ML20153E947
| ML20153E947 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 09/24/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20153E919 | List: |
| References | |
| 50-289-98-301, NUDOCS 9809280302 | |
| Download: ML20153E947 (110) | |
See also: IR 05000289/1998301
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U. S. NUCLEAR REGULATORY COMMISSION
REGION 1
Docket No: 50-289
Report No: '50-289/98-301
License No: DPR-50
Licensee: General Public Utilities Nuclear
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Facility: Three Mile Island Unit 1 Nuclear Station ]
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Location: Middletown, Pennsylvania !
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Dates: August 24- 27,1998
Chief Examiner: P. Bissett, Senior Operations Engineer / Examiner
Examiners: D. McNeil, Senior Operations Examiner (Rlll)
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Approved by: Richard J. Conte, Chief ;
Operator Licensing and Human Performance Branch l
Division of Reactor Safety i
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9809280302 98o924 '
hDR ADOCK 05000209
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EXECUTIVE SUMMARY
Three Mile Island Unit 1 Nuclear Station
inspection Report No. 50-289/98-301
Ooerations
Four senior reactor operator (SRO) applicants passed'all portions of the initial license
examination.
The applicants performed well on both the written and operating portions of the
examination. The applicants were well prepared for the examination, indicating that the
facility thoroughly evaluated the knowledge and ability of each candidate in an effort to
determine their readiness to sit for an initial NRC senior reactor operator examination.
, Crew communications and crew briefings during the simulator scenario portion of the
- examinations were good.
The training department did an excellent job in following the guidance set forth in the
examiner standards during the development of the examinationc. With few exceptions,
excellent attention to detail prevailed throughout the examination development process.
On the quality of the original exam submittal by the TMl training department, several
administrative questions were deemed direct lookup questions and, as a result,
administrative job performance measures (JPMs) were substituted for these questions.
These administrative JPMs were appropriately developed by the TMI training department to
test the knowledge level of the applicants in the administrative area. Overall, examination
nye!ity was good, with the majority of changes having dealt with editorial changes or
enhancements.
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Report Details
1. Operations
05 Operator Training and Qualifications
05.1 Senior Reactor Operator Initial Examinatioas
a. Scope
The examinations were prepared by Three Mile Island (TMI) Unit 1 personnel in
accordance with the guidelines in interim Revision 8, of NUREG-1021, " Operator
Licensing Examination Standards for Power Reactors." The initial operator licensing
examinations were administered to four senior reactor operator (SRO) applicants.
One of the SRO applicants was an SRO instant, whereas the other three applicants
were SRO upgrades. The NRC administered the operating portion of the
examinations, whereas the written examinations were administered by the TMI
- training organization following the completion of the operating examinations.
b. Observations and Findinas
Gradina and Results
The results of the SRO examinations are summarized below:
SRO Pass / Fail
Written 4/O
Operating 4/O
Overall 4/0
Preoarations
The written examination, job performance measures (JPMs), including follow-up
questions, and simulator scenarios were developed by TMI personnel in accordance
with NUREG-1021. All individuals involved signed a security agreement once the
development of the examination commenced. TMI personnel also validated the
examination prior to their submitting it to the NRC. During the exam preparation
week of August 10,1998, the NRC subsequently reviewed and validated, together
with TMI personnel, all portions of the proposed examinations.
For the most part, the written exam, JPMs and scenarios required only minor
changes. However, the examiners determined that several of the administrative
questions were essentially direct look-up questions After discussion with TMl
training representatives, it was agreed upon that administrative JPMs would be
developed to replace those administrative questions that were not considered
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appropriate by the examiners. TMI subsequently developed several administrative
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JPMs that met the guidelines of NUREG-1021. A couple of JPM follow-up
questions were also deemed to fall into the category of a direct lookup question.
These questions were subsequently replaced with higher cognitive thinking level
questions.
Administration and Performance
The written portion of the examine * ion was administered by TMI training personnel
on August 27,1998, and consist ( f 100 multiple choice questions. There were
no comments by either the NRC or TMI concerning the validity of questions on the
written examination, however one question was determined to have had the wrong
answer designated in both the answer key and supporting documentation. TMl
training personnel and the NRC subsequently verified the validity of the answer and
appropriately changed the answer key.
The operating portion of the examination was conducted from August 24 - 25,
1998, and consisted of thrt,e simulator scenarios for each applicant. Also, ten
JPMs were administered to the SRO instant applicant, whereas the three SRO
upgrade applicants were required to perform only five JPMs. All JPMs were
followed up with two system-related questions. As previously mentioned,
l administrative JPMs were developed and administered to all applicants to evaluate
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the administrative requirement portion of the examination.
Simulator and JPM performance by the SRO applicants was good. Communications
was also good, including the use of repeat backs. Crew briefings were held when
time permitted and was deemed necessary by the SRO. Procedural usage was
appropriate throughout the conduct of all scenarios. Good control board awareness
by all of the candidates was evident throughout each of the four scenarios observed
by the NRC examiners.
For feedback to the initial licensed operator training program, the examiners noted,
as did the facility, that questions 23,47,65, and 70 were missed by three or more
of the applicants. These questions, respectively, dealt with power level vs rod
realignment; bus power supplies to major DC loads; automatic protective actions for
the waste gas decay system; and, reactor bui! ding access authorization
requirements. Also, three of four applicants failed to correctly answer one JPM
follow-up question which called for the identification of inoperable control rods as a
result of control rod misalignment.
c. Conclusions
The four SRO applicants passed all portions of the initiallicense examination. The
applicants performed well on both the written and operating portions of the
! examination. The applicants were well prepared for the examination, indicating that
the facility thoroughly evaluated the knowledge and ability of each candidate in an
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' effort to determine their readiness to sit for an initial NRC senior reactor operator
examination. Crew communications and crew briefings during the simulator
- scenario portion of the examinations were good.
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, The training department did an excellent job in following the guid6 ace set forth in
the examiner standards during the development of the examinations. With few
. exceptions, excellent attention to detail prevailed throughout the examination
i development process.
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On the quality of the original exam submittal by the TMI training department,
- several administrative questions were deemed direct lookup questions; and, as a
result, administrative job performance measures (JPMs) were substituted for these
questions. These administrative JPMs were appropriately developed by the TMl
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j training department to test the knowledge level of the applicants in the
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administrative area. Overall, examination quality was good, with the majority of
- changes having dealt with cditorial changes or enhancements.
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- E8 Review of the FSAR
While performing the preexamination activities discussed in this report, the
inspectors reviewed applicable portions of the FSAR, that related to the selected
examination questions or topic areas. No discrepancies were noted.
V. Manaaement Meetinas
X1 Exit Meeting Summary
On August 26,1998 the NRC discussed their observations regarding the examination with
Three Mile Island Unit 1 operations and training management representatives. The
examiner discussed generic candidate performance, as observed during the administration
of the simulator scenarios and job performance measures.
The NRC also expressed their appreciation for the cooperation and assistance that was
provided during both the preparation and examination week by licensed operator training
and operations personnel.
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PARTIAL LIST OF PERSONS CONTACTED
Three Mile Island
D. Boltz, Instructor, TMI Training
R. Hess, Plant Training Manager
J. Langenbach, Vice President, Director TMI
M. Ross, Director, Operations and Maintenance
R. Parnell, Operations Training Manager
NflG
P. Bissett, Senior Operations Enginar/ Examiner
D. McNeil, Senior Operations Examiner
Attachments:
1. Three Mile Island Unit 1 SRO Written Examination w/ Answer Key
2. Simulation Facility Report
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Attachment 1
TMI-1 SRO WRITTEN EXAMINATION W/ ANSWER KEY
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l Mr. J. Langenbach - 3
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ij Distribution w/enci and Attachments 1-2:
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DRS Master Examination File
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Nuclear Safety Information Center (NSIC)
V. Curley, DRS
Distribution w/ encl: w/o Attachments 1-2:
Region 1 Docket Room (with concurrences)
NRC Resident inspector
H. Miller, RA/W. Axelson, DRA
P. Eselgroth, DRP
l N. Perry, DRP l
D. Haverkamp, DRP
C. O'Daniell, DRP
J. Wiggins, DRS
L. Nicholson, DRS ;
P. Bissett, Chief Examiner, DRS ;
l DRS File
Distribution w/ encl: w/o Attachments 1-2 (VIA E-MAIL):
B. McCabe, OEDO
C. Thomas, PD1-3, NRR
T. Colburn, PD1-3, NRR
R. Eaton,'PDI-3, NRR
R. Correia, NRR
F. Talbot, NRR '
DOCDESK
Inspection Program Branch, NRR (IPAS)
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U.S Nuclear Regulatory Commission
Site-Specific
Written Examination
Applicant Information-
Name: Region: I -
Date: August 27,1998 Facility / Unit: Three Mile Island Unit 1
License Level: SRO Reactor Type: B&W-177
Start Time: Finish Time:
I instructions
Use the answer sheets provided to document your answers. Staple this cover sheet on top
of the answer sheets. The passing grade requires a final grade of at least 80.00 percent.
Examination papers will be collected four hours after the examination starts. ,
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Applicant Certification
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All work done on this examination is my own. I have neither given nor received aid.
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- Applicant's Signature
Results
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Examination Value 100 Points
Applicant's Score Points
Applicant's Grade Percent
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- 001 (1.00)
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Selected Pressurizer level signal indicates off-scale low. An evaluation must be conducted to
determine if the transmitter has failed low or if it is accurately displaying a low pressurizer level
i condition. Which ONE (1) parameter can be used IMMEDIATELY to make this determination?
- A Pressurizer spray valve posdon
B. RCS pressure
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C. Makeup tank level '
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D.- Pressurizerwater temperature
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QUESTION: 002 (1.00) l
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Current plant conditions are: "
- The reactor is tripped. !
- RCS subcooled margin is zero.
Which ONE (1) action results in increasing RCS subcooling margin?
A. Dscrease RCS pressurizerlevel
B. Decrease RCS hot leg flow
C. Increase RCS loop pressure
D. Increase RCS hot leg temperature
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QUESTION: 003 (1.00)
Reactor power is at 100% when the controlling RCS pressure channel, RC3A-PT1, fails
instantaneously LOW. With NO operator action, which ONE (1) statement describes the
SHORT-TERM plant response?
A. Reactor trip occurs on high RCS pressure
B. SASS shdts control to RC38-PT1 RCS pressure channel to stabilize the plant
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C. Pressurizer heaters energize from RC3A-PT1 to increase RCS pressure
D. Reactor trips and Safety injection is actuated on low RCS prassure
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QUESTION: 004 (1.00)
The reactor is super critical and just entenng the intermediate range (4000 cps on the Source
Range) when detector compensating voltage to NI-3 is lost. Which ONE (1) statement explains )
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A. Ni-3 would be unaffected at this low power level.
( B. NI-3 would indicate higher than NI-4. l
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D. NI-3 would go off scale low before NI-4 if a reactor trip occurred while at 1E-8
amps.
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QUESTION: 005 (1.00)
The plant is at 100% power when control room indications reveal the following:
- Reactor power is DECREASING
- RCS pressure is INCREASING
- Main Steam safety valves are OPEN
- MS-V-3s and MS-V-4s are OPEN
- Indicating lights on Panel SS-1 are GREEN for the breakers for the Middletown 1092,
Jackson 1051 and 500 kV tie lines
. indicating lights on Panel SS-1 for the Middletown 1091 breaker switches are GREEN
add YELLOW
- Indicating lights for both main generator breakers are RED
- Main generator electrical megawatts are 56 MW
Which ONE (1) event le described by these symptoms? -
A. Loss of 230 kV substation DC
B. Load rejection
C. Auxiliary transformer fault
D. Loss of station power
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-QUESTION: 006 (1.00)
Which ONE (1) statement desenbos the requirements for an individual to be allowed to receive a
TEDE dose greater than 4000 mrem per year, excluding a planned special exposure?
A. . A special RWP is wntten covenng the individual to be permitted to exceed 4000
mrem
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B. Approval from RadCon/ Safety Director and Site Director
C. Approval from the Presdent, GPU Ntelear
D. Notification of the NRC l
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QUESTION: 007 (1.00)
Given a Switching and Tagging Request to remove a fire system heat detector in the EG-Y-1 A
diesel room from service, identify ONE (1) action is required to compensate for this detector
being removed from service. l
~A. Establish a fire watch PATROL within one hour to inspect the diesel room at least
ONCE PER HOUR. l
B. Station a CONTINUOUS fire WATCH in EG-Y-1 A diesel room WITHIN ONE
HOUR.
C. START EG-Y I A to perform the one-hour surveillance to verify OPERABILITY.
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D. Restore detector to operable status WITHIN 14 DAYS or commence plant i
shutdown to hot shutdown.
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QUESTION: 008 (1.00)
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Which ONE (1) statement describes the purpose of the protective action guidelines?
A. Protect plant workers from receiving excessive radiation exposures in excess of
10CFR20 limits
B. Prevent radioactive releases from exceeding 10CFR20 limits
C. Recommend sheltering or evacuation for the general population
- - D. Determine if potassium iodide tablets should be administered to reduce thyroid
dose .
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QUESTION: 009 (1.00)
With the plant operating at 50% power, BOTH intermediate Range NI detectors fail LOW. Which
- ONE (1) statement desenbes the required action (s).
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l A. Coininue power operatons but limit power to 50%.
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B. Continue power operations, power may be increased.
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! C. Immediately take achon to place the unit in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
D. Take adion within one hour to restore at least one Intermediate Range channel to
operable status or place the unit in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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QUESTION: 010. (1.00)
10 CFR 50.54 (x) specifically allows " reasonable action that departs from a license condition or
a technical specification in an emergency when this action is immediately needed to protect ;
the public health and safety." Select the MINIMUM position that may approve 10 CFR 50.54 i
(x) actions.
A. . Director Operations & Maintenance
B. Plant Operations Director
C. Licensed Senior Reactor Operator
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D. Licensed Reactor Operator
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QUESTION: 011 (1.00)
In addition to the person having the clearance, which ONE (1) of the following must grant
j permission to change the condition of BLUE tagged ES equipment?
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A. - Director Operations and Maintenance
i B. Plant Operations Director
,- C. Duty Shift Supervisor / Shift Foreman
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D. Licensed Control Room Operator
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, .QUESTION: 012 (1.00)
I Following a reactor trip, the followng conditions exist
- A OTSG level is 87" and decreasing slowly
- B OTSG level is 82" and decreasing slowly
- MFW flowis O gpm
MFWvalve D/P is O psig
- RCS pressure is 1725 psig and stable
- MUT levelis 62 inches
- - PZR levelis 35 inches
- MU flowis 60 gpm
Which ONE (1) statement desenbes the action required per AP 1210-1, Reactor Trip?
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A. Increase MFW pump speed
B. Open MU-V-14A or MU-V-148 as necessary. l
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C. Initiate HPl.
D. Start second Makeup Pump and open MU-V-217.
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QUESTION: 013 (1.00)
Current plant conditions are:
- Reactor is operating at 80% power.
- RM-L-1 (RC Letdown) has increased to the ALERT setpoint.
- Chemistry analysis indicates dose equivalent 1-131 concentration is 0.28 uci/gm.
l - RCS specific activity is 220 uci/gm.
- E-BAR is 0.5.
Which ONE (1) statement describes the required Tech Spec actions?
A. Reduce RCS activity to less than the Tech Spec limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot
shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. !
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l B. Reduce RCS activity to less than Tech Spec limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot
l shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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C. Initiate actions to place the unit in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. !
i D. Reduce RCS activity to less than Tech Spec limit within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or place the l
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QUESTION: 014 (1.00)
Current plant conditions are:
- Reactor is operating at 100% power.
- RCS Tavg is constant at 579'F.
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- Make up taak level is decreasing slowly - MU-V-17 is in MANUAL control.
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- Letdown flow has been constant at 45 gpm.
- RCP total seal injection flow is 38 gpm (normal)- MU-V-32 is in AUTO.
- RCP labyrinth seal D/P indicators show low off-scale (negative).
- Auriliary Building airborne activity is increasing.
Which ONE (1) statement describes the cause for the abnormal conditions?
A. RCP seal #1 leak-off flow is aligned to the Auxiliary Building sump.
B. RCP total seal injection flow transmitter has failed.
C. RCP seal injection flow is not reaching the RCPs.
D. RCP seal #1 leak-off flow has been isolated by closure of MU-V-26.
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[ - QU:ESTION: 015 (1.00)
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i. Some reactor trip situations require large volumes of makeup water for RCS inventory control
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simultaneous with the need to emergency borate the core. Which ONE (1) source should be
used for this condition?
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i A. 4% BAMT (CA-T-8)
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- B. Concentrated Waste Storage Tank (WDT-6A/B)
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l- C. RC Bleed Tank 1C (WDL-T-1C)
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- - D. BWST (DH-T-1)
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( . QUESTION: O_16 (1.00)
It is necessary to evacuate the control room due to a serious fire in the relay room Which ONE
i' (1) required action must be performed prior to exiting the control room?
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A. Perform notifications for an ALERT.
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i B. Stat EG-Y-1 A and EG-Y-18.
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j C. Trip the MFW pumps.
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l o. close MU-v-3.
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QUESTION: 017 -(1.00)
RCPs are bumped during inadequate cora heat removal condi .i.ms. Which ONE (1) statemeni
desenbos the reason for this action?
A.- Decrease RCS pressure
B. Induce OTSG heat transfer
C. Prevent RCS 'wwentory loss .
D. Increase OTSG pressure
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L QUESTION: 018 (1.00) ~
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A small break LOCA is in progress. Which ONE (1) set of cordtions requires tnpping all RCPs?
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A .579'F 1800 psig '
B. 537'F. 1300 psig
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C. 525'F 1000 psig
D. 473*F 800 psig
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QUESTION: 019 (1.00)
Which ONE (1) condition requires a Core Flood Tank to be declared inoperable?
. A. -- Boron is 2290 ppm.
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B. Pressure is 620 psig. l
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C.' Level is 14 ft.
D. Temperature is 100 degrees F.
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QUESTION: 020 (1.00)
Plant conditions requre manual reactor trip. Upon depressing the Tiip M DSS
pushbuttons, the reactor does not trip (reactor power remains at 100%). Which ONE (1)
statement desenbes the required action?
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A. Place the EHC pump control switches in P-T-L and open EHC-FV-1.
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B. Place the diamond rod control station in manual and reduce reactor power.
l C. Initiate HPl and maintain primary to secondary heat transfer until power level is less
L than 10%.
D. I
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Transfer FW to manual to control OTSG levels.
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l QUESTION: 021 (1.00)
BS-P-1B is runnog for surveillance. Which ONE (1) corstion would result in automatic trip of I
the BS-P-1B breaker?
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A. 1 A Aux Transformer fault with Auto Transfer of loads to 1 B Aux Transformer
B. 18 Aux Transformer fault with Auto Transfer of loads to 1 A Aux Transformer
C. Fault downstream of 1P 480v Bus low side feeder breaker )
D. Fault downstream of 1S 480v Bus low side feeder breaker
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QUESTION: . 022 (1.00)
Current plant conditions are:
- Pump down of RCS is in progress.
- CH-P-1 A is in service
-
- Low DH Pump Flow annunciator C-1-7 is actuated.
- RCS temperature is increasin0-
- Discharge pressure for DH-P-1 A is unstable
- Motor amperage irdcation for DH-P-1 A is unstable. .
- Noise is reported in DH-P-1 A vault. l
'
Whch ONE (1) statement desenbes the required operator actions?
A. Place Loop B Decay Heat Removal System in service, and trip DH-P-1 A.
,
B. Trip DH-P-1 A, and do NOT restart this pump until appropriate actions have been
completed.
C. Place Loop B Decay Heat Removal System in service, and leave DH-P-1 A running
until conditions stabilize.
D. Start one Makeup Pump and do NOT trip this pump until incore Thermocouple
temperatures stabilize.
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QUESTION: 023 (1.00) l
Current plant conditions are: l
- Reactoris operating at 100% power.
- At 0715 a regulating rod became stuck and misaligned by 12" from the group average.
- At 0750 (current time) the problem causing the rod to be stuck is corrected and the rod !
- is ready to be realigned.
What is the MAXIMUM PERMISSIBLE power level at which the control rod may be realigned
with the group?
A. 45% i
l
B. 55 %
C. 60%
D. 100% i
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QUESTION: 024 (1,00)
Current RC-P-1 A conditions: :
. Number one seal leak off flow indication is 5.8 gpm. j
- Periodic RCP shaR vibration ALERT alarms are actuating - alarms can be reset without :
immediate re actuation. ;
- Bentley-Nevada vibration readings are ranging between 14 and 18 mils.
- Number one Seal leak-off temperature indication is 197'F.
- Radial Bearing temperature indication is 170*F. 1
- High Standpipe level alarm is clear. !
,
Which ONE (1) failure could cause the above indications?
A. Seal #1 1
B. Seal #2
l
C. Seal #3
D. Labyrinth seal l
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QUESTION: 025 (1.00)
Current plant conditxms are:
i
- Reactoris operating at 95% power. i
- RCS pressure is 2150 psig and decreasing slowly.
- Pzr level is 200 inches and decreasing slowly.
- MU TANK LEVEL LO alarm is actuated. ,
- SEAL INJECTION FLOW LOW alarm is actuated. l
'
- Total RCP seal injection flow indication has decreased to 22 gpm.
Which ONE (1) abnormal condition could result in these indications?
- - A. RCP sealfailure l
B. MU-V-17 failed open
C. Sealinjection line leak
D. Makeup line;eak
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QUESTION: 026 (1.00)
l
Current plant conditions are:
- Reactor is operating at 100% power.
- INST AIR PRESS LOW TURBINE AREA alarm is actuated.
- INSTRUMENT AIR PRESS LOW AUX BLDG AREA alarm is actuated
- Instrument Air pressure is 58 psig.
- Secondary plant is stable
Which ONE (1) statement describes the required actions? ,
1
A. Manually trip the reactor and perform the immediate Manual Actions of 12101, ;
Reactor Trip. '
i
B. Dispatch operators to start backup instrument air compressors. '
C. Maintain power at present level and make plant page and radio announcement to
all personnel using instrument air to stop use immediately. I
D. Cross connect instrument and service air headers until cause for low header
pressure is determined and corrected.
1
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, . . _ _ . . . _ . __ ._.... _ _ .__ _ .._. _ _._.. _ __.__. _ _ _ __.__._ . _ _ _ .._. _.__ ,
l- )
QUESTION: 027 (1.00)
- Current plant conditions are:
- ESAS actuation (A & B) occurred 1 minute ago.
Which ONE (1) statement describes the operation of the RB fans and coolers
(AH-E-1N1B/1C)?
A. Fans run in slow speed with river water flowing through the emergency cooling
coils.
B. Fans run in fast speed with river water flowing through the emergency cooling
coils.
C. Fans run in slow speed with river water flowing through the normal and
emergency cooling coils.
l
l D. Fans run in fast speed with river water flowing through the normal and
emergency cooling coils.
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QUESTION: 028 (1.00) :
l
Initial plant conditions are: !
- Reactoris shutdown.
'
- RCS temperature is 250*F and stable.
.
'
- RCS pressure is 250 psig and stable.
- Pressurizer level is 200 inches and stable.
- . . DR-P-1 A, DC-P-1 A, and DH-P-1 A are operating.
'
- DC-V-2A is closed, and DC-V-65A is open
- Operators have just initiated DHR flow through the "A" DHR Cooler.
The following parameters are now changing:
i
- Pressurizer level is slowly decreasing. !
-
- LT-109 indicates DC-T-1 A is increasing. l
- RM-L-2 count rate is increasing.
Identify the ONE (1) cause for the above conditions: j
A.
DHCCW temperature is increasing due to energy transfer from the DHRS.
i B. DC-T-1 A fill valve (DC-V-19A) is failing open.
- . C. DHR cooler is leaking into the DHCCWS. l
. D. LT-109 is failing high.
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QUESTION: 029 (1.00)
1
initial plant conditions are:
l
Reactor is operating at 100% power.
- CO-P-1 A and CO-P-1B are running, CO-P-1C is in Normal-After-Stop.
- CO-P-2A and CO-P-28 are running, CO-P-2C is in Normal-After-Stop. ,
- CO-P-1B trips on an electrical fault. '
- After a period of two (2) seconds, CO-P-1C automatically starts.
Which ONE (1) statement describes the response for these conditions?
A. One condensate booster pump will trip, both main feed pumps remain running,
with an ICS runback.
B. One condensate booster pump and one main feed pump will trip, with an ICS ,
runback. l
C. One main feed pump will trip, both condensate booster pumps remain running,
with an ICS runback.
D. Both main feed pumps trip, one condensate booster pump trips and the reactor
trips.
T
. _ _ . . _ . _ . . _ - . _ _ . _ . _ _ . .- _ -. -.__ _..__ _ __
QUESTION: 030 (1.00)
Current plant conditions are:
- The reactor is operating at 95% power.
- Control Rod exorcising is in progress.
- Rod Control, FeuJwater, and Reactor-Steam Generator Master controls are in MANUAL.
- An "OUT-INHIBIT * condition is illuminated on the Diamond rod control panel.
- A rapid reduction in power level has just occurred with fluctuations in RCS temperature,
pressure and pressunzerlevel.
- Asymmetric Rod alarm is actuated.
- Current NI readings are as follows:
M-5 N14 NI-7 NI-8
90% 96% 97 % 96 %
Identify tlw ONE (1) cause for the above conditions:
A. Ni detector power stpply fault
B. Partial insertion of rods during exercising
C. Azimuthal xenon oscillation
D. Dropped rod
.
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QUESTION: 031 (1.00)
Current plant conditiens are:
- Reactor is operating at 100% power.
- RCS pressure is 2155 psig.
- CRO closed RC-V-2 one minute ago due to suspected PORV leakage.
- RC Drain Tank pressure is 5.0 psig.
If the PORV is leaking, what is the expected tailpipe temperature?
A. 162 degrees F
B. 212 degrees F
C. 228 degrees F
D. 267 degrees F
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QUESTION: 032 (1.00)
Current plant conditions are:
- Reactor is at cold shutdown condition.
-
"A" Decay Heat Removal string is operating.
- Decay Heat Closed Cooling flow through the Decay Heat Removal cooler is throttled to
maintain the RCS at 130*F.
- Total loss of instrument Air (0 psig) occurred.
i
Which ONE (1) statement describes the response of the cooling system and subsequent effect
on RCS temperature for this situation? i
.A Closure of DC-V45A (Cooler bypass) AND DC-V-2A (Cooler inlet) results in
RCS heatup.
B. Opening of DC-V45A (Cooler bypass) AND DC-V-2A (Cooler inlet) results in i
RCS cooldown.
C. Closure of DC-V-2A (Cooler inlet) results in RCS heatup.
D. Opening of DC-V-2A (Cooler inlet) results in RCS cooldown.
.
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QUESTION: 033 (1.00)
Whch ONE (1) statement desenbos the requirements for performing an independent Verifcation
for an opea valve (located in a High Radiation Area) that is required to be closed?
A. One individual closes the va!ve, SECOND individual verifies from an independent
'
remote position DEMAND indicator that the valve closed
B. One individual closes the valve, SAME individual independently uses a remote
position DEMAND indicator to verify valve is closed.
C. One individual closes valve, SECOND individual independently verifies the valve is
closed (locally).
'
D. One individual closes valve (locally), SAME individual verifies from an independent
remote POSITION indicator that the valve is closed.
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QUESTION: 034 (1.00)
l
Current plant conditions are: l
- Reactoris tripped.
- Large break LOCA is in progress.
- RCS pressure is 540 psig. . ;
- - Reactor Building pressure is 35 psig.
t
When ESAS actuates properly, which ONE (1) statement describes the expected lineup for the
listed support systems?
l A. Seal injection is isolated to the RCPs since they should be tripped.
l
B. NSCC is isolated to the RCP motors since they should be tripped.
C. NSCC is aligned to the RCP motors to support pump operation if needed.
I
'
D. ICCW is aligned to the RCDT cooler to prevent flashing in the tank.
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QUESTION: 035 (1.00)
During a radioactive gaseous release from a Waste Gas Decay Tank, RM-A-7 HIGH alarm
actuated. Whis ONE (1) statement describes required automatic actions for this condition?
A. Trips AH-E-11
B. Trips AH-E-10 & 11, closes WDG-V-47, and starts MAP-5 iodine Sampler
<
C. Closes WDG-V-47
D. Closes WDG-V-47 and starts MAP-5 iodine sampler
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. _ .__. _ _ _ _ . _ _._ _ .._ _._ _ _ ._ _ . _ _ . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
_
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QUESTION: 036 (1.00)
Which ONE (1) statement is the basis for the power imbalance RPS trip envelop for nuclear
overpower?
l
A. Assure acceptable power distribution is maintained for control rod misalignment
'
analysis.
B. Assure Nuclear Peaking Factors are within limits in the event of a cold water
l accident
!
C. Assure transient protection (minimum DNBR) is maintained for loss of coolant flow
events.
l D. Assure uniform fuel burn-up over core life.
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QUESTION: 037 (1.00)
Current plant conditions are:
- Large break LOCA is in progress.
- RB pressure and temperature are elevated.
Which ONE (1) statement describes the cause and effect for erroneous pressurizar level
indication during LOCA conditions? '
A. Level indicates high due to RB depressurization effects by RB spray. '
B. Level indicates low due to RB depressurization effects by RB spray.
C. Level indicates high due to reference leg boiling.
D. Level indicates low due to reference leg boiling.
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. - . - -. . - . - .--. -. .- . . . - - . - . . - . - - .
QUESTION: 038 (1.00)
Emergency feedwater pump suction can be lined up directly from the hotwell. After vacuum is I
' broken, the Emergency Alignment pushbutton is depressed to realign the following valves: l
!
'
l
- Condensate Reject Valve CO-V4 l
- Normal Makeup Valve to the Hotwell CO-V-7 '
i - Emergency Makeup Valve to the Hotwell CO-V-8
Which ONE (1) statement describes valve response to operation of this pushbutton?
A. CO-V-6 and CO-V-7 open, CO-V-8 closes.
B. CO-V4 and CO-V-7 close, CO-V-8 opens. ;
1
'
C. CO-V4 opens, CO-V-7 and CO-V-8 closes.
D. CO-V-6 and CO-V-8 open, and CO-V-7 closes.
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L QUESTION: 039 '(1.00)
l
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Which ONE (1) accident situation would result in a direct Main Feedwater isolation by HSPS?
A. Foodwater line break outside RB at FW-V-17NB
B. Steam line rupture outside RB
, C. Large break LOCA
l
l D. OTSG tube rupture
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QUESTION: 040 (1.00)
2
' l
Which ONE (1) statement describes the restrictions on operation of TMI-1230KV switchyard I
auxiliary transformer disconnect switches?
A. Opening operations limited to normal load current interruption.
B. Opening operations limited to isolation of energized transformers (unloaded).
.
C. Opening operations limited to isolation of transformers currently de-energized.
D. Closing operations ONLY after synchronization is completed due to possible
,
arcing. i
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QUESTION: 041 (1.00)
Current plant conditions are:
l- - Reactor power is 12%.
! - Main turbine is on line in manual control.
- Unit load demand is 12%.
l - Turbine Bypass Valves are fully closed in automatic control.
l - Steam header pressure is 885 psig.
!
! Without operator actions, which ONE (1) statement describes the response of the turbine bypass
l valves to a pressure increase in both OTSGs to 960 psig?
!
A. Valves remain closed, since the pressure is less than setpoint plus selected bias. l
l
B. Valves open to reduce pressure to setpoint plus 10 psig.
C. Valves open to reduce pressure to setpoint plus 75 psig.
'
- D. Valves open to reduce pressure to setpoint plus 125 psig.
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QUESTION: 042 (1.00)
An anticipatory reactor trip due to a turbine trip, or a trip of both feedwater pumps is designed
j to prevent which ONE (1) condition?
A. Challenges to steam generator tube integrity
' B. Exceeding core thermal limits (KW/ft limits)
C. Challenges to the PORV and pressurizer code safeties
D. Exceeding core DNBR limits
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. - . -.- .. - .~.- . - .. - _- _ _ -. . - . . - . . . - . . . . - - - . - . -
- QUESTION: 043 (1.00)
Current plant conditions are:
- Plant has experienced a reactor trip and loss of offsite power.
- EG-Y-1B diesel failed to automatically START.
Which ONE (;) condition will prevent automatic startup of the diesel generator?
A. The exciter Auto Manual switch in the control room is in the Manual position.
B. The Emergency Bypass selector switch at the EDG breaker cubicle is in the
Emergency position.
C. The Unit / Parallel switch is in the Parallel position.
D. 1B Diesel Auto-Standby / Manual-Exercise switch in the control room is in the
Manual Exercise position.
.
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QUESTION: 044 (1.00)
Which ONE (1) statement describes the purpose of the emergency diesel generator govemor
speed droop adjustment.
A. Limits voltage changes during load changes when running in Unit.
B. Adjusts engine response to load changes.
4
C. Limits maximum engine load.
.
. D. Prevents engine overspeed during initial start.
.
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' QUESTION: 045' (1.00)
Current plant conditions are:
-Unit is operating at 100% power.
-Total loss of Nuclear Services Closed Cooling Water occurred 5 minutes ago.
-Attempts to start NS-P-1A/B/C were unsuccessful.
Which ONE (1) condition requires the opesrator to trip Reactor Coolant Pumps?
l
A. Motor stator temperature indication is 140 degrees C.
' B. Motor bearing temperature indication is 180 degrees F.
C. Motor thrust beanng temperature indication is 200 degrees F.
D. Seal #1 leak-off temperature indication is 220 degrees F. !
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._. ___._._.._ . _ _ .__ _ _ . . . _ _ . . _ ..._. . _._ . _ . .. ___. _
. QUESTION: 046 (1.00)
An unexpeded illness of a CRO has occurred Wdh acton being taken to corred the situation,
which ONE (1) conddion desenbos the maximum time the shift crew may remain below minimum
shnng requirements?
A. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with RCS Taw < 200*F
B. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at all RCS temperatures
C. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with RCS taw > 200*F
D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if during shift relief regardless of RCS temperature
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i QUESTION: 047 (1.00) )
l Current plant conditions are. i
i
-
Reactor is at hot shutdown condition. 1
-
All RCPs are operating.
-
Plant electrical configuration is normal for power operating conditions.
l -
MU-P-1B is operating.
l
-
CW-P 1 A and CW-P-1C are not operating. ;
i l
l A tagging request requires 1E 125/250 VDC Bus to be removed from service. Which ONE (1)
! statement describes impact of de-energizing 1E 125/250 VDC Bus?
!
A. MU-P-1 A will not start manually (from the Control Room) or automatically.
i B. MU-P-1C will not start manually (from the Control Room) or automatically.
!
l C. CW-P-1 A will not start manually (from the Control Room).
D. CW-P-1C will not start manually (from the Control Room). l
l l
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._. _ . . . . _ . _ _ _ . . _ .. _ _ _ _ . _ _ __ . _ . _ _ _ _ _ . . . _ _ . _ _ _ . _ .. _.
QUESTION: 048 (1.00)
Current plant conditions are:
- Time is ten minutes after reactor trip due to loss of both Main Feedwater Pumps.
- EF-P-2A is ope ating, EF-P-1 and EF-P-2B are not operating.
- Tu is 585*F and slowly increasing.
- RCS pressure is 2300 psig and slowly increas'ng.
- All RCPs are operating.
- OTSG 1 A level is 25 inches.
- OTSG 18 levelis 0 inches.
- OTSG 1A pressure is stable at 1010 psig.
- OTSG 18 pressure is 800 psig and decreasing.
- RCS heat up rate is +75 F/Hr.
Which ONE (1) action is required concerning operation of the RCPs?
' A. Stop 1 RCP per loop.
B. Stop 3 RCPs.
C. Stop 4 RCPs.
D. -Continue to oper:te 4 RCPs.
-
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__ _. - . . . . . _ . _ .... _ -.--- _ ___. -. --- - _ - - . . - - ~.. - - - . _
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l QUESTION: 049 .
(1.00)
1
.
To support plant start-up, the valve line-up of a system is required to be modifed until work is
'
completed on that system. From the list below, select ONE (1) choice that identifies the )
position (s) required to approve this change in the valve line up?
I
l A. Shift Supervisor
B. One CRO and one Shift Supervisor / Shift Foreman with an SRO license
i
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C. One CRO and the on-shift STA
D. System Engineer
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QUESTION: 050 (1.00) l
l
A Tech Spec Surveillance is satisfactorily completed upon retuming a plant component to service
that has a Reguietory Retest tag. It is now permissible to remove the Regulatory Retest tag.
From the list below, select ONE (1) choice that identifies the person who is responsible to ensure
the Retest Tag book is property closed out. j
1
A. CRO who completed the test. ]
B. Responsible System Engineer witnessing the test.
C. IST Coordmator
D. Shift Foreman l
l
QUESTION: 051 (1.00)
Which ONE (1) Abnornwl Transient procedure has the highest priority during an emergency
situation?
A. Fvraneive Primary to Secchy Heat Transfer
B. Lack of Primary to Secondary Heat Transfer
C. Loss of 25'F Subcooling Margin
D. Steam GeneratorTube Leak
.
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QUESTION: 052 (1.00)
'IVhich ONE (1) statement desenbos the reason for manually tripping or verifying turbine trip in the
(mmediato Actions of ATP 1210-1 Reactor Trip?
' A. Ensures OTSGs are no longer cross connected through the Main Steam lines.
'
B. Rar* ram Main Feedwater flow requirements.'
C. Minimizes steam generator tube-to-shell delta-T.
D. Prevents an uncontrolled RCS cooldown.
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QUESTION: 053 (1.00)
Current plant conditions are:
- The reactor is operating at 50% power with rod control in automatic.
- Uncontrolled withdrawal of group 7 control rods is occurring.
- Control rod withdrawal command does not exist.
Which ONE (1) statemord describes required action for these conditions?
A. Select group 7and tum the Single Select switch to ALL.
B. Dispatch an operator to pull the fuses on group 7 programmer motor.
C. Depress the in-Limit-Bypass pushbutton and attempt to insert group 7.
D. Select Sequence Override at the CRD operator's console.
!
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- QUESTION: 054 (1.00)
Which ONE (1) condition requires initiation of HPl cooling according to the Abnormal Trarsient
procedures?
A. Pressunzer level cannot be maintained greater than 200 inches with the reactor at
100% power.
B. RCS subcooling is 30*F.
C. . Post trip RCS pressure is 1750 psig with incore exit thermocouples indicating 560
degrees.
D. Neither OTSG is available as a heat sink.
,
b
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-QUESTION: 055 (1.00)
'
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Completion of the Containment integrity Checklist is in progress in preparation to begin refueling
operations. Which ONE (1) condition prevents start of refueling operations?
A. RB purge is in progress.
B. MU-V-25 and MU-V-26 are physically removed for rebuilding.
,.
C. Service Airis in use inside the RB.
D. One door of the RB equipment hatch is open
,
_ .. _. _ _ _ _ _. - _ _ ~. . . - _ _ . . . _ . _ . . _ . _ . . . _ . __._ _ _.
QUESTION: 056 (1.00)
A point source in the Auxiliary Building is reading 500 mrem /hr at 2 feet. Two options exist to
complete a mandatory task near this radiation source:
Option 1: Operator X can complete the task in 30 minutes working at a distance of 4
feet from the point source.
Option 2: Operators Y and Z, using a special extension tool can complete the same
task in 75 minutes at a distance of 8 feet from the point source.
Which ONE (1) statement describes personnel exposure for completion of this task in accordance
with ALARA guidance?
A. Option 1 will expose Operator X to 62.5 mrem.
B. Option 1 will expose Operator X to 125 mrem.
C. Option 2 will expose each operator Y and Z to 39 mrem.
.
D. Option 2 will expose each operator Y and Z to 156 mrem.
.
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I QUESTION: 057 -(1.00)
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j Current plant conditions are:
- Reactor is operating at 20% power.
4 - CO-P-1 A and CO-P-2A are operating.
3
- One Main Feedwater Pump is operating.-
, - The following alarms are actuated simultaneously:
- . A-18 Battery 1B Discharging
~
- A-2-8 Battery Charger 1B/1D/1F Trouble 1
- A-3-8 inverter 1B/1D Inverter System Trouble I
! - H&V A3-2-3 Cont. Bldg. Batt. Cnargers B Damper Tbl. Fire-Smoke
- AA-3-2 7 KV Bus Trouble
-
.s
- AA-3-5 480V BOP Bus Trouble
Which ONE (1)_ action is required for this condition?
l A. ~ . Close DC tie switches to provide an alternate source of DC power.
1 B. Open suction valve (VA-V-58) for VA-P-1 B.
l C. Reduce power to within the reduced capability of the condensate system.
D. Transfer Alterex Excitation System DC power supply to B side DC power.-
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QUESTION: 058 (1.00)
An emergency event has been declared. Which ONE (1) statement describes the maximum time
limits for initial notification of the NRC, state, and local agencies?
A. 15 minutes for NRC notification,15 mir>utes for state and local notifications
B. 30 minutes for NRC notification,15 minutes for state and local notifications
C. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for NRC notification,15 minutes for state and local notifcations
D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for NRC notification, 30 minutes for state and local notifications
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QUESTION: 059 (1.00)
Which ONE (1) condition will actuate the CRD PATTERN ASYMMETR C annunciator?
-
A. Safety rods greater than 7 inches (5%) from group average position as determined
by absolute position indication.
. B. Safety or regulating rods greater than 9 inches (6.5%) from group average position
as determined by relative position indcation.
,
C. Safety or regulating rods greater than 7 inches (5%) from group average position as
determined by absolute position indication.
.
D. Reguieting rods greater than 9 inches (6.5%) from group average position as
determined by relative position indication
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QUESTION: 060 (1.00)
Cunent plant conditions are:
- Reactoris operating at 100% power.
'
- ICS is in full automatic.
- AT is ZERO.
- Selected Loop A Main Feedwater Flow instrument calibration is slowly drifting low- 2%
decrease (linear) over the past hour.
- SASS actuation has not occurred
Which ONE (1) statement describes the comparison between INITIAL and FINAL Feedwater
. Loop ACTUAL FLOWS at the end of the one hour?
A. Loop A FW flow same as initial flow, Loop B FW flow same as initial flow.
B. Loop A FW flow decreased by 1% and Loop B FW flow increased by 1%.
C. Loop A FW flow decreased by 2%, Loop B FW flow increased by 2%.
D. Loop A FW flow increased by 2%, Loop B FW flow decreased by 2%.
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QUESTION: 061 (1.00)
The crew has reduced RCS pressure to minimize subcooling margin in accordance with ATP
L 1210-5, OTSG Tube Leakage. Which ONE (1) statement describes the reason for minimizing
l subcoolmg margin?
A.- Minimize RCS leakage through the leaking OTSG tube.
- B. Minimize time required for cooldown of the RCS.
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C. Minimize potential of lifting Main Steam safety valves.
D.- Minimize tensile stresses on affected OTSG tubes.
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QUESTION: 062 (1.00)
.
,
Data contained on survey sheet:
- 2,500 DPW100 cm2 . betagunma i
- 10 DPW100 cm2alpha
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- 450 mRemthr general area
. - 470 mrem /hr on the surface of a tank
. . Airborne activity < 10% of a!! DACs !
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Which ONE (1) posting is required for these conditions?
A. Contaminated /High Radiation Area
B. Contaminated / Radiation Area
5 D. Airborne Radiation Area
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j QUESTION: 063 (1.00)
Current plant conditions are:
! - Numerous fire alarms are actuated on panels HVB, PLA, and PLB.
j - Fire dampers AH-D-4 and AH-D-5 have automatically closed.
j - Air tunnel deluge systems have actuated.
-
- Air tunnel halon system has actuated.
- Aux and Fuel Handling Building ventilation supply fans have tripped.
.
Which ONE (1) statement describes the action (s) required for this condition?
- A. Trip Aux and Fuel Handling Building ventilation exhaust fans.
l B. Actuate Relay Room CO2 system manually.
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! C. Trip reactor.
!
D. Start three fire pumps.
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l QUESTION: 064 (1.00)
Auxiliary Operator erroneously irwtiates a liquid release from A WECST. The Release Permit is 1
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for B WECST. Which ONE (1) cordton causes automatic termination of the accidental release
of radioactive liquid?
A. RM-L-7 in ALERT alarm QB loss of sample flow through RM-L4
B. RM-L4 HIGH alarm QB high tank release rate
]
3 C. High MDCT emuent flow QB low tank release rate j
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D. RM-L-6 ALERT alarm QR_RM-L-7 ALERT alarm
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QUESTION: 065 (1.00)
Waste Gas Decay Tank relief valve WDG-V-36 has opened due to high tank pressure. This
valve is now failed open. Which ONE (1) statement describes automatic action (s) initiated by the
(. Radiation Mondoring system related to this accidental gaseous release?
A. Trips AH-E-14NC (B/D)
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B. Trips AH-E-10 ANQ AH-E-11
C. Trips AH-E-10 ONLY
I D. Trips AH-E-11 ONLY
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QUESTION: 066
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(1.00)
The reactor is operating at 100% power. Which ONE (1) list identifies systems / tanks which
~ ALL require sampling and analysis for boron concentration FOLLOWING EACH MAKEUP
rather than relying solely upon a prescribed sample schedule or frequency?
A. . BWST, Core Flood Tanks, Spent Fuel Pool
B. Reactor Coolant System, BWST, Core Flood Tanks
.
C. Core Flood Tanks, Spent Fuel Pool, Boric Acid Mix Tank or Reclaimed Boric Acid
Tank
D. Reactor Coolant System, Core Flood Tanks, Spent Fuel Pool
.
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QUESTION: 067 (1.00)
Sequence of events:
- Reactor is operating'at 35% power. i
- FW-P-1 A is in operation; FW-P-1 B trip has not been reset, l
- Steam line rupture occurs upstream of Main Steam isolation Valve MS-V1 A. l
- The reactor and turbine trip.
- Main turbine Stop Valves 1-4 fail to close.
Which ONE (1) statement describes why only one OTSG depressurizes as a direct result of
the steam line break?
A. HSPS isolates Main Feedwater to OTSG 1 A. i
B. MS-V-1 A and MS-V-1 B are stop check valves that prevent back flow from
OTSG 18.
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C. Closure of turbine Control Valves 1-4 results in separation of the two steam
generators. l
D. Automatic open command for EF-P-1 steam supply valve MS-V-138 is delayed
for 40 seconds following MS-V-13A automatic operation.
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QUESTION: 068 (1.00)
Current plant conditions are:
- A reac'or trip has occurred. !
- Three control rods failed to fully insert into the core. 1
- Power Range Nis are off scale low. j
Which ONE (1) statement describes the required action (s) for this condition?
A. Emergency borate the RCS. l
.
B. Maintain primary to secondary heat transfer. ;
C. De-energize 1G and il 480 volt buses.
.
D. Perform immediate actions of EP 1202-8, CRD Equipment Failure for stuck rods.
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QUESTION: 069 (1.00)
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Current plant conditions are: 1
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- Reactoris operating at 100% power.
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- 6 CW Pumps are operating.
- Wwter conditions exist.-
Which ONE (1) statement describes when the CRO is required to manually trip the turbine?
Assume no automatic reactor trip.
A. Loss of one CW Pump
B. Condenser pressure is 8.7 inches Hg absolute i
C. Loss of the Gland Steam Exhauster
D. Condenser pressure is 7.7 inches Hg absolute
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QUESTION: 070 (1.00)
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Current plant conditions are:
- Reactor is operating at 75% power.
- RC-P-1C is shutdown due to high motor winding temperatures.
- Entry into the R8 is required to venfy the NSCCW valve line up for RC-P-1C.
- R8 purge must be initiated prior to the entry.
- D-Rings will not be entered. ;
Which ONE (1) position is authorized to approve this entry into the RB7
A. Director of Operations and Maintenance
B. Plant Operations Director
C. Shift Supervisor
1
D. Group Rad Con Supervisor
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QUESTION: 071 - (1.00)
.
.. .Which ONE (1) statement does NOT satisfy the Superheat Determination / Limit Rule as defined in
ATP-1210-107
A. 25*F of superheat as determined by the most conservative of the two s*veling
.
margin meters on panel PCL
!. B. 25*F of superheat as determined by the plant computer
l
i C. 25*F of superheat as determined by the average of 5 highest operable incore
j thermocouples and RCS wide range pressure
.
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D. 25 F of superheat as determined by the highest operable BIRO incore
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thermocouple and RCS wide range pressure
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QUESTION: 072 (1.00)
Current plant conditions are:
- Refueling operations are in progress.
'
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While lowering a spent fuel assembly into the upender, the assembly drops into the upender !
basket. The fuel handling bridge operator suspects the fuel assembly is damaged.
Whch ONE (1) set of statements describes Refueling Supervisor actions required for this event?
A. Stop all fuel movement. Further handling of the damaged assembly is allowed only
-
under your supervision with Core Load Engineer concurrence
B. Direct the dropped fuel assembly to remain in the basket. Continue refueling
- operations using the alternate transfer mechanism.
. C. Stop all fuel movement. Further handling of any fuel is allowed after obtaining
approval from the Director O&M. I
D. Direct the dropped fuel assembly to be transferred out of the RB to the Spent
Fuel Pool. Continue refueling operations using the alternate transfer mechanism.
.
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QUESTION: 073 (1.00) l
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Initial Plant Conditions
1
- Plant is in the Cold Shutdown condition. I
- RB purge is in progress.
- A indioactive liquid release is in progrees from A WECST.
- RM-A-4 (FH Building exhaust monitor) interlock control switch is in Defeat for I&C
calibration surveillance.
Twenty galions of highly radioactive liquid is accidentally spillad in the Fuel Handling Building. I
Which ONE (1) statement desenbes required automatic action (s) related directly to this spill !
incident? l
A. AH-V-1NB/C/D close to terminate R8 purge.
B. AH-E-14NC (B/D) trip to terminate exhaust ventilation flow.
C. AH-E-10 and AH-E-11 trip to terminate supply ventilation flow.
D. WDL-V-257 closes to terminate liquid release
1
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QUESTION: 074
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(1.00)
,
Current plant conditions are:
- Tube leak exists on OTSG 1 A.
- Plant cooldown is in progress per ATP 1210-5.
- BWST levelis 48 feet.
., - Tu is 500*F.
-- RCS pressure is 980 psig. j
Which ONE (1) statement describes the reason why OTSG 1 A should NOT be isolated?
A. Buildup of radioactive water in OTSG 1 A
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- B. Extension of time needed to reach Cold Shutdown
C. Increased probability of lifting Main Steam safety valve
. D. Isolation of one source of steam to EF-P-1
.
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QUESTION: 075 (1.00) ]
1
Current plant conditions are: i
I
- Refueling operations are in progress. l
- FTC water level is 24 feet above the fuel assemblies. i
'
- BWST is drained to 5 feet.
- RCS temperature is 90 degrees. 1
- Loop A DHR is in service. l
- Repair work is in progress on OTSG 1 A and OTSG 1B. !
Which ONE (1) statement desenbes Tech Spec requirements for the DHR System during these
conditions?
e
A. One DHR string is required to be operable; it is required to be operating. 4
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, B. Two DHR strings are required to be operable; one is required to be operating. I
C. One DHR string is required to be operable; it is not required to be operating.
D. Two DHR strings are required to be operable; neither is required to be operating.
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QUESTION: 076 (1.00)
Current plant conditions are:
- Alarge break LOCA has occurred.
- RCS pressure is 30 psig.
- Operator actions have been performed up to this point in accordance with ATPs.
- BWST levelis 6 ft. 4 inches.
Which ONE (1) statement describes the required sequence of operator actions to be performed
based upon the conditions described above?
A. Open the suction of the DH and BS pumps from the RB Sump (DH-V4A/B).
B. Close the suction of the DH pumps from the NaOH Tank (BS-V-2NB), then open
DH pump suction from the RB Sump (DH-V-6A/B).
C. Open DH and BS pump suction from the RB Sump (b, , V-6NB), then close BWST
outlet valves (DH-V-5A/B), then close NaOH tank Outlet Valves (3S-V-2NB).
D. Verify open DH and BS pump suctions from the RB sump (DH-V-6NB), then close
BWST outlet valves (DH-V-5NB), then close NaOH Tank Outlet Valves (BS-V-
2NB.
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QUESTION: 077 (1.00)
The reactor is operating at 100% power. Which ONE (1) statement explains the reason for
entering a Tech Spec time clock?
A. BS-V-1 A does not open automatically during ESAS testing.
B. Sodium Hydroxide Storage Tank NaOH concentration is 10.4%.
C. BS-V-49A (BS-V-2A inlet isolation valve) is not locked closed.
D. Sodium Hydroxide Storage Tank level is 8 ft. 2 inches lower than BWST level.
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QUESTION: 078 (1.00) l
Current plant conditions are:
l
- Reactor startup (approach to criticality) is in progress. 1
- Control rods reach upper ECP limit prior to achieving criticality.
'
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Which ONE (1) statement describes requirements for this condition? !
A. Commence immediate boration to achieve 1% Ak/k suberitical condition; reactor
startup may not be continued until authonzed by Nuclear Engineering.
B. Stop rod withdrawal and recalculate critical boron concentration; commence slow
controlled dilution to achieve criticality.
C. Trip the reactor; re-commence reactor startup after authorized by Plant Operations
Director.
D. Insert rods to achieve at least 1% Ak/k subcritical condition; notify Nuclear l
Engineering to evaluate ECP conditions prior to re-commencing startup.
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QUESTION: 079 (1.00)
Which ONE (1) condition will DIRECTLY close the letdown isolation valve, MU-V-37
. A. High letdown temperature
B. 30# RB pressure ESAS signal j
C. ReactorTrip isolation
D. High Makeup dem'nsralizer D/P
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QUESTION: 080 (1.00)
Which ONE (1) document is NOT required to be reviewed prior to assuming shift duties as the
licensed Shift Foreman? -
A. ESAS Checklist
B. Locked Valve Book
C. TCN/STP Book
C. Revision Review Book
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QUESTION: 081 (1.00)
Which ONE (1) statement describes a function of the Technical Support Center (TSC)?
A. Perform off-site dose prediction calculations.
B. Approve official press releases. i
C. Notify off-site agencies for event re-classification.
D. Perform backup RCS leakrate calculations.
.
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QUESTION: 082 (1.00)
Which ONE (1) statement explains the reason (s) EFW is injected into the OTSG near the top
of the tube bundle? i
A. Reduce the thermal stress on the lower tubesheet since the upper tubesheet can
withstand a higher thermal stress than the lower tube sheet.
i
B. Reduce the thermal stress on the lower tubesheet and elevate the thermal center I
of the OTSG.
C. Reduce the thermal stress on the upper shell (steam exit region) and elevate the
thermal center of the OTSG '
D. Reduce the thermal stress on the upper tubesheet and MFW nozzles.
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QUESTION: 083. (1.00)
The OTSG maximum allowable secondary pressure shall be limited to less than when
OTSG shell temperature is below .
A. 100 psig,100*F
B. .100 psig,200*F
C. 200 psig,100*F
D. 200 psig,200*F
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QUESTION: 084 (1.00)
The SBO diesel has the capability to energize 4160V bus through cross-tie breakers
via start aruf loading.
A. - Only C or D, manual
B. C ,D w E, ado
C. Only D or E, auto
D. C, D or E, manual !
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QUESTION: 085 (1.00)
An RCS sample is being drawn when RM-G-18 goes into HIGH alarm. . Which ONE (1) statement
describes the automatic actions associated with RM-G-18 for this situation.
A. CA-V-4A and CA-V-5A close.
B. CA-V-4B and CA-V4B close
C. CA-V-2 and CA-V-13 close.
D. The Control Building ventilation system is shifted to emergency recirculation mode.
,
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QUESTION: 086 (1.00)
-
Linear signal failure rates are desenbod in each condition listed. Which ONE (1) condition will
cause an automatic MANSFER by the Smart Auto Signal Selector (SASS) dunng full power
operation? I
- A. Selected RCS narrow range pressure instrument drifts to produce a difference amar
- of 4% over a five minute period l
B. Selected "B" OTSG pressure fails high at a rate of 10%/sec.
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- C. Selected Loop A Feedwater Flow instrument drift produces a difference error of 5%
over 10 minutes.
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D. Selected Ni Power Range Channel fails low at a rate of 10%/30 seconds.
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QUESTION: 087 (1.00)
A reactor trip from 100% power has occurred.
- All RCPs were tripped 60 minutes ago.
- Steam flow and feedwater flow have been verifed.
- incore thermocouples are tracking Tw.
Which ONE (1) set of plant conditions indicates natural circulation is occurring?
"
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Pressure Temperature Temperature
A. 700 psig, decreasing 540 deg F, increasing 550 deg F, stable
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B. 750 psig, stable 513 deg F, stable 545 deg F, decreasing
C. 800 psig, stable 550 deg F, decreasing 570 deg F, increasing
D. 940 psig, increasing 540 deg F, increasing 600 deg F, stable
1
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l QUESTION: 088 (1.00)
Which ONE (1) statement describes the function of the PORV NDTT Key Lock Switch on
Control Room panel PCR7
A. In the AUTO position, setpoint is 2450 psig ONLY if RCS temperature is BELOW
275'F.
B. In the OFF position, setpoint is 485 psig at ANY RCS temperature.
C. In the AUTO position, setpoint is 485 psig ONLY if RCS temperature is ABOVE
275'F.
D. In the OFF position, setpoint is 2450 psig at ANY RCS temperature
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QUESTION: 089 (1.00)
Which ONE (1) statement describes the operation of RC-V-1 Pressurizer Spray Valve?
A. . In Manual control, the JOG circuit directs movement of the valve using momentary
contact pushbuttons; valve position is limited to 40% open position.
B. In Manual control, the JOG circuit directs movement of the valve using momentary
contact pushbuttons; valve can be opened to 100% open position.
C. In Automatic control, valve operation is controlled by RCS pressure conditions;
valve is opened to 60% open position in response to open commands.
i D. In Automatic control, valve operation is controlled by RCS pressure conditions;
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valve is opened to 100% open position in response to open commLnds.
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QUESTION: 090 (1.00)
Which ONE (1) condition actuates the control rod withdrawal Out-Inhibit?
A. Sequence Fault condition due to 30% overlap between 2 successive regulating
rod groups
B. ICS neutron error signal of-5%
C. Reactor power at 56% with Asymmetric Rod Fault
. D. Startup rate at +3.2 DPM on one Intermediate Range NI channel
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! QUESTION: 091 (1.00)
l Current plant conditions are:
(
- Plant is shutdown in preparation for a Refueling Outage.
l - RB purge is in progress for one hour. ,
- RB sump liquid is being gravity drained to the Auxiliary Building sump.
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l Which ONE (1) statement describes required actions for actuation if RM-A-9G HIGH alarm?
i A. WDL-V-534 and WDL-V-535 RB sump drain line isolation valves close. ,
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B. Kidney Filter Fan (AH-E-101) trips. ;
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' C. RC Drain Tank vent isolation valvss (WDG-V-3/4) close.
l D. RCS letdown isolation valves (MU-V-2N2B) close.
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QUESTION: 092 (1.00)
The reactor is operating at 100% power. Which ONE (1) condition causes automatic start of ALL
Emergency Feedwater Pumps?
.
l A. FW-P-1 A hydraulic oil pressure at 68 psig, E loss of two RCPs in RCS Loop A
1
! B. Loss of two RCr 4 in RCS Loop B M 24 inches OTSG 1B Startup Range level
C. RB pressure at 5 psig, QB 9 inches OTSG 1 B Startup Range level
D. .FW-P 1B hydraulic oil pressure at 74 psig, QB 24 inches 1B OTSG Startup Range
level
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- ... . - -. -. . - - . . - - - - . - _ - - - - - - - . - . . . . . - . ..
QUESTION: 093 (1.00)
Current plant conditions are:
- Reactor is at cold shutdown condition.
l - DHR Loop A is operating.
- PARTIAL loss of ICS/NNI Hand Power (DH AUTO SUBFEED) has occurred.
Whch ONE (1) statement describes information the operator can use to quickly determine if
any or all Control Room DHR Loop A analog meters are inoperable due to LOSS OF DH
AUTO POWER?
!- A. Analog meters Fall LOW when instrument power is lost.
B. Flow instrument fluctuations (flow noise) STOP when instrument power is !ost.
C. EP 1202-40 Loss of ICS Hand and Auto Power enclosures identify specific
meters affected by loss of DH AUTO POWER.
D. Plastic labels glued onto the control console identify DHR analog meter power
supplies.
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QUESTION: 094 (1.00)
I
Current plant conditions are:
'
- Reactor is operating at 100% power.
- ICS is in full Automatic.
4
- Due to relay actuations, a trip of ALL 230 KV LINE Breakers EXCEPT those associated
with the 1091 line (Middletown Junction) occurs.
Which ONE (1) statement describes the expected plant design response to these conditions?
A. Plant remains at 100% load,1 A & 1B Aux Transformers remain energized.
B. Plant remains at 100% load,1 A Aux Transformers is de-energized,1B Aux
Transformer remains energized, canying all station loads.
. C. Automatic load reduction to 50% due to inadequate line capacity,1 A & 1B Aux
Transformers remain energized.
,
D. Reactor Trip, loss of off-site power.
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QUESTION: 095 (1.00)
During refueling operations RM-G-6, the Auxiliary Fuel Handling Bridge monitor, becomes
' inoperable. Which ONE (1) condition allows Refueling Operations to continue?
A. Install a suitable portable instrument of comparable range and sensitivity.
B. Verify RM-G-7 is operable on the adjacent Fuel Handling Bridge.
C. Obtain the approval of the Fuel Handling Supervisor.
D. Obtain the approval of the Radiological Controls GRCS.
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QUESTION: 096 (1.00)
Current plant conditions are:
- A LOCA has occurred.
,
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- Off-site power is being supplied.
- Hydrogen Recombiner is in service due to elevated RB hydrogen concentration.
- Plant electrical systems are aligned normally.
- No equipment failures have occurred.
Control Room staff has just been informed that the Hydrogen Recombiner reaction chamber
temperature has steadily decreased over the last half-hour from 1300 F dowri to 1280 F.
Which ONE (1) statement describes the reason for the slowly decreasing Reaction Chamber
temperature?
A. RB Pressure decreasing.
B. RB Pressure is increasing.
C. Hydrogen concentration is decreasing.
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D. Hydrogen concentration is increasing.
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QUESTION: 097 (1.00)
Current plant conditions are:
- Reactoris tripped from full power.
- HPl has been manually actuated due to LOCA.
- RCS Subcooling Margin is 3 degrees F.
- 4 RCPs are running.
After stopping one RCP in each loop, the primary CRO was distracted. No other control
manipulations have taken place for the last 4 minutes. Which ONE (1) statement describes
required actions for this situation?
A. Trip remaining RCPs, verify EFW, and raise OTSG levels to 75-85%.
B. Trip remaining RCPs, verify EFW, and raise OTSG levels to 50%.
C. Trip one RCP, start EFW, and raise OTSG levels to 50%.
D. Continue operation of both RCPs, start EFW, and raise OTSG levels to 75%-85%.
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QUESTION: 098 (1.00)
Current plant conditions are:
l - Reactoris operating at 90% power
'
- All RCPs are operating.
- High Vibration alarm actuates on RC-P-1 A (15 mils and slowly increasing).
l - Reset of vibration alarm has failed to clear the alarm.
1
Which ONE (1) statement describes the required actions required for this situation?
! A. Start the oil lift pump, then if vibration does not decrease trip the RCP.
l B. Trip RC-P-1 A, then reduce power as necessary to stabilize plant.
C. Reduce power to 50% - 75%, then trip RC-P-1 A.
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D. Close #1 Seal Leakoff valve (MU-V-33A), within 5 minutes, reduce power to 50% -
75%, and trip the affected pump within 30 minutes.
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QUESTION: 099 (1.00) j
Current plant conditions are:
- Reactoristnpped
- Manual HPl and Manual 4 PSI ES actuation on BOTH trains. !
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, - Automatic ESAS actuation has NOT occurred
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Refer to the attached control panel drawing. The control room operator momentarily !
depresses the following pushbuttons:
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- 2 of 3 Enable \ Defeat pushbuttons on the 4 PSI Manual actuations (ONCE) for both i
trains
- All 3 Enable / Defeat pushbuttons on the HPI Manual actuations (ONCE) for both trains ;
Which ONE (1) statement describes the status of the Manual ES signals to affected plant
,
components (pumps, valves, etc.)?
A. BOTH the Manual 4 PSI and the Manual HPI ES actuation signals ARE
DEFEATED for ALL affected plant components.
B. The Manual 4 PSI ES actuation signals ARE DEFEATED for ALL components
, affected by the 4 PSI actuation - but the Manual HPl actuation signals ARE NOT
DEFEATED for ALL components affected by the HPl ES actuation.
C. The Manual HPl ES actuation signals ARE DEFEATED for ALL components l
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affected by the HPl actuation - but the Manual 4 PSI actuation signals ARE NOT
DEFEATED for ALL components affected by ths 4 PSI ES actuation.
D. BOTH the Manual 4 PSI AND the Manual HPl signals ARE NOT DEFEATED for
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ALL affected plant components.
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QUESTION: 100 (1.00)
Current plant conditions are:
- Readoris operating at 90% power
- EG-Y-1 A is being started for surveillance
- Synchroscope indicator is at 12:00 and is not moving in e,ther forward or reverse direction.
- Indications at time of EG-Y-1 A breaker closure:
- Generator load is 0.15 MW.
- Generator reactive load is - 0.3 MVars.
After closing the output breaker, the operator pauses for five seconds to monitor generator output
conditions. Which ONE (1) statement describes the status of the output breaker and generator
FIVE SECONDS after the breaker was initially closed?
A. Output breaker is closed with generator load slowly increasing above 0.15 MW.
B. Output breaker is open due to over-current relay trip protection.
C. Output breaker is closed with generator output at 0.15 MW.
D. Output breaker is open due to reverse power trip relay protection.
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TMl UNIT 1 SRO LICENSING EXAMINATION
8/27/98
ANSWER KEY
1. . . . ... B 26. .. .. A 51. ... . C 76.....D
- 2.... ... C 27. ..... A 52. ... . D 77. . . . . . .A
3... ... B 28. .. .. C 53. ..... D 78... ...D
4. . . .... B 29. ..... B 54. ..... D 79.... . . . A
5.. .. ... B 30. .. .. D 55. .. . B 80. . . . . . . B
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6....... B 31. ... .. C 56. ... . A 81. . . . . . . D
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7. . ... . . A 32. ..... D 57. ..... C 82. . . . . . . B
8. .. .... C 33. . ... C 58. ..... C 83.. ....C
9. .. .. .. B 34. ... . B 59. ..... C 84. . . . . . . D
10. .... C 35. ..... C 60. ... . A 85. . . . . . .C
11. .. .. C 36. ..... C 61. .. .. A 86. . . . . . . B
12. .... D 37. ..... C 62. ..... A 87... . .. . B
13..... C 38.....B 63. . ... D 88.. ....D l
14. .... C 39. ..... B 64. .. .. B 8 9... . ... B
15. .. .. D 40. ..... B 6 5. ..... B 90... ...D
16.. .. C 41. ... .. B 66. ..... A 91... ... . A
17. .... B 42. ...C 67. ..... B 92. . . . . . . C
18. .... C 43. ..... D 68. .. .. A 93.. ....D
19. .... C 44. ..... B 69. ..... B 94.. .... . A
20....C 45. ..... C 70. ..... B 95.. ...A
21....A 46. ..... B 71. ..... A 96.... ..C
. 22. .. .. B 47.....A 72. ..... A 97.....D
23. .... D 48.....A 73. ..... C 98.. ..C
24..... A 4 9. . . .. B 74. . . . B 99.......C
25. .... D 50. . .. D 75. ..... C 100.....D
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Attachment 2
SIMULATION FACILITY REPORT
Facility Licensee: Three Mile Island Unit 1
Facility Docket No: 50-289
Operating Tests Administered from: Auoust 24-25,1998
This form is used only to report simulator observations. These observations do not -
constitute audit or inspection findings and are not, without further verification and review,
indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC
certification or approval of the simulation facility other than to provide information that
May be used in future evaluations. No licensee action is required in response to these
observations.
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No simulator deficiencies, that affected the scenario examinations or JPMs, were identified
during the conduct of the examination.
. . .
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