selection in
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058 a b c d 081 a b c d
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060 : a b c d 083 - a b c d
061 & b c d- 084 a b c d
062 ~ a b c d 085 a b c d
. 063 a : b - c d 086 a b c d
- 064 a b . c ~ d 087 a b . c d
065 :a b c d 088 a b c d
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ANSWER SHEET- .
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(*"*"**" END OF EXAMINATION """"")
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
During the administration of this examination the following rules apply:
1. Cheating on the examination means an automatic denial of your application and could
result in more severe penalties.
2. After the examination has been completed, you must sign the statement on the cover
Sheet indicating that the work is your own and you have not received or given assistance
in completing the examination. This must be done after you complete the examination.
3. Restroom trips are to be limited and only one applicant at a time may leave. You must
avoid all contacts with anyone outside the examination room to avoid even the
appearance or possibility of cheating.
4. Use black ink or dark pencil ONLY to facilitate legible reproductions.
5. Print your name in the blank provided in the upper right-hand corner of the examination
cover sheet and each answer sheet.
6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED
AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
7. The point value for each question is indicated in parentheses after the question.
8. If the intent of a question is unclear, ask questions of the examiner only.
9. When turning in your examination, assemble the completed examinction with
examination questions, examination aids and answer sheets. In addition, turn in all scrap
paper.
10. Ensure all information you wish to have evaluated as part of your answer is on your
answer sheet. Scrap paper will be disposed of immediately following the examination.
11. To pass the examination, you must achieve a grade of 80.00% or greater.
12. There is a time limit of four (4) hours for completion of the examination.
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13. When you are done and have turned in your examination, leave the examination area
(EXAMINER WILL DEFINE THE AREA). If you are found in this area while the
examination is still in progress, your license may be denied or revoked.
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QUESTION: 001 (1.00)
The reactor is at 100% power when alarm " PUMP A INMER SEAL LKG
HI/LO" annunciates. Concurrently, it is noted that the "A"
Reactor Recirculation pump No. 2 seal pressure is decreasing
toward zero and both seal cavity temperatures are increasing.
WHICH ONE of the following characterize the indications on the
"A" Reactor Recirculation pump shaft seal assembly?
a. Plugging of the No.1 internal orifice
b. Plugging of the No.2 internal orifice
c. Failure of the No.1 seal
d. Failure of the No.2 seal
GUESTION: 002 (1.00)
The reactor automatically scrams due to low reactor vessel
level.
WHICH ONE of the following describe how the scram pilot solenoids
and back-up scram valves initially respond to the scram? -
a. The scram pilot solenoids deenergize.
The backup scram valves deenergize.
b. The scram pilot solenoids energize.
The backup scram valves energize,
c. The scram pilot solenoids energize.
The backup scram valves deenergize.
d. The scram pilot solenoids deenergize.
The backup scram valves energize.
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OUESTION: 003-(1.00)
WHICH ONE of-the following describe the problem that could occur
if no flow exists through this reactor water cleanup filter demins
with the RWCU system in service?
a. Drywell/ Torus differential pressure could decrease due
to decreased RBCCW heat load,
b. A vacuum in the system could be drawn ultimately
resulting in resin entering the vessel on the next
system startup,
c. Non-regenerative heat exchanger outlet temperature will
increase to the isolation setpoint.
d. Water hammer would occur when the RWCU filter demins
are placed in service.
QUESTION: 004 (1.00)
vermont Yankee is at 85% power when a loss of air occurs to the
"A" feedwater reg valve.
WHICH ONE of-the following describe the affect on t al feedwater
flow and level control? Assume reactor power is not changed and
no operator action is taken.
a. Total feedwater flow will remain the same.
Neither feedwater reg valve will respond to feedwater
level control signals,
b. Total feedwater flow will remain the same.
Only the "B" feedwater reg valve will respond to
feedwater level control signals.
c. Total feedwater flow will increase.
Neither feedwater reg valve will respond to feedwater
level control signals.
d. Total feedwater flow will decrease.
Neither feedwater reg valve will respond to feedwater
level control signals.
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QUESTION: 005 (1.00)
.The reactor is at 100% power. The "A" Residual Heat Removal
(RHR) pump is in suppression pool cooling when a small break LOCA
causes a high drywell pressure sigt.al reactor scram.
WHICH ONE of the following describes the response _of the "A" RHR
loop?
a. The "A" pump trips and does not automatically restart,
b. The "A" pump trips and restarts 5 seconds later
operating on minimum flow only.
c. The "A" pump continues to run on minimum flow only.
/. . The "A" pump continues to run in the suppression pool
cooling mode.
QUESTION. 006 (1.00)
WHICH ONE of the following correctly describes the normal
alignment of the 125VDC distribution system?
a. DC-1 supplies diesel 1B normal control power.
DC-2 supplies bus DC-3.
b. DC-1 supplies diesel 1B alternate control power.
DC-2 supplies bus DC-3.
c. DC-1 supplies bus DC-3.
DC-2 supplies diesel 1B normal control power.
d. DC-1 supplies bus DC-3.
DC-2 supplies diesel 1B alternate control power.
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OUESTIOth 007 (1.00)
Operator logs indicate that the fire pump diesel engine fuel
storage tank contains 140 gallons of fuel.
WHICH ONE of the following actions are required?
.a. Establish a backup fire suppression water system within
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and submit a special report as specified in-
T.S. 6.7.C.2.
b. Restore the component to_ operable status within 7 days
or submit a report within the next 30 days to the
Commission as specified in T.S. 6.7.C.2.
c. Ensure the fuel. oil storage tank contains at least 150
gallons of fuel within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. No further action is
required.
d. Establish a continuous fire watch within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and
ensure the fuel oil storage tank contains at least 150
gallons of fuel within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
QUESTIOth 008 (1.00)
Vermont Yankee is performing a plant startup. Intermediate Range
Monitors (IRMs) are all reading on range 5 or 6.
NHICH ONE of the following describe how a loss of the -24VDC
supply to the +/- 24 VDC system would affect the operation of the
IRMs.
a. IRM indicated reactor power would be higher than actual
power,
b. IRM indicated reactor power would be lower than actual
power.
c. IRMs would continue to indicate properly as power is
increased,
d. -IRM indicators would remain as is, no change would be
observed as power is increased.
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QUESTION: 009 (1.00)
WHICH ONE of the following conditions will generate an insert
block in the Rod Worth Minimizer system?
a. When a withdraw block occurs due to a withdraw error on
the selected control rod and total steam and feed flow
are below 20%.
b. When the third insert error occurs and total steam and
feed flow indicate 28%.
c. When the low power setpoint is reached while inserting
control rods in the currently latched sequence,
d. When a withdraw block occurs due to a withdraw error
and any control rod other than the withdraw error rod
is selected.
QUESTION: 010 (1.00)
Given the following conditions:
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Reactor power increase is in progress from 30%
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Control rod 42-19 is selected for withdrawal
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An LPRM string next to the selected control rod is
WHICH CNE of the following describes the response of the Rod
Block Monitor?
a. A rod block will be initiated due to a down scale trip.
b. The " push to setup" light will be illuminated.
c. A rod block will initiated due to an inop trip.
d. The rod block monitor will be bypassed.
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QUESTION: 011 (1.00)
A plant transient occurs resulting in a successful reactor scram-
and appropriate PCIS isolations. The following conditions exist:
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Torus sprays are in service
- Torus venting using CAD is in progress
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Drywell oressure is 15 psig and steady
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Drywell temperature is 250 degrees F. and increasing
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Torus pressure is 14 psig and steady
- Torus-level is 11 feet
- Drywell and Torus H2 are both at 0.4%
WHICH ONE of the following actions is correct?
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a. Initiate d _-_.rywell' sprays ___
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b. Execute section RPV-ED
c. Restart drywell cooling
d. Secure venting the torus via. CAD
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QUESTION: 012
The "A" diesel. generator is paralleled to the grid for-the
monthly readiness demonstration and has been at full load,
2700KW, for 15 minutes.
WHICH ONE of the following would be correct if a loss of offsite
powered occurred at this time.
a. The diesel output breaker opens and recloses after
breaker 3T1 trips open. The-diesel remains running,
b. The diesel output breaker remains closed and breaker
3T1 trips open. The diesel remains running.
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c. The diEserontydt~ breaker openo and the diesel trips.
After breaker 3T1 trips open, the diesel auto starts,
d. The diesel output breaker opens and the diesel remains
running. After breaker 3T1 trips open, the diesel
output breaker must be manually closed.
QUESTION: 013 (1.00)
WHICH ONE of the following will cause the full core display red
DRIFT ligr.c to illuminate?
a. A control rod closing both an odd and even numbered
reed switch simultaneously with a rod motion signal
present.
b. A control rod closing both an odd and even numbered
reed switch simultaneously with no rod motion signal
present.
c. A control rod closing only an odd numbered reed switch
with no rod motion signal present.
d. A control rod closing only an odd numbered reed switch
with a rod motion signal present.
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QUESTION: 014 (1.00)
Prior to placing RHR in shutdown cooling, a flush from the
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condensate transfer system to-the reactor vessel is performed.
WHICH ONE of the following describe the purpose of the flush?
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a. To preheat the shutdown cooling system piping and
prevent injection of suppression pool water into the
reactor.
b. To displace / collapse-the steam / vapor voids in the
piping highpoints.
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c. To obtain.a representative sample.to analyze for RHR
service water system leakage into the RHR system.
d. To preheat the shutdown cooling piping downstream of
the RHR pump discharge valve.
QUESTION: 015 (1.00)
The refueling platform is over the fuel pool.
WHICH ONE of the following, by itself, will prevent the refueling
platform moving over the vessel,
a. Mode switch in startup
b. Mode switch in refuel
c. One control rod not full in
d, Any refueling hoist loaded
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. QUESTIONi 016 71.00)-
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- - WHICH ONE:of thetfollowing describesTh_ow a fully? inserted-TIP I
will withdraw from--theicore'when a GroupJII-primary! containment-
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- isolation _ signalL occurs?- ,
a.- 1 Probe withdraws at-slow speed-until within:2 feet-.of-
- the~ shield'then shifts to-fast.. speed until completely-. ;
withdrawn,
' b ~. Probe. withdraws at slow speed until completely
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withdrawn.
c. Probe withdraws at. fast speed?until within 2 feet of~
_ _ _thenshield then shifts to slow speid until completelyf_
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withdrawn. 7i
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.d. Probe' withdraws at fast speed until completely
withdrawn.
QUESTIONi 017. (1.00)
. A reactor scram occurs and.the scram-inlet valve for one control'
rod fails'to open.
- WHICH ONE of-the following describe the effeet of this-fallure?-
a. The control rod fails to scram and its white scram
light on the full core display does not illuminate.
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b. ' The control rod fails to-scram and its white scram
. light on the: full core display illuminates.
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c. The control $Cocrams
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and its white scram light on-the
full core display-does'not_ illuminate.
d. The control' rod scrams and cits . white scram light.- on. the
full' core display. illuminates.
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QUESTION: 018 (1,00)
WHICH ONE of the following describe how a displaced jet pump
-riser will affect core plate delta p and recirculation flow in
the loop containing the displaced jet pump riser?
a. Sudden decrease in core plate delta p.
Sudden increase in recirc drive flow,
b. Sudden decrease in-core plate delta p.
Sudden decrease in recire drive flow,
c. Sudden increase in core plate delta p.
Sudden increase in recire drive flow.
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dnStidden~ increase in ~cbrd~ plate delt~a~ p.
Sudden decrease in recire drive flow.
QUESTION: 019 (1,00)
The plant is at 100% power. While test closing an inboard MSIV, a
loss of vital AC power occurs.
WHICH ONE of the following describe the response of the MSIVs?
a. The selected MSIV will continue to test close. All
other inboard MSIVs remain open.
b. The selected MSIV will continue to test close. All
other inboard MSIVs will close.
c. The selected MSIV will reopen. All other inboard tiSIVs
remain open.
d. All inboard MSIVs fast close.
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QUESTION: 020 (1.00)
The following plant conditions exist:
- Reactor power at~90%
- Power increase in progress with recirc pumps
WHICH ONE of the following describe how the loss of 120 Volts AC
to the r? circ mg set voltage regulator will affect recirc pump ~
speed, if at all, and why?
a. speed decreases due to a decrease in exciter field
-current.
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b. speed _ decreases.due_to_a. decrease _in_ exciter output
voltage,
c. speed decreases due to a decrease in generator output
voltage,
d. no affect because 120 Volts AC is only used during mg
set startup sequence.
QUESTION: 021 (1.00)
The plant is at 100% reactor pouer. A variable leg break occurs-
on ECCS reactor vessel level instruments LT-2-3-72A and
LT-2-3-72C.
WHICH ONE of the following describe how the ECCS pumps and ADS
low-low vessel level logic relays will respond? Assume no
operator action is taken.
a. All ECCS pumps start.
ADS low-low vessel level logic relays energize,
b. All ECCS pumps start.
ADS is not affected.
c. No ECCS pumps start.
ADS low-low vessel level logic relays energize.
d. No ECCS pumps start.
ADS is not affected.
.
.
17
- QUESTION:- 022 (1.00)-
RCIC auto initiated and was injecting at rated flow into the
reactor vessel. Subsequently, the RCIC minimum flow valve, RCIC-
27, inadvertently opened and went full open.
WHICH ONE_of.the following describe the change in RCIC speed and
flow after the transient has stabilized?
a. Speed increases to attempt to maintain the 400 gpm
flow setpoint.
h. Speed decreases to attempt to maintain the 400 apm
flow setpoint.
-
-_ -
c. Speed remains the same and flow decreases.
d. Speed remnins the same and flow increases.
QUESTION: 023 (1.00)
A reactor starcup is in progress. The following data for SRM A
was obtained with no control rod motion.
TIME SRM A (counts per second)
10:10 750
10:11 960
10:12 1350
10:13 1920
10:14 2700
10:15 3840
The reactor was declared critical at 10:11.
WHICH ONE of the following time period bands contain the
calculated reactor period?
a. 85-90 seconds
b. 170-175 seconds
c. 260-265 seconds
d. 345-350 seconds
.
.
18
OUESTION: 024 (1.00)
When initiating SLC, the control switch on CRP 9-5 was taken to
"SYS 1" and the 'A" SLC pump started as expected. The control
switch was then taken to "SYS 2" but the "B" pump failed to start
due to a motor thermal overload failure.
WHICH ONE of the following describes the response of the squib
valves and RWCU-system isolation valves?
a. Only the "A' squib valve fired.
The inboard and outboard RWCU isolation valves closed,
b. Only the "A' squib valv( fired.
_ . _ _ _
Gnly the inboard RMCU isolation valve closed,
c. Both squib valves fired.
Only the inboard RWCU isolation valve closed,
d. Both squib valves fired.
The inboard and outboard RWCU isolation valves closed.
QUESTION: 025 (1.00)
Due to degraded plant conditions HPCI initiated and was injecting
to the-reactor vessel. While injecting, HPCI isolated on low
steam supply pressure due to a spurious trip signal from the
steam supply pressure transmitter. I&C was immediately notified
to reset the trip.
WHICH ONE of the following describe the action (s), if any, that
must be taken to allow HPCI to restart? Assume reactor pressure
is 900 psig, the trip is reset, and the initiation signal is
still present,
a. None,
b. Only depress the isolation reset pushbutton on CRP 9-3.
c. Only depress the auto initiation reset pushbutton on
CRP 9-3.
d. Depress the isolation reset followed by the auto
initiation reset pushbuttons on CRP 9-3.
i
. - -._ _ . _ _ _ . -
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LOUESTlON:(026 (1.00) -
- Given'thelfollowing~. conditions::
-
-Drywell-pressure 3,0 psig_.
- Reactor water level 80 inches
, 7
Both parameters have been at those values for 3.5 minutes. -You?
then place all the low pressure ECCS pumps in pullHto_-lock.
.
WHICH ONE of the-following describes ADS response?.-
.
'
a.- l ADS-blowdown' continues.
7 .~ AD57 blowdown-is terminated. ~
c. ' ADS-blowdown is terminated but will' resume when the 120
-second timer' times out.
d .~ / ADS blowdown-is terminated but-will resume when the 8
minute timer-times,out.
.
~ QUESTION: .'027 (1.00)-
E
'
The following: plant conditions exist:-
-- reactor--power is at 55%
- "A" and "B" reactor feedwater pumps are in service-
- "C" reactor feedwater pump is in-AUTO
WHICH ONE of the following describe the status of the reactor l
feedwater.9 umps if a loss of. power occurs to 4-KV Bus.#2? Assume _ !
no operator action-and_feedpump suction pressure is > 150 psig. _
l
a. "A" running, -
"B"'not running, "C"cnot running
b. "A":not_ running,="B" not running,'"C" running
c. "A" not running,
- "B" running, "C" running
.-d.-- "A" running,- "B" running, "C":not running
!
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'OUESTIONj . 028 (1' 00)
.
'
.
.
!
L Both . reactor; building railroarc doorsi are- open.
p WHICH ONE 'of the ;following would constitute- a violation of tecliL 5
ispecs?.
'
-
t
~
La.- -Insequence' critical? testing.-.- -
l
3
b. ..
-
New fuelsbeing inspected and< channeled.- -
'
- . ' _ . LPPM atring being--' replaced..
Reactor coolant temperature of:195/ degrees F. with the,
~
'd. i ,
' main-steam line bypasses and drains ~open'to the ,!
>
-
c o n d e n s e r . .. -
,
+
.
.
\
, 1 QUESTION .029 ,(1.00)-
WHICH10NE of'the following-describes the power supplies to the
-
>
- APRMs-immediately following.a station' blackout (no offsite AC,-no
. _
-EDGs available)?
-a. Channels "A", "
C", and "E" are'. battery powered-
Channels "B", "
D", and "F" lost all power
E b. Channels "A", "
C", and "E" lost all power
Channels "B", " D", and "F" are battery powered
c. Channels "A", "
C", and "E" are battery powered
Channels "B", " D",-and "F" are battery powered .
~
-
' d' .-
~ Channels "A", " C", and "E" lost-all power ,
Channels'"B", " D", and "F" lost all power-
g
-
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.
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.
.
21
QUESTION: .030 (1.00)
-
The reactor is at 90% power. . The BOP operator inadvertently
-takes the control switch for the bypass valve opening jack to
RAISE and holds it until the raise limit is reached.
WHICH ONE of the following describe the turbine control and
bypass-valve response?
a. Control valves throttle close to raise reactor
pressure, and bypass valves remain closed,
b. Control valves open to the Speed / Load changer setpoint
and then che bypass valves start to open.
c. Bypass valves open and-the control valves throttle
closed to maintain reactor pressure,
d. Control valves throttle open to lower reactor pressure,
and bypass valves remain closed.
QUESTION: 031 (1.00)
WHICH ONE cf the following describe the operational concern if a:
SRV bellows leakage alarm is received?
a. SRV may not automatically open at the proper setpoint
on high reactor pressure,
b. False indications of an SRV open on the acoustic
instrumentation,
c. SRV will open if bellows pressure drops below
10 psig.
d. Leakage directly into the torus air space if the SRV
opens.
.
, ., .. . .- . . . ~ . - ~ - . . . . . - .. .- ..-.. - .
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-
- . QUESTIONiz 032 = (1.00):
- The
- -:following1 plant; conditions-exist . -
.
'
- .- ~ Reactor /startup-is in progrees
-
.
'
Reactor;' power-is 10% '
_
~
~ .
.
_.. .
.
.
" '
-
'"B" StandbyjGas: Treatment syetem,is;inop due'to n.
sfailed damper and the LCO wa9_ entered 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />s-ago. -
- Repair. time:-isTexpected in d-hours.
The;"A"iEDGl capability.-test was completed at the endlof;th's- d
previous!: shift. As you review the_ paperwork you determine that
theJEDG must be--deplared inoperable due to unsatisfactory
'
acceptance-criteria.- ,
-:
WHICH'ONE of: the following actions-are required'per: technical
' '
~
i specificatiens?!
a. The-reactor'must be in Hot Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. -Startup may continue provided-the "A"'EDG is operable + -
within the following 7 days. >
c. :The reactor must'be:in cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d. :The reactor must be'in-cold Shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />..
.
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.,
23
QUESTION: 033 (1.00)
The following plant conditions exist at 1300 on 9/2/97
-
A LOCA and a_-loss of normal power (LNP) have occurred
-
"A" EDG failed to start
-
Drywell pressure is 3 psig and rising
-
Reactor pressure is 350 psig and decreasing
Two miwites later, BUS #4 is energized from the Vernon Tie.
-
Drywell pressure is 3.5 psig and rising
-
Reactor pressure is 250 psig and decreasing
WiiICH ONE of the following describe the "A" core _ spray pump and-
injection valve response when BUS #4 is reenergized?
a. pump starts immediately
injection valve opens immediately
b. pump starts immediately
injection valve opens in 10 seconds
c. pump starts in 10 seconds
injection valve opens immediately-
d. pump starts in 10 seconds
injection valve opens in 10 seconds
T
-.
.
24
-.
QUESTION: 034 (1.00)
WHICH ONE of the following conditions describes a' situation in
which drywell to suppression chamber delta p would exceed a
limiting condition for operation? Assume reactor power is/was
90% unless otherwise stated.
a. At 0200 on 9/2/97 delta p was reduced to < 1.7 psi
for operability testing of the HPCI
system pump.
At 0500 on 9/2/97 delta p was restored to > 1.7 pai.
b. At 1700 on 8/11/97 delta p was reduced to <1.7 psi in
anticipation of commencing a cold
shutdown at 0600 on 8/12/97. I
c. At 0700 on 8/27/97 operating temperature and pressure
were reached during a startup.
At 0500 on 8/28/97 delta p was >1.7 psi,
d. At 0300 on 8/5/97 delta p was reduced to < 1.7 psi
for testing of the inboard RCIC
steam isolation valve.
At 04GO.on 8/5/97 delta p was restored to > 1.7 psi.
QUESTION: 035 (1.00)
The plant is operating at 100% reactor power and all equipment is
in its normal alignment. A loss of voltage occurs on the
Instrument AC Bus.
WHICH ONE of the following actions will occur?
a. PCIS Group 1 inboard logic will be deenergized and the
MSIVs will close.
b. PCIS Group 2 outboard logic will be deenergized causing
valves LRW-83 and LRW-95 to close.
c. PCIS Group 3 inboard logic will be deenergized causing
the "B" SBGT train to start.
d. PCIS Group 4 outboard logic will be deenergized but no
actions will occur.
- - -. _. - . . _ . . . _ _ . _ _ . . _ _ . .
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OUESTION: 036-(1,00)1
.
_
> Given;theifollowing conditions:. l ,
A: breaker on!4KV. Bus 311st-closed bntithe:Lred; supervisory ;. ~
!
- 1ight iscNOT illuminated. ;
.
The' red ~ supervisory:lampjis,NOT.l burned out.-
. WHICH ONE of the following-describe:the effect on the breaker?
,
a.: The breaker.-:will only. trip under fault conditions,
b.' LThe-breaker-will not trip under fault-conditions.
-TherDC control' power was swappedito alternate.
'
=
- c.
-d. ?The trip _ circuit energized,xbut the breaker remained'
closed. . ,
,
OUESTION: 037 (1.00)
.A large. break LOCA has occurred at 1200. When attempting to
throttle.the LPCI injection flow through LPCI LOOP INJECTION
- valve-.RHR-27A'immediately following the LPCI initiation signal,
-
the valve did not move.
'
-WHICH ONE of the-following-describes the earliest time and
additional conditions that must be satisfied to allow valve
movement in the closed direction?
a. At 1201 place the UPS FDR BLOCK keylock switch in
" BLOCK"
- b. .
At 1205 place.the-UPS FDR BLOCK keylock switch in
-
" BLOCK" ' ,
.c. :- At 1201Lverify the-UPS FDR~ BLOCK keylock switch is-in
" NORMAL"'
2d. -At 1205~ verify the UPS FDR BLOCK.keylock switch is in
" NORMAL":
t
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, 26_
, c QUESTION: 1038 71.00).
, -
iIrradiatedi. fuel is;in-the" reactor vesseltand*the?reabtoriisiin
-
- ~
Lcold shutdown ~. ,
WHICH ONE of the following describes the-conditions,;rifJany, whenL :
r
Lall' core-'and containment ec,oling' subsystems may-be1 inoperable?-
a. - There are no. conditions under'which al1~: core and
containment cooling systems may-be inoperable.
b. JAll= core and containment-cooling systems;may be .l '
'in' operable:as long as no work is permitted-which has_
-
the potential for draining the reactor; vessel. ,
=c.- LAll-core and containment cooling systems may.1be- '
- inoperable as long as-all control-rods are fully
-inserted and electrically disabled.
d. All core and containment cooling systems may.be
inoperable if the decay heat production is calculated ,
to be less than the conductive-heat-dissipation rate.
.
QUESTION:- 039 (1,00)
l: The:EDGrisioperating in parallel with another-source when an EDG* '
overspeed trip occurs.
4 . .
.WHICH:ONE.of the following correctly describes the relay that
~
will initially.cause the EDG output breaker to open?
4
.
a '. Bhutdown relay
.b. Reverse power relay
'c.- -Generator phase. differential
( .d . Loss of field; relay
.
c.- .2
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.-
- - QUESTION: 1040i(1.00):
-WHICH ONElof the following. describes an APRM GAF of-1.02-
~
^
-
,
-a. The - APRM channel .is indicating' a higher percent, power
than' core. thermal power. _
b.- The APRM' channel is-indicatingea higher percent power
_
'than;the= sum =of'itsKLPRM inputs..
c.; The APRM, channel is--indicating a lower percent ~ power :r
than core thermal power. ;
d. The APRM channel is. indicating a lower percent power
than the sum of;its LPRM inputs.- ,
s
OUESTION: 041-(1.00)
While the reactor is at-power and the drywell is being vented:
through SBGT,-the operating reactor building exhaust fan trips- '
and the backup fails to start.
WHICH-ONE of the-following describes the response of reactor :
-
building.-delta p and containment venting? -
- a. reactor building delta p will. increase and
- drywell venting will isolate
b. reactor building delta p will increase and drywell
venting will-not isolate
c. reactor building delta p will decrease and drywell
venting will isolate
d. reactor building' delta p will decrease and drywell
venting will not isolate
.
',-
.
_..l- _,_m .L,.. . . , , . - - . ,. , . - , , -
..
.
28
OUESTION: 042 (1.00)
A control room trend recorder for a parameter that is required to
be continuously monitored has become inoperable.
WHICH_ONE of the following describes operator actions in response
to the inoperable trend recorder?
a. Attach a yellow sticker to the recorder.
Log the parameter every 15 minutes,
b. Attach a yellow sticker to the reccrder.
Log the parameter every 60 minutes,
cg. Attach a caution tag to the recorder.
Log the parameter every 15 minutes.
Af. Allaen a caution tag to the recorder.
-
LGQ the perua.0Las 6V4sy 60 F.inutc;.
QUESTION: 043 (1.00)
When taking actions in the EOP's, WHICH ONE of the following,. at
all times, requires reference to operating procedures before,
during, or after its performance?
a. resetting an EDG lockout
b. placing torus cooling in service
c. resetting main generator lockouts
d. operating the main turbine bypass jack
.
.
29
OUESTION: 044 (1.00)
A valve lineup verification requires access to an area with dose
ratas of 100 mrem /hr.
WHAT is the_ maximum time permitted by AP0155 * Current System
Valve and Breaker Lineup and Identification," to perform the
verification without reliance on transferring data from the
previous lineup in the current system lineup book?
a. 6 minutes
L. 12 niiiiutes
c. 20 minutes
,
d. 30 minutes
QUESTION: 045 (1 00)
WHICH ONE of the following individuals is responsible for
determining that pump operability requirements are met when the
determination is based on test data obtained to satisfy IST -
requirements?
a. Operations Engineering Analyst
b. Operations / Maintenance personnel performing the test
c. Senior Operations Engineer
d. -shift Engineer
._ . _ _ - . .._.._.-._.s. , . . _ . _ _ . - . . _- - . _ . ._ . .- _ . _
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-
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.
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,
. QUESTION:- v46 ;(1.00);
"
- 1
[Given[the;follbwing" conditions::
- -Stacknflowlis'5000Tscfm.. =-
-- The reactor.~scrammedi10_ hours ago.-
. .;
- Stack high-range isireading:20,0001mr/hr;; ;
-. . The windi direction is ' fron. -190 degrees. <
- 1- -The wind; speed-is 5: mph'.--
_
-
t
.
'
WHICH ONE-of'the following bands contains the: predicted dosi rate
'et::theEsite boutidery?
^
'
-
_ _
a, 5-to110"mr/hr ,
.
-#-* b. }10(to 50 mr/hr I
r- c. Si,0-to i:0 r.r/hr ?
-d. >100 to'500 mr/hr-
-
_
-?
L'
~ QUESTION: 047 (1.00)
7
- '
- A1 job?to1be = performed while the plant is shutdown requires a Hot . -
-
Work-Control Permit.t _-The job is expected to be complete over two
,
days.
.WHICH'ONE of the following describes the maximum time limit tht
a=.. Hot Work Control ~ Permit can--be. issued and active for this job?
d
a. for-duration of the job
b. -24 hours '
E c. .'two--operating shifts-
d. one-operating = shift
.
4
1
_
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-_. , - _. . . , , . _ _ _ . . _ , ,_- ,
- ._. ._. . _ __ __ _ _ _ _ . _ _ - . . ._ ._
.
31
,
QUESTION: 048 (1.00)
A White Tag from a valve has been cleared and its present
position is closed. The governing procedure appendix requires
the valve to be open.
WHICH ONE of the follcwing conditions, if any, permit the valve
to be in the closed position?
a. None,
b. If its current position agrees with a tagging order to
be hung on the r. xt shif t.
'
c. If the current position is documented in the control
room log.
d. If the current condition is documented in the lineup
deviation book.
QUESTION: 049 (1.00)
WHICH ONE of the following approaches to performing a job should
be used based on ALARA considerations?
a. One individual performing the job in a 60 Mrem /hr field
for 60 minutes.
b. One individual installing temporary shielding in a
60 Mrem /hr field for 30 minutes and then performing the
job in a 6 Mrem /hr field for 60 mirutes.
c. Two individuals performi.g the job in a 60 Mrem /hr
field for 35 minutes. ,
-d. Two individuals installing temporary shielding in a
60 Mrem /hr field for 15 minutes and then these
indiv33uals performing the job in a 6 Mram/hr field for
35 minuces.
. .-. .
-. .- . ._ -. ._ .
._. . _ . - _-
.
32
OUESTICRh 050 (1.00)
WHICH ONE of the following individuals is required to recalve a
whole body count based on the VY Internal Exposure Monitoring
Program?
a. A rad worker receives 4 mrem / day CEDE for 4 consecutive
days.
5. A rad worker receives 1 mrem / day CEDE for 30
consecutive days.
c. A declared pregnant woman is predicted to receive 50
mrem TEDE for the duration of her pregnancy.
d. A rad worker is predicted to receive an intake of 3 DAC
hours during a refueling outage.
QUESTICRh 051 (1.00)
An individual performing a continuous fire watch due to hot work
in the area requires a 15 minute break.
WHICH ONE of the following describes fire watch coverage during
the break?
a. Notify the roving fire watch patrol to pass by the area
at least once during the 15 minutes.
b. A formal relisf with another qualified fire watch must
be performed prior to the break,
c. Any individual performing work in the area can support
the fire watch requirements during the 15 minutes,
d. The fire watch can be suspended for up to 15 minutes if
hot work in the area is also suspended.
J
-_ . .. .-
.
.
33
QUESTION: 052 (1.00)-
As the refuel floor eperator, you are removing a fuel channel.
The channel fastener has been removed and you are d.n the process
of removing the channel. While lowering the prep machine, the
channel holder tool indicator is indicating a red color.
WilICH ONE of the following describes the significance of the red
color on the indicator?
a. Excessive breakaway load has been reached,
b. Excessive peak load is being approached.
c. The channel is moving freely and easily,
d. Separation of the channel and fuel assembly has
occurred.
QUESTION: 053 (1.00)
WHICH ONE of the following describen the maximum amount of
radioactive material that may be stored in an outside undiked
tank and when a representative sample of the tank's contents must
be analyzed after the material is added?
a. 1-0 curies and i day
b. 1.0 curies and 1 week
c. 10 curies and 1 day
d. 10 curies and 1 week
___ , . _ - - --
. _
- _
t
34
OUESTIC N: 054 (1.00)
!
WHICH ONE of the following is correct regarding the Minimum Zero-
Injection RPV Water Level?
.
a. Precludes clad temperature in the uncovered portion of
the core from exceeding 1800 degrees F. without
injection.
b. Precludes clad temperature in the uncovered portion of
the core from exceeding 1500 degrees F. without
injection.
c. Precludes clad temperature in the uncovered portion of
the core from exceeding 1800 degrees F. with injection.
d. Precludes clad temperature in the uncovered portion of
the core from exceeding 1500 degrees F. with injection.
QUESTIC N: 055 (1.00)
An Unusual Event (UE) has been declared which was not immediately
rectified.
When the UE conditions no longer exist, WHICH ONE.of the
following is the lowest level emergency plan position-that can
authorize termination of the UE?
a. Shift Supervisor / Plant Emergency Director
b. TSC Coordinator
c. OSC Coordinator
d. Site Recovery Manager
, -
- --
- _ _ . . _ __ _
I
.
r
35
r
OUESTION: 056 (1.00)
A minor change to a previously approved minor modificati mn (MM)
is required.
WHICH ONE of the following describe approvals required and the '
, need for a safety evaluation?
a. Safety evaluation is not required.
Requires the approval of only the shift supervisor.
b. Perform a safety evaluation. [
Requires the approval of only the shift supervisor.
c. Safety evaluation is not required. :
Requires the approval of the shift supervisor and
implementing department head.
d. Perform a safety evaluation.
= Requires the approval of the shift supervisor and
implementing department head.
QUESTION: 057 (1.00)
During refueling, primary containment is to be purged through the
SBGT.
WHICH ONE of the following describe the tech spec sampling
requirements?
a. No grab sample is required.
b. A grab sample is required prior to purging.
c. A grab sample is required during purging,
d. A grab sample is required prior to and during purging,
t
.-s, ,. - -.. .
- - - . - - _ _
.
.
36
OUESTION: 058 (1.00)
.The plant is at 65% power when a loss of stator water cooling
Occurs.
WHICH ONE of the following describes the automatic response of
reactor power and the turbine bypass valves? Assume no operator
actions are taken.
a. Reactor power increases.
Bypass valves open.
b h. Reactor power increases.
Bypass valves remain closed.
ch. Reactor power remains constant.
Bypass valves remain closed.
Jh. Reactor power remains constant.
Bypass valves open.
QUESTION: 059 (1.00)
'
The plant is operating at 100% reactor power when the "A" reactor
recire pump trips. The operator closes the "4" recirc pump
discharge valve.
Ilow is core flow determined?
a. Directly from Loop "B" flow indication on CRP 9-4.
b. By subtracting Loop "A" flow on CRP 9-4 from Total Core
flow on CRP 9-5.
c. By adding Loop "A" flow on CRP 9-4 and Total Core flow
on CRP 9-5.
d. Directly from Total Core Flow recorder on CRP 9-5.
- .- - - _ . - ,- _
_-
.
.
37
OVESTION: 060 (1,00)
The plant is operating at 28% reactor power, all systems
operable, when a loss of voltage occurs on 4KV buses #1,2,3 and
4. The reactor scrams.
WH10H ONE of the following caused the reactor scram?
a. High reactor pressure
b. Turbine control valve fast closure
Low RPV water level
d. Loss of power to RPS
QUESTION: 061 (1.00)
Given the following conditions:
-
Reactor power is 100%
-
Plant has been operating continually for 125 days
-
All systems are operable
-
No LCOs are in effect
WHICH ONE of the following describes the effects on the MSIVs and
SBGT of a continuing decrease in instrument air header pressure?
Assume reactor water level during the transient is maintained at
greater than 140 inches,
a. Outboard MSIVs close
SBGT valves line up for initiation
b, Inboard MSIVs close
SBGT valves line up for initiation
c. Outboard MSIVs close
SBGT initiates
d. Inboard MSIVs close
SBGT initiates
- - . - - - .
4
38
OUESTION: 062 (1.00)
WHICH ONE of the following conditions requires RPV-ED?
Assume a primary system is discharging into the areas listed. ,
a. NE corner room - 232' area temperature is 195 deg.F.
NE corner room - 213' area temperature is 195 deg.F.
b. TIP room radiation level is 1_ rem /hr.
Torus catwalk radiation level is 100 mr/hr.
c. Torus room SW area temperature is 280 degrees F.
Torus room NW area temperature is 270 degrees F.
d. SE corner room - 213' area temperature is 190 deg.F.
NE corner room - 232' area temperature is 192 deg.F.
QUESTION: 063 (1.00)
A loss of all RPV level indication due to high drywell
temperature has occurred. The reactor was successfully scrammed
at 1020. The following conditions have existed since 1120.
- 3 SRVs manually opened
- RPV pressure - steady at 120 psig
-
Torus water level - 12.7 feet
-
Torus pressure - 10 psig and stable
-
DW pressure - 2 psig and stable
-
DW temperature - 195 degrees f. and stable
-
Core spray pump "B" injecting
-
RPV water level instrumentation is available
It is now 1150. WHICH ONE of the following actions should be
taken?
a. Continue to inject to establish RPV pressure above the
minimum alternate RPV flooding pressure,
b. Immediately terminate RPV injection for a maximum of 5
minutes or until RPV level indication is restored.
c. Continue to inject but at 1200 terminate RPV injection
for a maximum of 5 minutes or until RPV level
indication is restored.
d.- Continue to-inject until RPV indication is restored.
.- . . . .
_ _ -_.
-
.
39 .
QUESTION: 064 (1.00)
Given the fellowing plant conditions:
-
A failure to scram occurred
- Reactor power is 20%
- Torus water temperature is 112 degrees F.
- MSIVs are closed
- 2 SRVs are cycling to control reactor pressure
- Drywell pressure is 2.2 PSIG
- RPV level is 50" and slowly lowering
- Bus #1 and #2 are deenergized
WilICil ONE of the following actions is required?
a. Terminate and prevent all injection to the RPV except
b. Maintain RPV water level between -31" and 177".
c. Maintain RPV water level between TAF and 177".
d. Reopen the MSIVs to reestablish the condenser as a heat
sink.
_
,
.
40
QUESTION: 065 (1.00)
The plant is operating at 70% reactor power when the "A" outboard
MS7V fails closed.
WHICH ONE of the following describes the response of the reactor?
Assume no operator action is taken,
a. Reactor power will decrease and stabilize at a lower
power.
RPV water level will decr'sase and then return to a
normal level,
b. Reactor power will decrease and stabilize at a lower
power.
RPV water level will increase and then return to a
normal level.
c. Reactor power will increase and stabilize at a higher
power.
RPV water level will increase and then return to a
normal level,
d. Reactor power will increase and stabilize at a higher
power.
RPV water level will decrease and then return to a
normal level.
.
w% 4 - , ,
. . - _ __ - - . - _ . _ . . . _ -
.
41
OUESTION: 060 (1.00) ;
WHICH ONE the following is the purpose of the Heat Capacity
Temperature Limit curve? :
a. To prevent dynamic pressure loads from exceeding the
structural limits of the suppression pool and submerged '
suppression chamber components during an emergency
depressurization.
b.- To prevent dynamic pressure loads from exceeding the
structural limits of the suppression pool and submerged
suppression chamber components during a design basis
LOCA.
c. To assure the blowdown energy from the RPV is within
the capacity of the primary containment vent before the
Primary Containment Pressure limit is exceeded. '
d. To= assure the Primary Containment Pressure limit is not
exceeded during a design basis LOCA after the blowdown
energy is transferred from the RPV to the containment.
QUESTION: 067 (1.00)
The plant is operating at 1001 reactor power when a "CRD HYD TEMP
HI" alarm is received.
WHICH ONE of the following could have caused this condition?
a. Eroded CRD cooling water orifice
b. Excessive drive piston leakage
c. CRD flow control valve fails open
d. Leaking scram inlet valve
.. - .-.
. -. . - . -. . . _ _ _ --_- _ _ _ _
.
42
OUESTION: 008 (1.00)
A fire protection header rupture has resulted in 13 inches of
water in the Torus area and RCIC room. All appropriate systems
have been isolated and all sump pumps are operating but level
remains at 13 inches.
W11IC11 ONE of the following actions is required?
a. A scram should be initiated.
b. Recirc should be run back to minimum. .
c. A normal shutdown should ue commenced. .
d. No action is required. :
QUESTION: 069 (1.00)
Given the following plant conditions:
- Reactor pressure 500 psig
-
Reactor level 125 inches
- Torus water temperature 182 degrees F.
W111CH ONE of the following bands contain the actual 11 eat Capacity
Level Limit?
a. 7.0 to 7.4 feet
b. 7.5 to 7.9 feet
c. 8.0 to 8.4 feet
d. 8.5 to 8.9 feet
_
_ . _ _ ._ _ _ _, , . _ . . _ .
-
-_
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QUESTION: 070 (1.00)
WHICH ONE of the following thermal limits are required to
mitigate power oscillations at high power / low flow conditions?
i
a. 7PLHGR
'
b. LHGR
c. CMFLPD ,
d. MCPR
OUESTION: 071 (1.00)
A LOCA has. occurred and the following conditions exist.
- Reactor pressure is 400 psig
-
Reactor is ahutdcwn
-
Drywell pressure is 8.5 psig
-
Drywell temperature is 270 degrees F.
-
Torus temperature is 100 degrees F.
- Instrument reference leg temperatures are
<300 degrees P
WHICH ONE of the following instruments would be a reliable
reactor vessel level indication under the listed conditions?
a. LT-57A indicates 80 inches
b. LT-57B indicates 83 inches
c. LT-69A indicates 81 inches
d. LT-68B indicates 79 inches
., , _ - . . - -
. . - --, . .
- - - - _ _ . .
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OUESTION: 072 (1.00) ,
During a major transient Torus level is approaching 22.0 feet.
WHICH ONE of the following is the concern if 22.8 feet is
exceeded?
a. SRV tailpipes will be submerged.
b. Torus vent path will be uncovered,
c. Torus to drywell vacuum breaker inlets are submerged,
d. Vent header drain lines will be submerged.
>
QUESTION: 073 (1.00)
The Advanced Offgas System radiation monitor, RAN-OG-3127 trips
on a valid Hi-Hi trip signal. The dryer skid and adsorber bypass
valves (OU 14 5,00-146) are closed.
WHICH ONE of the following describes the stack isolation valve,
OG FCV-11, response to the radiation monitor trip?
a. The valve will remain open,
b. The valve will close concurrent with the trip signal.
c. The valve will close after the trip signal has been
present for 2 minutes.
d. The valve will close after the trip signal has been
present for 30 minutes.
. - ..
- _
-
__
0
45
OUESTION: 074 0.00)
Due to a transient an offsite release is in progress. A sample
analysia of the discharge as well as a projected offsite dose
calculation has been done with the following'results.
- Noble gas discharge at the site boundary will result in
a dose rate of 2 rem / year total body
-
The projected duration of the gaseous release at the
site boundary will result in a TEDE of 150 mrem -
- The projected duration of the gaseous release at the
site boundary will result in a thyroid CDE of 400 mrem
WHAT is the emergency plan classification of this event?
'
a. Unusual Event
b. Alert
c. Site Area Emergency
d. General Emergency
GUESTION: 075 0,00)
Due to an ATWS condition you are venting the control rod over
piston volume for control rod insertion.
HOW does venting the over piston volume aid in rod insertion?
a. A delta p is established across the drive piston equal
to reactor pressure,
b. A higher delta p is established to assist the CRD pump
drive pressure,
c. Pressure is equalized between the scram discharge
volume and the reactor,
d. A lower delta p is established between the drive piston
and the scram discharge volume.
. _ . . - -.
_ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
46 ,
OUESTICHh 076 (1.00) c
WHICH ONE of the following CRD mechanism components assists in
preventing drive damage during a reactor scram?
a. The collet finger to index tube notch asssembly
b. The velocity limiter
c. The piston tube buffer hole assembly
d. The guide cap
OUESTICHh 077 (1.00)
An unexpected increase in condenser backpressure occurs. Reactor
power is 90%.
WHICH ONE of the following is the required immediate operator
action for this event?
a. Reduca reactor power to maintain condenser vacuum below
4 inches HgA by inserting control rods.
b. Reduce reactor power to maintain condenser vacuum below
4 inches HgA by reducing recire flow,
c. Reduce reactor power to maintain condenser vacuum below
5 inches HgA by inserting control rods.
d. Reduce reactor power to maintain condenser vacuum below
5 inches HgA by reducing recire flow.
__ _
_
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OUESTION: 078 (1.00)
t
Given the following conditions:
-
Reactor power 100%
-
All systems operable
- All systems in normal alignment
A loss of DC-1 occurs. You notice that the 'A" Recirc MG Drive
Motor current meter is pegged high.
!
WilICll ONE of the following describes the actions you are
required to perform in the order listed?
a. De-energize and/or verify Bus #1 is de-energized.
Scram the reactor.
Manually trip the main turbine using MTS-2 pushbutton.
b. De-energize and/or verify Bus #1 is de-energized. ,
Scram the reactor.
Manually trip the main turbine using MTS-1/3
pushbutton.
c. Scram the reactor.
Manually trip the main turbine using MTS-2 pushbutton.
De-energize and/or verify Bus #1 is de-energized.
d. Scram the reactor.
Manually trip the main turbine using MTS-1/3
pushbutton.
De-energize and/or verify Bus #1 is de-energized.
.__ _
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48
OUESTKUh 079 (1 00)
Given the following plaat conditions:
- Drywell temperature is 300 degrees F. and rising
-
Drywell pressure is 44 psig and rising
-
Torus level is 23 feet and stable
-
Torus pressure is 43 psig and rising
- CAD Torus vent path is operable
-
- Reactor pressure is 60 psig and stable
- SBGT is jn service
- Reactor level is below TAF and stable
All attempts you have made per the EOPs to this point have been
unsuccessful in reducing the rise in containment pressure.
WHICH ONE of the following actions is now required to vent
-
containment?
a. Vent the drywell via the SBGT system
b. Vent the drywell via the CAD system
c. Vent the torus via the SBGT system
d. Vent the torus via the CAD system
QUESTKRh 080 (1.00)
The reactor is at 100% power. All systems are operable and in
their normal alignment.
WHICH ONE of the following would be the plant response to a down
scale failure of the steam flow summer to the feedwater level
control system? Assume no operator action is taken,
a. Vessel level control will automatically swap to single
element control and control level in the normal band,
b. Vessel level will increase until the main turbine
trips.
c. Vessel level will decrease until the reactor scrams,
d. Vessel level control will remain in 3 element control
and control level in the normal band.
..
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'
OUESTION: 081 (1.00)
The plant is at 100% reactor power. The EPR is controlling
pressure. All systems are opera'aJ e.
WHICH ONE of the following immediate operator actions are
required if an unexplaine' increase in reactor pressure occurs
and the EPR fails to respond automatically?
a. manually lower the EPR set point as necessary
b. scram the reactor
c. verify the MPR takes control
d. use the bypass valve opening jack to control p) essure
OUESTION: 092 (1.00)
Refueling is in progress. The reactor head has been removed.
The reactor cavity has been flooded. RHR pamp "B" is operating
in shutdown cocling.
WHICH ONE of the following automatic responses will occur
if Fuel Pool Level falls to it's low low setpoint? (Consider only
actions triggered directly by fuel pool level)
a. NFPC isolation valves FPC-220 and FPC-221 close,
no other responses occur.
b. NFPC isolation valves FPC-220 and FPC-221 close
causing the NFPC pumps to trip on low suction pressure.
c. Only NFPC isolation valve FPC-221 closes causing the
NFPC pumps to trip on low suction pressure,
d. Only NFPC isolation valve FPC-220 closes causing the
NFPC pumps to trip on low suction pressure.
_ _ __ _ - _ _ _ _ _ _ - _ __ _________ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . ___
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50
OUESTl0th 083 (1.00)
The control room must be abandoned due to a fire in the cable
vault. The plant must be placed in cold shutdown. ,
WilICH ONE of the following deceribes how the "A" EDG LNP start
logic and RCIC trip systems are affected when control is taken at
their respective alternate / local control panels?
a. RCIC - all automatic trips are bypassed EXCEPT
ovarspeed
"A" EDG - No effect on LNP start logic
b. RCIC - all automatic trips are bypassed EXCEPT
"A" EDG - LNP start logic is disabled c
c. RCIC - all automatic trips are bypassed INCLUDING
"A" EDG - No effect on LNP start logic
d. RCIC - all automatic trips are bypassed INCLUDING
"A" EDG - LNP start logic is disabled
QUESTK)th 084 (1.00)
It has been noted that drywell pressure and temperatures have
-
been slowly increasing over the past 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. A primary
containment atmospheric sample indicates a hydrogen concentration
of 2.2%.
WHICH ONE of the following actions are required per the EOPs?
Assume all other reactor parameters are normal and all systems
are operable.
a. Vent the drywell and the torus
b. Vent the drywell and purge the torus
c. Vent the torus and purge the drywell
d. Purge the torus and the drywell
. _ _ _ _ _ , _ . . .
_ _ - _ _ _ . __ _ . _ . - _ _ - . . _ _ _ _ _ . _ . - _ _ _
,
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51
QUESTION: 085 (1.00)
WHICH ONE of the following describe how reactor pressure will
respond when manually swapping from the EPR to the MPR? Assume
reactor power is 100%
a. Slightly decrease and remain at the lower value until
operator action is taken.
b. Slightly increase and remain at the higher value until
operator action is taken.
c. Slightly decrease and slowly return to its original
value with no operator action,
d. Slightly increase and slowly return to its original
value with no operator action.
.
QUESTION: 080 (1.00)
The plant was shutdown for refueling and has been in shutdown
cooling for the past 2 days. Reactor vessel level is currently
190 inches and reactor coo' ant temperature is 150 degrees. I&C
han performed tne cold calibratior, of vessel level instruments.
WHICH ONE of following instruments is preferred to determine
reactor vessel level under the above conditions?
a. Shroud level inoicators LI 2-3-91A/B
b. Wide range recorder LR 6-98
c. Transient recorder LR 2-0-68B
d Feedwater level indicators LI 6-94A/B
f
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OVESTK)N: 087 (1.00) ,
WHICH ONE of the following will cause a loss of shutdown cooling?
a. Reactor pressure at 150 psig
b. Drywell pressure at 2.25 psig
c. A TCIS Group VI isolation occurs
d. Reactor vessel level at 145 inches
OUESTK)N: 088 (1.00)
Due to loss of RBCCW, alternate cooling to the RHR pump coolers -
is being established.
WHICH ONE of the following describes the operational concern when
the operator opens valves ^W-36A(B), SW Loop A(B) X-ties to
Alternate Cooling?
a. RHR SW pump runout may occur,
b. RBCCW piping may be overpressurized. ,
c. Cooler RHR SW water flowing through the fuel pool
cooling heat exchangers may decrease fuel pool
temperature below its lower limit,
d. RHR SW flow will not be available to CRD.
_ _ _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ _ - _ _ ______ _____ _______ - - _ _ _ _ - - _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ - - _ - _ _ - _ _ _ _ _ _ _ - _ - _ _ _ _ _ _
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53
OUESTICHO 089 (1.00)
WHICH ONE of the following describes the operaticn of the i
torus /drywell vacuum breakers?
'
a. Valves open at approximately 0.25 pold to relieve
pressure from the drywell to the torus
b. Valves open at approximately 0.25 paid to relieve
pressure from the torus to the drywell
c. Valves open at approximately 0.5 paid to relieve
pressure from the drywell to the torus
d. Valves open at approximately 0.5 psid to relieve
pressure from the torus to the drywell
OUESTIC4h 090 (1.00)
WHICH ONE of the following describes how RPS is designed to
protect against steam line isolation transients at full reactor
power?
a. APRM neutron flux is the primary scram signal.
High reactor pressure is the backup scram signal,
b. MSIV closure is the primary scram signal.
High reactor pressure is the backup scram signal,
c. High reactor preacure is the primary scram signal.
APRM neutron flux is the backup scram signal.
d. High reactor pressure is the primary scram signal.
MSIV closure is the backup scram signal.
_
. - . .. _
_ _
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54
OUESTION: 091 (1.00)
A loss of drywell cooling results in a drywell I cessure reaching
2.6 psig.
WHICH ONE of the following describes EDG, RCIC, and RWCU
response?
a. EDGs - running and loaded
RCIC - running and injecting
RWCU - pumps tripped i
b. EDGs - running and NOT loaded
RCIC - not affected
RWCU - not affected
c. EDGs - running and loaded
RCIC - not affected
RWCU - pumps tripped
d. EDGs - running and NOT loaded
RCIC - running and injecting
RWCU - not affected ,
QUES 110N: 092 (1.00)
ARM #11, Reactor Building RWCU panel, has been alarmir.g downscale
intermittently over the last 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> which has resulted in its
being declared inoperable. All other ARMS are operable.
WHICH ONE of the following actions should be taken due to the
a. An RWP should be written to cover the affected area.
b. The control switch for the ARM should be taken to and
left in the "OFF" position.
c. An RP technician should survey the area on a twice
weekly basis,
d. The control switch for the ARM should be periodically
taken to " OPERATE" to check indication.
, . _ _ - . , ._ _
._
.
55
GUESTION: 093 (1.00)
The MSIVs have closed and the reactor has scrammed due to a loss
of condenser vacuum. Current plant conditions are as follows:
-
All rods in
-
Mode switch in shutdown
- Reactor level is 145 inches
- Reactor pressure is 900 psig
-
Drywell pressure is 1.4 psig
- Torus water temperature is 115 degrees F.
WHICH ONE of the following is the Technical specification action
required at thia time?
a. Power operation shall not be resumed until the pool
temperature is reduced below 100 degrees F.
b. Reactor pressure shall be depressurized to less than
200 psig at normal cooldown rates.
c. The reactor shall be in a cold shutdown condition
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
d. Maintain Primary Containment integrity until pool
temperature is redaced below 100 degrees F.
_- . .- _ _ . _ . . .
.
.
56
QUESTION: 094 11.00)
While reviewing control room logs prior to assuming the shift you
note that containment pressure has decreased over the previous 2
shifts. It is now 1.85 poig.
WHICH ONE of the changes in 3 variables will 1 REDUCE indicated
containment pressure?.
a. Decrease in torus water level
Decrease in barometric pressure
Decrease in RBCCW temperature
b. Decrease in torus water level
Increase in barometric pressure
Increase in RBCCW temperature
c. Decrease in torus water level
Increase in barometric pressure
Decrease in RBCCW temperature
d. Increase in torus water level
Increase in barometric pressure
,
Decrease in RBCCW temperature
OUESTION: 095 (1.00)
Given the following plant conditions:
- Reactor is shutdown, all rods in
-
RPV pressure is O psig
-
Torus level is 11.1 feet
-
Core Spray system "A" is injecting to the vessel
-
Cote Spray system "B" is injecting to the vessel
-
CRD is injecting to the vess_1
-
RPV level is -25" and decreasing
The operator is now at step J. of OE3102 page 1 of 3.
WHICH ONE of the following actions is required per EOPs?
a. Perform Primary containment flooding
b. Perform RPV-ED
c. Perform Steam Cooling
d. Line up alternate ir.jection systems
._ _. - _ ,-. - _ _ _ -. - _ _ . - .
.. .- . - . . - - ..
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57
OUESTION: 096 (1.00)
OT 3118 "Recirc pump Trip", directs that the discharge va)ve of
the tripped pump be reopened several minutes after it is closed.
WHICH ONE of the following ir. the reason for taking this action?
a. It helps allow the pump shaft to settle properly on its
thrust bearing.
b.- It restores valid jot pump flow indication.
c. It restores valid core flow indication.
d. It helps maintain the idle loop's temperatures.
QUESTION: 097 (1.00)
The reactor was operating at 90% power when one of the two
operating feedwater pumps trips. The feedwater pump in auto
fails to start and the reactor water low water level scram fails.
WHICH ONE of the statements below describes the response of the
reactor recirc pumps assuming no operator action is taken?
a. Both reactor recirc pumps immediately trip,
b. Both reactor recire pumps will trip immediately upon
reactor vessel level decreasing to 82.5 inches.
c. Both reactor recire pumps will trip 10 seconds
after reactor vessel level decreases to 82.5 inches.
d. Both reactor recire pumps will runback to minimum in 10
seconds.
___ _ _
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_ . - _ -.
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OUESTION: 098 (1.00) l
WHICH ONE of the following statements below is the basis for the
sizing of the main steam safety /reitef valves?
a. To ensure downcomer and torus dynamic loads will not
exceed design limits under niaximum pressure /flov safety
valve actuation. ,
b. To ensure containment pressuxe and temperature design
limits are not exceeded in the event of one safety
valve failure,
c. To ensure steam line piping will not exceed system
design pressure limits upon a load reject and turbine
trip at 100% reactor power. 6
d. To ensure reactor vessel dome pressure will ot exceed
design pressure limits if all the MSIVs isolated at
once.
QUESTION: 099 (1.00)
The reactor was operating at 100% power when a reactor scram
occurred. Only one half of the control rods went full in due to
an undetected high level in the scram discharge volume. Reactor
power is currently 8%.
WHICH ONE of the following methods would be most effective in
inserting control rods at this time?
a. Manually initiate ARI
b. Manually insert control rods
c. De-energize the scram solenoid
..- -- . . - - - _
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59
QUESTION: 100 (1.00)
WHICH ONE of the following will occur due to an electrical fault
on Bus-97
a. Half scram
b. Start.of "A" EDG
.
c. Group II PCIS isolation
. d. Loss of vital AC
..
END OF EXAM
-. . _ _ _ . . . __ .
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VERMONT YANKEE - SRO EXAM - ANSWER KEY - 9/2/97
QUES ANS QUES ANS- QUES lANS QUES ANS
1. A 26. B 51.~ B 76. C
2. D 27. D 52. A 77. D
3. - B 28. A 53. -- D g
78. D.
.-n.
4. B 29. A 54.- A 79. B
__
5. C- 30. B 55. B 60. - C
6. A 31. A 56. C 81. C
7. B 32. C 57. A 82. B
8. B 33. C 58. A 83. A
9. D 34, D 59. D 84. C
__
10. D- 35. C 60. D 85. A
11. A 36. B 61, A 86. B
12. B 37. B 62. C 87. A
13. C 38. B 63. C 88. 8
it . 39. B 64. A 89. D
15. A 40. C 65. D LJL _ h)Ib
16. C 41. D 66. C 91. B
17. C 42. B 67, B 92. D
18. A 43. A 38. C 93. A
I 19. C 44. B 69. C 94. C
20. D 45. D 70. D 95. B
21. C 46. B 71. B 96. D
22. A 47. A 72. C 97. B
23. B 48. D 73. D 98. D
24. D 40. B 74. C 99. B l
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25. A 50. A 75. A 100. A
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ATTACHMENT 2
Facility Comments on Written Examination
,
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VERMONT YANKEE
(Q . NUCLEAR POWER COkPORATION
185 Old Ferry Road, Brattleboro, VT- 053017002
(802) 257 5271 September 10,1997
'
TDL 97-028
Regional Administrator, Region i
ATTN: Glenn Meyer
U.S. Nuclear Regulatory Commission
475 Allendale Road
King of Prussia, Pa. 19406-1415
References: (a)- License No. DPR-28 (Docket No. 50-271)
Subject: Comments on NRC SRO Written Examination
Attachments: (1) Question 59 Comments and References
(2) Question 69 Commen:s and References
(3) Question 74 Comments and References
(4) Question 90 Comments and References
In accordance with NUREG 1021, " Operator Licensing Examination Standards for Power
Reactors," ES 402, " Administering Initial Wntten Examinations," Vermont Yankee personnel
have reviewed the SRO initial Licensed Operator wiitten examination that was administered on
September 2,1997. Attached are four (4) questions with facility comments.
If you have any further questions, please contact Mr. Scott T. Brown, Operations Training
Supervisor, in our Brattleboro office at (802) 258-4163.
Sincerely,
VERMONT YANKEE NUCLEAR POWER CORPORATION
Gregory A, Marek
Plant Manager
c: USNRC Resident inspector - VYNPS
Document Control Desk
Mr. Don Florek, USNRC Lead Examiner
_ _ _ .
'
1
VERMONT YANKEE NUCLEAR POWER CORPORATION
.
,
- A1TACHMENT l'
Exam Question 59
Comments: This question relates to core flow indications following a recire pump trip and -
closure of the isolation valve. i
There are two issues relative to this question:
(1) The wording in the question stem, "How is core flow determined?" can be
interpreted in two different ways. The question can either be interpreted:
a. "how does the circuitry determine core flow?"
b. "how does the operator determine core flow based upon available
indications?"
(2)- The lesson plan taught to this class (previous revision) [REF: LOT-00-202
Rev.14 Section IV.B.4 (page 46 of 59)] incorrectly identified the loop
flow indicators and recorders on CRP 9-4 as indicating the sum of jet
pump flows associated with the individual recite loop. The lesson plan
was revised to provide correct information (Revision 15 06/03/97);
however, this was well after the class had received systems training on the
recirculation system (01/07/97).
,
Irrespective of the classroom training, if the question is interpreted as indicated in
(1)a. above, a conclusion could be drawn that "By subtracting Loop "A" flow on
CRP 9-4 from Total Core flow on CRP 9-5" [ Answer B] is the correct answer.
If the question is interpreted as indicated in (1)b. above, "Directly from Total Core
Flow recorder on CRP 9-5" [ Answer D) is the correct answer [REF: LOT-00-202
Rev 14. Sections IV.B.4 and 5 (pages 46-47 of 59)].
,
Recommendation: Accept both B and D as correct answers.
Attachntent 1 Page 1 of 4
_ . . _ . . _ - . . _ _ _ _ _ _. _._._ _ _ . _ . _ , _ . . _ .-
,r ,
..- -
4
4
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4
-,
.
.-
QUE5710N:' 059 (1.00)
"
The-plant.i's operating at.100% reactor. power'when_the *A" reactor-
recire pump
recire pump trips. The' operator. closes.the "A"
-dischurge. valve.
HOW is core. flow. determined?
a. ..Directly from Loop *B" -
flow indication on CRPJ9-4.
b. By-subtracting Loop
'
"A" -
flow on CRP 9-4 from Total Core
flow on CRP 9-5.
c. By adding Loop "A" -
flow on CRP 9-4 and Total Core flow
on CRP-9-5.
-
d. Directly from Total' Core Flow recorder on CRP 9-5.
,
i
?
4
Attachment 1 Page 2 of 4
.-. , . . . _ . _. _ . - _ - _ . ,. _ . - .-_ - _ _
. .. _.
.
LOT-00-202 ,
Rev.14,11/95
Page 46 of.59
OUTLINE NOTES
3. - Four jet pumps, one in each quadrant are
individually instrumented (1,6,11, and 16)
a. The individual D/P instruments are
calibrated for flow prior to pump
installation
b. Allows for calibration of all jet pump flow
indicators, using these as a reference
4. Total core flow is determined as follows with TRANSPARENCY 7
both recirc loops operating: TRANSPARENCY 7a
TRANSPARENCY 7b
a. Sums the flow from the five jet pumps in
each quadrant
- - b. Sums the flow from each quadrant
,
associated with one recirc loop, displays
'
this flow for each recirc loop on CRP 9-4
meters and CRP 9-4 recorder
c. Sums the flow signal from each recirc
loop recorder and displays this flow as
total core flow on the 9-5
5. Total core flow is determined as follows with
,ess
'
than two recirc loops operating:
a. If the recirc pump discharge valve comes Necessary since
off its open seat gr the MG se_t field reverse flow will
breaker is open, the idle loop's flow provide a d/p signal
signal is subtracted from the operating that appears to be
loop's flow signal forward flow
b. Subtracts core bypass flow to develop a Core flow indicator
total jet pump flow that represents "real" would be high
core flow
Attachment 1 Page 3 of 4
-- . . . - . - ,
' LOT-00 202
Rev.14,11/95
Page 47 of 59
,
OUTLINE NOTES
c. - When the field breaker is closed apj;( the CWD 729
discharge valve is_ fully open, bcth loops -
are once again added-
d. With both loops out of service, they are
summed as if they were operating
normally
V. SPEED CONTROL SYSTEM
'
-A. Flowpath (General) TRANSPARENCY 14
1. Recirc master controller receives a demand
signal for a change in pump speed from the
operator
2. The demand signalis processed via the
speed controller and scoop tube positioner to
the fluid coupler
3. The fluid coupler varies the MG set slip
effecting a change in generator speed
4. Recirc pump speed changes, changing core
flow and hence reactor power
5. As MG set speed changes the MG set tach.
generator provides a feedback signal to the
speed control system. It also provides a
speed signal to the voltage regulator to
,
ensure 70 volts / cycle-
. B. Component Description
1. Computation Module
a. Computes the difference between actual TRANSPARENCY 16
reactor steaming rate and desired TRANSPARENCY 15
steaming rate based on speed load is a simpiiiied version
changer setting of 16
b. Input to master controller auto circuit
Attachment 1 Page 4 of 4
.
. . _ .. . .- - - ,
. .
NERMONT YANKEE' NUCLEAR POWER CORPORATION -
,
'
,
ATTACHMENT 2 ;
- Exam Question 69
Comments: The question tests the use of the Heat Capacity Temperature Limit and Heat.
Capacity Level Limit curves by requiring the applicant to determine the actual
Heat Capacity Level Limit for a given set of degraded conditions.
As noted on the attached EOP charts (REF: OE 3104), a small change in the.
~
.
. values interpreted by: the applicant could result in a different and more
conservative Heat Capacity Level Limit, and therefore a different answer,
t
Recommendation: Accept both C and D as correct answers.
F
4
4
Attachment 2 Page 1 of 4
_ ,_, _ _
- - . _ . - -
'
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4
QUESTION:' 069 (1.00) l
_ ,
Given theLfollowing plant conditions:
'
- ' Reactor pressure-500 psig;
-
-- Reactor-level 125 inches
- Torus water temperature.182 degrees F.
WHICH ONE of the following bands contain the actual Heat Capacity
Level Limit?
a.. 7.0 to 7.4 fact
b. 7.5 to 7.9 feet
c. 8.0 to:8.4 feet
d. - 8.5 to 8.9 fegt
-
.
.
.
.
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A1TACHMENT 3
Exam Question 74
Comments:- This question tests the applicant's ability to classify an off-site radiation release _-
condition. The question intended that the Radiological Conditions section of the
Emergency Plan Classification and Action Level Scheme [REF: AP 3125 Rev.15
= Appendix A- (page l'of 2)] would be used to classify the event.
.
There are only two ways to get high radiation levels at the fence line a) an event
involving the movement of highly radioactive material outside of the reactor
building, or b) a transient (accident) within the reactor building. The stem of the
question eliminated the first possibility.
Section 14.6 of the FSAR " Analysis of Design Basis Accidents" provides tables
for exposures at the fence line for VY's Design Basis Accidents. The tables for
the DBA-LOCA (an elevated release) and the DBA-Refueling Accident (a ground
release) are attached. For the accidents the tables show the worst case dose
received at fence line to bc:
DBA-LOCA
2hr exposure from the time of the accident = 0.04mR
24hr exposure from the time of the accident = 2mR
DBA-Refueling Accident
2hr exposure from the time of the accident = 4mR
24hr exposure from the time of the accident = 22mR
.
The ' worst case life-time thyroid exposure for either of these accidents is 34mR.
The stated site boundary TEDE in the question is 7 to 75 times higher than the
Design Basis Accidents. The stated thyroid CDE at the boundary is 11 to 1400
,
times higher than that analyzed for the Design Basis Accident.
Based upon the dose levels given in the question stem, the applicant could
interpret that an off-site release resulting in the general public receiving such
levels of both whole body exposure (due to noble gasses) and thyroid dose (due
to Iodine), as indicative of a fuel clad boundary failure, reactor coolant boundary i
I
failure, and primary containment barrier failure. - As such, the applicant could
conclude, based upon the " General" category of the Fission Product Barrier Matrix
[REF: AP 3125 Rev 15 App:ndix B (page 1 of 1)], that a General Emergency
declaration is appropriate.
.
i
- Recommendation
- -- Accept both C and D as correct answers.
'
l l
Attachment 3 Page 1'of 8 l
.
- - .
45-
QUESTION:~ 074 (1.001-
Due to a ' transient an off site release is in progress. A sample
analysis of the' discharge as well as a projected offsite dose
calculation has been done with the following results.
.
Noble gas discharge at the-site boundary will result in
a dose rate of 2 rem / year total body
.
- The projected duration of the gasecus release at the
sits boundary will result in a TEDE of 150 mrem
- The projected duration of the gaseous release at the
site boundary will result in a thyroid CDE of 400 mrem
WHAT is the emergency plan' classification of this event?
a. Unusual Event
b. Alert
c. Site Area Emergency
d. General Emergency
.
Attachment 3 Page 2 of 8
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Attachment 3 Page 4 of 8
_
, + , , , . . , _w-..y , - - ,v-, .<,,_.%,w. , . . - - , - . . . . , . - _ -w .. - - - , . ,
. .y.-
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..
e:
?? mti' 14.6.8A
Loss-of-Coolant Accident - Radiolocical Effects
2-Hour Dose
Meteoroloaical conditions
Distance N-5 U-1 U-5
VS-1 MS-1 W-1
(Miles) .
Passina Cloud Whole Body Dese (Rem)
3.0E-05 5.2E-06 3.6E-05 5.8E-06
,
1/11* 2.92-05 2.9E-05 4.5E-06
'2.3E-05 2.6E-05 3 .7 E-0 6 3.1E-05
1/2 2.3 E-05 1.6E-05 2.5E-06
'
1.6E-05 1.6E-05 2.0E-05 2.8E-06
1- 1.3E-06 3.0E-07
4.5E-06 4.9E-06 2.8E-06 6.9E-07
S
7.9E-07 2.7E-07 3.2E-07 1.0E-07
10 2.2E-06 2.3E-06
. Lifetime Thyroid Dose (Rem)
,
2.8E-13 8.7E-19 1.5E-06 9.9E-08
.1/11' O. 1.4E-14 1.0E-06
2.2E-10 8.9E-07 3 .7 E-08 5.3E-06
1/2 0. 2.3E-06 5.2E-07
7 . 4E-31 1.6E-08 2.5E-06 3.9E-07
1 1.8E-07 4.7E-08
6.2E-14 5.1E-07 4.4E-07 1.2E-07
5 6.3E-08 1.7E-08
<
1.7E-10 5. 0E-07 1.ft'"C7 4.5E-08
!
10
,
1
- Site Boundary (283 insters)
Meteoroloav Wind Soeed (M/S)
VS-1 Very Stable' 1
MS-1 Moderately Stable i
Neutral 1
N-1
Neutral 5
N-5
Unstable 1
U-1
-Unstable 5
U-5
NOTE: 2.9E-05 = 2.9 x 10 5
.
Revision 13
Attachment 3 Page 5 of 8
. .
, -
- - . .=.-. . .. . . _ -
.
. ,
I&gLE 14.6.81
Loss-of-Coolant = Accident - Radioloaical Effects '
24-Hour Dese
-
Meteorolooical Conditions
Dictance .
N-5 U-1 U-5
(Miles) VS-1 MS-1 N-1
Passina Cloud Whole Body Ebse (Rem)
1.6E-03 2.9E-04 2.0E-03 3.2E-04
1/11* 1.6E-03 1.6E-03
1.4E-03 2.0E-04 1.7E-03 2.5E-04
1/2 1.3E-03 1.3E-03
8.8E-04 1.1E-03 1.5E-04 8.8E-04 1.4E-04
1 8.8E-04 1.6E-05
2.7E-04 1.6E-04 3.8E-05 7.0E-05
5 2.5E-04 5.7E-06
1.3E-04 4.3E-05 1.5E-05 1.8E-05
10 1.2E-04
Lifetime Thyroid Dose (Rem) .
1.5E-11 4.6E-17 7.8E-05 5.2E-06
1/11* 0. . 7.3E-13
4.7E-05 2.0E-06 2.8E-04 5.4E-05
1/2 0. 1.2E-08 *
1.3E-04 2.0E-05 1.2E-04 2.7E-05
1 3.9E-29 8.3E-07
2.3E-05 6.3E-06 9,5E-06 2.5E-06
5 3.2E-12 2.7 E-0 5
8.5E-06 2.4E-06 3.3E-06 8.7E-07
10 9.2E-09 2.6E-05
,
e site Boundary (283 meters)
Meteoroloav Wind Speed (M/S)
VS-1 Very Stable 1
MS-1 Moderately stable 1
N-1 Neutral 1
N-5 Neutral 5
Unstable 1
U-1
5
-
. U-5 Unstable
NOTE: 1.6E-03 = 1.6 x 10-8
Revision l'
Attachment 3 Page 6 of 8
- - . . - _.-
-
.
TABLE 14.6.11A
Refuelina Accident - Radiolooical Ef fects,
2-Hour Dose
.
Meteoro1ocical conditions
Distance U-1 U-5
(Miles) VS-1 MS.1 N-1 N-5
Passino Cloud Whole Body Dose (Rese)
3.1E-03 3.1T-03 5.5E-04 3.8E-03 6.1E-04
1/11* 3.1E-03
2.4E-03 2.4E-03 2.7E-03 3.9E-04 3.3E-03 4.7E-04
1/2 2.6E-04
1 1.7E-03 1.7E-03 2.1E-03 2.9E-04 1.7E-03 .
4.8E-04 5.2E-04 3.0E-04 7.2E-05 1.3E-04 3.1E-05
5
2.5E-04 8.3E-05 2.8E-05 3.4E-05 1.1E-05
10 2.3E-04
W etime 7avroid Dose (Rem)
1.3E-11 2.6E-10 8.1E-16 1.4E-03 9.2E-05
1/11* 0.
4.9E-03 9.7E-04
1/2 0, 2.1E-07 8.4E-04 3.5E-05
6.9E-28 1.5E-05 2.3E-03 3.6E-04 2.1E-03 4.8E-04
1
5.8E-11 4.7E-04 4.1E-04 1.1E-04 1.7 E-0 4 4.4E-05
5
4.7E-04 1.5E-04 4.2E-05 5.92-05 1.5E-05
10 1.6E-07
<
- Site Boundary (283 meters)
Meteoroloav Wind Soeed (M/S)
VS-1 ,Very Stable 1
MS.1 Moderately Stable 1
N-1 Neutral 1
N-5 Neutral 5
U-1 Unstable 1
U-5 Unstable 5
NOTE: 3.1E-03 = 3.1 x 10.s
Revision 13
Attachment 3 Page 7 of 8
-_
- - _ ._ . . _ _
_t
e
Tymts 14,g,113
>
gggpolina Accident - Radioloaical Effects >
24-Huur Dese
'
-
~Metooroloaical conditions
Dictance. U-5
(Miles) VS-1 'MS-1 N-1 N-5 U-1
'Passino Cloud Whole Body Dose (Real
1.8E-02 1.8E-02 3.2E-03 2.2E-02 3.5E-01
1/11* 1.8E-02
1.4E-02 1.6E-02 2.2E-03 1.9E-02 2.7E-03
1.4R-02
,
1/2 9.6E-03 1.5E-03
,
9.6E-03 9.6E-03 1.2E-02 1.7E-03
1 1.8E-04
,
2.7E-03 3.0E-03 ~1.7E-03 4.1E-04 7.6E-04
5
1.4E-03 4.8E-04 1.6E-04 1.9E-04 6.2E-05
10= 1.3E-03
6
feifetime Thyroid Dose (Rem)
4.9E-11 1.8E-09 5.6E-15 9.7E-03 6.4E-04
1/11* 0.
2.4E-04 3.4E-02 6.7E-03
1.5E-06 5.8E-03
1/2 0.
1.5E-02 3.4E-03
1 4,. 8 E-27 1.9E-04 1.6E-02 2.5E-03
3.3E-03 .2.9E-03 7.8E-04 1.2E-03 3.0E-04
5 4.0E-10
3.3E-03 1.1E-03 3.0E-04 4.1E-04 1.1E-04
10 1.1E ,06
- Site poundary (283 meters)
Meteoroloav Wind speed (M/s)
-vs-1 Very stable 1
MS-1 Moderately stable 1
N-1 Neutral 1
N-5 Neutral 5
U-1 Unctable 1
U-5 Unstable 5
4
NOTE:- 1.8E-02 = 1.8 x 10
- '
Revision 13
Attachment 3 Page 8 of 8
.
VCRMONT Y ANKEB NtJCLEAR POWER CORPORATION
..
ATTACHMENT 4 -
Exam Question 90
'
Commen+s: This question tests the applicant's knowledge of_ the bases for_ the reactor
protection signals that protect the reactor during-a Main Steam Line isc% tion
event. There are two issues relative to this question:
(1) Technical Specifications [REF: TS Bases 2.1.0 Amendment 84 (page 171)
states that the Main Steam Line Isolation Valve (MSIV) Closure Scran -
" anticipates the pressure and flux transients." in addition, the FSAR [REF: _
F5/." Section 14.5.1.3.1 (page 14.5-4)) specifically refers to the high
neutron flux scram as a backup / indirect means of shutting down the
reactor.
As an anticipatory signal, the MSIV Closure S: ram is thus the primary -
protection for this event. Answers A, C, and D are therefore incorrect.
(2) Technical Specifications [REF: TS Bases 1.2 and 2.2 Amendment 18
(page 19)] states that the indirect scram signal for r.n MSIV closure is -
APRM High Flux, with High Pressure as a backup to the APRM Hirh
Flux Scram. The High Pressure Scram is therefore one of the backup
rignals for an MSIV closure event.
Based upon the above listed items, Answer B is the only correct answer.
Recommendation: Change correct answer to B.
Attachment 4 Page 1 of 5 -
. . . . . - -. .. .-. . . .-._ . -
,
- ,
- QUESTION: 090 (1.00).
~
WHICH CNE of the following describes how RPS is designed to r
protect against steam-line isolation transients at. full reactor
power?
a. ' APRM neutron _ flux is the primary scram signal.
High reactor pressure is the backup scram signal.
b. MSIV closure is the primary scram signal.
High reactor pressure is the backup scram signal.
c. High reactor pressure is the primary scram signal.
APRM neutron flux is the backup scram signal.
d. High reactor pressure is the primary scram signal.
MSIV closure'is the backup scram signal.
,
.
4
.
.
,
_
Attachment 4 Page 2 of 5
.- . - . .- -
. _
,.
.. e
'
ggfJ,i 2.1 -(Cont'd)
metal-water. reaction to less than it, to assure that core geomet:y
remains intact.
The design of'the ECCS components to meet the above criteria was
the maximum break
dependent on three previously set parameters:
size, the low water level scram setpoint, and the ECCS initiation
setpoint. To lower the ECCS initiation setpoint would nowToprevent raise the
the ECCS components from meeting their design criteria. it would
ECCS iniciation setpoint would be in a safe direction, but
reduce the margin established to prevent actuation of the ECCS during
normal operation or during normally expected transients.
E. Turbine Stoo Valve Closure Scram Trio Settina
The turbine stop valve closure scram trip anticipates the pressure,
neutron flux and heat flux increase that could result from rapid-
closure of the turbide stop valves. With a scram trip setting of
<10% of valve closure from full open, the resultant increase in
surface heat flux is limited such that MCPR remains above the fuel
cladding integrity safety limit even during the worst case transient
that assumes the turbine bypass is closed. This scram is bypassed
when turbine steam flow is belov 30% of rated, as measured by turbine
first stage pressure.
F. Turbine control Valve Fast Closure Scram _
The control fast closure screa is provided to limit the rapid
in valve
-
increase pressure and neutron flux resulting from fast closure of
the turbine control valves due to a load rejection coincident with
failure of the bypass system. This transient in less severe than the
( turbine stop valve closure with failure of the bypass valves and
therefore adequate margin exists.
C. Main Steam Line Isolation Valve closure Scram
The isolation valve closure scram anticipates the pressure and flux
transients which occur during normal or inadvertent isolation valve
closure. With the scram setpoint at 10% of valve closure, there is
no increase in neutron flux.
H. Reactor Coolant Low Pressure Initiation of Main Steam Isolation valve
.
Closure _
l
The low pressure isolation of the main steam lines at 800 psig is
.provided to give protection against rapid reactor- depressurization
and the resulting rapid cooldown of the vessel. Advantage is taken
-
of the scram
valves featurttowhich
are closed, occurs
provide when the
the reactor main steam
shutdown linehigh
so that isolation
power
operation at low reactor pressure does not occur. Operation of-the
' reaccor at pressures lower than 800 psig requires that the reactor
mode switch be in the startup position where protection of the fuel
cladding integrity safety limit is provided by the IRM high neutron
flux scram.
Thus, the combination of main steam line low pressure isolation and
isolation valve closure scram assures the available of neutron fuel scra:n
protection over the entire range of applicability of the
cladding integrity safety limit.
.
17
Amendment No. M , M , 84
Attachment 4 Page 3 of 5
.
r-
. . .. .- ., . -. . .. . - - - . - .-- - .- . - - - . - - .
e a
,
,
.- 'i
.
l
'14.5.1.3.1:_ closure of All Main steam Line Isolation valves-
Se'AsME'Soiler and Pressure Vessel Code requires overpressure protection for ,
each vessel designed to meet' Code Section III. For the' plant, the transient-
I produced by the fast closure (3.0 seconds) of all main steam line isolation
valves represents the most severe abnormal operational' transient resulting.in
a ruclear system pressure rise ~when direct scrams are-ignored. The code
overpressure protection analysis hypothetically assumes the failure of the '
'
-direct' isolation valve position scram. Se reactor is shutdown by the backup, -
' indirect, high neutron flux scram. This event can be' categorized as.a core
4 dynamic event-for analysis purposes.,,
.
'
-Analysis of the event demonstrates that the installed safety valve capacity of
L 28.35% of rated flow, L in conjunction with relief capacity of 49.7% of rated
flow, limits the peak Nuclear System pressure at vessel. invert to less than
4 l 1,375 psig. The margin to the ASME Code limit assures adequate protection
i cgninst excessive enterpressurization of the Nuclear system process barrier
even for,this~ hypothetical isolation event. Table 14.5.1 lists the peak-
,
. I values of1the key process variables for this transient. Figures 14.5-5 and.
.14.5-6 graphically show the results produced by this simulated analysis. (
4
14.5.2 jyenes Pesultina in a Reactor vessel water Temperature Decrease
,.
Events that result directly in a reactor vessel water te'aperature decrease are
those that either increase the flow of cold water to the vessel or reduce the
temperature.of water being delivered to the vessel. The events that result in
the most severe transients in this category are the following: a
i 1. Loss of a Feedwater Beater
2. Shutdown Cooling (RERS) Malfunction - Decreasing Temperature
,
3. Inadvertent Pump Start
Loss of Stator cooling
~
l 4.
i
14.E42.1 Loss of a Feedwater Heater
I A feedwater heater can be lost in at least two ways: (1) if the steam
- extraction line to the heater is shut, the heat supply to the heater is-
'
l-removed,(producing a. gradual cooling of the feedwater, and (2) a bypass line-
- 'in usually provided so that the feedwater flow can be passed around rather
than through-the-heater. In either case, the reactor vessel receives cooler
- feedwater which produces an increase in core inlet subcooling. Due to the
negative void reactivity coefficient, an increase in' core power results. The
} ,
14.5-4 Revision 13
> Attachment 4 Page 4 of 5
.. _ _ - - __. _._ _ _ . . . _ - . _
.
- ___ _
o
o-
VYt1FS
(.
AblU.'
.1.2--REAc*~JR Coot. ANT SYSTEM
The reactor coolant system is an important barrier in the prevention of
uncontrolled release of fission products. It is essential that cr.e
integrity of this system be protected by establishing a pressure 1 Lait to
be observed for all operating conditions and whenever there is irradiated'
fuel in the reactor vessel.
The pressure saf ety 1 Lait of :1335 psig as measured by the vessel steaa
space pressure indicator is equivalen to 1375 psig at the lowest
elevation of the reactor coolant system, The 1375 psig value is derived
from the design pressures of the reactor pressure vessel, and the coolant
system piping. The respective design pressures _are 1250 psig at 575'F
and 1143 psig at 560*F. The pressure safety limit was chosen as the
lower of the pressure transients permitted by the applicable design
codest ASME Boiler and Pressure Vessel Code, Section III-A for the
pressure vessel, ASME Boiler and Pressure Vessel Code Section III-C for ,
the recirculation pump casing, and USASI B31.1 Code for the reactor
coolant system piping. The ASME Boiler and Pressure Vessel Code permits
pressure transients up to los over design pressure
(110% x 1250 = 1375 psig), and the USASI code permits pressuru transient:
up to 20% over the design pressure (120% x 1148 = 1378 psig).
The safety valves are sized to prevent exceeding t"s pressure "essel code
limit for the worst-case isolation (prescurization) event (MSIV closure)
assuming indire:c (neutron flux) scram.
,
2.2 REACTOR Coot. ANT SYSTD4
(. .The settings on the reactor high pressure scram, reactor coolant system
relief and safety valves, have been established to assure never reach:ng
the reactor coolant system pressure safety limit as well as assuring the
system prassure does not exceed the range of the fuel cladding integrity
safety limit. In addition to preventing power operation above 1055 psig,
the pressure scram backs up the APP.M neutron flux scram for steam line
isolation type transier.ts. (See FSAR Section 14.5 and Supplement 2 to
Proposed Change No. 14, Novamber 12, 1973.)
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Amendment 1ks. 18 19
_ Attachment 4 Pahe 5 of 5
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ATTACHMENT 3
NRC Resolution of Facility Comments
QUESTION 59 Disagree with VY comment. As stated iri LOT 00 202 Rev.14 page
46 step 5.a, "the idle loop's flow signal is subtracted from the
operating loop's flow signal." Answer "B" is incorrect in that
subtracting the LOOP "A" flow on CRP 9 4 from TOTAL CORE FLOW
on CRP 9 5 is not the same as idle loop's flow signal subtracted from
the operating loop's flow signal. LOT-00-216 Rev.13 page 19 step
4.c. states "When one recirculation loop is idle, the f!ow indication is
automatically _ subtracted from the running loop's jet pump flow."
Neither reference supports subtractir.g the idle loop flow from total
core flow as answer "B" states. Therefore, the only correct answer
is "D" There was no change to the answer key.
OucSTION 69 Disagree with VY comment. The value given for reactor pressure in
the question stem,500 psig, clearly places the associated value for
Torus Water Temperature at > 210 degrees on the HCTL curve. With
this value transposed to the HCLL curve, the only correct answer is
"C". A small change in the values interpreted sould still place you in
the HCLL band of 8.0 to 8.4 feet. There was no change to the
answer key.
QUESTION 74 Disagree with VY comment. The question stem gave no indication of
the status of the fuel clad, reactor coolant, or primary containment.
The answer is based only on the information stated in the question
stem. An SAE is the correct classification. Additionally, the
classification level of an SAE for site boundary radiological dose as
per AP3125 must have it's basis in an analysis other than the DBA
LOCA or Refueling Accident, otherwise, those boundary doses would
be listed under the GE column of AP3125. There was no change tn
the answer key.
QUESTION 90 Disagree with VY comment. A review of technical specifications and
the UFSAR showed statements tnat appear to conflict. VY UFSAR
Rev.12 Sect.7.2.3.6.6 "The scram initiated by main steam line
isolation closure anticipates a reactor vessel low water level scram."
VY T.S. Bases 2.1.G, page 17 "The isolation valve closure scram -
anticipates the pressure and flux transients which occur during normat
or inadvertent isolation valve closure."
Due to your comments and the conflicting statements listed above,
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the question will be deleted from the exam. No correct answer can
be assuredly referenced. The answer key was changed to reflect the
deletion.
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ATTACHMENT 4
SIMULATION FACILITY REPORT
Facility License: DPR 28
Facility Docket No: 50-271
Operating Test Administration: September 3-4,1997
This form is to 'oe used only to report observations. These observations do not constitute
audit or inspection findings and are not, without further verification and review, indicative
of a noncompliance with 10 CFR 55.45(b). These observations do not affect NRC
eertification or approval of the simulation facility other than to provide information that
..ey be used in future evaluations. No licent ee action is required in response to these
[' observations.
ITEM DESCRIPTION
1. During a LOCA at atmospheric pressure, HPCIindicated that it was operating
even though there was insufficient pressure to operate HPCI and HPCl was
isolated.
2. During a low power ATWS the MSlVs went closed for no apparent reason
when the simulator instructor inserted a malfunction to fail the standby liquid
control pump suction line.
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