ML20195H922
| ML20195H922 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 11/17/1998 |
| From: | Hurley L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-445-98-301, 50-446-98-301, NUDOCS 9811240150 | |
| Download: ML20195H922 (200) | |
See also: IR 05000445/1998301
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h, - UNITED STATED
[ g NUCLEAR REGULATORY COMMISSION
L E
REGION IV
F
44 [ 611 RYAN PLAZA DRIVE, SUITE 400
,e, ARLINGTON. TEXAS 76011-8064
9*****
, November 17,1998
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NOTE TO: NRC Docum'ont Control Desk !
Mall Stop O-5-D-24
FROM: Laura Hurley,' Licensing Assistant
Operations Bic.,m, Region IV
SUBJECT: OPERATOR LICENSl!!G EXAMINATIONS ADMINISTERED ON
JUNE 19-26,1998, /J COMANCHE PEAK STEAM ELECTRIC STATION,
UNITS 1 AND 2.
DOCKETS #50-445/446
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On June 19-26,1998, Operator Licensing Examinations were administered at tne
referenced facility. Attached you will find the following information for processing
through NUDOCS and distribution to the NRC staff, including the NRC PDR:
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Item #1 - a) Facility submitted outline and the initial exam submittal for distribution !
under R:CS Code A070.
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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.
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If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations
Branch, Region IV at (817) 860-8253.
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- July 10,1998
Mr. C. L. Terry
1TU Electric.
. Senior, Vice President & Principal Nuclear Officer
ATTN: Regulatory Affairs Department
i P O. Box 1002
_ Glen Rose, Texas 76043
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SUBJECT: NRC INSPECTION REPORT 50-445/98-301; 50-446/98-301 i
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Dear Mr. Terry:
From June 19-26,1998, an operator licensing certification inspection was conducted at your
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Comanche Peak Steam Electric Station, Units 1 and 2, reactor facilities. The enclosed report
presents the scope and results of that inspection.
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The inspection included an evaluation of three applicants for a senior reactor operator upgrade i
, license and six applicants for a reactor operator license. We determined that all applicants . 1
satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
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[ While the submitted written examination was adequate and the integrated plant operations
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i portion of the operating test was excellent, the administrative topics and control room systems
and facility walkthrough portions of the operating test of the examination submitted by your
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staff were inadequate for administration, as submitted. Required changes involved revision or
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replacement of the Administrative Section A4 questions of both reactor operator sets,
L substantial modification or change from open to closed reference of 16 of 40 task questions
and replacement of three job performance measures to achieve a cognitive level of difficulty , i
' which would discriminate at the required knowledge level. !
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These revisions were necessary to satisfy the NRC's individual test item expectations,
- consistent with the NRC's Examination Standards. We reviewed our specific concems and
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' recommendations with your staff and believe they now have a better understanding of the
- '.NRC's examination expectations. We commend your staff for their prompt and effective
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' efforts to enhance the examinations with no adverse effect on the examination schedule.
Consequently, no further response to address this matter is required. r
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l In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its
enclosure will be placed in the NRC Public Document Room (PDR).
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TU Electric -2-
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- Should you have any questions concerning this inspection, we will be pleased to discuss them
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Sincerely,
/s/
John L. Pellet, Chief
Operations Branch
- Division of Reactor Safety
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Docket Nos.: 50-445;50-446 '
) License Nos.: NPF-87; NPF-89 I
Enclosure:
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NRC Inspection Report
50-445/98-301;50-446/98-301
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cc w/ enclosure and Attachments 1-2:
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Matt Sunseri, Director
4 Nuclear Training
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TU Electric
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P.O. Box 1002
Glen Rose, Texas' 76043 , j
cc w/ enclosure and Attachment 1 only:
Mr. Roger D. Walker .
TU Electric
Regulatory Affairs Manager
, P.O. Box 1002
,
Glen Rose, Texas 76043
Juanita Ellis
.g . President - CASE
1426 South Polk Street
Dallas, Texas 75224 1
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TU Electric I
Bethesda Licensing
3 Metro Center, Suite 610
Bethesda, Maryland 20814
George L. Edgar, Esq.
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. TU Electric -3-
1800 M. Street, NW
Washington, D.C. 20036
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G. R. Bynog, Program Manager /
Chief Inspector
' Texas Department of Licensing & Regulation
Boiler Division
P.O. Box 12157, Capitol Station
Honorable Dale McPherson
County Judge
P.O. Box 851 1
Glen Rose, Texas 76043
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Texas Radiation Control Program Director
1100 West 49th Street
. John Howard, Director
Environmental and Natural Resources Policy
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Office of the Govemor
P.O. Box 12428
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TU Electric -4
E-Mail report to T. Frye (TJF)
E-Mail report to D. Lange (DJL) ,
E-Mail report to NRR Event Tracking System (IPAS) l
E-Mail report to Document Control Desk (DOCDESK) l
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bec to DCD (IE01)(IE42)'
bec distrib. by RIV w/ enclosure and Attachment 1 only:
Regional Administrator Resident inspector (2)
DRS Director DRS Deputy Director
DRP Director DRS-PSB
Branch Chief (DRP/A) MIS System
Project Engineer (DRP/A) RIV File
Branch Chief (DRP/TSS)
bec w/ enclosure and Attachment 1-2:
R. Gallo (HOLB/NRR)(MS:9D4)
L. Hurley
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DOCUMENT NAME: R:\_CPSES\CP301rp.mem
To receive copy of document. Indicate in box:"C" = Copy Wthout enclosures "E" = Copy Wth enclosures "N" = No copy
RIV:SRE:OB E C:OB C:PBA C:OB
MEMurphy/Imb JLPellet JITapia JLPellet
07/08/98 07/09/98 07/09/98 07/09/98
OFFICIAL RECORD COPY
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g_NCLOSURE j
U.s. NUCLEAR REGULATORY COMMISSION
REGION IV
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Docket Nos.: 50-445;50-446
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Report No.: -50-445/98-301;50-446/98-301
Licensee: TU Electric
Facility: Comanche Peak Steam Electric Station, Units 1 and 2 '
Location: FM-56 i
Glen Rose, Texas !
Dates: June 19-26,1998 i
inspectors: Michael E. Murphy, Chief Examiner
Steve L. McCrory, Senior Reactor Engineer, Examiner / inspector
Ryan E. Lantz, Reactor Engineer, Examiner / Inspector ,
John L. Pellet, Chief, Operations Branch
Accompanying i
Personnel: Lawrence Vick, Reactor Engineer, Operator Licensing and Human
Factors Branch, Office of Nuclear Reactor Regulation j
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Approved By: John L. Pellet, Chief, Operations Branch
Division of Reactor Safety
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ATTACHMENTS:
Attachment 1: SupplementalInformation
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Attachment 2: Final Written Examinations and Answer Keys
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EXECUTIVE SUMMARY
Comanche Peak Steam Electric Station, Units 1 and 2
NRC Inspection Report 50-445/98-301; 50-446/98-301
NRC examiners evaluated the competency of three senior operator and six reactor operator
license applicants for issuance of operating licenses at the Comanche Peak Steam Electric
Station. The licensee developed the initial license examinations using NUREG-1021, Interim
Revision 8, January 1997. NRC examiners reviewed, and approved the examinations. The
initial written examinations were administered to all nine applicants on June 19,1998, by
facility proctors in accordance with the guidance in NUREG-1021, Interim Revision 8. The
NRC examiners administered the operating tests on June 22-25,1998.
Operations
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All nine (six reactor operators and three senior operators) license applicants passed
their examinations. The applicants exhibited good oversight, peer checking and
effective communications (Sections 04.1, 04.2).
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The administrative topics and control room systems and facility walk-through test
materials were inadequate for administration as submitted (Section 05.1).
. The licensee's staff was highly responsive to replacement and revision
recommendations developed during the review process. No significant changes to
examination materials were required as a result of administration (Section 05.1).
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- Report Details
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' Summary of Plant Status
The plant operated at essentie'if ')0
t percent power on both units for the duration of this
- inspection.
l. Operations
04' Operator Knowicdga and Performance -
04.1 Initial Written Examination j
a. Insoection Scope
On June 19,1998, the facility licensee proctored the administration of the written
examinations approved by the NRC to six individuals who had applied for initial reactor
operator licenses, and three individuals who had applied for initial upgrade senior
operator licenses. The licensee proposed grades for the written examinations and ,
evaluated the results for question validity and generic weaknesses. The examiners !
revlewed the licensee's results.
b. Observations and Findinas
The minimum passing score was 80 percent. The scores for the written examination
ranged from 87 to 98 percent. The overall average score was 93 percent. The
licensee's post-administration analysis identified that Questions 4 and 23, common to ,
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both the reactor operator and senior operator examinations, were missed by more than 1
50 percent of the applicants. Analysis indicated this was due to isolated knowledge l
weaknesses. No broad training or knowledge weaknesses were identified during
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review of applicant performance on the administered examinations. There were no
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post ' examination comments or changes to the written examination.
c. Conclusions
All nine applicants passed the written examinations. No broad knowledge or training
weaknesses were identified as a result of evaluation of the graded examinations.
04.2 initial Ooeratina Test
a. Inspectiori Scope
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The examination team administered the various portions of the operating examination l
to the nine applicants on June 22-25,1998. Each applicant participated in two dynamic
simulator scenarios. Each reactor operator applicant received a walkthrough test,
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which consisted of ten system and four administrative areas. The upgrade senior
reactor operator applicants were tested in five system and four administrative areas.
b. Observations and Findinas ;
All applicants passed all portions of the operating test. Overall, the applicants
performed well in the dynamic simulator scenarios with good oversight, peer checking,
and effective communications noted by the examiners. The applicants displayed good
knowledge of technical specifications and facility abnormal and emergency procedures.
However, during one scenario, the applicants in the position of control room supervisor
and reactor operator failed to monitor reactor plant pressure, resulting in automatic
actuation of protection systems. This was considered to be a weakness in observation
of plant conditions but was limited to the single cited instance. The plant was
maintained within its design envelope.
The applicants performed well on the walkthrough and administrative sections of the
examination.
c. Conclusions
All nine applicants passed the operating tests. The applicants exhibited good
oversight, peer checking, and effective communications.
05 Operator Training and Qualification
O5.1 Initial Licensino Examination Development
The facility licensee developed the initial licensing examination in accordance with
guidance provided in NUREG-1021, Interim Revision 8, " Operating Licensing
Examination Standards, For Power Reactors, January 1997."
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05.1.1 Examination Outline
a. Inspection Scope
The facility licensee submitted the initial examination outlines on March 10,1998. The
chief examiner reviewed the submittal against the requirements of NUREG-1021,
Interim Revision 8.
b. Observations and Findinas
The chief examiner determined that the initial examination outlines satisfied NRC
requirements. The chief examiner advised the licensee to enhance the simulator
scenarios by replacing some component and instrument failures to add discriminatory
value and rea! ism within each scenario from one event to the other.
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! c. Conclusions I
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The licensee submitted an adequate examination outline.
i' 05.1.2 Examination Packaae
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{ a. Inspection Scope l
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The draft written examination was transmitted by the licensee to the NRC on April 13,
1998. The licensee submitted the completed final examination package on June 4,
- 1998. The chief examiner reviewed the submittals against the requirements of
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NUREG-1021, Interim Revision 8.
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The draft written examination contained 125 questions,75 of which were designated to
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be included on both reactor operator and senior reactor operator examinations. All of
the questions were developed for this examination. The draft examination was
i considered technically valid, to discriminate at the proper level, and responsive to the
outline submitted by the licensee on March 10,1998. However, the chief examiner
provided enhancement suggestions for approximately one-third of the questions. The l
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suggestions generally related to clarity of the question and stem, not soliciting a single '
i answer, inadvertent cues, and distractor plausibility. After discussion of the chief
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examiner's suggestions, the licensee modified the examinations as agreed. The chief
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examiner concurred with the resolution of the suggestions and the final product.
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i The licensee submitted two dynamic scenarios and one backup scenario, which was "
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not used during the examination. The chief examiner made suggestions to enhance
j the examination quality by replacing some component and instrument failures to better
! discriminate applicant performance. Other comments, which the licensee incorporated,
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included editorial and enhancements to facilitate administration, such as adding more
i detailed expected actions, primarily for the unit supervisor. The licensee initiated minor
! additional editorial enhancements to facilitate the time-line running of the scenarios
! during the preparation week onsite on June 10-11,1998.
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To support the systems portion of the operating test,25 job performance measures
l with 2 followup questions each were submitted. The chief examiner provided
! comments concerning enhancement of the test, which were incorporated. The chief
j- examiner challenged the critical step assignments for some of the tasks and the
k licensee revised these critical step assignments. Also, the licensee revised or replaced
approximately 9 questions in response to the chief examiner's enhancement )
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suggestions. However, during the examination week pre-administration review, it was
determined that further changes were necessary. This review identified that 16 of the
40 questions required either substantial modification or change from open to closed
reference. Further, there were three job performance measures that were identified as
nondiscriminatory and required replacement. These items were reviewed with the
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- licensee and after agreement on the needed revisions, the licensee promptly instituted
- the changes.
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The licensee submitted two sets of administrative tasks and questions to cover the
administrative section of the operating test for the reactor operator applicants and one ,
set of administrative tasks for the senior operator upgrade applicants. To facilitate j
administration, some minor changes were made to some administrative tasks during i
the review process. However, during the examination week pre-administration review,
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it was determined that the Section A4 questions of both reactor operator sets required j
revision or replacement because they were too easy or not related to the subject area l
of amergency preparedness. These items were also reviewed with the licensee and
after agreement on the needed revisions, the licensee promptly instituted the changes.
c. Conclusions
The administrative topics and control room systems and facility walkthrough test
materials were considered inadequate for administration as submitted. However, the
licensee's staff was highly responsive to replacement and revision recommendations
developed during the week of administration, outside the normal review process. No
significant changes to examination materials were required as a result of
administration.
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05.1.3 Licensina Conditions
a. Inspection Scope
The chief examiner verified that the applicants had conducted the proper number of
reactivity manipulations for licensed operator qualification and that the scope of the
manipulations were adequate in accordance with NUREG-1021, Interim Revision 8.
b. Observations and Findinas
The chief examiner verified that the facility licensee properly identified the required five
significant reactivity manipulations on the reactor operator final application. The chief
examiner also verified that the facility had properly documented these manipulations
and that they were significant in accordance with NRC Information Notice 97-67.
c. Conclusions
The facility's program was adequate to ensure that initial applicants for reactor operator
licenses met licensing conditions for performance of significant reactivity manipulations.
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O5.2 Simulation Facility Performance
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a. Insoection Scope I
The examiners observed simulator performance with regard to fidelity durir,g the
examination validation and administration.
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b. Observations and Findinas )
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The simulator performance was good. No fidelity problems were noted. The licensee's l
simulator support staff was very efficient and greatly enhanced the examination l
schedule. Tum around times between scenarios and job performance measures were
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very fast. This eliminated dead time and helped ease applicant stress levels. l
c. Conclusions l
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The simulator and simulator staff supported the examinations well. No fidelity issues
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- V. Manaaement Meetinas
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X1 Exit Meeting Summary
3 The examiners presented the inspection results to members of the licensee
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management at the conclusion of the inspection on June 25,1998. The licensee
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acknowledged the findings presented.
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The licensee did not identify as prnprietary any information or materia!s examined
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ATTACHMENT 1
SUPPLEMENTAL INFORMATION
PARTIAL LIST OF PERSONS CONTACTED
Licensee
M. Blevins, Vice President, Nuclear Operations
S. Falley, Training Supervisor
W. Guidemond, Shift Operations Manager
P. Presby, Training instructor
C. Rice, Licensed Operator instructor
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M. Sunseri, Nuclear Training Manager
NRC
H. Freeman, Resident inspector
L. Vick, Reactor Engineer, Operating Licensing Branch, Office of Nuclear Reactor Regulation
INSPECTION PROCEDURES USED
NUREG-1021, NUREG-1021, Interim Revision 8, " Operating Licensing Examination
Standards, For Power Reactors, January 1997"
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ATTACHMENT 2
FINAL WRITTEN EXAMINATIONS AND ANSWER KEYS
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p U.S. Nuclear Regulatory Commission
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Site-Specific ,
' Written Examination-
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Applicant Information
Name: Region: IV
Date: June 19. 1998 Facility / Unit: Comanche Peak 1/2
License Level: R0 Reactor Type: W
Start Time: Finish Time:
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Instructions i
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 j
hours after the examination starts. :
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Applicant Certification i
All work done on this examination is my own. I have neither given nor
received aid.
Applicant's Signature
Results
Examination Value Points
Applicant's Score Points
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Applicant's Grade Percent
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i- 1. A severe weather condition has caused structural damage to switchyard transmission equipment requiring both
units to operate on Emergency Diesel power for an exteeded period. Unit I is operating on a single diesel
generator due to mechanical problems with the remaining diesel.
Assuming Unit 1 operated within Technical Specification guidelines before the incident, which one of the below
- correctly describes the minimum design capacity.of the diesel fuel oil system for a single Unit I diesel generator
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under these conditions?
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a. Operate for seven days at continucus load rating
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b. . Operate for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at maximum overload rating
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c. Operate for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at maximum overload rating
d. Operate for one day at continuous load rating
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2,z in accordance with STA-656," Radiation Work Control", which ONE of the following statements is correct
regarding logging out of the RP computer system when preparing to leave the RCA?
a. When logging out of the RP computer, input the areas entered while in the RCA
.- b., .When logging out of the RP computer, specify both the RWP/ GAP and Task numbers
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c. When logging out of the RP computer, insert your hand into the hand-reader for verification
d. When logging out of the RP computer, slide your RCA Access Card through the bar code reader
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3; Which ONE of the below correctly describes the reason why the Containment Purge Air Supply and Exhaust
valves are required to be locked and administratively controlled during plant operations?
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a.~- To prevent a spurious Containment ventilation isolation
b. To prev.ent the' admission of unfiltered air into Containment ,
c. The valve actuators do not have penetration conductor overcurrent devices
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'd. - The valves are not guaranteed to close during a LOCA or steam line break accident -
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4. A liquid radioactive discharge is scheduled for your shift. Under these conditions, which ONE of the below is
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correct regarding the operation of the Liquid Radwaste to Circulating Water Monitor X-RE-52537
a. ' It can be monitored in the Control Room on the RM-23 during the release
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b. . It cannot be monitored in the Control Room if both the PC-lls are inoperable ,
c. It uses a Geiger-Mueller detecting element in a lead-shielded sample chamber
d. It _will automatically close X-RV 5253 at the ALERT setpoint to terminate the release
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' 5 L While operating at steady state conditions at 75% power, the Unit 2 Reactor Operator reports that control rods a
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stepping outward in AUTOMATIC. The operator places the rods in MANUAL and rod motion stops.
- Which one of the following is a possible cause of this rod motion?
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. a.. A Loop T-cold fails high ,
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- b. NIS channel N-42 fails high
, c. . First Stage Pressure transmitter, PT-505, fails high
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d. An N16 channel fails high
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W Reactor power is being maintained at approximately 6% prior to placing the Main Turbine on-line. Intermediate
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Range (IR) channel N-35 is in the TRIP BYPASS posiden, with all required bistables tripped for troubleshootmg !
due a failure which occurredjust after entering Mode 1.
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Which one of the following will occur if the N-35 Control Power fuses were to blow at this time?
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a. An overhead annunciator for IR detector high voltage will occur
, b. The reactor will immediately trip on IR high
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c. An overhead annunciator for IR detector compensating voltage will occur
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d. Both Source Range instruments will automatic'ally re-energize
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7.' Safety injection pump train A has been tagged out for motor bearing replacement. A Safety Injection
subsequently occurs due to Large Break Loss of Coolant Accident (LBLOCA) inside containment. Two hours
later, the train B SI pump fails. All cther equipment functions as designed for the duration of the accident,
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.Which one oT he following de:cribes how the loss of both Si pumps will affect the ability of the crew to mitigate
. . the effects of this accident?
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a. A transition to EOS-1.1, Safety injection Termination, will be required upon receipt of the RWST 'ow-low
level alarm
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b. - Both trains of RHR will remain aligned for cold leg injection during the alignment in EOS- 1.3, Trans'er to
i Cold Leg Recirculation
{ c. . Actions will be necessary to restore at least one S1 Pump to service in order to achieve hot leg injection pe
EOS-1.4, Transfei to HL Recirculation
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- d. EOS-1.'4, Transfer to HL Recirculation, provides for aligning one CCP for hot leg injection when neither SI
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Pump is operable -
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8. - An automatic reactor trip and safety injection has occurred on Unit 2 as a result oflowering RCS pressure. The
operators note the following conditions:
- Pressurizer pressure dropping prior to and following the SI
e - RCS average temperature stable prior to and following the SI
e Pressurizer level rising prior to the SI and rising following the SI .
Initially which ONE of the following accidents would result in these conditions?
a. Steamline break
b. Double-ended hot leg break
c. Stuck open pressurizer safety valve
d.' . 4' inch break on a RCS cold leg
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9. A reactor trip has occurred on Unit 2 due to a loss ofoffsite power. Th- crew has completed most of the actions
of EOS 0.1, Reactor Trip Response, and are verifying natural circulation flow. When adjusting steam dumping
rate to control natural circulation, the operators also adjust AFW flow to all of the SGs.
Which ONE of the following correctly explains why narrow range level is re-esteblished in all SGs?
. .
.
a. To maintain symmetric cooling of the RCS
b. To flood all SGs for subsequent entry into Mode 5
c. SG wide range level indication is lost on loss of offsite power
d. Top of SG tubes on all SGs must be covered for natural circulation to occur
.
-- . -- . ~ . .. ~ . . . - _ . . - . - . - . . - . - - . . - - - - - . . . . ._. - .~,- -.
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10. During operation at power with the Reactor Trip Breakers (RTBs) closed a loss of 125 VDC to one of the RTBs
.
,
occurs. '
'
q Which one of the following correctly describes how the rea : tor trip breaker (RTB) will be affected by loss of the ;
' 125 VDC power?
l
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. .
. \
a. It trips open due to loss of power '.o the shunt coii '
b. It trips open' due to less of power to the undervoltage coil
c. It is not capable of tripping on a shunt trip
d. It is not capable of tripping on an undervoltage trip _ l
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11. In the event of a Steam Generator Tube Rupture (SGTR), the assumption is made that operators will isolate
~ Auxiliary Feed Water (AFW) flow to the a'Tected SG within 10 minutes.
Which one of the following is the basis for the action and time limit?
,
. a. -
To coniserve CST inventory ,
b. . Minimize the probability of SG overfill
.
cc To maximize the time the AFW pumps are on recirculation
d. To limit the cooling effect on the SG so steamline pressure remains high '
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12. During operation at power steam generator tube leakage is detected and estimated at 250 gpm by the reactor
4
operator. The fe' lowing plant indications existed at that time: '
,
- RCS pressure - 2200 p' sig and lowering
n Reactor Power- 80%
SG Pressures- 1000 psig
,
- PZR Level-42% and lowering
The unit is tripped end plant parameters following the trip are:
RCS pressure- 1700 psig and lowering
Reactor Power-0%
SG Pressures- 1100 psig
PZR Level- 13% i
,
' Based on the two sets of given data, which ONE of the below describes the effect on primary-to-secondary
leakage? .
I
Leakage following the trip is
l
a. one half of the initial leak rate or about 125 gpm.
b. essentially equal to the initial leak rate or about 250 gpm.
c. approximately 70% of the initial leak rate or about 175 gpm.
d. ' One third of the initialleak rate or about 83 gpm.
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13. During clearance or valve positioning activities, which ONE of the following conditions would allow independent
verification of equipment status to be waived in accordance with STA-694," Station Verification Activities"?
. a.- A clearance requires removal of a gag on a Main Steam Safety valve
b. The valve verification would result in radiation exposure of 120 mrem
c. ' A clearance requires installation of a grounding strap on a non-safety related 480V breaker
d. The valve verification requires entry into containment during fuel movement and would result in a
radiation exposure of 5 mrem -
.
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.
.
14, Unit I is conducting refueling activities with the core being offloaded to the spent fuel pool for determination of
suspected leaking assemblies. All Fuel Handling Building systems are operational and correctly aligned for
refueling operations.
l During movement of one of the assemblies, the Fuel Building radiation monitor, X-RE-6272, alarms. Which ,
4
-. ONE of the following INITI A.L operator actions is required in this situation? ,
4
a. Evacuate the Fuel Building
i
b.
-
Start Pre-Access Filtration System
- c. Ensure Containment Ventilation Isolation occurs
l-: d. . Verify Fuel Building Ventilation System automatically shifts to the Isolate mode
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15. Unit 2 is currently operating at 100% when a control rod in bank C partially drops. A QPTRjust performed by
the RO indicates a Quadrant Power Tilt Ratio (QPTR) of 1.07.
- Usi ng the following statement from Technical Specification (3/4.2.4), determine the course of action the operators i
mv4t take given this scenario. '
"Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each I e of
QUADRANT POWER TILT RATIO in excess of 1.' i
a. Reduce thermal power to 97%
b. Reduce thermal' power to 89%
-
,
c. - Reduce thermal power to 79%
' d.' Reduce thermal power to 71% .
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. . _ _ . . . ._. _ . . . _ - . _ ._ _.._._ _ _ . _ _ . . _ _ . __ _ , . _ . _ .
16. Regarding Technical Specification SAFETY LIMITS, which ONE of the following core limitations does the OT
N16 reactor trip prevent exceeding?
!
a. Power Density (KW/A) ,
,
, . b. Departure from Nucleate Boiling (DNB) . ,
4 c. Total Core Power (NSSS Power Limit) . 4
j d. Axial Flux Difference (AFD)
.
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17. 'Which ONE of the following lists of personnel satisfies the requirement for the Fire Brigade complement in
accordance with STA-727," Fire Brigade"?
a. One Fire Brigade Leader (Reactor Operator); 2 Maintenance Mechanics (Nozzleman); 2 Maintenance
Electricians (Hoseman).
'
. b.. 'The Shift Manager (Fire Brigade Leader); 2 Plant Equipment Operators (Nozzleman); 1 Security
Personnel (Hoseman).
c. One Fire Brigade Leader; 3 Plant Equipment Operators (Nozzleman/Hoseman); 1 Safety Services
(Hoseman).
d. One Fire Brigade Leader; 2 Security Personnel (Nozzleman); 4 Maintenance Mechanics (Hoseman).
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18. An event has occurred on Unit I and operators are conducting EOP-l.0A, '.' Loss of Reactor or Secondary -
.
~ Coolant", when the below parameters are observed:
.
All SG pressures- 800 psig and stable
1 All SG levels - being controlled at 10% NR
'~.
PZR level- off-scale. low
Containment Pressure- 16 psig
RWST level-40%
. RCS pressure- 180 psig and stable
. Based on these indications, Which ONE of the following procedures would the operators enter next to mitigate the
event in progress?
al EOS-l.2A," Post-LOCA Cooldown and Depressurization"
l
. b. -
- EOS-1.1 A, "SI Termination"
c. ECA-l.l A, " Loss of Emergency Coolant Recirculation"
i
' ~ d. EOS-1.3 A, " Transfer to Cold Leg Recirculation" !
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19. A sontrol rod in Control Bank D partially drops into the core with the below conditions: I
!
-
Reactor power 60%
-
Tave 0.5 'F below Tref - 1
-
Control Bank D group demand counters at 180 l
-
. \
After the dropped rod, the unit stabilizes at the following conditions: '
!
-
Reactor power 60%
,
-
Tave 2.5 F below Tref .
- - Control Bank D group demand counters at 180 .
1 Which ONE of the below is correct regarding the effects on the Cold Leg temperature (Tc) and Shutdown Margin
(SDM) from the onset of the event?
a. Both Tc and SDM have decreased
b. Both Tc and SDM have remained the same
c. Te has decreased and SDM has remained the same
d. Tc has remained the same and the SDM has decreased
,
2
..- , . - - . _ . . - . . _ . . _ . . _ _ - . - . - - . - - . - . _ . - - - . - . . , . -
'
20. A small break LOCA has occurreu nd operators have implemented EOP-0.0A," Reactor Trip or Safety
Injection". In this situation, which ONE of the below statements indicates the basis for tripping the RCPs if
minimum RCS subcooling is lost and SI flow has been established? ,
!
, a. Prevent excessive depletion of RCS inventory through a small break leading to severe core uncovery if the.
RCPs were later tripped.
b. Prevent damage to the RCP and RCP seal package due to possibility of two-phase flow in RCS.
c. Prevent physical damage to RCPs and RCS due to stresses associated with pumping a two-phase mixture.
d. To further decrease RCS pressure, enhancing ESF systems ability to inject borated liquid into RCS.
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21. Which ONE of the following is an operational implication ofmaintaining Rod Insertion Limits?
"
- Maintaining Rod Insertion Limits ensures
a. rod tip fretting is minimized."
.
.
b. proper bank overlap is maintained."
'
c, effects of rod drops are minimized." '
d. minimum shutdown margin is maintained."
.
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22. Operators are conducting a plant cooldown for a refueling outage and have reached plant conditions required to
allow blocking of the low PZR pressure Safety Injection (SI) signal.
I
During the subsequent heatup and startup of the plant following the outage, which ONE of the below will unblock j
the automatic low PZR pressure SI signal? i
.
l
a. When 2 out of 3 PZR pressure channels are greater than the P-11 setpoint of 1960 psig !
b. When 3 out of 4 PZR pressure channels are greater than the auto Si setpoint of 1820 psig
l
c. When the control room operator manually unbiccks the signal as directed in the heatup/startup procedure
d. When BOTH reactor trip breakers are closed, removing the Si blocking feature provided by the P-4 I
interlock
,
.
.,
3
- 23. Which'ONE of the following specifies the minimum number of core exit thermocouples (CETs) which must be
operable per Technical Specification Table 3.3-67
, ,
a. 2 per quadrant
b. .4 per quadrant
.
- c. '6 per quadrant -
' d. 8 per quadrant
. -
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24. Unit 2 is in Mode 3 at 375 'F when the reactor operator observes a PZR PORV open with RCS pressure dropping
rapidly. Subsequent investigation reveals wide range temperature instrument TE-413A, HL 1 WR TEMP failed
low.
Which ONE of the below accurately describes the response of the Low Temperature Overpressure Protection
(LTOP) system to the observed indication.s?
a. The LTOP system is operating correctly; the average of four loop temperatures has been reduced low
enough to cause arming and opening of the train associated PORV
b. The LTOP system is not operating correctly; the failed temperature input should have reduced the average
'
temperature input to the associated train and redundant backup to the opposite train causing both PORVs
to open
c. The LTOP system is operating correctly; the auctioneered low temperature plus exceeding the calculated
equivalent pressure should have opened the train associated PORV
d. The LTOP system is not operating correctly; no PORV should have opened because no input was
received from the redundant train temperature instrument
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- _ . -. . . - - ._-. .- .- . - . .
25. Unit 1 is operating with the below piant conditions:
4- - Reactor power 99%
- PZR level 60%
- Letdown flow 75 gpm .
, - The PD pump is in service
- All controls are in automatic
, A 40 gpm charging line leak exists outside containment.
Assuming NO operator action is taken, which ONE of the following identifies the potential consequence of the
event: PZR level drops to 17%, letdown isolates and PZR heaters tum off; and the
4
a. Reactor trips on high PZR level
l
- 4 b. Reactor trips on high PZR pressure I
'
c. PZR level maintains at 60%, no reactor trip
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d. PZR pressure maintains at 2235 psig, no reactor tnp '
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. ... . - - -
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26. An event has occurred which has actuated the Containment Spray system. The following plant conditions are
observed:
-
Containment Spray discharge flow is 900 gpm
-
Containment Spray Sump Suction valves are CLOSED
,
-
Containment Spray Recirculation Valves are OPEN
-
Containment Spray 11eader Isolation valves are OPEN
'
Which ONE of the below describes the operation of the Containment Spray system based on these observations?
a. Not operating correctly; the Recirculation valves should not be OPEN until-1200 gpm
b. Not operating correctly; the Recirculation valves should not be OPEN with sump suction valves
.
CLOSED
c. Not operating correctly; the Recirculation valves should be CLOSED with the Header Isolation valves
'
OPEN
d. Operating correctly; valve interlocks have been met and discharge flow is less than setpoint
1
. . . _ . -... - -, ... .. . ._. .. .. . - . . . - . . . . . . . . . . . .
27. A design basis LOCA has occurred on Unit 2. Assume all equipment operates as designed.
Determine which ONE of the below is the approximate time for RWST level to decrease to the cold leg
recirculation transfer criteria level.
, a. 15 minutes ., ,
i
b. ' 30 minutes !
c. ! 40 minutes I
d. 50 minutes
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28. During performance of EOS-L3, " Transfer to Cold Leb Recirculation", operators are attempting to open valves
' 8804A and B, RHR Discharge to Safety injection / Charging Pump suction.
Considering each answer separately, which ONE of the below should be considered for the interlock logic for this
situation?
'
a. RHR suction from RCS isoIr4ed
b. - RHR suction from RWSTisolated ,
c. RHR suction from Containment sump isolated
d. - Safety injection pump suction from RWST isolated
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29. Which ONE of the below automatic reactor trip signals is described by the following statement taken from the ,
. BASES of Technical Specifications, Section 2.0, Limiting Safety System Settings?
" Prevents water relief through the pressurizer safety valves."
a. PZR Level High ,
b. PZR Pressure Low
c. .> SG Water Level Low
d. . Source Range Flux High '
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30. Unit 2 is conducting a plant shutdown with reactor power currently at 15%. As directed by the procedure in
effect, the RO selects the steam pressure mode of operation. Unknown to the operator, 'he controller output has
failed to 100%.
Which ONE of the below plant responses will occur if no operator action i; taken?
~
a.. All steam dumps will arm, but remain closed. Reactor power will not be affected.
b. All steam dumps will open until Tave reaches 5'F above Tref. Reactor power will not be affected. -
l
c. All steam dumps will remain closed until steam pressure reaches 1160 psig. All steam dumps will cycle at !
.1160 psig and reactor power will decrease. '
d. All steam dumps will open and reactor power will rise. When Tave reaches 553 F all dump valves will
clo'se and reactor power will decrease. l
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31. Unit 1 is' operating at 100% power with the below conditions:
-
PZR pressure control system is in automatic maintaining 2235 psig
-
PT-455 is selected as the pressure input to the PZR pressure master controller
r..
-
PT-455 has failed full scale high
w - NO operator action is taken for the event .
.-
Which ONE of the following describes the initial system response to this failure?
, a. PZR PORV PCV-455A opens and PZR PORV PCV-456 remains closed
,
b. - PZR PORV PCV-455A remains closed and PZR PORV PCV-456 opens
4
' d. PZR PORV PCV-455A and PZR PORV PCV-456 both open j
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32.~ Which of the following is an indication of vortexing at the suction of the RHR pump during reduced inventory
conditions?
,
a. Erratic pump amps
b. RHR suction reliefJifting
c. Decreasing RCS temperature
d. Constant pump discharge pressure
-
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33. Unit 2 is operating at full power when a total loss of Main Feedwater occurs coincident with a failure of the
reactor to trip (ATWS).
' For this condition, what is the bases for the operator action to trip the main turbine within 30 seconds of the
event?
'
a. Prevent a safety injection on low steam line pressure ;
b. Initiate an alternate means of reactor trip from a turbine trip j
c. Conserve remaining steam generator inventory
d. Initiate an alternate means of reactor trip from high RCS pressure
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34. Unit I control room operators are investigating a potential malfunction of the Reactor Coolant Makeup system
based on abnormal indications following an RCS makeup evolution. The RO suspects the pot setting for 1-FK-
110, Boric Acid Blender Flow Control, may have been set wrong.
Given RCS boron concentration is to be maintained at 500 ppm, with BAST boron concentration at 7000 ppm and
, ' a blended flowrate of 127 gpm, determine the proper, setting for 1-FK-110.
,
a. 2.27
- b. . 2.85 <
c. 8.55
I d.'- 9.07
.
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35.' A safety injection has occurred coincident with a :oss of off-site power, Bus leal has decaergized due to an 86-1 -;
lockout. i)etermine from the table below the high head pumps that would be available.
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PDP- CCP1 CCP2 )
,
,
a. ON ON OFF
b. ON OFF OFF
c. OFF ON ON
I
, d. OFF OFF ON
.
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. .
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36. Unit 1 is operating with the following conditions: I
b,
--
90% power
-
Both Main Feedwater pumps in service
-
Both Condensate pumps in service
-
Main Turbine. controls have been shifted to MHC
-
All other control sysems in normal automatic alignment
k
' - Under these conditions, which ONE of the below is expected to occur if condensate pump 1-01 were to trip and
. no operator actions were taken?
3 a. Main Feedwater Pump 1-01 will trip and unit load will decrease to 60% at 35%/ minute
b. Both Main Feedwater Pumps will trip and the reactor will trip on SG NR level at <25%
, c.
. .
Unit load will decrease to 60% at 35%/ minute and both Main Feedwater pumps will continue to operate
d.
Main Feedwater pump 1-01 will trip and unit load will remaia at 90% with SG level remaining on
program
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37. Which ONE of the following conditions would cause loss of Component Cooling Water (CCW) flow to the
Ventilation Chillers?
a. A Reactor Trip signal
i
b. A Containment Isolation Phase B signal
!
- - c. A Containment Ventilation Isolation signal
d. A Containment Isolation Phase A signal
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38. A large steamline break has occurred inside Containment, Operators have been unable to close any .MSIV. The
following plant conditions exist:
- All SG levels:- <5% NR
- All SG pressures: 800# and decreasing
." Operators bas e been directed by the ERGS in progress to maintain a minimum AFW flow of 100 gpm to each SG.
i Which ONE of the following is the basis and operational implication of performing this action?
'
- a. . prevents steam p ,. erator tube dryout
b. maintains a verifiable RCS cooldown rate
!
c. maintains steam pressure above safety injection setpoint
- d. ensures steam generator levels will remain above 5% narrow range
c.
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39, Containment Integrity is required by Technical Specifications while in Mode 1. Which ONE of the following
. would be defined as a loss of Technical Specification required containment integrity while in Mode !?
a. Containment temperature is 112 T
.
, , b. . The plant vent' radiation monitor is not operable ,
,
I c. - One door of the Personnel Hatch is open for egress
. d. An automatic containment isolation valve has failed open
'
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. . __ . _.. . . _ _ .. __ _ - _ _ _..,.__.. _ ._ . . - _ _ . _.._. _.. .
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40, if a reactor trip were to occur from 100% power with all control systems in normal automatic alignment, which
ONE of the below would disable operation of the Steam Dump system immediately aller the trip?
,
a. - A subsequent failure of turbine header pressure PT-506 high
b. A coincident failure of steam header pressure PT-507 high ,
c. . A failure ofone condenser vacuum switch input to Steam Dumps
. d. . A subsequent failure of turbine header pressure PT-505 low
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41. Unit I was operating at 25% power when a problem developed with #2 RCP requiring the pump to be tripped.
Assuming the unit is stabilized at 25% with three RCPs running following the transient, which ONE of the below
represents expected secondary plant conditions?
a. Feed flow to.#1 SG will be equal to original flow . ,
b. Steam flow from #3 SG will be 1 and 1/3 times original flow
c. Fred flow to #4 SG will be 2/3 times original flow
d. Steam flow from #4 SG will be twice the original flow
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42. Control room operators determine RCDT level is rising during operation in Mode 3. .An investigation is begun to
i
determine the source ofin-leakage. i
. 1
g :.Which ONE of the below represents a list where ALL of the items are potential in-leakage sources? .
'
a. SI Accumulator drains, PRT, and GWPS drains i
,
b. ' Valve leakoffs, CVCS excess letdown, and RCP seals
2 c. PRT, GWPS drains, and valve leakofTs
d. CVCS excess letdown, RCP seals, and SI Accumulator drains
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43. Unit 2 is operating at 100% power with all control systems in automatic when a PZR spray valve inadvertently
opens.
Which ONE of the below correctly lists the initial response of PZR parameters for this event?
PZR Level PZR Temperature PZR Pressure
,
a. INCREASE DECREASE DECREASE
b. DECREASE INCREASE DECREASE
..
- c. DECREASE DECREASE- INCREASE
d. INCREASE INCREASE DECREASE
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44. Unit 2 reactor has tripped due to a loss of off-site power. Natural circulation flow has been established.
The present plant conditions are:
- PZR level 50%.
,- All SG pressures are ~995 psig.-
,
- RCS subcooling is 87 degrees F.
'
Given Steam Tables, what should RCS loop wide range cold leg temperatures be indicating?
a. 480 - 484 degrees F.
b. 486 - 490 degrees F.
c. 544 - 548 degrees F.
d. 550 - 554 degrees F.
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45. During the performance of EOS-1.2," Post-LOCA Cooldown and Depressurization," h is desirable to have only ,
one RCP running. Why only one RCP7
a. _ One RCP provides the dp required to provide letdown. Additional RCPs would add unnecessary heat
load, l
b. One IkCP is desired for spray and RCS heat transport to the SGs. Idditional RCPs wculd add !
unnecessary heat load. I
c. One RCP is needed for RCS heat transport to the SGs. Additional RCPs cou!d overload the electrical l
power supply.
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d. One RCP is desired for spray and RCS mixing. Additional RCPs would strain the plant electrical power ;
supply in the post-LOCA condition.
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. 46. Radiation alarms and confirmatory sample results indicate that RCS activity has exceeded Technical Specification
limits.
'
In addition to a reactor shutdown, which one of the following actions is taken to minimize the likelihood of a
radioactive release to the environment in the event that a Steam Generator Tube Rupture were to occur with the
elevated RCS activity?-
a. Isolate the CVCS demineralizers
b. The RCS is cooled down to <S00 F i
.
.
c. ' Steam Generator Blowdown is secured
.
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b" d. - All' Main Steam Isolation Valves are closed
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.47. Which ONE of the following describes the adverse consequences of continuing charging flow after isolating
letdown with the plant in. Mode !?
a. High temperature at the inlet to the mixed bed demineralizer.
b. - High temperature at the inlet to the letdown heat exchanger.
. c. High thermal stress at the regenerative heat exchanger tube walls.
d. ~ High thermal stress'at the charging line connection to the RCS.
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48. Unit 2 is in Mode 5. The shift is in the process of drawing a bubble in the pressurizer. Pressurizer level hasjust
started to come on scale when a complete loss ofinstrument air occurs.
~ Which ONE of the following describes the plant response with no operator action?
, a. The RCS rapidly depressurizes with max.imum letdown and no charging flow.
b. The plant will hold pressure until the heaters trip on low Pressurizer level.
4
- c. l Charging flow increases and RCS pressure increases until a PZR PORV opens.
d. The plant slowly depressurizes due to inventory loss through the RCP seal leakoff.
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"49.- The Containment Ventilation System (CVS) is comprised of several subsystems, each having particular functions
in the event of specific plant conditions or events
' Which one of the below lists the safety bus electrical loading response of the following three (3) CVS subsystem
to a Blackout (BO) condition?
'
NUTE: Acronyms used
,
- Reactor Coolant Pipe Penetration Cooling System (RCPPCS)
- Containment Air Cooling and Recirculation System (CACRS) '
- Control Rod Drive Mechanism Ventilation System (CRDMVS)
a. Only the CACRS is reloaded in response to a BO signal
b. ~All three systems are sequentially loaded in response to a BO signal
c. Only the CRDMVS is sequentially reloaded in response a BO signal ,
d. Only the CACRS and CRDMVS are sequentially loaded in response to a BO signal
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50. Which ONE of the choices below identifies the minimum Spent Fuel Pool boron concentration by design,
necessary to maintain a Keffless than 0.957
a. ., 2200 ppm
~ b. 2000 ppm
c. 1600 ppm _
d. Oppm
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' 51. Charcoal filters provided on containment air processing systems are designed to remove which ONE of the
.
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> following radioactive isotopes?
a. . Xenon (Xe)
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b; Strontium (Sr) ,
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c.' . Iodine (1)
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d. Krypton (Kr) .
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52. A Plant Equipment Operator (PEO) is dispatched to respond to a Train A Emergency Diesel Generator trouble
alarm. Upon investigation oflocal alarms and indications, the PEO reports both the left and right bank starting air
' pressure is low and has decreased to less than 150 psig.
Which ONE of the following identifies the signals which could start the diesel if the signal was actuated under
these conditions? ,
,
LOCAL MANUAL SAFETY BLACKOUT REMOTE MANUAL
_
NORMAL START INJECTION START EMERGENCY START
a YES NO YES NO
b YES YES NO YES
c NO YES YES YES l
d NO YES YES NO l
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53. A refueling outage is in progress on Unit 2. A Containment purge is in progress to prepare the containment
- ; environment for personnel entry when a 3R SHTDN FLUX HI alarm is received in the control room.
.
Which ONE of the below conditions could be the cause of this alarm?
i-
, a. Or.e Source Range instrument channel has increased to 4 times background
, b. One Source Range instrument channel has increased to 5 times background
c. Both Source Range instrument channels have increased to 2 times background
d. ' Both Source Range instrument channels have increased to 3 times background
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54. With the plant operating at 100% power, a CNDS VAC LO alarm is received and operators observe vacuum at 22
inches and lowering.
Which ONE of the below operator actions is required in this situation?
a. , Open steam dumps to control Tave ,
b. - Start all available Condenser Vacuum pumps
- c. Stop ONE main feed pump to reduce the amount of steam entering the condenser
d. Manually trip the turbine since it failed to trip automatically at the current vacuum
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- 55. When testing Unit 1 Main Steam Isolation Valve #1 (MSIV-1) from the control room, which ONE of the
following conditions will actuate the MSIV #1 TEST FAILED alarm?
, a. MSIV-1 fails to reach 80% open in 10 seconds or less
b. MSIV-1 fai.ls to reach 90% open in 20 seconds or less
c. -MSIV-1 closes 10% and fails to return to full open
.
d. MSIV-1 closes more than 20% during the test
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56. Initial plant conditions are as follows:
1
-
Reactor shutdown
-
RCS boron concentration - 800 ppm
-
Rods are fully inserted ,
It is necessary to add 750 pcm of' negative reactivity to achieve the desired shutdown margin. N' hat is the fina',
RCS boron concentration at the desired condition?
Boron worth =-7.5 pcm/ ppm
Boration = 10 gallons / ppm
Rod Worth = 5 pcm/ step
a. 850 ppm
b. . 875 ppm
c. 900 ppm -
d. 925 ppm
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< 57t A Large Break LOCA has occurred on Unit 1. Hydrogen concentration in Containment has reached the level
required to place the Hydrogen Recombiners in service. Present Containment pressure is 4 psig. Given the
_
.following:
-
1 Post LOCA Pressure (psia) Pressure Factor (cp)
14.7 - 1.14
18.7 1.22
22.7 1.36
24 7 1.45
Reference Power Value:
Train A = 45.86 -
Train B = 45.57
.Recombiner Power Setting = Pressure Factor (cp) x Referene Power I
1
Using the information provided, which ONE of the following is the correct power setting for the Train A 'l
recombiner? l
a.- 52.2 KW
b. ~ 55.9 KW
' c. 62.3 KW
d. 66.4 KW
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58. Which ONE of the below describes the effects of the recombiner in the' Gaseous Waste Processing system?
a. CCW FCV automatically closes on low recombiner water level
c. Inlet Oxygen is automatically terminated to prevent forming a flammable mixture
d. . Oxygen is automatically injected into the gas stream after exiting the recombiner to further reduce
hydrogen concentration
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59. One control rod indicates 10 steps futher out than its bank.
'
Which ONE of the below indications would confirm an actual misalignment of the rod versus a failure of the
individual rod position detector?
4
a. A m. ore negative Axial Flux Difference (AFD) for the power range detector in the immediate vicinity of
- the suspected rod.
, b. A Quadrant Power Tilt Ratio (QPTR) of 1.03 for the core quadrant containing the suspected rod
I
c. A more positive Axial Flux Difference (AFD) for the power range detector opposite the quadrant (180 F )
containing the suspected rod
d. A Quadrant Power Tilt Ratio (QPTR) of 1.03 for the core quadrant opposite the one containing the
suspected rod "
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- 60'. Which ONE of the following would result in a 125 VDC SWITCil PNL 1ED2 TRBL alarm on CB-117
.
a.- - Ground on bus LED 2 -
b) Low AC volts to BC1ED2-2
c. SWBD IED2 feeder breaker open
d. Blown control power fuse on #1 RCP breaker
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61. The plant is operating at full power when a large break LOCA occurs. I
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-Which ONE of the following situations would, ifit occurred, have the greatest negative impact on reducing
containment radiation levels?
' l
a. A failure of all intermediate and,high head Si pumps -l
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b. ' A failure of all high head and one train oflow head SI pumps !
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- c. A failure of all Containment Spray and one train of high head Si pumps I
d. A failure of all intermediate head Si and one train of Containment Spray pumps
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62. A Large Break LOCA has occurred and several critical ECCS components have failed to operate leading to an
inadequate Core Cooling (ICC) condition.
How would indicated Source Range counts change as the downcomer voids? ,
i
, " Source Range count rate would initially... .
a. decrease due to higher coolant density".
.b. increase due to increased neutron leakage".
- c. increase due to increased boron concentration".
d. decrease due to improved neutron moderation in steam",
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63. Which ONE of the below statements accurately describes the Digital Radiation Monitoring system?
jt
- a. All RM-80s provide input into the RM-23s
b. The RM-80 centralizes all radiation data in the plant
c. The PC-I ls allow the operator to exercise control over the RM-80s
' d. The two PC-Ils in the control room each receive inputs from ONLY one half of the RM 80s -
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64. Unit 1 is at 75%. Refueling Water Storage Tank parameters have been observed as follows:
- Leve193%
e Boron Concentration 2500 ppm
- Water Temperature 96 'F
Which ONE of the following describes the required actions?
a.
Restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> due
to Boron Concentration being out of specification
b.
Restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> due
to low RWST level
c. Restore the tank to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT STANDBY within the next 6
hours due to Boron Concentration being out of specification
d. Restore the tank to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT STANDBY within the next 6 .
hours due to low RWST level !
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L 65. Plant conditions:
- Unit load is 590 MWe
, - RCP #1 frame vibration hasjust increased to 7 mils and shaft vibration hasjust increased to 27 mils.
!
Which ONE of the following contains the proper actions required to be performed?
a. Stop # 1 RCP, then trip the reactor and enter EOP-0.0A, " Reactor Trip or Safety Injection".
b. Stop #1 RCP, manually control #1 SG level as necessary and commence a unit shutdown to Mode 3 within
one hour, ;
I
c. Trip the reactor and enter EOP-0.0A, " Reactor Trip or Safety Injection"; then stop #1 RCP.
! d. Trip the reactor and enter EOP-0.0A, " Reactor Trip or Safety Injection"; operate RCPs as directed in the
EOPs.
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66. With the plant operating at 90% power with all control systems in automatic, an I&C Technician error causes a
failure high of Feedwater Header Pressure transmitter PT-508.
Assuming NO operator action is taken, which ONE of the following is correct regarding plant response to the
failure?
a. All SG levels will initially increase and then return to normal programmed level
b. All SG levels will initially decicase and then return to normal rcogrammed level
c. All SG levels will increese and the unit will trip on a turbire: trip >P4
d. All SG levels will decrease and the unit will trip on Low-Low SG level
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4- 67. A Safety injection (SI) actuation has occurred coincident with a 6.9 KV safeguards bus fault preventing the
associated safety equipment from loading onto the bus?
.
Which ONE of the below correctly completes the following statement regarding the operating limits for the
associated Emergency Diesel Generator?-
"Due to the loss of Service Water cooling, the diesel should be...
a. stopped within 15 minutes".
,e
,i 15. ~ stopped within 25 minutes".
'
c. stopped within 35 iainutes."
d. stopped within 45 minutes."
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68. You are required to. perform a verification'of a valve alignment in a plant area containing a radioactive hotspot.
. On the radiation entry permit the hot spot is indicated as 300 mrem /hr when measured 18 inches from the location
. of the radiation source.
If you estimate you will be approximately 3 feet from the source when you perform the valve alignment check,
.
which ONE.of the below is the correct estimate of the radiation field you,will be exposed to?
. a. 30 mrem /hr
b. 75 mrem /hr
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c. 150 mrem /hr
'-
d. 300 mrem /hr l
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69. Which ONE of the conditions or situations described below would require action to be taken in one hour or less in
accordance with Technical Specifications?
a. . A Cold Leg Accumulator boron concentration is reported as 2250 ppm while operating in Mode 1
b. .While operating in Mode 2 a Main Steam Line Safety valve is found leaking and must be gagged
c. AFD is determined to be outside the target band for more than one hour while operating at 100% power
d. UNIDENTIFIED LEAKAGE from the Reactor Coolant System is determined to be 7 gpm with the unit
,
in Mode 3
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. 70. If the Train B CCW pump trips, which of the following is a required initial action of ABN-5027
. ~a. ' Verify adequate RCP Thermal Barrier lix cooling flow
' b. - Verify Train B SSW pump did not trip
'
c. Verify Safety Chiller Recire pump u-05 is running
d. Verify CCW Hx outlet flow is less than 17,500 gpm
.
71. Unit I is operating at 100% power. A Plant Computer alarm occurs for RCP l 01. The Reactor Operator observes
the following parameter:
Motor Stator Winding Temperature 270"F
Motor Upper Radial Bearing Temperature 160 F
Motor Upper Thrust Bearing Temperature ,163 F
Lower Scal. Water Bearing Temperature 240 *F
' Shaft Vibration 12 mils
Frame Vibration 2 mils
Which of the following indicates the reason the operator must trip the reactor:
a. Motor Stator Winding Temperature High
b. Motor Upper Thrust Bearing Temperature High
c. Lower Seal Water Bearing Temperature High
d .' Shaft Vibration High
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72. Unit 1 is operating with the following conditions:
e i
Reactor power 50% i
- ; 2 condensate pumps nmning
ei 2 cire water pumps running
- All other systems and components in automatic .
1
1
In this situnion, which ONE of the following conditions would result in a trip of a main feedwater pump
assuming no operator actions are taken?
1
a. A spurious reactor trip
j
b. A condensate pump trips on overcurrent
c. A selected SG level channel fails low
'
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d. A heater drain pump trips on overcurrent
I
73, Unit 2 is operating at 200 MWe when control room operators observe a simultaneous trip of the main turbine
generator and the running main feedwater pump.
Which ONE of the below plant conditions would have resulted in this transient?
a. Conde.nser vacuum degraded to 20 inches ,
b. The operating Main Feedwater pump tripped
c. A Hi-liilevel occurred in one SG
d. Loss of all condensate pumps occurred
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74. Following a small break loss of coolant accident inside containment concurrent with a faulted SG inside
containment, the following conditions are noted:
-* - RCS subcooling is indicating 10 F
e- RCS pressure is 1380 psig
- All ECCS pumps are operating
- The crew is implementing FRZ-0.1," Response to High Containment Pressure"
- Which one of the following correctly describes why the reactor coolant pumps are tripped under these conditions?
a. Prevent RCP seal damage due to loss of CCW flow
b. Prevent RCP seal damage due to loss of seal injection flow
c. Prevent RCP motor winding damage due to loss of CCW flow
/'
d. - Prevent RCP motor winding damage due to loss of seal injection flow
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75. An out of specification condition is identified on the Unit 1 OWI-104, Reactor Operator logs. Which of the
. following describes how this reading should be recorded?
&
a. If the OOS reading can be restored to normal, restore the reading and record the new reading; ifit cannot
be restored, record the OOS reading
b. Record the OOS reading and circl' e in red ink; note the reason for the OOS reading and corrective actions
taken and notify the Shift Manager or Unit Supervisor
c. Record the OOS reading and the initials of the logkeeper; then enter the exact time of the reading in the
comments section of the log and notify the Shift Manager or Unit Supervisor
d. Record the OOS reading and place an asterisk by the reading; then notify the Shift Manager or Unit
Supervisor
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76. Unit I is operating in Mode 5 with the RCS filled and train A RHR in service providing decay heat removal. -
In this situation which ONE of the below alarm conditions most appropriately indicates the need to use section
4.0, Loss of RCS temperature / flow control, of ABN-104A, Residual Heat Removal System Malfunction?
- a. RHR PMP 1 SUCT VLVS NOT FULLY OPEN
'b. RHRP 1/2 OVRLOAD/ TRIP
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77. In reference to the Critical Safety Function Status Trees, which ONE of the following statements is FALSE?
a. If a Yellow condition is diagnosed, the control room operators may choose whether to continue with the
optimal recovery in progress or to mitiate action to restore the critical safety function.
b. . The Integrity safety function has priority over the Containment critical safety function.
c. Orange path actions for Shutdown Margin take priority over Red path actions for Heat Sink. ,
d. . Critical Safety Function Status Trees apply upon completion of step 3 of EOS-1.3 A, " Transfer to Cold
Leg Recirculation".
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78.' While operating at 100% power with all controls in automatic Power Range Channel N41 fails high. Assuming
- . NO operator action is taken, which ONE of the below describes the effect of the malfunction on the Rod Control
,_ System?
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! a. Rods step in until average Taverfref error signal matches the error signal from turbine power / reactor
power mismatch; rods do.not step out
4
b. Rods step out until bank D rod stop is encountered; rods step in to reduce Taverfref error signal with the
deadband
.
,
c. Rods step in until average Taverfref error signal matches the error signal from turbine power / reactor
, power mismatch; rods step out to reduce Tavertref .r.w signal with the deadband -
d. - Rods step out until Taverfref error signal matches Wblue power / reactor power mismatch error signal:
rods step in to reduce Tave/ Tref error signal with the deadband 1
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79. A main steam line break occurs, resulting in a reactor trip and safety injection. Which ONE of the following
. makes this event a Pressurized Thermal Shock (PTS) concern of the reactor vessel?
a. A rapid heatup followed by a rapid pressurization
b. A rapid depressuriza. tion followed by a rapid heatup ,
c. A rapid cooldown followed by a rapid pressurization
' d. A rapid depressurization followed by a rapid cooldown
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80, The control room has been evacuated due to a fire. Operators have transferred unit control to the Remote
Shutdown Pane! and are monitoring plant conditions. !
'
Which ONE of the following plant parameters is NOT displayed on the Remote Shutdown Panel for this
situation?
E
a. CST level
bi Source Range neutron flux
c. Steam generator pressure
d. RCS flow
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- 81, A reactor trip has occurred from full power and operators are responding. Assume all equipment operates as
~
designed. '
Which ONE of the below must be performed to verify the turbine has tripped in accordance with EOP 0.0A,
" Reactor Trip or Safety Injection".
. 1
i
a. - Verify trip fluid indicates O psig
b. Verify all turbine stop valves are closed
c. - Verify main generator breakers are open
d. . Verify P-13, Turbine $10% Power, light is lit
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82. The unit is at 100% and VCT gas pressure is lowered from 35 psig to 15 psig.
,
Which ONE of the below describes the effect on the plant of this evolution?
'
a. . Letdown flow decreases
. .
.
b. A high RCP standpipe level alarm will actuate
'
c. Flow rates through the RCP No. 2 seals will decrease
d. Charging pump NPSH will be inadequate
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83. Compensation for the effects of gamma current on excore NIS detectors is used in several ranges of flux
monitoring. In the power range, however, compensation is NOT performed.
Which ONE of the below statements describes why gamma compensation is not required for the power range
,
circuitry?
'
a. The power range detectors are located in areas oflow gamma flux and not susceptible to the effects of
gamma-induced currents
b. The power range detectors are surrounded by sufficient lead shielding to reduce the gamma efTect to
insignificant levels
c. Gamma-induced current is relatively insignificant and also proportional to neutron current in the power
range
d. Once the power range is entered, gamma flux in the vicinity of the excore detectors is relatively constant
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l 84. Which ONE of the below completes the following statement comparing the Unit I and Unit 2 Steam Generator
water level ESF-related setpoints?
' Unit 2 SG LOW-LOW LVL REACTOR TRIP setpoint is and the Unit 2 SG Hi-H1 LVL P-14
. TURBINE trip setpoint is , ,
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a. Lower, higher '
b. - Lower, lower
c c. Higher, higher
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85. Unit 2 is operating at 50% power with all controls in automatic when power is lost to 2PC4118 VAC Instrument
. bus. The following control switch alignments existed:
- . PZR pressure control selected to 455/456
- - PZR level control selected to 459/460
Which ONE of the following describes the effects of the loss of 2PC4 with this control alignment?
a. C-5 blocks auto rod withdrawal
' bc Letdown is isolated causing PZR level to rise
ci Charging flow control valve FCV-121 fails open
d. Automatic operation of PCV-455A is lost
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86. Which ONE of the following control room indications would be the MOST useful immediately following an
event to discriminate between a large steamline break in containment and a large LOCA inside containment?
a.-' Containment sump levels
, . b. Pressurizer level ,
c. Containment radiation levels .
d. Power Range NIS
.
87. A reactor startup is in progress on Unit 2. The Intermediate Range (IR) channel I hasjust generated the P-6
_
permissive and channel 2 indicates 1 E-11 amps. Both source range channels indicate approximately 5 E4 cps.
SELECT the statement that describes what these readings indicate.
a. - Both IR channels are undercompensated,
'
b. ~lR channel I is undercompensated.'
c. IR channel 2 is overcompensated.
d. Both'IR channels are overcompensated. l
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'88. A loss of all AC power has occurred and operators are conducting EOPs to mitigate the transient.
Which ONE of the following plant indications would be used by the operators to determine the need to reduce
battery loads in this situation?
a. Battery specific gravity ,
b. DC bus voltage
c. Battery cell voltage
d.' DC bus current
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89. Unit 1 is operating at full power with all control systems in automatic. If the controlling PZR level channel
detector were to develop a leak on the reference leg, which ONE of the below describes how the CVCS system
would respond? -
" Indicated PZR Icvel for the failed channel would . VCT level would , and actual PZR level
would ". .
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a. Increase, decrease, decrease
.
b. Increase, increase, decrease
c. Decrease, increase, decrease
d. Decrease, decrease, increase
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90. Unit 1 i's operating at 100% power when Instrument Air pressure begins to decrease. Identify the condition below
requiring a reactor trip. j
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. a. Compressors 1-01 and X-02 are tripped and cannot be reset.
b. - INSTR AIR HDR PRESS LO alarm is received
c. CCW Surge Tank Makeup valve LV-4500 opens
d. Instrument air pressure decreases to 35 psig
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91' A loss of off-site power has occurred and operators are conducting a natural circulation cooldown on Unit 2.
During the cooldown, the reactor operator is directed to depressurize the RCS by establishing auxiliary spray to
the PZR from CVCS.
Which ONE of the below would indicate the presence of a steam bubble in the reactor vessel head during the
,' auxiliary spray evolution?
a. - Rapid rise in PZR level when spray initiated
, b. Rapid lowering of PZR pressure when spray initiated
c. Rapid lowering of PZR level when spray initiated
d. Rapid rise in PZR pressure when spray initiated
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i 92. The following unit operating conditions exist: i
-
Reactor tripped 'l
- -
Main feedwater isolated
-
Both Main Feedwater pumps tripped
~
Which ONE of the below conditions will result in the plant indications observed? :
1
a. Level in at least one SG decreased below 25%
I
i - b. A Train B AMSAC signal is inadvertently actuated
c. PZR pressure decreased below I820 psig on 2 or more channels
d. A reactor trip occurred and 2 or more Tave channels decreased below 557 F
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l 93. The following are start signals for the Auxiliary Feedwater (AFW) pumps:
i. 1. LO-LO level 2 of 4 detectors on 1 of 4 SGs !
2. LO-LO level 2 of 4 detectors on 2 of 4 SGs ,
3.- AMSAC
L 4. Trip of both Main.Feedwater Pumps *
l 5. Safety injection signal
, 6. Blackout sigual
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Which ONE of the below lists will start BOTH the motor driven and turbine driven AFW pumps? ,
a. 2, 3, 5
b. 2,3,6
. c. 1, 4, 6
' d.' 1, 3, 5 <
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94. A Gaseous Waste Processing System release is in progress.
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~Which ONE of the following malfunctions could result in the release exceeding the limits on the release permit? l
a. Auxiliary Building Vent Duct monitor XRE-5701 fails as is
'
b. Loss of control power to systen$ discharge valve HCV-014
- . , ' c. . Isolation of air to system discharge valve HCV-014
d. Vent Stack #1 radiation monitor fails high
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- 95. During a plant heatup, operators observe the Pressurizer Relief Tank (PRT) level rising.
Which ONE of the following systems or components should the operator investigate that discharges directly to
,
the Pressurizer ReliefTank?
, a. RCP #2 Seal Leakoff
b. ' CVCS Letdown Relief valve
c. Reactor Vessel Flange LeakofT
~ d. Excess Letdown Heat Exchanger CCW Relief valve
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96. Unit'l is at 1% Reactor Power with Auxiliary Feedwater in service. Due to a mechanical maintenance error, air is
isolated to the AFW flow control valve for #1 SG. Select ONE of the below that best describes the response of #1
SG water level and mass.
4
a.- - Level increases due to feed-flow greater than steam flow. S/G mass remains the same.
o - b. L'evel increases due to feed-flow greater than steam-flow. S/G hass increases.
-
c. Level decreases due to feed-flow less than steam-flow. S/G mass remains the same.
~d. Level decreases due to feed-flow less than steam-flow. S/G mass decreases. .
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97. Unit 1 is at 70% with all systems in automatic. A grid fault causes both main generator output breakers
(8000/8010) to trip open.
. Which of the following describes the expected response of the plant?
a. Rods will aptomatically step in at 48 spm
b. Steam dumps will maintain power steady at 70%
c. The main turbine and reactor will automatically trip
d. The main turbine will automatically runback to 60%
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98. Given the following conditions:
- Unit I hasjust experienced a Reactor trip from 60% power. .t
- Power Range NIs are at 0% and current is decreasing in the Intermediate Range. I
,
- Feedwater Regulating Valves are OPEN.
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- Steam Dump Valves are controlling Tavg at no-load.
Which ONE of the following permissives has failed to perform its intended function?
a. P-4, Rx Trip Permissive
b. P-7, Rx and Turbine 210% Power
c. P-10, Rx 210% Power
d. P-12, Tave Lo-Lo
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- 99. A technician error caused a Unit I reactor trip. During step 1 Expected Response of EOP-0.0 you note that six (6)
control rods are indicating fully withdrawn. All other conditions are as expected for a reactor trip.
' Select the correct course of action for current plant con'ditions. >
a. The reactor trip is not confirmed. Perform the RNO by manually initiating a reactor trip,
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b. Initiate a 1560 gallon emergency boration for each of the rods not fully inserted. ,
c. The reactor trip is confirmed. The stuck rods will be addressed in other sections of the EOP network. !
- d. The SRO will direct a transition to FRS-0.1, Response to Nuclear Power Generation.
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100. " Bullets"(solid black dots) preceding steps in an Abnormal Conditions Procedure indicate which of the
following?
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a. The steps may be performed in any order
, , b. - The steps must be performed simultaneously ,
, c. The steps must be performed in the order listed
d. The steps should be performed only if directed by the Shift Manager / Unit Supervisor
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U.S. Nuclear Regulatory Commission
'
Site-Stuecific
.
Written Examination
-
. Applicant Information
- Name: Region: IV
.
] Date: June 19. 1998 Facility / Unit: Comanche Peak 1/2
License Level: SR0 Reactor Type: W
l
Start Time: Finish Time:
Instructions i
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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 i
grade of at least 80.00 percent. Examination papers will be collected four '
hours after the examination starts.
J
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Applicant Certification l
All work done on this examination is my own. I have neither given nor
received aid.
Applicant's Signature
Results
Examination Value Points
Applicant's Score Points
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Applicant's Grade Percent
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' l. A severe weather condition has caused structural damage to switchyard transmission equipment requiring both
units to operate on Emergency Diesel power for an extended period. Unit 1 is operating on a single diesel
generator due to mcchanical problems with the remaining diesel.
Assuming Unit I operated within Technical Specification guidelines before the incident, which one of the below
correctly describes the minimum design capacity of the diesel fuel oil system for a single Unit I diesel generator
under these conditions?
i
a. . Operate for seven days at continuous load rating
b. Operate for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at' maximum overload rating '
c. Operate for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at maximum overload rating
-d. Operate for one day at continuous load rating i
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- 2. In accordance with STA-656," Radiation Work Control", which ONE of the following statements is correct
regarding logging out of the RP computer system wben preparing to leave the RCA?
a. When logging out of the RP computer, input the areas entered while in the RCA
b. When logging out of the RP computer, specify both the RWP/ GAP and Task numbers
c. . When logging out of the RP computer, insert your hand into the hand-reader for verification
i
d. When logging out of the RP computer, slide your RCA Access Card through the bar code reader
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3. Which ONE of the below correctly describes the reason why the Containment Purge Air Supply and Exhaust
valves are required to be locked and administratively controlled during plant operations?
.
a. To prevent a spurious Containment ventilation isolation
"
b. To prevent the admission of unfiltered air into Containment
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c. The valve actuators do not have penetration conductor overcurrent devices
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, d. The valves are not guaranteed to close during a LOCA or steam line break accident
.
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4. A liquid radioactive discharge is scheduled for your shift Under these conditions, which ONE of the below is
correct regarding the operation of the Liquid Radwaste to Circulating Water Monitor X-RE-52537
a. It can be monitored in the Control Room on the RM-23 during the release
b. It cannot be monitored in the Control Room if both the PC-1 is are inoperable
c. It uses a Geiger-Mueller detecting element in a lead-shielded sample chamber
d. It will automatically close X-RV-5253 at the ALERT setpoint to terminate the release
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5. . While operating at steady state conditions at 75% power, the Unit 2 Reactor Operator reports that control rods are l
stcpping outward in AUTOMATIC.- The operator places the rods in MANUAL and rod motion stops.
Which one of the following is a possible cause of this rod motion?
a. . A Loop T-cold fails high ,
' b ' NIS channel N-42 fails high j
,
c. First Stage Pressure transmitter, PT-505, fails high
d. An N16 channel fails high -
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- 6. Reactor power is being maintained at approximately 6% prior to placing the Main Turbine on-line. Intermediate
Range (IR) channel N-35 is in the TRIP BYPASS position, with all required bistables tripped for troubleshooting
-
due a failure which occurredjust after entering Mode 1.
Which one of the following will occur if the N-35 Control Power fuses were to blow at this time?
a. An overhead annunciator for IR detector high voltage will occur
b. The reactor will immediately trip on IR high
'
c. An overhead annunciator for IR detector compensating voltage will occur
- d. Both Source Range instruments will automatically re-energize
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7.' Safety injection pump train A has been tagged out for motor bearing replacement. A Safety Injection
subsequently occurs due to Large Break Loss of Coolant Accident (LBLOCA) inside containment. Two hours
later, the train B Si pump fails. All other equipment functions as designed for the duration of the accident.
Which one of the following describes how the loss of both Si pumps will affect the ability of the crew to mitigate
the effects of this accident? ,
a. A transition to EOS-l.1, Safety injection Termination, will be required upon receipt of the RWST low-low
level alarm
b. Both trains of RilR will remain aligned for cold leg injection during the alignment in EOS- 1.3, Transfer to
Cold Leg Recirculation
c. Actions will be necessary to restore at least one SI Pump to service in order to achieve hot leg injection per
EOS-l .4, Transfer to IIL Recirculation
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d. EOS-1.4, Transfer to llL Recirculation, provides for aligning one CCP for hot leg injection when neither SI
Pump is operable
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- 8. An automatic reactor trip and safety injection has occurred on Unit 2 as a result oflowering RCS pressure. The
operators note the following conditions:
Pressurizer pressure dropping prior to and following the St .
..
. RCS average temperature stable prior to and following tfic Si
e Pressurizer level rising prior to the Si and rising following the S1
.' e ' Reactor power stable prior to the Si and dropping following the S1
Initially which ONE of the following accidents would result in these conditions? ,
a.' Steamline break -
,
b. Double-ended hot leg break
c. Stuck open pressurizer safety valve I
d. ' 4 inch break on a RCS cold leg !
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- 9. A reactor trip has occurred on Unit 2 due to a loss of offsite power. The crew has completed mo'st of the actions
of EOS 0.I, Reactor Trip Response, and are verifying natural circulation flow. When adjusting steam dumping
rate to control natural circulation, the operators also adjust AFW flow to all of the SGs.
Which ONE of the following correctly explains why narrow range level is re-established in all SGs?
.
a. To maintain symmetric cooling of the RCS
b. ' To flood all SGs for subsequent entry into Mode 5
c. SG wide range level indication is lost on loss of offsite power
' d. Top of SG tubes on all SGs must be covered for natural circulation to occur
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CPSES MO Exam
.10. During operatiw at power with the Reactor Trip Breakers (RTBs) closed a loss of 125 VDC to one of the RTBs
. occurs.
- Which one of the following correctly describes how the reactor trip breaker (RTB) will be affected by loss of the
' 125 VDC power?
.
..
a. It trips open due to loss of power to the shunt coil
b. It trips open due to loss of power to the undervoltage coil
. c. It is not capable of tripping on a shunt trip
.'d. It is not capable of tripping on an undervoltage trip
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' i 1.' In the event of a Steam Generator Tube Rupture (SGTR), the assumption is made that operators will isolate
- Auxiliary Feed Water (AFW) flow to the affected SG within 10 minutes.
. Which one of the following is the basis for the action and time limit?
a. - To conserve CST inventory ,
b. Minimize the probability of SG overfill
,
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c. To maximize the time the AFW pumps are on recirculation
4 d. To limit the cooling effect on the SG so steamline pressure remains high
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- 12. During operation at power steam generator tube leakage is detected and estimated at 250 gpm by the reactor
operator. The following plant indications existed at that time:
RCS pressure-2200 psig and lowering
Reactor Power- 80%
SG Pressures- 1600 psig , ,
~ PZR Level--42% and lowering
The unit is tripped and plant parameters following the trip are:
RCS pressure- 1700 psig and lowering
Reactor Power-0%
SG Pressures- 1100 psig
PZR Level- 13%
- Based on the two sets of given data, which ONE of the below describes the effect on primary-to-secondary
leakage?
Leakage following the trip is
a. one half of the initial leak rate or about 125 gpm.
b. essentially equal to the initial leak rate or about 250 gpm.
c. approximately 70% of the initial leak rate or about 175 gpm.
- d. One third of the initial leak rate or about 83 gpm.
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r: 13. During clearance or valve positioning activities, which ONE of the following conditions would allow independent
verification of equipment status to be waived in accordance with STA-694," Station Verification Activities"?
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a. A clearance requires removal of a gag on a Main Steam Safety valve
-
b. The valve verification would r.esult in radiation exposure of 120 mrem
c. ' ' A clearance requires installation of a grounding strap on a non-safety related 480V breaker
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d. The valve verification requires entry into containment during fuel movement and would result in a
radiation exposure of 5 mrem
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14. Unit 1 is conducting refueling activities with the core being offloaded to the spent fuel pool for determination of
suspected leaking assemblies.- All Fuel Handling Building systems are operational and correctly aligned for
. refueling operations.
During movement of one of the assemblies, the Fuel Building radiation monitor, X-RE-6272, alarms. Which
.
ONE of the following INITIAL operator actions is required in this situation?
'
a.' Evacuate the Fuel Building ,
b. - Start Pre-Access Filtration System
c. Ensure Containment Ventilation Isolation occurs
d. Verify Fuel Building Ventilation System automatically shifts to the Isolate mode
.
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15. Unit 2 is cun'ently operating at 100% when a control rod in bank C partially drops. A QPTR just performed by
the RO indict'ns a Quadrant Power Tilt Ratio (QPTR) of 1.07.'
Using the following statement from Technical Specification (3/4.2.4), determine the course of action the operators
must take given this scenario.
"Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reduce THERMAL POWER at least 3% from ikATED THERMAL POWER for each 1% of
QUADRANT POWER TILT RATIO in excess of 1."
a. Reduce thermal power to 97%
b. Reduce thermal power to 89%
c. Reduce thermal power to 79%
d.- Reduce thermal power to 71%
'
. 16.' Regarding Technical Specification SAFETY LIMITS, which ONE of the following core limitations does the OT
N16 reactor trip prevent exceeding?
a. Power Density (KW/ft)
b. --' Departure from Nucleate Boiling (DNB) ,
c. Total Core Power (NSSS Power Limit)
d. Axial Flux Difference (AFD) -
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117. Which ONE of the following lists of personnel satisfies the requirement for the Fire Brigade complement in
accordance with STA-727," Fire Brigade"?
a. One Fire Brigade Leader (Reactor Operator); 2 Mai. unance Mechanics (Nonteman); 2 Maintenance
Electricians (Hoseman).
~
b;
The Shift Manager (Fire Brigade Leader); 2 Plant Equipment Operators (Nonieman); 1 Security
Personnel (Hoseman).
'
c. One Fire Brigade Leader; 3 Plant Equipment Operators (Nonleman/Hoseman); 1 Safety Services
(Hoseman).
d.' >
One Fire Brigade Leader; 2 Security Personnel (Nonleman); 4 Maintenance Mechanics (Hoseman).
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'18. An event has occurred on Unit I and operators are conducting EOP-1.0A," Loss of Reactor or Secondary
Coolant", when the below parameters are observed:
All SG pressures- 800 psig and stable
'
All SG levels - being controlled at 10% NR
PZR level-off-scale low
Containment Pressure- 16 psig
'
RWST level-40% +
RCS pressure- 180 psig and stable
Based on these indications, which ONE of the following procedures would the operators enter next to mitigate the
event in progress?
a. EOS-1.2A," Post-LOCA Cooldown and Depressurization"
b. EOS-1.1 A, "Si Termination" +
c. ECA-1.l A, " Loss of Emergency Coolant Recirculation"
,
d. EOS-1.3 A, " Transfer to Cold Leg Recirculation"
- - -- - . . - , - - -.. . . _ . .. -. .~ . - . . _
- 19.- A control rod in Control Bank D partially drops into the core with the below conditions:
-
- Reactor power 60%
-
Tave 0.5 'F below Tref
-
Control Bank D group demand counters at 180
.
After the dropped rod, the unit stabilizes at the following conditions:
-
Reactor power 60%
,
-
Tave 2.5 'F below Tref - l
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Control Bank D group demand counters at 180
-)
Which ONE of the below is correct regarding the effects on the Cold Leg temperature (Tc) and Shutdown Margin 1
- (SDM) from the onset of the esent? .
l
a. Both Tc and SDM have decrea:ed )
b. Both Tc and SDM have reianined the same
c. Tc has decreased and SDM has remained the same I
d. Tc has remained the same and the SDM has decreased ;
,
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20. A small break LOCA has occurred and operators have implemented EOP-0.0A," Reactor Trip or Safety i
'
Injection". In this situation, which ONE of the below statements indicates the basis for tripping the RCPs if
, minimum RCS subcooling is lost and Si flow has been established?
a. , Prevent excessive depletion of RCS inventory through a small break leading to severe core uncovery if the
RCPs were later tripped.
b. Prevent da'mage to the RCP and RCP seal package due to possibility of two-phase flow in RCS.
c. Prevent physical damege to RCPs and RCS due to stresses associated with pomping a two-phase mixture,
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d. To funher decrease RCS pressure, enhaacing ESF systems ability to inject borated liquid into RCS.
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- 21.- Which ONE of the following is an operational implication of maintaining Rod insertion Limits?
. ' " Maintaining Rod Insertion Limits ensures
,
a. rod tip fretting is minimized."
"
b/ proper bank overlap is maintained."
c. etTects of rod drops are minimized."
d. minimum shutdown margin is maintained."
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22. Operators are conducting a plant cooldown for a refueling outage'and have reached plant conditions required to
, allow blocking of the low PZR pressure Safety injection (SI) signal.
' During the subsequent heatup and startup of the plant following the outage, which ONE of the below will unblock
4
the automatic low PZR pressure SI signal?
'
'
a. When 2 out of 3 PZR pressure channels are greater than the P-11 setpoint of 1960 psig '
b. When 3 out of 4 PZR pressure channels are greater than the auto SI setpoint of 1820 psig
c. When the control room operator manually unblocks the signal as directed in the heatup/startup procedure
'
d. When BOTH reactor trip breakers are closed, removing the SI blocking feature provided by the P-4
,
intedock -
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. 23.. Which ONE of the following specifies the minimum number ofcom exit thermocouples (CETs) which must be
'
- operable per Technical Specification Table 3.3-67 .
a. 2 per quadrant
b. 4 per quadrant ,
,
c. 6 per quadrant
d. 8 per quadrant
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24. Unit 2 is in Mode 3 at 375 *F when the reactor operator observes a PZR PORV open with RCS pressure dropping
rapidly. Subsequent investigation reveals wide range temperature instrument TE-413A, HL 1 WR TEMP failed '
low.
Which ONE of the below accurately describes the response of the Low Temperature Overpressure Protection !
,
(LTOP) system to the observed indicat:ons?
,
1
a. The LTOP system is operating correctly; the average of four loop temperatures has been reduced low
enough to cause arming and opening of the train associated PORV
b. The LTOP system is not operating correctly; the failed temperature input should have reduced the average ,
. temperature input to the associated train and redundant backup to the opposite train causing both PORVs l
to open j
i
c. The LTOP system is operating correctly; the auctioneered low temperature plus exceeding the calculated i
equivalent pressure should have opened the train associated PORV
d. The LTOP system is not operating correctly; no PORV should have opened because no input was !
received from the redundant train temperature instrument
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25. Unit 1 is operating with the below plant conditions:
- Reactor power 99%
- PZR level 60%
- Letdown flow 75 gpm
- The.PD pump is in service
- All controls are in automatic
,
A 40 gpm charging line leak exists outside containment.
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Assuming NO operator action is taken, which ONE of the following identifies the potential consequence of the ;
. event: PZR level drops to 17%, letdown isolates and PZR heaters turn off; and the
1
a. Reactor trips on high PZR level !
b. Reactor trips on high PZR pressure
,
c. PZR level maintains at 60%, no reactor trip .
d. PZR pressure naintains at 2235 psig, no reactor trip
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26. An event has occurred which has actuated the Containment Spray system. The following plant conditions are
observed:
-
Containment Spray discharge flow is 900 gpm
- -
Containment Spray Sump Suction valves are CLOSED
-
Containment Spray Recirculation Valves are OPEN ,
-
Containment Spray Header Isolation valves are OPEN '
Which ONE of the below describes the operation of the Containment Spray system based on these observations?
4
a. Not operating correctly; the Recirculation valves should not be OPEN until ~1200 gpm
b. Not operating correctly; the ' Recirculation valves should not be OPEN with sump suction valves
CLOSED '
c. Not operating correctly; the Recirculation valves should be CLOSED with the Header Isolation valves
4- OPEN
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d. Operating correctly; valve interlocks have been met and discharge flow is less than setpoint l
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27.- A design basis LOCA has occurred on Unit 2. A'ssume all equipment operates as designed.
.
Determine which ONE of the below is the approximate time for RWST level to decrease to the cold leg
recirculation transfer criteria level,
a. 15 minutes ,
b. 30 minutes
c. 40 minutes
' d. 50 minutes
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28, During performance of EOS-l.3," Transfer to Cold Leg Recirculation", operators are attempting to open valves
. , .
- 8804A and 'B, RHR Discharge to Safety Injection / Charging Pump suction.
Considering each answer separately, which ONE of the below should be considered for the interlock logic for this
situation?
a. RilR suction from RCS isolated
b. RHR suction from RWSTisolated
c. . RHR suction from Containment sump isolated
d.' Safety Injection pump suction from RWST isolated
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29. Which ONE of the below automatic reactor trip signals is described by the following statement taken from the
BASES of Technical Specifications, Section 2.0, Limiting Safety System Settings?
" Prevents water relief through the pressurizer safety valves."
,a. PZR Level High .
b. PZR Pressure Low
c. SG Water Level Low
~ d. -Source Range Flux High
- . _ . - . __ - . _ . . . . . . .._._ - . . . _ - . _ . . _ . . _ _ _ _ _ _ . . _ . _ ._ .. _ _.
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30. Unit 2 is conducting a plant shutdown with reactor power currently at 15%. As directed by the procedure in
effect, the RO selects the steam pressure mode of operation. Unknown to the operator, the controller output has )
failed to 100%. l
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' Which ONE of the below plant responses will occur if no operator action is taken? !
a. All steam dumps will arm, but remain closed. Reactor power will'not be affected.
b. All steam dumps will open until Tave reaches 5*F above Tref. Reactor power will not be affected.
- c. All steam dumps will remain closed until steam pressure reaches 1160 psig. All steam dumps will cycle at
1160 psig and reactor power will decrease.
~
- d. All steam dumps will open and reactor power will rise. When Tave reaches 553*F all dump valves will
close and reactor power will decrease.
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31. Unit 1 is operating at 100% power with the below conditions:
-
PZR pressure control system is in automatic maintaining 2235 psig
-
PT-455 is selected as the pressure input to' the PZR pressure master controller
-
PT-455 has failed full scale high
-
NO operator action is taken for the event
.
Which ONE of the following describes the initial system response to this failure?
a. PZR PORV PCV-455A opens and PZR PORV PCV-456 remains closed
b. PZR PORV PCV-455A remains closed and PZR PORV PCV-456 opens
c. PZR PORV PCV-455A and PZR PORV PCV-456 both remain closed
d. PZR PORV PCV-455A and PZR PORV PCV-456 both open
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32.' Which of the following is an indication of vortexing at the suction of the RilR pump during reduced inventory
conditions? -
a. Erratic pump amps
L b. krik suction relieflifling ,
c. Decreasing RCS temperature
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' d. - Constant pump discharge pressure
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33. Unit 2 is operating at full power when a total loss of Main Feedwater occurs coincident with a failure of the
reactor to trip (ATWS).
For this condition, what is the bases for the operator action to trip the main turbine within 30 seconds of the
event?
.
'
a. Prevent a safety injection on low steam line pressure
b. Initiate an alternate means of reactor trip from a turbine trip
c. Conserve remaining steam generator inventory
d.- ' Initiate an alternate means of reactor trip from high RCS pressure
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34. Unit I control room operators are inve'stigating a potential malfunction of the Reactor Coolant Makeup system
.b'ased on abnormal indications following an RCS makeup evolution. The RO suspects the pot setting for 1-FK-
110, Boric Acid Blender Flow Control, may have been set wrong.
- Given RCS boron concentration is to be maintained at 500 ppm, with BAST boron concentration at 7000 ppm and
a blended flowrate of 127 gpm, determine the proper setting for 1-FK-110.
a. - 2.27
b. ' 2.85
c. 8.55
. c d. 9.07
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35. A safety injection has occurred coincident with a loss of off-site power Bus leal has deenergized due to an 86-1
lockout. Determine from the table below the high head pumps that would be available.
PDP CCPl CCP2
.
a. ON ON OFF
b.- ON OFF OFF
c. OFF ON ON
d. OFF OFF ON
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36. Unit 1 is operating with the following conditions:
'
-
90% power
-
Both Main Feedwater pumps in service I
-
Both Condensate pumps in service
-
Main Turbine controls have been shifted to M11C ,
-
-
All other control systems in normal automatic alignment
.
Under these conditions, which ONE of the below is expected to occur if condensate pump 1-01 were to trip and
no operator actions were taken?
a. Main Feedwater Pump 1-01 will trip and unit load will decrease to 60% at 35%/ minute
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b. Both Main Feedwater Pumps will trip and the reactor will trip on SG NR level at <25%
,
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c. Unit load will decrease to 60% at 35%/ minute and both Main Feedwater pumps will continue to operate
program
,
d
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37.. Which ONE of the following conditions would cause loss of Component Cooling Water (CCW) flow to the
Ventilation Chillers? ,
,
a. A Reactor Trip signal
c_
= b. A Containment Isolation Phase B sign.at
c. A Containment Ventilation Isolation signal
d. ~A Containment Isolation Phase A signal
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38, A large steamline break has occurred inside Containment. Operators have been unable to close any MSIV. The !
following plant conditions exist: I
All SG levels:
'
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<5% NR
-
All SG pressures: 800# and decreasing
' Operators have been directed by the ERGS in progress to maintain a minimum AFW flow of100 gpm to e
Which ONE of the following is the basis and operational implication of performing this action? 1
a. prevents steam generator tube dryout
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b. maintains a verifiable RCS cooldown rate I
c. maintains steam pressure above safety injection setpoint
I
d. ensures steam generator levels will remain above 5% narrow range
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- 39. Containment Integrity is required by Technical Specifications while in Mode 1. Which ONE of the following .
,
would be defined as a loss ofTechnical Specification required containment integrity while in Mode !?
a. Containnient temperature is 112 'F ,
' b. l The plant vent radiation monitor is not operable ,
c. One door of the Personnel Hatch is open for egress
d. An automatic containment isolation valve has failed open
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,
. 40. If a reactor trip were to occur from 100% power with all control systems in normal automatic alignment, which
ONE of the below would disable operation of the Steam Dump system immediately after the trip?
a. , A subsequent failure of turbine header pressure PT-506 high
. b. A coincident failure ofucam header pressure PT-507 high
- - '
c. A failum of one condenser vacuum switch input to Steam Dumps
~
d. A subsequent failure of turbine header pressure PT-505 low
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41, Unit I was operating at 25% power when a problem developed with #2 RCP requiring the pump to be tripped.
- Assuming the unit is stabilized at 25% with three RCPs running following the transient, which ONE of the below
represents expected secondary plant conditions?
.
'
. a. Feed flow to #1 SG will be equa,1 to original flow ,
i
b. ' Steam flow from #3 SG will be I and 1/3 times original flow
. c. Feed flow to #4 SG will be 2/3 times original flow
,
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i d. Steam flow from #4 SG will be twice the original flow
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.42. Control room operators determine RCDT level is rising during operation in Mode 3. An investigation is begun to
.
determine the source ofin-leakage.
1
2
Which ONE of the below represents a list where ALL of the items are potential in leakage sources?
i , a. St Accumulator drains, PRT, and GWPS drains
,
b. Valve leakoffs, CVCS excess letdown, and RCP seals
c. PRT;GWPS drains,and valve leakoffs
.
d. CVCS cxcess letdown, RCP seals, and SI Accumulator drains
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43. Unit 2 is operating at 100% power with all control systems in automatic when a PZR spray valve inadvertently
opens.
Which ONE of the below correctly lists the initial response of PZR parameters for this event?
PZR Level P2R Temperature PZR Pressure
a.' INCREASE DECREASE DECREASE
.
b. DECREASE INCREASE DECREASE
c. DECREASE DECREASE INCREASE
l
d. INCREASE INCREASE DECREASE
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-44. Unit 2 reactor has tripped due to a loss of off-site power. Natural circulation flow has been established .
' The present plant conditions are:
- PZR level 50%.
- All SG pressures are ~995 psig.
- RCS subcooling is 87 degrees F.
' Given Steam Tables, what should RCS loop wide range cold leg temperatures be indicating?
'
a. 480 '- 484 degrees F.
. b. 486 - 490 degrees F.
c. 544 - 548 degrees F.
d. 550 - 554 degrees F.
.
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45. During the performance of EOS-1.2, " Post-LOCA Cooldown and Depressurization," it is desirable to have only ,
one RCP running. Why only one RCP't I
a.
One RCP provides the dp required to provide letdown. Additional RCPs would add unnecessary heat l
load. l
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b. One RCP is desired for spray and RCS heat transpon to the SGs. Additional RCPs would add l
unnecessary heat load.
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. c. One RCP is needed for RCS heat transport to the SGs. Additional RCPs could overload the electrical i
power supply.
d. One RCP is desired for spray and RCS mixing. Additional RCPs would strain the plant electrical power
supply in the post LOCA condition.
.
a
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46. Radiation alarms and confirmatory sample results indicate that RCS activity has exceeded Technical Specification
limits.-
~ In addition to a reactor shutdown, which one of the following actions is taken to minimize the likelihood of a
radioactive n: lease to the environment in the event that a Steam Generator Tube Rupture were to occur with the
>
elevated RCS activity?
,
~
- a. isolate the CVCS demineralizers
,
b. The RCS is cooled down to <500*F
c. Steam Generator Blowdown is secured
1
d. ' All Main Steam Isolation Valves are closed
1
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47. Which ONE of the following describus the adverse consequences of continuing charging flow after isolating
letdown with the plant in Mode !?
a. High temperature at the inlet to the mixed bed demineralizer.
, b. . High temperature at the inlet to the letdown heat exchanger.
c. High thermal stress at the regenerative heat exchanger tube walls.
~ d. High thermal stress at the charging line connection to the RCS.
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48. Unit 2 is in Mode 5. The shift is in the process of drawing a bubble in the pressurizer. Pressurizer level hasjust
started to come on scale when a complete loss ofinstrument air occurs.
Which ONE of the following describes the plant response with no operator action?
a. The RCS rapidly depressurizes with maximum letdown and no. charging flow.
b. The plant will hold pressure until the heaters trip on low Pressurizer level.
c. Charging flow increases and RCS pressure increases until a PZR PCRV opens.
d. The plant slowly depressurizes due to inventory loss through the RCP seal leakoff.
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49. The Containment Ventilation System (CVS) is comprised of several subsystems, each having particular functions
in the event of specific plant conditions or events.
Which one of the below lists the safety bus electrical loading response of the following three (3) CVS subsystems
to a Blackout (BO) condition?
.
.
NOTE: Acronyms used
- Reactor Coolant Pipe Penetration Cooling System (RCPPCS)
,
- Containment Air Cooling and Recirculation System (CACRS)
- Control Rod Drive Mechanism Ventilation System (CRDMVS)
,
a. ~ ~~ Only the CACRS is reloaded in response to a BO signal
b. All three systems are sequentially loaded in response to a BO signal
c. Only the CRDMVS is sequentially reloaded in response a BO signal
d. Only the CACRS and CRDMVS are sequentially loaded in response to a BO signal
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1CPSES SRO Exam
50. Which ONE of the c$ ices below identifies the minimum Spent Fuel Pool boron concentration by design,
necessary to maintain a Keffless than 0.95?
a. 2200 ppm :
b. : 2000 ppm ,
c. 1600 ppm
d. O ppm . '
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S t. Charcoal filters provided on containment air processing systems are designed to remove which ONE of the
following radioactive isotopes?
a. Xenon (Xe)
, b. Strontium (Sr)- .
.
c. ' Iodine (1) . ,
- d. . Krypton (Kr)
,
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52. A Plant Equipment Operator (PEO) is dispatched to respond to a Train A Emergency Diesel Generator trouble
alarm. Upon investigation oflocal alarms and indications, the PEO reports both the left and right bank starting air
pressure is low and has decreased to less than 150 psig.
Which ONE of the following identifies the signals which could start the diesel if the signal was actuated under
these conditions? .
LOCAL MANUAL SAFETY BLACKOUT REMOTE MANUAL
NORMAL START INJECTION - START EMERGENCY START
a YES NO YES NO
b YES YES NO YES
c NO YES YES YES
d NO YES YES NO
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53. A refueling outage is in progress on Unif 2. A Containment purge is in progress to prepare the containment
environment for personnel entry when r. SR SliTDN FLUX 111 alarm is received in the control room.
Which ONE of the below conditions .:ould be the cause of this alarm?
a. - One Source Range instrument channel has increased to 4 times background
b. . One Source Range instrument channel has increased to 5 times background
c. Both Source Range instrument channels have increased to 2 times background
d. Both Source Range instrument channels have increased to 3 times background
,
.
. . . . - .. . .- .~ - - , . - - . . . . . . .
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- 54. With the plant operating at 100% power, a CNDS VAC LO alarm is received and operators observe vacuum at 22
inches and lowering.
. Which ONE of the below operator actions is required in this situation?
a. Open steam dumps to control Tave .
b. Start all available Condenser Vacuum pumps
F c. Stop ONE main feed pump to reduce the amount ofsteam entering the condenser
d.L Manually trip the turbine since it failed to trip automatically at the current vacuum
7
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55. When testing Unit 1 Main Steam Isolation Valve #1 (MSIV-1) from the control room, which ONE of the
.
I
following conditions will actuate the MSIV #1 TEST FAILED alarm?
a. MSIV-1 fails to reach 80% open in 10 seconds or less
b. MSIV-1 fails to reach 90% open in 20 seconds or less
c. MSIV-1 closes 10% and fails to return to full open
d. MSIV-1 closes more than 20% 6!:ing the test
,
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56. Initial plant conditions are as follows:
' Reactor shutdown
-
Rods are fully inserted
'
it is necessary to add 750 pcm of negative reactivity to achieve the desired shutdown margin. What is the final
RCS boron concentration at the desired condition?
Boron worth -7.5 pcm/ ppm
Boration = 10 gallons / ppm ~
,
Rod Worth = 5 pcm/ step
a. 850 ppm
b. 875 ppm
.
c. 900 ppm -
.
- d. 925 ppm
,
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57. A Large Break LOCA has occurred on Unit 1. Hydrogen concentration in Containment has reached the level
required to place the Hydrogen Recombiners in service. Present Containment pressure is 4 psig. Given the
following:
Post LOCA Pressure (psia) Pressure Factor (cp) !
1.4.7 1.14 ,i
18.7 1.22
22.7 1.36
24.7 1.45
Reference Power Value:
Train A = 45.86
Train B = 45.57
Recombiner Power Setting = Pressure Factor (cp) x Reference Power
Using the information provided, which ONE of the following is the correct power setting for the Train A
recombiner7
a. 52.2 KW
b. 55.9 KW
c. ' 62.3 KW '
d. 66.4 KW
.
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'58. .Which ONE of the below describes the effects of the recombiner in the Gaseous Waste Processing system?
b. - Helium is automatically injected into the feed to purge out excess Hydrogen
c. - Inlet O'xygen is automatically terminated to prevent forming a flamniable mixture
i. d. - ~ Oxygen is automatically injected into the gas stream after exiting the recombiner to further reduce
hydrogen concentration
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59. One control rod indicates 10 steps farther out than its bank.
.
Which ONE of the below indications would confirm an actual misalignment of the rod versus a failure of the
individual rod position detector?
a. A more negative Axi.al Flux Difference (AFD) for the power range detector in the.immediate vicinity of ,
the suspected rod.
b. A Quadrant Power Tilt Ratio (QPTR) of 1.03 for the core quadrant containing the suspected rod
c. A more positive Axial Flux Difference (AFD) for the power range detector opposite the quadrant (180 F )
containing the suspected rod
d. A Quadrant Power Tilt Ratio (QPTR) of 1.03 for the core quadrant opposite the one containing the
suspected rod
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60.~ Which ONE of the following would result in a 125 VDC SWITCH PNL LED 2 TRBL alarm on CB-117
a. - Ground on bus IED2
b. , Low AC volts to BC1ED2-2 -
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. c. . SWBD LED 2 feeder breaker open
d. Blown control power fuse on #1 RCP breaker
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61. The plant is operating at full power when a large break LOCA occurs.
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, Which ONE of the following situations would, ifit occurred, have the greatest negative impact on reducing
containment radiation levels?
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- a. A failure of all intermediate and high head,SI pumps 1
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. b. A failure of all high head and one train oflow head SI pumps l
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c. A failure of all Containment Spray and one train of high head SI pumps
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d. A failure of all intermediate head Si and one train of Containment Spray pumps i
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62. A Large Break LOCA has occurred and several critical ECCS components have failed to operate leading to an
inadequate Core Cooling (ICC) condition.
How would indicated Source Range counts change as the downcomer voids?
" Source Range count rate would initially... ,
,
! a. decrease due to higher coolant density".
- b. increase due to increased neutron leakage".
c. increase due to increased boron concentration".
d. decrease due to improved neutron moderation in steam".
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63. .Which ONE of the below statements accurately describes the Digital Radiation Monitoring system?
a. ~ All RM-80s provide input into ths RM-23s
b. The RM-80 centralizes all radiation data in the plant
c. The PC'- l is allow the operator to exercise control over the RM-80s
. d. The two PC-Ils in the control room each receive inputs from ONLY one-half of the RM-80s -
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64. Unit 1 is at 75% Refueling Water Storage Tank parameters have been observed as follows: f
e Leve193%
e Boron Concentration 2500 ppm
e ' Water Temperature 96 'F *
Which ONE of the following describes the required actions?
a. Restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> due
to Boron Concentration being out of specification -
b. Restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> due
to low RWST level
c. Restore the tank to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT STANDBY within the next 6
hours due to Boron Concentration being out ofspecification
d. Restore the tank to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT STANDBY within the next 6
hours due to low RWST level
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2 65. Plant conditions:
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- Unit load is 590 MWe
- RCP #1 frame vibration hasjust increased to 7 mils and shaft vibration hasjust increased to 27 mils.
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Which ONE of the following contains the proper actions required to be performed?
a. Stop #1 RCP, then trip the reactor and enter EOP-0.0A, " Reactor Trip or Safety injection".
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b. Stop #1 RCP, manually control #1 SG level as necessary and commence a unit shutdown to Mode 3 within
one hour.
j. c Trip the reactor and enter EOP-0.0A, " Reactor Trip or Safety injection"; then stop #1 RCP. ,
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d.~ Trip the reactor and enter EOP-0.0A, " Reactor Trip or Safety Injection"; operate RCPs as directed in the
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EOPs.
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66. With the plant operating at 90% power with all control systems in automatic, an I&C Technician error causes a
failure high of Feedwater Header Pressure transmitter PT-508. :
Assuming NO operator action is taken, which ONE of the following is correct regarding plant response to the
._
failure? '
'
a. ~ All SG levels will initially increase and then retbrn to normal programmed level
b. 'All SG levels will initially decrease and then return to normal programmed level
c.' 'All SG levels will increase and the unit will trip on a turbine trip >P-9
d. . All SG levels will decrease and the unit will trip on Low-Low SG level -
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67. A Safety Injection (SI) actuation has occurred coincident with a 6.9 KV safeguards bus fault preventing the
associated safety equipment from loading onto the bus?
Which ONE of the below correctly completes the following statement regarding the operating limits for the
associated Emergency Diesel Generator?
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"Due to the loss of Service Water cooling, the diesel should be...
a. stopped within 15 minutes",
b. stopped within 25 minutes".
c. stopped within 35 minutes."
- d. . stopped within 45 minutes."
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. 68. You are required to perform a verification of a valve alignment in a plant area containing a radioactive hotspot.
On the radiation entry permit the hot spot is indicated as 300 mrem /hr when measured 18 inches from the location
- of the radiation source.
Ifyou estimate you will be approximately 3 feet from the source when you perform the valve alignment check,
which ONE of the below is the correct estimate of the radiation field you will be exposed to?
a. - 30 mrem /hr
b. 75 mrem /hr .-
c. 150 mrem /hr
'd. 300 mrem /hr
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69. Which ONE of the conditions'or situations described below would require action to be taken in one hour or less in
accordance with Technical Specifications?
a.-
A Cold Leg Accumulator boron concentration is reported as 2250 ppm while operating in Mode 1
b. While operating in Mode 2 a Main Steam Line Safety valve is found leaking and must be. gagged
c. AFD is determined to be outside the target band for more than one hour while operating at 100% power
d. UNIDENTIFIED LEAKAGE from the Reactor Coolant System is detennined to be 7 gpm with the unit
in Mode 3
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70. If the Train B CCW pump trips, which of the following is a required initial action of ABN-5027
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, a. Verify adequate RCP Thermal Barrier Hx cooling flow
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-.b. - VerifyTrain B SSW pump did not tnp ;
c. Verify Safety Chiller Recire pump u-05 is running
-
d.
-
' Verify CCW Hx outlet flow is less than 17,500 gpm '
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. 71. Unit 1 is operating at 100% power. A Plant Computer alarm occurs for RCP l-01, The Reactor Operator observes
the followmg parameter: ,
Motor Stator Winding Temperature 270'F
Motor Upper Radial Bearing Temperature 160'F
Mptor Upper Thrust Bearing Temperature 163 F ,
Lower Seal Water Bearing Temperature 240 *F
Shaft Vibration 12 mils ,
Frame Vibration 2 mils
Which of the following indicates the reason the operator must 1:ip the reactor:
,
a. Motor Stator Winding Temperature High
b. Motor Upper Thrust Bearing Temperature High
,
c. Lower Seal Water Bearing Temperature High
d. Shaft Vibration High
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72. Unit 1. is operating with the following conditions:
- Reactor power 50%
- * 2 condensate pumps running
e 2 cire water pumps rtmning
~
e All other systems and components in automatic
In this situation, which ONE of the following conditions would result in a trip of a main feedwater pump
. assuming no operator actions are taken?
a. A spurious reactor trip
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b. A condensate pump trips on overcurrent
c. A selected SG level channel fails low
d. A heater drain pump trips on overcurrent
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73. Unit 2 is operating at 200 MWe when control room operators observe a simultaneous trip of the main turbine
generator and the running main feedwater pump.
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Which ONE of the below plant conditions would have resulted in this transient?
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a. Condenser vacuum degraded to 20 inches
b. The operating Main Feedwater pump tripped l
c. A Ili-Ililevel occurred in one SG
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d. Loss of all condensate pumps occurred l
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- 74. While conducting the procedure to align a Main Feedwater pump for warmup, an RO encounters a problem with -
- the procedure content. Upon subsequent review it is determined that a required change to the SOP which does not
change the intent of the procedure and is needed immediately.
Which ONE of the following identifies the person (s) who may review and approve this laterim Approved
4
Change? - ,
,
a. - A non-licensed operations staff person knowledgeable in the related area and the on-duty RO.
b. The on-duty US and the SRO assigned to the CPC.
. c. . The on-duty RO and the Operations Manager,
d. The Duty Manager.
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75. A plant incident occurs on your shift requiring declaration of a General Emergency. I
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Acting as the Emergency Coordinator, which ONE of the following can you NOT delegate in this situation?
a. Conduct personnel accountability within the Protected Area
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b. . Making Protective Action recommendations to the offsite authorities '
c. Direct requests for corporate support to the Executive Vice President, Nuclear Operations
'd. Approve shift schedules that support long-term emergency response i
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- 76. During operation at power an alarm is received indicative of an introduction of smoke from outside into the
Control Room area.
For this situation, which ONE of the below indicates the correct operator action in response to this condition?
a. . Manually shift the Control Room ventilation system to the Ernergency Recirculation Mode
b. - Ensure the Control Room Ventilation system automatically shifted to the Emergency Recirculation Mode
c. . Manually shift the Control Room ventilation system to the Isolation Mode
.
d. Ensure the Control Room Ventilation system automatically shifted to the Isolation Mode
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77.' Unit I was operating at 100% power when a loss of off-site power occurred. Reactor core delta-T was
approximately 60 *F at the time of the reactor trip.
Which ONE of the below describes the response of core delta-T over the next several hours assuming off-site
power is not restored and no plant cooldown is initiated?
a. . Decrease over time from an initial value of 50-60 *F established shortly after the trip
b. Remain constant at 50-60 *F as both decay heat and r.atural cliculation flow decrease
- c. Increase over time as natural circulation flow is reduced due to decay heat lowering
d. Remain constant at 50-60 *F due to RCS and SG temperatures being maintained constant
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78. The following plant conditions exist:
- The unit has been tripped for 20 minutes
- IR channel N35 has decreased and stabilized at 1.2E-11 amps.
- IR channel N36 has decreased and stabilized at 9.9E-9 amps.
Which ONE of the following describes the probable cause and action to be taken for these conditions?
a. IR channel N35 is over compensated; continue with the shutdown.
~ b. IR channel N35 is under compensated; unblock the Source Range nuclear instruments.
c. - IR channel N36 is under compensated; unblock the Source Range nuclear instruments.
d. IR channel N36 is over compensated; continue with the shutdown.
. - . . . . - . . . _ -.. . -. . . . . - . . . - .
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79. The applicability statement for EOP-0.0A, " Reactor Trip or Safety injection", states this procedure is used for
events occurring in Modes 1,2, and 3.
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Which ONE of the below describes the applicability of EOP-0.0A while in Mode 4?
a. EOP. 0.0A cannot be used in Mode 4, EOS-0.0A, "Rediagnosis", should be used in this situation
.
- b.
- EOP-0.0A can be used if a step by step evaluation is made to determine if each step or action is applicable
c. EOP-0.0A can be used in Mode 4 only if directed by the Critical Safety Function Status Trees (CSFSTs)
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- d. EOP-0.0A cannot be used in Mode 4; FRS-0.2," Response to Inadequate Shutdown Margin" should be ;
used in this situation
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- 80. Which of the following specifies the ODA-102," Conduct ofOperations" shift manning requirements, given that
Unit I is in mode 3 and Unit 2 is in mode 47 Assume individuals filling given positions are not qualified to fill
multiple positions.
1-SM,
'
a. 2-US, 4-RO, 8-PEO, 1STA
[ b. None-SM, 3 US, 3-RO, 10-PEO, 1-STA
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c. 1-SM, 3-US, 4-RO, 5-PEO, None-STA
d. I-SM, - 2 US, 3-RO, 3-PEO, 1-STA
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81. Dissolved hydrogen in the Reactor Coolant System (RCS) is found to be outside normal limits, if a chemistry
remedial action level 2 has been reached, which ONE of the following actions is appropriate?
'
a. Shutdown the reactor and cooldown the plant to 250 F as rapidly as plant conditions will permit
b. Initiate corrective action to return the hydroger parameter to normal limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i
. c. - Immediately begin increased monitoring and trending of the RCS hydrogen parameter '
d. 1
- Initiate corrective action to return the hydrogen parameter to normal limits within 7 days ,
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82. During performance of ECA-0.0, " Loss of All AC Power," the SGs are depressurized to 270 psig. What is the
purpose of the depressurization? .
. a. Reduce RCS temperature to prevent inadvertent criticality.
'
b. Reduce decay heat load to miniinize possibility of SG dryout,
c. Reduce SG pressure and temperature to prevent chemical hideout return.
d. Reduce RCS pressure to reduce seal leakage and minimize RCS inventory loss.
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83. Procedure FRH-0.1 A, " Response to Loss of Secondary IIeat Sink", directs operators to stop all RCPs if actions to
restore AFW flow are not successful.
Why are the RCPs stopped in this situation?
,
a. Minimize the possibility of a tube rupture when AFW is eventually restored to the steam generators
b. Conserve reactor coolant inventory by reducing seal leakoff
c;
Obtain increased safety injection flow by decreasing RCS cold leg pressure
d. Conserve steam generator inventory by reducing RCS heat input
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84. While refueling, the water level has decreased to 21 feet above the reactor vessel flange and containment area
radiation monitors indicate increased radiation. Which ONE of the following describes the Tech Spec action
required for this condition? .
a. Place at least two RHR loops in operation.
'
- b. Raise cavity water level to at least 22 feet ab'ove the reactor vessel flange.
c. . Suspend all operations involving moyen ent of fuel assemblies within containmei.t. !
. d. , Open the transfer tube gate valve to equalize cavity and transfer canal water levels.
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85. Uait 2 is experiencing a rapid loss ofinstrument Air header pressure. If a complete loss of header pressure !
occurs, control of which ONE of the following valves will be immediately lost?
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a. AFW flow controlvalves
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c. SG Atmospheric relief valves
' d. - Safety Chilled Water Condenser CCW regulating valves
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' 86. In accordance with Technical Specifications, which ONE of the following conditions will result in the declaration
of an INOPERABLE control rod?
.
a. AFD exceeding operating limits
b. - A rod bottom light remains extinguished after a trip
c. One Contro'l Bank D rod trippable but cannot be moved electrically
d. 'One Control Bank D rod indicates 210 steps with Control Bank D demand at 200 steps.
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87. A plant event has occurred which resulted in an inadvertent 51 signal.
. Operators have transitioned to EOS-1.1," Safety injection Termination", and are preparing to reduce Si flow.
Which ONE of the below ECCS pmnps is the FIRST to be stopped in the procedure?
a.' RHR pump . .
. b. Charging pump
. .c. _ Si pump
- d. Varies based on exact plant response
.
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> 88. Unit 1 is operating at 80% power with all control systems in automatic when the operators observe control rods i
stepping in at 72 steps per minute for no apparent reason. ,
- Which ONE of the following should be investigated as a potential cause for the unwarranted rod motion?
> a. A Tave channel failed low ,
b. Turbine lleader Pressure PT-505 failed high
i
c. Loss of vital instrument bus IPCI
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d. A power range channel failed low
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89. Which ONE of the following is a requirement for the release of a Waste Gas Decay Tank?
,
a. . A g,u ,mlease shall not be initiated if the oxygen content exceeds 1%
-
b. A Waste Las Decay Tank release shall be completed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of sampling the tank
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i c. No containnent release may be in progress during a Waste Gas Decay Tank release
d. The original release permit may be used to reinitiate the release following a valid hi rad automatic
termination
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90. Unit 2 is operating at 75% power when a trip and SI occur. Forty minutes after the trip the following conditions
are indicated:
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All SG levels 40% NR
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Containment pressure 4 psig
-
RCS subcooling 0F i
- Power Range Nis 0% l
-
Intermediate Range SUR -0.1 dpm j
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CETs 225'F
- RVLIS 11" above core plate light lit
i
Which ONE of the following Critical Safety Function paths is in effect? .
a. Transition to FRS-0.lB, " Response to Nuclear Power Generation /ATWT" l
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b. Transition to FRC-0.lB," Response to inadequate Core Cooling"
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c. Transition to FRil-0.lB," Response to Loss of Secondary Heat Sink"
d. Transition to FRP-0.1B," Response to imminent Pressurized Thermal Shock Condition"
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91. A reactor startup is in progress on Unit 2. Operators are performing steps oflPO-002, Plant Startup,
in accordance with the procedure, which ONE of the below is the expected Source Range counts when the
Intermediate Range instruments comes on scale?
a. . SE2 cps ,
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b. .5E3 cps
c. SE4 cps
d .- ' SES cps -
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92, A 6.9 KV breaker, normally controlled from the control room is racked in and closed when DC control power to
the breaker is lost.
Which ONE of the below is correct concerning the present situation?
a. The breaker can be rem.otely opened once from the control room
b. The breaker can only be opened locally at the cubicle
c. All protective trip features on the breaker remain functional
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d. The breaker must be racked out of the cubicle to open the breaker
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93.' At 1200 on June 12,1998, with the unit in Mode 1, it is determined that a required surveillance on a Technical
Specification contponent was not performed within the required time schedule. The ACTION statement for the
component requires the unit to be placed in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if found INOPERABLE.
If all requirements of Technical Specifications are adhered to, what is the latest date and time the unit must be in ;
Mode 37 ,
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a. 1800 on June 12
b. 0000 on June 13
c. 0600 on June 13
d. 1800 on June 13
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94. Which ONE of the below fire suppression systems is used to mitigate a fire in the cable spreading room?
a. . Halon' system
b. Deluge-type system
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t. c. Wet pipe sprinkler system
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- d. Wet standpipe system
.
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95. Unit 2 is operating with the following plant conditions when a PZR level is observed rising above program:
-
50% power _ .
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All control systems for PZR pressure and level control in automatic
-
Rod control in manual
-
Train A.CCP in service j
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No plant power changes in progress i
In this situation, which ONE of the below could be the cause of the deviation alarm?
a. -Tcold channel TE-421B failed high
b. PZR pressure controlling channel PT-455 failed low
c, _ Power Range channel N41 failed high !
d. PZR level controlling channel LT-459 failed high
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96. While operating in Mode 3, a mechanical fault on the Condensate Storage Tank (CST) has emptied the tank.
- AFW operation is required to maintain a heat sink. Which ONE of the below correctly lists alternate methods of
providing a source of water for AFW in this condition?
a. Install prefabricated spool pieces from the Service Water system to the suction of the AFW pumps.
b. Establish mak'eup to the CST using existing valves and piping from the fire water system
c. - Install prefabricated hoses and fittings from the Demineralized Water storage system to makeup to the
- CST.
d. Establish a suction source to AFW from existing valves and piping from the Service Water system
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97. An I&C Technician reports to the control room that he will be taking a liquid process radiation monitor off-line to
.
perform calibration checks in accordance with a scheduled shift work package,
Which ONE of the below indications will be available on the control room radiation monitor CRT that the
monitor has been taken off-line?
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4 : a. The associated monitor will be displayed in yellow '
' b. A background magenta border will be flashing for the monitor !
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c. The monitor will be displayed in green but will not respond to commands
d. The monitor designation will be displayed in white -
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98. Which ONE of the following design features provides the interlock that prevents raising irradiated fuel with the
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b. A weight sensing device ,
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- c. The upper travellimit switch
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d.' The' lower travel limit switch ;
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99. Unit 2 is operating with the following plant conditions when a PZR high level deviation alanu is received:
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10% power :
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~ All control systems are in automatic except Rod Control is in manual
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Train B CCP in service
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a a. Main Steam Header pressure PT-507 failed low
b. Main Steamline pressure PT-508 failed low
.. c. Turbine impulse pressure PT-506 failed high
- d. ' Turbine impulse pressure PT-505 failed high
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100.An event has occurred and operators are conducting EOP-1.0, " Loss of Reactor or Secondary Coolant". At the
appropriate transition pomt, operators are directed to perform an alignment to the Hot Leg Recirculation Mode.
Which UNE of the bel w indicated to the operators that the transition was required?
a. A predetermined reactor vessel level ,
- b. The amount ofinventory remaining in the RWST
- c. The amount of time elapsed during the accident
d. An RCS boron value obtained via PASS sampling
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