IR 05000289/1998301
| ML20155J188 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 11/04/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-289-98-301, NUDOCS 9811120053 | |
| Download: ML20155J188 (1) | |
Text
_ . _ _ - . _ _ _ . . - _ - - -
Nevern bete f /9'W NOTE T0: NRC DOCUMENT CONTROL DESK MAIL STOP 0-5-D-24 FROM: , LICENSING ASSISTANT OPERATING LICENSING BRANCH _ REGION I SUBJECT: OPERATOR LICENSING EXAMINATION ADMINISTERED QN hunns$ 24-27 N97, AT bree No hhnN D0bKETNO.g[-Aff ON ona0'24-2 7 /9ff0PERATOR LICENSING EXAMINATIONS WERE ADMINISTERED ATTH(REFERENCED' FACILITY. ATTACHED YOU WILL FIND THE FOLLOWING INFORMATION FOR PROCESSING THROUGH NUDOCS AND DISTRIBUTION TO THE NRC STAFF, INCLUDING THE NRC PDR.
Item #1 a) FACILITY SUBMITTED OUTLINE AND INITIAL EXAM SUBMITTAL DESIGNATED FOR DISTRIBUTION UNDER RIDS CODE A070.
-b) AS GIVEN OPERATING EXAMINATION, DESIGNATED FOR DISTRIBUTION UNDER RIDS CODE A070.
Item #2 EXAMINATION REPORT WITH THE AS GIVEN WRITTEN EXAMINATION ATTACHED, DESIGNATED FOR_ DISTRIBUTION UNDER RIDS CODE IE42.
R K O 29
)
I e
, - - . - .
'
% P Eto
'$ ** UNITED STATES j '
g NUCLEAR REGULATORY COMMISSION j REGION 475 ALLENDALE MOAD
\*; ..**- &g ' KING OF PRUSSIA, PENNSYLVANIA 19406 1415 September 24, 1998
,
Mr. ^
Vice President and Director ;
[ GPU Nuclear, Inc. i
Three Mile Island Nuclear Station .;
i- P.O. Box 480 I
- Middletown, PA 17057-0480 l
I
' SUBJECT: ~ THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR INITIAL EXAMINATION REPORT NO. 50-289/98-301 i !
e
Dear Mr. Langenbach:
-
e i
, This report transmits the findings of the senior reactor operator (SRO) licensing operating ,
! examinations, conducted by NRC examiners, during the week of August 24 - 27,1998 at I the Three Mile Island Unit 1 Nuclear Station. Based on the results of the examinations, all
,
four SRO applicants passed all portions of the examination. At the conclusion of tho
,- operating examination, Mr. P. Bissett discussed the preliminary findings with members of your staff. i i
! The examinations addressed areas important to public health and safety and were
-
developed and administered under interim Revision 8 of the Examiner Standards (NUREG- ;
. e '1021). All portions of the examinations were developed by Three Mile Island (TMI)
personnel, while the NRC provided oversight and final approval prior to the administration
. of the examinations. The operating portion of the examinations were administered by the j
.
- NRC, whereas TMI training personnel subsequently administered the NRC-approved, written portion of the examinations. -
,'
In accordance with 10 CFR 2.790 cf the NRC's " Rules of Practice," a copy of this letter
- ' and its' enclosures will be placed in the NRC Public Document Room.
l
,
,
"
g J '
.
. - . . --
?
I
'
Mr. No reply to this letter is required, but should you have any questions regarding these examinations, please contact me at 610-337-5183or by E-mail at RJC@NRC. GOV..
Sincerely, i
Richard J. Conte, Chief
! Operator Licensing and Human Performance Branch Division of Reactor Safety i
Docket No. 50-289 i
Enclosure:
Initial Examination Report No. 50-289/98-301 l
REGION 1 ]
1 s Docket No: 50-289 Report No: 50-289/98-301 i
License No: DPR 50 n
Licensee: General Public Utilities Nuclear Facility: Three Mile Island Unit 1 Nuclear Station Location: Middletown, Pennsylvania Dates: August 24 - 27,1998 Chief Examiner: P. Bissett, Senior Operations Engineer / Examiner Examiners: .D. McNeil, Senior Operations Examiner (Rill)
,
Approved by: Richard J. Conte, Chief Operator Licensing and Human Performance Branch Division of Reactor Safety 1
. .
P
'
,
- - - - _
. -
. . . . . . _ _ . . - . . . - . - . - - - - - . - - - - - - - . . - _ . . . ._- -
L I
l
' l
'
EXECUTIVE SUMMARY
, Three Mile Island Unit 1 Nuclear Station inspection Report No. 50-289/98-301 i Operations
'
Four senior reactor operator (SRO) applicants passed all portions of the initial license examination, i
.,
{ The applicants performed well on both the written and operating portions of the i 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 examinations. 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 j test the knowledge level of the applicants in the administrative area. Overall, examination quality was good, with the majority of changes having dealt with editorial changes or enhancements. ,
.
l l
l I
I i
l
)
.
ii ;
>
.
.
Report Details 1. Operatigna
.
05 Operator Training and Qualifications 05.1 Senior Reactor Operator Initial Examinations a. Eg.gns 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 Findinos Gradino and Results The results of the SRO examinations are summarized below:
SRO Pass / Fail Written 4/0 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. Allindividuals 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 exem preparation week of August 10,1998, the NRC subsequently reviewed and validated, together with TMl 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 TMI training representatives, it was agreed upon that administrative JPMs would be developed to replace those administrative questions that were not considered *
appropriate by the examiners. TMl subsequently developed several administrative
_ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ . _____.______ _ . _ _ __
i .
I
'
a l
.
! 2 l
l JPMs that met the guidelines of NUREG-1021. A couple of JPM follow-up l questions were also deemed to fall into the category of a direct lookup question.
L These questions were subsequently replaced with higher cognitive thinking level
! questions.
'
,
Administration and Performance
. The written portion of the examination was administered by TMI training personnel
- on August 27,1998, and consisted of 100 multiple choice questions. There were
- no comments by either the NRC or TMI concerning the validity of questions on the I 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 l appropriately changed the answer key.
!
e The operating portion of the examination was conducted from August 24 - 25, i 1998, and consisted of three simulator scenarios for each applicant. Also, ten l JPMs were administered to the SRO instant applicant, whereas the three SRO i upgrade applicants were required to perform only five JPMs. All JPMs were followed up with two system-related questions. As previously mentioned, administrative JPMs were developed and administered to all applicants to evaluate 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 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 building 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 i
'
result of control rod misalignment.
c. Conclusions The four 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 examinations. With few exceptions, excellent attention to detail prevailed throughout the examination development process.
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 TMI training department to test the knowledge level of the applicants in the administrative area. Overall, examination quality was good, with the majority of changes having dealt with editorial changes or enhancements.
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.
l l
.
I I
1
_ _ _ _ . _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _
__ . _ _ _ _ .. - _ __ _ _ . _ . . . _ . _ . - . _ _ . _ . _ _ . . _ _ _ _ . __ _. .. .. .
.
i'
.
.
PARTIAL LIST OF PERSONS CONTACTED
- Three Mile Island i
D. Boltz, Instructor, TMl Training R. Hess, Plant Training Manager
,
J. Langenbach, Vice President, Director TMI M. Ross, Director, Operations and Maintenance R. Parnell, Operations Training Manager
'
!-
P. Bissett, Senior Operations Engineer / Examiner D. McNeil, Senior Operations Examiner 4 Attachments:
1. Three Mile Island Unit 1 SRO Written Examination w/ Answer Key .
2. Simulation Facility Report
,
j
,
- .
l l
.
. - .
.
O Attachment 1 TMI 1 SRO WRITTEN EXAMINATION W/ ANSWER KEY
l l
l l
l l
l l
l t
. .
. _ . . _ . . . . _ . . . _ _ .. - . . . . . _ . _ _ . _ . _ _ . _ . _ ___ _ _ . _ _
- -
.
- Mr. Distribution w/enci and Attachments 1-2:
DRS Master Examination File PUBLIC .
Nuclear Safety Information Center (NSIC)
V. Curley, DRS Distribution w/ encl: w/o Attachinente 1-2: I Region 1 Docket Room (with concurrences) I NRC Resident inspector H. Miller, RA/W. Axelson, DRA P. Eselgroth, ORP N. Perry, DRP D. Haverkamp, DRP C. O'Daniell, DRP J. Wiggins, DRS j L. Nicholson, DRS l P.= Bissett, Chief Examiner, DRS !
DRS OL Facility File j DRS File
.
Distribution w/ encl: w/o Attachments 1-2 (VIA E-MAIL):
B. McCabe, OEDO l 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)
l
\
\
. _ . . - - _ - - - . .- - . - . - - - . _..-.-.- - --- --
.
, ,
.
U S Nuclear Regulatory Commission Site-Specific Written Examination Appliant information Name: Region: l l Date: August 27,1998 Facility / Unit Three Mile Island Unit 1 License Level: SRO Reacier Type: B&W-177 Start Time: Finish Time:
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. I Examination papers will be collected four hours after the examination starts. <
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results Examination Value 100 Points i Applicant's Score Points Applicant's Grade Percent l
l l
.
. __ _ . _ . _ . _ _ _ . _ _ . _ . - _ _ _ _ _ _ _ _ . _ . _ . _ _ . _ _ . . . _ _ . . _ _
- .
-
l
-
- QUESTION
- 001 (1.00)
! 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
) condition. Which ONE (1) parameter can be used IMMEDIATELY to make this determination?
A. Pressurizer spray valve posibon B. RCS pressure !'
.1 C. Makeup tank level t-
D. Pressurizerwater temperature ,
l
- i
!
a l
,
'
!
,
-
'
i i'
I l
.
T i
i
!
i
}
r i
i
1
i
'
i
l
. . . - - -
. . _ . . ... . . _ _ _ _ _ . . ._ _ _ _ . . _ _ . . - . _ . _ . _ _ _ _ _ _ . _ _ _ . _ . . . _ . . . . . . _ .
.
1.
.
QUESTION: 002 (1.00)
Current plant conditions are:
- The reactor is tripped.
- RCS = W margin is zero.
Whid1 ONE (1) uction results in increasing RCS subcooling margin?
A. Deasese RCS pressunzerlevel B. Decrease RCS hotleg flow C. Increase RCS loop pressure D. Increase RCS hotleg temperature I
)
J a
i
!
'~ ___ _ __ ._ _ _ _ _ _
_ _ . _ . . _ . _ _ . . _ _._. . _ . _ _ _ _ _
_ - . _ _ _ _ _ . . _ . _ _ _ _ _ . . . _ . _ _
.
.
j . .
l QUESTION: 003 (1.00)
'
'
Reactor power is at 100% when the cs4@@ RCS pressure channel, RC3A-PT1, fails instantaneously LOW. With NO operator action, which ONE (1) statement describes the
- SHORT-TERM plant response?
.
- A. Reador trip occurs on high RCS pressure A
B. SASS shifts control to RC38-PT1 RCS pressure channel to stabilize the plant i
C. Pressurizer heaters energize from RC3A-PT1 to increase RCS pressure
! D. Reactor trips and Safety inischon is aduated on low RCS pressure ,
l
! ;
I
\
,
i e
.
l
!-
i
l l
l l
l
,
e
4 h
'
i
'
1 i
l
. - - _ _ . . . . - - - . . _ . . _ - . . - . . - . . - . - . -
.
~
QUESTION: 004 (1.00)
!
The reactor is super entical and just entering the intermediate ranDe (4000 cps on the Source l
Range) when datedor compensating voltage to Ni-3 is lost. Which ONE (1) statement explains the effect this loss of compensating voltage will have on NI-3 indication? l A. NI-3 would be unaffected at this low power level.
B. Ni-3 would indicate higher than NIA.
C. NI-3 would come on scale some time after NI-4.
D. NI-3 would go off scale low before NI-4 if a reactor trip occurred while at 1 E-8
!
l l
l ( ,
,
l
l I
!
.
l
!
. -
. . . _ . . . _ . _ . _ _ _ _ _ . _ . _ . _ _ . . _ . _ _ . _ _ _ . _ _ . _ . _ . _ _ _ _ . _ _ _ . _ . _ . . _ . _
. .
QUESTION: 005 (1.00)
!
The plant is at 100% power when control room indications reveal the following:
- Reactor poweris DECREASING 4 - 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
'
EDd YELLOW
- - Indmating lights for both main generator breakers are RED i - Main generator electrical megawatts are 56 MW
!
.
j Which ONE (1) event is described by these symptoms?
j A. Loss of 230 kV suostation DC l
l B. Load rejection C. Auxiliary transformer fault
l D. Loss of station power
.
'
i
i
!
!
I
1 l
I i
1 l
$
l A
l r.- -
- . _ . _ . _ . . . _ _ _ . . _ _ . . _ . . . _ _ . _ _ . _ . . _ . _ . _ . _ . - _ _ _ _ _ _ . _ _ _ . _ _ . . . .
.
.
QUESTION: 006 (1.00)
Whkh 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 written oovenng the individual b be permitted to exceed 4000 mRom B. Approval from RadCon/ Safety Director and Site Director C. Approval from the President, GPU Nudear D. Notdication of the NRC
,
,
.. . _ _ _ . . _ _ _ _ _ _ _ _ . . . _ _ _ _ _ . _ _ _ . . . _ . . _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ . . . .
<
,
> .
- QUESTION
- 007' (1.00)
$ Given a Switching and Ta0ging Request to remove a fire system heat detector in the EG-Y-1 A l diesel room from service, idenbfy ONE (1) action is required to compensate for this detector l being removed from service.
'
.
A. Establish a fire watch PATROL within one hour to inspect the diesel room at least ONCE PER HOUR.
B. Station a CONTINUOUS fire WATCH in EG-Y-1 A diesel room WITHIH ONE j HOUR.
. C. START EG-Y-1 A to perform the one-hour surveillance to verify OPERABILITY.
!
j D. Restore detector to operable status WITHIN 14 DAYS or commence plant i shutdown to hot shutdown.
i
.
!
e
.
i
!.
!
.
l l
)
-. . . . _ - . . - . . .-.. - . . _ . . _ _ _ - . - . - . . . _ _ - - ..-
.
I
.
QUESTION: 008 (1.00)
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 rhive releases from exceeding 10CFR20 limits C. Recommend sheltering or eve A for the general population D. Determine if potassium iodKle tablets should be administered to reduce thyroid dose
_ _ _ _ _ - _ _ - _ _ _ _ - _ _ _ - - _ _ - _ - _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __________ . - .
. . - -- ~ - . _ . . - - - . . . . - - . . . - - - - - . - . . - . - . . . - . . - . . . .
.
'
.
t
- -
QUESTION: 009 (1.00)
Wdh the plant operahng at 50% power, BOTH intermediate Range NI detectors fail LOW. Which
- ONE (1) statement desenbos the required adion(s).
! A Conhnue power operations but limit power to 50%.
! B. Continue power operations, power may be increased.
f
! C. Immediately take adion 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. Take action wdhin one hour b 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 /> i
!
!
e e
!
t I
ii b
i I
!
!
i
!
i
!
i
!
,
d
!
!
!
!
(
i
!
$
!
!
i i,
__ ,__ , _
_
.
.
QUESTION: 010 (1.00) 1 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 (x) actions.
A. Director Operations & Maintenance l B. Plant Operations Director
C. Licensed Swiior Reactor Operator D. Licensed Reactor Operator
'
l l
'
1 i
l l
l
l l
i l
,
-__
.. . .. - . . _ .. - - . _ . .. -. - . - . . - _ - . - . = - .-.. .
i l QUESTION: 011 (1.00)
in addition to the person having the clearance, which ONE (1) of the followmg must grant permission to change the condition of BLUE tagged ES equipment?
l
A. Director Operations and Maintenance '
,
B. Plant Operations Director j
.,
j C. Duty Shift Supervisor / Shift Foreman
. ,
4 D. Licensed Control Room Operator l i l l
!
l i
,
A
3
i i
4
i
)
i i
i
!
J
!
i.
'
.
.
QUESTION: 012 (1.00)
Following a reactor trip, the following corxhons exist
- A OTSG level is 87" and demunsing slowly
- B OTSG level is 82' and decreasing slowly
- MFWflowis O gpm
- MFW volve DIP is O poig
- RCS pressure is 1725 psig and stable
- MUT levelis 82 inches
- PZR levelis 35 in$es
- MU flowis 60 gpm Which ONE (1) statemert desenbos the action requred per AP 1210-1, Rascior Trip?
A. Increase MFW pump speed.
B. Open MU-V-14A or MU-V-148 as necessary.
C. - Initiate HPl.
f D. Start second Makeup Pump and open MU-V-217.
L
l l
l
,
. __.. . _ _. _ _ .. _ _ _ _.. _ ._. _- . _ _ _ ..__ _ _ _ _ _ _ . . . _ _ . . . _ _ . . _
.
.
- QUESTION: 013 (1.00)
Current plant conditions are:
- - Reactor is operatog at 80% power.
, - RM4.-1 (RC Letdown) has increased to the ALERT setpoint. .
,
!
'
- Chemistry ereis indicates dose equivalent I-131 concentration is 0.28 uci/gm.
- RCS speedic 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 i shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
i 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
- - shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. l
j- 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 />.
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 reactor in cold shutdown.
I
l
!
i
!
!
e i
!
i i
I
. - . . - . . _ - - . - . - -. - . _ . . . - . - . - . - . - . - . - - . _ _ . - . . . . . . . .
.
.
.OUESTION: 014 (1.00)
Current plant conditions are:
- Reactoris operating at 1uu a power.
- RCS Tavg is constant at 579'F.
- Make up tank level is deusasing slowly - MU-V-17 is in MANUAL control.
- Letdown flow has been constant at 45 gpm. ~
- RCP total seal injection flow is 38 gpm (normal)- MU-V-32 is in AUTO.
- RCP labynnth seal D/P indicators show low off-scale (negative).
- Auxiliary Building airbome activity is increasing i
l 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.
!-
l
.
. _ . . _ . . . _ . _ . - . ... . . _ _ . . . . _ _ . . . -
_ _ . _ . . . . . _ _ . _ _ _ _ . . _ _ _ . . _ _ .
'
- QUESTION
- 015 (1.00)
Some reactor trip situations require large volumes of makeup water for RCS inventory control l l simultaneous with the need to emergency borate the core. Whid1 ONE (1) source should be used for this condition? l
- A. 4% BAMT (CA-T-8) l
l B. Concentrated Waste Storage Tank (WDT 6A/B) i i 1 l C. RC Blood Tank 1C (WDL-T-1C) l
'
D. BWST (DH-T-1)
'
!
.
h I
4
,
l
.
,
d
$
t h
!
a M
i i
E i
-
!
'
QUESTION: 016 (1.00) !
!
It is necessary b evacuate the control room due to a serious fire in the relay room. Whid1 ONE l (1) required action must be performed prior to exiting the control room?
A. Perform notdications for an ALERT.
B. Start EG-Y-1 A and EG-Y-18. ;
C. Trip the MFWpumps
. D. Close MU-V-3.
.-. - -. __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _
._ _ _- . _ _ _ . . _ . . . . _ - _ . _ _ . . _ _ _ _ _ _ . _ . _ _ . _ . _ . . . _ _ . _ _ _ _ . _ . . ___ ..._.__.
.
'
l
-
i l QUESTION: 017 (1.00)
RCPs are bumped during inadequate core heat removal conditions. Which ONE (1) statement desaibes the reason for this edion?
! A. Decrease RCS pressure i
B. Induce OTSG heet transfer
! C. Provwt RCS inventory loss
<
D. Increase OTSG pressure
.
e
4
h
i
<
f
!
'I f
f
.
!
.
A k
j
-
. - - ,n
'
,
!
-
QUESTION: 018 (1.00)
A small break LOCA is in progress. Which ONE (1) set of conchtions requires tripping all RCPs?
RCS TEMP RCS PRESS A. 579'F 1800 psig '
B. 537'F 1300 psig C. 525'F 1000 poig
'D. 473*F 800 psig
,
t
!.
'
.- - . _ _. . . - - - - _ _ - . - . .. - - . .. . _ - . . . . . - . - .
.
.
.
t
QUESTION: 019 (1.00)
.
Which ONE (1) condition requires a Core Flood Tank to be dedared inoperable?
.
A. Boron is 2290 ppm.
B. Pressure is 620 poig.
C. Levelis 14 ft
,
- D. Temperature is 100 degrees F.
i a
J t
e
.
.
QUESTION: 020 (1.00)
Plant conditions require manual reador trip. Upon depressing the Trip 6tiQ DSS pushbuttons, the reador does not trip (reactor power remains at 100%). WNch ONE (1)
statement desenbos the required adion?
A. Place the EHC pump control switdes in P-T-L and open EHC FV-1. ;
B. Place the diamond rod control stabon in manual and redum reador power.
C. Initiate HPI and maintain pnmary to secondary heat transfer until power level is less than 10%.
D. Transfer FW to manual to control OTSG levels.
I i
l
- -- . _ . .- .. - - _ . - . . . . - . - . . .. - -_ _. . - .. __ - - - . . _ _ . _ . .
d
'
.
QUESTION: 021 (1.00)
' BS-P-1B is running for surveillance. Which ONE (1) condition would result in automatic trip of the BS-P-1B breaker?
l l A. 1 A Aux Transformer fault with Auto Transfer of loads to 1B Aux Transformer !
i l B. 1B Aux Transformer fault with Auto Transfer of loads to 1 A Aux Transfonner l 1 C. Fault downstream of 1P 480v Bus low side feeder breaker
'
D. Fault downstream of 1S 480v Bus low side feeder breaker
i s
_ _ . . _
- . . _ _ _ . ._ _._._ _ _ __._ _ . _
.
.
QUESTION: 022 (1.00)
Current plant cordtions are:
- Pump down of RCS is in progress.
- DH-P-1 Ais in service
- Low DH Pump Flow annunciator C-1-7 is aduated. -
- RCStemperatureisincreasing
- Discharge pressure for DH-P-1 A is unstable
- Motorampera0sindicationforDH-P-1Aisunstable.
- Noise is reported in DH-P-1 A vault.
Which ONE (1) statement desenbos the required operator adions?
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 appropnate adions 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.
!
i
i i
- _ - - . _ _ _ _ _-+ _ . _ _ e- --
. -.. - . - . . . . _ . . . . - . - . . - . . . _ - - . . . - - . . . - _ . .
.
.
! QUESTION: 023 (1.00)
. Current plant conditions are:
d
- Reactor is operating at 100% power.
'
_
'
- At 0715 a regulating rod became stud 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 %
,
B. 55 %
]
C. 60%
, D. 100 %
,
t a
t
,
s
,
4
,
, + , - - ,
_ ._ . _ . _ _ _ _ . . _ _ _ . . . _ _ . ... ~ .. _ ...__ _ ._ _ . _ . _ . _ .. _ _ ... . . _ _ _ _ _ _._ _.._ _ . ..
..
!
=
i l QUESTION: 024 (1.00)
Current RC-P-1 A conditions:
- Number one seal leak off flow irdcation is 5.8 gpm.
! - Periodic RCP shaft vibration ALERT alarms are actuating - alarms can be reset without immediate reactuation.
- - Bentley-Nevada vibration readings are ranging between 14 and 18 mils.
- Number one Seal leak off temperature indicebon is ig7'F.
- - Redal Beanng temperature indication is 170*F.
- High Standpipe level alarm is dear.
Which ONE (1) failure could cause the above indications?
,.
A. Seal #1 ,
B. Seal #2 C. Seal #3 D. Labyrinth seal
.
. . - - . _ _ _ . . . _ _ . _ _ _ . _ . _ _ . _ . . _ _ _ . _ . . . _ . _ . . _ . _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ . _ _
- .
l' .
!.
f-
'
QUESTION: 025 (1.00)
} Current plant conchtions are:
I
. - Reactoris operahng at 95% power.
- RCS pressure is 2150 poig and decreasing slowly.
!
- Par level is 200 indies and deamesing slowly.
l - MU TANK LEVEL LO alarm is actuated.
. - SEAL INJECTION FLOW LOW alarm is actueled.
- Total RCP seal injection flow indcation has decreased to 22 gpm.
Which ONE (1) abnormal condition could result in these irth?
l A. RCP sealfailure
l B. MU-V-17 failed open l
l l C. Sealinjection line leak i
.
D. Makeup line leak k
,
d
i j,
j-
!
!
. . . _ -. - - . . . . - . - . - . ~ . _ . - - _ . _ . - . _ -
!
.
.
QUESTION: 026 (1.00)
Current plant conditions are:
- Roadoris 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?
, A. Manually trip the reactor and perform the immediate Manual Actions of 1:2101, Reactor Trip.
B. Dispatch operators to start backup instnJment air compressors.
C. Maintain power at present level and make plant page and radio announcement to l all personnel using instrument air to stop use immediately. I l
D. Cross connect instrument and service air headers until cause for low heada-pressure is determined and corrected.
l
. - - . . . . - . . - _ _ - _ _ _ _ . - - . . _ . - -.-._. .-. _ - . __ - . .
,
,
,
{ QUESTIOld: 027 (1.00)
J Current plant conditions are:
'
- ,
j - 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.
i B. Fans run in fast speed with river water flowing through the emergency cooling coils.
i j C. Fans run in slow speed with river water flowing through the normal and j emergency cooling coils. ,
!
'
- D. Fans run in fast speed with fiver water flowing through the normal and emergency cooling coils. !
l 4 ,
, !
i l
< I i ,
i l
'
,
i l l
4 i
,,
' '
)
\
l l
I I
l
l l
I l
l
,
-. __ - - - - . . _ - . . - .. - - - . _ _ - - . . ._-. - _. -.- . -. -
.
.
QUESTION: 028 (1.00)
Initial plant aanditions are:
- Reactoris shutdown.
- RCS temperature is 250*F and stable.
- RCS pressure is 250 peig 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 danging:
- Pressurizer level is slowly decreasing.
- LT-109 indicates DC-T-1 A is increasing.
- RM-L-2 count rate is increasing.
Identify the ONE (1) cause for the above conditions:
A. DHCCW temperature is increasing due to energy transfer from the DHRS.
B. DC-T-1 A fill valve (DC-V-19A) is failing open.
C. DHR cooler la leaking into the DHCCWS.
D. LT-109 is failing high.
i i
-. _ - . . . - _ _ _ . _ . _ _ _ _ _ _ _ _ ___._ . _ ._ _ _ ._.._ _ __ _ _ _ _ . _ _ ._
.
i
,-
,
QUESTION: 029 (1.00)
i initial plant conditions are:
! - Reactoris operating at 100% power.
,
- CO-P-1 A and CO P-1B are runnin0. CO-P-1C is in Normel After-Stop.
- - CO P-2A and CO-P-2B are running. CO-P-2C is in Normal-After-Stop.
i - CO-P-1B trips on an elecMeal 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 runbeck.
i.
! B. One condensate booster pump and one main feed pump will trip, with an ICS
'
- runback.
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. l
!
,
l
.
I
..
_ _._. _ .
- . _ . . . _ _ _ _ . _ _ _ . . _ _ . _ _ _ _ _ . _ _ _ _ _ . _ . _ _ _ . . . _ _ _ . _ _ . _ - _ -
.
.
QUESTION: 030 (1.00)
currervt plant constions are:
- The reactor is operating at 96% power.
- Control Rod exercising is in progress.
- ! Rod Control, Feedwater, and Reador-Steam Generator Master controls are in MANUAL
- An "OUT-INHIBIT * constion is illuminated on the Diamond rod control panel.
- A rapid reduction in power twel has just occurred with fluctuations in RCS temperature, pressure and pressurizerlevel.
- Asymmetric Rod alarm is stated.
- Current NI readings are as follows:
M M Ub1 M 90% 96 % 97 % 96 %
Identify the ONE (1) cause for the above conditions:
A. Ni detector power supply fau!t 3. Partial insertion of rods during exercising l
C. Azimuthal xenon oscillation l
D. Dropped rod L
l
. _ . _ ___ _ _ . . . . . . . . . _ _ . _ _ . . . _ _ . _ . . _ . . _ _ . _ . . . . . . _ _ . _ _ . . . . . _ . . _ _ _ _ . . _ . . . . _ _
.
t
'
.
l QUESTION: 031 (1.00)
,
Current plant conditions are:
,
! - Reactor is operating at 100% power.
! - RCS pressure is 2155 psig.
i - CRO closed RC-V-2 one minute ago due to suspected PORV leakage.
- RC Drein Tank pressure is 5.0 peig.
N the PORV is leaking. what is the expected tailpipe temperature?
l A 162 degrees F B 212 degrees F s
C. 228 degrees F
- '
D. 267 degrees F i
,
!
1 i
j.
i,
s i
n
, . , , _ . . _ ,
_ . - . . -. - - . -.. . - - - - . -. . - - - - . - _ - - -. - . _ _ -.
..
(
.
QUESTION: 032 (1.00)
Current plant conditions are:
- Reactor is at cold shutdown condition. l
- "A" Decay Heat Removal string is operating.
- Decay Heat Closed Cooling flow through the Decay Heat Removal cooler is throttled to j maintain the RCS at 130*F.
- Total loss of instrument Air (0 psig) occurred.
Which ONE (1) statement describes the response of the cooling system and subsequent effect on RCS temperature for this situabon?
'
A. Closure of DC-V 65A (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 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.
. . _ _ _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . __ ._ .
. . - _ _ _ . _ _ _ . . . _ . . . . . _ _ _ _ _ _ . _ . _ . _ . _ _ . _ . - - _ . _ _ . . _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ - . . _ . . .
d
.
I
'
, .
! QUESTION: 033 (1.00)
l Which ONE (1) statement desates the requirements for performing an independent Venfication
for an open valve (located in a High Radiation Area) that is required to be closed?
!
!
A. One individual closes the valve, SECOND individual venfies from an independent
- remote position DEMAND indcator that the valve ciosed.
I B. One individual closes the valve, SAME individual independently uses a remote 1 l
position DEMAND indcator to venfy valve is closed. !
l
C. One individual closes valve, SECOND individual independently venfies the valve is l
i closed (localy).
l
!
D. One individual closes valve (locally), SAME individual verifies from an independent I remote POSITION indcator that the valve is closed
!
I
,
<
.
- -.
_ . _ _ . . _ _ _ _ _ . _ . _ . _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ . . _
,
QUESTION: 034 (1.00)
Current plant conditions are:
- Reactor is tripped.
- Large break LOCA is in progress.
- RCS pressure is 540 psig.
- Reactor Budding pressure is 35 psig.
When ESAS actuates property, whidi ONE (1) statement desenbos the expected lineup for the listed support systems?
A. Seal iriection is isolated to the RCPs since they should be tripped 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.
D. ICCW is aligned to the RCDT cooler to prevent flashing in the tank.
I
<
____ _ _ _ _ _--
._ _ _ . . . . _ . . _ . _ _ . . _ _ _ . _ _ . . _ . _ . _ _ _ . _ _ _ . ~ . _ _ _ _ . _ . _ _ . _ _ _
,
.
I
.
QUESTION: 035 (1.00)
i .
During a radioactive gaseous release from a Weste Gas Decay Tank, RM-A-7 HIGH alarm
actuated. Whidi ONE (1) statement describes required automatic actions for this condition?
A. Trips AH-E-11 B. Trips AH-E-10 & 11, cioses WDG-V47, and starts MAP-5 iodine Sampler C. Closes WDG-V-47 i
D. Closes WDG-V 47 and starts MAP-5 iodine sampler i
a
!
i
!
.
i-
,
!
l I
!
_ _ __. _..____ _ . . . _ _ _ . _ _ _ _ . . ____._.._._ _...._. __ .._ ._
!
.
QUESTION: 036 (1.00)
Which ONE (1) statement is the basis for the power-imbalance RPS trip envelop for nuclear overpower?
A. Assure acceptable power distribution is maintained for control rod misalignment analysis.
B. Assure Nucieer Peaking Factors are within limits in the event of a cold water accident C. Assure transient protechon (minimum DNBR) is maintained for loss of coolant flow events.
D. Assure uniform fuel burn-up over core life.
l
,
l l
l l
l l
'
._- _.
,
.
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 effed for erroneous pressurizer level indicaten during LOCA conditions?
A. Level indicates high due to RB depressurization effects by RB spray.
B. Level irdcates 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.
!
l
I
._
.- - -- -. . .. - . . - - - - _ . _ . - - - . = - - . - - . . _ . -
,
.
QUESTION: 038 (1.00)
I Emergency feedwater pump suction can be lined up directly from the hotwell. After vacuum is broken, the Emergency Alignment pushbutton is depressed to realign the following valves: l
- Condensate RejectValve CO-V4 l
.
- Normal Makeup Valve to the Hotwell CO-V-7 i
!
- Emergency Makeup Valve to the Hotwell CO-V4 i
'
Which ONE (1) statement describes valve response to operation of this pushbutton?
i
,
i A. CO-V4 and CO-V-7 open, CO-V4 closes
I B. CO-V4 and CO-V-7 close, CO-V4 opens.
C. CO-V4 opens, CO-V-7 and CO-V4 closes.
D. CO-V4 and CO-V4 open, and CO-V-7 closes.
\
l
.. -_ . . _ . - . ._ . . _.. _ . . _ . _ _ _ _ . _ _ . _ _ _. . -.. _ _ _ _ . . . . _
.
.j
.
'
QUESTION: 039 (1.00)
Which ONE (1) accident situation would result in a direct Main Feedwater isolation by HSPS?
f A.- Feedwater line break outside RB at FW-V-17NB B. Steam line rupture outside RB
-
C. Large break LOCA
.
D. OTSG tube rupture
,
l l
l l
l
!
_. ._= _ .. - . - .. . - _ . - - .. ._ - --
,
'
.
l QUESTION: 040 (1.00)
i Which ONE (1) statement describes the restrictions on operation of TMI-1230KV switchyard auxiliary transformer disconnect switches?
- A. Opening operations limited to normal load current interruption.
! B. Opening openstions limited to isolation of energized transformers (unloaded).
j C. Opening operations limited to isolation of transformers currently de energized.
D. Closing operations ONLY after synchronizaten is completed due to possible arcing
,
i
.
)
l l
_ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ -____ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ - _ - _ _ _
. .
.
.
@ QUESTION: 041 (1.00)
Current plant conditions are:
- Reador poweris 12%.
- Main turbine is on line in manuel control.
- Unit load demand is 12%.
- Turbine Bypass Valves are fully closed in automatic control.
- Steam header pressure is 885 poig.
Without operator actions, which ONE (1) statement desenbos the response of the turbine bypass valves to a pressare increase in both OTSGs to 980 poig?
A. Valves remain closed, since the pressure is less than setpoint plus selected bias.
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.
.
I
.
. .. ..
.
. -. .. . . . - . - . . - - . _ . _ - . - . . - . .
, 1 0 .
l
- 1 i
I
-
(
.
.
QUESTION: 042 (1.00)
- An anticipatory reactor trip due to a turbine trip, or a trip of both feedwater pumps is designed
- to prevent which ONE (1) condition?
l A. Challenges to steam generator tube integrity l ,
! B. E+:::ing core thermal limits (KW#t limits)
.
- C. Challenges to the PORV and pressurizer code safeties D. Exceeding core DNSR limits
!
,
E p
,
l' -
. .. . _. _ ___ _. _ __. _ ._ _ __. _ .._ _ _ _ . _ _ _ . . _
,
+
.
QUESTION: 043 ~ (1.00)
i
. 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 (1) 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 swit& 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.
i
l l
l
l
.. - . - . . .- .. _ . - - . . . - . - - - . - - - - - - - . - . . - . - . . _ . - . . .
i
.
QUESTION: 044 (1.00)
Which ONE (1) statement desenbes the purpose of the emergency diesel generator govemor speed droop aquatment A. Limits voltage dienges during load changes when running in Unit.
B. Adjusts engine response to load changes.
C. Limits maximum engine load.
D. Prevents engine overspeed during initial start.
.
- . . . _ . - _ . _ .. . - _ - . . . - , - - . . . . - . . - - . - - . _ - . _
.
.
QUESTION: 045 (1.00)
Current plant conditions are:
-Unit is operahng at 100% power.
-Total loss of Nuclear Sendees Closed Coohng Water occurred 5 minutes ago.
-Attempts to start NS-P-1 A/B/C were ur=-W.
"
Whid) ONE (1) condsbon requres the operator to tiip Reactor Coolant Pumps?
!
A. Motor stator temperature indication is 140 degrees C.
B. Motor bearing temperature irhiari is 180 degrees F.
C. Motor thrust beanng temperature indicabon is 200 degrees F.
. D. Seal #1 leak-off temperature indication is 220 degrees F.
I
)
i i i l
!
!
.
.
QUESTION: 046 (1.00)
An unegoded illness of a CRO has ocx:urred With schon being taken to correct the situation, which ONE (1) conchtion desenbes the mammum time the shift crew may remain below minimum staffing 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 < 2007 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 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
,
C.
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
l l
l l
l
,
_ _ _. .._ . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . . ._ _
- ,
'
.
'
QUESTION: 047 (1.00)
Current plant conditions are:
'
- Reactor is at hot shutdown condition. ,
l
- - All RCPs are operating.
'
- Plant elodrical configuration is normal for power operating conditions.
- - MU-P-1B is operating.
- - CW-P 1 A and CW-P-1C are not operating.
,
A tagging request requires 1E 125/250 VDC Bus to be removed from service. Which ONE (1)
statement desenbos impact of de-energimng 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.
!
'
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).
k
,
l
,
A
!- 1 l
,
4 !
!
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _
_
j
. _ .
. _ . . . -- _ - - - - . - - . _ - - . _
.-..-..._..- - . - . - . ~ _ . - _ - . - . .
'
. l
- l
- .
I 4 QUESTION: 048 (1.00)
{ Current plant cordtions are:
!
- Time is ten minutes after reactor trip due to loss of both Main Feedwater Pumps.
- . - EF#-2A is operating, EF-P-1 and EF#-2B are not operating.
r -- Tu is 585V and slowly increasing. ,
i - RCS pressure is 2300 peig and slowly increasing. )
- All RCPs are operating.
-
l:
- OT8G 1Alevelis 25 inches.
-
.
, - OT8G 18 levelis 0 inches. n.3 l - OTSG 1 A psessure is stable at 1010 peig.
'
- - OTSG 1B pressure is 800 peig and decreasing.
.
- - RCS heat up rate is +75TMr.
,
Which ONE (1) action is required concerning opershon of the RCPs?
!
- A. Stop 1 RCP per loop.
.
B. Stop 3 RCPs.
C. Stop 4 RCPs.
,
D. . Continue to operate 4 RCPs. ;
k
.- -_ _ .. .. . _ . . _ _ _ _ _ _ - _ . .. .. _ __
Q
.
l l QUESTION: 049 (1.00)
To support plant start-up, the valve kne-up of a system is required to be modified until work is completed on that system. From the list below, select ONE (1) choice that identifies the position (s) required to approve this dian0s in the valve line up?
A. Shift Supervisor
'
B. One CRO and one Shift Supervisor / Shift Foreman w'.th an SRO license C. One CRO and the on-shift STA D. System Ergineer l
!
l i
!
!
!
l
!
i
1
,
.. . . .. . .- _ . - - - - . . - . - - . = .- .
.
.
QUESTION: 050 (1.00) -
A Tech Spec Surveillance is satisfactorily completed upon retuming a plant component to service that has a Regulatory Retest tag. It is now permissbie 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 properly closed out.
A. CRO who completed the test.
B. Responsible System Engineer wineesing the test.
C. IST Coordinator D. Shift Foreman l
l l
I
,
,
I
.
QUESTION: 051 (1.00)
Which ONE (1) Abnormal Transient procedure has the highest prionty during an emergency situation?
A. Excessive Primary to Secondary Heat Transfer B. Led of Primary to Secondary Heat Transfer C. Loss of 25'F mh% Margin D. Steam GeneratorTube Leak l
.
.
__ _... - _ _ _ _ _ _ _ _ _ _ _ - .. _ _._ __ _ _ . . _ _ _ - . _ _ _ . _ . _ _ _ _ _ - _ . .
__
.
f QUESTION: 052 (1.00) '
Which ONE (1) statement desonbos the reason for manually tripping or venfyng turbine trip in the immediate Actions of ATP 12101 ReedorTrip?
A. Ensures OTSGs are no lon0er cross conneded through the Main Steam lines.
B. Reduces Mein Feedwater flow requirements
.
C. Moimizes steem generator tubebehell delta-T.
D. Prevents an uria &uiled RCS cooldown.
l l
)
l l
'
,
,
QUESTION: 053 (1.00)
cument 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) statement describes required action for these conditions?
A. Select group 7and turn the Single Select switch to ALL.
l'
B. Dispatch 6.1 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. l l
l l
l
,
!
l l
..- - _. - . - - , . . _ _ _ _ _ ,
.
.
-QUESTION: 054 (1.00)
Which ONE (1) condition requires initiabon of HPl cooling according to the Abnormal Transient 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 poig with incore exit timrmocouples indicating 560 degrees.
D. Neither OTSG is available as a heat sink.
L l
i
.
. . - - . - - . - - .-
.
'
,
I QUESTION: 055' (1.00)
Completion of the Containment integrity Checklist is in progress in preparation to begin refueling operations. Which ONE (1) con & tion 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 equipnent hats is open.
l l
I l
__ _ _ . _ . . __ _ _ _ ___._ ______ _______. _ ____.___ ______ _ _____
.
.
QUESTION: 056 (1.00)
A point source in the Amokary Building is reading 500 mrem /hr at 2 feet. Two opbons exist to
complete a mandatory task near this radiabon source:
Ophon 1: Operator X can complete the task in 30 minutes working at a distance of 4 feet from the posit 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.
,
Whdi 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 eact) operator Y and Z to 39 mrem.
.
D. Option 2 will expose each operator Y and Z to 156 mrem.
._. . . ... .. ._ _ . - - - . . . -. -_ ._
.
.
QUESTION: 057 (1.00)
Current plant conditions are:
. - Reactor is operating at 20% power.
- CO-P-1 A and CO P-2A are operating.
- One Main Feedwater Pump is operatog.
- The following alarms are actuated simultaneously:
- A-1-8 Battery 1B Disdarging
- A-2 8 Battery Charger 1B/1DI1F Trouble
- A-38 inverter 1B/1D Inverter System Trouble
- PRF 1-1-1 CRDM Bkr Test Trouble
- H&V A3-2-3 Cont. Bldg. Batt. Chargers B Damper Tbl. Fire-Smoke
- AA-3-2 7 KV Bus Trouble
-
- AA-3-5 480V BOP Bus Trouble Which ONE (1) action is required for this condition?
A. Close DC tie switches to provide an alternate source of DC power.
B. Open suction valve (VA-V-58) for VA-P-1 B.
! 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.
1
4
j
!
!
l l
l
)
._ _- ._.._ _ __ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ . _ _ _ . . . _ - . - .
.
.
QUESTION: 058 (1.00)
An emergency event has been declared. Whidi ONE (1) statement describes the maximum time limits for initial notificabon of the NRC, state, and local agencies?
A 15 minutes for NRC not6 cation,15 minutes for state and local nobficabons B. 30 minutes for NRC noticahon,15 minutes for state and local notdcations C. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for NRC notAcetion,15 minutes for state and local nobfications D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for NRC not6 cation, 30 minutes for state and local nobfications i.
l l
l l
l l
. _ _. . _ _. ___ _ _ . . .
I
-
l QUESTION: 059 (1.00)
Which ONE (1) condition will actuate the CRD PATTERN ASYMMETRIC annunciator?
A Safety rods greater than 7 inches (5%) from group average position as determined by ahare da posebon indcation, B. Safety or regulating rods greater than 9 inches (6.5%) from group average position as determined by relative posebon indcation.
C. Safety or regulating rods greater than 7 inches (5%) from group average position as determined by absolute possbon indicahon D. Regulating rods greater than 9 inches (6.5%) from group average position as determined by relative position indication
-
- - . - -.- .
. _ . _ . _ - _ - . _ . - - .. - . - ___-.. _ - . - . - - .-.. - - - - . -
.
.
QUESTION: 060 (1.00)
Current plant conditions are:
- Reactoris operating at 100% power.
- ICS is in full automatic.
- AT. is ZERO.
Ral=4M Loop A Mein Feedwater Flow instrument calibration is slowly dnfting low - 2%
decrease (linear) over the past hour.
- SASS actuation has not occurred.
Which ONE (1) statement desaibes the compenson between INITIAL and FINAL Feedwater Loop ACTUAL FLOWS at the end of the one hour?
A. Loop A FW ficw 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 irnessed by 2%, Loop B FW flow decreased by 2%.
l
._ -
. ..
.
. _ . . -. . - - - _ . . - - . . .. . - . . - .- - - - . - . - - .
.
.
QUESTION: 061 (1.00)
'
The crew has reduced RCS pressure to minimize subrmoling margin in accordance with ATP !
12145, OTSG Tube Leakage. Which ONE (1) statement describes the reason for minimizing :
i subcooling margin? !
A. Minimize RCS leakage through the leaking OTSG tube. I
B. Minimize time required for cooldown of tlw RCS. i i C. Minimize potential of lifting Main Steam safety valves.
!
D. Minimize tensile stresses on affected OTSG tubes. l
'
I l
.
i A
$
I
I
!
i J i
i J
.1
. _. ___ .. .. _ _ . _ . . . . . . _ . . _ _ . . . __ _ _ _ .____ _ _ _ _ _ _ . . . . _ . _
.
.
QUESTION: 062 (1.00)
Data contained on survey sheet
- 2,500 DPW100 cm2. betagamma
- 10 DPW100 cm8 elpha
- 450 mRommr general ares
- 470 mRomer on the surface of a tank
- Airbome activity < 10% of all DACs l Whis ONE (1) pc.urg is required for these conditions?
,
A. Contaminated /High Radiation Area B. Contaminated / Radiation Area C. High Radiation Area D. Airbome Radiation Area
. _ - _. -. . - . - - - - . - . - . . . - . ..- .-.- ...
<
.
'
QUESTION: 063 (1.00)
', Current plant conditions are:
- Numerous fire alarms are actuated on panels HVB, PLA, and PLB.
. - Fire dampers M+04 and AH-04 have automatically closed.
,
- Air tunnel deluge systems have arewed
- Air tunnel halon system has actuated.
- Aux and Fuel Handing Buildmg 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.
B. Actuate Relay Room CO2 system manually, i
l C. Trip reactor, i D. Start three fire pumps. 1
j
1 4 I
!
< \
l l
.
I l
l
.
.
QUESTION: 064 (1.00)
Auxiliary Operator erroneously irutiates a liquid release from A WECST. The Release Permit is for B WECST. Whid) ONE (1) condtion causes automahc termination d the acx:idental release d rhive liquist A. RM4.-7 in ALERT alarm QS loss of sample flow through RM4.4 B. RM4.4 HIGH alarm Q,8 high tank release rate c. High MDCT emuont flow QB low tank release rate D. RM-L4 ALERT alarm QB,.RM4.-7 ALERT alarm l
l
l l
!
l l
!
_ _ _ . _ _ - - _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ - - _ - _ _ _ _ _ _ - _ _ - - - _ - - - _ _ . , ..
. __ . __ _. . . _ . __ _~ ._ ___ __.-___-._ __
,
.
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. Whkh ONE (1) statement desaibes automatic action (s) initiated by the !
Radiation Monitoring system related to this accidental gaseous release?
A. Trips AH-E-14A/C (8/D)
B. Trips AH-E-10 M AH-E-11 C. Trips AH-E-10 ONLY D. Trips AH-E-11 ONLY i
l l
!
f
(
l l
l
.
- - . -- - . ... . - - - - - ._.-. .. . . - _ - _ _-- . _ _
.
.
QUESTON: 066 (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 prescnbod 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 i
.. . - _ . - . . - .
. -. .. - . - - - - - _ _ . ~ _ _ . . ,
,
i
.
QUESTION: 067- (1.00)
Sequence of events:
- - Reactor is operating at 35% power.
- FW-P-1 A is in operation; FW-P-1B trip has not been reset.
- Steam line rupture ocaJrs upstroom of Main Steam isolation Valve MS-V1 A.
- The reactor and turbine trip.
- Main turbine Stop Valves 1-4 fail to close.
Which ONE (1) statement desenbos why only one OTSG depressurizes as a direct result of the steam line break?
A.- HSPS isolates Main Feedwater to OTSG 1 A.
B. MS-V-1 A and MS-V-1B are stop check valves that prevent back flow from OTSG 1B.
C. Closure of turbine Control Valves 1-4 results in separation of the two steam generators.
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.
l
,
i
l
,
.
' QUESTION: 068 (1.00)
Currerd plant conditions are:
- A roedor trip has occurred.
.
- Three control rods faded to fully insert into the core.
- Power Range Nis are of scalelow.
Whch ONE (1) statement desenbes the required action (s) for this cordtson?
- A. Emergency borate the RCS.
. B. Maintain primary to secondary heat transfer.
C. De energize 1G and 1L 480 voit buses.
D. ' Perform immediate actions of EP 1202-8, CRD Equipment Failure for stuck rods.
l'
!
.
l
'
l
. . .. . . _ - - . = _ . . . _ . . - . . - - . - .
.
.
QUESTION: 069 (1.00)
Current plant conditions are:
- Reactor is operahn0 at 100% power.
- 6 CW Pumps are operahnD
- Wrterconddions exist Which ONE (1) statement desoibes 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 C. Loss of the Gland Steam Exhauster D. Condenser pressure is 7.7 inches Hg absolute
_- . - _ . . . . . . _ . _ _ . . _ . _ _ - _ . . . . _ _ . _ . ~ _ . ._ _ . _ _ _ _ _ _ . _ _ _ . . . _ _ . _ _ _ _ . _ . _ _ . . __
.
.
QUESTION: 070 (1.00)
Current plant conditions are:
- Reactor is operating at 75% power.
- RC-P-1C is shutdown due to high motor winding temperatures.
- Entry into the RB is required to verify the NSCCW valve line up for RC-P-10.
- RB purge must be initiated prior to the entry.
- D-Rings will not be ordered.
Which ONE (1) position is authorized to approve this entry into the RB7 A. Diredor of Operations and Maintenance B. Plant Operations Director C. Shift Supervisor D. Group Rad Con Supervisor
l I
l f
'
.
!
l
. . . . . . . . - . . ._
.,
.
- ..- . .. --- . ..
'
,
i
.
- QUESTION
- 071 (1.00)
Which ONE (1) statement does NOT satisfy the Superheat Determination / Limit Rule as derned in ATP-1210-107
- A. 25*F W superheet as determined by he most oor1servative of the two a Wing i- margin meters on panel PCL B. 25'F of superheet as determined by the plant computer
.
C. 25'F of superhost as determined by the average of 5 highest operable incore
thermocouples and RCS wide rargo pressure 25'F of superheat as determined by the highest operable BIRO incore
'
D.
I thermocouple and RCS wide range pressure
l i
.
)
'
i l
l
l l
k
t
.
QUESTION: 072 (1.00)
Current plant conditions are:
- Refueling operations are in progress.
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.
Which ONE (1) set of statements desenbes 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. Dired the dropped fuel assembly to remain in the basket. Continue refueling operabons using the alternate transfer medumism.
C. Stop all fuel movement. Further handling of any fuel is allowed after obtaining approval from the Director O&M D. Direct the dropped fuel assembly to be transferred out of the RB to the Spent Fuel Pool. Continue refueling operations using the attemate transfer mechanism.
-
- .- - __ _ _. - . - . . . . _ = . - - . . - - - . - - - - . . . - _ _ . - . - . . -
, ,
i
,
.
QUESTION: 073 (1.00)
{
! Instial Plant Cordbons
i - Plant is in the Cold Shutdown condition l - RB purge is in progress.
- A rarGaar4ive liquid release is > aogress from A WECST.
l - RlHA (FH Building exhaust e enstor) interiock control switch is in Defeat for l&C i calibration surveillance.
!
- Twenty gallons of highly rhive liquid is accidentally spilled in the Fuel Handling Building j Which ONE (1) statement desenbos required automatic action (s) related directly to this spill incident?
,
! A. AH-V-1NB/C/D close to terminate RB purge
B. AH-E-14NC (B/D) trip to terminate exhaust ventilation flow.
l C. AH-E-10 and AH-E-11 trip to terminate supply ventilation flow.
l
l D. WDL-V-257 closes to tarminate liquid release.
i
i l
i.
.
i
$
!
!
.
.
.
i
._ _
,
.
QUESTION: 074 (1.00)
Current plant condetions are:
- Tube leak exists on OTSG 1A.
- Plant cooldown is in progress per ATP 1210-5.
- BWST levelis 48 feet.
- Tu is 500*F.
RCS pressure is 900 poig.
< Which ONE (1) statement desenbos the reason why OTSG 1 A should NOT be isolated?
A. Buildup of radioactive water in OTSG 1 A l
B. Extension of time needed to reach Cold Shutdown j
C. Increased probability of lifting Main Steam safety valve D. Isolation of one source of steam to EF-P-1 l
l l
.
.
__ _ _ _ _ _ _ _ _ . _ . _ . _ _ . _ _ _ _ . _ _ _ _ _ . _ _ _ . . . _ _ _ . _ _ _ . - _
,
,
I e
1 l
- ,
l i - QUESTION: 075 (1.00) l i ,
'
Cunont plant conditions are:
!
' - Refueling operations are in progress. l
'
- FTC water level is 24 feet above the fuel assemblies.
- BWST is drained to 5 feet. l l - RCS temperature is 90 degrees. .
- Loop A DHR is in service. I
- Repair work is in progress on OTSG 1 A and OTSG 1B.
Which ONE (1) statement describes Tech Spec requirements for the DHR System during these oordtions? l l
A. One DHR stnng is required to be operable; it is required to be operating.
C. One DHR stnng is required to be ope ebie; it is not required to be operating.
D. Two DHR sinngs are required to be operable; neither is required to be operating.
l
,
l
!
_ ._ .. - _ . . _ . . _ _ . . _ . . _ . _ _ _ . __ . _ _ . __ ._ .
,
,
! .
'
QUESTION: 076 (1.00)
s
! Current plant conditions are:
- Alarge break LOCA has ocx:urred.
- - RCS pressure is 30 peig.
j - Operator actions have been performed tp to this point in occordance with ATPs.
! - BWST levelis 6 R. 4 inches.
<
!
Which ONE (1) statement desenbos the required sequence of operator actions to be performed based upon the conditions described above? l i l A. Open the suction of the DH and BS pumps from the RB Sump (DH-V-6A/B).
- B. Close the suchon of the DH pumps from the NaOH Tank (BS-V-2A/B), then open l
'
DH pump suction from the RB Sump (DH-V-6A/B).
s Open DH and BS pump suchon from the RB Sump (DH-V-6A/B), then close BWST ;
C.
outlet valves (DH-V-5A/B), then close NaOH tank Outlet Valves (BS-V-2A/B). l D. Verify open DH and BS pump suctions from the RB sump (DH-V-6A/B), then close !
BWST outlet valves (DH-V-5A/B), then close NaOH Tank Outlet Valves (03-V-2A/B.
l l
-. .. . - . - .- . . . - - - _ . - - . - - . . - _ . - - - - - . . - - . . - -
.:
,
-
- QUESTION: 077 (1.00)
The reactor is operating at 100% power. Which ONE (1) statement explains the reason for i
.
entering a Tech Spec time clock?
l 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.
j' D. Sodium Hydroxide Storage Tank level is 8 ft. 2 inches lower than BWST level.
,
h
!
!
I k
'
.
e
!
I
.
!
!
<
_- . ..-.. - . . _ - . . _ - . .. . _ - - - _ _ - - - . . _ ~ . - - . _ .
s
!' ,
'
I l QUESTION: 078 (1.00)
!
. Current plant conditions are:
i - Reactor startup (approach to criticality) is in progress.
- - Control rods reach upper ECP limit prior to achieving criticality.
f Which ONE (1) statement desenbos requirements for this cordtion?
A. Commence immodata boration to adhvo 1% Ak/k subentical cxanddion; reador i startup may not be continued until authonzed by Nuclear Engineering l B. Stop rod withdrawal and re dda critcal boron concentration; commence slow ,
controlled dilubon to achieve entcality. i
,
a l 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
!
Engineering to evaluate ECP conditions prior to re-commencing startup.
i l
I i
_ _ .__ _ _ . _ _ . _ . _ _ - . - _ _ _ . . . . . _ .
.
.
QUESTION: 079 (1.00)
,
,
Which ONE (1) conddion will DIRECTLY close the letdown isolation valve, MU-V-37 A. High letdown temperature B. 30# RB pressure ESAS signal
.
ReactorTrip Isolation
'
C.
D. High Makeup domineralizar D/P
i
!
.
J
.
E i
<
_ .. .. ...- - _ . .- .. - .-.- . . - . - . - .
.
.
QUESTION: 080 (1.00)
Which ONE (1) document is NOT required to be reviewed prior to assuming shif t duties as the licensed Shift Foreman?
A. ESAS Checklist B. Locked Valve Book C. TCN/STP Book C. Revision Revie.v Book l
l l
l l
t
- . , -
. - - - . -.- . _ . .-.. . . - . - . - .-. . . . ._ _ .. ._. . - - . _ - . . . - . .
,
<
.,
'
.
QUESTION: 081 (1.00)
i
- Whis ONE (1) statement describes a function of the Technical Support Center (TSC)?
.
A. Perform on-site dose prodction calculations.
-
,
B. Approve omcial press releasos.
l C. Notdy on-site agencies for event to-classdication.
r
.
D. Perform backup RCS leakrate calculations.
.
.
d j
l l
i
!
i.
-
!
!
!
k s
s
.
<
t
!
E
,
J
$
5 .
.
= , , . - - - .- - .,
- - _ . . -. . - -. - - - . - - - . . - .- - - .- -. . . .- - .
.
!
.
f QUESTION: 082 (1.00)
l Which ONE (1) statement explains the reason (s) EFW is injected into the OTSG near the top of the tube bundle?
-
A. Reduce the thermal stress on the lower tubesheet since the upper tubesheet can vnthstand a higher thermal stress then the lower tube sheet.
! B. Reduce the thermal stress on the lower tubesheet and elevate the thermal center
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.
i
,
j l
~. - . _ _ _ . . - _ . _ _ _ _ . . _ . _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ . . _ , _ _ _ _
l- ,
i
! - ,
i QUESTION: 083 (1.00)
!
The OTSG maximum allowable secondary pressure shall be limited to less than when
,
OTSG shell temperature is below .
-
A. 100 poig,100*F
' '
B. 100 psig,200*F C. 200 poig,100*F D. 200 psig,200'F l
.
l l
,
QUESTION: 084 (1.00)
The SBO diesel has the capability to energize 4160V bus through cross-tie breakers 1 start and loading. l via 1 A. Only C or D, manual B. C , D or E, auto C. Only D or E, auto
'
D. C, D or E, manual
,
i l
L
(
l l
l
- ._ . _ . . . _ . . _ _ . . _ _ _ _ _ _ . _ _ . _ . . . _ . _ . _ _ . . _ . . . . _ . _ . . ._._ _ _ _ _ . _
l
'
l i
l
'
,
l QUESTION: 085- (1.00)
'
An RCS sample is being drawn when RM-G-18 goes into HIGH alarm. Which ONE (1) statement
desenbos the automatic actions a==MdM with RM G-18 for this situation.
! A. CA-V-4A and CA-V-5A close.
.
B. CA-V-48 and CA-V-58 close.
C. CA-V-2 and CA-V-13 close.
l D. The Control Buildog ventilation system is shdted to emergency recirculation mode.
$
.
l d '
- .
A i
e
!
.
i
,'
.,
i
i t-f i
.
- . - . . . . _ _ ,
.-- . _ . .-. - .- . _ . .
. _ . - - . - . . . .
.
.
QUESTION: 086 (1.03)
Linear signal falure rates are desabed in sad 1 conddion listed. Which ONE (1) condition will cause an automabc TRANSFER by the Smart Auto Signal Selector (SASS) dump full power operation?
wad RCS nemm range pressure instrument drifts to produce a difference error ;
A.
of 4% over a five minute period.
B. Selected "If OTSG pressure fails high at a rate of 10%/sec.
C. Selected Loop A Foodwater Flow instrument drift produces a difference error of 5%
over 10 minutes.
D. Selected NI Power Range Channel fails low at a rate of 10%/30 secondo.
l
. _ _ . . . - . . . ._ _ . _ _ __ _._- _ _. . _.
t 9
! QUESTION: 087 (1.00)
A reactor trip from 100% power has occurred.
- All RCPs were tripped 60 minutes a00.
Steam flow and feedwater flow have been venfied.
- incore tiwTrcmg:: are tracking Tw. !
Which ONE (1) set of plant conditKms indlCates natural Circulation is Occurring?
. OTSG RCS Cold Leg RCS Hot Leg
. Pressure Temperature Temperature A. 700 psig, decreasing 540 deg F, increasing 550 deg F, stable l
B. 750 psig, stable 513 deg F, stable 545 deg F, decreasing {
l
C. 800 psig, stable 550 deg F, decreasing 570 deg F, increasing I
!
D. 940 psig, increasing 540 deg F, increasing 600 deg F, stable
I
.
,
_ .._ _ _ _ . _ _ _ . . _ . . _ . _ _ . _ . - _ . . . _ . _ _ . _ . . _ _ _ _ _ _ _ . _ _ _ . _ . _ _ _ _ _ . _ .
'
a .
QUESTION: 088 (1.00)
e Which ONE (1) statement describes the function of the PORV NDTT Key Lock Switch on Control Room panel PCR7 j A. In the AUTO position, setpoint is 2450 psig ONLY if RCS temperature is BELOW i 275'F.
B. In the OFF posdion, setpoint is 485 psig at ANY RCS temperature.
I 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 f
i i
'.
i l l
<
l I
l j
_ _ . . . . _ . _ _ _ . _ _ . _ . . . _ _ _ _ _ . _ . . _ _ _ _ . _ . _ . _ . _ . . _ _ _ _ . _ _ _ _ _ . _ . . _ _ _ _ _ _ _ _ _ .. _ . _ _ .. _ ._
-
!
.
)
.
.f QUESTION: 089 (1.00)
Whi& ONE (1) statement desenbos the operation of RC-V-1 Pressurizer Spray Valve?
i
- A. In Manuel control, the JOG circuit drects movement of the valve using nxxnentary j contact pushbuttons; valve position is limited to 40% open poshion.
i i B. In Manual control, he JOG circud drects movement of the valve using momentary
contact pushbuttons; valve can be opened to 100% open position.
i
! C. In Automatic control, valve operation is controlled by RCS pressure conditions; valve is opened to 80% open position in response to open commands.
l
D. In Automatic control, valve operation is controlled by RCS pressure conditions; l- valve is opened to 100% open position in response to open commands.
l
\
l l
- l
- ,
I
i i
J i
i
..
!
!
!
i
i i
i I
!
i e
^
a
, , ,~ __ _ _ ._ . . , . . , . _ , . . . _ . - . . . , = _
. _ - -_
. . . . _ . ._ _ ____.-._....__.._ .._.__ _ ___ _ . . . . . . _ - _
. ,
.
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 groupe B. ICS neutron error signal of-5%
C. Reador power at 56% with Asymrnetric Rod Fault D. Startup rate at +3.2 DPM on one intermediate Range Ni channel l
l
!
l l
-
-. . . . . .. . - . . _ . - - . _ . . - . . . .. _. _. ..- . - . . . - -
, ,
.
. QUESTION: 091 (1.00)
Current plant conditions are:
.
! - Plant is shutdown in preparation for a Refueling Outage.
i - RB purge is in progress for one hour.
'
- RB sump liquid is being gravity drained to the Auxiliary Building sump.
Which ONE (1) statement desuibes required actions for actuation if RM-A-9G HIGH alarm?
A. WDL-V-534 and WDL-V-535 RB sump drain line isolation valves close.
- B. Kidney Filter Fan (AH-E-101) trips.
.
C. RC Drain Tank vent isolation valves (WDG-V-3/4) close.
l D. RCS letdown isolation valves (MU-V-2N28) close.
i l
.
!
,
a W
I i
.
- l l
!
!
_. . . _ _ _ _ . _ . _ _ . . _ _ . _ _ . . . _ . _ . _ . _ _ . . . _ _ _ - _ _ . _ _ _ _ _ _ . _ _ _ _ _ . _ _ ,
t
.
QUESTION: 092 (1.00)
The reactor is operating at 100% power. Which ONE (1) condition causes automatic start of ALL Emergency Feedwater Pumps?
A. FW-P-1 A hydraulic oil pressure at 68 peig. AND loss of two RCPs in RCS Loop A B. Loss of two RCPs in RCS Loop B 6NQ 24 inses OTSG 18 Startup Range level C. RB pressure at 5 poig, QB 9 indes OTSG 18 Startup Range level D. FWP 1B hydraulic oil pressure at 74 psig, QB 24 inches 1B OTSG Startup Range level i
l l
l l
,
. . _ _ . , , _ .
. _ _ _ . . _ _ _ _ _ _ . _ _ _ _ . . _ _ _ _ . . . _ . _ . _ . . _ . . . . _ _ _ _ . - . _ _ . _ _ _ . _ _ _ _
.
i i .
l QUESTION: 093 (1.00)
- Current plant conditions are
i - Reactor is at cold shutdown conddion.
!
- DHR Loop A is operating.
- PARTIAL loss of ICS/NNI Hand Power (DH AUTO SUBFEED) has occurred.
Which 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 l AUTO POWER?
A. Analog meters Fall LOW when instrument power is lost.
f l B. Flow instrument fluctuations (flow noise) STOP when instrument power is lost. )
i
! C. EP 1202-40 Loss of ICS Hand and Auto Power enclosures identify specific i meters affected by loss of DH AUTO POWER. l i :
D. Plastic labels glued onto the control console identify DHR analog meter power l
, supplies. )
- i i
i
.
. _.__ .
.
QUESTION: 094 (1.00)
Current plant conditions are:
- Reactor is operating at 100% power. 1
'
- ICS is in full Automatic.
- Due to relay actuations, a trip of ALL 230 KV LINE Breakers EXCEPT those associated l
!
with the 1001 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,18 Aux !
Transformer remains energized, canying all station loads.
C. Automatic load reduction to 50% due to inadequate line capacity,1 A & 18 Aux Transformers remain energized.
D. Reactor Trip, loss of off-site power.
,
I T
- , . . - _ . - ~
. _ __ . .._._ .. ._._._ _ _ _ _ _ _ _ . . _ . . _ . . _ . - _ _ _ . . _ _ _ . _ . . _ _ _ _ . _ . . _
4
.
QUESTION: 095 (1.00)
,
During refueling operations RM-G4, the Auxiliary Fuel Handling Bridge monitor, becomes inoperable. Which ONE (1) condition akws Refueling Operations to continue?
.
A. Install a suitable portable instrument of comparable range and sensitivity.
. B. Verify RM-G-7 is operable on the a4acont Fuel Handling Bridge.
C. Obtain the approval of the Fuel Handling Supervisor.
! D. Obtain the approval of the Radiological Controls GRCS.-
,
!
$
i
.
>
,
.
.
.
l j'
i
!
)
.
- 4
!
, , _ _ _ . -._
_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ . _ _ . _ _ _ _ _ _ _
-
-_
.
.
QUESTION: 096 (1.00)
Currerd plant conditions are:
- A LOCA has occurred.
- On-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 star has just been indbrmed that the Hydrogen Recombiner reaction chamber temperature has steadily decreased over the last half hour from 1300*F down to 1280*F.
Which ONE (1) statement descnbos the reason for the slowly decreasing Reaction Chamber temperature?
A. RB Pressure decreasing B. RB Pressure is increasing.
C. Hydrogen concentraton is decreasing.
D. Hydrogen concentration is increasing.
.
'
I
.m. - -- . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
-
. - . -- . - - - . . - . . . - - . - - - . - . - - . _ - - - . - . . . . - . - - - . , _ . - . . . . .
.
.
QUESTION: 097 (1.00)
Current plant conditions are:
t - Reactoris tnpped from full power.
- HPI has been manually actuated due to LOCA.
- RCS Sheme gnMargin 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 Whk:h ONE (1) statement desenbos requred actions for this situation?
A. Trip remainog 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%.
l D. Continue operation of both RCPs, start EFW, and raise OTSG levels to 75%-85%. <
'
I
!
.
l l
l l
l.
l
!
!
l l
_ . . . . . _ . -
__
_ _ _ . . _ . . _ _ _ . _ _ _ . _ . . _ _ _ _ _ . _ _ . . . . . _
.
QUESTION: 098 (1.00)
Current plant conditions are:
- Reactoris operating at 90% power
- All RCPs are operatin0-
- High Vibration alarm actuates on RCf-1 A(15 mils and slowly increasing).
- Roset of vibration alarm has failed to clear the alarm.
Whis ONE (1) statement desalbes the required adions required for this situation?
A. Start the oil lift pump, then if vibration does not decrease trip the RCP.
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.
D. Close #1 Seal Leakoff valve (MU-V-33A), within 5 minutes, reduce power to 50% -
75%, and trip the affeded pump within 30 minutes.
l
.
l
. , .-- . - _ . . - - - . - - . . - - - . _ . _ . - _ - _ - . . - - - .
')
'
i
! QUESTION: 099 (1.00)
l Current plant conditions are:
- Randorisinpped.
j - Automatic ESAS aduation has NOT occurred.
Refer to the attached control panel drawing. The control room operator momentarily
depresses the following pushbutions:
- - 2 of 3 Enablemefeat pushbuttons on the 4 PSI Manual actuations (ONCE) for both
,
trains
- All 3 Enable / Defeat pushbuttons on the HPl Manual actuations (ONCE) for both trains i
Which ONE (1) statement describes the status of the Manual ES signa ls to affected plant
- components (pumps, valves, etc.)?
! A. BOTH the Manual 4 PSI and the Manual HPl 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 i DEFEATED for ALL components affected by the HPI ES actuation.
!
'
C. The Manual HPI ES actuation signals ARE DEFEATED for ALL components affected by the HPI actuation - but the Manual 4 PSI actuation signals ARE NOT DEFEATED for ALL components affected by the 4 PSI ES actuation.
,
l . D. BOTH the Manual 4 PSI AND the Manual HPl signals ARE NOT DEFEATED for i ALL affected plant components.
.
l
!
!
, _
-_. . . _ ____ _ _ . _ _ _ _ ._._ _ _ _.___ _ __ _ _ _ _ ._
$
,
.
i
,
a,sm,..
.
.
.
_.__. . _ . ._ . _.__.
.
MANUAL ES ACTUATION 1EOIE RC press I,wna "L'h"M " % .a
- - -
. *
, ,_._
- .
._e-
, P$[
RB PRESS bm II'fk pggy ,ryggy I'b *
.
at
.
'8* * *
._._ - -
PSC PU EAT
-
rsemaa / ENABLE Patmea RB PRESS gpg py ,
.
giant:1 g!gg '
?n" e t. - e - e -- . e.- e aj l
l l
l i
, ,,, , w-w-- - '"
. ____ _ _ _ _ _ . _ - - - - .. - ._. - _ _ . . _ . . . _ . _ . _ . _ _ . _ _ . _ _ _ _ . _ _ . . _ _ . . _ _
i e l
l QUESTION: 100 (1.00)
i Current plant condtions are: 1 l
- Reedoris operahno at 90% power
- EG-Y-1 A is being started for survedience. 1
- Synchroscope indicator is at 12:00 and is not moving in either forward or reverse direchon
- Ir% at time of EG-Y-1 A breaker closure: I
- Generator load is 0.15 MW. l
- Generator reactive load is - 0.3 MVars.
After closing the output breaker, the operator pauses for five somnds to monitor generator output conddsons. Whid1 ONE (1) statement desmbos the status of the output breaker and generator FIVE SECONDS after the breaker was irutially closed?
A. Output breaker is closed with generator load slowly increasing above 0.15 MW.
B. Output breaker is open due to over <:urrent relay trip protection.
,
C. Output breaker is closed with generator output at 0.15 MW.
D. Output breaker is open due to revorse power trip relay protection.
I i
i i
8 I
a a
__ __ _ _ __ _ _ _ _ _ _ . - _ -v.-
. .. . . .
.
. . .
. .
s e
TMI UNIT 1 SRO LICENSING EXAMINATION 8/27/98 ANSWER KEY 1. ...... B 26. ..... A 51. _ C 76.......D y
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 6. ...... B 31. ..... C 56. ..... A 81.. .....D 7....... A 32. ..... D 57. ..... C 82. . . . . . . B 8. ...... C - 33. ..... C 58. ..... C 83. . ... .. C 9. ...... B 34. ... .. B 59. ..... C 84.... ..D i
10. .... C 35.....C 60. ..... A 85... ...C l
11.. .C 36. ..... C 61. ..... A 86.. ....B 12. .... D 37. .....C 62. ..... A 87. . . . . .. B 13. .... C 38. ..... B 63. ..... D 88... ..D 14..... C 39. ..... B 64. ..... B 8 9..... . . B
' 15. .... D 40.....B 65. ..... 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 49. ..... B 74. ..... B 99.. ...C i
'
25. .... D 50. ..... D 7 5. ..... C 100.....D I
_ _ _ _ _
.
_ _ . _ _ . . _ . . _ .. . _ _ _ . _ _ . _ _ _ _. _ -
t e
l Attachment 2 i
SIMULATION FACILITY REPORT Facility Licensee: Three Mile Island Unit 1 Facility Docket No: 50-289 Operating Tests Administered from: Auaust 24-25,1998 This form is used only to report simulator observations. These observations do not
- constitute au fit or inspection findings and are not, without further verification and review, i 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 la required in response to these observations.
No simulator deficiencies, that affected the scenario examinations or JPMs, were identified
,
during the conduct of the examination.
i
$
r D
e *
,