ML20154L410: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot insert)
 
(StriderTol Bot change)
 
Line 1: Line 1:
#REDIRECT [[IR 05000498/1998301]]
{{Adams
| number = ML20154L410
| issue date = 10/09/1998
| title = Forwards NRC Operator Licensing Exam Repts 50-498/98-301 & 50-499/98-301,including Completed Graded Tests Administered on 980602-04
| author name = Hurley L
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name =
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000498, 05000499
| license number =
| contact person =
| document report number = 50-498-98-301, 50-499-98-301, NUDOCS 9810190282
| document type = INTERNAL OR EXTERNAL MEMORANDUM, MEMORANDUMS-CORRESPONDENCE
| page count = 104
}}
See also: [[see also::IR 05000498/1998301]]
 
=Text=
{{#Wiki_filter:. -    - - . - . _ . . . . - - . . . . . . ~ . - - . - . . , . .
pr                                                                            .
      fuzu
  y              g                                  UNITED STATES
  7              g                    NUCLEAR REGULATORY COMMISSION
  O              8                                    REGION IV
                                              611 AYAN PLAZA DRIVE, suite 400
    e            [
      +,p***** g
                                                ARLINGTON, TEXAS 76011-8064
l                                                    October 9,1998
l        NOTE TO:        NRC Document Control Desk
!                          Mail Stop O-5-D-24
          FROM:            Laura Hurley, Licensing Assistant
l                          Operations Branch, Region IV
l        SUBJECT:        OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON
l'                        JUNE 2-4,1998, AT SOUTH TEXAS PROJECT
                            DOCKETS #60 498/499 '
                                                                                                                                                      .
          On' June 2 4,1998, Operator Licensing Examinations were administered at the
l
<
        - referenced facility. Attached you will find the following information for processing
          through NUDOCS and distribution to the NRC staff, including the NRC PDR:
,
          item #1 - a) Facility submitted outline and the initial exam submittal for distribution
!
                            under RIDS Code A070.
,                      b) As given operating examination, designated for distribution under RIDS
!
                            Code A070.                                                                                                    .. fus, sis
l                                                                                ,p                                                  n:'
!          Item #2 '- -    Examination Report with the as given written examination attached,
;                          designated for distribution under RIDS Code IE42.
i
i
l
          If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations
l          Branch, Region IV at (817) 860-8253.
l
('
I
l
I
;
                V
                          A      b    PDR
                                          99
                                                  (
I
L
\
r
(                                                                                          -
 
          .                  -    -.  -    .    -          - . . -    . - - ~    ~-          .        . .-
                                                                                                                    !
j      odf' '"- g
                .
                                                    UNITED $T ATES
                                                                                                                    '
    f..          . '$,                NUCLEAR REGULATORY COMMISSION
            "
              *
    $                  k                              REGION IV                                                    ;
f    $                                    611 RY AN PLAZA ORIVE. SUITE 400
l
-
      %..''.a/
          ..
                                              AR LINGTON. T E XAS 76011 8064
                                                    sut i4 id                                                      !
i
I
L
        William T. Cottle, President and
            Chief Executive Officer                                                                                a
      ' STP Nuclear Operating Company
l        P.O. Box 289
        Wadsworth, Texas 77483
        SUBJECT: NRC INSPECTION REPORT 50-498/98-301; 50-499/98-301
l        Dear Mr. Cottle:
i
l        From June 1-4,1998, an operator licensing certifi:ation inspection was conducted at your South
        Texas Project Electric Generating Station, Units I and 2 reactor facilities. The enclosed report
        presents the scope and results of that inspection
        This inspection included an evaluation of six appli; ants for senior operator licenses. We
,
        determined that five of the six applicants satisfied the requirements of 10 CFR Part 55 and the
;        licenses have been issued.
        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),
        Should you have any questions concerning this inspection, we will be pleased to discuss them
        with you.
l                                                        Sincerely,
                                                              :D
!                                                        John L. Ptsliet, Chief                        l
                                                                                                            /
                                                        Operations Branch                                I
                                                                                                          '
                                                        Division of Reactor Safety                      ,
          Docket Nos.: 50-498;50-499
          License Nos.: NPF-76; NPF-80
  :
i
4
  !
                          '
                                                                                                                    5
                            ,
 
                . . .  _    . _ _ _        .. _.    . . __ . . _ . . . . _ _
  ?
!
i-
!
      STP Nuclear Operating Company              -2-
                                                                              l
      Enclosure:
      NRC Inspection Report
      50-498/98-301; 50-499/98-301-
      cc w/ enclosure and Attachments 1-4:
      Ken Coates, Manager, Nuclear Training                                    !
    ' STP Nuclear Operating Company
      P.O. Box 289                                                            I
    Wadsworth, Texas 77483
    - cc w/ enclosure and Attachments 1-3:
      Lawrence E. Martin, Vice President
      Nuclear Assurance & Licensing
    STP Nuclear Operating Company
    P.O. Box 289
    Wadsworth, Texas 77483
i                                                                            ;
,
'
                                                                              1
    A. Ramirez/C. M. Canady                                                  ;
    City of Austin
'
L
l
    Electric Utility Department
    721 Barton Springs Road
    . Austin, Texas 78704
    Mr. M. T. Hardt/Mr. W. C. Gunst
    City Public Service Board
    P.O. Box 1771
    San Antonio, Texas 78296
    D. G. Tees /R. L. Balcom
    Houston Lighting & Power Company
    P 0. Box 1700
    Houston, Texas 77251
    Jon C. Wood
    Matthews & Branscomb
    One Alamo Center
      106 S. St. Mary's Street, Suite 700
    San Antonio, Texas 78205-3692
    Jack R. Newman, Esq.
      Morgan, Lewis & Bockius
;    1800 M. Street, N.W.                                                    ;
'
    Washington, D.C. 20036-5869                                              !
-
                                                                              .
!                                                                              .
f
l                                                                              l
l                                                                              ,
                                                                              '
>
 
  .  ..    .    .    ..      . _ . . . . - - . .. -. _ - - . ... . . .._.. _ ..  _ . . _  .
i      STP Nuclear Operating Company            -3-
                                                                                              .
;
'
      Mr. G. E. Vaughn/Mr. C. A. Johnson                                          '
                                                                                                  ,
      Central Power & Light Company
      P.O. Box 289                                                                                ;
      Mall Code: N5012
l    - Wadsworth, Texas 77483
      INPO
      Records Center-
      700 Galleria Parkway                                                                        ,
      Atlanta, Georgia 30339A957
      Bureau of Radiation Control
      State of Texas                                                                              ,
      1100 West 49th Street
      Austin, Texas 78756
      Mr. Jim Calloway                                                                            .
    . Texas Public Utility Commission
      William B. Travis Building
      1701 North Congress Avenue
      P.O. Box 13326
      Austin, Texas 78701-3326
      John Howard, Director
      Environmental and Natural Resources Policy
    - Office of the Governor
      P.O. Box 12428
      Austin, Texas 78711
      Judge, Matagorda County
      Matagorda County Courthouse
      1700 Seventh Street
      Bay City Texas 77414
1
I
I
1
:
l
:
                                                                                                  Y
 
  , .                                      . .        ~      .      . ~ . -_.        -..      -                - -      . - -  . ~ . .
r                                                                                                                                        .
l
L      STP Nuclear Operating Company                              -4-
l
                                                                                                                                ~
l      E-Mail report to T. Frye (TJF)
;.    E-Mail report to D. Lange (DJL)
      E-Mail report to NRR Event Tracking System (IPAS)
      E-Mail report to Document Control Desk (DOCDESK)
      bec to DCD (IE01)(IE42)
,
      bec distrib. by RIV w/ enclosure and Attachments 1-3:
      Regional Administrator                                  Resident inspector
      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)                                  R. Bachmann, OGC (MS: 15-B-18)
      bec w/ enclosure and Attachments 1-4:
      R. Gallo (HOLB/NRR) (MS: 9D4)
      L. Hurley
                                                                                                                                            I
                                                                                                                                          i
                                                                                                                                            i
                                                                                                                                            i
      DOCUMENT NAME: R:\ stp\st8301rp.hfb                                                                                                  ;
      To receive copy of document, indicate in box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy
l        RIV:SRl(:OQ / a      ,
                                      RE:OB        .    N 'C:OB              d    C:PBA _ o #            C:OB            g                )
l        HFBundf(llVb4 JF            RELantz '      k        JLPellet M            JITapia %), JLPellet g                                  !
'
                            "                                                      07/[y98 l ')          07/j98 # ~
        07#l/98 \                    078/98      1
                                                              07M/98 *
            '                                                                                  V
                                                    OFFICIAL RECORD COPY
                                                                                                                                            l
:
!
 
              .-.      .-        . __        .-    _ ._
                                      ENCLOSURE
                      U.S. NUCLEAR REGULATORY COMMISSION -
                                        REGION IV
Docket Nos.:        50-498;50-499
License Nos.:      NPF-76; NPF-80
Report No.:        '50-498/98-301; 50-499/98-301
Licensee:          STP Nuclear Operating Company
Facility:          South Texas Project Electric Generating Station, Units 1 and 2
Location:          FM 521 - 8 miles west of Wadsworth
                    Wadsworth, Texas
Dates:              June 1-4,1998
Inspectors:        Howard F. Bundy, Chief Examiner
                    Ryan E. Lantz, Reactor Engineer, Examiner / inspector
Approved By:        John L. Pellet, Chief, Operations Branch
                    Division of Reactor Safety
ATTACHMENTS:
Attachment 1:    Supplemental Information
Attachment 2:    Simulation Facility Report
Attachment 3:    Post-Examination Comments
Attachment 4:    Final Written Examination and Answer Key
  [[~}                    i
 
      _      _    .        . _ _ _ _ _ _.            _    _    _ . _ _ _ _ _ .            __      _    __ _
,.
                                                                                      -
1
l
                                                        2-
                                          E'XECUTIVE SUMMARY
                      South Texas Project Electric Generating Station, Units 1 and 2
                          NRC Inspection Report 50-498/98-301; 50-499/98-301
    NRC examiners evaluated the competency of six senior operator applicants for issuance of
  operating licenses at the South Texas Project Electric Generating Station, Units 1 and 2. The
    licensee developed the initial license examinations using the guidance in NUREG-1021, Interim
    Revision 8. January 1997. NRC examiners reviewed and approved the examinations. The
  initial written examinations were administered to all six applicants on -June 30,1998, by facility
  proctors in accordance with the guidance in NUREG-1021, Interim Revision 8. The NRC
  examiners administered the operating tests June 2-4,1998.
  Ooerations
  .        Five of the six applicants for senior operator licenses passed their examinations.
            Applicants demonstrated effective oversight and good communication techniques during
            the dynamic scenarios. A common applicant performance weakness involving clearance
            order review was identified in that the majority of the applicants failed to or were slow in
            identifying that component cooling water had not been designated for isolation for
            overhaul of a residual heat removal pump (Sections 04.1 and 04.2).
  .        The examination submitted was adequate for administr$ ion and required only limited
            enhancement and editorial corrections. The senior operators assigned for examination
            validation provided valuable enhancement suggestions. The licensee staff was highly
            responsive to incorporating enhancement suggestions developed during the review
            process (Section 05.1).
!
,
                                                                                                              ,
 
    . . _ _ . _      m    . _ _ _ _ _ _ . _ . _ _ _ _ _ _                          - _ _      . . . _ . . _ _ _
                                                                                                                  . _ _ . . . _!
{                                                                                                                                ,
L
l
!.
!
                                                                    3-
                                                                                                                                  l
                                                            Report Details
                  Summarv of Plant Status -
                  Both units operated at.100 percent power for the duration of this inspection.                                  l
{                                                                -
                                                                                                                                l
l                                                            l. Operations                                                      l
l
p                04      Operator Knowledge and Performance
                  04.1    Initial Written Examination
l                ' a.    Inspect:on Scone
                          On June'30,1998, the facility licensee proctored the administration of the written
l                        examination, approved by the chief examiner and NRC Region IV supervision, to six
l                        individuals who had applied for upgrading their reactor operator licenses to senior
l                        operator licenses. The licensee proposed grading for the written examinations and
l-                      E evaluated the results for question validity and generic weaknesses. The examiners
                          reviewed the licensee's results.                                                                      .
                                                                                                                                j
L                  b..    Observations and Findings                                                                              :
                          Five of the six applicants passed the written examination. Written examination scores
                          ranged from a low of 73 to a high of 91 percent with an average of 84 percent overall and            .!
  ~
                          a lowest passing score of percent. Greater than 50 percent of the applicants missed                    l
l                        16 questions. Post-examination review by the licensee indicated that the questions-
l                        missed were primarily due to isolated knowledge and training weaknesses. However, no
                          broad scoped commonalities were determined from the missed question analysis. The                      l
                        . licensee's post-examination review resulted in the recommendation that the answer for                i
;.                        Question 57 be changed to accept both choices 'A' and 'D' as correct. The proposed
'
                          change was accepted based on the technical merits and the examinations were
                          evaluated incorporating this change. The licensee's post examination ' comments are                    ;
                          included in Attachment 3.'                                                                            t
                  c.    Conclusions                                                                                            )
l                        Five of the six license applicants passed the written examinations. No broad knowledge
!                        or training weaknesses were identified as a result of evaluation of the graded
                          examinations.
L
I
,
)
t-
                                                                                                                                I
                                                                                                                              J
 
                            -                          .        -    .                        _
                                                -4-
04 2 initial Ooeratina Tesj
a.  Insoection Scoce
    The examination team administered the various parts of the operating examination to the
    six applicants on June 2-4,1998. Each applicant participated in three scenarios.
    However, each applicant was formally evaluated for senior operator license only for the
    scenario in which he participated as the control room senior operator in charge. Each
    applicant also received a control room and facilities walkthrough test, which consisted of
    five tasks with followup questions, and an administrative test, which consisted of five
    tasks in four administrative areas.
b.  Observations and Findinas
    All applicants passed all sections of the operating test. The examiners noted appropriate
    use of peer and self-checking practices in all areas of the examinations. When
    evaluated in the senior operator in charge position for the dynamic scenarios, all
    applicants demonstrated effective oversight and good communication techniques. The
    applicants displayed effective application of technical specifications and emergency and
    abnormal procedures.
    The applicants generally performed well on the systems and facility walkthrough and
    administrative tasks. However, on one administrative task the majority of the applicants
    failed to or were slow in identifying that component cooling water had not been
    designated for isolation as reviewed during review of a clearance order for overhaul of a
    residual heat removal pump.
c.  Conclusions
    All six applicants passed all parts of the operating test. Applicants demonstrated
    effective oversight and good communication techniques during the dynamic scenarios.
    The exarniners identified a common applicant performance weakness involving
    clearance order review in that the majority of the applicants failed to or were slow in
    identifying that component cooling water had not been designated for iso!ation for
    overhaul of a residual heat removal pump.
05    Operator Training and Qualification
05.1  Initial Licensino Examination Develooment
      The facility licensee developed the initiallicensing examination in accordance with
      NUREG-1021, Interim Revision 8. " Operator Licensing Examination Standards for Power
      Reactors."
 
              -      -              -        - -.      .          .        -_        .
I
                                                        5
    05.1.1 Examination Outline
  ..  a.    Insoection Scoce
            The facility licensee submitted the initial examination outline on December 17,1997.
            The chief examiner reviewed the submittal against the requirements of NUREG-1021,
            Interim Revision 8.
      b.    Observations and Findinas
            Region IV approved the initial examination outline with minor comments for
            enhancement, which were promptly resolved, and advised the licensee to proceed with
            examination development.                                                                      i
      c.    Conclusions
            The licensee submitted an examination outline in a timely manner, which required only
            minor revisions for enhancement.
    05.1.2 Examination Packs,1g                                                                          i
      a.    Insoection Scooe
            The facility licensee submitted the completed draft examination package on
            February 27,1998. Prior to formal submittal of the examination package, meetings were        I
            held in the Region IV office on January 27 and February 12,1998, to discuss
            examination development issues. These meetings were attended by licensee
            representatives J. Calvert, Operations Training Manager, and M. DeFrees, Licensed
            Operator Training Supervisor, and the chief examiner and operations branch chief. The
            chief examiner reviewed the submittal against the requirements of NUREG-1021, Interim        .
            Revision 8. An onsite review of the revised examination was conducted during the              l
            period March 20-23,1998.                                                                      !
                                                                                                          I
      b.    Observations and Findinas                                                                      !
                                                                                                          !
            The draft written examination contained 100 questions. The questions were                      l
            predominantly new for this examination. The draft examination was considered
            technically valid, to discriminate at the proper level, and responsive to the sample plan      ,
            submitted by the licensee on December 17,1997. The chief examiner provided                    l
            enhancement suggestions on 11 questions, which were appropriately incorporated by
            the licensee. The suggested enhancements generally related to clarity of the question
            stem and distractor plausibility. The chief examiner also identified an error in the ans. r
            for original Question 73, which related to the definition of core alterations. The license
            replaced this question and initiated a station condition report to investigate a potential
            discrepancy in fuel handling procedures, in addition, the licensee changed the wording
                                                                                                        ]
 
                                                                                            ._
                                            6-
    on several other questions in response to a generic chief examiner comment regarding
    minor grammatical errors. Finally, prior to administration, the licensee modified
  Question 37 to reflect deletion of the original reference and enhanced the distractors on
  Question 48. Minor editorial changes were made to 4 other questions.
  The operating test consisted of dynamic scenarios, administrative tasks, and system
  tasks with followup questions. The licensee submitted four dynamic scenarios, including
    one backup scenario, which was not used during the examination. The submitted
    scenarios conformed to NUREG-1021. The licensee subsequently incorporated several
  minor enhancement suggestions provided by the chief examiner and the licensee
  validation crew during the week of March 20,1998.
  To support the administrative section of the operating test the licensee submitted five
  administrative tasks. Although the submitted tasks were satisfactory test items,
  enhancements, and clarifications were necessary for each task in response to chief
  examiner questions and comments. Also, during the onsite review it was discovered the
  answer for Job Performance Measure A3, " Review a Tagout," was not complete and
  further changes were necessary, in addition, the onsite validation time for Job
  Performance Measure A3 was in excess of one hour, which required a reduction in task
  scope. Enhancements were made to otherjob performance measures during the onsite
  review.
  To support the systems walkthrough section of the operating test, the licensee submitted
  five system job performance measures with two followup questions associated with each.
  The chief examiner provided only one editorial comment on the job performance
  measures. However, several enhancements were made to the job performance
  measures during the onsite review. Also, it was necessary to modify the cues to clarify
  the scope of the task for Job Performance Measure 3," Place a Class 1E 125V DC
  Battery Charger in Service."
  Many of the enhancements made to the operating test during the onsite review were
  directly attributable to comments and suggestions by the validation crew, which was
  comprised of senior operator members of an off-duty shift crew.
c. Conclusions
  The examination submitted was adequate for administration and required only limited
    enhancement and editorial corrections. The meetings with licensee training
    representatives in the Region IV office were instrumental in achieving a satisfactory
    examination submittal. The senior operators assigned for examination validation
    provided valuable enhancement suggestions. The licensee staff was highly responsive
    to incorporating enhancement suggestions developed during the review process.
 
    - _. _ . - _ - _ _ . . . _ . _ . - . _ _ _ . _ _ . _ . _ _ , _ _                              - . . _    .- _ __ ._.
                                                                                                                            :
                                                                                                                          ,
.J
                                                                      7-
                                                                                                                          )
            05.2. Simulation Facility Performance -
            a.          Insoection Scone
*
                                                                    .
                                                                                                                          ;
                      - The examiners observed simulator performance with regard to fidelity during the                  ,
                        examination validation and administration.                                                        t
.
            b;          Observations and Findinas                                                                        :
                        The simulation facility supported the validation and administration of the examination
                        well. However, as discussed in Attachment 2, there were general failures on two
                        occasions for one scenario during the onsite review and an identical failure during                !
                        examination administration. The failures had minimalimpact on the review effort.
i                        However, the failure during examination administration required use of the backup                ,
                        scenario for one crew. No other fidelity problems were observed.
                                                                                                                          l
,
                                                                                                                          :
            c.          Conclusions
i                      The simulator and simulator staff supported the examinations well. General simulator
j.                    ' failures had minimal impact on examination review and slight impact on examination
  '
                        administration.                                                                                  l
  !
                                                                                                                          ;
          05.3          Examination Security                                                                              ;
'                                                                                                                          !
            a.          Scooe
                                                                                                                          I
*                        The examiners reviewed examination security both during on site preparation week and              l
,
                      - examination administration week for compliance with NUREG-1021 requirements.                      :
                                                                                                                          i
                                                                                                                          '
              b. .      Observations and Findinas
.                                                                                                                          i
;                                                                                                                          *
'
                        Members of the licensee's operations and training staff signed onto the NUREG-1021
                                                                                                                          '
,
                        approved examination security agreement acknowledging their responsibilities for
                        examination security. The licensee maintained secure areas for examination
:                        development, review, validation, and administration. Signs were conspicuously posted to
i
                        avoid inadvertent unauthorized access, and doors were maintained locked with good key
                        control to ensure proper access to sensitive areas. Applicants were maintained under              ,
l                        constant supervision and were always escorted to and from examination points.
                        Simulator security was strictly complied with.
,
            c.          Conclusiorts
j                        Effective examination security was maintained.
.
  4
_
.
:
*
                                                  -.
                                                                                                                        a
 
  . _  . . _ .    _          _    .-. _ _ _ _      _ _ . . _ . . _ _ . . ~ . . - _ . _ . _ ..    . . .
                                                -8-
                                        V. Management Meetings
    X1  Exit Meeting Summary
        The chief examiner presented the preliminary inspection results to members of the
        licensee management at the conclusitan of the operating test administration on June 4,
        1998. Final results of the inspection were presented by Mr. John Pellet during a
        telephone conference with Mr. Mike DeFrees on July 8,1998. The licensee
        acknowledged the findings presented.
                                                                                                          ,
                                                                                                          .
        The licensee did not identify as proprietary any information or materials examined during
        the inspection.
,
                                                                                                            l
                                                                                                            l
                                                                                                            l
                                                                                                            l
                                                                                                        ,
 
-                    -        .-      .      -.  --      - --.        . - -    .    .
                                                                                            . - - . .
                                                                                                      4
                                            ATTACHMENT 1
                                                                                                        '
                                    SUPPLEMENTAL INFORMATION
                                                                                                        '
                              PARTIAL LIST OF PERSONS CONTACTED
  - Licensee
    J. Calvert, Manager, Operations Training
  G. Chitwood, Examination Lead
    K. Coates, Manager, Nuclear Training
    M. DeFrees, Lead Instructor, Licensed Operator Training
  W. Dowdy, Unit 2 Operations Manager
  J. Lovell, Manager, Generation Support
  K. Struble, instructor, Licensed Operator Training                                                    l
  NEC
  G. Guerra, Resident inspector
                                                                                                          i
                                  INSPECTION PROCEDURE USED                                              l
  NUREG-1021              " Operator Licensing Examination Standards for Power Reactors,"
                          Interim Revision 8
 
  - --      -    -.    ..          .  _ .        . . _ - - - - . . - - - - -                - - - . - -    .
                                                                                                                  !
                                                                                                                !
                                                                                                                  ,
                                                                                                                  i
                                                    ATTACHMENT 2                                                  -
                                            SIMULATION FACILITY REPORT
        Facility Licensee:    STP Nuclear Operating Company .
        Facility Docket-      50-498; 50-499
        Operating Examinations Administered at: STP Electric Generating Station, Units 1 and 2
        Operating Examinations Administered on: June 2-4,1998 -                                                  r
        These observations do not constitute audit or inspection findings and are not, without further
        verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations
        do not affect NRC certification or approval of the simulation facility, other than to provide
        information, which may be used in future evaluations. No license action is required in response
i
        to these observations.
        Deficiency identified Durina Examination Preoaration
        .
                During scenario validation, RCFC 11 A outlet air temperature indicator was reading
                150' F when actual temperature was 62* F. Replaced sticking meter to correct
                erroneous reading.
        .
                The F.B. exhaust filter train isolation dampers took an unrealistically long time to close
                during realignment of F.B. VAC while performing POP 05-EO-EOOO, i.e.,13 to
                17 seconds versus 7 seconds expected in the plant. The stroke time was adjusted in
                the simulator.
        .
                During a LOCA the simulator stopped responding and the audible simulator trouble alarm
                actuated on two occasions. On the third attempt, the simulator performed as expected.
        Deficiency identified Durina Examination Administration
        .
                Despite the simulator staff's unsuccessful attempts to reproduce the simulator failure
                several times prior to examination administration, it recurred at the identical point in the
                LOCA scenario during applicant evaluation.
l
1
l
l:
l
l
l
!
l
                            .    - .                .                          .        .. .-              - -
 
                                ATTACHMENT 3
                            SOUTH TEXAS PROJECT
                            WRITTEN EXAM (6/30/98)
                    POST-EXAM APPLICANT COMMENTS
QUESTION 57 ANSWER 'D'
REFERENCES:    OPOP05-EO-F002, Core Cooling Critical Safety Function Status Tree
                OPOP05-EO-FRC3, Response to Saturated Core Cooling, Step 2
COMMENT:        Use of High Head injection (Answer 'S') is also an ' appropriate method' to
                use under the specified circumstances.
RESOLUTION:    Comment accepted. Accept choices 'A' and 'D'. Choice 'A' was originally
              thought to be incorrect based on not being applicable unless RCS
              pressure was below 1745 psig. The question stem specified an RCS
              pressure of 1785 psig. However, OPOP05-EO-FRC3, Response to
                                      .
              Saturated Core Cooling, Step 2b, has the operators establish HHSI flow
              regardless of w? ether pressure is above or below 1745 psig. The
              importance of being above or below 1745 psig is related to whether
              charging flow should be established, not HHSI flow.
                                                                        ,/
                                                                      /
 
                  e            A  A 4 a w - - a e. . 4 a
l
1
l
                    ATTACHMENT 4
l
    FINAL WRITTEN EXAMINATION AND ANSWER KEY
l
l
l
                                                          >
<
> .
 
    ,  -                -.                    -          . -            . . - . -    -
                                                                                                  _ , _ - - . .      .,
                                                                                                                                                                ,
                                                                                                                                    ,
                                                                                                    *                ~'                                '
                                                                                                                                  ll'    ;                [ f.
i                                                                                                            .                      .
:                                                                                                                                      -                  ,
                                                                                                                                            .
                                                                                                                            *                                        !
                                                                                                                                                  .t'
                                                                                                                                                                      ,
l
                                                          STP SRO EXAM KEY                                                                      :
                                                                                                                                                        sc ..:l
                                                                                                                                                          . :. .
I
'
                                                                                                                                                                      ;
                                                                                                                                                  4-
;
                                                                                                                                              .
                                                                                                                                                                      l
            1)'      A                                                35) D                                                    69)      B                            !
            2)      C                                                36) A                                                    70)      B                            l
l-          3)      D                                                37) A                '
                                                                                                                                71 ) -.  A'                          !
            4)      B                                                38) A                                              .    72) B                                  l'
            5)      .D                                                39) C                                                    73)' D'
            6)      A                                                40) D                                                    74) B
            7)      B                                                41) A                                                    75) -C                                l
!
            8)      C                                                42) B                                                    76) A
            9)      D                                                43) C                                                    77) C                                j
            10) C                                                    44) D                                                    78) C                                  i
            11) A                                                    45) C                                                    79) B
            12) D                                                    46) C                                                    80) A                                  !
            13)-B                                                    47) B.                                                  81) B                                  !
            14) C                                                    48) B                                                    82) D                              'i
            15) B                                                    49) D                                                    83) C                                  l
l-          16) A                                                    50) C                                                  '84)      B                          j
            17) D                                                    51) B                                                    85)      C                        '{
            18) D                                                    52) D                                                    86)      B                            ;
.
            19) A                                                    53) C                                                    87)      C                            l
            20) A                                                    54) C                                                    88)      C
                                                                                                                                                      *
                                                                                                                                                                      l
            21) A                                                    55) B                                                    89)      C                            !
            .22) D                                                    56) B                                                    90)      D                            i
            23) A                                                    57) D or A                A pta r                        91)      g                          j
            24) D                                                    58) B                                                    92)      C                          '!
            25) B                                                    59) A                                                    93)      A.                  -
                                                                                                                                                                      {
            26) D                                                    60) D                                                    94)      B                            l
            27) A                                                    61) B                                                    95)      A                            ]
l            28) A                                                    62) B                                                    96)      D                            j
l            29) B                                                    63) D                                                    97)      B                            i
            30) C                                                    64) C                                                    98)      D                            i
          '31) B                                                      65) A                                                    99)      A
('        '32) A                                                      66) D                                                    100) C                                  1
            33) A                                                    67) D                                                                                            !
            34) C                                                    68) D
i
j            - - - - . . . . . . . . . . - - . . - - - - - - - . . . . . . . . . . . . . - - - . . . . . . . . . . . . . . - -
            ANSWER USAGE BY LETTER:                                  A) 25
                                                                      B) 26
;-                                                                    C) 24
                                                                      D) 25                                                                                            l
;
                                                                                                                                                                        I
.
l
1
  -
                                              >
\
'
      _      .          . . . .                            -    -      _ _ .- .            .                                                          - . . .
 
  _    ,                    _ _ _
                                                                                                      l
  ES-401                        Site-Sptcific Written Examination                  Form ES-401-7
                                            Cover Sheet
                                                                                                      I
                                                                                                      1
                                                                                                      i
                    U.S. Nuclear Regulatory Commission
                                        Site-Specific '
                              Written Examination                                                    l
                                                                                                      l
                                    ' Applicant information.
  Name:                                            Region:                          IV-              ,
                                                                                                      1
  Date:                    6/30/98-              Facility / Unit:      South Texas Project        !
  License Level:'                  SRO'            Reactor Type:                    W'              I
  Start Time:                                      Finish Time:
                                                                                                    )
                                            Instructions
  Use the answer sheets provided to document your answers. Staple this cover
      ~
                                                                                                    -
  sheet on top of the answer sheets. The passing grade requires a final grade of c
  f.t least 80.00 percent. Examination papers will be collected five hours after the
  examination starts.
                                      Applicant Certification
  All work done on this examination is my own. I have neither given nor received
  aid.
                                                        Applicant's Signature
                                              Results
  Examination Value                                                                      Points
  ' Applicant's Score                                                                      Points
l  Applicant's Grade                                                                      Percent
  NUREG-1021                                                      Interim Rev. 8, January 1997
L                                                                                                  ./
 
      .        .          m. _ _ _ __ _  m            _ _ .
                                                                  .
                                                                                      .        . . . . _ ~ _ _
                                                                                                              .
.
l
l
                                              STP Senior Reactor Operator Exam ... Page 1 of 100              j
l
l QUESTION 1:
i                                                                                                              l
'
  A fire'has been detected in the Post Accident Sampling System (PASS) Room. The                              i
  security card reader controlling access to the Fuel Handling Building is inoperable,
  preventing entry.
                                                                                                              :
  Which ONE of the following individuals may be issued a Vital Area Security Key                              l
  from the Controlled Key Locker to access this area in accordance with
! OPOP01-ZA-0019, Operation Key Control?
  A. Fire Brigade Leader                                        ,
  B. Fire Protection Coordinator
  C. Health Physics Supervisor
  D. Chemical Analysis Supervisor                                                                              :
:
                                                                                                              I
                                                                                                              !
                                                                                                              l
l
t
!
l
                      .
 
      .    - .        . - . . .    ~ . - . . . .  . - - . . .          . - - . . - - . . . - _ - . .            - . . - - . . . . - . - . .      -
  '
l'                                                              STP Senior Reactor Operator Exam ... Page 2 of 100
                QUESTION 2:
;
                Consider the following plant conditions:
                .  Plant was initially at 100% power
                .  RCS leak through a weld crack on a pressurizer level detector lower tap                                                          l
l                  occurred
!
                .  A plant shutdown has been initiated
                Based on this information, the NRC must be notified within                                followed by a                            !
              ' 30 day report (LER).                                                                                                                !
                A.15 minutes
                B. 30 minutes                                                                                                                        l
                C.1 hour
                                                                                                                                                      !
                D. 4 hours
        '
        -                                                                                                                                      -
                                                                                                                                                      l
                                                                                                                                                  ~
!
l
[
    '
!
I.
,
i
.
                                        .                        --                                  -
                                                                                                        ..            .-                    .
 
    .    .        ._            __        _ _ -_. _ . _ _        _ _ .  - - -        _ _ . . _  _ _.. . ._.
  -
                                                        STP Senior Reactor Operator Exam ... Page 3 of loo
        QUESTION 3                                                                                              .
                                                                                                                l
l
        Given the following:
1
'
        .    Both Units are at full power with one hour left until shift turnover.
        *    Current shift manning is as follows,:
i
                                    UNIT 1                                        UNIT 2
            SS                        1                                            1                          i
            US              2 (1 is Field Sup.)                2 (1 is Field Sup, and STA)
i          RO                        2                    2 (Extra RO left due to emergency)
            PO                        4                                            6
!
                                                                                                                ,
                                                                                                                i
        .    The Unit 1 Unit Supervisor, Field Supervisor and Reactor Operator are injured                      l
,
            during a midshift plant tour and are taken to first aid for treatment and                          i
!            observation
        Per Technical Specifications, which ONE of the following actions must be taken if
        this condition exists until shift turnover?
!                                                                                                              i
l      A. Unit 1 minimum shift crew composition is met and no further action is needed.
,                                                                                                              1
l      B. Unit 1 minimum shift crew composition is met provided the Unit 2 STA
l            assumes the Unit 1 Unit Supervisor position.                                                      ;
                                                                                                                <
i
!
l      C. Unit 1 minimum shift crew composition is NOT met and the unit must enter
l            T.S. 3.0.3 if the proper crew composition is not met within 2 hours.
1
1
'
      'D. Unit 1 minimum shift crew composition is NOT met and no further action is
            required other than the continual efforts to restore crew composition.
1 -
i
i
:
1
 
          , - .  .    . . - . . - - -.      . . . . . .-...- . .  . ..        . . . . .      . . . . . - . . . . . _ . . . . . - . .
!
l
                                              STP Senior Reactor Operator Exam ... Page 4 of.100
      QUESTION 4:
      Unit 1 is operating at 100% power when a valid High Alarm is received on
      RT-8013, RCB Purge Exhaust Monitor.
      Which ONE of the following radiation monitors is now. inoperable?.
      A. RCB High Range Ra'd iation Monitor (RT-8051)                                                                                    )
    ' B. RCB Atmosphere Radiation Monitor (RT-8011)
      C. Unit Vent Stack Monitor (RT-8010A)
      D. GWPS Outlet Radiation Monitor (RT-8032)
..
  6                                                                                                              --
                                                                                                                                          ;
                                                                                                                                          !
:
!
                                                                                                                                          !
                                                                                                                                          ;
i
      .        . _ . .                                            - ~ _ _ - .. - - . . - . , _ .
 
4
                                          STP Senior Reactor Operator Exam ... Page 5 of 100
  QUESTION 5:
  The unit is at 2% power during a plant startup following a refueling outage.
  Which ONE of the following conditions requires entry into a Technical Specification
  action statement?
.
  A. Core Physics testing has determined that the Axial Flux Difference is outside
    the target hand specified in the COLR.
  B. Primary containment internal pressure is at -0.05 psig during Supplemei.tary
'
      Purge operation.
, C. "A" Safety injection Accumulator nitrogen pressure is 660 psig.
i
  D. Intermediate Range Nuclear instrument Channel N35 has experienced a loss of
;    instrument power.
                                                                                              !
                                                                                              )
 
  _      . . .  _ . - _ _ _ . . _ . _ . _ _ . . .
J
l                                                        STP Senior Reactor Operator Exam ... Page 6 of 100
      QUESTION 6:
      On a loss of al; main feedwater at power, an AFW actuation may be initiated by
      the AMSAC system.
      - During a power decrease, which ONE of the following signals causes the C20
      - AMSAC BLOCKED permissive status light to illuminate?
                '
      A. < 30% Turbine Power
'
:
.      B. '< 30% Reactor Power
,
      C. < 4d% Turbine Power
      D. < 40% Reactor Power
i
.
    .                                                                                                        .
!
5
1
                                                    . . .
 
  -- . .          - = - . -  - .  - - _ . . - .. . . .  .- . . . . . . .          -. .- . .. .  . --_. -. -. -.-. -
l
i-
I                                                          STP Senior Reactor Operator Exam ... Page 7 of 100
l
l.        ' QUESTION 7:
!
          ~ Unit 1 is operating at 100% power with all systems in their normal lineup. A small
          RCS leak develops in the Chemical Volume and Control System (CVCS) with the
l'        fo!!owing Control Room indications:
l          *  Pressurizer level indicates 50% and'is decreasing
j          +  RCS pressure indicates 2220 psig and is stable
i          a  ' Containment pressure is normal
l          *  OUTLET PRESSURE PI-0135 indicates O psig
          *  PRT PRESS Hi alarm is illuminated
i.        *  LETDN HX OUTL FLOW Hl/LO alarm is illuminated.
          *1  ERFDADS Points (2) CVCS LTDN AREA TEMP TRN Hi are in alarm
i-        Based on these indications, which ONE of the following AUTOMATIC actions has
          taken place?
                            ~
l;        A. TCV-0143, Letdown Temperature Divert Valve, positioned to the VCT
r
'
          B. MOV-0023/0024, Letdown Line Containment Isolction Valves, closed
          C. FV-0011, Letdown Header Orifice Isolation Valve, closed
          D. PCV-0135, Letdown Pressure Control Valve, opened
,
i
l
t
f
                                  .
                                                          y                  m-,-.rr-                        m
 
                        _  __  .    _ _ _ . .  . - _ .  . _ . _ - . -    . . . _ . . _ _ _ _ _ __        _. _
                                                                                                                    . .,
                                              .                                                                          I
                                                        STP Senior Reactor Operator Exam ... Page 8 of 100
l                                                                                                                          I
l      QUESTION 8:
      Given the following:
i
l      *  Pressurizer variable heaters are energized                                                                      l
l      .  Pressurizer backup heaters are in AUTO control                                                                  l
      =  RCS pressure is at 2000 psig
      Assuming no operator action, if pressurizer level decreases to 14%, then is
!      restored to 25%, which ONE of the following describes the condition of the
      _ pressurizer heaters?                                                                                      ,
      A. Backup and variable heaters will be off.                                                                        4
                                                                                                                          i
      B. Backup and variable heaters will be on.
l      C. Backup heaters will be on and variable heaters will be off.
'
                                                                                                                          i
      D. Backup heaters will be off and variable heaters will be on,                                                      j
                                                                                                                          :
                                                                                                                          l
i
    .                                                                                                      *
l
l
                                                                                                                    .
I
!
l
,
4
                                                                                                                        .
  i                                                                                                                        .
 
    - . . . .                - -      -~ _-. - - - . _ . .            - -      -  - .          ...    .. ,  .- -
l                                                            STP Senior Reactor Operator Exam ... Page 9 of 100
              QUESTION 9:
              Unit 2 is preparing to fill the reactor cavity for refueling. A few minutes after LHSI
l            pump 2A is started, RHR PUMP 2A DISCH FLOW LO and RHR PUMP 2A TRIP
i            alarms are received.
l
              Which ONE of the following is the reason for the trip of the operating RHR pump?
              A. RHR relief valve has opened causing discharge flow to decrease to < 925 gpm
                for > 5 seconds.
              B. Start of LHSI pump 2A has caused RHR pump 2A mini flow recirc valve to
l                close and discharge flow has decreased to < 925 gpm for > 10 seconds.
              C. RHR inlet pressure has increased to > 350 psig causing the inlet isolation
                valves to close reducing discharge flow to < 925 gpm for > 10 seconds.
              D. Start of LHSI pump 2A has caused RHR pump 2A to become dead headed and
                discharge flow has decreased to < 925 gpm for > 5 seconds.
                                                                                                                    !
.
i
(
,
  '
 
    , . . .        -    ...  . . ~ . ~ . ~ .. - . .- - - _ .. . - . . . ~ _                  -. . - .    . ~ , . .- . . . --. - . ~ . . .
,
P
[                                                                STP Senior Reactor Operator Exam ... Page 10 of 100
!
(              QUESTION 10:
                When in Mid-Loop Operations, whicii ONE of the following set of conditions
                indicate that an RHR Pump is AIR BOUND?
,.
                A. Pump Current - HIGH
f
                  _ Pump Flow'- LOW
l
'
                B. Pump Current - HIGH
                    Pump Flow - HIGH
l              C. Pump Current - LOW
!-                  Pump Flow - LOW
                D. Pump Current - FLUCTUATING
                  Pump Flow - FLUCTUATING
!
t
l
,
                                                                                                                                            4
                                                                                                                                                  4.
i
                                                                                                                    o
i
                                                                                                                                              '
\
!
l-
.
i'
                                                                                                                                                      I
i
i
  ~
l
            .-
                                                                                    .        .                                                  ., ,
 
  ...            ..        -    . - - . - - - _ _ -              -  . - . . - . . -                  - - . -  .- __
                                                                    '
                                                                                                                        ,
                                                      STP Senior Reactor Operator Exam ... Page 11 of 100
f-
        QUESTION 11:-
        Unit 1 refueling is in progress. The refueling machine operator is inserting a fuel
j      assembly into the In-Containment storage area.
l      Which ONE of the following cautions should be observed while lowering the fuel
l      ' assembly into the In-Containment storage racks?
t
3'      A. Use a slow speed while entering the storage racks until the bottom of the fuel                                [
            assembly has entered.                                                                                      i
        B. Monitor the weight indicator continuously for an unexpected decrease of 50
            pounds which would require operators to stop lowering the fuel assembly.
        C. Use the Gripper Mast Down (red) light to determine the assembly is fully
          .
            lowered and can be disengaged.                                                                              ,
        D. Monitor the Z-axis tape continuously to ensure the mast is not lowered below
          the mast disengagement position.
                                                                                                                        .
                                                                                                                          <
                                                                                                                          i
                                                        , . _ , ,                    _ . . . . _ . . _        _
 
    _ _ _ _ . . .          .  . . . . - _ _    ._ _ ._. . .        ._    _._.__.......m.              ._ . - =._ _      _
.
                                                                    STP Senior Reactor Operator Exam ... Page 12 of 100
          QUESTION 12:
          Given the following:
          *        . Unit 1 is in Mode 6.
          .        50 fuel assemblies have been reloaded after a complete core off-load.
          *        RHR Train A is in service maintaining RCS temperature 90'F.
          4 RHR Train A loop boron concentration at last sample: 2950 ppm.
          . . Source Range NI channe! 31 indicates 10 cps and is selected for audible count -
'
                    rate indication.
          . ~ Source Range NI channel 32 indicates 5 cps.
          *        Both Extended Range NI detectors are operable
          in accordance with OPOP08-FH-0009, Core Refueling, which ONE of the following
          conditions requires suspension of Core Alte- 7tions and an investigation by the Core
          Loading Supervisor prior to resuming Core , r cerations?
          A. Source Range NI channel 31 increases to 40 cps.
          B. Updated log reading on RHR Train A inlet temperature indicates 98 F.
  +
          C. Source Range NI channel 32 suffers a loss of instrument power.                                              *
                                                                                                                                ,
          D. Updated RHR Train A boron concentrations on two consecutive samples
                    indicate 2920 ppm.
                                                                                                                            .
                                                                                                                          ~
1
$
_
$
  m                      -                  _ -              _ _ _                          _            -. -
 
                                        -    - . . . . . _ _ . . . . - - . . ..  . . . . . . . . . . .-
                                                                                                          l
                                          STP Senior Reactor Operator Exam ... Page 13 of 100            ;
QUESTION 13
                                                                                                          i
Given the following:
* ' Unit 1 has just experienced a Loss of All AC Power
*  OPOPO5-EO-ECOO is in progress
*  An RCS leak inside containment causes RCB pressure to rise to 11.0 psig.
*  Efforts are underway to restore power to at least one 4160 VAC ESF bus.                              '
.  The STA reports that an ORANGE Path condition exists on Containment
    Integrity Critical Safety Function (CSF) Status Tree.                                                4
Which ONE of the following describes the action (s) that should be taken?
A. OPOP05-EO-FRZ1 should be entered to address the loss of containment
    integrity safety function.
B. OPOP05-EO-FRZ1 should NOT be entered because it is assumed that at least
    one ESF bus will be energized during any Functional Restoration Procedure.
                                                                                                          "
C. OPOP05-EO-FRZ1 should be entered only when containment pressure rises to
    indicate a RED path on the Containment integrity CSF.
D. OPOP05-EO-FRZ1 should NOT be entered because the loss of power makes the                              !
    critical safety function status monitoring inaccurate.
                                                                                                          ;
                                                                                                          I
                                                                                                          l
 
  ..          .  . ~ _ . =                                  .        . . . . - . - . - - .
                                                                                            . . - _ . . - . - - . .      . - . .
i                                      .
                                                                                                                                  l
                                                STP Senior Reactor Operator Exam ... Page 14 of 100                              t
                                                                                                                                  '
                                                                                                                                  '
        QUESTION 14:
        OPOPO5-EO-EC00, Loss of All AC Power contains the immediate action step:
        " Verify Turbine Trip."                                                                                                  ,
                                                                                                                                  .
        Which ONE of'the following statements describes the BASIS for this step?
        A. Prevent an inadvertent Safety injection
        B. Prevent an inadvertent Main Steam Line Isolation
        C. Prevent an uncontrolled cooldown of the RCS
        D. Prevent steam generator inventory depletion                                                                            >
                                                                                                                                  .
                                                                                                                                  !
                                                                                                                                    ,
      .                                                                                                              .
                                                                                                                                  )
                                                                                                                                  ,
                                                                                                                                  f
                                                                                                                                  i
                                                                                                                                  ?
                                                                                                                        %
:
l-
l
!
                                                                                                                          -
..
                                                                                                                r  4
 
  . .    . -. .          .._    _ __          _.        _ . . _ . - __                        . ..
I
                                                                        STP Senior Reactor Operator Exam ... Page 15 of 100    q
                    QUESTION 15:-                                                                                                ;
1
                    Given the following:                                                                                        j
                                                                                                                                i
,
                -
                    .  .The reactor has tripped from 100% power operation                                                      ;
L.                  . Three (3) Control Rods did not fully insert                                                                l
!                  *    Reactor power is at 1E-6% and decreasing .                                                            i
!                  .    Boration is in progress at 25 gpm
i                  . - Reactor Trip Response (OPOP05-EO-ES01) is~in progress                                                    !
                                                                                                                                !
l
                  ' The Suberiticality ' Critical Safety Function (CSF):
i      ..
                - A. is satisfied and boration can be secured.
                                                                                                                                i
                    B. is satisfied, but boration must continue until shutdown margin is established.                            ;
l
                    C. is NOT satisfied because boration must be greater than 30 gpm per Tech
l.                      ' Specs.
                                                                                                                                i
j              ' D.' is NOT satisfied because more than one (1) control rod is not fuliy inserted.
                                                                                                                                l
l                                                                                                                                1
l
                                                                                                                                l
                                                                                                                                l
l
l                                                                                                                                .
                                                                                                                                !
..
I.
!.
I
                                                                                                                                  I
    4          m                    n-. 4                                                                                  - ,
 
                                          STP Senior Reactor Operator Exam ... Page 16 of 100
  QUESTION 16:
  OPOPO5-EO-FRS1, Response to Nuclear Power Generation - ATWS, Step 4.g.
  directs the operator to check pressurizer pressure less than 2335 psig.
  Which ONE of the following is the basis for this step?
  A. Verify RCS pressure is not high enough to reduce charging injection into the
      RCS.
  B. Verify RCS pressure will not challenge a pressurizer PORV.
  C. Verify RCS pressure has stabilized fellowing the initial heatup and insurge into              ;
      the pressurizer.
  D. Verify pressurizer PORVs are functioning to limit the RCS pressure increase.                  ;
.                                                                                              .
                                                                                                  !
                                                                                                  .
                                                                                                A
                                                                                                    l
                                                                                                  ,
                                                        . .                  .
 
  .    _    _-      __    _-  .  . . . ~ _ _ _ _  _.  ._.m._-._        _ _    _ - ____ _  -_m  . _ . _ . . _ .__ _ _ _
,
~
                                                                                                                              J
j                                                  STP Senior Reactor Operato. 8Exem ... Page 17 of 100
,
      QUESTION 17:
      A condition has occurred which warrants entry into OPOPO4-CV-0003, Emergency
:    Boration. While attempting to borate using the Normal Emergency Boration
      flowpath, both boric acid pumps fail to start.
:
      Which ONE of the following PAIRS of Emergency Boretion flowpaths are used in                                            l
                                                                                                                              l
      this condition?
      A. Through Alternate Boration isolation MOV-0218A OR
I        Emergency Boration via GravitV Feed
!    8. Through Alternate Boration Isolation MOV-0218A OR
          Through Manual Alternate immediate Borate 1(2)-CV-0221
s
      C. Through Manual Alternate immediate Borate 1(2)-CV-0221 OR
          Emergency Borate from the RWST                                                                                      ;
      D. Emergency Boration via Gravity Feed OR
          Emergency Borate from the RWST                                                                                      ;
                                                                                                                              :
                                                                                                                              i
                                                                                                                              l
                                                                                                                              l
                                                                                                                              l
                                                                                                                              l
                                                                                                                              )
    5
                                                                                                                            -
                                                                -__ - - -
 
.        -.            - - -      = . . . . . _ . . _ .        .. .        .  .        -. . - -        .
                                                                                                              l
                                                                                                              I
                                                  STP Senior Reactor Operator Exam ... Page 18 of 100
                                                                                                              1
    QUESTION 18:
                                                                                                              {
    Given the following:
    .  Unit 1 is operating at 90% power, steady-state, following a power increase                              ;
                                                                                                              '
    .
      Chemical Analysis reports that RCS samples indicate increasing RCS activity
    .  RT-8039, CVCS Letdown Failed Fuel Monitor is NOT in alarm                                              !
    .
      Decontamination Factor (DF) across the inservice mixed bed demineralizer is
      adequate
    .  CVCS letdown flow is 140 gpm
    Which of the following actions, directed by Chemical Analysis, are in accordance
    with OPOPO4-RC-0001, High Rer.ctor Coolant System Activity?
    A. Maintain present letdown flowrate and place the alternate mixed bed
      demineralizer in service.
    B. Maintain present letdown flowrate and place the cation demineralizers in
      service.
    C. Maximize letdown flow and place the alternate mixed bed demineralizer in
  -
      service.                                                                                        -
    D. Maximize letdown flow and place the cation demineralizers in service.
                                                                                                          .
                                                                                                            }
 
                                                              . .            ...    .      . .-- --
                                      9
                                          STP Senior Reactor Operator Exam ... Page 19 of 100
  QUESTION 19:
                                                                                                        I
  Given the following:                                                                                  j
  .  Unit 2 is in Mode 4
  .  Gaseous Waste discharge from the VCT is in progress
  a  GWPS Inlet Header valve (FV-4657) and Discharge flow valve (FV-4671)
    handswitches are in the NORMAL position at the control panel
  .  A high alarm condition exists on RT-8032, GWPS Outlet
  Which ONE of the following is the final position of the GWPS Inlet Header
  (FV-4657) and Discharge Flow (FV-4671) valves?
          Inlet Header Valve (FV-4657)      Diacharge Flow Valve (FV-4671)                              I
    A.                CLOSED                              CLOSED
    S.                CLOSED                                OPEN
    C.                OPEN                                CLOSED
    D.                OPEN                                  OPEN                                        I
                                                                                                        l
                                                                                                        l
                                                                                                        I
                                                                                                        ;
                                                                                                        i
                                                                                                      .
-                                                                                                    J
 
. . _ . . . _ _ _.              . ._ .    _ . . . . _ - . _ . _ . . . _ _ . _ . __                _ _ _ . _ . _ _ _ . . . . _ . . _ _ . _ . - . . _
                                                                                                                                        _
                                                                                                                                                                  - . . . . .
                                                                            _.
                                                                                                                                                                                l
                                                                                                                                                                                l
                                                                                                                                                                                '
l
l                                                                                    STP Senior Reactor Operator Exam ... Page 20 of 100                                      j
                    QUESTION 20:                                                                                                                                                I
                    - A Small Break LOCA has occurred, primary plant conditions have been stabilized                                                                            !
                    and Sl has been terminated in accordance with OPOP05-EO-ES11, Si Termination.                                                                              ;
                    Operators have just restored MINIMUM charging flow and continue performing                                                                                  !
                    ES11 when the following conditions are identified:                                                                                                          l
                                                                                                                                                                                l
                    *    RCS subcooling - 45*F and stable
j                    *    Pressurizer level - 5% and stable                                                                                                                    ~
l'                  *    Adverse Containment Conditions do NOT exist                                                                                                          ,
                                                                                                                                                                                !
                    Which ONE of the following actions will the operators perform?
                                                                                                                                                                                :
                                                                                                                                                                                l
                    A. Increase charging flow to restore pressurizer level and continue with
                          OPOP05-EO-ES11, Si Termination.                                                                                                                      i
1                      .
                                                                                                                                                                                \
l                    B. Operate Si pumps and continue with OPOP05-EO-ES11, Si Termination.                                                                                      1
                                                                                                                                                                                J
                                                                                                                                                                                i
i
l                    C. Operate Si pumps and transition to OPOP05-EO-E010, Loss of Reactor or                                                                                    l
l=                        Secondary Coolant.                                                                                                                                    I
                                                                                                                                                                                !
(.
l
                    D. Manually initiate Si and transition to OPOP05-EO-EOOO, Reactor Trip or Safety                                                            -
                                                                                                                                                                                L
l                        Injection.
                                                                                                                                                                                i
                                                                                                                                                                                )
                                                                                                                                                                                1
                                                                                                                                                                        -      :
l
i,
c
9
                                                                                                                                                                                !
i.
                                                                                                                                                                                .
                                                                                                                                                      - . , , .
 
    . .. .      _    _ . _ . .    _ ~ . -  .    .m  _-    _ - . . _ _ _ . . _ . . _ m . . _ . _ . . _ . .. ._. . . _ .- ..
                                                                                                                                ,
                                                                                                                                i
                                                                                                                                i
,
                                                    STP Senior Reactor Operator Exam ... Page 21 of 100
                                                                                                                                l
        QUESTION 21:
                    ~
                                                                                                                                i
,
        Given the following-
                                                                                                                                ,
                                                                                                                                '
        .  Unit 1 was operating at steady-state 100% power
        a  Reactor Trip and Safety injection actuated from low pressurizer pressure
        .
            RCS pressure decreased rapidly to 1150 psig and is now increasing very slowly                                      '
        .. Pressurizer level decreased to 8% initially and then increased rapidly to stabilize
            at 100%
        *  Reactor Containment Building process radiation monitors are alarming                                              ;
                                                                                                                                I
        Which ONE of the following Loss of Primary Coolant events is the cause for these                                        I
        indications?
      ' A. Pressurizer spray line rupture
        B. Pressurizer surge line rupture
        C. Reactor head vent line rupture
        D. Incore lower guide tube rupture                                                                                    l
                                                                                                                                l
                                                                                                                                l
                                                                                                                                1
                                                                                                                                I
                                                                                                                                i
  w        -                    ..              _
                                                            .
                                                                                                                              l
 
  _.  . _ . - . ~ _ _ . . . . _ ...._..... _ ... _ _ _ _ _.- _ _ _ _ _ _                              m. _ _ _ . . - _ . . . _    . . . _ . _ < .
.                                                                                                                                                    ?
,
;                                                                        . STP Senior Reactor Operator Exam ... Page 22 of 100
'
          QUESTION 22:
;i        A Large Break LOCA occurred 3 hours ago and the following plant conditions
          presently exist:                                                                                                                          ,
'
          *      RCS pressure is 350 psig
,
          *      Containment pressure is 8 psig                                  _
                                                                                                                                                      '
l          *      Containment Hydrogen concentration is 0.4%-
i          * SG NR levels are: 1 A - 32%; 1B - 40%; 1C - 28%; 1D - 12%
          =      Total Auxiliary Feedwater flow is 200 gpm
          *      Spent Fuel Pool temperature is'145 F
,
        ' Which 'ONE of the following actions is REQUIRED to be performed in accordance
          with OPOPO5-EO-E010, Loss of Reactor or Secondary Coolant?                                                                              -*
          A.~ Stop LHSI pumps and place them in auto
3
          B. Place Hydrogen Recombiner(s) in service                                                                                                ;
i-
.          C. Increase total AFW flow to greater than 576 gpm
i
,
      -
          D. Place Spent Fuel Pool Cooling in service                                                                              -
:
                                                                                                                                          &
                                                                                                                                                      l
                                                                                                                                                      1
                                                                                                                                                      I
                                                                                                                                                      l
                                                                                                                                                      i
                                                                                                                                                      1
                                                                                                                                  -                  >
 
      .      .        --    . . . ~ = . . - -    ~ .      . .    .- - -
,
.
i
                                                      STP Senior Reactor Operator Exam ... Page 23 of 100
          QUESTION 23:
)
          Given the following:
              Unit 2 is at 50% power
                      -
          *
,
;        e  Rod control is in manual
          a.~ T REF/AUCT T AVG DEV annunciator is in alarm
:
              Rod control Temperature error meter (TI-0612) indicates + 10 degrees
                                                                            -
          *
;
          a-  P-13 Channel 1 bistable light is extinguished
          *  P-13 Channel 2 bistable light is illuminated
      ,
          Which ONE of the following instrument failures has occurred?
          A. Pressure Transmitter 505 has failed low
l
          B. T-hot instrument 413 has failed high
.
  '
          C. Pressure Transmitter 506 has failed low
.l
i        . D. T-cold instrument 412 has failed high
;
-
.
?
    .
 
    .. . - - . . .        ~ .. ..-    . . - . - . . - . - . _ . . -            . . - . . _ _ - . . . _ -
i                                                                                                            ,
                                                                                                                                    j
                                                                                                                                    ,
                                                                                                                                    I
                                                                      STP Senior Reactor Operator Exam ... Page 24 of 1o0
            QUESTION 24:                                                                                                            !
            Given the following:                                                                                                    .
            .    Unit 1 is at 100% power                                                                                          l
            .    VCT LEVEL Hl/LO alarm is received                                                                                l
            *    VCT LEVEL LO-LO alarm is received                                                                                i
,
            .    VCT level (on CP-OG4) is indicating 85%
l            . VCT pressure (on CP-OO4) is indicating 20 psig
            .    LCV'112A, Divert valve to RHT is indicating intermediate
                                                          '
                                                                                                                                    l
l            Which ONE of the following instrument failures is the cause of the plant                                                .
I            conditions?
l
L            A. VCT Level Transmitter 112 has failed high
l
            B. VCT Level Transmitter 113 has failed high
l            C. VCT Level Transmitter 112 has failed low
l
l-          D. VCT Level Transmitter 113 has failed low
                                                                                                                          4
.
                                                                                                                                    '
                                                                                                                            .
!
.
  l
I
;
:
i
!
!
l
l
                    ,  .
                                                                            - , ,                          -                . . . '
 
    .  - - - - -    -          . - .
                                              STP Senior Reactor operator Exam ... Page 25 of 10o
      QUESTION 25:
      In OPOPO5-EO-FRC1, Response to inadequate Core Cooling which ONE of the
      following methods is attempted FIRST, based on it being the most effective
      approach to recover the core and restore core cooling?
      A. Restart of a Reactor Coolant Pump
      B. Reinitiation of High Head Safety injection
      C. Rapid depressurization of intact Steam Generators
      D. Opening Pressurizer PORVs and Reactor Head Vent Valves
l
i
.
l .
{
                                                                                                  /
 
  .    .;.-            . ~,,    . . - . . . . - . .  . . . . . - - . .  _ . . . . -  . - . - - . - . . - - . -  . . . . .    . .
                                                    -
                                                                                                                                    i
                                                                                                                                    !
                                                              STP Senior Reactor Operator Exam ... Page 26 of 100                  )
                                                                                                                                    !
    QUESTION 26:
    Step 1 of OPOP05-EO-FRP1, Response to imminent Pressurized Thermal Shock
    Condition, has the operator check that LHSI flow is greater than 500 gpm if RCS                                                i
    pressure is less than 415 psig.                                                                                                1
    This step is based on:
    A. ensuring adequate mixing in the cold leg downcomer region during natural
          circulation conditions,                                                                                                  j
                                                                                                                                    ;
                                                                                                                                    i
    B. preventing core exit temperatures from exceeding the required temperature to
          place RHR in service.
    C. ensuring adequate Low Head Safety injection cooling prior to Accumulator
          isolation.
    D. preventing implementation of Pressurized Thermal Shock (PTS) actions if a
          Large Break LOCA has occurred.
                                                                                                                                    l
                                                                                                                            .
                                                                                                                                    i
                                                                                                                                    l
                                                                                                                              .
.
 
  '
                                                STP Senior Reactor Operator Exam ... Page 17 of 100
      QUESTION 27:
'
      A reactor trip with a loss of ALL feedwater has occurred. The crew has entered
      OPOP05-EO-FRH1, Response to Loss of Secondary Heat Sink and RCS Bleed and
l
      Feed Criteria have been met. Attempts to establish RCS Bleed and Feed produce
      the following conditions:
      .  One (1) HHSI Ptmp - RUNNING
      =
          One (1) Pressurizer PpRV - OPEN
      =  Two (2) Reactor Vessel Head Vent paths - OPEN
      Based on these conditions, determine the adequacy of the RCS Bleed and Feed.
      A. Only the RCS Bleed path is NOT adequate
i
      B. Only the RCS Feed path is NOT adequate
      C. Both RCS Bleed and Feed paths are NOT adequate
l      D. Both RCS Bleed and Feed paths are adequate
,
!
!
l
l
!'
h
4
    .
 
    . . . . . . . . . . _ _ _ . . _ - . . . . . . _ . . . . _ . - _ , _ . . - _
                                _.                                              - . _ . _ . _ _ . . .      . . _  _ . _ _ _ . . - . . . . _ . _ . . - _ . . _ _ . . . _ _ . - _ .
                                                                                                                  !
,
                                                                                                      STP Senior Reactor Operator Exam ... Page 28 of 100
                                  QUESTION 28:
.
                                  Unit 1 has experienced a reactor trip due to an RCS leak inside containment.
                                  .  .0POPO5-EO-EO10, Loss of Reactor or Secondary Coolant, is in progress.                                                                          I
                                  e  . Containment pressure is 12 psig decreasing at 0.25 psig per minute.
                                  *    RCS temperature is 570*F
                                  e    SG pressures are: 1 A - 1230 psig; 18 - 1225 psig; 1C - 1350 psig; 1D - 1150
                                        psig
                                  a    SG NR levels are: 1 A - 45%; 1B - 28%; 1C - 75%; 1D - 34%.
                                  .    OPOP05-EO-FRH2, Response to Steam Generator Overpressure, has been
                                        entered.
                                  *    An operator has been dispatched to the 1C IVC to investigate.
                                  For the existing plant conditions, the Unit Supervisor should:
                                  A. NOT direct the operator to locally open the 1C SG PORV due to the potential of
                                        water hammer in the steam piping.
                                  B. NOT direct the operator to locally open the 1C SG PORV due to the excessive                                                                      i
                                        differential pressure across the valve.                                                                                                    -
                                                                                                                                                                                        .
                                  C. Direct the operator to open the 1C SG PORV to decrease pressure below the
                                        SG PORV setpoint.                                                                                                                              ;
                                  D. Direct the operator to assist maintenance in manually lifting a 1C steam safety                                                                '
                                        valve to decrease pressure.
                                                                                                                                                                                        !
                                                                                                                                                                                        l
                                                                                                                                                                                        1
                                                                                                                                                                                        i
                                                                                                                                                                                        ;
                            4
  .
 
    .- .-.              . - . .    . . . - --        - _ . -  -        _ . - _ . - . . ~ . . . . - . - - . . ~ _ . . . _ _ _ _
l-
                                                                                                                                  I
                                                    STP Senior Reactor Operator Exam ... Page 29 of 100                            ,
                                                                                                                                  ;
i
                                                                                                                                  !
'
            O.UESTION 29:
            In OPOPO5-EO-ES12, Post LOCA Cooldown and Depressurization, an Si flow
            reduction sequence is performed to minimize leakage from the RCS while still                                          i
            maintaining adequate coolant inventory. An RCS Subcooling Table, which takes
            into account current pump status, is used to determine when the HHSI pumps are                                        '
            stopped.
:                                                                                                                                  i
            Which of the following pump's status are accounted for in this RCS Subcooling
            Table?
            A. RCPs, HHSI Pumps and LHSI Pumps
            B. CCPs, HHS! Pumps and RCPs
            C. CCPs, HHSI Pumps and LHSI Pumps
            D. CCPs, LHSI Pumps and RCPs                                                                                            j
                                                                                                                                    .
                                                                                                                                    l
l
l
i
.
I
4
i
  '
(
,
 
  . _
          -    .  . _ - .-                .  -        .--    .    . .        ...          .        ..  .
                                                                                                                i
                                                    STP Senior Reactor Operator Exam ... Page 30 of 100          ;
                                                                                                                '
            QUESTION 30:
            The following plant conditions exist:
      .
            *  Small Break LOCA has occurred                                /
                                                                        *j'
            .  Adverse Containment conditions exist                                                              I
            a  OPOP05-EO-EO10, Loss of Reactor or Secondary Coolant, is in progress
            *  Operators are checking whether RCS cooldown and depressurization is required                      )
                                                                                                                  1
            Which ONE of the following criteria warrants transition to OPOP05-EO-ES12, Post
            LOCA Cooldown and Depressurization?
            A. Pressurizer Levelless than 8%
            B. RCS Subcooling less than 35*F
            C. RCS Pressure greater than 415 psig
            D. Two Hot Leg Temperatures greater than 420 F
        .
                                                                                                          -
l                                                                                                                l
                                                                                                              *
!
.
5
l
l
l
 
,
    .,  __ _.              . . _ _ . . _
                                _
                                            . _ ~ . .  . . _ . _ - . - _ . - __ _ _ . . _ . - _ _ _ . _ _ . . . . _ _ _ _ _ . . _ _ . .            _ . . . _ . , _ _ .
                                                                                                                                                    _
              $
                                                                        ' STP Senior Reactor Operator Exam ... Page 31 of loo
!        QUESTION 31:
        Given the following:                                                                                                                                            j
<                                                                                                                                                                        !
        *      Unit 1 is at 25% power and stable
        e      SG 1 A level is indicating 83% NR
        e      SG 1B level is indicating 78% NR
        ..      SG 1C level is indicating 89% NR
        *      SG 1D level is indicating 80% NR
        e      All systems are in their normal alignment for the current power level
        Which ONE of the following lists the automatic actions that should directly result                                                                              i
        from the above conditions?                                                                                                                                    '!
  .
        A. Reactor trip, turbine trip, feedwater isolation bypass valves close and low
                power feed regulating valves close
        B. Turbine trip, operating feed pumps trip, feedwater isolation valves'close and                                                                                i
                feedwater regulating valves close                                                                                                                        l
                                                                                                                                                                          1
        C. Reactor trip, operating feed pumps trip, feedwater isolation valves close and                                                                                j
'
                low power feed regulating valves close
l        D.' Turbine trip, operating feed pumps trip, feedwater isolation bypass valves close
                and feedwater regulating valves close
l
!
                                          .
                                            - , - . -            y                ~ --.                                              . . - . . ,
 
. _ .          _ _ _ - _ . . . . _ _ _ . _ . . _ . . _ - . _ _ . _ . _ . . ~        -.- - _ _ _ _ . _ _ _ . _ . _ _ _ . . _ .    ...
                                                                                                                                      \
                                                                                                                                      l
                                                                                                                                      .
                                                              STP Senior Reactor Operator Exam ... Page 32 of 100
        QUESTION 32:                                                                                                                  ;
                                                                                                                                      l
                                                                                                                                      '
        During a unit shutdown, all 4 pressurizer pressure channels slowly decrease and
        stabilize at approximately 1960 psig activating P-11. The operator manually blocks
        the low pressure Si signal. Over the next 15 minutes, the pressure channels have
        the following trends:                                                                                                        l
                                                                                                                                      1
        *  PT-455 - Slowly increases to 1984 psig, then lowers to 1860 psig                                                          '
        *  PT-456 - Slowly increases to 1987 psig, then lowers to 1856 psig                                                          ,
                                                                                                                                      '
        *  PT-457 - Slowly increases to 1986 psig, then lowers to 1858 psig
        *  PT-458 - Slowly increases to 1983 psig, then lowers to 1855 psig
        Assuming all bistables trip EXACTLY on their setpoint, which ONE of tha following                                            ;
        describes what will happen during the pressure transient?
        A. P-11 will reset and a safety injection will occur
        B.  P-11 will NOT reset and a safety injection will NOT occur
                                                                                                                                      l
                                                                                                                                      !
        C. P-11 will reset but a safety injection will NOT occur
                                                                                                                                      !
      - D. P-11 will NOT reset but a safety injection will occur                                                              -
                                                                                                                                      ;
                                                                                                                                      I
                                                                                                                                      !
                                                                                                                                -    !
                                                                                                                                      l
                                                                                                                                      1
                                                                                                                                      l
                                                                                                                                      !
                                                                                                                                  .
 
              ..
                                        STP Senior Reactor Operator Exam ... Page 33 of 100
                                                                                            ,
QUESTION 33:
                                                                                            I
Which ONE of the following alarm conditions is indicative of a simultaneous Digital        I'
Rod Position Indication (DRPI) Data A AND Data B failure?
LEGEND: General Warning (GW), Rod Bottom (RB), Urgent (Urg), Trouble (Troub),
Rod Supervisory Monitor Rod Position Trouble Alarm (Rod Supv Mntr Rod Pos
Troub), Flashing (FLASH) and illuminated (LIT)
                                                                                            >
      DRPI      DRPI      DRPI      RPI (05M3)      DRPI          Rod Supv
      GW        RB        Urgent    Trouble        Non-Urg        Mntr Rod Pos
      Light      Light    Alarm    Alarm          Alarm        Troub Alarm
A.    FLASH      LIT      LIT      LIT            OFF            LIT
  B.  FLASH      FLASH    LIT      OFF            LIT            LIT
  C.  LIT        FLASH    LIT      LIT            LIT            OFF
  D.  LIT        LIT      OFF      LIT            LIT            LIT
                                                                                            l
                                                                                            )
                                                                                            l
                                                                                            1
                                                                                            !
                                                                                            !
 
            .                                            . - . -  .    .      _              ..  .
                                                                                                      ,
l
                                            STP Senior Reactor operator Exam ... Page 34 of 100
  QUESTION 34:
  Given the following conditions:
  .  Unit 1 is at 75% power
  .  Rod Control is in MANUAL
  *  Turbine Feedback Control is in IMP IN
  *  PR NI Channel 41 indicates 5% less than other PR NI Channels
! .  "D" Control Bank Rod D-12 DRPI Rod Bottom Light Lit
  .  The following Control Room Annunciators are illuminated:
    *  SM03/F-4, ROD BOTTOM
      *
        SM03/D-5, ROD SUPV MNTR ROD POSITION TRBL                                                      i
    *  SM02/A-6, TREF /AUCT T AVG DEV
    *  SM03/C-3, PR CHANNEL DEV
                                                                                                        l
  Assuming no operator actions, which ONE of the following correctly shows how
  the final steady state plant conditions will compare to the steady state plant
  conditions that existed prior to the event?
,
            Average Core Power          Generator MW                      RCS Tavg
                                                                                            -
                                                                                                  ^
    A.            Lower                      Same                        Lower
    B.            Higher                    Lower                        Same
    C.            Same                        Same                        Lower
                                                                                                    '
    D.            Same                      Lower                        Same
l
:
I
              .
 
  ._ .      . _ . . . . = . _ , . . . _ _ .__ _  _ _ . . _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ . _    _-  _ . _ . _ _ _ _
                                                STP Senior Reactor Operator Exam ... Page 35 of loo
QUESTION 35:
The following plant conditions exist:                                                                              t
*      Unit 1 is operating at 90% power
.      I&C has determined that TWO control rods are untrippable
Which ONE of the following operator actions is required per OPOPO4-RS-0001,
Control Rod Malfunction?
A. Maintain affected banks within 12 steps of the untrippable rods
B. Borate the RCS until Cold Shutdown boron concentration is achieved
C. Shutdown the reactor to Mode 3 within six hours
D. Trip the reactor and initiate emergency boration
                                                                                                                    i
                                                                                                                    l
                                                                                                                    1
                                                                                                                    l
                              .
 
  . _ . . . .          .      . . -  - . . _ . . - . .-- . ,..-.- - .-.~ . -                    - . .  _ .    _ . . . . . . .
,
                                                              STP Senior Reactor operator Exam ... Page 36 of 100
              - QUESTION 36:
                Given the following:
                .  Unit 1 is in Mode 1 at 75% power
                *  Rod control is in AUTO
                e  Control Bank D at 235 steps
                *  Main turbine control in IMP OUT
                *  Control rods continuously withdraw for no apparent reason
                =  Operators place rod controlin MANUAL and rod motion stops
                *  Operators determine that Control Bank "D" rods have withdrawn a total of 10
                  steps
l              -Which ONE of the following conditions would exist?
l
l
                A. Axial Flux Difference (AFD) would become more positive /less negative.
                B. Overpower Delta Temperature trip (OPDT) setpoint would increase.
;-
                C. Nuclear Enthalpy Rise Hot Channel Factor (HCF) would decrease.
1
L              D. Quadrant Power Tilt Ratio (OPTR) would increase.
r
!
,
!
i
l
l
l
,
i
 
  . - . . _        . _ _ . . . _ _ . _  _ . _ _ . . . . _ _ _ _ _ _ _ _ _ _ _ _                        _ , _ _ . . - . _ _ _ _ . _ _ - . . _ _ . . _ _ _ . .
                                                                                  STP Senior Reactor Operator Exam ... Page 37 of 100
              QUESTION 37:                                                                                  g
              Unit 1 unidentified leakage has increased to 0.5 gpm and containment entry is
;            . being planned to identify the source of the increased leakage.                                              .                                      ,
l
l            'In accordance with OPSP03-XC-0002A, Partial Containment inspection, which ONE
l,            of the following is a requirement associated with the containment entry and close-
'
              out inspection?
              A. The Shift Supervisor shall resolve any question that arises concerning what
,
                  constitutes loose debris and/or material.
l
              B. A Reactor Operator shall be the Test Coordinator for the containment entry and
l                close-out inspection.
!-
              C. A Health Physics Technician shall perform Form 1, Containment Pre-Entry
l                Briefing.
I              D. The Health Physics Supervisor shall be the only point of contact for determining
                  what tools may be left in the RCR,
                                                                                                                                                              .
(
l
                                                                                                                                                                ~
l
l
l
l
l
!
i
.'
                                        .
                                                                                                                              Y
 
  - .  . . - . .      .    .  - - . -
                                          - . . - . . . ~ . - - - - . .            . - ,_. ~ _ -              - - . -.. - - - .. -.
                                                -
.                                                                                                                                    ,
                                                                                                                                      I
                                                                                                                                      r
;
                                                                    STP Senior Reactor Operator Exam ... Page 38 of 100
                                                                                                                                      f
.
      QUESTION 38:
;
      Given the following:
                                                                                                                                      '
,
t      *      During a midcycle reactor startup the Reactor Containment Building (RCB)                                                :
L            CNTMT PRESS Hl/LO alarm annunciates with RCB pressure at +0.3 psig.
l
      =    T.S. 3.6.1.4 for Containment Systems Internal Pressure is entered.                                                      l
!      + It is deterniined that a Supplementary Containment Purge will be performed to
              lower the RCG pressure.                                                                                                <
l      =      Four RCFC Units are in operation
;
      *    The US notes that the Reactor' Containment Purge Release Permit has expired.                                            4
                                                                                                                                      )
      Which ONE of the following describes the action (s) to be taken to lower RCB
      pressure?
      A. The purge may be started provided Chemical Analysis concurs and a new purge
              request is initiated.
      B. The purge may be started provided Chemical Analysis concurs and a continuous                                                !
              sample is performed during the purge.
      C. The purge may NOT be started until a valid purge permit is completed and
              delivered to the control room.
      D. The purge may NOT be started until a valid purge permit is completed and
              additional RCFCs are started,
                                                                                                                                        i
                                                                                                                                        !
                                                                                                                                      j
 
  - . . . .  -    .  .    . . . ~ . . . . . . . . . - . . . . - - - . - . . . - . . . _ - . . . . . . - ~ - . - . . . . ~ . -  -  .
                                                                                                                                        i
                                                                                                                                      I
                                                                                                                                      '
                                                                          STP Senior Reactor Operator Exam ... Page 39 of 100
              QUESTION 39:
.
              The following plant conditions exist:
                                                                                                                                      4
                                                                                                                                      '
'
'
                *  Small break LOCA occurred 12 hours ago                                                    .
                *  Containment pressure is 3.7 psig                                                                                    ,
                *  Containment temperature is 215"F
:            . * OPOP05-EO-FRZ3, Response to High Containment Radiation Level, is entered
              Which ONE of the following actions is required in accordance with FRZ37
.
:              A. Place the Containment Carbon Filter Units in service.                                                                !
!
;
                B. Ensure all Reactor Containment Fan Coolers are running.
)              C. Ensure Containment Ventilation isolation has occurred.                                                              )
'
                                                                                                                                        i
;              D. Sample containment atmosphere activity using the PASS.                                                              1
1
                                                                                                                                        1
:                                                                                                                                        I
                                                                                                                                .
                                                                                                                                    A
                                                                                                                                        !
                                                                                                                                        l
                                                                                                                                        l
                                                                                                                                        l
                                                                                                                                      ,
 
.                  . -  ._.                .    .      __._    .            . . _ ..    - .
                                        STP Senior Reactor Operator Exam ... Page 40 of 100
  QUESTION 40:                                                                                  !
                                                                                                l
  Given the following:
  *  Unit 1 is at 100% power                                                                  !
  .  ECW and CCW pumps 1 A are running
  a  ECW and CCW pump 1B control switch is in AUTO and their ECW/CCW                          '
      selector switch is in standby
  .  ECW and CCW pump 1C are not running and their ECW/CCW selector switch is
      in OFF
  A ground on 4.16 KV bus E1 A causes a bus It ": ut. The standby ECW/CCW train
  will start when the:
  A.1 A ECW and CCW pump breakers open                                                          i
  B. CCW common header discharge pressure decreases to 80 psig
  C.1 A ECW pump discharge valve begins closing
  D. ECW trains 1 A and 1C discharge pressure decreases to 20 psig
 
    . - . - . . ~ - .        . , . .    . . - .  . ~ . .  . . - . -  .    . ~ . . . _ . . . . . . . . . - - . . . _ _ , . . . -
'
                                                                                                                                    1
                                                                                                                                    i
l
4
                                                          STP Senior Reactor Operator Exam ... Page 41 of 100                        I
:
;        QUESTION 41:
                                                                                                                                    1
                                                                                                                                    1
:.        Given the following:                                                                                                      !
4
          a    ' Unit 1 is in Mode 5                                                                                                l
1        e      RHR train 1 A is in service maintaining RCS temperature 130 F                                                        !
          *-    CCP 1 A is in service -                                                                                              I
!        *      CCW trains 1 A and 1C are in service                                                                                l
          *      Low pressure letdown supplied from RHR train 1 A
          Tiie control room receives the CCW SURGE TK LVL LO and CCW SRG TK
b        COMPART A LVL LO alarms. Surge tank levelin the train A compartment is
:        currently 48% and decreasing.
1
:        'Which ONE of the following describes the Control Room response to existing plant                                          l
'                                                                                                                                    I
          conditions?
-
                                                                                                                                      l
l        A. Trip CCW pump 1 A, place RHR train 1C in service, and secure RHR train 1 A.                                            )
                                                                                                                                    '
i
;        B. Ensure the CCW Surge Tank Reactor Makeup Water Storage Tank makeup
                valve is open, monitor CCW surge tank level and trip CCW pump 1 A if level
  .
                decreases to < 38E                                                                                          .      !
        - C. Trip CCW pump 1 A, place CCP 1B in service and secure CCP 1 A.
          D. Ensure the train 1 A CCW to RHR heat exchanger inlet valve is open, place RHR
                train 1C in service, and align low pressure letdown to RHR train 1C.                                              -
                                                                                                                                    :
                                      ,
                                                                                                                                    i
 
          . ~ - -      . - . . _ .    =  . . . . = . . . . . _ . .    -              . ~ . . . . . . -
                                            STP Senior Reactor Operator Exam ... Page 42 of 100
    QUESTION 42:
    A plant shutdown to Hot Standby is in progress with reactor power well below the
    POAH. As control bank rods are being inserted, the following conditions occur:
    .  Permissive P-6 clears at the expected setpoint
    .  Both Source Range Nls fail to automatically re-energize
    .  Reactor power is decreased to 5.0 E-11 amps on the Intermediate Range and
      the Primary RO is directed to manually energize the Source Range Nis.
    Which ONE of the following Source Range NI readings is expected at this reactor
    power level?
    A. > 5.0 E4 and s 5.0 E5
't
    B. > 5.0 E3 and s 5.0 E4
    C. > 5.0 E2 and s 5.0 E3
    D. > 5.0 E1 and s 5.0 E2
:
  ,
                                                                                                          ,
                                                      w
 
                      -.  . .    - . .  . .      .    - _ = . - . _ _ . _ _ ._ . . _    - __  .
                                                                                                      l
                                                                                                        l
                                              STP Senior Reactor Operator Exam ... Page 43 of 100
  QUESTION 43:                                                                                        1
  Unit 1 is performing a reactor startup after a refueling outage of 25 days. The
  ICRR has continuously predicted criticality within the + or - 500 pcm band.
  During the last 50 step pull the reactor was determined to be critipal at 60 steps
  on control bank C (Below LO-LO Insertion Limit).
  One of the required actions is to                    to ensure that                            .
  A. commence emergency boration; acceptable temperature limits are maintained.
  B. trip the reactor; reactivity transients associated with accidents are within
    acceptable limits.
  C. commence emergency boration; shutdown margin limits are maintained.
  D. trip the reactor; heat flux hot channel factor remains within limits.                              ,
                                                                                                        l
                                                                                                      ;
.
                                                                                                    A
                                                                                                      ,
                          4
 
    ..-. .      _ . ~ . . ~  , . -    . . . . . - . .  ~ . _ . . - _ - - -        - . - . _ - - _        , - - . . - .. . -
                                                      -
                                                                                                                                .
                                                                                                                                !
                                                                                                                                l
                                                                                                                                .
                                                            STP Senior Reactor Operator Exam ... Page 44 of 100                :
t.
            QUESTION 44:                                                                                                        !
            A major steam line break inside containment has caused a reactor trip, safety
            injection and containment spray actuation.. The Control Room operators have just                                  ,
            completed the immediate actions of OPOP05-EO-EOOO, Reactor Trip or Safety                                          I
            Injection, when the secondary operator notes that 1D steam generator level is                                      !
            increasing in an uncontrolled manner.                                                                              l
l
            In regards to the AFW pumps, which ONE of the following statements describes                                        :
            the correct operator response to this event?                                                                        j
                                                                                                                                ;
!          A. Verify all Motor Driven AFW pumps are running then secure #14 AFW pump.                                          !
                                                                                                                                i
,.
            B. Continue to feed steam generator 1D to a level of > 14% then secure #14                                          l
                                                                                                                                '
l              AFW pump.
l
l          C. Verify at least two Motor Driven AFW pumps are running then secure #14 AFW                                    'l
l              pump.
l
l          D. Continue to feed steam generator 1D to a level of > 34% then secure #14
                AFW pump.
i
:
I
                                                                                                                                ;
j.                                                                                                                              ,
,
l
l
l
l
I
i.
l
l
.
(
  '
l
 
  ..~ ,    . , _ .  . . .        .. .          .. .    . - . . .          .-  - . ~        . . . . - -
!
l
                                                                                                              i
l-
i
                                                    STP Senior Reactor Operator Exam ... Page 45 of 100
I                                                                                                            !
;-        QUESTION 45:
t.                                                                                                            ,
,        Assuming all plant systems function as designed, which ONE of the following                        !
'
          Critical Safety Functions would have the HIGHEST priority condition during a -
          postulated " Uncontrolled Depressurization of ALL Steam Generators?"                                ;
        - A. Core Cooling (C)
          B. Heat Sink (H)
        .C. Integrity (P)                                                                                    ;
                                                                                                              .
          D. Inventory (1)                                                                                    .
                                                                                                              '
                                                                                                              ,
,                                                                                                            ,
                                                                                                            . .
                                                                                                              !
                                                                                                              5
                                                                                                              .I
                                                                                                              1
                                                                                                                l
                                                                                                                l
                                                                                                                l
                                                                                                                1
l-                                                                                                              !
l                                                                                                              I
f
l
                                                                                                                1
 
  ~ ~ . .-.      ..          .- . - - - _. - . - . .                    . - . . ~ _        . - . . . . - . . .-
4
4
i
!
4
                                                      STP Senior Reactor Operator Exam ... Page 46 of 100
                                                                                                                  ,
        QUESTION 46:
        Given the following:
                                                                                                                  '
*
        *  A Reactor Startup is being performed following a midcycle outage.
        *  Reactor Power has been stabilized at 1E-8 amps                                                        !
.
        *  RCS temperature is at the no-load value
l        .  Critical data has been taken.
;
        *  Prior to any additional control rod movement, a single SG Safety Valve on SG                          ;
i
            1D fails opens and remains open.
j        *  RCS Tavg decreases 9"F and reactor power starts to increase.
:
        Which ONE of the following states the correct action, required to satisfy Technical                      ;
!
        Specification LCO(s)?
j        A. Reduce power range high flux high trip setpoint to 87% rated thermal power.                            .
                                                                                                                  '
!
,
        B. Restore SG safety valve to operable status prior to entering Mode 1.                                  ;
                                                                                                                  '
1
        C. Restore RCS Tavg within 15 minutes or be in Mode 3 within the next 15
,
            minutes.
,
l        D. Immediately initiate boration to restore Shutdown Margin.
:
!
i
                  .
                                                                                                                  l
,
N
n                                                                                                                  !
4
4
                                                                                                                    i
 
  . ._. .m    . _ _ .      -  _ . _ . - . ._ -  _ ___.    ...._ ._. ._      _ _ _ _ _ _ _        _ . . _ _ _ . . _ _
                                                STP Senior Reactor Operator Exam ... Page 47 of 100
          QUESTION 47:
          Unit 1 is operating at 85% power when the Plant Computer AFD monitor alarm
          actuates. Which ONE of the following actions will comply with Technical
          Specification requirements? -                                        .
          A. Within 30 minutes restore AFD to within limits.
          B. Within 30 minutes reduce power to s 50% and reduce Power Range neutron
            setpoint to s 55% within 4 hours.
          C. Within 15 minutes restore AFD to within limits.
          D. Within 15 minutes reduce power at least 3% from rated thermal power for ever
            1% that AFD exceeds the limit and similarly reduce Power Range neutron
            setpoint within 4 hours.
.
                                                                                                                        s
                                                                                                                          !
                                                                                                                          :
                                                                                                                          l
                                                                                                                          l
                                                                                                                          !
                                ..                              _                          __
 
                                        STP Senior Reactor Operator Exam ... Page 48 of 100
QUESTION 48:                                                                                ,
While performing a control rod operability surveillance at 100% power, one control
bank "D" rod is found to be immovable due to mechanical interference.
Which ONE of the following actions shall be taken within one hour?
A. Determine the position of the immovable rod using the movable incore
  detectors.
B. Determine that the shutdown margin requirement is satisfied.
C. Reduce power to less than 90% of rated thermal power.
D. Reposition all Control Bank "D" rods to within 12 steps of the misaligned rod.
 
                                          STP Senior Reactor Operator Exam ... Page 49 of 10o
  QUESTION 49:
  Which ONE of the following is an interlock associated with supplying power to a
  4.16 KV ESF bus?
  A. Diesel Generator output breaker cannot be closed manually un!'ess the
    emergency bus 1L (2L) feedor breaker supplying the respective ESF transformer
    is open.
  B. Diesel Generator output breaker will not close automatically unless diesel
    generator speed is 2 500 rpm and diesel generator voltage is 2 3744 volts.
  C. Standby bus 13.8 KV ESF transformer feeder breaker is interlocked with its
    disconnect switch such that the disconnect must be open before the feeder
    breaker can be opened.
  D. Standby bus 13.8 KV ESF transformer feeder breaker is interlocked with the
      13.8 KV Emergency bus ESF transformer feeder breaker to prevent both of
    them being closed at the same time.
.                                                                                            .
                                                                                                4
 
                                  .
                                        STP Senior Reactor Operator Exam ... Page 50 of 100
QUESTION 50:
On a loss of 125 VDC Class 1E power to the Emergency Diesel Generator, which
ONE of the following components loses power?
A. Control Power for Test Mode
B. Local Control Panel Indicating Lights
C. Voltage Regulator Field Flash Circuit
D. Standby Fuel Oil Booster Pump
                                                                                              l
                                                                                              1
                                                                                              !
                                                                                              l
                                                                                              ,
                                                                                            -l
                                                                                              l
                                                                                              l
                                                                                              ,
 
  .  . . - . . .          . . - . . . . . .. ._    . . . - , . . -      . . ~ . . . _ - . .    . . ~ - . . - -        - _    .
!
l
                                                                                                                                    t
                                                                      STP Senior Reactor Operator Exam . . Page 51 of 1o0
                                                                                                                                    ,
                    QUESTION 51:
                    Unit 1 is operating at full power, steady-state conditions. A periodic load test is
i                    being performed on ESF Diesel Generator #13 in accordance with
                    OPSP03-DG-0003. ESF Diesel Generator #13 has been paralleled..with 4.16 KV
                    Bus E1C and is carrying 2,000 KW of real load. A Feedwater line break inside
                    containment occurs and containment pressure increases to 15 psig.
                    Which ONE of the following describes the response of the "C" ESF load sequencer
                    and ESF Diesel Generator #137                                                                                  ,
                    A. The Diesel Generator remains paralleled to the 4.16 KV bus and the Mode l
l                      timer sequences equipment onto the 4.16 KV bus.
                                                                                                                                    ,
                    B. The Diesel Generator output breaker is tripped open and the Mode i timer
;                      sequences equipment onto the 4.16 KV bus.
                    C. The normal feeder breaker AND the Diesel Generator output breaker to 4.16 KV
L                      bus E1C are tripped open and the Mode Ill timer sequences equipment onto the
,
                        Diesel Generator.
1
            -
                    D. The normal feeder breaker to 4.16 KV bus E1C is tripped open and the Mode II                          -
                        timer sequences equipment onto the Diesel Generator.
!
                                                                                                                                ,
L
l
l
I
.
!
.'
l
!
    ,            .
 
  . .- .          .- - . . . . -  .    .    . . -    .          - . . ~ _ ~ - - . - . -    ...-        _ . . -  _ _ ~
                                                                                                                      l
        .
                                                            STP Senior Reactor Operator Exam ... Page 52 of 100
          QUESTION 52:
          . A steam generator tube rupture has occurred on Steam Generator (SG) 18. The
          ruptured SG has been isolated in accordance with OPOP05-EO-EO30, Steam
          Generator Tube Rupture, and RCS cooldown and depressurization are in progress.
          SG 1B pressure is currently 910 psig and normal containment parameters exist.
          Which ONE of the following conditions warrant stopping the RCS depressurization?
          A. RCS Pressure is 900 psig
              PRZR Level is 4%
              Subcooling is 70*F
          B. RCS Pressure is 915 psig
              PRZR Level is 32%                                                                                      !
              Subcooling is 55*F                                                                                      !
          C. RCS Pre .ure s *.35 psig
              PRZR Lt. el is 5's%
              Subcookn.) ie *O*F
            D. RCS Pressure is 950 psig
              PRZR Level is 78%                                                                                      ,
              Subcooling is 85 F                                                                                      l
                                                                                                                      l
                                                                                                                      l
                                                                                                                      ll
                                                                                                                      ;
i-
!                                                                                                                      1
,
                                                                                                                      !
                                                                                                                      I
                                                                                                                      l
                                                                                                                      l
l
l              ,.                                . _ . . - . ,
 
              .    _ _ . ._    .    _ _ _ , _ _ _ . _    . _  . .  _ _ _ _ -            __  _      _    _ __  _ .
                                                        STP Senior Reactor Operator Exam ... Page 53 of 100
      QUESTION 53:
      A tube leak has occurred in 1D Steam Generator. The Unit is currently performing
      a rapid plant shutdown for repair.
      Which ONE of the following Process and Effluent radiation monitors performs an
      automatic function that will prevent a potential release to the environment during
    .this event?
      A. Condenser Air Removal System Monitor (RT-8027)
      8. Steam Generator 1D Blowdown Monitor (RT-8025)
      C. Turbine Generator Building Drain Monitor (RT-8041)
      D. Main Steam Line "D" Monitor (RT-8049)
                                                                                                            .
  s.
                                                                                                                  %
                                                                                                                        I
L                                                                                                                        l
,
                                                                                                                        l
 
                      .. _ .. - - _            _.                _ _ . _        ._ . _ _    __ . _
                                                                                                  l
                                                                                                  4
                                                                                                  i
                                          STP Senior Reactor Operator Exam ... Page 54 of 100
  QUESTION 54:
                                                                                                  1
  Following a LOCA, hydrogen concentration in containment has increased slowly                    !
  over five days reaching 3.5% by volume.
  Which ONE of the following actions will be taken?                                              l
                    .                                                                              1
  A. Both electric hydrogen recombiners will be placed in service immediately.                    I
                                                                                                  l
                                                                                                  ,
  B. Both electric hydrogen recombiners will be placed in service when hydrogen                  I
      concentration reaches 4.0% by volume.
  C. One electric hydrogen recombiner will be placed in service immediately.
  D. One electric hydrogen recombiner will be placed in service when hydrogen
      concentration exceeds 4.0% by volume.
                                                                                                  l
                                    -
4
4
i
                  ?
 
  ~ . . . . . .              - ,      . . . - , - - . - - - - - ~ - - - . - . . .                            . . - - - ~. . ..-- . - . ._..- ---
                                                                                                                                                        i
i
I
                                                                                    STP Senior Reactor Operator Exam ... Page 55 of 100
l                    QUESTION 55:
l
l.                  Given the following:
                    *  A Large Break LOCA has occurred                                                                  .
                    *  OPOP05-EO-EOOO, Reactor Trip / Safety injection, has been completed and the
l                        operators are transitioning to OPOP05-EO-EO10, Loss of Reactor or Secondary
                        Coolant.
                    *  The STA reports a Red path on Integrity and the operators transition to
                        OPOP05-EO-FRP1, Response to Imminent Pressurized Thermal Shock Condition.
                    *  While the crew is performing OPOP05-EO-FRP1, the STA reports that the cold                                                    J
                        leg recirculation switchover criteria are satisfied per OPOP05-EO-ES13, Transfer
                        to Cold Leg Recirculation.
                    Which ONE of the following describes the actions required to be taken by the crew
                    at this time?
                    A. Stop FRP1, perform all steps of ES13 and then return to FRP1.
                    B. Stop FRP1, perform the first 5 steps of ES13 and then return to FRP1.
                    C. Continue with FRP1 until completion, then perform all steps of ES13 and then                                                -
                        return to E010.
                  . D. Continue with FRP1 until completion, then perform the first 5 steps of ES13
                        and then return to E010.
                                                                                                                                                      .
I
!
                                                  .
                ,r-        -      ,                                                  - , , - ~        ,    y-
 
    . _ . . .      _m    .    __  _.____..__m.    _ _ _ . . _ -___  _ _ . _ _ _ _ . . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _    _
                                                          ~
                                                                                                                                          >
                                                                                                                                          I
'
                                                                    STP Senior Reactor Operator Exam ... Page 56 of 100
                                                                                                                                          l
                  QUESTION 56:                                                                                                            >
                  The plant has experienced a LOCA outside Containment and is currently on Cold                                          l
                  Leg Recirculation.- The, crew has transitioned to OPOP05-EO-EC11, Loss of
                  Emergency Coolant Recirculation as the LOCA could NOT be isolated in OPOP05-
                  EO-EC12, LOCA Outside Containment.
i                  in an attempt to conserve RWST inventory, the crew must stop backflow from the
L                  RWST to tne containment sump in accordance with EC11.
                  Which ONE of the following is the-correct sequence to perform this task?
  '
                                                                                                                                          }
                  A.1. Stop the LHSI, HHSI and CS Pumps
i                      2. Open the Si Pump Mini Flow Valves
                              .
                        3. Open the RWST to SI Suction Header Valves                                                                        ;
                        4. Close the Containment Sump Suction Valves
                        5. Start the LHSI, HHSI and CS Pumps as necessary                                                                  :
                  B.1. Stop the LHSI, HHSI and CS Pumps                                                                                  !
                                            _
                        2. Close the Containment Sump Suction Valves                                                                        ;
                                                                                                                                            '
                        3. Open the RWST to SI Suction Header Valves
                      '4. Open the S1 Pump Mini Flow Valves
                        5. Start the LHSI, HHSI and CS Pumps as necessary
l                  C.1. Stop the LHSI, HHSI and CS Pumps,
                        2. Open the RWST to SI Suction Header Valves
                            .
                      ' 3. Close the Containment Sump Suction Valves
                        4. Open the S1 Pump Mini Flow Valves
L                        5. Start the LHSI, HHSI and CS Pumps as necessary
                  D.1. Stop the LHSI, HHSI and CS Pumps
                        2. Open the Si Pump Mini Flow Valves
                        3. Close the Containment Sump Suction Valves                                                                      i
                        4. Open the RWST to SI Suction Header Valves
j
                        5. Start the LHSI, HHSI and CS Pumps as necessary
i
t
                                                                                                                                            l
              _ ._                                                                                                                  -
                                                                                                                                      .
                                                                                                                                          ,
 
    ..      ...~            . . . . .      -.              . - - .  _    _ - - . . - ~ -. -- - -  ..... . . - -
!
I
!
                                                          STP Senior Reactor Operator Exam ... Page 57 of 100
        QUESTION 57:
        Given the fo? lowing:
i
'
        *  . Unit 1 has experienced a small break LOCA
        *  RCS pressure is stable at 1785 psig
!      *  RVWL plenum is indicating 15%
        *  CETs are indicating 621 F
;
l
        In accordance with OPOP05-EO-FRC3, Response to Saturated Core Cooling, which
l      ONE of the following is the appropriate method for reestablishing desired
        subcooling?
        A. Establish conditions for HHSI flow
;      B. Establish conditions for LHSI flow
:
        C. Establish conditions for Accumulator injection
        D. Establish conditions for maximum charging flow
                                                                                                                    .
!  .
l
;
!
l
!
l
l
I~
,
,
f
f
$
l.
I
I
Y
                                                , . - , .            n                              ,  ,
 
,    . . . _          _ ..      - _ _ . _ - _ _ _ _    .. _ ___. _ ___ _ _ . _ . _.
                                                                                  -                _ _ . _ . . _ . . _ _ _ . . . _ . .
                                                                STP Senior Reactor Operator Exam ... Page 58 of 100
    QUESTION 58:
l    Given the following:
    .- Unit 1 is at 100% power
    a        All systems in normal operation
    a        Primary RO reports an increase in the Number 2 Seal Leakoff on all RCPs
!
:
'
    Considering each condition separately, wl:ich GNE of the following conditions is
    the cause for the increased leakoff flow?
                                                                                                                                        l
    A. Decrease in charging flow
    B. Increase in VCT pressure                                                                                                        ;
    C. Decrease in level of the Number 3 Seal Standpipe
    D. Increase in Seal Injection flowrate
l
l
  '
!
l.
                                                    ._
 
    . . . _ _ .  .    ..            .-. -. .    . ~.. ..- - ........-.-. - - -                ~ . -- .... -    .  .  .
                                                                                                                            _
!
                                                              STP Senior Reactor Operator Exam ... Page 54 of 100
                  QUESTION 59:
                . in accordance with OPOP02-RC-0004, Operation of Reactor Coolant Pumps, which
                _.ONE of the following conditions requires tripping the reactor and the affected                          '
                                                                                                                            ,
l-                reactor ecolant pump?- (Censider each condition separately)
                  A. #1 seal water inlet temperature - 250*F
                      .
                  B. Motor stator winding temperature - 300 F                                                              ;
                                                                                                                            l
                  C. #1 Seal Delta-Pressure - 350 psid                                                                    j
                  D. Shaft Vibration - 10 mils                                                                            i
                                                                                                                            J
                                                                                                                            i
                                                                                                                            l
l
                                                                                                                            I
l                                                                                                                    .
                                                                                                                            i
                                                                                                                        *
                                                                                                                            ;
  .-
!
i
4
                                    r
 
                                  - . . ,      .    ~ -            ~~    ._      .- ..      . - -
                                                                                                    1
-
                                                                                                    1
                                            STP Senior Reactor Operator Exam ... Page 60 of 100    i
i
i QUESTION 60:
  Unit 2 is operating at 50% power when a RCP 2A NO 2 SEAL LKF FLOW Hi alarm
  is received. After investigation it has been determined that RCP 2A # 3 seal has
  NOT failed. In accordance with OPOPO4-RC-0002, RCP Off Normal, the control                        l
l room operators are directed to monitor:
  A. RCP 2A vibration and if it starts increasing perform a plant shutdown then
      secure RCP 2A.
  B. RCP 2A standpipe fill rate and if it starts increasing perform a plant shutdown                i
                                                                                                    '
      then secure RCP 2A.
  C. RCP 2A for 48 hours to determine if # 2 seal has not fully seated.
                                                                                                    !
  D. RCP 2A vibration and if limits are exceeded trip the reactor, trip the turbine, and
      secure RCP 2A.                                                                                i
                                                                                                    i
                                                                                                    I
l
l
                                                                                                    l
I
                                                                                                    i
                                                                                                    !
                                                                                                    !
i
                                                                                                    l
                                                                                                    l
                                                                                                    l
 
  ,              --            .    . - -      --__        . . _ . - . . .          _ - . .        ._. _
                                                                                                          I
                                              STP Senior Reactor Operator Exam ... Page 61 of 100
      QUESTION 61:
      Given the following:
      .  Unit 1 is at 25% power
      .  CCP 1B is out of service for maintenance
      .  CCP 1 A has tripped and the field supervisor is investigating
      .  PDP will NOT start
      .  RCP 1C has experienced an RCP 1C UPPR OIL RSVR LVL Hl/LO annunciator
      *  RCP 1D CCW to thermal barrier temperature sensor (TS-4638) fails high
      Which ONE of the following is the proper control room operators' response to the
      above plant conditions?
      A. Commence a power reduction to < 10% for performance of a containment
        entry to investigate problem with RCP 1C.
      B. Within one minute trip the reactor and trip RCP 1D.
      C. Commence a power reduction to < 10% and trip RCP 1D.
    -
      D. Within one minute trip the reactor and trip RCP 1C.                                      -
r
 
  - -      ~  -  -      . .. . . . -  .. .-_ - .., --- .~ .-.-...- - .- - -.-                                          .    .  .      . . . -
                                                                            .
                                                                                      STP Senior Reactor Operator Exam ... Page 62 of 100
                QUESTION 62:
l
                During operations at 95% power with pressurizer level at 53%, the Tave input
                (Auctioneered High) to the pressurizer level controller fails low.
                Which ONE of the following is an indication that this failure has occurred?
l              ' A. Backup heaters deenergize, charging flow control valve opens, low level
l                  deviation alarm actuates.                                                                                                        l
l
!              B. Backup heaters energize, charging flow control valve closes, high level deviation
!
                    alarm actuates.
!              .C. Backup heaters energize, charging flow control valve _ opens, low level deviation
l                  alarm actuates.
l
                D. Backup heaters deenergize, charging flow control valve closes, high level                                                        ;
i                  deviation alarm actuates.                                                                                                        j
                                                                                                                                                    i
;
l
                                                                                                                                                    ,
!
!
I
!
l'
!
.
i
)._
<
l
      . . ,                          -  -                ,.                      .-.                                                    -
                                                                                                                                                  ,
                                                                                                                                                    I
 
    .- . .    _-.          . .  -.    - . . . - . - - _ ~ . _          - .    . . -    . -    .-      ..          . . . - -.
t-
                                                                                                                                    I
l
t.
'                                                              STP Senior Rec? tor 0,,erator Exam ... Page 63 of 100
                                                                                                                                    l
          QUESTION 63:
          Given the following:                                                                                                      i
          a  Containment pressure transmitter PT-937 has been declared inoperable.
l-
          *  The required Technical Specification actions have been taken to remove PT-937                                          l
              from service.
                                                                                                                                    l
          Under these conditions, which ONE of the following statements describes the
          coincidence for a Containment Spray Actuation to occur and the actions that will
          result in this coincidence?                                                                                              !
          A.1/3 coincidence after the channel is placed in the TRIP condition.
i          B.1/3 coincidence after the channel is placed in the TEST BYPASS condition.
l
          C. 2/3 coincidence after the channel is placed in tite TRIP. condition.
          D. 2/3 coincidence after the channel is placed in the TEST BYPASS condition.                                              I
                                                                                                                                    ;
                                                                                                                          e
                                                                                                                                    i
,
                                                                                                                                    !
                                                                                                                                ,
'
                                      ..                                                                                            l
!
                                                                                                                                    l
                                                                                                                                    ;
  '
i
?
                        *
                              -                      -
                                                                                    ,_                  -    -      .      - ,. .
 
  .                      _          --_          . . _    _ ~    . -    . _ . . . _ _ . . _ _ _
                                                                          .                    _ _ . - . . . .
                                              STP Senior Reactor Operator Exam ... Page 64 of 100
      QUESTION 64:
      Per OPGPO3-ZF-0018, Fire Protection Operability Requirements, which ONE of the
l
'
      following identifies the minimum operability requirements for the Fire Protection
      Water Supply System?
    .
              Storage Tank Volume        Number of Pumps              Pump Capacity
                  gallons - each                                                  in gpm
        A.          200,000                          1                            2500
l        B.          250,000                          2                            2000
l        C.          300,000                          2                            2500
        D.          350,000                          1                            2000
!
!
,
!
                                                                                                                1
                                                                                                                l
l
;
l
!
!
r
 
    ._. .              -          .-    . . . . . - - . .    --        . . .  - - - -    . .  - .-.
                                                          STP Senior Reactor Operator Exam ... Page 65 of 100
.
            QUESTION 65:
            Which ONE of the following STPEGS Fire Protection Sprinkler Systems provides
            defense against Class "C" Fires?
            - A. Cable Tray sprinklers'
:
            B. Carbon Filter sprinklers
:
            C. DG FOST sprinklers
*
            D. Turbine Bearing sprinklers
.
                                                                                                1
..
.
          O
$
                                                                                                              .
O
i'
1
,
k
 
            .    ..  -              - - . -      . . _ _ . - .          .  .      ._        .. .
.
                                              STP Senior Reactor Operator Exam ... Page 66 of 100
    QUESTION 66:
    Given the following:
    *  Unit 1 is at 100% power and Unit 2 is in Mode 6 during a refueling outage.
    *  Fire Hydrant (FH) #13 serving the Essential Cooling Water (ECW) Intake
      Structure area is declared inoperable.
-
    Which ONE of the following actions must be taken due to this inoperable Fire
    Hydrant?
i  A. Restore FH-13 to operable status within 7 days or take the actions of T.S.
      3.0.3 due to EC'N system inoperability.
;  B. Establish a continuous fire watch within 1 hour for the Unit 1 ECW intake
<
      structure and components.
    C. Establish an hourly fire watch for the Unit 1 ECW intake structure and
  ,
      components.
    D. Provide additional lengths of 21/2 inch file hose from an operable Fire Hydrant                '
      within 1 hour to service the ECW intake Structure.
                                                                                                      l
                                                                                                      I
                                                                                                      /
 
  _ . ._. _ _ _ _ _          . ._ _ _ _ __._ _ _ .__. _ _ _ _ . _ _ _ _ . . _ .. _ _ _ . . . __ _ _ _ . . . . . ,
                                                                                                STP Senior Reactor Operator Exam ... Page 67 of 100
            QUEST ON 67:
            A Unit 1 Reactor Trip from 100% power has occurred due to a loss of off-site
            power caused by severe weather and tornado damage to the switchyard and the
            138KV Blessing Line. A Tornado Warning is still in effect.
            Given the following:
l
l
            a      Train B 4.16 KV Bus is energized from SDG #12.                                                                                      ;
j        '.        Train C 4.16 KV Bus is deenergized due to a bus overcurrent lockout.
;
            *        SDG#11 was secured due to a failed turbocharger thrust bearing and
l                  turbocharger damage.
            * - Restoration of any off-site power source is NOT expected for at least 6 to 8
l
l                    hours.
l'          =        Restoration of SDG #11 is NOT expected for at least 10 hours.
l          .- Restoration of Train C 4.16 KV Bus is NOT expected for at least 6 hours.
          ' Based on these indications, which ONE of the following statements is
          .
            recommended per OPOPO4-AE-0001, Loss of Any 13.8 KV or 4.16 KV Bus?
,
            A. Cross-connect Train A QB C 4.16 KV Bus to Train B 4.16 KV Bus to restore
l                    power to a CCP for RCS inventory control.                                                                                      -
!
!          B. Cross-connect Train A QB C 4.16 KV Bus to Train B 4.16 KV Bus.to restore
                    power to Pressurizer Heaters for RCS pressure control.
l          C. Do NOT Cross-connect Train A QB C 4.16 KV Bus to Train B 4.16 KV Bus                                                                  -
                    because the only SDG would be put at risk of overloading.
            D. Do NOT Cross-connect Train A QB C 4.16 KV Bus to Train B 4.16 KV Bus
,
                    because severe weather conditions exist.
,
i
.
i
I
}:
                                                                                                                  .        _  __
                                                                                                                                                -.
 
                                  .
                                          STP Senior Reactor Operator Exam ... Page 68 of 100 )
QUESTION 68:
Given the following:
*  Unit 1 is in Mode 5
*  RCS temperature is 180 F
a  RCS pressure is 225 psig
*  RHR Train 1 A is in service maintaining RCS temperature (with both RHR HX
    BYPASS FLOW CONTROL and OUTLET TEMPERATURE CONTROL valves in a
    throttled position)
+    125 VDC system E1 A11 TROUBLE annunciator is received
.  E1 A11 bus voltage indicates 0 volts
Which ONE of the following describes how the operating RHR train will respond?
                                                                                              l
            BYP FLOW CONT            OUTL TEMP CONT                      RCS
                (FCV-0851)              (FCV-0864)              TEMPERATURE
  A.              OPENS                    CLOSES                    INCREASES
  B.              OPENS                  NO CHANGE                  INCREASES
  C.          NO CHANGE                    OPENS                    DECREASES
  D.              CLOSES                    OPENS                    DECREASES
 
                                                                                                    1
                                                                                                    l
                                                                                                    i
                                                                                                    '
                                        STP Senior Reactor Operator Exam ... Page 69 of 100
                                                                                                    i
QUESTION 69:
Unit 2 was at 100% power, steady state conditions, when an electrical
disturbance occurred that produced the following Control Room indications:
.  2A Steam Generator PORV controller is NOT illuminated and cannot be operated
    from the Control Room.
.  Ventilation Monitors RT-8012, RT-8033, and RT-8035 are indicating Magenta
    color on the RM-11.                                                                            1
.  All three trains of the Control Room Envelope, Fuel Handling Building, and
    Containment HVAC are in the ACTUATED condition.
.  Charging Flow Control valve FCV-205 is in Automatic and closed.
.  Several QDPS Plasma displays in the Control Room are blank.
Which ONE of the following power distribution panels has lost power?                                ;
A.120 Volt AC Class 1E Panel DP1201
B.120 Volt AC Class 1E Panel DP001
C.120 Volt AC Class 1E Panel DP1202
                                                                                                .
D.120 Volt AC Class 1E Panel DP002
                                                                                                  ~
                                                                                            .-
 
    . . ~ . .  .        .~,- - . ~    . . . . _ , . - . - . - . . . . .      - ~ . - . -  .~    .    .    ..      ._ .-
!
I                                                                STP Senior Reactor operator Exam ... Page 70 of 100
l
              QUESTION 70:
l              Considering the following time line:                                                                        ,
l
              =  1200 - Large Break LOCA occurs
              .  1214 - Alert is declared, based on leakrate greater than makeup capability
              .  1232 - Loss of all HHSI and LHSI pumps is reported
              +  1251 - Containment hydrogen concentration > 0.5 %
              +  1300 - Sight Area Emergency is declared, based on high RCB radiation
              .  1343 - Core Damage confirmed by failed fuel monitor and RCS sample results
              =  1420 - General Emergency is declared, based on' rapid decrease in RCB pressure
:              Which ONE of the following is the EARLIEST time at which a Protective Action
l              Recommendation (PAR) is REQUIRED to be issued? -
              A.1315
              B.1358
              C.1403
              D.1435                                                                                                        ,
                                                                                                                            ;
                                                                                                                            l
,
!
!'
I
f
1
,
  ,
 
                                                                                                          l
                                                  STP Senior Reactor Operator Exam ... Page 71 of 100
t
:    - QUESTION 71:
.
.      Maintenance requests an Equipment Clearance Order to perform adjustment and
      testing of limit and torque switches for a motor. operated valve (MOV).
                                                                                ~
!      In order to perform the required testing, the MOV:
>      A. can only be operated under a Test Tag by the Operational Authority when
4
          requested by the individual or classification performing the maintenance.
[      B.' can only be operated under a Caution Tag by the Acceptor to whom the
j          Caution Tag is issued and only after Operational Authority approval.                          l
I
      C. shall be tagged with both a Caution Tag and a Test Tag with the Caution Tag
          stating that removal of the Test Tag permits operational testing.
;
;      D. shall be tagged with both a Danger Tag and a Test Tag, with the Danger Tag
!          removed prior to any MOV operation during the tssting.
l-
I
  +,                                                                                                .
,
!                                                                                                      ~
 
, . _ _ . _ _ _ . . .            . _ . _ _ . . . . . _ _ . . . . _ _ . . .  - . _ . . _ _ _ . . . _ . _ _ _ _ _ _ _ . , ._...._.. ___ _ ..._.. _ __ _ .
i
1
                                                                            STP Senior Reactor Operator Exam ... Page 72 of 100
                  QUESTION 72:
                Which of the following situations require Independent Verification?
                  1. Initial system valve lineup of TGB Fire Protection System following an outage.
                ' 2. Removing electrical grounding device from the 13.8 KV switchgear following
                        breaker maintenance.
                  3. Functional test has verified the resetting of overcurrent trip relay setpoint on
                        RCB Chiller 11 B breaker,1 D2/6, following electrical troubleshooting.
                  4. Throttling and locking CV-0198 prior to Mode 4 entry per Technical
                        Specifications.
                        A.1,2,3
                        B.1,2,4
                        C.1,3,4
                        D.2,3,4
l
l
l
t-
h
I
                  - - .    .
                                                                          ,                  _
 
                                                  .. . . .  . . . - - . . ~ .              . . - - . - .
*
.
^
                                              ' STP Senior Reactor Operator Exam ... Page 73 of 100
                                                .
                                                                                                              i
:                                                                                                            !
,
      QUESTION 73:-                                                                                          ;
                                                                                                              I
*
      Given the following plant conditions:
:  -a  Unit 1 was operating at 100% power when a loss of offsite power occurred.
'
      .  The following indications' are observed while performing the immediate actions
          of OPOP05-EO-EOOO, Reactor-Trip or Safety injection:                                                ,
'            * All Rod Bottom Lights are NOT LIT
                                                                                                              l
              *
                  Reactor Trip Breaker R is OPEN -                                                            i
                                                                                                              '
              *
                  Reactor Trip Breaker S is CLOSED
              *
                  Bypass Breakers R and S are OPEN
              *
.
                  N35 and N36 IR Channels indicate decreasing neutron power
!    *  The previously observed indications do NOT change after attempting a manual
)        trip from both reactor trip switches                                                                j
.
      Which ONE of the following actions should be taken by the crew?
!.
;    A. Go to OPOP05-EO-FRS1, Response to Nuclear Power Generation - ATWS.
        .
                                                                                                              l
<
                                                                                                              1
      B. Dispatch an operator to shift DRPl to the alternate power source, verify all Rod
.
          Bottom Lights are lit, then continue on with EOOO.
                                                                                                          *
!                                                                                                              !
      C. Open 480V Load Center 1K1 and 1L1 feeder breakers, then continue on with                            ;
          EOOO.
      D. Continue on with the immediate actions of EOOO.
i                                                                                                          ~
                                                                                                              ,
*
'                                                                                                            l
                                                                                                              l
i
#
  ,
!
                                                                                                              i
                                                                                                              l
                                                                                                              !
                                                                                                              i
                              .
 
                            . . _ .          __    _ _. _        ..    -            .    _. . . _ .
                                    .
                                          STP Senior Reactor operator Exam ... Page 74 of too
  QUESTION 74:
l Given the following:
!
j .  A Reactor Trip occurs due to a Loss of Offsite Power.
  a  The ESF diesel generators have all started and restored power to their ESF
,
    buses.
!
  *  The Control Room crew has just completed the immediate Actions of
    OPOP05-EO-EOOO, Reactor Trip or Safety injection.
  Which ONE of the following correctly identifies the status of containment cooling?
  A. The RCFCs are NOT running and CCW is flowing through the cooling coils.
  B. The RCFCs are running and there is NO flow through the cooling coils.
  C. The RCFCs are running and CCW is flowing through the cooling coils.
  D. The RCFCs are NOT running and there is NO flow through the cooling coils.
l
I                                                                                                    '
I                                                                                                    l
l
                                                                                                      l
                                                                                                      !
 
      -
            -
                        ~_.-                                                              ~.  . . - . . .
I
                                                                                                                ]
'
                                            STP Senior Reactor Operator Exam ... Page 75 of 100                ;
    QUESTION 75:
!
    Unit 1 is in Mode 4, coming out of a refueling outage, when a loss of instrument                            j
    air occurs.
                                                                          .,
    Given the following:
    *  .PCS temperature is 300"F
,
    a    tiarrow Range SG levels; 1 A = 40%,1B = 35%,1C = 55%,                  1D = 20%
    *    SG PORVs are all operable
    *    All Motor-driven AFW Pumps are operable
    * ' RHR Train status; 1 A and 1C - available,1B inservice and supplying Low
                            '
        Pressure Letdown
    *    St Pumps 'are in the normal Mode 4 alignment
    *    Both Pressurizer PORVs are operable
    .    Vacuum has been established in the Main Condenser
    * -Three Circulating Water pumps are running
    Which ONE of the following RHR valve alignment methods are Control Room                                    i
    operators directed to use for RCS temperature controlin accordance with                                    l
    OPOPO4-IA-0001, Loss of Instrument Air, Addendum 67
  .
                                                                                                            -
    A. LOOP B TC INJ MOV-0031B - open, cycle RHR B Heat Exchanger OUTL TEMP
        CONT HCV-0865 as needed for temperature control.
    B. MINI FLOW MOV-0067B - open, cycle RHR B Heat Exchanger BYP FLOW CONT
        FCV-0852 as needed for temperature control.                                                          -
    C. MINI FLOW MOV-0067B - open, cycle LOOP B TC INJ MOV-0031B as needed
                                                                                                                '
        for temperature control.
    D. LOOP B TC INJ MOV-0031B - open, cycle RHR B Heat Exchanger CCW OUTL
        FV-4548 as needed for temperature control.
                                                                                                                l
i                                                                                                              l
;
                                                          .                                            .,
 
  . . _ . _      ..  . _ . = . . . . . .    ._ ._  _.    _ . _ . . _ _ . _      ______ ...__ __              . _ _ _ _ ,_
                                                                                                                          ,
                                                                                                                          1
                                                          STP Senior Reactor Operator Exam . . Page 76 of 100
            QUESTION 76:
            Given the following:
            .  Unit 2 is at 100% power, steady-state conditions
            .  CCP 2B is operating, CCP 2A is available
            .  CHG FLOW Hl/LO alarm is received on Lampbox 04M8
            .  Operators determine that CCP 2B has tripped
            .  The Unit Supervisor directs the Primary RO to start CCP 2A                                                  l
            Which ONE of the following valve alignments should be performed prior to starting
            CCP 2A in accordance with the alarm response instructions?
,
                    MOV-8377A, 2A CCP              FCV-201, 2A CCP              FCV-205, Charging
i                    Discharge Isolation              Miniflow Valve          Flow Control Valve
              A.                        OPEN                OPEN                                CLOSED
              B.                    CLOSED                  OPEN                                OPEN
              C.                        OPEN              CLOSED                                  OPEN
              D.                    CLOSED                  OPEN                                CLOSED
 
  ..    . . . . . _ _
                -
                                  . . _ . _ . _ _ _ . _ , _ . . _ . - _ _              _.- --. .___ _ . _ .. . _ _ _ _ _ . . - . _ . _ _      _ - . - .
                                                                                                                                                          l
                                                                            STP Senior Reactor Operator Exam ... Page 77 of 100
              QUESTION 77:
            Given the following:
          .*          Unit 1 is in Mode 5                                                                                  .
                                                                                                                              7
              * RCS boron concentration is 2800 ppm
              .        Rods are locked in the Rapid Refueling position
          --*        Preparations are being made for refueling operations
          . A fire has been reported in the Train C ESF Switchgear room and it appears to be
            . spreading toward the Rod Drive MG Set Room.
              In accordance with OPOPO4-ZO-0008, Fire / Explosion, which ONE of the following
              describes the operator actions to be taken? -
            ' A. Open the reactor trip breakers and ensure all rods are fully inserted
              B.' Place Rod Holdout Power Supply to OFF and ensure all rods are fully inserted
              C.: Direct I&C to restore rod control, then fully insert all control rods
      -
              D. Initiate emergency boration until RCS boron concentration is a 3000 ppm-                                                  -
                                                                                                                                                    -    I
    .
                                                                                                                                            *
.
                                                                          T      -m                                                  -  -              .
 
                                                                                            1
                                                                                            l
                                                                                            l
                                        STP Senior Reactor Operator Exam ... Page 78 of 100
QUESTION 78:
While responding to an inadequate Core Cooling situation the operators are
directed to reestablish some form of core cooling.
Which ONE of the following statements describes the INITIAL response of the
CETs, and the reason for this response, if the operators restarted the safety
injection pumps?
A. Increase due to saturated steam being forced out of the core.
B. Decrease due to saturated steam forming a frothy two phase mixture.
                                                                                            '
C. Increase due to superheated steam being forced out of the core.
D. Decrease due to superheated steam causing injected water to boil forming a
    frothy two phase mixture.
                                                                                            :
                                                                                  .
 
                                                                                .        ..
                                              STP Senior Reactor Operator Exam ... Page /9 of 100
  QUESTION 79:
  The following plant conditions exist:
  ( Loss of seal cooling (CCW AND seal injection) to all RCPs              -
  a  f'anual reactor trip _and trip of all RCPs
  .  Performed OPOP05-EO-E000, Reactor Trip and Safety injection and
    OPOP05-EO 'dS02, Natural Circulation Cooldown
  .  Transitioned to OPOP05-EO-ESO3, Natural Circulation Cooldown with Steam
    Void in Vessel due to limited inventory in the AFWST
  *  RVWL - Upper Head is 80%
  .  Pressurizer level is 30%
  *  RCS Subcooling is 90*F
  While performing ES03, seal cooling is restored to RCP D.
  Which ONE of the following actions must be performed prior to starting RCP D7
  A. Decrease pressurizer level to accommodate void growth
  B. Increase pressurizer level to accommodate void collapse
.
                                                                                                  .
  C. Decrease subcooling to accommodate void growth
  D. Increase subcooling to accommodate void collapse
                                                                                                    .
 
  .--      - -              --      _  _ ._- - _ . .    . ~ _ =          = . .  - . ~ .        _ . _ _
                                        -
!-                                                                                                                l
!                                                                                                                1
I
                                                STP Senior Reactor Operator Exam ... Page 8o of 100            q
    QUESTION 80:
!
    A plant startup is in progress with main turbine roll commencing and reactor power
    at 15%. All procedural requirements of OPOP03-ZG-0005, Plant Startup to 100%
l  have been met for the current power level. Power range channel N-44 is out of
l  service due to a failed detector.
    Which ONE of the following reactor trip signals will cause the reactor to trip under
    these conditions?-
    A. Pressurizer low pressure
    B. Single loop loss of flow
    C. Intermediate Range high flux
    D. Turbine trip
                                                                                                                  1
                                                                                                                  I
                                                                                                                J
                                                                                                                  l
                                                                                                                l
                                                                                                                f
                                                                                                                l
i
                                                                                                                !
l
                                                                                                            .
                                                                                                              _
 
          .  ..          . _ _ . . . . ... . - .    - - - -. - . - . . . _ . .    - - - . . - .      . -
!
                                                                                                              ,
                                                  STP Senior Reactor operator Exam ... Page 81 of 100
.
    QUESTION 81:
    Given the following:
    *  Unit 1 is at 100% power                                                                                !
                                                                                ..
    *  Reactor power is slowly increasing
    *  RCS Tave is 588'F and slowly decreasing
    *  Turbine control is in IMP OUT
    Which ONE of the following conditions is causing the above plant transient?
    A. Main Turbine Governor valve # 3 has failed closed.
    B. . Extraction steam to a feedwater heater string has been isolated.
    C. Main Steam to Deaerator valve has failed closed.
    D. Circulating Water pump #14 has tripped.
  .                                                                                                    .
                                                                                                              ,
                                                                                                          e
!
!
!
.
                                                                                                            )
 
                                                                          . _ _ _ _ _ _ _ . _ _ _ _ _ _ - _ . . . . . . _ . _
  .
                                            STP Senior Reactor Operator Exam ... Page 82 of 100
    QUESTION 82:
                                                                                                                              i
    Given the following:
    *  Unit 1 has tripped from 100%
    *  All steam generator levels indicate approximately 50% WR
    *  A fault has left Vital DC bus E1D11 deenergized
    Which ONE of the following describes the effect this event has on the status of
    #14 AFW pump, its associated AFW REG valve and AFW OCIV?
                                                                                                                              i
    A. Pump will start, REG valve will remain closed and OCIV will remain open                                                )
    B. Pump will NOT start, REG valve will remain closed and OCIV will remain open
    C. Pump will start, REG valve will remain open and OCIV will remain closed
    D. Pump will NOT start, REG valve will remain open and OCIV will remain closed
l
i
'
i
I
:
I
,
 
    - _.  ..    _.    .      .  . - . _    _ - . . = .      . .  , .  _.    -    -    . . - .  .      -  -  ..
!
                                                            STP Senior Reactor Operator Exam ... Page 83 of 100
              QUESTION'83:
              Given the following:                                                                                        !
                                                                                                                          '
I            *  Unit 1 is in MODE 5 with RCS cooldown to refueling conditions in progress.
              .  The MEAB watch enters the FHB and discovers the Spent Fuel Pool (SFP) Area
                Radiation Monitors RT-8090 and RT-8091 in the HIGH ALARM condition.
l            *  SFP level is 66 feet 61/2 inches and has decreased 1/4 inch in the last 24                                ;
I                hours.
              *  SFP temperature is 90 F and steady.                                                                      :
              *  No other abnormal plant or alarm conditions exist.
              Which ONE of the following actions must be taken?
              A. Notify Control Room of SFP level and commence makeup to restore level.
l
              B. Verify FHB HVAC is operating in the Emergency Mode.
l
              C. Notify personnelin the affected area of the High Radiation alarm.
        -
              D. Verify SFP Gate Seal Air Pressure is in the normal band (30-34 psig).                          -
i
[
t
l                                                                                                                    *
,
l
!
                                                                                                                          i
!
!
!
f
{
1
l
!
  '
,
E
 
      . - - . . _ _-      __    __._.m._  .m . _ _ . . . . . . _ -_ _ . . _ . . _ _ . . _ . . - . - . . . _ - . - _ __ - . . . _ . . _ . _ - . . . _ . . _ . . _ _ _ . . _
  .-
                                                                            STP Senior Reactor Operator Exam ... Page 84 of 100
                      QUESTION 84:
                      It has been determined that Spent Fuel Pool level needs to be raised to account for
                      loses due to evaporation.
                      Per OPOP02-FC-0001, Spent Fuel Pool Cooling and Cleanup System, which ONE of
                      the following source combinations is used to makeup to the SFP to compensate for
!-                    normal EVAPORATION?
                      A.-      Reactor Makeup Water System or Refueling Water Storage Tank
                      B.      Reactor Makeup Water System or Demineralized Water System
                    -C.      Boron Recycle System or Refueling Water Storage Tank
l
l                    D.    ' Boron Recycle System or Demineralized Water System
!-
!                                                                                                                                                                            l
n
l
l
I
f
l
l
l
:                                                                                                                                                                            1
!                                                                                                                                                                            l
;                                                                                                                                                                            '
I
                                                                                                                                                                              !
                                                                                                                                                                              I
                                                                                                                                                                              l
                                                                                                                                                                              l
i
i
k                                                                                                                                                                          .
 
                                      . . . _ _ . _    ___. .        _ _ _ _ = _ . . . _ _ _ _ _ __ __  _ _  _
                                                    STP Senior Reactor Operator Exam ... Page 85 of 100
    QUESTION 85:
    Upon loading a spent fuel assembly into the reactor core the Core Loading
    Supervisor notifies the Control Room that an excessive amount of gas bubbles are
    emerging from an adjacent fuel assembly and that RCB 68' Area Monitor RT-8099
    Just went into High alarm. The Control Room performs the actions of
    OPOPO4-FH-0001, Fuel Handling Accident, and directs the operation of the
    Containment Carbon Filter Units.
    Which one of the following statements is correct?
    A. The Containment Carbon Filter Units are NOT required to be in operation until
      RT-8099 is determined to have a VALID High alarm.
    B. The Containment Carbon Filter Units are NOT required to be in operation until
      MORE than one RCB Area Monitor is determined to have a valid High alarm.
'
    C. The Containment Carbon Filter Units are required to be in operation if radiation
      levels or sampling indicates fuel cladding damage has occurred.
  .
,
    D. The Containment Carbon Filter Units are required to be in operation anytime
                                                                                                        -
      OPOP04-FH-0001 is entered.
                                                                                                                    l
                                                                                                              s
r
                                                                                                                    l
                                                                                                                    l
l
?                                                                                                                    l
l                                                                                                                    l
l
,
l
l                                                                                                                  )
 
                                    .
                                          STP Senior Reactor Operator Exam ... Page 86 of 100
  QUESTION 86:
  OPSP03-ZO-0028, Operator Logs, requires that RCFC outlet temperatures be
  maintained less than or equal to 75 F and greater than or equal to 50 F.
  Which ONE of the following r,tatements is the reason for these temperature limits?
        Hioh Temoerature Basis          Low Temoerature Basis
  A.  CRDM Cooling                    SI Accumulator Nil Ductility Temp.
  B.  LB LOCA Peak Clad Temp.          Si Accumulator Nil Ductility Temp.
  C.  CRDM Cooling                    SI Accumulator Boron Solubility
,
  D.  LB LOCA Peak Clad Temp.          SI Accumulator Boron Solubility
l
;
l
                                                                                                1
                                                                                                :
                                                                                                )
l                                                                                              l
.
A
                                                                                                l
                                                                                                l
                                                                                                l
l
!
l
l-
I
t                                                                                            )
 
  . - .- - .-- .-. ., -.                                ~.- _ .            .-    .. ~~ ~. .. . - - . . . -        . - . -  _
i-
i
l
I                                                                STP Senior Reactor Operator Exam ... Page 87 of 100
              QUESTION 87:
              A point source in the RCA is reading 500 mrem /hr at 1 foot. There are two                                          j
              options being considered to perform a job in the area.
                                                                                                                                  '
                                                                                                            '
                  Option #1: Two operators working together are capable of com[leting the job in
                                    20 minutes at 4 feet from the source.
1
                  Option #2: One operator can complete the job in 80 minutes at 8 feet from the
                                    source.
                                                                                                                                  I
i            Which ONE of the following is the preferred option AND consistent with the goals                                    !
!            of the ALARA program?
l                        A. Option #1 - each operator's exposure is 10.0-10.5 mrem
l
                          B. Option #2 - the operator's exposure is 5.8-C.2 mrem
I
                          C. Option #2 - the operator's exposure is 10.0-10.5 mrem
                          D. Option #1 - each operator's exposure is 5.8-6.2 mrem
l
                                                                                                                            s
                                            e
l
:
l
i
L
o
i
i
:
                                                                                                                                }
 
      .    . _        _ . . _ _ _ . . . _ . .    _ , . _ _ .  ...___._-..-      ._-  _ .  - . _ _ _ _ _ - _    ~._. . _ . . . .
l
i.
I.
I                                                            STP Senior Reactor Operator Exam ... Page 88 of 100
l
l    QUESTIOrJ 88:
!
l    The maximum dose rate in a room is 1200 mrem /hr one foot from a certain valve.
l    According to the STPEGS Radiological Access and Work Controls procedure, this
    room should be classified as which ONE of the following?
:
l    A. High Radiation Area
:    B. High Radiation Exclusion Area
l    C. Locked High Radiation Area
l
l    D. Very High Radiation Area
!'o
i
l
,
I
:
:
f                                                                                                                                I
I                                                                                                                                I
l
;
                                                                                                                                ;
                                                                                                                                i
              e
l
t'                                                                                                                              1
4                                                                                                                                i
i
.
I
                                                                                                                                I
                                                                                                                                l
  '
                                              .-            -
 
    - - - - . -.- .~ -._ -. -                    - -- - - -.. -                - -            - -.              . ..  . .  ..-_,
  '
                                                                  STP Senior Reactor Operator Exam ... Page 89 of 100
l
                    ' QUESTION 89:
l-                  ' Given the following:--
(                    *    Unit 1 is operating at 100% power.                                .._
i                    e 1CCW surge tank level is 58% and decreasing at 1 % per minute
l                    *    Crew enters OPOPO4-CC-OOO1, Loss of Component Cooling Water
l_
                      e    Operators trip the reactor and all RCPs
I                    *  . CCW pump 1 A is out-of-service for maintenance
                      *    CCW pump.1B trips on thermal overload
                      *    .CCW pump 1C is running                                                                                l
l                    Which ONE of the following describes the charging pump (s) wiiich may be
l                    operated under the current plant configuration?
:
l                    A.1 A Centrifugal Charging Pump ONLY
;
                                                                                                                                  l
l                    B.1B Centrifugal Charging Pump ONLY                                                                          1
l.                    C.1 A Centrifugal Charging Pump OR Positive Displacement Charging Pump
                .
                      D.1B Centrifugal Charging Pump OR Positive Displacement Charging Pump                          -
                                                                                                                                  l
                                                                                                                          .
                                                                                                                                  J
,
~
                                                                                                                                  \
;
!                                                                                                                                  !
1
!
i
t
4
1
  '
 
    . _ _ . . _ .            _  _. .      . _ - _ _ . _-_=_-.~.__.__.m._                  _ . _ _ _ _ . _ _ . _ . . _ _ . _ - _ . _ . ._
i
                                                                                                                                          I
I
<
                                                                                                                                          l
                                                                                                                                          ,
                                                                                                                                          1
    -
                                                            STP Senior Reactor Operator Exam ... Page 90 of 100                          l
                QUESTION 90:
                Which ONE of the following conditions w5 warrant entry into OPOPO4-CC-0001,
                Loss of Component Cooling Water?
                A. Trip of running CCW pump (s) and standby CCW pump fails to start.
                B. Loss of ECW flow to all CCW Heat Exchangers.
t                                                                                                                                        1
!              C.-Vital CCW Header Supply valves automatically close.
                D. Leak in CCW system greater than makeup capacity.
                                                                                                                                          l
'
                                                                                                                                          .
                                                                                                                                          l
I
t.
!
l
'.
i
r
  '
i
I
                                                                                                                                          i
!
 
  --    .    . . . -    .  - - - - - .    . . . . __- - - .    -.        ~.    .    . . .
                                                                                                      !
a                                                                                                      i
                                                STP Senior Reactor Operator Exam ... Page 91 of 100
!
l    QUESTION 91:
      A modification to the Containment Spray System is to be made and a 10CFR50.59
:    Evaluation is to be performed.
.                                                                            ,-
l    Which ONE of the following describes the 10CFR50.59 process?
5
i    A. The modification undergoes a Preliminary Screening followed by a Final
l        Screening then a Unreviewed Safety Question Evaluation before it is reviewed
#
        by the PORC.
7
      B. The modification undergoes a Unreviewed Safety Question Evaluation followed
}-      by a Preliminary Screening then a Final Screening before it is reviewed by the                j
        PORC.                                                                                        l
                                                                                                      :
                                                                                                        l
l    C. The modification undergoes a Final Screening followed by a Unreviewed Safety                  ;
j        Question Evaluation then it is reviewed by PORC prior to approval by the Plant                ;
*
        Manager.
      D. The modification undergoes a Unreviewed Safety Question Evaluation followed                  ,
,        by a Preliminary Screening then it is reviewed by PORC prior to approval by the
*
        Plant Manager.                                                                              -
l
i
                                                                                                      l
1
I
                                                                                                      l
                            4
 
          . . .    .-      - ._        ..            -.          . . . - . ..      -.      . _ _ . . .
,
                                      -
                                                                                                            I
                                                                                                            ll
                                              STP Senior Reactor Operator Exam ... Page 92 of 100
                                                                                                              l
  QUESTION 92:                                                                                              l
  Unit 1 is in Mode 1 at 100% power when the following occurs:
  .  1300 hours - Fire reported in the Unit 1 Relay Room and fire brigade dispatched
  .  1307 hours - Fire out of control and smoke infiltrating control room atmosphere
  .  1308 hours - OPOPO4-ZO-0001, Control Room Evacuation procedure is entered                              ,
    1310 hours - Control Room actions of OPOPO4-ZO-0001 are complete
                                                                                                              '
  a
  .  1311 hours - All personnel have evacuated the Unit 1 Control Room
  .  1318 hours - ASP and all Remote Shutdown Stations are manned
  .  1325 hours - Communications established between ASP and Remote Stations
  .  1330 hours - Control of SG 1D PORV and #14 AFW Pump are transferred to
    ASP                                                                                                    '
  .  1338 hours - All Switchgear Room Alignment Addenda complete
  Based on the above timeline, which ONE of the following Emergency
  Classifications is appropriate for this event?
  A. Unusual Event
  B. Alert
  C. Site Area Emergency
                                                                                                            .
  D. General Emergency
l
l
l
                                                                                                            i
 
      .  _.        ._ _    __    ._  _ _ _ _ _ . _ .          _              .      ._    ._      _ __ .  ._
(
l-                                                                                                                l
l
l                                                      STP Senior Reactor Operator Exam ... Page 93 of 100
1
    QUESTION 93:
i
    As the Shift Supervisor acting as the Emergency Director, which ONE of the
    following Emergency Director responsibilities and authorities may you delegate?
                                                                                    ...
    A. Approving required communications with the NRC.
    B. Approving required notifications to the State and County.
    C. Approving radiological exposures in excess of 10CFR20 limits.
l
    D. Approving departure from license conditions per 10CFR50.54(x).
                                                                                                                  l
                                                                                                                  j
                                                                                                                  l
  .
                                                                                                              *
                                                                                                                ,
l
i
a
'
 
                  .    .    .    .  . . ~ . . . . . .    .        . - - .                    ... .
l
l                                                  STP Senior Reactor Operator Exam ... Page 94 of 100
      QUESTION 94:
      Given the following:
l
      .  Unit 1 is at 100% power
      a  Containment pressure is indicating +0.4 psig
      *  Preparations for performing a containment supplementary purge are in progress.
      .  Noble gas concentration inside containment is 5.5 E-04 ci/cc
      Which ONE of the following actions will be taken to prevent actuation of CVI?
      A. Ensure all RCFCs are in operation to provide mixing of the containment
          atmosphere.
      B. Increase the High alarm setpoint on the RCB Purge Exhaust radiation monitors
          (RT-8012 and RT-8013).
      C. Ensure that containment carbon units are in operation to reduce noble gas
          concentration.
1
'
      D. Increase the High alarm setpoint on Containment Atmosphere radiation monitor
l        (RT-8011).
(
!
                                                                                                      l
l
!
4
  i
    -
 
    ._ .-      ~-            -.  - .--. - -. -  -. - -..      .. _ .. -..- _ . ~ _ -... - ..._.
                                                                                        .
                                                                                                      ~. . . - . - . - - . -
                                                        STP Senior R actor Operator Exam ... Page 95 of 100
            QUESTION 95:
            The following plant conditions exist:
            *  Reactor Power is 55%                                                              .:-
.
            *  Main condenser vacuum is 26 inches Hg and decreasing
l
          ' Which ONE of the following describes the appropriate operator action per
;          OPOPO4-CR-0001, Loss of Condenser Vacuum?                                                                        j
            A. Ensure the reactor and main turbine are tripped at 22 inches Hg.                                              l
            B. Ensure the reactor and main turbine are tripped at 25 inches Hg.
!
            C. Ensure the main turbine is tripped at 22 inches Hg.
                                                                                                                              l
            D. Ensure the main turbine is tripped at 25 inches Hg.
J
I
                                                                                                                        ..
J
f
                                                                                                                              I
  n
 
                                                                                                l
                                        STP Senior Reactor Operator Exam . . Page 96 of 100
QUESTION 96:                                                                                    l
                                                                                                1
Unit 2 is operating at 100% power when all circulating water pumps trip.
                                                                                                1
Assuming no operator action, except for AFW flow control, and all systems                      i
function as designed, which ONE of the following corresponds to plant conditions                l
                                                                                                l
after the plant has stabilized?
              RCS TAVG          SG Pressures
A.              567 F            1185 psig
B.              567 F            1225 psig
C.              571 F            1185 psig
D.              571 F            1225 psig
                                                                                            _A
 
  . . .  .      .. ..      .~..  -.      . . . . . - - - . . -                _ - . . . . .    ---..      - -- -
;
4
                                                            STP Senior Reactor Operator Exam ... Page 97 of 100
                                                                                                                      l
-
        QUESTION 97:
                                                                                                                      ,
1      A pressurizer PORV has failed open causing a manual reactor trip and safety
#
        injection. The control room operators are responding per OPOP05-EO-EOOO,                                    !
;        Reactor Trip or Safety injection.                                                      ..
                                                                                                                      j
        Which ONE of the following is an indication that the PRT rupture disk has
;        ruptured?
        A. Containment pressure is increasing at 5 psi per minute.
        B. PRT temperature is decreasing.
        C. PRT level spikes to 100% then decreases to 0%.                                                            ;
        D. RCB dew point is decreasing.
                                                                                                                      :
                                                                                                                      )
                                                                                                                -
                                                                                                                      i
                                                                                                                      i
                                                                                                                      i
 
                                    .
                                          STP Senior Reactor Operator Exam ... Page 98 of 10o
  QUESTION 98:
  A High alarm is received on RT-8032, GWPS Outlet Radiation Monitor, and the
  following actions are performed by the crew:
  .  OPOPO4-RA-0001, Radiation Monitoring System Alarm Response is entered
  .  Alarm is verified to be VALID
  .  Health Physics and Chemical Analysis are notified
  Which ONE of the following radiation monitors should be checked for increasing
  readings in accordance with OPOPO4-RA-00017
  A. RT-8031, GWPS Inlet
  B. RT-8058, MAB 10' Area Monitor
  C. RT-8029, MAB Ventilation
  D. RT-8010A, Unit Vent Stack
i
 
  -      ._.._.-_..____m
                  -        .-    . . _ . _ . . _ _      . _ . . . _ . _ . . . . . _ _. __
                                                                                                _ . . , . _ . . _ _ . _ . . _ . . .,
,
'
                                                                                                                                    !
                                                                                                                                    !
l
l                                                    STP Senior Reactor Operator Exam ... Page 99 of 100                          i
    QUESTION 99:                                                                                                                    ,
(    Plant Operators are performing a transfer of the Floor Drain Tank to the Waste
    Monitor Tank per OPOP02-WL-0003, Floor Drain Tank Operation...
    Which ONE of the following is a condition that the procedure cautions could occur
l
    if procedural steps are performed improperly?
                                                                                                                                    !
l    A. Contamination of non-contaminated systems
    B. Exceeding STP airborne limits
                                                                                                                                    t
    C. Contamination of personnel
l
    D. Exceeding STP personnel dose limits
                                                                                                                    .
,
l'
,
                                                                                                                                      I
,
                                                                                                                                      ,
t
                                                                                                                                      l
                                                                                                                                      l
i-
l
l
                                                                                                                                      l
                                                                                                                                      !
!
                                -                    - - -
                                                                                            , _
 
                                                      .
                                            STP Senior Reactor Operator Exam ... Page 100 of 100
  QUESTION 100:
  OPOP02-CD-0001, Condensate System, requires that the " Condensate pump
  discharge common vent line isolation valve and the Condensate pump 13 vent at                  .
                                                                                                  '
  the miniflow check valve" be open prior to starting Condensate pump 13.
  Which ONE of the following statements is the reason for opening these two valves                I
  . prior to starting the pump?
  A. To flush out corrosion particles that collect in the low flow areas of recirculation        :
      line check valve CD-0740.                                                                  l
                                                                                                  t
  B. To extend the life of recirculation valve CD-FV-7016 by eliminating air pockets              ;
      that promote erosion.                                                                      l
                                                                                                  .
  C. To minimize the occurrence of hydraulic transients in the Condensate pump #13                ,
      recirculation line.
                                                                                                  ,
  D. To ensure that no steam binding will occur in the Condensate pump (Iow                      ,
      pressure, low temperature impelier boiling).                                              !
,
                                                                                                  ;
                                                                                                  !
                                                                                                  I
                                                                                                  .
                                                                                                  h
                                                                                                  ,
                                                                                                  h
                                                                                                  I
                                                                                                  t
                                                                                                  k
                                                                                                  i
                                                                                                  e
                                                                                                  :
}}

Latest revision as of 21:26, 10 April 2022

Forwards NRC Operator Licensing Exam Repts 50-498/98-301 & 50-499/98-301,including Completed Graded Tests Administered on 980602-04
ML20154L410
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 10/09/1998
From: Hurley L
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
50-498-98-301, 50-499-98-301, NUDOCS 9810190282
Download: ML20154L410 (104)
v • d • e


See also: IR 05000498/1998301

Text

{{#Wiki_filter:. - - - . - . _ . . . . - - . . . . . . ~ . - - . - . . , . . pr . 

     fuzu
  y              g                                  UNITED STATES
  7               g                    NUCLEAR REGULATORY COMMISSION
  O               8                                     REGION IV
                                              611 AYAN PLAZA DRIVE, suite 400
   e            [
     +,p***** g
                                               ARLINGTON, TEXAS 76011-8064

l October 9,1998 l NOTE TO: NRC Document Control Desk ! Mail Stop O-5-D-24 

         FROM:            Laura Hurley, Licensing Assistant

l Operations Branch, Region IV l SUBJECT: OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON l' JUNE 2-4,1998, AT SOUTH TEXAS PROJECT 

                           DOCKETS #60 498/499 '
                                                                                                                                                      .
         On' June 2 4,1998, Operator Licensing Examinations were administered at the

l < 

        - referenced facility. Attached you will find the following information for processing
         through NUDOCS and distribution to the NRC staff, including the NRC PDR:

, 

         item #1 - a) Facility submitted outline and the initial exam submittal for distribution

! 

                           under RIDS Code A070.

, b) As given operating examination, designated for distribution under RIDS ! 

                           Code A070.                                                                                                     .. fus, sis

l ,p n:' ! Item #2 '- - Examination Report with the as given written examination attached,

designated for distribution under RIDS Code IE42.

i i l 

          If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations

l Branch, Region IV at (817) 860-8253. l (' I l I 

;
                V
                         A       b     PDR
                                          99
                                                 (

I L \ r ( - 

         .                   -     -.   -     .     -           - . . -     . - - ~     ~-          .        . .-
                                                                                                                   !

j odf' '"- g 

               .
                                                    UNITED $T ATES
                                                                                                                   '
    f..          . '$,                NUCLEAR REGULATORY COMMISSION
           "
             *
   $                   k                              REGION IV                                                    ;

f $ 611 RY AN PLAZA ORIVE. SUITE 400 l - 

     %...a/
         ..
                                             AR LINGTON. T E XAS 76011 8064
                                                    sut i4 id                                                      !

i I L 

        William T. Cottle, President and
           Chief Executive Officer                                                                                a
      ' STP Nuclear Operating Company

l P.O. Box 289 

        Wadsworth, Texas 77483
        SUBJECT: NRC INSPECTION REPORT 50-498/98-301; 50-499/98-301

l Dear Mr. Cottle: i l From June 1-4,1998, an operator licensing certifi:ation inspection was conducted at your South 

        Texas Project Electric Generating Station, Units I and 2 reactor facilities. The enclosed report
        presents the scope and results of that inspection
        This inspection included an evaluation of six appli; ants for senior operator licenses. We

, 

        determined that five of the six applicants satisfied the requirements of 10 CFR Part 55 and the
licenses have been issued.
        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),
        Should you have any questions concerning this inspection, we will be pleased to discuss them
        with you.

l Sincerely, 

                                                              :D

! John L. Ptsliet, Chief l 

                                                                                                           /
                                                        Operations Branch                                 I
                                                                                                          '
                                                        Division of Reactor Safety                      ,
         Docket Nos.: 50-498;50-499
         License Nos.: NPF-76; NPF-80
 :

i 4 

 !
                         '
                                                                                                                   5
                           ,

               . . .  _    . _ _ _        .. _.     . . __ . . _ . . . . _ _
 ?

! i- ! 

     STP Nuclear Operating Company              -2-
                                                                              l
     Enclosure:
     NRC Inspection Report
      50-498/98-301; 50-499/98-301-
     cc w/ enclosure and Attachments 1-4:
     Ken Coates, Manager, Nuclear Training                                    !
   ' STP Nuclear Operating Company
     P.O. Box 289                                                             I
    Wadsworth, Texas 77483
   - cc w/ enclosure and Attachments 1-3:
     Lawrence E. Martin, Vice President
     Nuclear Assurance & Licensing
    STP Nuclear Operating Company
    P.O. Box 289
    Wadsworth, Texas 77483

i  ; , ' 

                                                                              1
    A. Ramirez/C. M. Canady                                                   ;
    City of Austin

' L l 

    Electric Utility Department
    721 Barton Springs Road
   . Austin, Texas 78704
    Mr. M. T. Hardt/Mr. W. C. Gunst
    City Public Service Board
    P.O. Box 1771
    San Antonio, Texas 78296
    D. G. Tees /R. L. Balcom
    Houston Lighting & Power Company
    P 0. Box 1700
    Houston, Texas 77251
    Jon C. Wood
    Matthews & Branscomb
    One Alamo Center
     106 S. St. Mary's Street, Suite 700
    San Antonio, Texas 78205-3692
    Jack R. Newman, Esq.
     Morgan, Lewis & Bockius
1800 M. Street, N.W.  ;

' 

    Washington, D.C. 20036-5869                                               !

- 

                                                                              .

! . f l l l , 

                                                                             '

> 

 .   ..    .     .     ..       . _ . . . . - - . .. -. _ - - . ... . . .._.. _ ..   _ . . _   .

i STP Nuclear Operating Company -3- 

                                                                                             .

' 

      Mr. G. E. Vaughn/Mr. C. A. Johnson                                           '
                                                                                                  ,
     Central Power & Light Company
      P.O. Box 289                                                                                ;
     Mall Code: N5012

l - Wadsworth, Texas 77483 

     INPO
     Records Center-
     700 Galleria Parkway                                                                         ,
     Atlanta, Georgia 30339A957
     Bureau of Radiation Control
     State of Texas                                                                               ,
     1100 West 49th Street
     Austin, Texas 78756
     Mr. Jim Calloway                                                                             .
   . Texas Public Utility Commission
     William B. Travis Building
     1701 North Congress Avenue
     P.O. Box 13326
     Austin, Texas 78701-3326
     John Howard, Director
     Environmental and Natural Resources Policy
   - Office of the Governor
     P.O. Box 12428
     Austin, Texas 78711
     Judge, Matagorda County
     Matagorda County Courthouse
     1700 Seventh Street
     Bay City Texas 77414

1 I I 1

l 

                                                                                                 Y

 , .                                      . .        ~      .      . ~ . -_.        -..       -                - -       . - -   . ~ . .

r . l L STP Nuclear Operating Company -4- l 

                                                                                                                               ~

l E-Mail report to T. Frye (TJF)

. E-Mail report to D. Lange (DJL)
      E-Mail report to NRR Event Tracking System (IPAS)
      E-Mail report to Document Control Desk (DOCDESK)
      bec to DCD (IE01)(IE42)

, 

      bec distrib. by RIV w/ enclosure and Attachments 1-3:
      Regional Administrator                                   Resident inspector
      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)                                   R. Bachmann, OGC (MS: 15-B-18)
      bec w/ enclosure and Attachments 1-4:
      R. Gallo (HOLB/NRR) (MS: 9D4)
      L. Hurley
                                                                                                                                           I
                                                                                                                                         i
                                                                                                                                           i
                                                                                                                                           i
      DOCUMENT NAME: R:\ stp\st8301rp.hfb                                                                                                  ;
      To receive copy of document, indicate in box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy

l RIV:SRl(:OQ / a , 

                                     RE:OB        .    N 'C:OB               d     C:PBA _ o #            C:OB            g                )

l HFBundf(llVb4 JF RELantz ' k JLPellet M JITapia %), JLPellet g  ! ' 

                           "                                                       07/[y98 l ')           07/j98 # ~
        07#l/98 \                    078/98       1
                                                              07M/98 *
            '                                                                                   V
                                                    OFFICIAL RECORD COPY
                                                                                                                                           l

! 

             .-.      .-        . __        .-    _ ._
                                      ENCLOSURE
                     U.S. NUCLEAR REGULATORY COMMISSION -
                                       REGION IV

Docket Nos.: 50-498;50-499 License Nos.: NPF-76; NPF-80 Report No.: '50-498/98-301; 50-499/98-301 Licensee: STP Nuclear Operating Company Facility: South Texas Project Electric Generating Station, Units 1 and 2 Location: FM 521 - 8 miles west of Wadsworth 

                   Wadsworth, Texas

Dates: June 1-4,1998 Inspectors: Howard F. Bundy, Chief Examiner 

                   Ryan E. Lantz, Reactor Engineer, Examiner / inspector

Approved By: John L. Pellet, Chief, Operations Branch 

                   Division of Reactor Safety

ATTACHMENTS: Attachment 1: Supplemental Information Attachment 2: Simulation Facility Report Attachment 3: Post-Examination Comments Attachment 4: Final Written Examination and Answer Key 

 [[~}                    i

     _      _    .         . _ _ _ _ _ _.            _     _    _ . _ _ _ _ _ .            __      _    __ _

,. 

                                                                                      -

1 l 

                                                       2-
                                          E'XECUTIVE SUMMARY
                      South Texas Project Electric Generating Station, Units 1 and 2
                          NRC Inspection Report 50-498/98-301; 50-499/98-301
   NRC examiners evaluated the competency of six senior operator applicants for issuance of
  operating licenses at the South Texas Project Electric Generating Station, Units 1 and 2. The
   licensee developed the initial license examinations using the guidance in NUREG-1021, Interim
   Revision 8. January 1997. NRC examiners reviewed and approved the examinations. The
  initial written examinations were administered to all six applicants on -June 30,1998, by facility
  proctors in accordance with the guidance in NUREG-1021, Interim Revision 8. The NRC
  examiners administered the operating tests June 2-4,1998.
  Ooerations
  .        Five of the six applicants for senior operator licenses passed their examinations.
           Applicants demonstrated effective oversight and good communication techniques during
           the dynamic scenarios. A common applicant performance weakness involving clearance
           order review was identified in that the majority of the applicants failed to or were slow in
           identifying that component cooling water had not been designated for isolation for
           overhaul of a residual heat removal pump (Sections 04.1 and 04.2).
  .        The examination submitted was adequate for administr$ ion and required only limited
           enhancement and editorial corrections. The senior operators assigned for examination
           validation provided valuable enhancement suggestions. The licensee staff was highly
           responsive to incorporating enhancement suggestions developed during the review
           process (Section 05.1).

! , 

                                                                                                             ,

    . . _ _ . _       m    . _ _ _ _ _ _ . _ . _ _ _ _ _ _                          - _ _      . . . _ . . _ _ _
                                                                                                                  . _ _ . . . _!

{ , L l !. ! 

                                                                   3-
                                                                                                                                 l
                                                            Report Details
                 Summarv of Plant Status -
                 Both units operated at.100 percent power for the duration of this inspection.                                   l

{ - 

                                                                                                                                l

l l. Operations l l p 04 Operator Knowledge and Performance 

                 04.1    Initial Written Examination

l ' a. Inspect:on Scone 

                         On June'30,1998, the facility licensee proctored the administration of the written

l examination, approved by the chief examiner and NRC Region IV supervision, to six l individuals who had applied for upgrading their reactor operator licenses to senior l operator licenses. The licensee proposed grading for the written examinations and l- E evaluated the results for question validity and generic weaknesses. The examiners 

                         reviewed the licensee's results.                                                                       .
                                                                                                                                j

L b.. Observations and Findings  : 

                         Five of the six applicants passed the written examination. Written examination scores
                         ranged from a low of 73 to a high of 91 percent with an average of 84 percent overall and             .!
  ~
                         a lowest passing score of percent. Greater than 50 percent of the applicants missed                     l

l 16 questions. Post-examination review by the licensee indicated that the questions- l missed were primarily due to isolated knowledge and training weaknesses. However, no 

                         broad scoped commonalities were determined from the missed question analysis. The                      l
                        . licensee's post-examination review resulted in the recommendation that the answer for                 i
. Question 57 be changed to accept both choices 'A' and 'D' as correct. The proposed

' 

                         change was accepted based on the technical merits and the examinations were
                         evaluated incorporating this change. The licensee's post examination ' comments are                     ;
                         included in Attachment 3.'                                                                             t
                  c.     Conclusions                                                                                            )

l Five of the six license applicants passed the written examinations. No broad knowledge ! or training weaknesses were identified as a result of evaluation of the graded 

                         examinations.

L I , ) t- 

                                                                                                                                I
                                                                                                                              J

                           -                          .        -     .                         _
                                               -4-

04 2 initial Ooeratina Tesj 

a.  Insoection Scoce
    The examination team administered the various parts of the operating examination to the
    six applicants on June 2-4,1998. Each applicant participated in three scenarios.
    However, each applicant was formally evaluated for senior operator license only for the
    scenario in which he participated as the control room senior operator in charge. Each
    applicant also received a control room and facilities walkthrough test, which consisted of
    five tasks with followup questions, and an administrative test, which consisted of five
    tasks in four administrative areas.
b.  Observations and Findinas
    All applicants passed all sections of the operating test. The examiners noted appropriate
    use of peer and self-checking practices in all areas of the examinations. When
    evaluated in the senior operator in charge position for the dynamic scenarios, all
    applicants demonstrated effective oversight and good communication techniques. The
    applicants displayed effective application of technical specifications and emergency and
    abnormal procedures.
    The applicants generally performed well on the systems and facility walkthrough and
    administrative tasks. However, on one administrative task the majority of the applicants
    failed to or were slow in identifying that component cooling water had not been
    designated for isolation as reviewed during review of a clearance order for overhaul of a
    residual heat removal pump.
c.  Conclusions
    All six applicants passed all parts of the operating test. Applicants demonstrated
    effective oversight and good communication techniques during the dynamic scenarios.
    The exarniners identified a common applicant performance weakness involving
    clearance order review in that the majority of the applicants failed to or were slow in
    identifying that component cooling water had not been designated for iso!ation for
    overhaul of a residual heat removal pump.

05 Operator Training and Qualification 05.1 Initial Licensino Examination Develooment 

     The facility licensee developed the initiallicensing examination in accordance with
     NUREG-1021, Interim Revision 8. " Operator Licensing Examination Standards for Power
     Reactors."

              -       -               -        - -.      .          .         -_        .

I 

                                                       5
    05.1.1 Examination Outline
 ..  a.     Insoection Scoce
           The facility licensee submitted the initial examination outline on December 17,1997.
            The chief examiner reviewed the submittal against the requirements of NUREG-1021,
            Interim Revision 8.
     b.    Observations and Findinas
           Region IV approved the initial examination outline with minor comments for
           enhancement, which were promptly resolved, and advised the licensee to proceed with
           examination development.                                                                      i
     c.    Conclusions
           The licensee submitted an examination outline in a timely manner, which required only
           minor revisions for enhancement.
    05.1.2 Examination Packs,1g                                                                          i
     a.    Insoection Scooe
           The facility licensee submitted the completed draft examination package on
           February 27,1998. Prior to formal submittal of the examination package, meetings were         I
           held in the Region IV office on January 27 and February 12,1998, to discuss
           examination development issues. These meetings were attended by licensee
           representatives J. Calvert, Operations Training Manager, and M. DeFrees, Licensed
           Operator Training Supervisor, and the chief examiner and operations branch chief. The
           chief examiner reviewed the submittal against the requirements of NUREG-1021, Interim         .
           Revision 8. An onsite review of the revised examination was conducted during the               l
           period March 20-23,1998.                                                                      !
                                                                                                          I
     b.    Observations and Findinas                                                                      !
                                                                                                         !
           The draft written examination contained 100 questions. The questions were                      l
           predominantly new for this examination. The draft examination was considered
           technically valid, to discriminate at the proper level, and responsive to the sample plan      ,
            submitted by the licensee on December 17,1997. The chief examiner provided                    l
            enhancement suggestions on 11 questions, which were appropriately incorporated by
           the licensee. The suggested enhancements generally related to clarity of the question
            stem and distractor plausibility. The chief examiner also identified an error in the ans. r
            for original Question 73, which related to the definition of core alterations. The license
            replaced this question and initiated a station condition report to investigate a potential
            discrepancy in fuel handling procedures, in addition, the licensee changed the wording
                                                                                                        ]

                                                                                            ._
                                            6-
   on several other questions in response to a generic chief examiner comment regarding
   minor grammatical errors. Finally, prior to administration, the licensee modified
  Question 37 to reflect deletion of the original reference and enhanced the distractors on
  Question 48. Minor editorial changes were made to 4 other questions.
  The operating test consisted of dynamic scenarios, administrative tasks, and system
  tasks with followup questions. The licensee submitted four dynamic scenarios, including
   one backup scenario, which was not used during the examination. The submitted
   scenarios conformed to NUREG-1021. The licensee subsequently incorporated several
  minor enhancement suggestions provided by the chief examiner and the licensee
  validation crew during the week of March 20,1998.
  To support the administrative section of the operating test the licensee submitted five
  administrative tasks. Although the submitted tasks were satisfactory test items,
  enhancements, and clarifications were necessary for each task in response to chief
  examiner questions and comments. Also, during the onsite review it was discovered the
  answer for Job Performance Measure A3, " Review a Tagout," was not complete and
  further changes were necessary, in addition, the onsite validation time for Job
  Performance Measure A3 was in excess of one hour, which required a reduction in task
  scope. Enhancements were made to otherjob performance measures during the onsite
  review.
  To support the systems walkthrough section of the operating test, the licensee submitted
  five system job performance measures with two followup questions associated with each.
  The chief examiner provided only one editorial comment on the job performance
  measures. However, several enhancements were made to the job performance
  measures during the onsite review. Also, it was necessary to modify the cues to clarify
  the scope of the task for Job Performance Measure 3," Place a Class 1E 125V DC
  Battery Charger in Service."
  Many of the enhancements made to the operating test during the onsite review were
  directly attributable to comments and suggestions by the validation crew, which was
  comprised of senior operator members of an off-duty shift crew.

c. Conclusions 

  The examination submitted was adequate for administration and required only limited
   enhancement and editorial corrections. The meetings with licensee training
   representatives in the Region IV office were instrumental in achieving a satisfactory
   examination submittal. The senior operators assigned for examination validation
   provided valuable enhancement suggestions. The licensee staff was highly responsive
   to incorporating enhancement suggestions developed during the review process.

   - _. _ . - _ - _ _ . . . _ . _ . - . _ _ _ . _ _ . _ . _ _ , _ _                              - . . _    .- _ __ ._.
                                                                                                                           :
                                                                                                                          ,

.J 

                                                                      7-
                                                                                                                          )
           05.2. Simulation Facility Performance -
            a.          Insoection Scone

                                                                    .
                                                                                                                          ;
                      - The examiners observed simulator performance with regard to fidelity during the                   ,
                        examination validation and administration.                                                        t

. 

            b;          Observations and Findinas                                                                         :
                        The simulation facility supported the validation and administration of the examination
                        well. However, as discussed in Attachment 2, there were general failures on two
                        occasions for one scenario during the onsite review and an identical failure during                !
                        examination administration. The failures had minimalimpact on the review effort.

i However, the failure during examination administration required use of the backup , 

                        scenario for one crew. No other fidelity problems were observed.
                                                                                                                          l

, 

                                                                                                                          :
            c.          Conclusions

i The simulator and simulator staff supported the examinations well. General simulator j. ' failures had minimal impact on examination review and slight impact on examination 

 '
                        administration.                                                                                   l
 !
                                                                                                                          ;
          05.3          Examination Security                                                                              ;

'  ! 

            a.          Scooe
                                                                                                                          I
  • The examiners reviewed examination security both during on site preparation week and l

, 

                      - examination administration week for compliance with NUREG-1021 requirements.                      :
                                                                                                                          i
                                                                                                                          '
             b. .       Observations and Findinas

. i

*

' 

                        Members of the licensee's operations and training staff signed onto the NUREG-1021
                                                                                                                          '

, 

                        approved examination security agreement acknowledging their responsibilities for
                        examination security. The licensee maintained secure areas for examination
development, review, validation, and administration. Signs were conspicuously posted to

i 

                        avoid inadvertent unauthorized access, and doors were maintained locked with good key
                        control to ensure proper access to sensitive areas. Applicants were maintained under              ,

l constant supervision and were always escorted to and from examination points. 

                        Simulator security was strictly complied with.

, 

            c.          Conclusiorts

j Effective examination security was maintained. . 

 4

_ . 

                                                  -.
                                                                                                                        a

 . _  . . _ .    _          _    .-. _ _ _ _      _ _ . . _ . . _ _ . . ~ . . - _ . _ . _ ..     . . .
                                                -8-
                                       V. Management Meetings
   X1  Exit Meeting Summary
       The chief examiner presented the preliminary inspection results to members of the
       licensee management at the conclusitan of the operating test administration on June 4,
        1998. Final results of the inspection were presented by Mr. John Pellet during a
       telephone conference with Mr. Mike DeFrees on July 8,1998. The licensee
       acknowledged the findings presented.
                                                                                                         ,
                                                                                                         .
       The licensee did not identify as proprietary any information or materials examined during
       the inspection.

, 

                                                                                                           l
                                                                                                           l
                                                                                                           l
                                                                                                           l
                                                                                                       ,

- - .- . -. -- - --. . - - . . 

                                                                                            . - - . .
                                                                                                      4
                                            ATTACHMENT 1
                                                                                                        '
                                   SUPPLEMENTAL INFORMATION
                                                                                                        '
                              PARTIAL LIST OF PERSONS CONTACTED
 - Licensee
   J. Calvert, Manager, Operations Training
  G. Chitwood, Examination Lead
   K. Coates, Manager, Nuclear Training
   M. DeFrees, Lead Instructor, Licensed Operator Training
  W. Dowdy, Unit 2 Operations Manager
  J. Lovell, Manager, Generation Support
  K. Struble, instructor, Licensed Operator Training                                                     l
  NEC
  G. Guerra, Resident inspector
                                                                                                         i
                                  INSPECTION PROCEDURE USED                                               l
  NUREG-1021              " Operator Licensing Examination Standards for Power Reactors,"
                          Interim Revision 8

  - --      -    -.     ..           .   _ .        . . _ - - - - . . - - - - -                 - - - . - -     .
                                                                                                                 !
                                                                                                                !
                                                                                                                 ,
                                                                                                                 i
                                                   ATTACHMENT 2                                                  -
                                           SIMULATION FACILITY REPORT
        Facility Licensee:     STP Nuclear Operating Company .
        Facility Docket-       50-498; 50-499
        Operating Examinations Administered at: STP Electric Generating Station, Units 1 and 2
        Operating Examinations Administered on: June 2-4,1998 -                                                   r
       These observations do not constitute audit or inspection findings and are not, without further
       verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations
       do not affect NRC certification or approval of the simulation facility, other than to provide
       information, which may be used in future evaluations. No license action is required in response

i 

       to these observations.
       Deficiency identified Durina Examination Preoaration
       .
                During scenario validation, RCFC 11 A outlet air temperature indicator was reading
                150' F when actual temperature was 62* F. Replaced sticking meter to correct
                erroneous reading.
       .
                The F.B. exhaust filter train isolation dampers took an unrealistically long time to close
                during realignment of F.B. VAC while performing POP 05-EO-EOOO, i.e.,13 to
                17 seconds versus 7 seconds expected in the plant. The stroke time was adjusted in
                the simulator.
       .
                During a LOCA the simulator stopped responding and the audible simulator trouble alarm
                actuated on two occasions. On the third attempt, the simulator performed as expected.
       Deficiency identified Durina Examination Administration
       .
                Despite the simulator staff's unsuccessful attempts to reproduce the simulator failure
                several times prior to examination administration, it recurred at the identical point in the
                LOCA scenario during applicant evaluation.

l 1 l l: l l l ! l 

                           .     - .                .                           .        .. .-              - -

                                ATTACHMENT 3
                           SOUTH TEXAS PROJECT
                           WRITTEN EXAM (6/30/98)
                    POST-EXAM APPLICANT COMMENTS

QUESTION 57 ANSWER 'D' REFERENCES: OPOP05-EO-F002, Core Cooling Critical Safety Function Status Tree 

               OPOP05-EO-FRC3, Response to Saturated Core Cooling, Step 2

COMMENT: Use of High Head injection (Answer 'S') is also an ' appropriate method' to 

               use under the specified circumstances.

RESOLUTION: Comment accepted. Accept choices 'A' and 'D'. Choice 'A' was originally 

              thought to be incorrect based on not being applicable unless RCS
              pressure was below 1745 psig. The question stem specified an RCS
              pressure of 1785 psig. However, OPOP05-EO-FRC3, Response to
                                     .
              Saturated Core Cooling, Step 2b, has the operators establish HHSI flow
              regardless of w? ether pressure is above or below 1745 psig. The
              importance of being above or below 1745 psig is related to whether
              charging flow should be established, not HHSI flow.
                                                                        ,/
                                                                      /

                 e             A   A 4 a w - - a e. . 4 a

l 1 l 

                   ATTACHMENT 4

l 

   FINAL WRITTEN EXAMINATION AND ANSWER KEY

l l l 

                                                          >

< > . 

    ,   -                 -.                    -           . -            . . - . -     -
                                                                                                  _ , _ - - . .       .,
                                                                                                                                                                ,
                                                                                                                                   ,
                                                                                                    *                 ~'                                '
                                                                                                                                 ll'     ;                [ f.

i . .

- ,
                                                                                                                                            .
                                                                                                                            *                                        !
                                                                                                                                                  .t'
                                                                                                                                                                      ,

l 

                                                          STP SRO EXAM KEY                                                                      :
                                                                                                                                                        sc ..:l
                                                                                                                                                          . :. .

I ' 

                                                                                                                                                                     ;
                                                                                                                                                  4-

                                                                                                                                              .
                                                                                                                                                                     l
            1)'      A                                                35) D                                                    69)       B                             !
            2)       C                                                36) A                                                    70)       B                            l

l- 3) D 37) A ' 

                                                                                                                               71 ) -.   A'                           !
            4)       B                                                38) A                                               .    72) B                                  l'
            5)      .D                                                39) C                                                    73)' D'
            6)       A                                                40) D                                                    74) B
            7)       B                                                41) A                                                    75) -C                                 l

! 

            8)       C                                                42) B                                                    76) A
            9)       D                                                43) C                                                    77) C                                 j
            10) C                                                     44) D                                                    78) C                                  i
            11) A                                                     45) C                                                    79) B
            12) D                                                     46) C                                                    80) A                                  !
            13)-B                                                     47) B.                                                   81) B                                  !
            14) C                                                     48) B                                                    82) D                               'i
            15) B                                                     49) D                                                    83) C                                   l

l- 16) A 50) C '84) B j 

            17) D                                                     51) B                                                    85)       C                         '{
            18) D                                                     52) D                                                    86)       B                             ;

. 

            19) A                                                     53) C                                                    87)       C                            l
            20) A                                                     54) C                                                    88)       C
                                                                                                                                                      *
                                                                                                                                                                      l
            21) A                                                     55) B                                                    89)       C                            !
           .22) D                                                     56) B                                                    90)       D                            i
            23) A                                                     57) D or A                A pta r                        91)       g                           j
            24) D                                                     58) B                                                    92)       C                          '!
            25) B                                                     59) A                                                    93)       A.                  -
                                                                                                                                                                     {
            26) D                                                     60) D                                                    94)       B                             l
            27) A                                                     61) B                                                    95)       A                             ]

l 28) A 62) B 96) D j l 29) B 63) D 97) B i 

            30) C                                                     64) C                                                    98)       D                             i
          '31) B                                                      65) A                                                    99)       A

(' '32) A 66) D 100) C 1 

            33) A                                                     67) D                                                                                            !
            34) C                                                     68) D

i j - - - - . . . . . . . . . . - - . . - - - - - - - . . . . . . . . . . . . . - - - . . . . . . . . . . . . . . - - 

           ANSWER USAGE BY LETTER:                                   A) 25
                                                                      B) 26
- C) 24
                                                                      D) 25                                                                                             l
;
                                                                                                                                                                       I

. l 1 

  -
                                              >

\ ' 

      _       .           . . . .                             -    -       _ _ .- .             .                                                          - . . .

  _    ,                     _ _ _
                                                                                                     l
  ES-401                        Site-Sptcific Written Examination                  Form ES-401-7
                                            Cover Sheet
                                                                                                     I
                                                                                                     1
                                                                                                     i
                   U.S. Nuclear Regulatory Commission
                                        Site-Specific '
                              Written Examination                                                    l
                                                                                                     l
                                    ' Applicant information.
  Name:                                            Region:                          IV-              ,
                                                                                                     1
  Date:                     6/30/98-               Facility / Unit:       South Texas Project        !
  License Level:'                  SRO'            Reactor Type:                    W'               I
  Start Time:                                      Finish Time:
                                                                                                    )
                                           Instructions
  Use the answer sheets provided to document your answers. Staple this cover
     ~
                                                                                                    -
  sheet on top of the answer sheets. The passing grade requires a final grade of c
 f.t least 80.00 percent. Examination papers will be collected five hours after the
  examination starts.
                                     Applicant Certification
  All work done on this examination is my own. I have neither given nor received
  aid.
                                                       Applicant's Signature
                                             Results
  Examination Value                                                                       Points
 ' Applicant's Score                                                                      Points

l Applicant's Grade Percent 

  NUREG-1021                                                      Interim Rev. 8, January 1997

L ./ 

      .        .           m. _ _ _ __ _   m             _ _ .
                                                                 .
                                                                                     .        . . . . _ ~ _ _
                                                                                                              .

. l l 

                                             STP Senior Reactor Operator Exam ... Page 1 of 100               j

l l QUESTION 1: i l ' 

 A fire'has been detected in the Post Accident Sampling System (PASS) Room. The                               i
 security card reader controlling access to the Fuel Handling Building is inoperable,
 preventing entry.
                                                                                                              :
 Which ONE of the following individuals may be issued a Vital Area Security Key                               l
 from the Controlled Key Locker to access this area in accordance with

! OPOP01-ZA-0019, Operation Key Control? 

 A. Fire Brigade Leader                                        ,
 B. Fire Protection Coordinator
 C. Health Physics Supervisor
 D. Chemical Analysis Supervisor                                                                              :

                                                                                                              I
                                                                                                              !
                                                                                                              l

l t ! l 

                      .

     .    - .         . - . . .     ~ . - . . . .  . - - . . .          . - - . . - - . . . - _ - . .            - . . - - . . . . - . - . .       -
  '

l' STP Senior Reactor Operator Exam ... Page 2 of 100 

                QUESTION 2:

                Consider the following plant conditions:
                .  Plant was initially at 100% power
                .  RCS leak through a weld crack on a pressurizer level detector lower tap                                                           l

l occurred ! 

                .  A plant shutdown has been initiated
                Based on this information, the NRC must be notified within                                 followed by a                             !
              ' 30 day report (LER).                                                                                                                 !
                A.15 minutes
                B. 30 minutes                                                                                                                        l
                C.1 hour
                                                                                                                                                     !
                D. 4 hours
       '
        -                                                                                                                                      -
                                                                                                                                                     l
                                                                                                                                                 ~

! l [ 

   '

! I. , i . 

                                       .                         --                                   -
                                                                                                        ..            .-                     .

    .    .        ._            __         _ _ -_. _ . _ _         _ _ .   - - -        _ _ . . _  _ _.. . ._.
  -
                                                       STP Senior Reactor Operator Exam ... Page 3 of loo
       QUESTION 3                                                                                              .
                                                                                                               l

l 

       Given the following:

1 ' 

       .    Both Units are at full power with one hour left until shift turnover.
       *    Current shift manning is as follows,:

i 

                                   UNIT 1                                        UNIT 2
           SS                         1                                            1                           i
           US               2 (1 is Field Sup.)                 2 (1 is Field Sup, and STA)

i RO 2 2 (Extra RO left due to emergency) 

           PO                        4                                             6

! 

                                                                                                               ,
                                                                                                               i
       .    The Unit 1 Unit Supervisor, Field Supervisor and Reactor Operator are injured                      l

, 

            during a midshift plant tour and are taken to first aid for treatment and                          i

! observation 

       Per Technical Specifications, which ONE of the following actions must be taken if
       this condition exists until shift turnover?

! i l A. Unit 1 minimum shift crew composition is met and no further action is needed. , 1 l B. Unit 1 minimum shift crew composition is met provided the Unit 2 STA l assumes the Unit 1 Unit Supervisor position.  ; 

                                                                                                               <

i ! l C. Unit 1 minimum shift crew composition is NOT met and the unit must enter l T.S. 3.0.3 if the proper crew composition is not met within 2 hours. 1 1 ' 

      'D. Unit 1 minimum shift crew composition is NOT met and no further action is
            required other than the continual efforts to restore crew composition.

1 - i i

1 

         , - .   .     . . - . . - - -.      . . . . . .-...- . .   . ..         . . . . .       . . . . . - . . . . . _ . . . . . - . .

! l 

                                              STP Senior Reactor Operator Exam ... Page 4 of.100
     QUESTION 4:
     Unit 1 is operating at 100% power when a valid High Alarm is received on
     RT-8013, RCB Purge Exhaust Monitor.
     Which ONE of the following radiation monitors is now. inoperable?.
     A. RCB High Range Ra'd iation Monitor (RT-8051)                                                                                     )
    ' B. RCB Atmosphere Radiation Monitor (RT-8011)
     C. Unit Vent Stack Monitor (RT-8010A)
     D. GWPS Outlet Radiation Monitor (RT-8032)

.. 

  6                                                                                                              --
                                                                                                                                         ;
                                                                                                                                         !

! 

                                                                                                                                         !
                                                                                                                                         ;

i 

      .        . _ . .                                            - ~ _ _ - .. - - . . - . , _ .

4 

                                          STP Senior Reactor Operator Exam ... Page 5 of 100
 QUESTION 5:
 The unit is at 2% power during a plant startup following a refueling outage.
 Which ONE of the following conditions requires entry into a Technical Specification
 action statement?

. 

 A. Core Physics testing has determined that the Axial Flux Difference is outside
    the target hand specified in the COLR.
 B. Primary containment internal pressure is at -0.05 psig during Supplemei.tary

' 

     Purge operation.

, C. "A" Safety injection Accumulator nitrogen pressure is 660 psig. i 

 D. Intermediate Range Nuclear instrument Channel N35 has experienced a loss of
instrument power.
                                                                                             !
                                                                                             )

 _       . . .   _ . - _ _ _ . . _ . _ . _ _ . . .

J l STP Senior Reactor Operator Exam ... Page 6 of 100 

      QUESTION 6:
      On a loss of al; main feedwater at power, an AFW actuation may be initiated by
      the AMSAC system.
     - During a power decrease, which ONE of the following signals causes the C20
     - AMSAC BLOCKED permissive status light to illuminate?
               '
      A. < 30% Turbine Power

'

. B. '< 30% Reactor Power , 

      C. < 4d% Turbine Power
      D. < 40% Reactor Power

i . 

   .                                                                                                        .

! 5 1 

                                                   . . .

  -- . .          - = - . -   - .   - - _ . . - .. . . .   .- . . . . . . .          -. .- . .. .   . --_. -. -. -.-. -

l i- I STP Senior Reactor Operator Exam ... Page 7 of 100 l l. ' QUESTION 7: ! 

         ~ Unit 1 is operating at 100% power with all systems in their normal lineup. A small
          RCS leak develops in the Chemical Volume and Control System (CVCS) with the

l' fo!!owing Control Room indications: l * Pressurizer level indicates 50% and'is decreasing j + RCS pressure indicates 2220 psig and is stable i a ' Containment pressure is normal l * OUTLET PRESSURE PI-0135 indicates O psig 

          *   PRT PRESS Hi alarm is illuminated

i. * LETDN HX OUTL FLOW Hl/LO alarm is illuminated. 

          *1  ERFDADS Points (2) CVCS LTDN AREA TEMP TRN Hi are in alarm

i- Based on these indications, which ONE of the following AUTOMATIC actions has 

          taken place?
                            ~

l; A. TCV-0143, Letdown Temperature Divert Valve, positioned to the VCT r ' 

          B. MOV-0023/0024, Letdown Line Containment Isolction Valves, closed
          C. FV-0011, Letdown Header Orifice Isolation Valve, closed
          D. PCV-0135, Letdown Pressure Control Valve, opened

, i l t f 

                                  .
                                                         y                  m-,-.rr-                        m

                       _   __  .    _ _ _ . .   . - _ .   . _ . _ - . -     . . . _ . . _ _ _ _ _ __        _. _
                                                                                                                    . .,
                                              .                                                                           I
                                                        STP Senior Reactor Operator Exam ... Page 8 of 100

l I l QUESTION 8: 

      Given the following:

i l * Pressurizer variable heaters are energized l l . Pressurizer backup heaters are in AUTO control l 

      =  RCS pressure is at 2000 psig
      Assuming no operator action, if pressurizer level decreases to 14%, then is

! restored to 25%, which ONE of the following describes the condition of the 

     _ pressurizer heaters?                                                                                      ,
      A. Backup and variable heaters will be off.                                                                         4
                                                                                                                          i
      B. Backup and variable heaters will be on.

l C. Backup heaters will be on and variable heaters will be off. ' 

                                                                                                                          i
      D. Backup heaters will be off and variable heaters will be on,                                                      j
                                                                                                                          :
                                                                                                                          l

i 

   .                                                                                                       *

l l 

                                                                                                                   .

I ! l , 4 

                                                                                                                        .
 i                                                                                                                        .

   - . . . .                 - -       -~ _-. - - - . _ . .             - -      -   - .          ...    .. ,  .- -

l STP Senior Reactor Operator Exam ... Page 9 of 100 

             QUESTION 9:
             Unit 2 is preparing to fill the reactor cavity for refueling. A few minutes after LHSI

l pump 2A is started, RHR PUMP 2A DISCH FLOW LO and RHR PUMP 2A TRIP i alarms are received. l 

             Which ONE of the following is the reason for the trip of the operating RHR pump?
             A. RHR relief valve has opened causing discharge flow to decrease to < 925 gpm
                for > 5 seconds.
             B. Start of LHSI pump 2A has caused RHR pump 2A mini flow recirc valve to

l close and discharge flow has decreased to < 925 gpm for > 10 seconds. 

             C. RHR inlet pressure has increased to > 350 psig causing the inlet isolation
                valves to close reducing discharge flow to < 925 gpm for > 10 seconds.
             D. Start of LHSI pump 2A has caused RHR pump 2A to become dead headed and
                discharge flow has decreased to < 925 gpm for > 5 seconds.
                                                                                                                    !

. i ( , 

 '

    , . . .         -    ...  . . ~ . ~ . ~ .. - . .- - - _ .. . - . . . ~ _                  -. . - .    . ~ , . .- . . . --. - . ~ . . .

, P [ STP Senior Reactor Operator Exam ... Page 10 of 100 ! ( QUESTION 10: 

               When in Mid-Loop Operations, whicii ONE of the following set of conditions
               indicate that an RHR Pump is AIR BOUND?

,. 

               A. Pump Current - HIGH

f 

                  _ Pump Flow'- LOW

l ' 

               B. Pump Current - HIGH
                   Pump Flow - HIGH

l C. Pump Current - LOW !- Pump Flow - LOW 

               D. Pump Current - FLUCTUATING
                  Pump Flow - FLUCTUATING

! t l , 

                                                                                                                                           4
                                                                                                                                                  4.

i 

                                                                                                                    o

i 

                                                                                                                                             '

\ ! l- . i' 

                                                                                                                                                     I

i i 

  ~

l 

            .-
                                                                                   .        .                                                  ., ,

  ...            ..         -    . - - . - - - _ _ -              -   . - . . - . . -                   - - . -   .- __
                                                                    '
                                                                                                                        ,
                                                     STP Senior Reactor Operator Exam ... Page 11 of 100

f- 

       QUESTION 11:-
       Unit 1 refueling is in progress. The refueling machine operator is inserting a fuel

j assembly into the In-Containment storage area. l Which ONE of the following cautions should be observed while lowering the fuel l ' assembly into the In-Containment storage racks? t 3' A. Use a slow speed while entering the storage racks until the bottom of the fuel [ 

            assembly has entered.                                                                                       i
       B. Monitor the weight indicator continuously for an unexpected decrease of 50
            pounds which would require operators to stop lowering the fuel assembly.
       C. Use the Gripper Mast Down (red) light to determine the assembly is fully
          .
           lowered and can be disengaged.                                                                               ,
       D. Monitor the Z-axis tape continuously to ensure the mast is not lowered below
          the mast disengagement position.
                                                                                                                        .
                                                                                                                         <
                                                                                                                         i
                                                        , . _ , ,                     _ . . . . _ . . _         _

    _ _ _ _ . . .          .   . . . . - _ _     ._ _ ._. . .         ._     _._.__.......m.               ._ . - =._ _       _

. 

                                                                    STP Senior Reactor Operator Exam ... Page 12 of 100
         QUESTION 12:
         Given the following:
         *        . Unit 1 is in Mode 6.
         .         50 fuel assemblies have been reloaded after a complete core off-load.
         *         RHR Train A is in service maintaining RCS temperature 90'F.
         4 RHR Train A loop boron concentration at last sample: 2950 ppm.
         . . Source Range NI channe! 31 indicates 10 cps and is selected for audible count -
'
                   rate indication.
         . ~ Source Range NI channel 32 indicates 5 cps.
         *         Both Extended Range NI detectors are operable
         in accordance with OPOP08-FH-0009, Core Refueling, which ONE of the following
         conditions requires suspension of Core Alte- 7tions and an investigation by the Core
         Loading Supervisor prior to resuming Core , r cerations?
         A. Source Range NI channel 31 increases to 40 cps.
         B. Updated log reading on RHR Train A inlet temperature indicates 98 F.
  +
         C. Source Range NI channel 32 suffers a loss of instrument power.                                              *
                                                                                                                                ,
         D. Updated RHR Train A boron concentrations on two consecutive samples
                   indicate 2920 ppm.
                                                                                                                            .
                                                                                                                          ~

1 $ _ $ 

  m                      -                   _ -              _ _ _                          _            -. -

                                       -     - . . . . . _ _ . . . . - - . . ..   . . . . . . . . . . .-
                                                                                                         l
                                          STP Senior Reactor Operator Exam ... Page 13 of 100            ;

QUESTION 13 

                                                                                                         i

Given the following:

  • ' Unit 1 has just experienced a Loss of All AC Power
  • OPOPO5-EO-ECOO is in progress
  • An RCS leak inside containment causes RCB pressure to rise to 11.0 psig.
  • Efforts are underway to restore power to at least one 4160 VAC ESF bus. '

. The STA reports that an ORANGE Path condition exists on Containment 

   Integrity Critical Safety Function (CSF) Status Tree.                                                 4

Which ONE of the following describes the action (s) that should be taken? A. OPOP05-EO-FRZ1 should be entered to address the loss of containment 

   integrity safety function.

B. OPOP05-EO-FRZ1 should NOT be entered because it is assumed that at least 

   one ESF bus will be energized during any Functional Restoration Procedure.
                                                                                                         "

C. OPOP05-EO-FRZ1 should be entered only when containment pressure rises to 

   indicate a RED path on the Containment integrity CSF.

D. OPOP05-EO-FRZ1 should NOT be entered because the loss of power makes the  ! 

   critical safety function status monitoring inaccurate.
                                                                                                         ;
                                                                                                         I
                                                                                                         l

  ..          .  . ~ _ . =                                  .        . . . . - . - . - - .
                                                                                           . . - _ . . - . - - . .       . - . .

i . 

                                                                                                                                 l
                                                STP Senior Reactor Operator Exam ... Page 14 of 100                              t
                                                                                                                                 '
                                                                                                                                 '
       QUESTION 14:
       OPOPO5-EO-EC00, Loss of All AC Power contains the immediate action step:
       " Verify Turbine Trip."                                                                                                   ,
                                                                                                                                 .
       Which ONE of'the following statements describes the BASIS for this step?
       A. Prevent an inadvertent Safety injection
       B. Prevent an inadvertent Main Steam Line Isolation
       C. Prevent an uncontrolled cooldown of the RCS
       D. Prevent steam generator inventory depletion                                                                             >
                                                                                                                                  .
                                                                                                                                  !
                                                                                                                                   ,
     .                                                                                                               .
                                                                                                                                 )
                                                                                                                                  ,
                                                                                                                                  f
                                                                                                                                  i
                                                                                                                                  ?
                                                                                                                       %

l- l ! 

                                                                                                                          -

.. 

                                                                                                                r  4

  . .    . -. .          .._    _ __           _.        _ . . _ . - __                        . ..

I 

                                                                        STP Senior Reactor Operator Exam ... Page 15 of 100     q
                   QUESTION 15:-                                                                                                ;

1 

                   Given the following:                                                                                         j
                                                                                                                                i

, 

               -
                   .   .The reactor has tripped from 100% power operation                                                       ;

L. . Three (3) Control Rods did not fully insert l ! * Reactor power is at 1E-6% and decreasing . i ! . Boration is in progress at 25 gpm i . - Reactor Trip Response (OPOP05-EO-ES01) is~in progress  ! 

                                                                                                                                !

l 

                 ' The Suberiticality ' Critical Safety Function (CSF):

i .. 

               - A. is satisfied and boration can be secured.
                                                                                                                                i
                   B. is satisfied, but boration must continue until shutdown margin is established.                            ;

l 

                   C. is NOT satisfied because boration must be greater than 30 gpm per Tech

l. ' Specs. 

                                                                                                                                i

j ' D.' is NOT satisfied because more than one (1) control rod is not fuliy inserted. 

                                                                                                                                l

l 1 l 

                                                                                                                                l
                                                                                                                                l

l l . 

                                                                                                                                !

.. I. !. I 

                                                                                                                                 I
    4          m                     n-. 4                                                                                  - ,

                                          STP Senior Reactor Operator Exam ... Page 16 of 100
 QUESTION 16:
 OPOPO5-EO-FRS1, Response to Nuclear Power Generation - ATWS, Step 4.g.
 directs the operator to check pressurizer pressure less than 2335 psig.
 Which ONE of the following is the basis for this step?
 A. Verify RCS pressure is not high enough to reduce charging injection into the
     RCS.
 B. Verify RCS pressure will not challenge a pressurizer PORV.
 C. Verify RCS pressure has stabilized fellowing the initial heatup and insurge into              ;
     the pressurizer.
 D. Verify pressurizer PORVs are functioning to limit the RCS pressure increase.                  ;

. . 

                                                                                                  !
                                                                                                  .
                                                                                                A
                                                                                                   l
                                                                                                  ,
                                                       . .                   .

 .     _    _-       __    _-   .  . . . ~ _ _ _ _   _.  ._.m._-._        _ _     _ - ____ _  -_m   . _ . _ . . _ .__ _ _ _

, ~ 

                                                                                                                              J

j STP Senior Reactor Operato. 8Exem ... Page 17 of 100 , 

     QUESTION 17:
     A condition has occurred which warrants entry into OPOPO4-CV-0003, Emergency
Boration. While attempting to borate using the Normal Emergency Boration
     flowpath, both boric acid pumps fail to start.

     Which ONE of the following PAIRS of Emergency Boretion flowpaths are used in                                             l
                                                                                                                              l
     this condition?
     A. Through Alternate Boration isolation MOV-0218A OR

I Emergency Boration via GravitV Feed ! 8. Through Alternate Boration Isolation MOV-0218A OR 

         Through Manual Alternate immediate Borate 1(2)-CV-0221

s 

     C. Through Manual Alternate immediate Borate 1(2)-CV-0221 OR
         Emergency Borate from the RWST                                                                                       ;
     D. Emergency Boration via Gravity Feed OR
         Emergency Borate from the RWST                                                                                       ;
                                                                                                                              :
                                                                                                                              i
                                                                                                                              l
                                                                                                                              l
                                                                                                                              l
                                                                                                                              l
                                                                                                                              )
   5
                                                                                                                            -
                                                                -__ - - -

. -. - - - = . . . . . _ . . _ . .. . . . -. . - - . 

                                                                                                              l
                                                                                                              I
                                                  STP Senior Reactor Operator Exam ... Page 18 of 100
                                                                                                              1
   QUESTION 18:
                                                                                                              {
   Given the following:
   .  Unit 1 is operating at 90% power, steady-state, following a power increase                              ;
                                                                                                              '
   .
      Chemical Analysis reports that RCS samples indicate increasing RCS activity
   .  RT-8039, CVCS Letdown Failed Fuel Monitor is NOT in alarm                                               !
   .
      Decontamination Factor (DF) across the inservice mixed bed demineralizer is
      adequate
   .  CVCS letdown flow is 140 gpm
   Which of the following actions, directed by Chemical Analysis, are in accordance
   with OPOPO4-RC-0001, High Rer.ctor Coolant System Activity?
   A. Maintain present letdown flowrate and place the alternate mixed bed
      demineralizer in service.
   B. Maintain present letdown flowrate and place the cation demineralizers in
      service.
   C. Maximize letdown flow and place the alternate mixed bed demineralizer in
 -
      service.                                                                                        -
   D. Maximize letdown flow and place the cation demineralizers in service.
                                                                                                          .
                                                                                                            }

                                                              . .            ...     .      . .-- --
                                      9
                                          STP Senior Reactor Operator Exam ... Page 19 of 100
 QUESTION 19:
                                                                                                        I
 Given the following:                                                                                   j
 .  Unit 2 is in Mode 4
 .  Gaseous Waste discharge from the VCT is in progress
 a  GWPS Inlet Header valve (FV-4657) and Discharge flow valve (FV-4671)
    handswitches are in the NORMAL position at the control panel
 .  A high alarm condition exists on RT-8032, GWPS Outlet
 Which ONE of the following is the final position of the GWPS Inlet Header
 (FV-4657) and Discharge Flow (FV-4671) valves?
          Inlet Header Valve (FV-4657)      Diacharge Flow Valve (FV-4671)                              I
   A.                CLOSED                               CLOSED
   S.                CLOSED                                 OPEN
   C.                 OPEN                                CLOSED
   D.                 OPEN                                  OPEN                                        I
                                                                                                        l
                                                                                                        l
                                                                                                        I
                                                                                                        ;
                                                                                                        i
                                                                                                      .

- J 

. . _ . . . _ _ _.               . ._ .     _ . . . . _ - . _ . _ . . . _ _ . _ . __                _ _ _ . _ . _ _ _ . . . . _ . . _ _ . _ . - . . _
                                                                                                                                        _
                                                                                                                                                                  - . . . . .
                                                                           _.
                                                                                                                                                                                l
                                                                                                                                                                               l
                                                                                                                                                                                '

l l STP Senior Reactor Operator Exam ... Page 20 of 100 j 

                    QUESTION 20:                                                                                                                                                I
                   - A Small Break LOCA has occurred, primary plant conditions have been stabilized                                                                            !
                    and Sl has been terminated in accordance with OPOP05-EO-ES11, Si Termination.                                                                              ;
                    Operators have just restored MINIMUM charging flow and continue performing                                                                                  !
                    ES11 when the following conditions are identified:                                                                                                          l
                                                                                                                                                                                l
                    *    RCS subcooling - 45*F and stable

j * Pressurizer level - 5% and stable ~ l' * Adverse Containment Conditions do NOT exist , 

                                                                                                                                                                               !
                    Which ONE of the following actions will the operators perform?
                                                                                                                                                                               :
                                                                                                                                                                                l
                    A. Increase charging flow to restore pressurizer level and continue with
                         OPOP05-EO-ES11, Si Termination.                                                                                                                       i

1 . 

                                                                                                                                                                               \

l B. Operate Si pumps and continue with OPOP05-EO-ES11, Si Termination. 1 

                                                                                                                                                                               J
                                                                                                                                                                               i

i l C. Operate Si pumps and transition to OPOP05-EO-E010, Loss of Reactor or l l= Secondary Coolant. I 

                                                                                                                                                                               !

(. l 

                    D. Manually initiate Si and transition to OPOP05-EO-EOOO, Reactor Trip or Safety                                                            -
                                                                                                                                                                               L

l Injection. 

                                                                                                                                                                               i
                                                                                                                                                                               )
                                                                                                                                                                                1
                                                                                                                                                                        -      :

l i, c 9 

                                                                                                                                                                                !

i. 

                                                                                                                                                                               .
                                                                                                                                                      - . , , .

   . .. .      _     _ . _ . .    _ ~ . -   .    .m   _-     _ - . . _ _ _ . . _ . . _ m . . _ . _ . . _ . .. ._. . . _ .- ..
                                                                                                                               ,
                                                                                                                                i
                                                                                                                               i

, 

                                                    STP Senior Reactor Operator Exam ... Page 21 of 100
                                                                                                                                l
        QUESTION 21:
                   ~
                                                                                                                                i

, 

        Given the following-
                                                                                                                               ,
                                                                                                                               '
        .   Unit 1 was operating at steady-state 100% power
        a   Reactor Trip and Safety injection actuated from low pressurizer pressure
        .
            RCS pressure decreased rapidly to 1150 psig and is now increasing very slowly                                       '
        .. Pressurizer level decreased to 8% initially and then increased rapidly to stabilize
            at 100%
        *   Reactor Containment Building process radiation monitors are alarming                                               ;
                                                                                                                               I
        Which ONE of the following Loss of Primary Coolant events is the cause for these                                        I
        indications?
      ' A. Pressurizer spray line rupture
        B. Pressurizer surge line rupture
        C. Reactor head vent line rupture
        D. Incore lower guide tube rupture                                                                                     l
                                                                                                                                l
                                                                                                                                l
                                                                                                                                1
                                                                                                                                I
                                                                                                                               i
 w        -                    ..              _
                                                           .
                                                                                                                              l

  _.   . _ . - . ~ _ _ . . . . _ ...._..... _ ... _ _ _ _ _.- _ _ _ _ _ _                               m. _ _ _ . . - _ . . . _     . . . _ . _ < .

.  ? , 

;                                                                         . STP Senior Reactor Operator Exam ... Page 22 of 100

' 

          QUESTION 22:
i A Large Break LOCA occurred 3 hours ago and the following plant conditions
          presently exist:                                                                                                                           ,

' 

          *       RCS pressure is 350 psig
,
          *       Containment pressure is 8 psig                                   _
                                                                                                                                                     '

l * Containment Hydrogen concentration is 0.4%- i * SG NR levels are: 1 A - 32%; 1B - 40%; 1C - 28%; 1D - 12% 

          =       Total Auxiliary Feedwater flow is 200 gpm
          *       Spent Fuel Pool temperature is'145 F

, 

        ' Which 'ONE of the following actions is REQUIRED to be performed in accordance
          with OPOPO5-EO-E010, Loss of Reactor or Secondary Coolant?                                                                              -*
          A.~ Stop LHSI pumps and place them in auto

3 

          B. Place Hydrogen Recombiner(s) in service                                                                                                 ;

i- . C. Increase total AFW flow to greater than 576 gpm i , 

     -
          D. Place Spent Fuel Pool Cooling in service                                                                              -
:
                                                                                                                                         &
                                                                                                                                                      l
                                                                                                                                                      1
                                                                                                                                                      I
                                                                                                                                                      l
                                                                                                                                                      i
                                                                                                                                                      1
                                                                                                                                 -                  >

      .       .         --    . . . ~ = . . - -    ~ .       . .    .- - -

, . i 

                                                     STP Senior Reactor Operator Exam ... Page 23 of 100
         QUESTION 23:

) 

         Given the following:
             Unit 2 is at 50% power
                      -
         *

,

e Rod control is in manual
         a.~ T REF/AUCT T AVG DEV annunciator is in alarm

             Rod control Temperature error meter (TI-0612) indicates + 10 degrees
                                                                           -
         *

         a-  P-13 Channel 1 bistable light is extinguished
         *   P-13 Channel 2 bistable light is illuminated
      ,
         Which ONE of the following instrument failures has occurred?
         A. Pressure Transmitter 505 has failed low

l 

         B. T-hot instrument 413 has failed high

. 

  '
         C. Pressure Transmitter 506 has failed low

.l i . D. T-cold instrument 412 has failed high

- . ? 

    .

   .. . - - . . .         ~ .. ..-     . . - . - . . - . - . _ . . -             . . - . . _ _ - . . . _ -

i , 

                                                                                                                                    j
                                                                                                                                    ,
                                                                                                                                    I
                                                                     STP Senior Reactor Operator Exam ... Page 24 of 1o0
            QUESTION 24:                                                                                                            !
            Given the following:                                                                                                    .
            .     Unit 1 is at 100% power                                                                                           l
            .     VCT LEVEL Hl/LO alarm is received                                                                                 l
            *     VCT LEVEL LO-LO alarm is received                                                                                 i

, 

            .     VCT level (on CP-OG4) is indicating 85%

l . VCT pressure (on CP-OO4) is indicating 20 psig 

            .     LCV'112A, Divert valve to RHT is indicating intermediate
                                                         '
                                                                                                                                    l

l Which ONE of the following instrument failures is the cause of the plant . I conditions? l L A. VCT Level Transmitter 112 has failed high l 

            B. VCT Level Transmitter 113 has failed high

l C. VCT Level Transmitter 112 has failed low l l- D. VCT Level Transmitter 113 has failed low 

                                                                                                                         4

. 

                                                                                                                                    '
                                                                                                                           .

! . 

 l

I

i ! ! l l 

                   ,   .
                                                                           - , ,                           -                 . . . '

   .   - - - - -    -          . - .
                                              STP Senior Reactor operator Exam ... Page 25 of 10o
     QUESTION 25:
     In OPOPO5-EO-FRC1, Response to inadequate Core Cooling which ONE of the
     following methods is attempted FIRST, based on it being the most effective
     approach to recover the core and restore core cooling?
     A. Restart of a Reactor Coolant Pump
     B. Reinitiation of High Head Safety injection
     C. Rapid depressurization of intact Steam Generators
     D. Opening Pressurizer PORVs and Reactor Head Vent Valves

l i . l . { 

                                                                                                  /

 .    .;.-            . ~,,    . . - . . . . - . .   . . . . . - - . .   _ . . . . -  . - . - - . - . . - - . -  . . . . .     . .
                                                   -
                                                                                                                                   i
                                                                                                                                   !
                                                             STP Senior Reactor Operator Exam ... Page 26 of 100                   )
                                                                                                                                   !
   QUESTION 26:
   Step 1 of OPOP05-EO-FRP1, Response to imminent Pressurized Thermal Shock
   Condition, has the operator check that LHSI flow is greater than 500 gpm if RCS                                                 i
   pressure is less than 415 psig.                                                                                                 1
   This step is based on:
   A. ensuring adequate mixing in the cold leg downcomer region during natural
         circulation conditions,                                                                                                   j
                                                                                                                                   ;
                                                                                                                                   i
   B. preventing core exit temperatures from exceeding the required temperature to
         place RHR in service.
   C. ensuring adequate Low Head Safety injection cooling prior to Accumulator
         isolation.
   D. preventing implementation of Pressurized Thermal Shock (PTS) actions if a
         Large Break LOCA has occurred.
                                                                                                                                   l
                                                                                                                           .
                                                                                                                                   i
                                                                                                                                    l
                                                                                                                             .

. 

  '
                                               STP Senior Reactor Operator Exam ... Page 17 of 100
      QUESTION 27:

' 

      A reactor trip with a loss of ALL feedwater has occurred. The crew has entered
      OPOP05-EO-FRH1, Response to Loss of Secondary Heat Sink and RCS Bleed and

l 

      Feed Criteria have been met. Attempts to establish RCS Bleed and Feed produce
      the following conditions:
      .  One (1) HHSI Ptmp - RUNNING
      =
         One (1) Pressurizer PpRV - OPEN
      =  Two (2) Reactor Vessel Head Vent paths - OPEN
      Based on these conditions, determine the adequacy of the RCS Bleed and Feed.
      A. Only the RCS Bleed path is NOT adequate

i 

      B. Only the RCS Feed path is NOT adequate
      C. Both RCS Bleed and Feed paths are NOT adequate

l D. Both RCS Bleed and Feed paths are adequate , ! ! l l !' h 4 

    .

   . . . . . . . . . . _ _ _ . . _ - . . . . . . _ . . . . _ . - _ , _ . . - _
                               _.                                              - . _ . _ . _ _ . . .       . . _   _ . _ _ _ . . - . . . . _ . _ . . - _ . . _ _ . . . _ _ . - _ .
                                                                                                                 !

, 

                                                                                                     STP Senior Reactor Operator Exam ... Page 28 of 100
                                  QUESTION 28:

. 

                                  Unit 1 has experienced a reactor trip due to an RCS leak inside containment.
                                  .   .0POPO5-EO-EO10, Loss of Reactor or Secondary Coolant, is in progress.                                                                           I
                                  e   . Containment pressure is 12 psig decreasing at 0.25 psig per minute.
                                  *     RCS temperature is 570*F
                                  e     SG pressures are: 1 A - 1230 psig; 18 - 1225 psig; 1C - 1350 psig; 1D - 1150
                                        psig
                                  a     SG NR levels are: 1 A - 45%; 1B - 28%; 1C - 75%; 1D - 34%.
                                  .     OPOP05-EO-FRH2, Response to Steam Generator Overpressure, has been
                                        entered.
                                  *     An operator has been dispatched to the 1C IVC to investigate.
                                  For the existing plant conditions, the Unit Supervisor should:
                                  A. NOT direct the operator to locally open the 1C SG PORV due to the potential of
                                        water hammer in the steam piping.
                                  B. NOT direct the operator to locally open the 1C SG PORV due to the excessive                                                                       i
                                        differential pressure across the valve.                                                                                                    -
                                                                                                                                                                                        .
                                  C. Direct the operator to open the 1C SG PORV to decrease pressure below the
                                        SG PORV setpoint.                                                                                                                              ;
                                  D. Direct the operator to assist maintenance in manually lifting a 1C steam safety                                                                 '
                                        valve to decrease pressure.
                                                                                                                                                                                        !
                                                                                                                                                                                        l
                                                                                                                                                                                       1
                                                                                                                                                                                        i
                                                                                                                                                                                        ;
                           4
 .

    .- .-.               . - . .    . . . - --        - _ . -   -         _ . - _ . - . . ~ . . . . - . - - . . ~ _ . . . _ _ _ _

l- 

                                                                                                                                  I
                                                   STP Senior Reactor Operator Exam ... Page 29 of 100                            ,
                                                                                                                                  ;

i 

                                                                                                                                  !

' 

           O.UESTION 29:
           In OPOPO5-EO-ES12, Post LOCA Cooldown and Depressurization, an Si flow
           reduction sequence is performed to minimize leakage from the RCS while still                                           i
           maintaining adequate coolant inventory. An RCS Subcooling Table, which takes
           into account current pump status, is used to determine when the HHSI pumps are                                         '
           stopped.
i
           Which of the following pump's status are accounted for in this RCS Subcooling
           Table?
           A. RCPs, HHSI Pumps and LHSI Pumps
           B. CCPs, HHS! Pumps and RCPs
           C. CCPs, HHSI Pumps and LHSI Pumps
           D. CCPs, LHSI Pumps and RCPs                                                                                            j
                                                                                                                                   .
                                                                                                                                   l

l l i . I 4 i 

  '

( , 

 . _
         -    .   . _ - .-                 .   -        .--    .    . .         ...          .         ..   .
                                                                                                                i
                                                   STP Senior Reactor Operator Exam ... Page 30 of 100          ;
                                                                                                                '
           QUESTION 30:
           The following plant conditions exist:
     .
           *  Small Break LOCA has occurred                                /
                                                                        *j'
           .  Adverse Containment conditions exist                                                              I
           a  OPOP05-EO-EO10, Loss of Reactor or Secondary Coolant, is in progress
           *  Operators are checking whether RCS cooldown and depressurization is required                       )
                                                                                                                 1
           Which ONE of the following criteria warrants transition to OPOP05-EO-ES12, Post
           LOCA Cooldown and Depressurization?
           A. Pressurizer Levelless than 8%
           B. RCS Subcooling less than 35*F
           C. RCS Pressure greater than 415 psig
           D. Two Hot Leg Temperatures greater than 420 F
       .
                                                                                                          -

l l 

                                                                                                              *

! . 5 l l l

, 

    .,  __ _.              . . _ _ . . _
                               _
                                           . _ ~ . .  . . _ . _ - . - _ . - __ _ _ . . _ . - _ _ _ . _ _ . . . . _ _ _ _ _ . . _ _ . .             _ . . . _ . , _ _ .
                                                                                                                                                    _
              $
                                                                       ' STP Senior Reactor Operator Exam ... Page 31 of loo

! QUESTION 31: 

        Given the following:                                                                                                                                             j

<  ! 

        *       Unit 1 is at 25% power and stable
        e       SG 1 A level is indicating 83% NR
        e       SG 1B level is indicating 78% NR
       ..       SG 1C level is indicating 89% NR
        *       SG 1D level is indicating 80% NR
        e       All systems are in their normal alignment for the current power level
        Which ONE of the following lists the automatic actions that should directly result                                                                               i
        from the above conditions?                                                                                                                                     '!
  .
        A. Reactor trip, turbine trip, feedwater isolation bypass valves close and low
                power feed regulating valves close
        B. Turbine trip, operating feed pumps trip, feedwater isolation valves'close and                                                                                 i
                feedwater regulating valves close                                                                                                                        l
                                                                                                                                                                         1
        C. Reactor trip, operating feed pumps trip, feedwater isolation valves close and                                                                                 j
'
                low power feed regulating valves close

l D.' Turbine trip, operating feed pumps trip, feedwater isolation bypass valves close 

                and feedwater regulating valves close

l ! 

                                         .
                                            - , - . -             y                ~ --.                                               . . - . . ,

. _ . _ _ _ - _ . . . . _ _ _ . _ . . _ . . _ - . _ _ . _ . _ . . ~ -.- - _ _ _ _ . _ _ _ . _ . _ _ _ . . _ . ... 

                                                                                                                                     \
                                                                                                                                     l
                                                                                                                                     .
                                                             STP Senior Reactor Operator Exam ... Page 32 of 100
       QUESTION 32:                                                                                                                  ;
                                                                                                                                     l
                                                                                                                                     '
       During a unit shutdown, all 4 pressurizer pressure channels slowly decrease and
       stabilize at approximately 1960 psig activating P-11. The operator manually blocks
       the low pressure Si signal. Over the next 15 minutes, the pressure channels have
       the following trends:                                                                                                         l
                                                                                                                                     1
       *  PT-455 - Slowly increases to 1984 psig, then lowers to 1860 psig                                                           '
       *  PT-456 - Slowly increases to 1987 psig, then lowers to 1856 psig                                                           ,
                                                                                                                                     '
       *  PT-457 - Slowly increases to 1986 psig, then lowers to 1858 psig
       *  PT-458 - Slowly increases to 1983 psig, then lowers to 1855 psig
       Assuming all bistables trip EXACTLY on their setpoint, which ONE of tha following                                             ;
       describes what will happen during the pressure transient?
       A. P-11 will reset and a safety injection will occur
       B.   P-11 will NOT reset and a safety injection will NOT occur
                                                                                                                                     l
                                                                                                                                     !
       C. P-11 will reset but a safety injection will NOT occur
                                                                                                                                     !
     - D. P-11 will NOT reset but a safety injection will occur                                                              -
                                                                                                                                     ;
                                                                                                                                     I
                                                                                                                                     !
                                                                                                                               -     !
                                                                                                                                     l
                                                                                                                                     1
                                                                                                                                     l
                                                                                                                                     !
                                                                                                                                  .

              ..
                                       STP Senior Reactor Operator Exam ... Page 33 of 100
                                                                                            ,

QUESTION 33: 

                                                                                           I

Which ONE of the following alarm conditions is indicative of a simultaneous Digital I' Rod Position Indication (DRPI) Data A AND Data B failure? LEGEND: General Warning (GW), Rod Bottom (RB), Urgent (Urg), Trouble (Troub), Rod Supervisory Monitor Rod Position Trouble Alarm (Rod Supv Mntr Rod Pos Troub), Flashing (FLASH) and illuminated (LIT) 

                                                                                           >
      DRPI       DRPI      DRPI      RPI (05M3)      DRPI           Rod Supv
      GW         RB        Urgent    Trouble         Non-Urg        Mntr Rod Pos
      Light      Light     Alarm     Alarm           Alarm         Troub Alarm
A.    FLASH      LIT       LIT       LIT             OFF            LIT
 B.   FLASH      FLASH     LIT       OFF             LIT            LIT
 C.   LIT        FLASH     LIT       LIT             LIT            OFF
 D.   LIT        LIT       OFF       LIT             LIT            LIT
                                                                                           l
                                                                                           )
                                                                                           l
                                                                                           1
                                                                                           !
                                                                                           !

            .                                            . - . -  .     .      _               ..   .
                                                                                                      ,

l 

                                           STP Senior Reactor operator Exam ... Page 34 of 100
 QUESTION 34:
 Given the following conditions:
 .  Unit 1 is at 75% power
 .  Rod Control is in MANUAL
 *  Turbine Feedback Control is in IMP IN
 *  PR NI Channel 41 indicates 5% less than other PR NI Channels

! . "D" Control Bank Rod D-12 DRPI Rod Bottom Light Lit 

 .  The following Control Room Annunciators are illuminated:
    *   SM03/F-4, ROD BOTTOM
     *
        SM03/D-5, ROD SUPV MNTR ROD POSITION TRBL                                                      i
    *   SM02/A-6, TREF /AUCT T AVG DEV
    *   SM03/C-3, PR CHANNEL DEV
                                                                                                       l
 Assuming no operator actions, which ONE of the following correctly shows how
 the final steady state plant conditions will compare to the steady state plant
 conditions that existed prior to the event?

, 

           Average Core Power           Generator MW                      RCS Tavg
                                                                                            -
                                                                                                  ^
   A.             Lower                       Same                         Lower
   B.             Higher                     Lower                         Same
   C.             Same                        Same                         Lower
                                                                                                    '
   D.             Same                       Lower                         Same

l

I 

              .

 ._ .      . _ . . . . = . _ , . . . _ _ .__ _  _ _ . . _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ . _    _-  _ . _ . _ _ _ _
                                               STP Senior Reactor Operator Exam ... Page 35 of loo

QUESTION 35: The following plant conditions exist: t

  • Unit 1 is operating at 90% power

. I&C has determined that TWO control rods are untrippable Which ONE of the following operator actions is required per OPOPO4-RS-0001, Control Rod Malfunction? A. Maintain affected banks within 12 steps of the untrippable rods B. Borate the RCS until Cold Shutdown boron concentration is achieved C. Shutdown the reactor to Mode 3 within six hours D. Trip the reactor and initiate emergency boration 

                                                                                                                    i
                                                                                                                    l
                                                                                                                   1
                                                                                                                    l
                              .

  . _ . . . .          .       . . -  - . . _ . . - . .-- . ,..-.- - .-.~ . -                     - . .  _ .     _ . . . . . . .

, 

                                                             STP Senior Reactor operator Exam ... Page 36 of 100
              - QUESTION 36:
               Given the following:
               .  Unit 1 is in Mode 1 at 75% power
               *  Rod control is in AUTO
               e  Control Bank D at 235 steps
               *  Main turbine control in IMP OUT
               *  Control rods continuously withdraw for no apparent reason
               =  Operators place rod controlin MANUAL and rod motion stops
               *  Operators determine that Control Bank "D" rods have withdrawn a total of 10
                  steps

l -Which ONE of the following conditions would exist? l l 

               A. Axial Flux Difference (AFD) would become more positive /less negative.
               B. Overpower Delta Temperature trip (OPDT) setpoint would increase.
-
               C. Nuclear Enthalpy Rise Hot Channel Factor (HCF) would decrease.

1 L D. Quadrant Power Tilt Ratio (OPTR) would increase. r ! , ! i l l l , i 

  . - . . _        . _ _ . . . _ _ . _   _ . _ _ . . . . _ _ _ _ _ _ _ _ _ _ _ _                        _ , _ _ . . - . _ _ _ _ . _ _ - . . _ _ . . _ _ _ . .
                                                                                 STP Senior Reactor Operator Exam ... Page 37 of 100
             QUESTION 37:                                                                                   g
             Unit 1 unidentified leakage has increased to 0.5 gpm and containment entry is
. being planned to identify the source of the increased leakage. . ,

l l 'In accordance with OPSP03-XC-0002A, Partial Containment inspection, which ONE l, of the following is a requirement associated with the containment entry and close- ' 

             out inspection?
             A. The Shift Supervisor shall resolve any question that arises concerning what

, 

                 constitutes loose debris and/or material.

l 

              B. A Reactor Operator shall be the Test Coordinator for the containment entry and

l close-out inspection. !- 

              C. A Health Physics Technician shall perform Form 1, Containment Pre-Entry

l Briefing. I D. The Health Physics Supervisor shall be the only point of contact for determining 

                 what tools may be left in the RCR,
                                                                                                                                                              .

( l 

                                                                                                                                                                ~

l l l l l ! 

i

.' 

                                       .
                                                                                                                             Y

  - .   . . - . .      .     .   - - . -
                                          - . . - . . . ~ . - - - - . .             . - ,_. ~ _ -              - - . -.. - - - .. -.
                                                -

. , 

                                                                                                                                     I
                                                                                                                                     r

                                                                    STP Senior Reactor Operator Exam ... Page 38 of 100
                                                                                                                                     f

. 

      QUESTION 38:

      Given the following:
                                                                                                                                     '
,

t * During a midcycle reactor startup the Reactor Containment Building (RCB)  : L CNTMT PRESS Hl/LO alarm annunciates with RCB pressure at +0.3 psig. l 

      =     T.S. 3.6.1.4 for Containment Systems Internal Pressure is entered.                                                       l

! + It is deterniined that a Supplementary Containment Purge will be performed to 

             lower the RCG pressure.                                                                                                 <

l = Four RCFC Units are in operation 

      *     The US notes that the Reactor' Containment Purge Release Permit has expired.                                             4
                                                                                                                                     )
      Which ONE of the following describes the action (s) to be taken to lower RCB
      pressure?
      A. The purge may be started provided Chemical Analysis concurs and a new purge
             request is initiated.
      B. The purge may be started provided Chemical Analysis concurs and a continuous                                                !
             sample is performed during the purge.
      C. The purge may NOT be started until a valid purge permit is completed and
             delivered to the control room.
      D. The purge may NOT be started until a valid purge permit is completed and
             additional RCFCs are started,
                                                                                                                                       i
                                                                                                                                       !
                                                                                                                                      j

  - . . . .  -     .   .    . . . ~ . . . . . . . . . - . . . . - - - . - . . . - . . . _ - . . . . . . - ~ - . - . . . . ~ . -   -   .
                                                                                                                                       i
                                                                                                                                      I
                                                                                                                                      '
                                                                         STP Senior Reactor Operator Exam ... Page 39 of 100
              QUESTION 39:

. 

              The following plant conditions exist:
                                                                                                                                      4
                                                                                                                                      '

' ' 

               *  Small break LOCA occurred 12 hours ago                                                    .
               *  Containment pressure is 3.7 psig                                                                                    ,
               *  Containment temperature is 215"F
. * OPOP05-EO-FRZ3, Response to High Containment Radiation Level, is entered
              Which ONE of the following actions is required in accordance with FRZ37
.
A. Place the Containment Carbon Filter Units in service.  !

! 

               B. Ensure all Reactor Containment Fan Coolers are running.

) C. Ensure Containment Ventilation isolation has occurred. ) ' 

                                                                                                                                       i
;              D. Sample containment atmosphere activity using the PASS.                                                               1

1 

                                                                                                                                       1
I
                                                                                                                                .
                                                                                                                                    A
                                                                                                                                       !
                                                                                                                                       l
                                                                                                                                       l
                                                                                                                                       l
                                                                                                                                      ,

. . - ._. . . __._ . . . _ .. - . 

                                       STP Senior Reactor Operator Exam ... Page 40 of 100
 QUESTION 40:                                                                                  !
                                                                                               l
 Given the following:
 *   Unit 1 is at 100% power                                                                   !
 .   ECW and CCW pumps 1 A are running
 a   ECW and CCW pump 1B control switch is in AUTO and their ECW/CCW                           '
     selector switch is in standby
 .   ECW and CCW pump 1C are not running and their ECW/CCW selector switch is
     in OFF
 A ground on 4.16 KV bus E1 A causes a bus It ": ut. The standby ECW/CCW train
 will start when the:
 A.1 A ECW and CCW pump breakers open                                                          i
 B. CCW common header discharge pressure decreases to 80 psig
 C.1 A ECW pump discharge valve begins closing
 D. ECW trains 1 A and 1C discharge pressure decreases to 20 psig

    . - . - . . ~ - .        . , . .    . . - .  . ~ . .   . . - . -  .     . ~ . . . _ . . . . . . . . . - - . . . _ _ , . . . -

' 

                                                                                                                                    1
                                                                                                                                    i

l 4 

                                                         STP Senior Reactor Operator Exam ... Page 41 of 100                        I
QUESTION 41
                                                                                                                                    1
                                                                                                                                    1
. Given the following:  !

4 

         a    ' Unit 1 is in Mode 5                                                                                                 l

1 e RHR train 1 A is in service maintaining RCS temperature 130 F  ! 

         *-     CCP 1 A is in service -                                                                                              I

! * CCW trains 1 A and 1C are in service l 

         *      Low pressure letdown supplied from RHR train 1 A
         Tiie control room receives the CCW SURGE TK LVL LO and CCW SRG TK

b COMPART A LVL LO alarms. Surge tank levelin the train A compartment is

currently 48% and decreasing.

1

'Which ONE of the following describes the Control Room response to existing plant l

' I 

         conditions?

- 

                                                                                                                                     l

l A. Trip CCW pump 1 A, place RHR train 1C in service, and secure RHR train 1 A. ) 

                                                                                                                                    '

i

B. Ensure the CCW Surge Tank Reactor Makeup Water Storage Tank makeup
                valve is open, monitor CCW surge tank level and trip CCW pump 1 A if level
  .
                decreases to < 38E                                                                                           .      !
       - C. Trip CCW pump 1 A, place CCP 1B in service and secure CCP 1 A.
         D. Ensure the train 1 A CCW to RHR heat exchanger inlet valve is open, place RHR
                train 1C in service, and align low pressure letdown to RHR train 1C.                                              -
                                                                                                                                    :
                                     ,
                                                                                                                                    i

          . ~ - -       . - . . _ .    =   . . . . = . . . . . _ . .     -              . ~ . . . . . . -
                                           STP Senior Reactor Operator Exam ... Page 42 of 100
   QUESTION 42:
   A plant shutdown to Hot Standby is in progress with reactor power well below the
   POAH. As control bank rods are being inserted, the following conditions occur:
   .  Permissive P-6 clears at the expected setpoint
   .  Both Source Range Nls fail to automatically re-energize
   .  Reactor power is decreased to 5.0 E-11 amps on the Intermediate Range and
      the Primary RO is directed to manually energize the Source Range Nis.
   Which ONE of the following Source Range NI readings is expected at this reactor
   power level?
   A. > 5.0 E4 and s 5.0 E5

't 

   B. > 5.0 E3 and s 5.0 E4
   C. > 5.0 E2 and s 5.0 E3
   D. > 5.0 E1 and s 5.0 E2

 ,
                                                                                                          ,
                                                     w

                      -.   . .    - . .  . .       .     - _ = . - . _ _ . _ _ ._ . . _    - __   .
                                                                                                      l
                                                                                                       l
                                             STP Senior Reactor Operator Exam ... Page 43 of 100
 QUESTION 43:                                                                                         1
 Unit 1 is performing a reactor startup after a refueling outage of 25 days. The
 ICRR has continuously predicted criticality within the + or - 500 pcm band.
 During the last 50 step pull the reactor was determined to be critipal at 60 steps
 on control bank C (Below LO-LO Insertion Limit).
 One of the required actions is to                    to ensure that                            .
 A. commence emergency boration; acceptable temperature limits are maintained.
 B. trip the reactor; reactivity transients associated with accidents are within
    acceptable limits.
 C. commence emergency boration; shutdown margin limits are maintained.
 D. trip the reactor; heat flux hot channel factor remains within limits.                              ,
                                                                                                       l
                                                                                                      ;

. 

                                                                                                    A
                                                                                                      ,
                         4

    ..-. .      _ . ~ . . ~  , . -    . . . . . - . .   ~ . _ . . - _ - - -        - . - . _ - - _        , - - . . - .. . -
                                                      -
                                                                                                                               .
                                                                                                                               !
                                                                                                                               l
                                                                                                                               .
                                                           STP Senior Reactor Operator Exam ... Page 44 of 100                 :

t. 

           QUESTION 44:                                                                                                        !
           A major steam line break inside containment has caused a reactor trip, safety
           injection and containment spray actuation.. The Control Room operators have just                                   ,
           completed the immediate actions of OPOP05-EO-EOOO, Reactor Trip or Safety                                           I
           Injection, when the secondary operator notes that 1D steam generator level is                                       !
           increasing in an uncontrolled manner.                                                                               l

l 

           In regards to the AFW pumps, which ONE of the following statements describes                                        :
           the correct operator response to this event?                                                                         j
                                                                                                                               ;

! A. Verify all Motor Driven AFW pumps are running then secure #14 AFW pump.  ! 

                                                                                                                                i

,. 

           B. Continue to feed steam generator 1D to a level of > 14% then secure #14                                           l
                                                                                                                                '

l AFW pump. l l C. Verify at least two Motor Driven AFW pumps are running then secure #14 AFW 'l l pump. l l D. Continue to feed steam generator 1D to a level of > 34% then secure #14 

               AFW pump.

i

I 

                                                                                                                               ;

j. , , l l l l I i. l l . ( 

  '

l 

  ..~ ,    . , _ .   . . .        .. .           .. .     . - . . .          .-   - . ~        . . . . - -

! l 

                                                                                                             i

l- i 

                                                   STP Senior Reactor Operator Exam ... Page 45 of 100

I  !

- QUESTION 45

t. , , Assuming all plant systems function as designed, which ONE of the following  ! ' 

         Critical Safety Functions would have the HIGHEST priority condition during a -
         postulated " Uncontrolled Depressurization of ALL Steam Generators?"                                ;
        - A. Core Cooling (C)
         B. Heat Sink (H)
        .C. Integrity (P)                                                                                    ;
                                                                                                             .
         D. Inventory (1)                                                                                    .
                                                                                                             '
                                                                                                             ,

, , 

                                                                                                           . .
                                                                                                             !
                                                                                                             5
                                                                                                             .I
                                                                                                              1
                                                                                                               l
                                                                                                               l
                                                                                                               l
                                                                                                               1

l-  ! l I f l 

                                                                                                               1

  ~ ~ . .-.      ..           .- . - - - _. - . - . .                    . - . . ~ _         . - . . . . - . . .-

4 4 i ! 4 

                                                      STP Senior Reactor Operator Exam ... Page 46 of 100
                                                                                                                  ,
        QUESTION 46:
        Given the following:
                                                                                                                  '

        *   A Reactor Startup is being performed following a midcycle outage.
        *   Reactor Power has been stabilized at 1E-8 amps                                                        !
.
        *   RCS temperature is at the no-load value

l . Critical data has been taken. 

        *   Prior to any additional control rod movement, a single SG Safety Valve on SG                          ;

i 

            1D fails opens and remains open.

j * RCS Tavg decreases 9"F and reactor power starts to increase. 

        Which ONE of the following states the correct action, required to satisfy Technical                       ;

! 

        Specification LCO(s)?

j A. Reduce power range high flux high trip setpoint to 87% rated thermal power. . 

                                                                                                                  '

! , 

        B. Restore SG safety valve to operable status prior to entering Mode 1.                                   ;
                                                                                                                  '

1 

        C. Restore RCS Tavg within 15 minutes or be in Mode 3 within the next 15

, 

            minutes.
,

l D. Immediately initiate boration to restore Shutdown Margin.

! i 

                  .
                                                                                                                  l
,

N n  ! 4 4 

                                                                                                                   i

 . ._. .m     . _ _ .      -  _ . _ . - . ._ -  _ ___.    ...._ ._. ._       _ _ _ _ _ _ _         _ . . _ _ _ . . _ _
                                                STP Senior Reactor Operator Exam ... Page 47 of 100
         QUESTION 47:
         Unit 1 is operating at 85% power when the Plant Computer AFD monitor alarm
         actuates. Which ONE of the following actions will comply with Technical
         Specification requirements? -                                         .
         A. Within 30 minutes restore AFD to within limits.
         B. Within 30 minutes reduce power to s 50% and reduce Power Range neutron
            setpoint to s 55% within 4 hours.
         C. Within 15 minutes restore AFD to within limits.
         D. Within 15 minutes reduce power at least 3% from rated thermal power for ever
            1% that AFD exceeds the limit and similarly reduce Power Range neutron
            setpoint within 4 hours.

. 

                                                                                                                       s
                                                                                                                         !
                                                                                                                         :
                                                                                                                         l
                                                                                                                         l
                                                                                                                         !
                               ..                               _                          __

                                       STP Senior Reactor Operator Exam ... Page 48 of 100

QUESTION 48: , While performing a control rod operability surveillance at 100% power, one control bank "D" rod is found to be immovable due to mechanical interference. Which ONE of the following actions shall be taken within one hour? A. Determine the position of the immovable rod using the movable incore 

  detectors.

B. Determine that the shutdown margin requirement is satisfied. C. Reduce power to less than 90% of rated thermal power. D. Reposition all Control Bank "D" rods to within 12 steps of the misaligned rod. 

                                         STP Senior Reactor Operator Exam ... Page 49 of 10o
 QUESTION 49:
 Which ONE of the following is an interlock associated with supplying power to a
 4.16 KV ESF bus?
 A. Diesel Generator output breaker cannot be closed manually un!'ess the
    emergency bus 1L (2L) feedor breaker supplying the respective ESF transformer
    is open.
 B. Diesel Generator output breaker will not close automatically unless diesel
    generator speed is 2 500 rpm and diesel generator voltage is 2 3744 volts.
 C. Standby bus 13.8 KV ESF transformer feeder breaker is interlocked with its
    disconnect switch such that the disconnect must be open before the feeder
    breaker can be opened.
 D. Standby bus 13.8 KV ESF transformer feeder breaker is interlocked with the
     13.8 KV Emergency bus ESF transformer feeder breaker to prevent both of
    them being closed at the same time.

. . 

                                                                                               4

                                 .
                                        STP Senior Reactor Operator Exam ... Page 50 of 100

QUESTION 50: On a loss of 125 VDC Class 1E power to the Emergency Diesel Generator, which ONE of the following components loses power? A. Control Power for Test Mode B. Local Control Panel Indicating Lights C. Voltage Regulator Field Flash Circuit D. Standby Fuel Oil Booster Pump 

                                                                                              l
                                                                                              1
                                                                                              !
                                                                                              l
                                                                                              ,
                                                                                            -l
                                                                                              l
                                                                                              l
                                                                                              ,

  .   . . - . . .           . . - . . . . . .. ._    . . . - , . . -      . . ~ . . . _ - . .    . . ~ - . . - -         - _     .

! l 

                                                                                                                                   t
                                                                     STP Senior Reactor Operator Exam . . Page 51 of 1o0
                                                                                                                                   ,
                    QUESTION 51:
                    Unit 1 is operating at full power, steady-state conditions. A periodic load test is

i being performed on ESF Diesel Generator #13 in accordance with 

                    OPSP03-DG-0003. ESF Diesel Generator #13 has been paralleled..with 4.16 KV
                    Bus E1C and is carrying 2,000 KW of real load. A Feedwater line break inside
                    containment occurs and containment pressure increases to 15 psig.
                    Which ONE of the following describes the response of the "C" ESF load sequencer
                    and ESF Diesel Generator #137                                                                                  ,
                    A. The Diesel Generator remains paralleled to the 4.16 KV bus and the Mode l

l timer sequences equipment onto the 4.16 KV bus. 

                                                                                                                                   ,
                    B. The Diesel Generator output breaker is tripped open and the Mode i timer
sequences equipment onto the 4.16 KV bus.
                    C. The normal feeder breaker AND the Diesel Generator output breaker to 4.16 KV

L bus E1C are tripped open and the Mode Ill timer sequences equipment onto the , 

                       Diesel Generator.

1 

            -
                    D. The normal feeder breaker to 4.16 KV bus E1C is tripped open and the Mode II                          -
                       timer sequences equipment onto the Diesel Generator.

! 

                                                                                                                               ,

L l l I . ! .' l ! 

    ,             .

  . .- .          .- - . . . . -  .    .    . . -    .          - . . ~ _ ~ - - . - . -    ...-        _ . . -  _ _ ~
                                                                                                                      l
       .
                                                            STP Senior Reactor Operator Exam ... Page 52 of 100
          QUESTION 52:
         . A steam generator tube rupture has occurred on Steam Generator (SG) 18. The
          ruptured SG has been isolated in accordance with OPOP05-EO-EO30, Steam
          Generator Tube Rupture, and RCS cooldown and depressurization are in progress.
          SG 1B pressure is currently 910 psig and normal containment parameters exist.
          Which ONE of the following conditions warrant stopping the RCS depressurization?
          A. RCS Pressure is 900 psig
              PRZR Level is 4%
              Subcooling is 70*F
          B. RCS Pressure is 915 psig
              PRZR Level is 32%                                                                                       !
              Subcooling is 55*F                                                                                      !
          C. RCS Pre .ure s *.35 psig
              PRZR Lt. el is 5's%
              Subcookn.) ie *O*F
           D. RCS Pressure is 950 psig
              PRZR Level is 78%                                                                                       ,
              Subcooling is 85 F                                                                                      l
                                                                                                                      l
                                                                                                                      l
                                                                                                                      ll
                                                                                                                      ;

i- ! 1 , 

                                                                                                                      !
                                                                                                                      I
                                                                                                                      l
                                                                                                                      l

l l ,. . _ . . - . , 

              .     _ _ . ._    .    _ _ _ , _ _ _ . _    . _   . .   _ _ _ _ -             __   _      _    _ __   _ .
                                                       STP Senior Reactor Operator Exam ... Page 53 of 100
     QUESTION 53:
     A tube leak has occurred in 1D Steam Generator. The Unit is currently performing
     a rapid plant shutdown for repair.
     Which ONE of the following Process and Effluent radiation monitors performs an
     automatic function that will prevent a potential release to the environment during
    .this event?
     A. Condenser Air Removal System Monitor (RT-8027)
     8. Steam Generator 1D Blowdown Monitor (RT-8025)
     C. Turbine Generator Building Drain Monitor (RT-8041)
     D. Main Steam Line "D" Monitor (RT-8049)
                                                                                                           .
 s.
                                                                                                                  %
                                                                                                                        I

L l , 

                                                                                                                        l

                     .. _ .. - - _            _.                _ _ . _        ._ . _ _    __ . _
                                                                                                  l
                                                                                                  4
                                                                                                  i
                                          STP Senior Reactor Operator Exam ... Page 54 of 100
  QUESTION 54:
                                                                                                  1
  Following a LOCA, hydrogen concentration in containment has increased slowly                    !
  over five days reaching 3.5% by volume.
  Which ONE of the following actions will be taken?                                               l
                   .                                                                              1
  A. Both electric hydrogen recombiners will be placed in service immediately.                    I
                                                                                                  l
                                                                                                  ,
  B. Both electric hydrogen recombiners will be placed in service when hydrogen                   I
     concentration reaches 4.0% by volume.
  C. One electric hydrogen recombiner will be placed in service immediately.
  D. One electric hydrogen recombiner will be placed in service when hydrogen
     concentration exceeds 4.0% by volume.
                                                                                                  l
                                   -

4 4 

i
                 ?

  ~ . . . . . .               - ,       . . . - , - - . - - - - - ~ - - - . - . . .                            . . - - - ~. . ..-- . - . ._..- ---
                                                                                                                                                       i

i I 

                                                                                    STP Senior Reactor Operator Exam ... Page 55 of 100

l QUESTION 55: l l. Given the following: 

                    *   A Large Break LOCA has occurred                                                                  .
                    *   OPOP05-EO-EOOO, Reactor Trip / Safety injection, has been completed and the

l operators are transitioning to OPOP05-EO-EO10, Loss of Reactor or Secondary 

                        Coolant.
                    *   The STA reports a Red path on Integrity and the operators transition to
                        OPOP05-EO-FRP1, Response to Imminent Pressurized Thermal Shock Condition.
                    *   While the crew is performing OPOP05-EO-FRP1, the STA reports that the cold                                                     J
                        leg recirculation switchover criteria are satisfied per OPOP05-EO-ES13, Transfer
                        to Cold Leg Recirculation.
                    Which ONE of the following describes the actions required to be taken by the crew
                    at this time?
                    A. Stop FRP1, perform all steps of ES13 and then return to FRP1.
                    B. Stop FRP1, perform the first 5 steps of ES13 and then return to FRP1.
                    C. Continue with FRP1 until completion, then perform all steps of ES13 and then                                                -
                        return to E010.
                 . D. Continue with FRP1 until completion, then perform the first 5 steps of ES13
                        and then return to E010.
                                                                                                                                                     .

I ! 

                                                  .
                ,r-        -      ,                                                  - , , - ~         ,    y-

   . _ . . .       _m    .     __  _.____..__m.    _ _ _ . . _ -___   _ _ . _ _ _ _ . . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _     _
                                                          ~
                                                                                                                                          >
                                                                                                                                          I

' 

                                                                    STP Senior Reactor Operator Exam ... Page 56 of 100
                                                                                                                                          l
                  QUESTION 56:                                                                                                            >
                  The plant has experienced a LOCA outside Containment and is currently on Cold                                           l
                  Leg Recirculation.- The, crew has transitioned to OPOP05-EO-EC11, Loss of
                  Emergency Coolant Recirculation as the LOCA could NOT be isolated in OPOP05-
                  EO-EC12, LOCA Outside Containment.

i in an attempt to conserve RWST inventory, the crew must stop backflow from the L RWST to tne containment sump in accordance with EC11. 

                  Which ONE of the following is the-correct sequence to perform this task?
 '
                                                                                                                                          }
                  A.1. Stop the LHSI, HHSI and CS Pumps

i 2. Open the Si Pump Mini Flow Valves 

                             .
                       3. Open the RWST to SI Suction Header Valves                                                                        ;
                       4. Close the Containment Sump Suction Valves
                       5. Start the LHSI, HHSI and CS Pumps as necessary                                                                  :
                  B.1. Stop the LHSI, HHSI and CS Pumps                                                                                   !
                                           _
                       2. Close the Containment Sump Suction Valves                                                                        ;
                                                                                                                                           '
                       3. Open the RWST to SI Suction Header Valves
                      '4. Open the S1 Pump Mini Flow Valves
                       5. Start the LHSI, HHSI and CS Pumps as necessary

l C.1. Stop the LHSI, HHSI and CS Pumps, 

                       2. Open the RWST to SI Suction Header Valves
                           .
                      ' 3. Close the Containment Sump Suction Valves
                       4. Open the S1 Pump Mini Flow Valves

L 5. Start the LHSI, HHSI and CS Pumps as necessary 

                  D.1. Stop the LHSI, HHSI and CS Pumps
                        2. Open the Si Pump Mini Flow Valves
                        3. Close the Containment Sump Suction Valves                                                                       i
                       4. Open the RWST to SI Suction Header Valves

j 

                        5. Start the LHSI, HHSI and CS Pumps as necessary

i t 

                                                                                                                                            l
             _ ._                                                                                                                   -
                                                                                                                                      .
                                                                                                                                          ,

    ..       ...~             . . . . .      -.              . - - .   _    _ - - . . - ~ -. -- - -   ..... . . - -

! I ! 

                                                         STP Senior Reactor Operator Exam ... Page 57 of 100
       QUESTION 57:
       Given the fo? lowing:

i ' 

       *  . Unit 1 has experienced a small break LOCA
       *   RCS pressure is stable at 1785 psig

! * RVWL plenum is indicating 15% 

       *   CETs are indicating 621 F

l 

       In accordance with OPOP05-EO-FRC3, Response to Saturated Core Cooling, which

l ONE of the following is the appropriate method for reestablishing desired 

       subcooling?
       A. Establish conditions for HHSI flow
B. Establish conditions for LHSI flow
       C. Establish conditions for Accumulator injection
       D. Establish conditions for maximum charging flow
                                                                                                                    .

! . l

! l ! l l I~ , 

,

f f 

$

l. I I Y 

                                               , . - , .             n                              ,  ,

,    . . . _          _ ..      - _ _ . _ - _ _ _ _    .. _ ___. _ ___ _ _ . _ . _.
                                                                                 -                 _ _ . _ . . _ . . _ _ _ . . . _ . .
                                                               STP Senior Reactor Operator Exam ... Page 58 of 100
    QUESTION 58:

l Given the following: 

    .- Unit 1 is at 100% power
    a        All systems in normal operation
    a        Primary RO reports an increase in the Number 2 Seal Leakoff on all RCPs

!

' 

    Considering each condition separately, wl:ich GNE of the following conditions is
    the cause for the increased leakoff flow?
                                                                                                                                       l
    A. Decrease in charging flow
    B. Increase in VCT pressure                                                                                                        ;
    C. Decrease in level of the Number 3 Seal Standpipe
    D. Increase in Seal Injection flowrate

l l 

  '

! l. 

                                                    ._

    . . . _ _ .   .    ..            .-. -. .     . ~.. ..- - ........-.-. - - -                ~ . -- .... -     .   .  .
                                                                                                                            _

! 

                                                              STP Senior Reactor Operator Exam ... Page 54 of 100
                  QUESTION 59:
                . in accordance with OPOP02-RC-0004, Operation of Reactor Coolant Pumps, which
                _.ONE of the following conditions requires tripping the reactor and the affected                           '
                                                                                                                            ,

l- reactor ecolant pump?- (Censider each condition separately) 

                 A. #1 seal water inlet temperature - 250*F
                     .
                  B. Motor stator winding temperature - 300 F                                                              ;
                                                                                                                           l
                  C. #1 Seal Delta-Pressure - 350 psid                                                                     j
                  D. Shaft Vibration - 10 mils                                                                             i
                                                                                                                           J
                                                                                                                           i
                                                                                                                           l

l 

                                                                                                                           I

l . 

                                                                                                                           i
                                                                                                                        *
                                                                                                                           ;
 .-

! i 4 

                                   r

                                  - . . ,      .     ~ -             ~~    ._      .- ..      . - -
                                                                                                    1

- 

                                                                                                    1
                                            STP Senior Reactor Operator Exam ... Page 60 of 100     i

i i QUESTION 60: 

 Unit 2 is operating at 50% power when a RCP 2A NO 2 SEAL LKF FLOW Hi alarm
 is received. After investigation it has been determined that RCP 2A # 3 seal has
 NOT failed. In accordance with OPOPO4-RC-0002, RCP Off Normal, the control                         l

l room operators are directed to monitor: 

 A. RCP 2A vibration and if it starts increasing perform a plant shutdown then
     secure RCP 2A.
 B. RCP 2A standpipe fill rate and if it starts increasing perform a plant shutdown                 i
                                                                                                    '
     then secure RCP 2A.
 C. RCP 2A for 48 hours to determine if # 2 seal has not fully seated.
                                                                                                    !
 D. RCP 2A vibration and if limits are exceeded trip the reactor, trip the turbine, and
     secure RCP 2A.                                                                                 i
                                                                                                    i
                                                                                                    I

l l 

                                                                                                    l

I 

                                                                                                    i
                                                                                                    !
                                                                                                    !

i 

                                                                                                    l
                                                                                                    l
                                                                                                    l

 ,              --            .    . - -      --__        . . _ . - . . .          _ - . .         ._. _
                                                                                                         I
                                             STP Senior Reactor Operator Exam ... Page 61 of 100
     QUESTION 61:
     Given the following:
     .  Unit 1 is at 25% power
     .  CCP 1B is out of service for maintenance
     .  CCP 1 A has tripped and the field supervisor is investigating
     .  PDP will NOT start
     .  RCP 1C has experienced an RCP 1C UPPR OIL RSVR LVL Hl/LO annunciator
     *  RCP 1D CCW to thermal barrier temperature sensor (TS-4638) fails high
     Which ONE of the following is the proper control room operators' response to the
     above plant conditions?
     A. Commence a power reduction to < 10% for performance of a containment
        entry to investigate problem with RCP 1C.
     B. Within one minute trip the reactor and trip RCP 1D.
     C. Commence a power reduction to < 10% and trip RCP 1D.
   -
     D. Within one minute trip the reactor and trip RCP 1C.                                      -

r 

  - -       ~   -   -      . .. . . . -   .. .-_ - .., --- .~ .-.-...- - .- - -.-                                           .    .   .      . . . -
                                                                            .
                                                                                      STP Senior Reactor Operator Exam ... Page 62 of 100
               QUESTION 62:

l 

               During operations at 95% power with pressurizer level at 53%, the Tave input
               (Auctioneered High) to the pressurizer level controller fails low.
               Which ONE of the following is an indication that this failure has occurred?

l ' A. Backup heaters deenergize, charging flow control valve opens, low level l deviation alarm actuates. l l ! B. Backup heaters energize, charging flow control valve closes, high level deviation ! 

                   alarm actuates.

! .C. Backup heaters energize, charging flow control valve _ opens, low level deviation l alarm actuates. l 

               D. Backup heaters deenergize, charging flow control valve closes, high level                                                         ;

i deviation alarm actuates. j 

                                                                                                                                                    i

l 

                                                                                                                                                    ,

! ! I ! l' ! . i )._ < l 

      . . ,                          -  -                 ,.                      .-.                                                     -
                                                                                                                                                  ,
                                                                                                                                                    I

   .- . .    _-.          . .   -.     - . . . - . - - _ ~ . _           - .    . . -     . -    .-       ..           . . . - -.

t- 

                                                                                                                                    I

l t. ' STP Senior Rec? tor 0,,erator Exam ... Page 63 of 100 

                                                                                                                                    l
          QUESTION 63:
          Given the following:                                                                                                      i
          a  Containment pressure transmitter PT-937 has been declared inoperable.

l- 

          *  The required Technical Specification actions have been taken to remove PT-937                                          l
             from service.
                                                                                                                                    l
          Under these conditions, which ONE of the following statements describes the
          coincidence for a Containment Spray Actuation to occur and the actions that will
          result in this coincidence?                                                                                               !
          A.1/3 coincidence after the channel is placed in the TRIP condition.

i B.1/3 coincidence after the channel is placed in the TEST BYPASS condition. l 

          C. 2/3 coincidence after the channel is placed in tite TRIP. condition.
          D. 2/3 coincidence after the channel is placed in the TEST BYPASS condition.                                              I
                                                                                                                                    ;
                                                                                                                         e
                                                                                                                                    i

, 

                                                                                                                                    !
                                                                                                                                ,

' 

                                      ..                                                                                            l

! 

                                                                                                                                    l
                                                                                                                                    ;
 '

i ? 

                       *
                              -                      -
                                                                                   ,_                  -     -       .       - ,. .

 .                      _          --_           . . _    _ ~    . -    . _ . . . _ _ . . _ _ _
                                                                          .                     _ _ . - . . . .
                                             STP Senior Reactor Operator Exam ... Page 64 of 100
     QUESTION 64:
     Per OPGPO3-ZF-0018, Fire Protection Operability Requirements, which ONE of the

l ' 

     following identifies the minimum operability requirements for the Fire Protection
     Water Supply System?
   .
              Storage Tank Volume        Number of Pumps               Pump Capacity
                  gallons - each                                                   in gpm
        A.           200,000                           1                             2500

l B. 250,000 2 2000 l C. 300,000 2 2500 

        D.           350,000                           1                             2000

! ! , ! 

                                                                                                                1
                                                                                                                l

l

l ! ! r 

   ._. .              -           .-    . . . . . - - . .     --         . . .  - - - -    . .   - .-.
                                                         STP Senior Reactor Operator Exam ... Page 65 of 100

. 

            QUESTION 65:
            Which ONE of the following STPEGS Fire Protection Sprinkler Systems provides
            defense against Class "C" Fires?
           - A. Cable Tray sprinklers'

            B. Carbon Filter sprinklers

            C. DG FOST sprinklers

            D. Turbine Bearing sprinklers

. 

                                                                                               1
..

. 

         O

$ 

                                                                                                             .

O i' 1 , k 

            .     ..  -               - - . -       . . _ _ . - .          .  .       ._        .. .

. 

                                              STP Senior Reactor Operator Exam ... Page 66 of 100
   QUESTION 66:
   Given the following:
   *  Unit 1 is at 100% power and Unit 2 is in Mode 6 during a refueling outage.
   *  Fire Hydrant (FH) #13 serving the Essential Cooling Water (ECW) Intake
      Structure area is declared inoperable.

- 

   Which ONE of the following actions must be taken due to this inoperable Fire
   Hydrant?

i A. Restore FH-13 to operable status within 7 days or take the actions of T.S. 

      3.0.3 due to EC'N system inoperability.
B. Establish a continuous fire watch within 1 hour for the Unit 1 ECW intake

< 

      structure and components.
   C. Establish an hourly fire watch for the Unit 1 ECW intake structure and
 ,
      components.
   D. Provide additional lengths of 21/2 inch file hose from an operable Fire Hydrant                 '
      within 1 hour to service the ECW intake Structure.
                                                                                                      l
                                                                                                      I
                                                                                                     /

  _ . ._. _ _ _ _ _          . ._ _ _ _ __._ _ _ .__. _ _ _ _ . _ _ _ _ . . _ .. _ _ _ . . . __ _ _ _ . . . . . ,
                                                                                               STP Senior Reactor Operator Exam ... Page 67 of 100
           QUEST ON 67:
           A Unit 1 Reactor Trip from 100% power has occurred due to a loss of off-site
           power caused by severe weather and tornado damage to the switchyard and the
           138KV Blessing Line. A Tornado Warning is still in effect.
           Given the following:

l l 

           a       Train B 4.16 KV Bus is energized from SDG #12.                                                                                      ;

j '. Train C 4.16 KV Bus is deenergized due to a bus overcurrent lockout. 

           *        SDG#11 was secured due to a failed turbocharger thrust bearing and

l turbocharger damage. 

           * - Restoration of any off-site power source is NOT expected for at least 6 to 8

l l hours. l' = Restoration of SDG #11 is NOT expected for at least 10 hours. l .- Restoration of Train C 4.16 KV Bus is NOT expected for at least 6 hours. 

         ' Based on these indications, which ONE of the following statements is
         .
           recommended per OPOPO4-AE-0001, Loss of Any 13.8 KV or 4.16 KV Bus?

, 

           A. Cross-connect Train A QB C 4.16 KV Bus to Train B 4.16 KV Bus to restore

l power to a CCP for RCS inventory control. - ! ! B. Cross-connect Train A QB C 4.16 KV Bus to Train B 4.16 KV Bus.to restore 

                    power to Pressurizer Heaters for RCS pressure control.

l C. Do NOT Cross-connect Train A QB C 4.16 KV Bus to Train B 4.16 KV Bus - 

                    because the only SDG would be put at risk of overloading.
           D. Do NOT Cross-connect Train A QB C 4.16 KV Bus to Train B 4.16 KV Bus

, 

                    because severe weather conditions exist.

, i . i I }: 

                                                                                                                  .         _  __
                                                                                                                                                -.

                                 .
                                         STP Senior Reactor Operator Exam ... Page 68 of 100 )

QUESTION 68: Given the following:

  • Unit 1 is in Mode 5
  • RCS temperature is 180 F

a RCS pressure is 225 psig

  • RHR Train 1 A is in service maintaining RCS temperature (with both RHR HX
   BYPASS FLOW CONTROL and OUTLET TEMPERATURE CONTROL valves in a
   throttled position)

+ 125 VDC system E1 A11 TROUBLE annunciator is received . E1 A11 bus voltage indicates 0 volts Which ONE of the following describes how the operating RHR train will respond? 

                                                                                             l
           BYP FLOW CONT            OUTL TEMP CONT                      RCS
               (FCV-0851)               (FCV-0864)               TEMPERATURE
 A.              OPENS                    CLOSES                    INCREASES
 B.              OPENS                  NO CHANGE                   INCREASES
 C.           NO CHANGE                    OPENS                    DECREASES
 D.              CLOSES                    OPENS                    DECREASES

                                                                                                   1
                                                                                                   l
                                                                                                   i
                                                                                                   '
                                        STP Senior Reactor Operator Exam ... Page 69 of 100
                                                                                                   i

QUESTION 69: Unit 2 was at 100% power, steady state conditions, when an electrical disturbance occurred that produced the following Control Room indications: . 2A Steam Generator PORV controller is NOT illuminated and cannot be operated 

   from the Control Room.

. Ventilation Monitors RT-8012, RT-8033, and RT-8035 are indicating Magenta 

   color on the RM-11.                                                                             1

. All three trains of the Control Room Envelope, Fuel Handling Building, and 

   Containment HVAC are in the ACTUATED condition.

. Charging Flow Control valve FCV-205 is in Automatic and closed. . Several QDPS Plasma displays in the Control Room are blank. Which ONE of the following power distribution panels has lost power?  ; A.120 Volt AC Class 1E Panel DP1201 B.120 Volt AC Class 1E Panel DP001 C.120 Volt AC Class 1E Panel DP1202 

                                                                                               .

D.120 Volt AC Class 1E Panel DP002 

                                                                                                 ~
                                                                                            .-

    . . ~ . .   .        .~,- - . ~    . . . . _ , . - . - . - . . . . .      - ~ . - . -  .~    .     .    ..       ._ .-

! I STP Senior Reactor operator Exam ... Page 70 of 100 l 

              QUESTION 70:

l Considering the following time line: , l 

              =   1200 - Large Break LOCA occurs
              .   1214 - Alert is declared, based on leakrate greater than makeup capability
              .   1232 - Loss of all HHSI and LHSI pumps is reported
              +   1251 - Containment hydrogen concentration > 0.5 %
              +   1300 - Sight Area Emergency is declared, based on high RCB radiation
              .   1343 - Core Damage confirmed by failed fuel monitor and RCS sample results
              =   1420 - General Emergency is declared, based on' rapid decrease in RCB pressure
Which ONE of the following is the EARLIEST time at which a Protective Action

l Recommendation (PAR) is REQUIRED to be issued? - 

              A.1315
              B.1358
              C.1403
              D.1435                                                                                                        ,
                                                                                                                            ;
                                                                                                                            l

, ! !' I f 1 , 

  ,

                                                                                                         l
                                                 STP Senior Reactor Operator Exam ... Page 71 of 100

t

- QUESTION 71:

. . Maintenance requests an Equipment Clearance Order to perform adjustment and 

      testing of limit and torque switches for a motor. operated valve (MOV).
                                                                               ~

! In order to perform the required testing, the MOV: > A. can only be operated under a Test Tag by the Operational Authority when 4 

          requested by the individual or classification performing the maintenance.

[ B.' can only be operated under a Caution Tag by the Acceptor to whom the j Caution Tag is issued and only after Operational Authority approval. l I 

      C. shall be tagged with both a Caution Tag and a Test Tag with the Caution Tag
          stating that removal of the Test Tag permits operational testing.
D. shall be tagged with both a Danger Tag and a Test Tag, with the Danger Tag

! removed prior to any MOV operation during the tssting. l- I 

  +,                                                                                                 .

, ! ~

, . _ _ . _ _ _ . . . . _ . _ _ . . . . . _ _ . . . . _ _ . . . - . _ . . _ _ _ . . . _ . _ _ _ _ _ _ _ . , ._...._.. ___ _ ..._.. _ __ _ . i 1 

                                                                            STP Senior Reactor Operator Exam ... Page 72 of 100
                 QUESTION 72:
                Which of the following situations require Independent Verification?
                 1. Initial system valve lineup of TGB Fire Protection System following an outage.
               ' 2. Removing electrical grounding device from the 13.8 KV switchgear following
                        breaker maintenance.
                 3. Functional test has verified the resetting of overcurrent trip relay setpoint on
                        RCB Chiller 11 B breaker,1 D2/6, following electrical troubleshooting.
                 4. Throttling and locking CV-0198 prior to Mode 4 entry per Technical
                        Specifications.
                        A.1,2,3
                        B.1,2,4
                        C.1,3,4
                        D.2,3,4

l l l t- h I 

                  - - .     .
                                                                          ,                  _

                                                 .. . . .  . . . - - . . ~ .               . . - - . - .

. ^ 

                                              ' STP Senior Reactor Operator Exam ... Page 73 of 100
                                               .
                                                                                                             i
!

, 

     QUESTION 73:-                                                                                           ;
                                                                                                             I

     Given the following plant conditions:
:   -a   Unit 1 was operating at 100% power when a loss of offsite power occurred.
'
     .   The following indications' are observed while performing the immediate actions
         of OPOP05-EO-EOOO, Reactor-Trip or Safety injection:                                                ,

' * All Rod Bottom Lights are NOT LIT 

                                                                                                             l
             *
                 Reactor Trip Breaker R is OPEN -                                                            i
                                                                                                              '
             *
                 Reactor Trip Breaker S is CLOSED
             *
                 Bypass Breakers R and S are OPEN
             *

. 

                 N35 and N36 IR Channels indicate decreasing neutron power

! * The previously observed indications do NOT change after attempting a manual ) trip from both reactor trip switches j . 

     Which ONE of the following actions should be taken by the crew?

!.

A. Go to OPOP05-EO-FRS1, Response to Nuclear Power Generation - ATWS.
       .
                                                                                                             l

< 

                                                                                                             1
     B. Dispatch an operator to shift DRPl to the alternate power source, verify all Rod
.
         Bottom Lights are lit, then continue on with EOOO.
                                                                                                         *

!  ! 

     C. Open 480V Load Center 1K1 and 1L1 feeder breakers, then continue on with                             ;
         EOOO.
     D. Continue on with the immediate actions of EOOO.

i ~ 

                                                                                                             ,

' l 

                                                                                                             l

i 

  ,

! 

                                                                                                             i
                                                                                                              l
                                                                                                             !
                                                                                                             i
                             .

                            . . _ .          __    _ _. _        ..     -            .    _. . . _ .
                                    .
                                         STP Senior Reactor operator Exam ... Page 74 of too
 QUESTION 74:

l Given the following: ! j . A Reactor Trip occurs due to a Loss of Offsite Power. 

 a  The ESF diesel generators have all started and restored power to their ESF

, 

    buses.

! 

 *  The Control Room crew has just completed the immediate Actions of
    OPOP05-EO-EOOO, Reactor Trip or Safety injection.
 Which ONE of the following correctly identifies the status of containment cooling?
 A. The RCFCs are NOT running and CCW is flowing through the cooling coils.
 B. The RCFCs are running and there is NO flow through the cooling coils.
 C. The RCFCs are running and CCW is flowing through the cooling coils.
 D. The RCFCs are NOT running and there is NO flow through the cooling coils.

l I ' I l l 

                                                                                                     l
                                                                                                     !

     -
           -
                       ~_.-                                                              ~.   . . - . . .

I 

                                                                                                               ]

' 

                                            STP Senior Reactor Operator Exam ... Page 75 of 100                ;
   QUESTION 75:

! 

   Unit 1 is in Mode 4, coming out of a refueling outage, when a loss of instrument                            j
   air occurs.
                                                                         .,
   Given the following:
   *   .PCS temperature is 300"F

, 

   a    tiarrow Range SG levels; 1 A = 40%,1B = 35%,1C = 55%,                   1D = 20%
   *    SG PORVs are all operable
   *    All Motor-driven AFW Pumps are operable
   * ' RHR Train status; 1 A and 1C - available,1B inservice and supplying Low
                            '
        Pressure Letdown
   *    St Pumps 'are in the normal Mode 4 alignment
   *    Both Pressurizer PORVs are operable
   .    Vacuum has been established in the Main Condenser
   * -Three Circulating Water pumps are running
   Which ONE of the following RHR valve alignment methods are Control Room                                     i
   operators directed to use for RCS temperature controlin accordance with                                     l
   OPOPO4-IA-0001, Loss of Instrument Air, Addendum 67
 .
                                                                                                           -
   A. LOOP B TC INJ MOV-0031B - open, cycle RHR B Heat Exchanger OUTL TEMP
        CONT HCV-0865 as needed for temperature control.
   B. MINI FLOW MOV-0067B - open, cycle RHR B Heat Exchanger BYP FLOW CONT
        FCV-0852 as needed for temperature control.                                                          -
   C. MINI FLOW MOV-0067B - open, cycle LOOP B TC INJ MOV-0031B as needed
                                                                                                               '
        for temperature control.
   D. LOOP B TC INJ MOV-0031B - open, cycle RHR B Heat Exchanger CCW OUTL
        FV-4548 as needed for temperature control.
                                                                                                               l

i l 

                                                          .                                             .,

 . . _ . _      ..   . _ . = . . . . . .    ._ ._  _.    _ . _ . . _ _ . _      ______ ...__ __              . _ _ _ _ ,_
                                                                                                                          ,
                                                                                                                          1
                                                          STP Senior Reactor Operator Exam . . Page 76 of 100
           QUESTION 76:
           Given the following:
           .  Unit 2 is at 100% power, steady-state conditions
           .  CCP 2B is operating, CCP 2A is available
           .  CHG FLOW Hl/LO alarm is received on Lampbox 04M8
           .  Operators determine that CCP 2B has tripped
           .  The Unit Supervisor directs the Primary RO to start CCP 2A                                                  l
           Which ONE of the following valve alignments should be performed prior to starting
           CCP 2A in accordance with the alarm response instructions?

, 

                   MOV-8377A, 2A CCP              FCV-201, 2A CCP              FCV-205, Charging

i Discharge Isolation Miniflow Valve Flow Control Valve 

             A.                        OPEN                 OPEN                                CLOSED
             B.                    CLOSED                   OPEN                                 OPEN
             C.                        OPEN              CLOSED                                  OPEN
             D.                    CLOSED                   OPEN                                CLOSED

 ..     . . . . . _ _
               -
                                 . . _ . _ . _ _ _ . _ , _ . . _ . - _ _              _.- --. .___ _ . _ .. . _ _ _ _ _ . . - . _ . _ _       _ - . - .
                                                                                                                                                          l
                                                                           STP Senior Reactor Operator Exam ... Page 77 of 100
             QUESTION 77:
            Given the following:
          .*          Unit 1 is in Mode 5                                                                                   .
                                                                                                                              7
             * RCS boron concentration is 2800 ppm
             .        Rods are locked in the Rapid Refueling position
          --*         Preparations are being made for refueling operations
          . A fire has been reported in the Train C ESF Switchgear room and it appears to be
           . spreading toward the Rod Drive MG Set Room.
             In accordance with OPOPO4-ZO-0008, Fire / Explosion, which ONE of the following
             describes the operator actions to be taken? -
           ' A. Open the reactor trip breakers and ensure all rods are fully inserted
             B.' Place Rod Holdout Power Supply to OFF and ensure all rods are fully inserted
             C.: Direct I&C to restore rod control, then fully insert all control rods
      -
             D. Initiate emergency boration until RCS boron concentration is a 3000 ppm-                                                  -
                                                                                                                                                    -     I
    .
                                                                                                                                            *

. 

                                                                         T      -m                                                   -  -               .

                                                                                           1
                                                                                           l
                                                                                           l
                                       STP Senior Reactor Operator Exam ... Page 78 of 100

QUESTION 78: While responding to an inadequate Core Cooling situation the operators are directed to reestablish some form of core cooling. Which ONE of the following statements describes the INITIAL response of the CETs, and the reason for this response, if the operators restarted the safety injection pumps? A. Increase due to saturated steam being forced out of the core. B. Decrease due to saturated steam forming a frothy two phase mixture. 

                                                                                           '

C. Increase due to superheated steam being forced out of the core. D. Decrease due to superheated steam causing injected water to boil forming a 

   frothy two phase mixture.
                                                                                           :
                                                                                  .

                                                                                .        ..
                                             STP Senior Reactor Operator Exam ... Page /9 of 100
 QUESTION 79:
 The following plant conditions exist:
 ( Loss of seal cooling (CCW AND seal injection) to all RCPs               -
 a  f'anual reactor trip _and trip of all RCPs
 .   Performed OPOP05-EO-E000, Reactor Trip and Safety injection and
    OPOP05-EO 'dS02, Natural Circulation Cooldown
 .  Transitioned to OPOP05-EO-ESO3, Natural Circulation Cooldown with Steam
    Void in Vessel due to limited inventory in the AFWST
 *   RVWL - Upper Head is 80%
 .   Pressurizer level is 30%
 *   RCS Subcooling is 90*F
 While performing ES03, seal cooling is restored to RCP D.
 Which ONE of the following actions must be performed prior to starting RCP D7
 A. Decrease pressurizer level to accommodate void growth
 B. Increase pressurizer level to accommodate void collapse

. 

                                                                                                 .
 C. Decrease subcooling to accommodate void growth
 D. Increase subcooling to accommodate void collapse
                                                                                                   .

  .--       - -              --      _   _ ._- - _ . .     . ~ _ =           = . .   - . ~ .        _ . _ _
                                       -

!- l ! 1 I 

                                                STP Senior Reactor Operator Exam ... Page 8o of 100             q
   QUESTION 80:

! 

   A plant startup is in progress with main turbine roll commencing and reactor power
   at 15%. All procedural requirements of OPOP03-ZG-0005, Plant Startup to 100%

l have been met for the current power level. Power range channel N-44 is out of l service due to a failed detector. 

   Which ONE of the following reactor trip signals will cause the reactor to trip under
   these conditions?-
   A. Pressurizer low pressure
    B. Single loop loss of flow
    C. Intermediate Range high flux
    D. Turbine trip
                                                                                                                 1
                                                                                                                 I
                                                                                                                J
                                                                                                                 l
                                                                                                                l
                                                                                                                f
                                                                                                                l

i 

                                                                                                                !

l 

                                                                                                            .
                                                                                                              _

          .   ..          . _ _ . . . . ... . - .    - - - -. - . - . . . _ . .    - - - . . - .      . -

! 

                                                                                                             ,
                                                  STP Senior Reactor operator Exam ... Page 81 of 100

. 

   QUESTION 81:
   Given the following:
   *   Unit 1 is at 100% power                                                                                !
                                                                                ..
   *   Reactor power is slowly increasing
   *   RCS Tave is 588'F and slowly decreasing
   *   Turbine control is in IMP OUT
   Which ONE of the following conditions is causing the above plant transient?
   A. Main Turbine Governor valve # 3 has failed closed.
   B. . Extraction steam to a feedwater heater string has been isolated.
   C. Main Steam to Deaerator valve has failed closed.
   D. Circulating Water pump #14 has tripped.
 .                                                                                                    .
                                                                                                              ,
                                                                                                          e

! ! ! . 

                                                                                                            )

                                                                          . _ _ _ _ _ _ _ . _ _ _ _ _ _ - _ . . . . . . _ . _
 .
                                            STP Senior Reactor Operator Exam ... Page 82 of 100
   QUESTION 82:
                                                                                                                              i
   Given the following:
   *  Unit 1 has tripped from 100%
   *  All steam generator levels indicate approximately 50% WR
   *  A fault has left Vital DC bus E1D11 deenergized
   Which ONE of the following describes the effect this event has on the status of
   #14 AFW pump, its associated AFW REG valve and AFW OCIV?
                                                                                                                              i
   A. Pump will start, REG valve will remain closed and OCIV will remain open                                                 )
   B. Pump will NOT start, REG valve will remain closed and OCIV will remain open
   C. Pump will start, REG valve will remain open and OCIV will remain closed
   D. Pump will NOT start, REG valve will remain open and OCIV will remain closed

l i ' i I

I , 

   - _.   ..     _.     .       .   . - . _    _ - . . = .      . .   , .   _.    -     -    . . - .  .       -   -   ..

! 

                                                           STP Senior Reactor Operator Exam ... Page 83 of 100
             QUESTION'83:
             Given the following:                                                                                        !
                                                                                                                         '

I * Unit 1 is in MODE 5 with RCS cooldown to refueling conditions in progress. 

             .  The MEAB watch enters the FHB and discovers the Spent Fuel Pool (SFP) Area
                Radiation Monitors RT-8090 and RT-8091 in the HIGH ALARM condition.

l * SFP level is 66 feet 61/2 inches and has decreased 1/4 inch in the last 24  ; I hours. 

             *  SFP temperature is 90 F and steady.                                                                      :
             *  No other abnormal plant or alarm conditions exist.
             Which ONE of the following actions must be taken?
             A. Notify Control Room of SFP level and commence makeup to restore level.

l 

             B. Verify FHB HVAC is operating in the Emergency Mode.

l 

             C. Notify personnelin the affected area of the High Radiation alarm.
        -
             D. Verify SFP Gate Seal Air Pressure is in the normal band (30-34 psig).                           -

i [ t l * , l ! 

                                                                                                                          i

! ! ! f { 1 l ! 

 '

, E 

     . - - . . _ _-      __     __._.m._  .m . _ _ . . . . . . _ -_ _ . . _ . . _ _ . . _ . . - . - . . . _ - . - _ __ - . . . _ . . _ . _ - . . . _ . . _ . . _ _ _ . . _
  .-
                                                                           STP Senior Reactor Operator Exam ... Page 84 of 100
                     QUESTION 84:
                     It has been determined that Spent Fuel Pool level needs to be raised to account for
                     loses due to evaporation.
                     Per OPOP02-FC-0001, Spent Fuel Pool Cooling and Cleanup System, which ONE of
                     the following source combinations is used to makeup to the SFP to compensate for

!- normal EVAPORATION? 

                     A.-      Reactor Makeup Water System or Refueling Water Storage Tank
                     B.       Reactor Makeup Water System or Demineralized Water System
                    -C.       Boron Recycle System or Refueling Water Storage Tank

l l D. ' Boron Recycle System or Demineralized Water System !- ! l n l l I f l l l

1

! l

'

I 

                                                                                                                                                                             !
                                                                                                                                                                             I
                                                                                                                                                                             l
                                                                                                                                                                             l

i i k . 

                                      . . . _ _ . _     ___. .        _ _ _ _ = _ . . . _ _ _ _ _ __ __   _ _   _
                                                    STP Senior Reactor Operator Exam ... Page 85 of 100
   QUESTION 85:
   Upon loading a spent fuel assembly into the reactor core the Core Loading
   Supervisor notifies the Control Room that an excessive amount of gas bubbles are
   emerging from an adjacent fuel assembly and that RCB 68' Area Monitor RT-8099
   Just went into High alarm. The Control Room performs the actions of
   OPOPO4-FH-0001, Fuel Handling Accident, and directs the operation of the
   Containment Carbon Filter Units.
   Which one of the following statements is correct?
   A. The Containment Carbon Filter Units are NOT required to be in operation until
      RT-8099 is determined to have a VALID High alarm.
   B. The Containment Carbon Filter Units are NOT required to be in operation until
      MORE than one RCB Area Monitor is determined to have a valid High alarm.

' 

   C. The Containment Carbon Filter Units are required to be in operation if radiation
      levels or sampling indicates fuel cladding damage has occurred.
 .

, 

   D. The Containment Carbon Filter Units are required to be in operation anytime
                                                                                                        -
      OPOP04-FH-0001 is entered.
                                                                                                                    l
                                                                                                              s

r 

                                                                                                                    l
                                                                                                                    l

l ? l l l l , l l ) 

                                   .
                                         STP Senior Reactor Operator Exam ... Page 86 of 100
  QUESTION 86:
  OPSP03-ZO-0028, Operator Logs, requires that RCFC outlet temperatures be
  maintained less than or equal to 75 F and greater than or equal to 50 F.
  Which ONE of the following r,tatements is the reason for these temperature limits?
        Hioh Temoerature Basis          Low Temoerature Basis
  A.   CRDM Cooling                     SI Accumulator Nil Ductility Temp.
  B.   LB LOCA Peak Clad Temp.          Si Accumulator Nil Ductility Temp.
  C.   CRDM Cooling                     SI Accumulator Boron Solubility

, 

  D.   LB LOCA Peak Clad Temp.          SI Accumulator Boron Solubility

l

l 

                                                                                               1
                                                                                               :
                                                                                               )

l l . A 

                                                                                               l
                                                                                               l
                                                                                               l

l ! l l- I t ) 

  . - .- - .-- .-. ., -.                                ~.- _ .             .-    .. ~~ ~. .. . - - . . . -         . - . -   _

i- i l I STP Senior Reactor Operator Exam ... Page 87 of 100 

             QUESTION 87:
             A point source in the RCA is reading 500 mrem /hr at 1 foot. There are two                                           j
             options being considered to perform a job in the area.
                                                                                                                                  '
                                                                                                            '
                  Option #1: Two operators working together are capable of com[leting the job in
                                   20 minutes at 4 feet from the source.

1 

                  Option #2: One operator can complete the job in 80 minutes at 8 feet from the
                                   source.
                                                                                                                                 I

i Which ONE of the following is the preferred option AND consistent with the goals  ! ! of the ALARA program? l A. Option #1 - each operator's exposure is 10.0-10.5 mrem l 

                         B. Option #2 - the operator's exposure is 5.8-C.2 mrem

I 

                         C. Option #2 - the operator's exposure is 10.0-10.5 mrem
                         D. Option #1 - each operator's exposure is 5.8-6.2 mrem

l 

                                                                                                                            s
                                            e

l

l i L 

o

i i 

                                                                                                                                }

     .    . _        _ . . _ _ _ . . . _ . .    _ , . _ _ .   ...___._-..-      ._-  _ .  - . _ _ _ _ _ - _    ~._. . _ . . . .

l i. I. I STP Senior Reactor Operator Exam ... Page 88 of 100 l l QUESTIOrJ 88: ! l The maximum dose rate in a room is 1200 mrem /hr one foot from a certain valve. l According to the STPEGS Radiological Access and Work Controls procedure, this 

    room should be classified as which ONE of the following?

l A. High Radiation Area

B. High Radiation Exclusion Area

l C. Locked High Radiation Area l l D. Very High Radiation Area !'o i l , I

f I I I l 

                                                                                                                                ;
                                                                                                                                i
              e

l t' 1 4 i i . I 

                                                                                                                                I
                                                                                                                                l
  '
                                             .-            -

   - - - - . -.- .~ -._ -. -                    - -- - - -.. -                - -             - -.              . ..  . .   ..-_,
 '
                                                                 STP Senior Reactor Operator Exam ... Page 89 of 100

l 

                   ' QUESTION 89:

l- ' Given the following:-- ( * Unit 1 is operating at 100% power. .._ i e 1CCW surge tank level is 58% and decreasing at 1 % per minute l * Crew enters OPOPO4-CC-OOO1, Loss of Component Cooling Water l_ 

                     e     Operators trip the reactor and all RCPs

I * . CCW pump 1 A is out-of-service for maintenance 

                     *     CCW pump.1B trips on thermal overload
                     *    .CCW pump 1C is running                                                                                 l

l Which ONE of the following describes the charging pump (s) wiiich may be l operated under the current plant configuration?

l A.1 A Centrifugal Charging Pump ONLY 

                                                                                                                                  l

l B.1B Centrifugal Charging Pump ONLY 1 l. C.1 A Centrifugal Charging Pump OR Positive Displacement Charging Pump 

                .
                     D.1B Centrifugal Charging Pump OR Positive Displacement Charging Pump                           -
                                                                                                                                  l
                                                                                                                          .
                                                                                                                                  J

, ~ 

                                                                                                                                  \

!  ! 1 ! i t 4 1 

 '

    . _ _ . . _ .            _  _. .      . _ - _ _ . _-_=_-.~.__.__.m._                  _ . _ _ _ _ . _ _ . _ . . _ _ . _ - _ . _ . ._

i 

                                                                                                                                         I

I < 

                                                                                                                                         l
                                                                                                                                         ,
                                                                                                                                         1
    -
                                                            STP Senior Reactor Operator Exam ... Page 90 of 100                          l
               QUESTION 90:
               Which ONE of the following conditions w5 warrant entry into OPOPO4-CC-0001,
               Loss of Component Cooling Water?
               A. Trip of running CCW pump (s) and standby CCW pump fails to start.
               B. Loss of ECW flow to all CCW Heat Exchangers.

t 1 ! C.-Vital CCW Header Supply valves automatically close. 

               D. Leak in CCW system greater than makeup capacity.
                                                                                                                                         l

' 

                                                                                                                                         .
                                                                                                                                         l

I t. ! l '. 

i
r
  '

i I 

                                                                                                                                         i

! 

  --     .     . . . -    .   - - - - - .    . . . . __- - - .     -.         ~.     .     . . .
                                                                                                      !

a i 

                                                STP Senior Reactor Operator Exam ... Page 91 of 100

! l QUESTION 91: 

     A modification to the Containment Spray System is to be made and a 10CFR50.59
Evaluation is to be performed.

. ,- l Which ONE of the following describes the 10CFR50.59 process? 5 i A. The modification undergoes a Preliminary Screening followed by a Final l Screening then a Unreviewed Safety Question Evaluation before it is reviewed 

        by the PORC.
7
     B. The modification undergoes a Unreviewed Safety Question Evaluation followed

}- by a Preliminary Screening then a Final Screening before it is reviewed by the j 

        PORC.                                                                                         l
                                                                                                      :
                                                                                                       l

l C. The modification undergoes a Final Screening followed by a Unreviewed Safety  ; j Question Evaluation then it is reviewed by PORC prior to approval by the Plant  ; 

        Manager.
     D. The modification undergoes a Unreviewed Safety Question Evaluation followed                   ,

, by a Preliminary Screening then it is reviewed by PORC prior to approval by the 

        Plant Manager.                                                                              -

l i 

                                                                                                      l

1 I 

                                                                                                      l
                            4

          . . .     .-       - ._         ..            -.           . . . - . ..       -.      . _ _ . . .

, 

                                     -
                                                                                                            I
                                                                                                            ll
                                             STP Senior Reactor Operator Exam ... Page 92 of 100
                                                                                                             l
 QUESTION 92:                                                                                               l
 Unit 1 is in Mode 1 at 100% power when the following occurs:
 .  1300 hours - Fire reported in the Unit 1 Relay Room and fire brigade dispatched
 .  1307 hours - Fire out of control and smoke infiltrating control room atmosphere
 .  1308 hours - OPOPO4-ZO-0001, Control Room Evacuation procedure is entered                               ,
    1310 hours - Control Room actions of OPOPO4-ZO-0001 are complete
                                                                                                             '
 a
 .  1311 hours - All personnel have evacuated the Unit 1 Control Room
 .  1318 hours - ASP and all Remote Shutdown Stations are manned
 .  1325 hours - Communications established between ASP and Remote Stations
 .  1330 hours - Control of SG 1D PORV and #14 AFW Pump are transferred to
    ASP                                                                                                     '
 .  1338 hours - All Switchgear Room Alignment Addenda complete
 Based on the above timeline, which ONE of the following Emergency
 Classifications is appropriate for this event?
 A. Unusual Event
 B. Alert
 C. Site Area Emergency
                                                                                                            .
 D. General Emergency

l l l 

                                                                                                            i

     .   _.        ._ _    __    ._   _ _ _ _ _ . _ .           _              .       ._     ._      _ __ .   ._

( l- l l l STP Senior Reactor Operator Exam ... Page 93 of 100 1 

    QUESTION 93:

i 

    As the Shift Supervisor acting as the Emergency Director, which ONE of the
    following Emergency Director responsibilities and authorities may you delegate?
                                                                                   ...
    A. Approving required communications with the NRC.
    B. Approving required notifications to the State and County.
    C. Approving radiological exposures in excess of 10CFR20 limits.

l 

    D. Approving departure from license conditions per 10CFR50.54(x).
                                                                                                                  l
                                                                                                                  j
                                                                                                                  l
  .
                                                                                                             *
                                                                                                                ,

l i a ' 

                  .     .    .    .   . . ~ . . . . . .    .         . - - .                    ... .

l l STP Senior Reactor Operator Exam ... Page 94 of 100 

     QUESTION 94:
     Given the following:

l 

     .   Unit 1 is at 100% power
     a   Containment pressure is indicating +0.4 psig
     *   Preparations for performing a containment supplementary purge are in progress.
     .   Noble gas concentration inside containment is 5.5 E-04 ci/cc
     Which ONE of the following actions will be taken to prevent actuation of CVI?
     A. Ensure all RCFCs are in operation to provide mixing of the containment
         atmosphere.
     B. Increase the High alarm setpoint on the RCB Purge Exhaust radiation monitors
         (RT-8012 and RT-8013).
     C. Ensure that containment carbon units are in operation to reduce noble gas
         concentration.

1 ' 

     D. Increase the High alarm setpoint on Containment Atmosphere radiation monitor

l (RT-8011). ( ! 

                                                                                                      l

l ! 4 

 i
   -

   ._ .-      ~-            -.   - .--. - -. -   -. - -..      .. _ .. -..- _ . ~ _ -... - ..._.
                                                                                        .
                                                                                                      ~. . . - . - . - - . -
                                                       STP Senior R actor Operator Exam ... Page 95 of 100
           QUESTION 95:
           The following plant conditions exist:
           *  Reactor Power is 55%                                                               .:-

. 

           *  Main condenser vacuum is 26 inches Hg and decreasing

l 

         ' Which ONE of the following describes the appropriate operator action per
OPOPO4-CR-0001, Loss of Condenser Vacuum? j
           A. Ensure the reactor and main turbine are tripped at 22 inches Hg.                                               l
           B. Ensure the reactor and main turbine are tripped at 25 inches Hg.

! 

           C. Ensure the main turbine is tripped at 22 inches Hg.
                                                                                                                             l
           D. Ensure the main turbine is tripped at 25 inches Hg.

J I 

                                                                                                                        ..

J f 

                                                                                                                             I
 n

                                                                                               l
                                        STP Senior Reactor Operator Exam . . Page 96 of 100

QUESTION 96: l 

                                                                                               1

Unit 2 is operating at 100% power when all circulating water pumps trip. 

                                                                                               1

Assuming no operator action, except for AFW flow control, and all systems i function as designed, which ONE of the following corresponds to plant conditions l 

                                                                                               l

after the plant has stabilized? 

             RCS TAVG          SG Pressures

A. 567 F 1185 psig B. 567 F 1225 psig C. 571 F 1185 psig D. 571 F 1225 psig 

                                                                                            _A

  . . .   .      .. ..       .~..   -.      . . . . . - - - . . -                _ - . . . . .    ---..       - -- -

4 

                                                           STP Senior Reactor Operator Exam ... Page 97 of 100
                                                                                                                     l

- 

        QUESTION 97:
                                                                                                                     ,
1       A pressurizer PORV has failed open causing a manual reactor trip and safety

        injection. The control room operators are responding per OPOP05-EO-EOOO,                                     !
Reactor Trip or Safety injection. ..
                                                                                                                     j
        Which ONE of the following is an indication that the PRT rupture disk has
ruptured?
        A. Containment pressure is increasing at 5 psi per minute.
        B. PRT temperature is decreasing.
        C. PRT level spikes to 100% then decreases to 0%.                                                            ;
        D. RCB dew point is decreasing.
                                                                                                                     :
                                                                                                                     )
                                                                                                                -
                                                                                                                     i
                                                                                                                     i
                                                                                                                     i

                                   .
                                         STP Senior Reactor Operator Exam ... Page 98 of 10o
 QUESTION 98:
 A High alarm is received on RT-8032, GWPS Outlet Radiation Monitor, and the
 following actions are performed by the crew:
 .   OPOPO4-RA-0001, Radiation Monitoring System Alarm Response is entered
 .   Alarm is verified to be VALID
 .   Health Physics and Chemical Analysis are notified
 Which ONE of the following radiation monitors should be checked for increasing
 readings in accordance with OPOPO4-RA-00017
 A. RT-8031, GWPS Inlet
 B. RT-8058, MAB 10' Area Monitor
 C. RT-8029, MAB Ventilation
 D. RT-8010A, Unit Vent Stack

i 

  -       ._.._.-_..____m
                  -         .-     . . _ . _ . . _ _       . _ . . . _ . _ . . . . . _ _. __
                                                                                               _ . . , . _ . . _ _ . _ . . _ . . .,

, ' 

                                                                                                                                   !
                                                                                                                                   !

l l STP Senior Reactor Operator Exam ... Page 99 of 100 i 

    QUESTION 99:                                                                                                                    ,

( Plant Operators are performing a transfer of the Floor Drain Tank to the Waste 

    Monitor Tank per OPOP02-WL-0003, Floor Drain Tank Operation...
    Which ONE of the following is a condition that the procedure cautions could occur

l 

    if procedural steps are performed improperly?
                                                                                                                                    !

l A. Contamination of non-contaminated systems 

    B. Exceeding STP airborne limits
                                                                                                                                    t
    C. Contamination of personnel

l 

    D. Exceeding STP personnel dose limits
                                                                                                                   .

, l' , 

                                                                                                                                     I

, 

                                                                                                                                     ,

t 

                                                                                                                                     l
                                                                                                                                     l

i- l l 

                                                                                                                                     l
                                                                                                                                     !

! 

                               -                     - - -
                                                                                           , _

                                                     .
                                            STP Senior Reactor Operator Exam ... Page 100 of 100
  QUESTION 100:
  OPOP02-CD-0001, Condensate System, requires that the " Condensate pump
  discharge common vent line isolation valve and the Condensate pump 13 vent at                   .
                                                                                                  '
  the miniflow check valve" be open prior to starting Condensate pump 13.
  Which ONE of the following statements is the reason for opening these two valves                I
 . prior to starting the pump?
  A. To flush out corrosion particles that collect in the low flow areas of recirculation         :
      line check valve CD-0740.                                                                   l
                                                                                                  t
  B. To extend the life of recirculation valve CD-FV-7016 by eliminating air pockets              ;
      that promote erosion.                                                                       l
                                                                                                  .
  C. To minimize the occurrence of hydraulic transients in the Condensate pump #13                ,
      recirculation line.
                                                                                                  ,
  D. To ensure that no steam binding will occur in the Condensate pump (Iow                       ,
      pressure, low temperature impelier boiling).                                               !

, 

                                                                                                  ;
                                                                                                 !
                                                                                                 I
                                                                                                  .
                                                                                                  h
                                                                                                 ,
                                                                                                  h
                                                                                                 I
                                                                                                 t
                                                                                                 k
                                                                                                 i
                                                                                                 e
                                                                                                 :

}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20154L410&oldid=3439077"