IR 05000266/1996303: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot insert)
 
(StriderTol Bot change)
Line 19: Line 19:


=Text=
=Text=
{{#Wiki_filter:
{{#Wiki_filter:.- - .-. . . _ . -. ._
.
 
U.S. NUCLEAR REGULATORY COMMISSION REGION lll Docket Nos: 50-266; 50-301 Licenses No: DPR-24; DPR-27 Reports No: 50-266/96303(OL); 50-l'31/96303(OL)
Licensee: Wisconsin Electric Power Company Facility: Point Beach Nuclear Plant Units 1 and 2 Location: 6610 Nuclear Road Two Rivers, WI 54241 Dates: October 7-11,1996; November 20-27,1996 Inspectors: J. Lennartz, Chief Examiner, Rlli P. Cataldo, Examiner, Rll!
T. Guilfoil, Examiner, Sonalysts In K. Parkinson, Examiner, Sonalysts In Approved by: M. Leach, Chief, Operator Licensing Branch Division of Reactor Safety i
      !
9701280303 970123 PDR ADOCK 05000266 V  PDR
 
. . . . _ ._ _ _ -. . _ _ . _ _ . _ _ _ . _ . _ _ _ _ . . _ _ . _ _ . _ _ _ .
      -
        . _ . . - _ _ _ . _ _ _ . _ _ . . _ . . . .
U.S. NUCLEAR REGULATORY COMMISSION
            /
i      REGION lll
      '
Docket Nos:  50-266; 50-301    \
DPR-24; DPR-27    ^'
Licenses No:
          *
          ,
l            '
Reports No:  50-266/96303(OL); 50-301/96303(OL)    i
            ,
          .. -
l  Licensee:  Wisconsin Electric Power Company I
!
Facility:  Point Beach Nuclear Plant Units 1 and 2 Location:  6610 Nuclear Road
!
Two Rivers, WI 54241 i
l  Dates:  October 7-11,1996; November 20-27,1996 Inspectors:  J. Lennartz, Chief Examiner, Rlll l      P. Cataldo, Examiner, Rlll
,
T. Guilfoil, Examiner, Sonalysts Inc.
 
l      K. Parkinson, Examiner, Sonalysts Inc.
 
Approved by:  M. Leach, Chief, Operator Licensing Branch Division of Reactor Safety
      .
:
l l
9701280303 970123 PDR V
ADOCK 05000266 PDR
_ - - - - .  - , , ___ . - . ___
 
.
.
EXECUTIVE SUMMARY Point Beach Nuclear Plant, Units 1 and 2 Examination Report 50-266-96303; 50-301-96303 Operator initiallicensing examinations were administered to five Reactor Operator (RO)
applicants and two Senior Reactor Operator (SRO) applicants during the week of October 7,1996. Additionally, an inspection was conducted from November 20-27, 1996, to assess procedure deficiencies and apparent operator training weaknesses identified during initial examination administratio Examination Results:
* Three RO license applicants and both SRO license applicants passed the examinations and will be issued licenses. Two RO license applicants failed the operating examination due to unsatisfactory performance on Administrative wb Performance Measure (JPM) tasks and were denied license Ooerations_;
* An identified error in Abnormal Operating Procedure 0.0, " Vital DC System j Malfunction," would have aligned 125 VDC Electrical Distribution System Battery 4 Chargers D-107 and D-108 outside design capabilities which could have resulted in damage to required DC loads. This was considered a violation of 10 CFR 50, Appendix B, Criterion V. (Section 03.1)
* Two applicants failed the examination due to unsatisfactory performance while two i other applicants demonstrated weaknesses on the Administrative JPM tasks I administere Additionally, unsatisfactory performance was demonstrated by at least one applicant on five of the six different Administrative JPM tasks administered which included: completing portions of the criticality checklist; proper method to independently verify valve position; and determination of Main Turbine loading limitation The demonstrated weaknesses illustrated an apparent lack of attention to detail, lack of procedural knowledge, and a failure to follow procedures for the Administrative JPM tasks administered. The applicants' inability to perform the Administrative JPM tasks administered on this examination was considered a weakness. (Section 05.3)
* Operator training regarding the operating limitations and differences between the 125 VDC system battery chargers was considered a weakness. (Section 05.5)
 
.
.
EXECUTIVE SUMMARY Point Beach Nuclear Plant, Units 1 and 2 Examination Report 50-266-96303; 50-301-96303 Operator initial licensing examinations were administered to five Reactor Operator (RO)
applicants and two Senior Reactor Operator (SRO) applicants during the week of October 7,1996. Additionally, an inspection was conducted from November 20-27, 1996, to assess procedure deficiencies and apparent operator training weaknesses identified during initial examination administratio Examination Results:
e Three RO license applicants and both SRO license applicants passed the examinations and will be issued licenses. Two RO license applicants failed the operating examination due to unsatisfactory performance on Administrative Job Performance Measure (JPM) tasks and were denied license Ooerations:
e An identified error in Abnormal Operating Procedure 0.0, " Vital DC System Malfunction," would have aligned 125 VDC Electrical Distribution System Battery Chargers D-107 and D-108 outside design capabilities which could have resulted in ,
damage to required DC loads. This was considered a violation of 10 CFR 50, l Appendix B, Criterion V. (Section 03.1)
l e Two applicants failed the examination due to unsatisfactory performance while two other applicants demonstrated weaknesses on the Administrative JPM tasks I administere Additionally, unsatisfactory performance was demonstrated by at least one applicant on five of the six different Administrative JPM tasks administered which included: completing portions of the criticality checklist; proper method to independently verify valve position; and determination of Main Turbine loading limitation The demonstrated weaknesses illustrated an apparent lack of attention to detail, lack of procedural knowledge, and a failure to follow procedures for the Administrative JPM tasks administered. The applicants' inability to perform the Administrative JPM tasks administered on this examination was considered a weakness. (Section 05.3)
e Operator training regarding the operating limitations and differences between the 125 VDC system battery chargers was considered a weakness. (Section 05.5)
 
- -. .-. - . _ - . . . . . - . . - . . ~ ~ - .
.
.
Reoort Details 1. Operations  j i
03 Operations Procedures and Documentation l
03.1 Ooerations Procedure Deficiencies    I
        !
] Insoection Scooe An inspection was conducted from November 20-27,1996, to assess procedure  l deficiencies identified during initial examination administratio I l
i Observations and Findinas    !
-
I The following procedure deficiencies were identified:  i The examiners identified that Abnormal Operating Procedure (AOP) 0.0,
  " Vital DC System Malfunction," Revision 9, Attachment B, "DC Distribution
;  Panel Power Supplies," would have aligned battery chargers D-107 and D-j  108 to DC busses D03 and D04 respectively without the related battery also connected.
 
I However, battery chargers D-107 and D-108 were designed such that they I
.
could not function properly if connected to a DC bus without a battery also j
 
connected. Additionally, if battery chargers D-107 and D-108 were aligned f  to a bus without the related battery, damage could occur to the DC loads connected to the bus. (This issue is discussed further in Section 05.5 of l
:  this report).
 
i l  The identified error in AOP-0.0 would have aligned 125 VDC system battery
 
chargers D-107 and D-108 outside design capabilities which could have
,
resulted in damage to related DC loads. This was considered a violation of j  10 CFR 50, Appendix B, Criterion V.
 
i The licensee took prompt corrective actions and issued a temporary procedure charge to AOP 0.0 the same day that the NRC identified the issue. A caution which indicated that battery chargers D-107, D-108 and D-
'
4  109 could not be aligned to a DC bus independent of a battery was added i
before the steps which restored power to bus DO3 and bus D04 in AOP 0.0, i  Attachment B.
: The examiners identified that Annunciator Response Books (ARB) 1CO41 A 3-5, Revision 3, dated August 4,1995, Section 7.0, and 2C04 2A 3-7, Revision 3, dated March 6,1995, Section 7.0, incorrectly identified the power range overpower rod stop logic as two out of four (2/4). The actual logic was one out of four (1/4) power range instruments reading greater than setpoin .
-
      \
The incorrectly stated logic confused one RO license applicant while performing a JPM to remove a failed power range nuclear instrument from j service. The applicant reviewed the annunciators that were lit for the task conditions prior to performing the task and then indicated to the examiner that annunciator 2CO4 2A 3-7, " Power Range Overpower Rod Stop" should not be lit since only one power range instrument was failed and the rod stop was a two out four logic. The applicant then referenced ARB 2C04 2A 3-7 to confirm that the rod stop was a two out of four logic. The examiner acknowledged the applicant's concern regarding an incorrect annunciator being lit and then prompted the applicant to continue with task performanc After the examiner identified the procedure error to the licensee, prompt I corrective action was taken to issue a procedure change request (ARB 96-003) to revise the AR l Conclusions    l The examiners concluded the following regarding the identifie i procedure deficiencies    ,
e The identified error in AOP-0.0 was considered a violation of 10 CFR 50, Appendix B, Criterion e The identified error in the ARB could result in some operator confusion, as I observed during examination administration, but would not result in any I inappropriate actions being taken and therefore, the safety consequences were mino l l
e The licensee had taken prompt corrective actions for the procedure deficiencies which were considered adequat Operator Training and Qualification I
05.1 General Comments
,
Operator initial licensing examinations were administered at the Point Beach Nuclear Plant during the week of October 7,1996, in accordance with NUREG-1021, Revision 7, " Operator Licensing Examiner Standards," to five Reactor Operator (RO)
applicants and two Senior Reactor Operator (SRO) applicant !
05.2 Examination Prenaration and Validation The NRC prepared the examination in accordance with NUREG 1021 guideline The examination material was validated at the Point Beach Nuclear Plant during the week of September 23,1996, with licensee training and operations personnel
- assigned to the examination team. Licensee support during examination development and validation was considered goo .
05.3 Examination Administration Dynamic Simulator Examination The examiners observed the following regarding the license applicants' performance during the dynamic simulator examination:
e The RO applicants routinely identified Technical Specification (T/S) entry conditions and were able to quickly reference and correctly apply applicable T/Ss when require e No common performanco deficiencies were observe The examiners observed the following regarding dynamic simulator examination administration:
e A simulator operatcr inserted the incorrect IC (initial conditions) set at the start of one dynamic examination. An experimental IC set was inserted that established the same general plant conditions, including reactor power level, as the IC set validated for the examination. However, the experimental IC set had a positive moderator temperature coefficient vice a negative l coefficient which the scenario was validated wit l This examination had been in progress for approximately 15 minutes when the simulator operator identified the error to the Chief Examiner. The examination was temporarily delayed at that point so the IC set that had l been previously validated for use on the examination was established. This l
delayed examination administration approximately 30 minutes. The I applicants were required to exit the simulator and wait for the correct setup to be established which added unnecessary stress to the applicants, JPM Walkthrouah Examination Unsatisfactor/ performance was demonstrated by at least one applicant on the following five Administrative JPM tasks:
e " Complete Selected Portions of the Criticality Checklist CL-1 A" e " Determine Main Turbine Loading Limitations for a Turbine Hot Plant Startup"    i e " Conduct a Valve Lineup Verification Check of Various Valves"  '
e " Conduct Personnel Monitoring For Radioactive Contamination" e " Perform Activation of the Communications Area in the Operations Support Center" Conclusions The examiners concluded that the applicants were not adequately prepared for the Administrative JPM tasks. The demonstrated weaknesses on the Administrative JPM tasks illustrated an apparent lack of attention to detail, lack of procedural knowledge, and a failure to follow procedures. The applicants difficulty in
 
.
l
.
performing the Administrative JPM tasks administered for this examination was ;
considered a weakness. Additionally, the examiners concluded that the license !
applicants were well prepared for the dynamic simulator examination and that the l ROs' knowledge of and ability to use technical specifications was goo i i
05.4 Post Examination Activities Written Examination The licensee submitted two written examination post review comments for l consideration and the NRC's post examination review identified 13 written I examination questions where a majority of the applicants failed to provide the correct response. The licensee's submitted comments with the NRC's resolution as
; well as the NRC's post review results were documented in Enclosure 2, " Post i Examination Comments and Review."
 
The examiners identified an additional generic knowledge weakness regarding the )
Rod Control system response to a failed high Power Range Nuclear trotrument '
(PRNI) (RO question 015; SRO question 014). Knowledge regarding the " Power j Range Overpower Rod Stop" logic of one out of four (1/4) PRNis reading 105% '
was needed to answer this question correctly. Five out of the seven applicants answered this question incorrectly. This knowledge weakness appeared related to the identified procedure error (Section 03.1.b.2) where the ARB incorrectly identified the power range overpower rod stop logic as two out of four (2/4).
 
The examiners reviewed the training material regarding the power range overpower :
rod stop logic in Training Handbook (TRHB),13.1, " Nuclear Instrumentation System," Revision 3, January 3,1992, and determined that the correct logic (1/4)
was referenced, Conclusions The examiners concluded, based on training material review, that the correct logic for the power range overpower rod stop was being taught. However, the ARB procedure error may have contributed to the knowledge weaknes .5 125 VDC Electrical Distribution System Insoection Scooe An inspection was conducted from November 20-27,1996, to assess an apparent operator training weakness identified during NRC post examination review of a JPM question regarding the 125 VDC station battery operability requirement Observations The safety-related 125 VDC electrical distribution system included four main DC distribution busses (D01, D02, D03, and D04), five station batteries (D-05, D-06, D-105, D-106 and D-305) and six battery chargers (D-07, D-08, D-09, D-107, D-108, and D-109).
 
_ __
.
l i
*
 
      ,
A JPM question required the applicants to determine the applicable T/S actions for the following set of conditions regarding the 125 VDC electrical distribution j system:      '
e Both units at 100% power l
e Safety-related station battery D-105 had been out of service (OOS) for 4 '
hours and safety-related station battery D-305 (swing battery) was also OO e      l Remainder of DC electrical distribution system was in a normal at power line u .
e Battery charger D-108 subsequently fail l Battery D-105 would normally be connected to DC distribution bus D03. Swing battery D-305, if operable, could also have been aligned to DC distribution Bus D03. However, the above conditions resulted in only battery charger D-107 aligned to DC distribution Bus D03 and only battery D-106 on bus D04 while all other DC l distribution busses had a related station battery and battery charger aligned to the l bu I Five applicants stated that battery charger D-109 had to be connected to DC bus D-04 within 2 hours per T/S 15.3.7.B.1.1 which stated "one of the four connected battery chargers may be inoperable for a period not to exceed 2 hours."
 
Additionally, the five applicants indicated that after battery charger D-109 was ,
connected to bus D-04 the plant could remain at power for 20 hours without having to take further actions per T/S 15.3.7.B.1.i which stated "one of the four connected safety-related station batteries may be inoperable for a period not exceeding 24 hours provided four battery chargers remain operable with one charger carrying the DC loads of each main DC distribution bus."
 
Those five applicants provided the expected response as determined by the licensee members assigned to the examination validation team. Additionally, the T/Ss implied that this condition was allowe However, two applicants stated that battery charger D-107 could not function if connected to bus D03 without a battery also connected, due to the charger's o.ssign. Therefore bus D03 did not have any power which would require a plant shutdown per T/S 15.3.0.B which stated "in the event a limiting condition for operation (LCO) cannot be satisfied because of equipment failures or limitations beyond those specified in the permissible conditions of the LCO, action shall be initiated within one hour to place the affected unit in hot shutdown within seven hours of entering this LCO."
 
The examiners determined that battery chargers D-107, D-108, and D-109 could not function properly if connected to their related DC distribution bus without a battery also connected due to the chargers' design. Battery chargers D-107, D-108, and D-109 could supply respective DC loads while maintaining the batteries at full charge as well as recharge a partially discharged battery while carrying normal loads which was in accordance with requirements described in the Final Safety Analysis Report (FSAR). However, if D-107, D-108, or D-109 were aligned to a
 
_ . ~  _ - - . . - - - _ _  - - . .
.
d      1
.
bus without a battery, damage could occur to DC loads connected to the bus. This limitation did not exist for battery chargers D-07, D-08, and D-0 The licensee wrote a condition report (CR 96 1246) regarding this issue and i summarized the battery charger differences and limitations in the Operations l Notebook which was used to inform the operating crews via required readin Further, the licensee stated that a T/S amendment was not necessary; however, I T/S 15.3.7 basis would be amended by May 1,1997, to clarify the T/ '
The examiners reviewed training material pertaining to the 125 VDC Electrical Distribution System which included the System Description (TRHB 12.7, Revision i 4), Electrical Distribution Integrated Operations Lesson Plans (LP2440, Revision 0; j LPO121, Revision 10), and the DC System Operation Lesson Plan (LP2449, Revision O). The differences between battery chargers D-107, D-108, D-109 and
: D-07, D-08, D-09, regarding their ability to function as designed if connected to a l
. DC bus without the related battery also connected was not identified in the training
.
material.
 
'
The examiners queried a few licensed operators regarding the battery charger limitations and differences. The operators questioned had varying degrees of knowledge regarding this issue. Some operators knew that battery chargers D-107, l D-108, and D-109 could not function if connected to a DC bus without a connected '
battery while other operators only indicated that they knew the chargers were designed differently and could not provide any specific I Conclusions
,
The licensee took prompt action to identify this issue to the operating crews via required reading. However, based on responses provided by operators questioned,
; the level of knowledge regarding this issue varied. Clarification of T/S 15.3.7 basis i
was necessary to remove the implication that T/Ss allcwed aligning the 125 VDC system outside design capabilities which the licensee stated would be completed by May 1,1997. Additionally, the examiners concluded that operator training regarding the operating limitations and differences between the 125 VDC system battery chargers was a weakness.
 
- 05.6 Simulator Fidelity The Examiners observed some simulator modeling deficiencies during examination administration which are documented in Enclosure 3, " Simulation Facility Report."
 
Two deficiencies identified had minor impact on examination administration (the inability to increase Emergency Diesel Generator KVARs, Enclosure 3, Item #3; and the inability to manually control "A" feedwater regulating valve, Enclosure 3, item
  #5). Additionally, one license applicant was confused by an indicated thermocouple tilt on the plant computer during a rod position indication malfunction (Enclosure 3, item #1) and incorrectly diagnosed the event as a dropped ro ,
 
.._ _ __ -- . .______ ___. .__ _ _- _.. . . . _ . . . .
'
i
,
.
'
The examincts concluded that the identified simulator deficiencies did not preclude
}  completion of valid evaluations of license applicant performance and that the impact on simulator training would be minimal.
 
L Manaaement Meetinas i
: X1 Exit Meeting Summary The examiners conducted an exit meeting with members of licensee management i  on October 11,1996, and on November 27,1996. The licensee acknowledged the j  findings presented and indicated that the materials reviewed were not considered
:  proprietary.
 
d
        !
:I i
i        l 1        !
+
l I
h.
 
$
i
!
1        1 e        1 I
 
I d
b I
.
e
 
j i
 
a
:
___
 
.- _ _ . . . . _ _ _ _ . . - - _ _ . _ _ _ . _ _ _ _ . _ _ _ _ . _ _ __ _  _ . _ . _ _ _ . . _ . . _ _ _ . _ _ _ . . __ .__ _
.
:
!'
-
i, PARTIAL LIST OF PERSONS CONTACTED i
Licensee j  S. Patulski, Site Vice President j  C. Cerovac, Assistant Operations Manager, Programs
.'
C. Gray, Assistarn Operations Manager K. Grote, initial Program Administrator
;  T. Guay, Regulatory Services Manager
:  R. Harper, Training Performance Shift Superintendent
:  R. Seizert, Training Manager l  P. Smith, Acting Operations Training Coordinator j  T. Staskal, Operations Manager
!
l  NRC A. McMurtray, Senior Resident inspector
 
l
 
i i
 
;          s q :.
g f4 i
 
4 i
!
i a
 
h
$
 
4
<
i
<
 
l
!
i
:
t
!
i s
!      10
;
;
      .,. ,    .
 
. _ . - ~ . - . - - . - . . -. - _ . - . - - . _ .. . . - . .
'
 
.
Enclosure 2 l
4  Post Examination Comrr'ents and Review
 
l Licensee Submitted Comments and NRC Resolution:
;  The licensee's written examination post review resulted in the following two
:  questions submitted for NRC review:
i
' QUESTION (RO EXAM 044/SRO EXAM 043):
i Given the following Unit 2 plant conditions:  I
  - Condenser vacuum is 25 inches H Generator load is 250 M Which one of the following is the MAXIMUM amount of time the turbine can l be operated (Operational Back Pressure region curve is attached)?
        '
a. Operation is prohibited, b.10 minute c.1 hou d. Operation is unrestricte .
ANSWER: %
REFERENCE: AOP-5A, Rev,4, pg.1; K/A 000051 A202 (3.9/4.1)
LICENSEE COMMENT:
The question was based on the appropriate use of the figure (provided as an attachment) from AOP-5A and the recognition that operation in the " avoid operation" region should be limited to no more than 10 minutes (discussion item 2.6 on page 1 of AOP-5A).
 
The given condenser information of 25 inches Hg (or 5 inches Hg absolute)
is also addressed in step 6.1 of AOP-SA which required reactor and turbine trip if condenser pressure lowers to 5 inches Hg absolute. This corresponds to answer "a" for this questio We recommend that both "a" and "b" be accepted for this questio l NRC RESOLUTION:
Licensee comment partially accepted. Based on the technical information provided there was only one correct answer for the question. A condenser vacuum of 25 inches Hg (5 inches absolute) required a reactor trip and turbine trip per AOP-5A and therefore, operation in the " avoid operation" region was prohibited. The answer key was changed to accept choice "a" as the correct respons >
 
.
. QUESTION (RO EXAM 029/SRO EXAM 028):
Which one of the following is the reason for the 30 second time delay associated with the AMSAC actuation relay?
i a. Permits operator override in the event of spurious actuatio b. Prevents spurious actuation caused by S/G shrink and swell, c. Allows time for reactor power to decrease below 40% before actuatio d. Allows time for the AFW pumps to recover S/G levels before actuatio ANSWER: REFERENCE: TRHB 13.3, Rev,4, pg. 37; K/A 012000K406 (3.2/3.5)
LICENSEE COMMENT:
The word " override" used in the correct answer implies a blocking or defeating of the AMSAC circuitry which would not be an appropriate response to an AMSAC actuation signal. The TRHB used as a reference for this question states that the 30 second delay is to " allow the operator time to correct the problem" such as starting the other feed pump, manually opening a feed reg valve, etc. None of the remaining distractors provide a correct response. We recommend that this question be thrown out due to there being no correct answe NRC RESOLUTION:
Concur with recommendation. This question was deleted from the examinatio B. NRC Post Written Examination Review:
The NRC's post examination review identified the following questions where at least 50% of the applicants answered the question incorrectly. These were considered generic weaknesses and are being provided to the Point Beach training staff for consideration and implementation into their SAT based program:
#
Question #  Descriotion of Knowledae Weakness RO 015; SRO 014 Control rod response in automatic following failure of a power range nuclear instrumen RO 030; SRO 029 RVLIS (Reactor Vessel Level Indication System) readings that would indicate the highest probability of core voidin RO 039; SRO 038 Symptoms of a stuck control rod requiring entry into AOP-6B,
  " Stuck Rod or Malfunctioning Position Indication."
 
      /
 
  . . . - ~ . . . - . . - .. - . . . . . , - . - . . - -
        .
.-    ,
.
RO 039; SRO 038 Required method to restore Reactor Coolant Pump (RCP) seal cooling following a station blackou RO 045; SRO 044 Plant status concerns that the Operators would be confronted  I with following a station blackout in excess of seven hour !
RO 050; SRO 049 Conditions that allow securing RCPs per CSP-C.1, " Response To inadequate Core Cooling."
 
RO 051; SRO 050 The immediate actions of EOP-0, " Reactor Trip or Safety injection," following the Main Turbine's failure to trip automatically and manuall RO 070; SRO 069 The preferred method to rapidly reduce turbine load in response to a casualty per the " Managers Expectations."
 
RO 077; SRO 076 The minimum Condensate Storage Tank (CST) volume required by T/Ss when only one CST is supplying both unit RO 079 The equipment that is reset by the rod control step counter reset switc RO 086 The FSAR design ratings for the station batterie RO 094 The pressurizer pressure control system response to a pressure detector failur SRO 095 The conditions when EOP-0.0, "Rodiagnosis," could be used based on operator judgemen I
 
3
_ - -
        ]
 
.- ~ - - - - . - - _ _ . . - - _ - _- - . - . - - . ... . . . . --
.
:
.
O
 
l      Enclosure 3
,
SIMULATION FACILITY REPORT
!  Facility Licensee: Point Beach Nuclear Plant i
Facility Licensee Docket Nos: 50-266; 50 301 i
Operating Tests Administered On: October 8-11,1996
:  This form is to be used only to report observations. These observations do not constitute
,  audit or inspection findings and are not, without further verification and review, indicative
'
of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC i  certification or approval of the simulation facility other than to provide information that l  may be used in future evaluations. No licensee act.. .., required in response to these
 
observations, i
j  While conducting the simulator portion of the operating tests, the following items were
;  observed:
i
,
ITEM    DESCRIPTION
!
l  1. Plant Process Computer System A thermocouple tilt was incorrectly indicated
!    during an individual rod position indication
 
malfunction for rod D-10. This was NRC j    identified. The licensee generated Simulator
*
'
Discrepancy Report (SDR) 96-0111 to track and troubleshoot.
 
!  2. Rod Control Master Counter The rod control master counter reading in the simulator booth indicated 556 while actual l    reading should have been 543 following 4    simulator reset. This was NRC identified. The licensee generated SDR 96-0112 to track and 4    troubleshoot.
:
.
3. Emergency Diesel Generator Technical Specification Test 81, " Emergency Diesel Generator G-01 Monthly," step 4.24 i    required G01 load at 2500 KW and KVARs at
 
1875-2025 for the one hour run. However, 1    KVARs could not be raised greater than 165 This was NRC identified. The licensee generated
'l    SDR 96-0113 to track and troubleshoot.
:
1  4. Plant Process Computer  The Plant Process Computer did not reinitialize i    several times following simulator reset. The
:    licensee had previously identified this problem
]    and troubleshooting efforts were being tracked j    by SDRs 96-0063 and 96-008 l
 
l l
 
,    ,    __
 
      . . - . . - - . -
e e
5. "A" Feedwater Regulating Valve Manual control of "A" Feedwater Regulating Valve should have been functional during an "A" steam generator controlling feedflow channel malfunction; however,it was not. This resulted in an unplanned reactor trip. The licensee had previously identified this but could not get the problem to repeat. The licensee reopened a  j previously closed Simulator Fidelity Report (SFR)  l
  #1121 to track and troubleshoo . Plasma Display  The Plasma Display tracked contrcl rod  ;
movement but incorrectly indicated rod position  !
6 steps different than actual. The licensee had previously identified this problem and tre 2bleshooting efforts were being tracked by SDR 96-008 . . . - . . . . . . - -.. . . - - . - .. ...~ . - -. ~. . - . - . - - - - - . . . - - . - . . - .
 
1 U. S. NUCLEAR' REGULATORY COMMISSION SITE SPECIFIC EXAMINATION REACTOR OPERATOR LICENSE REGION 3
 
,. ,  CANDIDATE'S NAME:  MASTER EXAMINATION FACILITY:    Point Beach 1 & 2 REACTOR TYPE:  PWR-WEC2
 
DATE ADMINISTERED:  96/10/07  I INSTRUCTIONS TO CANDIDATE:
Usa.the answer sheets provided to document your answers. Staple this cover shsat on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at lecst 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE  %
44. 6e hd=0 0 0. 0 0 -    % TOTALS FINAL GRADE All work done on this examination is my ow I have neither given nor i received ai I Candidate's Signature
        ,
 
_ . . _. - ._  . _ _ _ . . . _ . _. _ _ .
:
,
PAGES 2, 3, AND 4, WERE USED AS ANSWER SHEETS
,
AND ARE NOT INCLUDED IN THIS MASTER EXAMINATION COP , .
 
*
.
l l
l l
l
 
_ _
Page 5 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
1. Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examinatio This must be done after you complete the examinatio . RSstroom trips are to be limited and only one applicant at a time may leave. You must avoid all contacts with anyone outside the examination i
room to avoid even the appearance or possibility of cheatin ,
4. Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provide USE ONT.Y THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
 
7. Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question pag . Use abbreviations only if they are commonly used in facility literatur Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it ou . The point value for each question is indicated in parentheses after the questio . Show all calculations, methods, or assumptions used to obtain an answer to
 
any short answer questions.
 
, 11. Partial credit may be given except on multiple choice questions. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . Proportional grading will be applied. Any additional wrong information that is provided may count against yo For example, if a question is worth one point and asks for four responses, each of which is worth 0.25
) points, and you give five responses, each of your responses will be worth 0.20 point If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answer . If the intent of a question is unclear, ask questions of the examiner onl . - . - - _ _ _
Page 6 14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition, turn in all scrap paper.
 
15. Ensure all information you wish to have evaluated as part of your answer is on your answer sheet. Scrap paper will be disposed of immediately following the examination.
 
~16. To pass #the examination. you must achieve a grade of 80% or greater.
 
17. There is a time limit of four (4) hours for completion of the examination.
 
18. When you are done and have turned in your examination, leave the examination i Grea (EXAMINER.WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoke .
      !
i l
 
      ;
i l
 
J
 
  . . . _ -- ~ - - -. . -
e i . REACTOR OPERATOR    Page 7
.
QUESTION: 001 (1.00)
Given the following Unit 2 plant conditions:
- A large break LOCA has occurre St t=0, an SI signal was generate '
- At t=30 seconds, a containment spray signal was generate Which ONE of the following describes the times at which containment 4 spray components will operate?
a. t=0 sec: spray pumps A and B start
~; t=30 sec: spray discharge valves 860A, B, C, D open t=1 min and 30 sec: NaOH addition valves 836A and B open i t*0 sec: spray discharge valves 860A, B, C, D open t=30 sec: spray pumps A and B start t=1 min and 30 sec: NaOH addition valves 836A and B open t=30 sec: spray discharge valves 860A, B, C, D open t=40 sec: spray pumps A and B start t=2 min and 30 sec: NaOH addition valves 836A and B open d. t=30 sec: spray pumps A and B start t=40 sec: spray discharge valves 860A, B, C, D open t=2 min and 30 sec: NaOH addition valves 836A and B open QUESTION: 002 (1.00)
Which ONE of the following describes the purpose of the back draft dampers installed in the Containment Air Recirculation System?
a. Prevent unit air backflow when the accident fan is running and the cooling fan is not, Prevent backflow in a cooling unit in the event of fire in Containmen c. Serve as a system air backflow damper in idle cooling units (both accident and cooling fans secured).
 
c. Serve as explosion dampers preventing duct work collapse during an acciden .- - =. _ - __ .- . - . .. .
REACTOR OPERATOR    Page 8 QUESTION: 003 (1.00)
Which ONE of the following explains why the steam generator level program is reduced at low power?
a. To minimize time delays in plant transient response due to
'" thermal lag",
-
b. To prevent thermal stratification above the U-tube c. To reduce the mass inventory available to boil off in the event of a steam brea d. To prevent low power level o.scillation due to level dominant control.
 
.
,
QUESTION: 004 (1.00)
Which ONE of the following is the final location for the Unit 1 Control Operator after leaving the control room during AOP-10A, " Control Room Inaccessibility", when the control room is inaccessible due to a fire?
a. Auxiliary Feed Pump Room l b. 4160 Vital Switchgear Room c. PAB Elevation 8' Emergency Diesel Generator Room T
QUESTION: 005 (1.00)
i Which ONE of the following would result if the reactor head to vessel inner "O" ring seal completely fails?
, Pressurizer Relief Tank level would increase.
 
! SI on high containment pressur c. SI on low PZR pressur Reactor Coolant Drain Tank level would increase.
 
.
v
 
. . . ._ .. . _ . - . . . _ . _ _ ..-.._ _..- _. _ .. _ _ __ ___ ._. _ . . . . . . _ . - - _ . -
 
4 REACTOR OPERATOR    Page  9 i
i j QUESTION: 006 (1.00)      l Givsn the following Unit 1 plant conditions:    l
.
l - Tavg is 552 deg F.
 
#
- Tref is 550 deg F.
 
! - Rod control is in' Automatic.
: - Control bank D is stepping OUT.
 
1        i Which ONE of the following describes the required IMMEDIATE ACTION (S)?
I a. Manually trip the reactor and perform EOP-0, Reactor Trip or i  Safety Injection immediate actions.
 
l'
b. Select Bank D rods and verify rod control is operable by driving Bank D in and ou c. Place the control bank selector switch to Manual.
 
.
! d. Maintain rods in automatic because the rod motion is in response to the Tavg/ Tref mismatch.
 
.
 
1
'
QUESTION: 007 (1.00)
Which ONE of the following conditions would require a reactor trip during a Loss of Component Cooling Water (CCW) ., according to AOP-9B,
" Loss of Component Cooling"?
a. CCW low flow alarms on the SI pump b. A CCW high radiation monitor alarm is receive ]
l c. CCW pump discharge low pressure alarm is receive d. Unable to maintain CCW Surge Tank leve .~  ,.
 
JREACTOR OPERATOR    Page 10 QUESTION: 008 (1.00)
According to 10 CFR 20, which ONE of the following is the definition of Total Effective Dose Equivalent (TEDE)?
a. It is the sum of the Deep Dose Equivalent (DDE) and the iCommitted Effective Dose Equivalent (CEDE).
 
b. It is the sum of the Deep Dose Equivalent (DDE) and the Committed Dose Equivalent (CDE).
 
c. It is the sum of the Shallow Dose Equivalent, Whole Body (SDE, WB) and the Committed Effective Dose Equivalent (CEDE). It is the sum of the Shallow Dose Equivalent, Max Extremity (SDE, ME) and the Deep Dose Equivalent (DDE).
 
QUESTION: 009 (1.00)
Given the following Unit 1 plant conditions:
- Unit 1 is in Cold. Shutdow RCP 1A motor is uncoupled from the pum RCS is filled and vente Maintenance is inspecting RCP 1A seal Which ONE of the following minimizes leakage of reactor coolant upward clong the RCP shaft?
a. The pump shaft mates with the thermal barrier casin ,
b. Nozzle dam installation prevents RCS water from entering the RCP shaft are c. Seal injection is maintained during this conditio d. Seal leakoff collects any RCS leakage up the shaft and directs it back to the VC REACTOR OPERATOR    Page 11 QUESTION: 010 (1.00)
Which'ONE of the following is the ALTERNATE power supply for the 2 P2A Ch9rging Pump?
a. B08
 
;
b. B09- B03 d. 2B03 QUESTION: 011- (1.00)
Which ONE of the following actuations WILL be caused by an AUTOMATIC SI, but WILL NOT be caused by a manual-SI?
a. Containment Spray b. Containment Isolation c. Closure of emergency diesel generator output breakers onto safeguards busses d. Trip of main feedwater pumps QUESTION: 012 (1.00)
Which ONE of the following safety injection actuation signals would be automatically unblocked if pressurizer pressure increased to 1800 psig Gfter safety injection had been manually blocked?
I'
a. Low pressurizer pressure onl b. Low pressurizer pressure and steamline low pressure onl :
c. Low pressurizer pressure, steamline low pressure, and containment high pressure onl d. Low pressurizer pressure, steamline low pressure, containment high pressure, and manua REACTOR OPERATOR    Page 12 QUESTION: 013 (1.00)
Which ONE of the following curves (shown on the following page) is indicative of an under compensated Intermediate Range. Nuclear Instrument?
a.1 Curve I b.. Curve II c. Curve III d. Curve IV QUESTION: 014' (1.00)
Which ONE of the following describes why the Source Range detectors are located at the lower quarter of the core rather than at core centerline?
a. The instrument tubes are angled in toward_the core, so this location places the detectors closer to the flux they are to detec b. Neutron flux is greater in the bottom half of the core during a-startu c. To allow a spare detector to be installed in the upper half of the instrument should the installed detector fai d. This position provides optimal cooling for the detector by natural circulation of air in the detector wel m A ,.4 a A.-,.J *..a d..1--_.,J A ah...J@_J-_J5 ._a u __4_4 ..___L., _m . A J - .e.- mJ m_2 ,a 4 l
.
l
,          1 i
i 10 4 1
'
i l
10 5 ELECTRICALLY ADJUSTED COMPENSATED ION CHAMBER G
e 104 E
TYPICAL SHUTDOWN CURVES
$ 10 4 E
o W
m m 108 Y
m
 
H
$ 104
; E
'
o U      / IV l
$ 1010          l lil i N 10 11      \  -
      \  NEUTRON SOURCE LEVEL ll N N
3 O 2
  .
 
  . .
6 8
    .  . . \ . . .
10  12 14  16 18 20 22 24 TIME AFTER REACTOR SHUTDOWN (MINUTES)
TYPICAL GAMMA COMPENSATED CURVE FOR A COMPENSATED lON CHAM"5R
  .
l l
l
'
.
I
{
 
  . . _  - _ . _ _ . __ ._ .
!
REACTOR OPERATOR    Page ~_3 QUESTION: 015 (1.00)    j l
i Giv:n the following Unit 2 plant conditions:  ;
- Unit is at 90% powe Rod control in automati Power Range NI N41 fails HIGH (top of scale).
 
Which ONE of the following describes the response of the control rods?
(A cume no operator action).
 
a. Rods initially drive IN and then drive OUT to match Tavg with Tre b. Rods initially drive IN and then STOP and remain at that position, c. Rods initially drive OUT and then Drive IN to match Tavg with Tre d. Rods initially drive OUT and then STOP and remain at that positio QUESTION: 016 (1.00)
Givsn the following Unit 1 plant conditions:
- Reactor is shutdow Reactor decay heat is being removed by natural circulatio RCS pressure is 1550 psi Average core thermocouple temperature is 402 degrees Which ONE of the following describes 'he approximate amount of subcooling that exists in the RCS? deg deg j deg deg l
      ,
 
- . . . - . . . - . . . _ . - - . . - - - _ - . - . - . . . _- - - - - . -
        ,
REACTOR OPERATOR    Page 14'
i        l i QUESTION: 017 (1.00)
Which ONE of the following describes the location of the core exit 1 thermocouples?      ,
1        1 a. They are-arranged in a horizontal plane just above the uppe i
' support plat ...
        ]
! b. They are arranged in strings extending from the upper support  j
'
plate to the core mid-plan '
i c. They are arranged in a horizontal plane just above the upper
!  core plate.
 
, d. They are arranged in strings extending from the upper core plate to the core mid-plane.
 
.
,
*
:
, QUESTION: 018 (1.00)
.
Which ONE of the following conditions is the reactor vessel cooling fans  j j (cavity cooling and control rod shroud fans) used to mitigate during a  l'
l natural circulation cooldown in accordance with EOP-0.2, " Natural l Circulation Cooldown"?
i a. Creep of the reactor vessel head flange bolts.
 
I b. Formation of a steam bubble in the reactor vessel hea ;
c. Catastrophic failure of the reactor vessel flange o-ring ~
) d. Damage to the Core Exit Thermocouples's electrical circuitry
 
from overheating.
 
;
k
    .
      --
      - .-
 
_ _
REACTOR OPERATOR    Page 15
      ,
 
      )
QUESTION: 019 (1.00)    !
, Which ONE of the following conditions will result in an automatic trip i of 911 operating Condensate Pumps?  l
'
a., Steam generator feed pump suction pressure at 179 psi )
b. Containment pressure at 5.2 psi !
c. Steam generator level at 73%. l d. Condenser hotwell level at 10 inche l QUESTION: 020 (1.00)
hhich ONE of the following conditions must be met in order to reopen a bypass feed regulating valve from the control room after a valid automatic isolation from Hi-Hi S/G level has occurred?
a. Momentarily depressing the feedwater control valve bypass reset pushbutton (FWCV BYPASS) once the closure signal has been l cleare l l
b. Locally reset the closing solenoid valves once the closure I signal has been cleared and rese j c. Go to manual on the feedwater control valve controller and open the valv l d. The condition which caused the isolation signal must be restored to its pre-trip valu REACTOR OPERATOR-    Page 16 QUESTION: 021 (1.00)
Which.ONE of the following components is the largest supply of gas to ths Waste Gas System during normal full power operations?
a. Volume Control Tank
\. -
b. CVCS Holdup Tank c. Pressurizer Relief Tank d. Reactor Coolant Drain Tank I
QUESTION: 022 (1.00)
Tho'high pressure sensing line on the RCS loop B flow instrument FT-414 break Which ONE of the following describes the resulting RCS flow indication? 1 a. Only one (1) RCS loop B flow instrument is affected (PT-414)
with HIGH flow indication, b. Only one (1) RCS loop B flow instrument is affected (PT-414)
with LOW flow indicatio c. All RCS loop B flow instruments are affected (PT-414, FT-415, FT-416) with HIGH flow indication,  j d. All RCS loop B flow instruments are affected (PT-414, FT-415, FT-416) with LOW flow indicatio ]
 
-REACTOR OPERATOR    Page 17 QUESTION: 023 (1.00)
Which ONE of the following is indicated if a light is ILLUMINATED on the SI/ Spray Ready Status Panel?  .
a. The component has lost AC powe ,
b..The component is.in an abnormal alignmen c. The component has lost DC control' powe d. The component is running / energize .
QUESTION: 024 (1.00)
Which ONE of'the following "A" SI Accumulator parameters needs to be corrected for the "A" SI Accumulator to be OPERABLE while the reactor is operating at 100% power?
a. Water Volume is 1130 cubic fee b. Pressure is 730 psi c. Boric acid concentration is 1910 pp d. Outlet Isolation valve is OPEN and the control switch RED position indicating light is NOT LI . - . . - . . - . . - . , . . - . . _ . - - . ~ . - . . . ... - - . . - - - ~ . - . . - , . . - . - - .
REACTOR OPERATOR      Page 18 '
s, e
l QUESTION: 025  (1.00)
i j Given the following Unit 1 plant conditions:
 
!
  -
Pressurizer pressure defeat sw, itch is in its normal position.
 
+  -
PT-449 (Yellow Channel), Pressurizer Pressure, has just failed
  ' LO li
;. Which ONE of the following describes the response of the pressurizer
 
,
pressure control system to this failure?
e
;  a. Only PORV 431C cannot operate automatically (PORV 430 operable).
 
l b. Only spray valve 431A closes (431B remains as is).
 
c. Both spray valves CLOSE full d. Both PORVs are PREVENTED from automatic operatio QUESTION: 026  (1.00)
      . .
i Which ONE of the following statements describes how plant operations are  ,
affected if Loop A RCS Wide Range Pressure instrument, PT-420, fails    '
HIGH during Low "emperature Overpressure Mitigation System operation?
a. Pressurizer PORV PCV-430 opens onl b. Pressurizer PORV PCV-430 opens and all pressurizer heaters  !
deenergize onl ,
        .
c. Pressurizer PORV PCV-430 opens and both pressurizer spray valves open onl d. Pressurizer PORV PCV-430 opens, all pressurizer heaters  i deenergize, and both pressurizer spray valves ope l
 
  . . _ _ . _ .
REACTOR OPERATOR    Page 19 QUESTION: 027 (1.00)
Giv:n the following Unit 2 plant conditions:
t      4
- Unit is at 50% powe l
- Rod control is in MANUA l
- Lo6p B cold leg temperature ';ector TE-401B fails hig No operator action is take l
      '
Which ONE of the following will be the steady-state pressurizer level?
a. 20%
b. 33% % %
,
QUESTION: 028 (1.00)
; Which ONE of the following conditions on LT-428 would result in an
; increase in indicated pressurizer level?
. a. A leak in the reference leg of the pressurizer level transmitter, LT / 2 8 .
 
b. Pressurizer liquid temperature increases.
 
! c. The reference leg for LT-428 cools down due to a decrease in containment temperatur d. Containment pressure increases to 0.3 psig; containment temperature remains constant.
 
,.
d f
I
 
l l
REACTOR OPERATOR    Page 20
    &ScjN  h I'20 QUESTION: 029- (1.00)-  [gt sM Which ONE of the following is the reason for the 3 d time delay associated with the AMSAC actuation relay?    l a. Permits operator. override in the- nt of spurious actuatio '- l      l b. Prevents spurious actuat caused by S/G shrink and swel .l l
c. Allows time for tor power to decrease below 40% before actuatio d._ Allo ime for the AFW pumps to recover S/G levels before uatio ,
QUESTION: 030 (1.00)
Which ONE of the following RVLIS readings indicates the highest probability of core voiding?
a. Wide Range reading 98 ft, with NO RCPs runnin b. Narrow Range reading 38 ft. with NO RCPs runnin c. Wide Range reading 65 ft. with ONE RCP runnin l d. Narrow Range reading 95 ft. With BOTH RCPs runnin l i
l l
l l
I
    . - - _ , , - . . -
 
REACTOR OPERATOR    Page 21 QUESTION: 031 (1.00)
Given the following Unit 1 plant conditions:
- Unit is operating at 75% steady state powe All systems are in automatic contro \The "A" S/G atmospheric valve fails ope Which ONE of the following describes the plant response to this condition? (Assume no operator action is taken.)
 
a. Turbine load decreases by 5%, reactor power remains stable at 75%.
b. Turbine governor valves open in response to lower steam header pressure to increase turbine load to 80%.
c. Control rods insert to maintain reactor power at 75%.
d. Control rods withdraw and raise reactor power to 80% where it stabilizes.
 
QUESTION: 032 (1.00)
Which ONE of the following describes the normal, emergency, and altornate emergency sources of power to Safeguards 4.16 KV bus 1A06?
Normal Emergency Alternate Emergency
______ _________ ___________________
a. Load Bus 1A04 EDG G03  EDG G04 b. Load Bus 1A04 EDG G01  EDG G02 c. Load Bus 1A03 EDG G01  EDG G04 d. Load Bus 1A03 EDG G03  EDG G02
 
l
 
-._m . _ - _ . _ . . _ . _ . _ . > _ _ . . _ _ . . _ . . . . ~ . _ . . . _ . . . _ _ _ . - _ . . _ .
    -    . . . . - . . . _ . - _ _ . _ . . _ . _ . . _ . . -
REACTOR' OPERATOR        Page 22
:
,
, QUESTION: 033  (1.00)
!
l Which ONE of the following will provide an automatic shutdown of the gas
-
turbine while operating in local' control?
l a. Low axial compressor suction pressure
  -
j t j b. Low lube oil level
!
c. Low lube oil pressure d. Low control air supply pressure QUESTION: b34  (1.00)-
Which ONE of the following is the maximum rated run time for Emergency Diesel Generator G01 loaded at 3050 KW?
 
a. 30 minute '
l i' hou ;
c. 24 hour hour ;
l QUESTION: 035  (1.00)
Which ONE of the following RMS displays must be selected to access alert and alarm setpoints?
a. Status Grid b. Sector Display c. HDSR d. Trend
 
. . . . . . . . _ _ . . -. _ _ _ _ . . _ . _ . . . . . . _ _ . . - - . . _ . _ _ . . _ . _ . . . _ ..__...__m____ _ _ _ _ . . _
REACTOR OPERATOR      Page 23 l
 
1 QUESTION: 036 (1.00)      :
Givsn the following Unit 2 plant conditions:
  - RHR is in servic RCS pressure is 320 psig and INCREASIN RCS temperature is 340 degrees F and INCREASIN ALL system lineups are in a normal shutdown configuration for-solid-plant operatio Which ONE of the following will act FIRST to prevent overpressurizing tha'RHR System?
a. Pressurizer PORVs will ope RHR return isolation valve will auto clos c. RHR pump hot leg relief valve RH-PCV-861C will ope d. RHR pumps discharge relief valve RH-PCV-861A will ope l QUESTION: 037 (1.00)
Which ONE of the following conditions will ARM the Turbine Crossover Steam Dump System?
a. Selecting MANUAL on the mode selector switch, b. AUTO selected on the mode selector switch and a 10% load drop in less than 120 second c. Going to TEST on the individual valves on the back of C0 d. Reaching 90 psig in the crossover steam header.
 
I I
l l          l
 
o i
:
!
      -
 
. _ __ . . - . . _ _ . . _ . _ _ _ _ _ . . . - . . _ _ . . _ . . _ . _ . _ _ _ .__ _. _ . _ .._.- .. _  _ _ - . _ . . _ . _ _
!
l REACTOR OPERATOR        Page 24
 
4
.
QUESTION: 038  (1.00)
i j Which ONE of the following is a symptom of a stuck control rod that would require entry into AOP-6B, " Stuck Rod Or Malfunctioning Position
; Indication", following a transient?
i i a.'An individual RPI in 8 step disagreement with the bank demand i  location.
 
I b. A variation in NIS instrumentation resulting in a quadrant tilt of 1.2%.
j -c. A variation in core outlet thermocouples of 8% relative to l  symmetric thermocouples.
 
!
! d. A> variation in axial flux of 1.2% of axial peak at any location j  relative to symmetrical trace.
 
i
:
[ QUESTION: 039  (1.00)
; Given the following Unit 1 plant conditions:
 
- A loss of all AC power has occurred.
 
'
- Power has been restored.
 
i - The crew has transitioned to ECA 0.1.
 
! - ECA 0.1 directs restoration of RCP seal cooling.
 
i a
j Which ONE of the following is the method required to restore RCP seal j cooling?
i 1 a. Seal injection flow is initiated to cool the seals to less than l  150 degrees F, then CCW flow to the RCP is established.
 
l l b. CCW flow is initiated to cool the seals to less than 150 degrees F, then seal injection flow is established.
 
l
 
! c. Seal injection flow is initiated to cool the seals to less than 190 degrees F, then CCW flow to the RCP is established.
 
l d. CCW flow is initiated to cool the seals to less than 190 degrees
{  F, then seal injection flow is established.
 
i
.
:
f
.
!
l
:
_    - . - , . , _ - - _ . . _ . _ . .  .
 
REACTOR OPERATOR    Page 25 QUESTION: 040 (1.00)
Givsn the following Unit 1 plant conditions:
- A plant trip has just occurre control rods are stuck out of the core following the tri Anremergency boration has been initiated by the reactor operator in accordance with EOP-0.1, " Reactor Trip Response."
 
Which ONE of the following lists the MINIMUM injected volume of boric ccid necessary to satisfy the required amount of.boration?
l a. 600 gallons ,200 gallons    !
-
I
      :
c. 2,400 gallons I
d. 3,000 gallons I
QUESTION: 041 (1.00)
Givan the following Unit 1 plant conditions:
- Steam generator A is faulted due to a feed line break outside of-containmen The crew is performing actions of EOP-2, " Faulted Steam Generator Isolation."
 
- The AFW system is in operatio i Which ONE of the following actions concerning the AFW pumps is required by EOP-2?
a. Shutdown the AFW Pumps immediatel l b. Maintain at least 50 gpm AFW flow to each S/G with narrow range levels less than 8%.
l c. Run the AFW Pumps only if less than 200 gpm is available to the ~
S/G Isolate the AFW Pumps from S/G A (steam and AFW flow).
:
i
      -
 
REACTOR OPERATOR    Page 26 I
l
      '
, QUESTION: 042 (1.00)
In cccordance with EOP-1, " Loss of Reactor or Secondary Coolant", which ONE of the following groups of parameters is required to be verified, in cddition to Pressurizer level, prior to terminating SI flow?
a. ' RCS subcooling, secondary heat sink, and containment pressure
 
b. Secondary heat sink, RCS pressure, and RCS subcooling c. RCS subcooling, RCS pressure, and containment radiation levels d. Secondary heat sink, containment pressure, and RCS pressure
 
.
QUESTION: 043 (1.00)
Given the following Unit 1 plant conditions:
- Unit is at 100% powe Condenser vacuum is decreasin Which ONE of the following condenser vacuum conditions will FIRST result in the loss of condenser steam dump availability?
a. 25" Hg vacuum b. 23" Hg vacuum c. 21" Hg vacuum d. 19" Hg vacuum i
,
I i
 
REACTOR OPERATOR    Page 27 QUESTION: 044 (1.00)
Given the following Unit 2 plant conditions:
- Condenser vacuum is 25 inches H Generator load is 250 M L:
Which ONE of the following is the MAXIMUM amount of time the turbine can be operated (Operational Back Pressure region curve is attached.)?
a. Operation is prohibited, minute hour, d. Operation is unrestricted.
 
QUESTION: 045 (1.00)
Which ONE of the following concerns would an operator most likely be confronted with during a total loss of AC power in excess of 7 hours?
a. Loss of secondary heat sink condition b. An unmonitored release of radioactivity c. Loss of containment integrity d. A Steam Generator overpressurization condition
 
  . - . . . - - -.  - .
_ - . _ . . - - . . - . . - _ - . . - . . - . - . - . _ - - - . - .-
l    POINT BEACH NUCLEAR. PLANT          AOP-5A
;
ABNORMAL OPERATING PROCEDURES          MAJOR l              Revision 4 l  LOSS OF CONDENSER-VACUUM          August 21,1995 ~
~
                      >
l l  @V FIGURE 1
          .
i
!,
.
j RECOMMENDED OPERATIONAL REGIONS FOR NUCLEAR UNITS
  , ' ,    WITH 2 DOUBLE FLOW LOW PRESSURE ENDS. ,  ,
                    ,
!    s  -  -
    .i
        - - - -  ~ - - -
            :__  __ =-~.
            :.__ -
    .. . . . - _
          -- : :::_ r-        -
c. : f_;: : g;
              .._~=..__=.+i. _
      ._
    :_ -  4 -
          -
_
          - _ ,  _.=  - --
_
                  -
                    .
                    : , ._f_ --__
    . . _  _ -  - _, -
          - -    - --
                  . _  ,  .
            . ,  _---  , ; , __ .; __ _ ; ,.  , _ . _ ,
    $''-  ,. _ _ ', -  . : _    ' -
_
                  '
                      ''n''
.    .
_ _ . - . , .        -
r i
          .. '            .-
}    1        -
              -
              -
    ._ 1 - f
_
_- _ _        r
_  C__ -_ _
s  ,= c
                -
T
_E : ' -_ r_ ::.  -... . . . 'r            1 _ iz . : --
        -
              - i__..--_    . _ ,    --.  - _
            -
              <
ie t      ,-    -
_
                  -- -__ .1 _ .-=-._  ___:_- _
;  W 4 = :- 7 __ - _ .. -  -
_5-  1 _-
            - '
            ---._,_'-'____-
                ,-
                  : ~ 21 __ __-
                      ---
                    ;; 3 :=
    >=
            . . - :-_'_ _ ._ i- , , , ,- , .    - ~_ ;_ -
      -
      - ._  __ ,  . . - .2        -] _ c1 F ri - : ,= -
1  3    --
        . , .  ,      -
                    -..~.=.--~_.=
r-.J: :"r.;_ - -T '
              -
!  _J  " ; ''- . :' _ :_ _ _ - 2    -
                  , '. . -. -
O      -_ _--c    >
;        y . ..          -=-~-r-  E ~ ._ /
a  01 ,-_  -  .
_=_,__._:-_=,:-_- . . .; - -
              -- ..
                - -
                  - - - -
l  C3
_
    :: .-,:_. < --  sy *
            .
                  - [_ -  - - -
                      = :r- ,
                - -: ~ _
              ~
          ,1
          ==--g - . i v ---
,  4 3 _: r: __  --
V _._.. --
i r" .. ;= - 27  --  i -:-
              ., ~' r- -
              "-
              - - _ :  .r_ -I ~.a ~;j";r;, --~
_2
                      "[J
 
2 W
    '
      --= ::_ ~ 1- =1
                  -
_
                    :- b
    ; _h
_.- Eg- -_E :
4                -
          -- -~-- 1 '.;i r e-: 3__ .
1  a- ..-.,--
      .
_ _ _r -2-- -  -
_______e  t _.=r, _. :  .. .. _ _ :
iE _-- _ =i
_ _ r --
r_-'~r:- -- _~: ; - . ~ '
_
I  D ,Z'--~._~        ~ -  212 - - :- L
                    ~~~
2,,, .E.:. :
i  -
  ,.. y *E - Jfi: r-  - " ; _ ~ - 1 _- - ~' Z r-    EE --  --  #. -~ T' "  -
E 2_ j . . Eg
                    =r_E_"~.- ,r:r :. ,: -: c_
                . .
              ,
lj gw,  w    12 -: - = e ;_ ,
          = ; =__.,s g y- r:._ .- .,.'  _,  ;,, .  : .;._ : =-;.1 35Er 35;;gggg E    _: .__.._:
              .,
    ' i _ i : ,-._  _:-          =
'
_
        ~--      _
s:__  E;_s=_resE
            - 1 :[. ,_:#~- = ~:
                    ~
_ "'d3
                      -
 
    < Ei_- _:E=r ,E 1:J;    i_ ~_~ _T~          _ _- jgEE3
-
*
a)
      =- - _ 2 2 g ,,, :_ _ ::1: -- : . , - ,__-  - _=  -_: =_ 1 ..r;..
_;------  : _ _u:-:7 _ _1 s. . _s .=- 4 : -.-.. _ . 1 - ...
              -:    :_  c =- . ;=-
    ' : , '-
              -2
_ ar : . ,- : =  : : , - .2: ==_=_ _ =
          :
          .      _
j
      . :-_:[ _r _;.--
L ::~ . . -  : '. _2
          . _r ; ; ..;, r _ - -.:,
              -
                ' : - _.- ~-i '
                  . J - ''_' 'E_.5:EE
                    -
    - - - :.-
    : _":    , ; _ , _,  ,  _ . . _  _: - - :  : _ .~: , .=; ==> __
]
;
    .
_ ..:._ ; :- .:; . . _ - ::_ .._ = .:, . . : :- - _ :: .    ;_. . . .r s; _ -.:: u;_._;;_:g._ - _-    _
3  . :,: .
      - gt _ . .. _ :.~.  :.ry ...: .. = ;:g; _ :_ _: :. j;: _. :. ...          . . . . . .- ,... :_: ;i = p: ;_: .:;;
    ,
      :
        .: : . : .- -- -
                ~
3                    ~
r_            *
g.-    ,
            ,_ . . '- r + .; . . . T. ! . . . . .......t..:. : . i. : . r . ., .:.._..... . .. _ . ;;. . .:: _:. :.' Eg
_
                      = ==
_ _ ::__
:
I 0  10 0  200  300  400  500  600  70c
,
GROSS LOAD-MW i
'                    393A421
.
l i
i a
d
:
!
>
T I
 
i
 
i i
 
i
!
  @
!
l j
          % dk J'or 6ui6        REFERENCE USE
,
i i
_ _ _ _ _ _ _ _ _  __  -  .  -    _  __ _      . . _ _ .
 
___ _ . . _ _ .. _ _
REACTOR OPERATOR    Page 28 f
QUESTION: 046 (1.00)
Given the following Unit 1 plant conditions:
- A loss of all AC power has occurre ECA-0.0, " Loss of All AC Power" is in effec Det ECA-0.0, certain Engineered Safeguards equipment control switches are placed in Pull-out.
 
J Which ONE of the following events is prevented by this switch alignment?
; a. An uncontrolled cooldown of the RCS and possible reactor restar b. An uncontrolled use of water that may be needed for long term cooldow c. An uncontrolled start of large loads on safeguards AC buses, d. An uncontrolled depressurization of the RC QUESTION: 047 (1.00)
l Which ONE of the following describes the response of Feedwater R0gulating Valves following loss of a RED or BLUE instrument bus? ,
 
a. May fail open; manual control will NOT be available to regain '
contro I b. May fail open; but manual contrcl will be available to regain control, l
c. May fail closed; manual control will NOT be available to regain I control, d. May fail closed; but manual control will be available to regain j control.
 
i i
 
,
m__ . _ . _ . = . . . . _ . . _ _ _ . _ . _ _ _ . . _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ . _ . . . _ .
          '
!
!
REACTOR OPERATOR        Page 29 l
          '
QUESTION: 048  (1.00)
'Which ONE of the following locations / equipment is protected by a Halon
. System?
a.'' Diesel' Generator rooms        I
  ,;
.b. Hydrogen Seal Oil package c. Service Water Pump area d. Auxiliary Feed Pump room
  .
QUESTION: 049  (1.00)
Technical Specifications require containment pressure to be between + psig and -2.0 psig during power operatio ~
 
Which ONE of the following is the MAXIMUM time containment pressure is'      l allowed to be outside this band per Technical Specifications before      ;
action must be initiated to shut down the plant?      i-a, 30 minutes        l b. 1 hour        i i
c. 6 hours hours l
t i
!
i
!
:
!
:
i
!
        -
        .- - -- - . . - _ .,,
 
    . - . - - . .. . - - . . . .. - ..
REACTOR OPERATOR    Page 30 QUESTION: 050 (1.00)
A LOCA has occurred and both RCPs have been started per CSP-C.1,  i
"Rasponse to Inadequate Core Cooling."
 
'
Given the,below list of criteria:
; I. Core cooling provided by low or high head SI
-
II. Narrow range reactor vessel level greater tl'an 25 feet j III. Any RCS hot leg loop less then 350 degrees 1
,
.IV. Core exit thermocouples less than 1200 degrees F
: Which ONE of the following gives the three conditions from the above
;
list that allow securing of the RCPs per CSP-C.17 a. I; II and III I, II and IV
_>
I I, III and I , II, III, and IV.
 
J QUESTION: 051 (1.00)
Given the following Unit 1 plant conditions:
- The Reactor tripped from 100% powe EOP-0, " Reactor Trip or Safety Injection" has been entere The main turbine did NOT trip as expecte MANUAL turbine trip is unsuccessfu Which ONE of the following is the NEXT action required per EOP-0?
a. Manually close MSIVs and Bypass valve b. Locally trip the main turbin c. Go to CSP-S.1, " Response to Nuclear Power Generation /ATWS."
 
d. Verify Safeguards Buses energize .  -  -
      ,
t 1      l
'
- REACTOR OPERATOR    Page 31 i
i  . l
! QUESTION: 052- (1.00)
l Which ONE of the following is the reason for the main feedwater l l icolation following a reactor. trip?
1' a. To prevent thermal shocking of the' steam generator (S/G) !
!  'tubesheet I
 
i b. To prevent excessive cooldown of the RCS.
 
i i
c. To prevent S/G overpressurizatio ;d. To preserve secondary water.for a subsequent RCS cooldown.
 
#
(
i'  .
l
! QUESTION: 053- (1.00)
-
Given the following Unit 1 plant conditions:
;
i - A reactor trip and safety. injection has occurre Pressurizer PORV 1RC-431C lifted and is stuck ope Which ONE of the following is the MAXIMUM pressure below which the PORV isolation valve must be shut?
a. 2450 psig psig c. 2225 psig psig
      !
      ,
  . , ~ . - - - , - - . - - - . - -- w
 
. . ~ . - ~ . . - .- -. - .. -- . _ . __-.- . .. -. . _. -  _ - . . - _ . _ . . -
_
,
 
7 i'
-REACTOR OPERATOR      Page 32
 
1 j QUESTION: 054 (1.00)
!
<
-Given the following Unit 1 plant conditions:
, - Unit has tripped from 100% due to a small break LOC l i
'
- Conditions have stabilized and operators are evaluating the    !
briteria for terminating SI.
 
'
'
--Adverse Containment conditions do NOT exis '
- A maximum of 50 gpm feedwater-flow is available to each S/G.
: . Which ONE of the following conditions.would PREVENT SI termination per j EOP-1.2, "Small Break LOCA Cooldown and Depressurization"?
a. Both steam generator levels ir^1cate .5% N !
 
a
; b. Pressurizer level indicates 11 I
 
j c. RCS subcooling'is 40 degrees Pressurizer pressure is 2050 psig.
 
i
:
;
j QUESTION: 055 (1.00)
l In procedure OP-5A, " Reactor Coolant Volume Control" there is a i PRECAUTION that states: "Do not secure letdown flow without also I securing charging flow ..."
i i- Which ONE of the following statements describes why charging flow should j also be isolated?.(Assume all systems are in a normal at power lineup.)
 
!.
i a a. VCT level will decrease until charging pump suction shifts to i  the RWST.
 
b. Reduce thermal shock on the Non-Regenerative Heat Exchange c. VCT level will decrease causing possible damage to the charging pumps.
 
!-
;
d. Reduce thermal shock on the charging penetration into the RCS.
 
4.
 
b
;
}
 
:
'
f
,
    -  - - c , - - --- _
 
. _ . . . . .. . _. _ _ _ _ . _ _ _ . _ _ _ . . . _ _ _ _ . . _ _ _ . . _ - , _ _ _ . _ _ _ _ _ . :
J I ' REACTOR OPERATOR      Page 33 J-f QUESTION: 056  (1.00)
j _Given the following Unit 1 plant conditions:
 
        '
i  - Unit is in Cold Shutdown with RHR Cooling in progress.
 
I  - The RCS is solid.
 
i
  - kHR flow is lost and CANNOT be restore l
  - All other systems and components are availabl l l
I Which ONE of the following methods of cooling will be utilized to remove
! the core decay heat?
*
a. Feed a S/G using an AFW pump, and bleed steam through the j  respective S/G atmospheric relief valv i i  b. Start a charging pump, with flow through an RHR heat exchanger,
:  and initiate Hot Leg Injection.
 
1  c. Feed the RCS with Safety Injection, use letdown to remove decay
,  heat.
:
;  d. Start a charging pump, with flow through an RHR heat exchanger,
>
and initiate Cold Leg Injection.
 
i
 
i QUESTION: 057  (1.00)
i Given the following Unit 1 plant conditions:
]
!  - A reactor trip occurred 30 minutes ago due to a Steam Generator
]  tube ruptur The crew is taking the actions of EOP-3, " Steam Generator Tube
'
Rupture".
 
- N-35_ indicates 1.4E-10 amps and N-36 indicates 1.6E-10 amps.
 
j  - Source Range Instruments are NOT energized.
 
! Which ONE of the following actions is required by the Reactor Operator?  ;
l
.
l  a. Allow N36 to decay to less than 1.5E-10 amps which will automatically energize the SR instrument b. Depress both SR RESET pushbutton c. Depress both IR BLOCK pushbutton d. Deenergize N3 _ _ . . . _ . _ _ _ . _ . _ _ . _ . _ _ - . _ _. .__ _- _ _ _ ____ _ _ _ _ _ _ .
i
 
REACTOR _ OPERATOR    Page 34
 
.:
'
QUESTION: 058 (1.00)
: Which ONE of the following conditions requires tripping the reactor per l AOP-1A, " Reactor Coolant Leak", for a confirmed steam generator tube i lenk? Assume appropriate actions have been taken per AOP-1A regarding chargingandletdownsystemoperation ,
a. Pressurizer level is at 11% and decreasing.
 
'
b. Narrow range S/G level is 62% and increasing.
 
!
I c. VCT level is at 17% in AUTO control and decreasing.
 
i d. 3/G Blowdown Isolation valve closes due to high radiatio ,
!
{ QUESTION: 059  (1.00)
i
* The crew is responding to a ruptured tube in 1B steam generator-(S/G)
using EOP-3, " Steam Generator Tube Rupture."
 
i i Given the following plant conditions:
;
- 1A S/G pressure is 950 psi B S/G pressure is 1050 psig.
 
l Which ONE of the following is the required core exit temperature that j the RCS must be cooled down to prior to depressurization? (Page 15 of
,
EOP-3 attached).
 
a. 505 degrees F
; b. 490 degrees F
 
*
c. 485 degrees F
-
d. 480 degrees F U.
:
.
I i
!
 
I
,
 
e  -  >,m ., , - ,
 
_ . _ _ _ _ _ _ _ __ _ _ _ . _ _ _ _ _  . . _ . _ . _ . . _ . .._ _ _ . . . . - . . _ _ . _ _ _ . . ._ _
'
d ,
t
,
POINT REACH NUCLEAR PLANT    EOP-3 Unit 1
  '~
EMERGENCY OPERATING PROCEDURES    MAJO Revision;19.
 
,
!
STEAM GENERATOR TUBE ~ RUPTURE ^    October;26, 1995..
:)
!
UNIT-1 STEP
        .
          - -
          ==.
;  ACTION / EXPECTED RESPONSE  RESPONSE NOT OBTAINED i
CAUTION:
;
iStep)  THE FOLLOWING STEP MAY REQUIRE A C00LDOWN OF MORE THAN 100 F IN DNE HOUR.
 
i NOTE:  RCP trip criteria do HQI apply af ter a RCS (Procedure) controlled cooldown has been initiated.
 
.
;
20 Initiate RCS Cooldown i
f Using ruptured steam l  generator pressure, detemine l
target"coretexit temperature.
 
.
 
)  Ruptured Steam Generator  Core Exit Pressure (psig)  Temperature (*F)
i
:
1100
    [460) 505 1000  [445] 490
;    900  [435) 480 g  800
    [420] 465
 
700
    [405) 450 600
    [390) 435 (    500
    [370] 415 400
    [350) 395 340
    [335) 380
          ,
    . . .
Step 20 Continued on Next Page .  . .
.    /
r=.
    %
1: m: :7 3 .
 
REACTOR OPERATOR    Page 35 QUESTION: 060 (1.00)
Which ONE of the following statements explains why AFW flowrate is procedurally restricted to 100 gpm when recovering steam generator (S/G)
Icvol if the level has fallen below 55 inches on the wide range indication?
,
a. To minimize water hammer to the S/G feed rin b. To prevent reactor restart from an excessive cooldow c. To minimize thermal stresses t) S/G components, d. To prevent exceeding reactor vessel cooldown rate limi l QUESTION: 061 (1.00)    l
      !
Which ONE'of the following is the basis for stopping all RCPs in CSP- j H.1, " Response to Loss of Secondary Heat Sink"?
a. It establishes natural circulation to enhance the blead and feed capability of safety injectio b. It extends the time available to restore feed flow before bleed and feed criteria is me c. It anticipates an RCS pressure decrease caused by spray valves opening when air is restored to containmen d. It anticipates an RCS pressure decrease caused by opening PZR PORVs during bleed and fee l REACTOR OPERATOR    Page 36 l-l QUESTION: 062 (1.00)'
l Loca of which ONE of the following distribution panels will result in a l DUAL PLANT TRIP?
a.,D13, b. D16 l
c. D18 i
l d. D21
,
l QUESTION: 063 (1.00)
l I With the Pressurizer Level Control Selector Switch in the NORMAL l
position, a pressurizer level instrument failure causes the following SEQUENTIAL plant event Charging flow is reduced to minimum.
 
I - Pressurizer level decrease Letdown. flow is secured and heaters turn of Pressurizer level increases until high level trip occurs.
 
l
! Which ONE of the following instrument failures could have occurred?
(Assume NO operator action)
a. Pressurizer level channel 428 (Blue) failed low Pressurizer level channel 428 (Blue) failed high c. Pressurizer level channel 427 (White) failed lo Pressurizer level channel 427 (White) failed hig __ _ _._. _ ._ _. _ _. _ . . _ . . . _ . . . . _ _ . . _ . _ . - _ . - _ . _ . _ - _ _ _ _ . . _ . . - _ _ _ . _
l REACTOR OPERATOR      Page 37
!
! QUESTION: 064 (1.00)      ;
;
'
Givsn the following Unit 2 plant conditions:      -
i
- The plant has tripped from 100% power due to a loss of off-site    !
powe The crew is required to verify natural circulation in EOP-0.1,
,
" Reactor Trip Response."
 
I i
Which ONE of the following parameters satisfies one of the criteria for i indication of natural circulation?
l
,
a. Steam generator pressures slowly trending upward.
 
,
b. RCS Cold leg temperatures are increasing.
 
!
l c. RCS subcooling is 36 degrees >
d. Core exit thermocouples at saturation temperature for steam generator pressur ,
QUESTION: 065 (1.00)      j Which ONE of the following " Procedure Usage Levels" allows performance of all activities from memory?
. Infrequent Use
! Information Use t
'
c. Refelence Use d. Continuous Use I
t
,
i I
,
l
,
l
'
      - -_
_  , _
_  -.
 
      )
REACTOR OPERATOR    Page 38 ]
l I
QUESTION: 066 (1.00)    l l
Th3 applicability statement for Procedure EOP-0, " Reactor Trip Or Safety i Injection", indicates that this procedure is used for initiating events ;
occurring where RCS hot leg temperature is greater than or equal to 350 '
d*grees $ .
Which ONE of the following describes the applicability of EOP-0 when RCS temperature is below 350 degrees F and a LOCA occurs?
a. EOP-0 cannot be used, procedure EOP-0.0, "Rediagnosis" is to be use b. EOP-0 cannot be used unless directed by the Critical Safety Function Status Trees, c. EOP-0 cannot be used unless a step-by-step evaluation is made to determine if each action is still applicabl d. EOP-0 cannot be used, Shutdown Emergency Procedures (SEPs) are to be used.
 
QUESTION: 067 (1.00)
Which ONE of the following is the MINIMUM number of shifts, per 10 CFR 55, on which you must actively perform operator or senior operator duties to maintain your license in an ACTIVE status? (ASSUME 8 hour shifts.)
 
a. 5 shifts per calendar quarter I shifts per calendar year  l
      '
l c. 7 shifts per calendar quarter 1 shifts per calendar year  l
 
REACTOR OPERATOR    Page 39 i
QUESTION: 068 (1.00)
Which ONE of the following is the LATEST time that alcohol consumption ;
would be allowed prior to assuming the watch at 1800 per NP-1.7.5, ;
      '
" Fitness for Duty Policy and Procedure"?
a.'0600    j b. 1000 l
      !
QUESTION: 069 (1.00)
A bomb threat has been received and the Duty shift Superintendent (DSS)
has initiated a plant evacuation. A suspicious device has been diccovered in an emergency diesel generator (D/G) room by the extra reactor operato Which ONE of the following states the actions to be taken by the individual discovering the1 device, in accordance with NP 1.7.10?
a. Immediately notify the DSS by two-way radio and then leave the D/G roo Immediately notify DCS by two-way radio and then leave the D/G roo c. Leave the D/G room, and then notify the DSS by phone, d. Leave the D/G room, and then notify DCS by phone, j
 
l REACTOR OPERATOR-    Page 40 i
i QUESTION: 070 (1.00)
Which ONE of the'following is the preferred method to rapidly reduce turbine load in response to a casualty per the Managers Expectations?
      !
l a. Operator Automatic Load Rate Contro \, :
b. Turbine manual using governor valve decreas c. Valve Position Limite l d. Turbine manual using Governor fas l
      \
QUESTION: 071 (1.00)
Which ONE of the following is an individual who is authorized to be a member-of the fire brigade and to perform operations health physics functions?'
a. Unit 1 Control Operator b. Unit 2 Control Operator c. Duty; Technical Advisor d. Extra Operator QUESTION: 072 (1.00)
Which ONE of the following describes the process of determining an instrument's accuracy by comparing the indication to other independent instrument channels measuring the same paramete a. Channel Calibration b. Channel Functional Test c. Channel Check d. Channel Verification l
      :
 
i REACTOR OPERATOR    Page 41
      :
i QUESTION: 073 (1.00)    ,
l
      '
An oncoming Reactor Operator has worked the following schedule:
Day: 1 2 3 4 5 6 7 8 9 10 11 12 13 14 ,
l Hrs: 12 8 12 8 8 8 8 12 8 12 12 12 12 ?
      >
 
Which ONE of the following is the MAXIMUM number of hours the individual i may work on day 14 without obtaining special authorization assuming the l operator has a minimum of 8 hours off between each shift?
a. 2 hours    ,
b. 4 hours c. 8 hours d. 12 hours QUESTION: 074 (1.00)
Which ONE of the following is correct regarding INDEPENDENT verification of control board sliders?
a. Must be done by the Unit Control Operato b. Must be done by visually observing the operating lin c. Can be accomplished using status lights, d. Can be accomplished using remote indicator .- .. ._ _
REACTOR OPERATOR-    Page 42
      !
.
,
QUESTION: 075 (1.00)
Which ONE of the following conditions is required in order to perform work on a system using " Positive Control" in lieu of danger tags?
a. A QC certified inspector is presen V :
b. Work is to be performed in a high radiation are c. Requires direct line of sight of the isolatio Isolation boundaries are locke i
      !
QUESTION: 076 (1.00)    )
 
Giv@n the following Unit 1 conditions:  ,
l
      '
.l - Unit is at 100% powe VCT level is 23%.
- All controls are in automati LT-112, VCT level transmitter, fails hig Which ONE of the following describes the FINAL ACTUAL VCT level? (Assume l no operator action.)    1
,
a. Increases to 78% and then diverts to the HU Increases to 28% where auto-makeup stop I c. Decreases to 17% where auto-makeup initiates, d. Decreases to 0% (empty).
 
'l
 
REACTOR OPERATOR    Page 43 QUESTION: 077 (1.00)
Givsn the following plant conditions:
- Both units are operatin The South Condensate Storage Tank (CST) is isolated for repair The North CST is selected for Auxiliary Feedwater Pump suction at both unit Which ONE of the following is the MINIMUM volume required in the North CST?
a. 13,000 gallons ,000 gallons ,000 gallons ,000 gallons QUESTION: 078 (1.00)
WHICH ONE of the following describes the initial output signal from the rod control system for a 12% step change in turbine load with power at 80% as compared to a 12% step change in turbine load with power at 40%?
Assume rods in automati The initial output for the step change at 80 % is: larger due to the response of the Variable gain unit, larger due to the response of the Power mismatch rate / lag uni smaller due to the response of the Variable gain uni smaller due to the response of the Power mismatch rate / lag uni .
 
_ . . . _ _ _ _ _ _ _ _ . _ . . _ _ . ._ _- . _ _ . _ _ . _ -. . .
l
' REACTOR OPERATOR    Page 44 i
QUESTION: 079 (1.00)
Which ONE of the following groups of equipment will be reset by the rod control step counter reset switch?
a. Slave cycler, master cycler, urgent alarm, RPI, NR-45 recorde b. P/A converter, master cycler, multiplexer, bank overlap unit trigger circui ,
 
c. Alarm circuits, master cycler, slave cycler, step counter, bank overlap unit, and P/A converter, d. Master cycler, slave cycler, step counter, P/A converter and urgent alar .
QUESTION: 080 (1.00)
Which ONE of the following describes the effect of VCT pressure being at 13 psig on the operation of the Reactor Coolant Pump (RCP) seals?
a. The pressure is too low and insufficient seal flow will be obtained from the #2 sea b. The pressure is too low and insufficient seal flott will be obtained through the #3 sea c. The pressure is too high and will force excessive flow through the #2 RCP sea l d. The pressure is too high and will force excessive flow through the #3 RCP sea i
 
l
 
  -. .
_ _ _ _ . _ _ _ . . _ . _ . _ . . _ _ . . . _ _ . _ . . - .
REACTOR OPERATOR      Page 45 QUESTION: 081 (1.00)
During a normal reduction in power using boration, which ONE of the following is the reason that additional pressurizer heaters should be ennrgized?
a.'Al* low an increased ramp rate for the power change b. Ensure positive pressurizer pressure control is estaIblished prior to starting the power change c. Maintain PZR pressure in normal operating range during the power change d. Equalize the reactor coolant system and Pressurizer boron concentrations QUESTION: 082 (1.00)
Which ONE of the following indications on the RMS Station Grid would be indicative of an RMS Area monitor that has failed LOW 7    >
Tha detector unit and channel number would be colored ...
a. light blue b. yellow c. violet d. white I
 
  . . . _ _ - _
REACTOR OPERATOR    Page 46 4 QUESTION: 083 (1.00)
Which ONE of the following ds used as the reactor power input to the Rod i Insortion Limit (RIL) compacer?    j l
a. Average Delta-T b. Auctioneered High Tavg    )
c. Turbine Impulse Pressure d. Auctioneered High Power Range NI
.
QUESTION: 084 (1.00)
Which ONE of the following describes the purpose of the Containment Purge System?
a. Reduce containment atmosphere activity levels prior to personnel entry for the monthly containment inspectio b. Remove airborne radioactivity from containment immediately following a LOC c. Provide additional cooling to containment during hot shutdown condition d. Remove airborne radioactivity from containment during outage QUESTION: 085 (1.00)
Which ONE of the following components is designed to limit the rate of affected S/G blowdown during a steam line break inside containment on Unit 27 a. Main Steam Line siz b. Main Steam Flow Ventur c. Main Steam Isolation Valv d. Main Steam Flow Limite . . - - . . . _ _ .
 
REACTOR OPERATOR    Page 47
 
l
 
l QUESTION: 086 (1.00)
Giv:n the following Unit 2 plant conditions:
- The reactor was operating at 100% power when a reactor trip occurre A.1oss of all AC power occurred at the time of the reactor tri I Which ONE of the following is the length of time, based on FSAR
  '
#
rsquirements, the DOS station battery is DESIGNED to carry its normal chutdown loads?
l a. 30 minutes    i hour c. 8 hours    l d. 24 hours
 
QUESTION: 087 (1.00)    i Tha operation of which ONE of the following systems / components ir !
interlocked with a SPECIFIED NUMBER [ minimum or maximum] of Circulating Water Pumps to be in operation?    .
I a. Steam dump system b. Control Room ventilation condenser c. Air ejectors
,
d. Main turbine above 50% rated load
,
 
1
 
_ _ _ . _ _ . _ . _ _ . _ _ . _ _ . _ _ _ _ _ . _ . _ _ _ _ . _ . - _ - _ _ .. _ . _ _ _ _ , _ . _ _ _ . .
        .
        .
REACTOR OPERATOR      Page 48 l
l
' QUESTION: 088  (1.00)
Which ONE of the following is the reason a nitrogen overpressure is mtintained in the Pressurizer Relief Tank (PRT)?
a. Reduce the amount of water in the PRT necessary to quench
  ' pressurizer relief valve discharge.
 
L  b. Reduce the potential for overpressurizing the PRT and prevent deformation of the rupture disks.
 
I  c. Reduce the potential for an explosive mixture of hydrogen and l
oxygen in the PRT.
 
!
d. Reduce the potential for corrosion in the PR :
i l        l l QUESTION: 089 (1.00)
Which ONE of=the following conditions would result in AUTOMATIC closure of the CCW surge tank vent valve, RCV-017, if open?
I a. High level in the CCW surge tank.
 
!  b. Low level in the CCW surge tan c..High radiation level at CCW surge tank vent.
 
l d. High radiation level at CCW pump suction.
 
l QUESTION: 090  (1.00)
Which ONE of the following describes the primary purpose of the activated charcoal filters in the Containment Cleanup System?
a. Remove radioactive iodine, b. Remove moisture and water droplet c. Remove particulate matter.
 
. d. Remove explosive gasses.
 
I i
;
.
  - . ,
 
I REACTOR OPERATOR    Page 49 QUESTION: 091 (1.00)
Which ONE of the following is the reason the Post Accident Containment Ventilation System (PACVS) is placed in service when hydrocan concentration is about 3%?
a.'To ensure containment hydrogen analyzer indications remain accurat b. To ensure explosive mixtures will not exist in the Auxiliary Building exhaust syste c. To ensure hydrogen sampling system capacity is not exceede I d. To ensure oxygen concentration is sufficient to support hydrogen recombiner operation.
 
QUESTION: 092 (1.00)
Which ONE of the following fuel handling components does NOT require l air / nitrogen'for operation?    l a. Fuel transfer car b. RCCA change fixture carriag c. Manipulator crane gripper assembl d. Control rod drive shaft unlatching too :
 
{
REACTOR OPERATOR    Page 50 QUESTION: 093 (1.00)
Which ONE of the following explains why the pressurizer spray valves rcmain operable on a loss of instrument air to Unit 1 containment?
a. A backup regulator from service air to Unit 1 containment
' instrument air supply will open to supply air to Unit 1
-
containmen b. A backup regulator from the installed nitrogen twelve pack is lined up when at power.
 
,
j c. A backup regulator from the Unit 2 containment instrument air
'
supply will open to supply air to Unit 1 containment, d. A backup regulator from an installed nitrogen bottle is lined up when at powe ;
I l
l l QUESTION: 094 (1.00)
Giv:n the following Unit 1 plant conditions:
i
- The plant is at 100% powe Pressurizer pressure is in AUTOMATIC contro The Pressurizer Pressure Channel Defeat switch is in the normal positio No operator action is taken.
 
I Which ONE of the following actions occurs when pressurizer pressure detector PT-431 fails LOW?
a. Pressurizer PORV (PCV-430) cycles to maintain pressure below the reactor trip setpoin I b. Pressurizer heaters and spray valves operate normally to I
      '
maintain pressurizer pressure.
 
I l Pressurizer pressure DECREASES, resulting in a LOW pressure )
l reactor tri Pressurizer pressure INCREASES, resulting in a HIGH pressure l reactor trip.
 
I
 
      !
t
 
  - . . . - . . . . - . . - - . _ . - - . - - - - - - -
_ REACTOR OPERATOR    Page 51 QUESTION: 095 (1.00)
Wh n recovering a dropped rod in control bank C, an " Urgent Failure" alarm is received. This condition normally blocks rod motion for the affscted bank in AUTO and MANUA Which ONElof the following reasons explains why the recovery of the dropped rod is possible?
a. Selecting the specific bank with the Bank Selector Switch overrides the bloc b. Rod motion is allowed since only 3 of 4 disconnect switches for the affected power cabinet are opened, c. Rod motion is allowed when the Demand Step Counter for the affected banks is reset to zer d. Resetting the Pulse-to-Analog (P-A) Converter removes the bloc QUESTION: 096 (1.00)
Which ONE of the following describes the flow path that allows REFLUX cooling to remove heat from the core?
a. Boiling in the core with steam flowing out the break to the sump for recirculation and then back to the hot leg b. Vapor bubbles formed in the core and condensing in the S/Gs to flow back to the core through the cold leg c. Boiling in the core and steam condensing in the S/Gs to flow back to the core through the hot leg !
d. Vapor bubbles formed in the core and condensing in the head area to flow back into the cor l
 
i REACTOR OPERATOR    Page 52 I
l
      ;
      ,
l QUESTION: 097 (1.00)    !
l Which ONE of the following conditions requires the main steam isolation l valves to be shut during the performance of CSP-S.1, " Response to Nuclear Power Generation /ATWS"?
'
a.1If a steam generator tube rupture is identifie I b. Upon steam generator level reaching the high level alar If RCS temperature decreases below 547 degrees d. Actions to trip and runback the turbine are unsuccessfu l l
QUESTION: 098 (1.00)    l l
Uhich ONE of the following conditions will result in the AUTOMATIC l closure of Waste Condensate Discharge valve WL-018?
l a. Waste Condensate Pumps tri Loss of Circulating Water flo c. High level alarm on the Waste Holdup Tan .
d. An alarm condition on Radiation Monitor Channel RE-21 QUESTION: 099 (1.00)
Which ONE of the following is the pressure at which IA-3014, " Service Air to Instrument Air Backup Valve" will BEGIN to open on a loss of 3 Instrument Air?
a. 90 psig    i i psig psig d. 75 psig l
 
l REACTOR OPERATOR    Page 53 l
QUESTION: 100 (1.00)
Which ONE of the following actions is required immediately by the escort-if a visitor becomes separated from the escort?  -
a. Make a page announcement to direct roving security personnel to Treport to last known locatlon of visito l b. Make a page announcement to direct the visitor to report to the
      '
control roo c. Notify security _to deenergize vital area door card readers, d. Notify Central Alarm Statio ,
i l
i
.
  (********** END OF EXAMINATION **********)
 
- . . . , , . - - . - . . .-. - . - . . . . - . - . . . . . . . . . . . . . _ . . - . . . . . . . . . . . . . . . . - . -
~ REACTOR' OPERATOR  <      Page 54
 
. ANSWER:  001 ('1. 0 0 )        l l
          ' I i
REFERENCE:
  *
  . g TRHB'10.12, Re , pg. 6; C1301AOT; Setpoint Document, Section 2.4; and Logic Sheet l l
KA 026000A301 [4. 3 /4. 5]        j
 
l 026000A301  ..(KA's)
  .
ANSWER:  002 (1.00) REFERENCE:          l TRHB 10.1, Rev. 4,  pg. 15 and C1805 COT and TRHB Figure 10.-l.24-
-KA 022000G007 [3.3/3.5)
          !
022000G007  ..(KA's)
i ANSWER:  003 (1.00)        j i !
l
,
REFERENCE:
l TRHB 13.7, Re , pg. 2.
 
l l KA 059000A302 [2.9/3.1)
          ,
059000A302  . . ( FA ' s )
r l
!.
l
!
,
. , -  - - -  ,.m-..
      -
 
e REi.CTOR OPERATOR    Page 55 ,
      !
      !
      !
' ANSWER: 004 ( l '. 0 0 ) ,
; REFERENCE:    ,
8 .
AOP-10A, Rev. 18, pg. 1 [Similar to question used on 94 RO exam.]  ;
KA 000068A101 [4.3/4.5]
000068A101 ..(KA's)
.
      '
ANSWER: 005 (1.00) t
      :
REFERENCE:
TRHB 10.15, Rev. 5, pg. 7 and C4313AOT and P&ID Sheets 541F091 Sheet 1 !
and.9645971 Sheet 1A KA 068000K107 [2.7/2.9]
      ,
068000K107 ..(KA's)
      .
ANSWER: 006 '(1.00)    ; f i
l
 
. .-. ~ .. . ... . - _ . ... .- . . - -. - - . - -. - .. - . - . - . -_ - .. . .... - .. .... . -..... ..
'
 
-
l
' REACTOR OPERATOR      Page'56  I f REFERENCE:
 
'
AOP-6C, Re , pg. 3 and LP 2441,.EO'2.1 KA~0000017.205.[4.4/4.6]
    .
(Similar question used on 95 SRO exam.]
.
v .
l  0000611205  ..(KA's)
;
;
?
1.
 
i . ANSWER:
 
007 (1.00)
! d.
 
j.
:
l-REFERENCE: AOP-9B, Re , p .
:
i
; '
KA.000026G010 [3.6/3.5]        4 i
(Similar to question on 95 SRO exam.]
:
;  000026G010  ..(KA's)
 
          !
:
I  008'
AMSWER:  (1.00)
-
l
; '
i l          l I
{ REFERENCE:
{ 10 CFR 20.1003, Definition [Similar question used on.95 SRO exam.]
l KA 194001K103 (2.8/3.4]        l
.
;
;
194001K103  ..(KA's)
 
.
 
l i
i i
 
$
>
,
  > , , - . _  -- . . .
 
    !
REACTOR OPERATOR    Page 57 l
    *
ANSWER: 009 (1.00)-
5.
 
REFERENCE:    .
5.- ?    i TRHB 10.2, Re I 5, pg. 1 i KA 003000K407 (3.2/3.41
 
003000K407 ..(KA's)  ,
l l
ANSWER: 010 (1.00)    j
    : l l
REFERENCE:
l
:TRHB 10.6, Re , pg. 71 and Master Data Book for B0 I KA-004000K203 [3.3/3.5]
004000K203 ..(KA's)
ANSWER: 011 (1.00)
b.
 
REFERENCE:
TRHB 10.16, Rev. 3, pg. 12 and C7304 COT and Logic Sheet 7 KA 013000A302 (4.1/4.2]    ,
I 013000A302 ..(KA's)
l
    .
 
REACTOR OPERATOR    -Page 58-ANSWER: 012 (1.00)
b.
 
REFERENCE:
t -
TRHB 13.4, Re , pg. 18 and C730 COT and Logic Sheet KA 013000A402 [4.3/4.4]
013000A402 ..(KA's)
.
ANSWER: 013 (1.00)
d.
 
REFERENCE:
TRHB 13.1, Rev. 3, Figure 13.1.19 and C1315 CO KA 015000A202 [3.1/3.5]    '
 
015000A202 ..(KA's)
ANSWER: 014 (1.00)
i b.
 
REFERENCE:
TRHB 13.1, Re , pg. 11 and C1310 CO KA 015000K601 [2.9/3.2]
015000K601 ..(KA's)
    -_ _ . _ _ _
 
  . . .. . - _ .. . . __
REACTOR OPERATOR    Page 59 ANSWER: 015 (1.00)
b.
 
REFERENCE: ,
TRHB 13.8, Re , pg. 6 and C0105 COT; and TRHB 13.8, pg. 19 and Logic Shtot 1 KA 015000K302 [3.3/3.5]
015000K302 ..(KA's)
ANSWER: 016 (1.00)
c.
 
REFERENCE:
Steam Tables KA 017020K502 [3.7/4.0]
017020K502 ..(KA's)
ANSWER: 017 (1.00)
C.
 
REFERENCE:
TRHB 13.2, Re , pg. 8 and TRHB Figure 10. KA 017020K102 [3.3/3.5]
017020K102 ..(KA's)
 
-. - - _ . ~ . - . . . . . . . . . . . - ~ . . - . . _ _ . . - . . . . . - . . - . - _ . - . .
1-
).          ;
'
 
1
; REACTOR OPERATO Page 60
 
.
j. ANSWER: 018 .(1.00)
. l a
1 REFERENCE:        *
-
s .        i
 
,
BG EOP-0.2, Rev. 13, pg. 7 and C7033 CO l
; KA 022000K404 [2.8/3.1]
.,
j hl
;
022000K404  ..(KA's)
i s
  #
J
$ ANSWER: 019 (1.00)
;          l I ,
 
,t          ,
?!
j= REFERENCE:
'
TREB 11.2, Re , pg. 11 and C3912 COT and Logic Sheet 24
. KA 056000A204 [2.6/2.8)
*
 
t          ;
j 056000A204  ..(KA's)
i
.          \
;
l ANSWER: 020 (1.00)
i t a.
 
i
,
; REFERENCE:
4 TRHB 13.7, Re , pgs 28 and 29 and C4210 COT and Logic Sheet 10 l.
 
i [EBQ 052-05-149A]
 
1 l KA'059000K419 [3.2/3.4]
{
; 059000K419  ..(KA's)
 
1
 
.
k
 
3
-
  .. .- ~  . -  . . . , . . . , - . ,
 
  . ._ _ . .- .. ._ . _ REACTOR. OPERATOR    Page 61
!
ANSWER: 021 (1.00) REFERENCE:
,
TRHB 10.15, Re , pg. 24 and C4504AOT and P&ID 648J972 Sheet 1 KA 071000A405 [2.6/2.6]
071000A405 ..(KA's)
AMSWER: 022 (1.00) REFERENCE:
TRHB 10.2, Re , Figure 10.2.1.B and P&ID 541F091 Sheet 1 KA 002000K606 [2.5/2.8)
002000K606 ..(KA's)
ANSWER: 023 (1.00) REFERENCE:
TRHB 10.8, Re , pg. 15 and Control Panel CO2 KA 006030A401 [4.4/4.4]
006030A401 ..(KA's)
 
  -. .- .. - - . .
.
REACTOR OPERATOR    Page 62 ANSWER: 024 (1.00)
.
c.
 
!
REFERENC,E: ,
TRHB 10.8, Re , pg. 18 and Technical Specification 15.3.3.
 
KA 006000G005 [3.5/4.2]    I 006000G005 ..(KA's)
l      l
 
ANSWER: 025 (1.00)
a.
 
j
'
REFERENCE:
TRHB 10.3, Re , Figure 10.3.7, Logic Sheet 18 and ICP 1 KA 010000A403 [4.0/3.8)
.
'
010000A403 ..(KA's)
:
ANSWER: 026 (1.00)
: a.
 
,
REFERENCE:    1
 
TRHB 13.5, Re , pg. 21, C0905C01, Logic Sheet 18 and P&ID 541F091 Sheet 1 KA 010000K301 [3.8/3.9)
010000x301 ..(KA's)
i
 
. _ _ - .. . . - _ _ _ . _ _ . _ . _ . . . . . - _ . . _ _ _ . _ _ _ _ _ _ _ _ . . . _ . . _ _ . . . . _ . . _ , .
          ,
s
.
] REACTOR OPERATOR        Page 63
          -
 
s I          i l ANSWER: 027 (1.00)        !
i          I
' l l
l REFERENCE:
j i; j TRHB 13.6, Re , pg. 7 and C1006 COT and Logic Sheet 18 KA 011000K604 (3.1/3.11
,
:
2 011000K604 ..(KA's)
!
  .
: ANSWER:
:
028 (1.00)
*
a.
 
l l REFERENCE:
TRHB 13.6, Re , p .
'
KA 011000A101 (3.5/3.6)
011000A101 ..(KA's)
WER: 029 (1.00)
          ! l
 
REFERENCE:  g)ggppy      I TRHB 13.3, Rev. 4 . wwW KA 012000K406 (3.2/ K406 ..(KA's)
    @
_
 
. _ . _ . . . . _ _ . _ _ . . - . . _ . . _ _ . . _ = _ _ _ _ . _ _ _ . - . _ . . _ . _ _ _ _ . _ . - . . _ _ _
REACTOR OPERATOR'        Page 64 ANSWER:  030 (1.00)
l- ' REFERENCE:
N-l BG CSP-I.3, Re , pg. KA 016000G015 [3.6/3.8)
016000G015  ..(KA's)-
i
!
ANSWER:  031 (1.00)
d.
 
!
i. REFERENCE:
!
TRHB 13.8, pg. 20.
 
l KA 035010k501 [3.4/3.9)
l 035010K501
          '
  ..(KA's)
l ANSWER:  032 (1.00)
a.
 
l REFERENCE:
          *
TRHB 12.5, Re , Figure 12.5.1 and Master Data Book KA 062000K201 [3.3/3.4)
i 062000K201  ..(KA's)
          '
i I
!
!
  . . _ _ -- - , _ . _ . ,  . _ . ,
        - _ - - _
 
. - . . ~ . . . - . . . . . . . . . . . - - . . - . . _ . . . . - . . - . . . . - _ . - . . . . . _ . - ..-- -....-. -... ..
 
[
i i REACTOR OPERATOR        Page 65 I
!          '
,
l          I j-. ANSWER: 033 (1.00) -
t l
l ,
 
REFERENCE:        ,
  ,
  . .        ;
TRHB 12.2, Re , pg. 38, C4105AOT, OI-110, Rev. 1, and PC-29, Rev. 29 Monthly Best KA 064000K402 (3.9/4.2]
064000K402  ..(KA's)
          ,
  '
          .
          (
ANSWER: 034 (1.00)
          ' r I
          ,
REFERENCE:
TRHB 12.8, Re , pg. 74 and OP-11A, EDG G01/G02 KA 064000A206 [2.9/3.31
 
1 064000A206  ..(KA's)
ANSWER: 035 (1.00) REFERENCE:
TRHB 13.12, Rev. 3, pg. 1 KA 073000A402 (3.7/3.7]
 
i
          '
073000A402  ..(KA's)
i I
i l
          !
I
 
REACTOR OPERATOR    Page 66 ANSWER: 036 (1.00)
a.
 
REFERENCE:
'-
TRHB 10.7, Rev. 7, pg.'17, C0506AOT and Setpoint Document,. Sections 2.3, 8.2 and 1 KA 005000K109 [3.6/3.9)
005000K109 ..(KA's)
.
ANSWER: 037- (1.00)
d.
 
REFERENCE:
.TRHB 13.10, Re , pg. 2; 62804 COT; Setpoint Decament 14.13; P&ID PBM-241;.OP-1c, Step 7. KA 041020K603 [2.7/2.9)
041020K603 ..(KA's)
ANSWER: 038 (1.00) l l
REFERENCE:      '
AOP-6B, Re , pg. 1 and LP2441, EO )
KA 000005A201 [3.3/4.1]
000005A201 ..(KA's)
      ,
  . . _ . - ,
  -
    - . , - - , . ,, . _ - , _ , , , 4
 
. _.. _ _ _ . . - . . . . . . . . . _ . _ , _ . _ . . . . . = _ . _- _ .
,      i e
-
i
<
, REACTOR OPERATOR    Page 67 i
!
i r
ANSWER:  039  (1.00) REFERENCE:  ,
ECA 0.1, Rev. 10, pg. 12    j KA 000015G007 [3.1/3.2]
000015G007  ..(KA's)
  .
ANSWER:  040  (1.00) REFERENCE:
EOP-0.1, Rev. 14, p .
KA 000024A205 [3.3/3.9)
000024A205  ..(KA's)
ANSWER:  041  (1.00) l I
REFERENCE:
 
EOP-2, Rev. 10, pg. j KA 000040A110 [4.1/4.11    !
      :
i 000040A110  ..(KA's)  l l
 
      !
      !
 
.
  .._....-.~ !
REACTOR OPERATOR    Page 68
      ,
      :
'
ANSWER: 042 (1.00)    !
  '
h b.
 
.l .
.
l REFERENCE:    )
s' .    :
1      l
; EOP-1, Rev. 19, Foldout pag ;
KA 000040A205 (4.1/4. 5 )
000040A205 ..(KA's)
i      I i      l a
  *
! ANS#ER: '043 (1.00)    i C.
 
l
:
} REFERENCE:
.
1.
 
j TRHB 13.9, Rev. 2, pg. 4 and Setpoint Document, Section 14.2.
 
1      i i
KA 000051K301 (2.8/3.1)    )
i
-
i 000051K301 ..(KA's)
      'I ANSWER: 044 (1.00)
.L /fvt/Sves &//#Cw$ h45 CE/ />P P4M t-iTY
*
FERENCE:
ls 5 r BM~v  A amw csmm car- :
AOP-5A,.Re , p .  ;
KA 000051A202 (3.9/4.1)    I 000051A202 ..(KA's)
 
. _ _ _ _ . . _ . . . . . _ . . _ _ . . - . . . _ . _ . . _ _ _ . _ _ . . _ . . _ . _ . - _ _ _ . _ _ _ _ . _ . _ . . . _ _ _ . _ _ . _ _ . .
.
            :
1            !
. ' REACTOR OPERATOR          Page-69 '
?
 
!
*
            .,
 
i
; ANSWER: 045 (1.00)
;
 
i i4
!~            l i            j 1 REFERENCE:
  , .
LP 0462, Re , pg. 7 and LO K302 [4.3/4.61 000055K302  ..(KA's)
AMSWER: 046 (1.00) . REFERENCE:
BG ECA-0.0, Rev. 14, pg. 14 and LP 0462, LO KA 000055A106 [4.1/4.5]
000055A106  ..(KA's)
ANSWER: 047 (1.00)
            !
i REFERENCE:
TR11B 12. 9, Re , pg. 1 A218 [3.1/3.1]
l 000057A218  ..(KA's)
 
I I
l REACTOR OPERATOR  Page 70 j ANSWER: 048 (1.00) I REFERENCE:
3 .
TRHB 11.14, Re , pg. 2 (Similar to EBQ 052-01-065A] I KA 000067K102 [3.1/3.9)
000067K102 ..(KA's)
ANSWER: 049 (1.00)
b.
 
REFERENCE:
Tech. Spec. 15.3.6, pg. 15.3.6- KA 000069G003 [3.3/3.9)
000069G003 ..(KA's)
ANSWER: 050 (1.00) ,
REFERENCE:
CSP-C.1, Rev. 12, pg. 2 KA 000074K201 (3.6/3.9)
000074K201 ..(KA's)
 
REACTOR OPERATOR  Page 71 ANSWER: 051 (1.00)
d.
 
REFERENCE:
,
EOP-0, Rev. 20, pg. 2 and LP 0405 LO 2.1 KA 000007A202 (4.3/4.6]
000007A202 ..(KA's)
ANSWER: 052 (1.00)
b.
 
REFERENCE:
BG EOP-0, Rev. 21, pg. 9 and LP 0405, LO 2.1 KA 000007K301 (4.0/4.6]
000007K301 ..(KA's)
ANSWER: 053 (1.00)
b.
 
REFERENCE:
EOP-0, Rev. 20, pg. 18 and LP 0405, LO KA 000008A101 (4.2/4.0)
    !
000008A101 ..(KA's)
    ;
 
REACTOR OPERATOR  Page 72-ANSWER: 054 ' ( 1. 0 0 )
 
REFERENCE:
s .
,
EOP-1.2, Rev. 13, Foldout Page-and LP 0435, LO KA 000009A234 [3.6/4.2]
000009A234 ..(KA's)
.
ANSWER: 055 (1.00) l REFERENCE:    !
OP-5A, Rev. 28, pg. 1 and THRB 10.6, Rev. 3, pg. KA 000022K307 [3.0/3.2]
    !
    ~
000022K307 ..(KA's)
ANSWER: 056 (1.00)  ) \
I REFERENCE:
SEP-1.1, Rev.-0, pg. KA 000025K101 [3.9/4.3]
    !
000025K101 ..(KA's)
 
REACTOR OPERATOR  Page 73 i
i
    !
i
' ANSWER: 057 (1.00)  j .
REFERENCE:
'
1  i EOP-3, Rev. 19, pg. 3 !
KA 000032A101 [3.1/3.4]  1
    :
000032A101 ..(KA's)  ;
ANSWER: 058 (1.00)  '
i REFERENCE:
AOP-1A, Rev. 8, pg. KA.000037G011 [3.9/4.1)
    ;
000037G011 ..(KA's)
s ANSWER: 059 (1.00) REFERENCE:
EOP-3, Rev. 19, pg. 15 and BG EOP-3, Step 20 (PROVIDE: EOP-3, pg. 15 of-36)
'KA 000038A136 (4.3/4.5)
000038A136 ..(KA's)
 
_ -_ - __ . - . _ - .
REACTOR OPERATOR    Page 74 i
    !
l l
ANSWER: 060 (1.00) REFERENCE:
s .
BG CSP-H.5, Re , pg. 4 and LP 1998, EO )
KA 000054K102 [3.6/4.2]    l 000054K102 ..(KA's)  l ANSWER: 061 (1.00) i l
REFERENCE:
 
BG CSP-H.1, Rev. 10, pg. i KA 000054K304 (4.4/4.6]
000054K304 ..(KA's)
I ANSWER: 062 (1.00)
a.
 
REFERENCE:
AOP-0.0, Re , pg. KA 000058A203 [3.5/3.9)
000058A203 ..(KA's)
 
_ _. _..- _.. _ __ _._ _ . .. _ _- _ _ _ _ _ _ _ _ . - _ . . . . _ _ . _ . _ _ , _ - . .
REACTOR OPERATOR      Page 75 ANSWER: 063 (1.00)
b.
 
REFERENCE:        ,
'
.:
TRHB 13.6, Rev. 1, pg. 6 and Figure 13.6.1 and Logic Sheet 1 KA 000028K202 [2.6/2.7)      ;
000028K202 . . ( YA ' s )
i ANSWER: b64 (1.00)      j l REFERENCE:
EOP-0.1, Rev. 14, pg. 1 KA 000056K101 [3.7/4.2]
000056K101 ..(KA's)
I
 
ANSWER: 065 (1.00)
b.
 
REFERENCE:
NP 1.1.4, Re , pg. KA 194001A101 [3.3/3.4]
194001A101 ..(KA's)      l l
l
 
        !
 
,__ _ . . . _ _ _ _ . . _ _ _ . _ ~ , . _ . _ _ . . _ . . . . . _ _ . ~ . _ _ . _ . ~ . . _ . - _ _ . _ _ . _ _ _ _ _ . .
\
!.
REACTOR OPERATOR      Page 76
,
ANSWER:  066- (1.00) t REFERENCE:
  \ -
=EOP-0, Rev. 20, pg. 1 and SEP-2, Rev. 2, pg. KA 194001A102 [4.1/3.9]
194001A102  ..(XA's)
ANSWER:  b67  (1.00) REFERENCE:
10 CFR 55.53 and OM 3.10, Rev. 5, pg. 1 KA 194001A103 [2.5/3.4]
194001A103  ..(KA's)
ANSWER:  068  (1.00) REFERENCE:        !
NP-1.7.5, Re , p . i KA 194001A103 [2.5/3.4)      j 194001A103  ..(KA's)
l
 
. _ _ _ . _ _ _ _ . _ _ . _ . _ . _ _ _ . . _ . _ _ . . . . _ _ . _ . _ _ . . _ . _ . _ . . . . . . _ _ . . - . - _ . . . . .
          -
          . . _ _ _ . . , -
            ,
,
REACTOR OPERATOR        Page 77 i
,
 
ANSWER:  069- (1.00)        i i
: .
;            i
            '
;
! REFERENCE:
4  3 .
l NP 1.7.10, Re , pg. 1.
 
KA 194001A105 [3.6/3.8)          >
i
  .
1            ,
, 194001A105  ..(KA's)
            '
i b
i e
i ANSWER:
.
070 (1.00)
 
1 b.
 
!
REFERENCE:  '
l
 
Manager's Expectations, 93-1.
:
l KA 194001A103 [2.5/3.4]          l 1            i i            ,
j 194001A103  ..(KA's)        I
;            l
:            ,
i
 
l ANSWER:  071 (1.00)
.
d.
 
>
REFERENCE:
l 'OM 3.10, Rev. 5, pg. KA 294001A110 [2.9/3.9]
194001A110  ..(KA's)
 
1
 
_ _ _ . . _ _ . . _ - _ . . _ . . . _ . _ . . . . _ _ _ _ _ _ _ . . _ . . . , _ . . _ . _ . _ _ _ . . _ _ . _ . _ . _ _ _ _ . . _ . _ _ _ _ _ _ . _ _ _ _ _
            , i 1            l REACTOR OPERATOR          Page 78 i
i
!
ANSWER: 072  (1.00)
i C.
 
i REFERENCE:
  * .
.
Tech. Spec. Definition 15.1.f.
 
a
! KA 194001A113 [4.3/4.1]
a
!-
'
194001A113  ..(KA's)
i ANSWER:  b73  (1.00)
.
b.
:
I
! REFERENCE:
l NP 1.6.6, Re , p .
!            l j KA 194001A103 [2.5/3.4]          l
 
;
l  194001A103  ..(KA's)
:
s
 
ANSWER: 074  (1.00) REFERENCE:
OM 3.17, Rev. O, pgs. 4 & KA 194001K101- [3.6/3.7)
l 194001K101  ..(KA's)        ,
            :
l I
 
REACTOR OPERATOR    Page 79
      '
ANSWER: 075 (1.00) l
      .
REFERENCE:    )
-
%.
l NP 1.9.15, Rev. 2, pg. 1 l KA 194001K102 [3.7/4.1)
194001K102 ..(KA's)
ANSWER: 076 (1.00)
d.
 
REFERENCE:
TRHB 10.6, Rev. 3, pg. 35 and C0413 COT and Setpoint Document,-Section KA 004010A105 [3.0/3.2]
i 004010A105 ..(KA's)
ANSWER: 077 (1.00)
C.
 
REFERENCE:
TRHB 11.4, Rev. 5, pg. 15 and Tech. Spec. 15. KA 061000G005 (3.3/4.0)
061000G005 ..(KA's)
 
    'i REACTOR OPERATOR  Page 80 ANSWER: 078 (1.00)
c.
 
REFERENCE:
s .
,
TRHB 13.8, Re , pg. 3,7 and C0105 CO KA 001000K403 [3.5/3.8)
001000K403 ..(KA's)
ANSWER: 079 (1.00)
'C.
 
REFERENCE:
TRHB 10.5, Rev. 4, pg. 1 KA 001010K604 [2.9/3.2)
001010K604 ..(KA's)
ANSWER: 080 (1.00)
a.
 
REFERENCE:
i TRHB 10.6, Rev. 3, pg. 34 and C5702 CO KA 003000A205 [2.5/2.8)
.003000A205 ..(KA's)
_ _ .
 
l REACTOR OPERATOR  Page 81
'
ANSWER: 081 (1.00) REFERENCE:
s .
OP-3A, Rev. 37, pg. KA 004000K601 [3.1/3.3]
004000K601 ..(KA's)
ANSWER: 082 (1.00) REFERENCE:
TRHB 13.12, Re , pg. 15 and C5005 CO KA 072000A202 (2.8/2.9]
t 072000A202 ..(KA's)
ANSWER: 083 (1.00)
'
a.
 
,
REFERENCE:
TRHB 13.13, Re , p .
KA 014000A103 [3.6/3.8)
014000A103 ..(KA's)
 
. . . _ . . _ _ . _ _ _ . _ _ _ _ . . _ . _ . . _ . . _ _  _ _ . _ . _ . . . _ . . - _ _ _ . . . _ . _ . . . . _ _ .
;
, REACTOR OPERATOR        Page 82 ANSWER:  084 (1.00)
o l
'd.
 
!
l
.
L REFERENCE:
  , .
TRHB 10.01, Re , pg. 17.
 
l KA 029000G004 [2.9/3.0)
L i
l  029000G004  ..(KA's)
,
l
,
  -
l ANSWER:  085 (1.00)
b.
 
;
i-l REFERENCE:
l l TRHB 11.1, Rev. 6, pg. KA 039000K101 [3.1/3.2]
039000K101  ..(KA's)
ANSWER:  086 (1.00)
b.
 
t
' REFERENCE:
l TRHB:12.7,.Re , pg. 7; and FSAR Section 8.2, pg. 8.2-19 l
KA 063000A101 [2.5/3.3]
l i
063000A101  ..(KA's)
L i            ;
;
i n
,
I I
. - -  ., , __
_ . - - . . . _ - , . . . _ _ __ -  , _
 
-. .. . . . . - . . - ~ . _ . ~ . _ . . . . . - _ . - . _ - - . . . . ~ . . - - . . . . . . . . _ .
I
,
REACTOR OPERATOR      Page 83
!
          '
:
!
l ANSWER: 087 (1.00)
i
; REFERENCE:
<
  \ :
! TRHB 13.9, Re , pg. 4.
 
.
l KA 075000A203 [2.5/2.7)
!
          ,
t 075000A203  ..(KA's)      !
i'
          :
 
l ANSWER: b88 (1.00)      ,
i c.
 
i a
: REFERENCE:
}
1 OP-4C, Rev. 13, pg. 1.
 
KA 007000G007 [2.9/3.1)
l 007000G007  ..(KA's)
 
i
.
ANSWER: 089 (1.00) REFERENCE:
TRHB 10.9, Re , pg. 5 and C0802AOT and RMS Alarm Setpoint and Rosponse Book
[Similar to EBQ 051-06-009A)
KA 008000A204 [3.3/3.5)
008000A204  ..(KA's)
-, .- _ - . . - _ _  - - ..-
 
.
- . - . . _ . . . . - - . . _ - . . - . . - .  . - - - . - . _ ~ . . - . _ . . - . . . . . . - . . . - . - . . . - -
)
! REACTOR OPERATOR        Page 84 T
d
] ANSWER:
090 (1.00)
; 9..
i j
; REFERENCE:
j- .
'
TRHB.10.1, Re , pg. 18.
 
! KA 027000K501 [3.1/3.4)
:
i j 027000K501  ..(KA's) i
!
 
ANSWER: 091 (1.00)
i.
 
q li j.
 
'
' REFERENCE:
.
[
J TRHB 10.11, Rev. 3, pg. 3.
 
(~
J KA 028000K502 [3.4/3.9)
j
! 028000K502  ..(KA's)        j
:            1
 
.,
 
,
ANSWER: 092 (1.00)
: i i
REFERENCE:
TRHB 10.13, Re , pg. KA.034000G009 [3.0/3.0]
034000G009  ..(KA's)
i l
          . . .
_ _ - _ , . . - - , . . . . - . _ _ -, , ._ _ . . - . _ . -
 
_ _ .__ ___ . - . . _ _ -
    . _ _ . _ _ . . . . _ . . _ _ - . . . _ . _ . - - . _ _ . . .__ . _ . . _ .
i
:
  ~
REACTOR OPERATOR      Page 85
:
i i
?
#
ANSWER: .093 (1.00)      l
>
          !
i d.
 
. REFERENCE:
i TRHB'11.18, Re , pg. [EBQ 052-06-005A]
1          I
'
f KA 078000K302 [3.4/3.6]
          '
,
u
'
'
;
078000K302  ..(KA's)      I
 
  ,
i i
i ANSWER:
 
094 (1.00)      j l i l' REFERENCE:
.TIUiB 10. 3, Rev. 3, Figure 10.3.7 and LP 2438 LO 2.1, Logic Sheet 18 and    -l
. the Setpoint Documen !
l
'          i
          '
l KA 000027A101 [4.0/3.9)
i
{ 000027A101  ..(KA's)
;
;
;          i
.          .
l ANSWER:
t 095 (1.00)
i '
i
{ REFERENCE:
,
AOP-6A, Re , pg. 10; AOP-6A BG pg 16.
 
i 104 000003A102 [3.6/3.4)
 
000003A102  ..(KA's)
,
 
i
$        . - -
    .-. _-  ,
 
REACTOR OPERATO Page 86 ANSWER: 096 (1.00)
c
, REFERENCE:
:. '.. *
1 LP 0435, Re , pg. 11.
 
I.
:
KA 000011K101 [4.1/4.4]
 
3 K101 ..(KA's)
i 4-Y
 
ANSWER: 097 (1.00)
l d.
 
J
!
REFERENCE:
.
! CSP-S.1, Rev. 13, pg. 2.
 
i.
 
i KA 000029K308 [3.6/3.8]
]
,
'
000029K308 ..(KA's)
' ANSWER: 098 (1.00) REFERENCE:
TRHB 13.12, Rev. 3, pg. 49; RMS Alarm Setpoint and Response Book KA 000059K201 [2.7/2.8)
.000059K201 ..(KA's)
 
. . _ . . _ . . _ . . . . . _ . _ _ . _ _ _ _ _ _ _ _ _ ~ . _ _ . . _ _ . _ _ _ . _ . . _ . . . . . _ . _  -._._
i          l 2,
 
. REACTOR OPERATOR        Page 87
,
'
l
 
4
:
l' ANSWER:
e 099 (1.00)
I b.
 
^
I'
.
,
REFERENCE:
s .
AOP-5B,~Rev. 10, pg. 2.
 
l    .
l
 
KA 000065A201 [2.9/3.2]        !
i
-
.
: 000065A201  . ,KA's)
;
*
ANSWER: kOO (1.00)      *
 
'd.
 
!
 
4
 
REFERENCE:
] NP 1.7.1, Rev. 1, pg. 3.
 
( KA 194001K105 [3.1/3.4]        ,
;-          l l
:
194001K105  ..(KA's)
 
i 4-
;
:
i l
!
:
 
i
          !
          ,
    (**********  END OF EXAMINATION **********)
      - , - - - - _ . . _ . - .- . _ . -_ _ -
 
e j  U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION
;
SENIOR OPERATOR LICENSE i  REGION 3
"
MASTER EXAMINATIO CANDIDATE'S NAME:  ~
4  FACILITY: Point Beach 1 & 2 REACTOR TYPE: PWR-WEC2 l  DATE ADMINISTERED: 96/10/07
;
!
INSTRUCTIONS TO CANDIDATE:
Use the answer sheets provided to document your answers. Staple this cover i sheet on top of the answer sheet Points for each question are indicated in
{- parentheses after the question. The passing grade reqaires a final grade of
: at least 80%. Examination papers will be picked up four (4) hours after the
 
examination starts.
 
}i  CANDIDATE'S l  TEST VALUE SCORE %
i
 
l 99fa
)qht_100.u0  % TOTALS
 
FINAL GRADE
!.
.1 j All work done on this examination is my ow I have neither given nor i
,
received aid.
 
!
I i    Candidate's Signature
 
3 i
i
;
k l'
 
3
.
l
 
.
PAGES 2, 3 AND 4 WERE USED AS ANSWER SHEETS AND ARE NOT INCLUDED IN THIS MASTER EXAMINATION COP !
    ,
    .
l l
    ,
@'
 
_ _ _ . . -
_ . _ _ _ - ... ~ _ _ > Page 5 l
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
 
l 2. After the examination has been completed, you must sign the statement on
! the cover sheet indicating that the work is your own and you have not
!
received or given assistance in completing the examinat' ion. This must be done after you complete the examinatio . Restroom trips are to be limited and only one applicant at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG . Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question page.
 
l 8. Use abbreviations only if they are commonly used in facility literature, l Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answe Write it out.
 
!
9 The point value for each question is indicated in parentheses after the queFtion.
 
l 10. Show all calculations, methods, or assumptions used to obtain an answer to j any short answer questions.
 
l 11. Partial credit may be given except on multiple choice question Therefore, j ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
 
l 12. Proportional grading Will be applied. Any additional wrong information that is provided may count against you. For example: if a question is worth one point and asks for four responses, each of wnich is worth 0.25 points, and you give five responses, each of your responses will be worth 0.20 point If one of your five responses is incorrect. 0.20 will be deducted and your total credit for that question will be 0.80 instead of i
1.00 even though you got the four correct answers, l
13. If the intent of a question is unclear, ask questions of the examiner only.
 
t e
 
_ _ _ _ _ . _ _ . _ _ _ _ . _ _ _ _ _ _  . __
Page 6 14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheet In addition, turn in all scrap paper.
 
i 15. Ensure all information you wish to have evaluated as part of your answer is on your answer shee Scrap paper will be disposed of immediately following the examination.
 
>
16. To pass the examination, you must achieve a grade of 80% or greate . There is a time limit of four (4) hours for completion of the examinatio . When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoked.
 
t l
l l
,
e
 
      '
t
:
I SENIOR REACTOR OPERATOR    Page 7
!
l l
l
      .
! QUESTION: 001 (1.00)
f l Which.ONE of the following describes the purpose of the back draft
    -
!. dampers installed in the Containment Air Recirculation System?
!
l a. Prevent unit air backflow when the accident fan is running and j the cooling fan is not.
 
, Prevent backflow in a cooling unit in the event of fire in l Containment.
 
i
,
c.. Serve as a system air backflow damper in idle cooling units l (both accident and cooling fans secured).
 
L d. Serve as explosion dampers preventing duct work collapse during
! an accident.
 
l l
QUESTION: 002 (1.00)
Which ONE of the following is the PREFERRED alternate AFW supply which must be aligned during a loss of. secondary coolant when the CST level l
'
decreases to.less than 8 feet, according to EOP-1, " Loss of Reactor or i Secondary Coolant"?    I l
a. Align. service water to AFW pump suctio b. Pump the hotwell to the CST using a condensate pump c. Align AFW pump suction to the fire syste d. Connect a fire hose to the CST drain valve and backfill the CST,
!
l
:
i
*
l. .
I i . _  . - . -
    . . . ,
 
SENIOR REACTOR OPERATOR    Page 8
. QUESTION: 003 (1.00)
Which ONE of the following explains why the steam generator level program is reduced at low power?
a. To minimize time delays in plant transient response due to
" thermal lag".
 
,
b. To prevent thermal stratification above the U-tube c. To reduce the mass inventory available to boil off in the event of a steam brea d. To prevent low power level oscillation due to level dominant contro QUESTION: 004 (1.00)
Given the following Unit 2 plant conditions:
- The plant has just been manually runback by 30% power within the last minut All systems are operable and in automati Actual Tavg is 9 degrees F higher than Tre Which ONE of the following describes the response of the steam dump valves?
a. Two dump valves full open and two modulating b. Four dump valves full open
      )
c. Four dump valves full open and two modulating Six dump valves full open and two modulating
.
 
      ;
SENIOR REACTOR OPERATOR    Page 9 ]
l QUESTION: 005 (1.00)    ;
Which ONE of the following is the final location for the Unit 1 Control Operator after leaving the control room during AOP-10A, " Control Room Inaccessibility", when the control room is inaccessible due to a fire?
a. Auxiliary Feed Pump Room b. 4160 Vital Switchgear Room c. PAB Elevation 8'
d. Emergency Diesel Generator Room i
QUESTION: 006 (1.00)
Which ONE of the following would result if the reactor head to vessel inner "O" ring seal completely fails?
a. Pressurizer Relief Tank level would increase, b. SI on high containment pressur c. SI on low PZR pressur d. Reactor Coolant Drain Tank level would increase.
 
QUESTION: 007 (1.00)
Which ONE of the following conditions would require a reactor trip during a Loss of Component Cooling Water (CCW), according to AOP-9B,
" Loss of Component Cooling"?
a. CCW low flow alarms on the SI pump b. A CCW high radiation monitor alarm is receive c. CCW pump discharge low pressure alarm is receive d. Unable to maintain CCW Surge Tank leve .
 
.. .__ -_ . _ . . . _ . _ _ . _ . _ _ . ... - . - _ . ___ _ -. __ _ _ .___._~. . . ___ _
e I
j 1 SEMIOR' REACTOR OPERATOR    Page 10 l
:
i i-QUESTION: 008 (1.00)
Given'the following Unit 1 plant conditions:
  '
;  - Unit 1 is in Cold Shutdown.
 
i  - RCP 1A motor is uncoupled from the pump.
 
{  - RCS is filled and vented.
 
  - Maintenance is inspecting RCP 1A seal Wh1ch ONE of the following minimizes leakage of reactor coolant upward
.
along the RCP shaft?
I a. The pump shaft mates with the thermal barrier casing.
 
.
b. Nozzle dam installation prevents RCS water from entering the RCP shaft area.
 
; Seal ~ injection is maintained during this condition.
 
' Seal leakoff collects any RCS leakage up the shaft and directs  {
it back to the VC !
i
        !
QUESTION: 009~ (1.00)
Which ONE of the following is the ALTERNATE power supply for the 2 P2A Charging Pump?
a. B08      i B09 c. 1B03 B03 t
        ,
  ..
 
. _ _ . _ . _ __ _ _ _ . . _  _ . _ . . _ . _ . . . ~ _ ~ . _ _ _ _ _ _ . _ _ . _ _ _ _ . . _ _ . _ _ _ . _ _ _
          !
 
J l SENIOR REACTOR OPERATOR-      Page 11
!
l QUESTION: 010  (1.00)
l l
Which ONE of the-following actuations WILL be. caused by an AUTOMATIC SI, but WILL NOT be caused by a manual SI?      i
!  a.' Containment Spray l          l b. Containment Isolation l  c. Closure of emergency diesel generator output breakers onto l  . safeguards busses
 
d. Trip of main feedwater pumps      !
I I
l QUESTION: 011  (1.00)
l Which ONE of the following safety injection actuation signals would be automatically unblocked if pressurizer pressure increased to 1800 psig after safety injection had been manually blocked?
a. Low pressurizer pressure onl I l          \
b. Low pressurizer pressure and steamline low pressure onl !
 
c. Low pressurizer pressure, steamline low pressure, and    !
containment high pressure onl l l          1 l  d. Low pressurizer pressure, steamline low pressure, containment    !
j  high pressure, and manua !
!          !
l
          ;
QUESTION: 012  (1.00)
l l- Which ONE of the following curves (shown on the following page) is    ,
'
indicative of an under compensated Intermediate Range Nuclear    !
Instrument?        .
a. Curve I l
b. Curve II c. Curve III
{
;  d. Curve IV
  *
!
!.
,._ . . . , - -
 
_ . - - . _- -. -..-. . - - . . . . ~ . . ~ . _ . . . _ . . ~ - - . - . - _ - . _ . . _ . . . . . - - . _ . . ~ . . . - -
a
;
          /!&
fY i
5 %/
,
e i
!
l i
a 10 4 1 i
i
.
10 5 Y i
1  G  ELECTRICALLY AOJUSTED COMPENSATED 13N CHAMBER
-
  $
w 104 I  Q.
 
*
k    TYPICAL SHUTDOWN CURVES m
J C 10 7
 
E o
a
-
w
  <
i  E i
'  e 104 W
w z
i b
w 104 E
a
 
U IV l  $ 1010 i            i ll1
~
l N      i
}  10 11      -
            )
      %  NEUTRON SOURCE LEVEL
!            j ll  g    l
  .
 
-
10 12  . .  . .  \-
 
        .
          .
2 4      .
          .
6  8 10 12 14 s
16 18 20 22 24
,
1    TIME
    * AFTER REACTOR SHUTDOWN (MINUTES)    l
-            l
            '
l
 
1
.
i  TYPICAL GAMMACOMPENSATED CURVE FOR A COMPENSATED lON CHAMBER    !
 
  -
 
i e
..
''.
 
<
,
 
i I
SENIOR REACTOR OPERATOR    Page 12 l l
i i
I l
QUESTION: 013 (1.00)
Which ONE of the following describes why the Source Range detectors are located at the lower quarter of the core rather than at core centerline?
a. The instrument tubes are angled in toward the core, so this location places the detectors closer to the flux they are to detect, b. Neutron flux is greater in the bottom half of the core during a startu c. To allow a spare detector to be installed in the upper half of the instrument should the installed detector fai d. This position provides optimal cooling for the detector by natural circulation of air in the detector wel QUESTION: 014 (1.00)
Given the following Unit 2 plant conditions:
- Unit is at 90% powe Rod control in automati Power Range NI N41 fails HIGH (top of scale).
 
Which ONE of the following describes the response of the control rods?
(Assume no operator action).
 
a. Rods initially drive IN and then drive OUT to match Tavg with Tref, b. Rods initially drive IN and then STOP and remain at that position, c. Rods initially drive OUT and then Drive IN to match Tavg with Tre d. Rods initially drive OUT and then STOP and remain at that positio .
 
SENIOR REACTOR OPERATOR    Page 13 QUESTION: .015 (1.00)
Given the following Unit 1 plant conditions:
      '
- Reactor is shutdow , Reactor-decay heat is being removed by natural circulatio RCS pressure is 1550 psi Average core thermocouple temperature is 402 degrees Which ONE of the following describes the approximate amount of subcooling.that exists in the RCS?
a. 50 deg b. 100 deg c. 200 deg deg F.
 
QUESTION: 016 (1.00)
      '
Which ONE of the following describes the location of the core exit thermocouples?
a. They are arranged in a horizontal plane just above the upper support plat b. They are arranged in strings extending from the upper support plate to the core mid-plan c. They are arranged in a horizontal plane just above the upper core plat d. They are arranged in strings extending from the upper core plate ,
to the core mid-plan l
..
      ;
_ - _ ._ . . _ . - __
 
SENIOR REACTOR. OPERATOR    Page 14 QUESTION: 017- .( 1. 0 0 )
Which ONE of the following conditions is the reactor vessel cooling fans (cavity cooling and control rod shroud fans) used to mitigate during a natural circulation cooldown in accordance with EOP-0.2, " Natural Circulation Cooldown"?
a. Creep of the reactor vessel head flange bolts, b. Formation of a steam bubble in the reactor vessel hea c. Catastrophic failure of the reactor vessel flange o-ring d. Damage to'the Core Exit Thermocouples's electrical circuitry from overheatin ,
i i
 
QUESTION: 018 (1.00)    ,
I Which ONE of the following conditions will result in an automatic trip '
of all operating Condensate Pumps?
      ]
a. Steam generator feed pump suction pressure at 179 psi ;
b. Containment pressure at 5.2 psi c. Steam generator level at 73%.
d. Condenser hotwell level at 10 inche i
 
. _ _ _ . _ ___. _ .. _ _. _ ._ _ . _ . - _ _ . _ . _ _ _ _ _ _ _ . _ . _ . _ . _ _ __ _ .. _ _ . . _ _ . . _ _ _ . _ _ _ _ . _
b
$ SENIOR REACTOR OPERATOR      Page 15 i
i
 
i QUESTION: 019  (1.00)
i Which ONE of the following conditions must be met in order to reopen a i bypass feed regulating valve from the control room after a valid
:
,
automatic isolation from Hi-Hi S/G level has occurred? a Momentarily depressing the feedwater control valve bypass reset i  pushbutton (FWCV BYPASS) once the closure signal has been j  cleare Locally reset the closing solenoid valves once the closureL
. signal has been cleared and reset.
 
i'
c. Go to manual on the feedwater control valve controller and open  i the valv l
 
l
        '
l  d. The condition which caused the isolation signal must be restored j  to its pre-trip valu QUESTION; .020  (1.00)
Which ONE of the following components is the largest supply of gas to    j the Waste Gas System during normal full power operations?
a. Volume Control Tank b. CVCS Holdup Tank Pressurizer Relief Tank d. Reactor Coolant Drain Tank      ;
i O
 
    . .  .- - .
SENIOR REACTOR OPERATOR    Page 16 QUESTION: 021 (1.00)
The high pressure sensing line on the RCS loop B flow instrument FT-414 break Which ONE of the following describes the resulting RCS flow indication?
l a. Only one (1) RCS loop B flow instrument is affected (PT-414) !
with HIGH flow indication, b. Only one (1) RCS loop B flow instrument is affected (FT-414)
with LOW flow indicatio c. All RCS loop B flow instruments are affected (PT-414, FT-415, FT-416) with HIGH flow indicatio d. All RCS loop B flow instruments are affected (PT-414, FT-415, FT-416) with LOW flow indicatio QUESTION: 022 (1.00)
Which ONE of the following is indicated if a light is ILLUMINATED on the SI/ Spray Ready Status Panel?
a. The component has lost AC powe b. The component is in an abnormal alignmen l c. The component has lost DC control powe I d. The component is running / energize !
l
)
e i
l
 
    . SENIOR REACTOR OPERATOR    Page 17 I
QUESTION: 023 (1.00)
Which ONE of the following "A" SI Accumulator parameters needs to be corrected for the "A" SI Accumulator to be OPERABLE while the reactor is operating at 100% power?
a. Water Volume is 1130 cubic fee b. Pressure is 730 psi c. Boric acid concentration is 1910 pp d. Outlet Isolation valve is OPEN and the control switch RED position indicating light is NOT LI QUESTION: 024 (1.00)
Given the following Unit 1 plant conditions:
-
Pressurizer pressure defeat switch is in its normal positio PT-449 (Yellow Channel), Pressurizer Pressure, has just failed LO Which ONE of the following describes the response of the pressurizer pressure control system to this failure?
a. Only PORV 431C cannot operate automatically (PORV 430 operable).
 
b. Only spray valve 431A closes (431B remains as is).
 
c. Both spray valves CLOSE fully, d. Both PORVs are PREVENTED from automatic operatio .
%
 
- - - . - . . - . . - . . _ _ . - . _ . . . . . . _ - _ - . - - . . . _ - . . -
        ,
SENIOR REACTOR OPERATOR    Page 18 I
QUESTION: 025 (1.00)
        ,
Which ONE of the following statements describes how plant operations are affected if Loop A RCS Wide Range Pressure instrument, PT-420, fails HIGH during Low Temperature Overpressure Mitigation System operation?
I
        '
a. Pressurizer PORV PCV-430 opens onl Pressurizer PORV PCV-430 opens and all pressurizer heaters deenergize onl ,
c. Pressurizer PORV PCV-430 opens and both pressurizer spray valves open onl Pressurizer PORV PCV-430 opens, all pressurizer heaters l  deenergize, and both pressurizer spray valves open.
 
i QUESTION: 026 (1.00)
-Given the following Unit 2 p1~ ant conditions:
!
  - Unit is at 50% powe '
  - Rod control is in MANUA Loop B cold leg temperature detector TE-401B fails high.
 
,  - No operator action is taken.
 
t l Mhich ONE of the following will be the steady-state pressurizer level?
l % %
c. 46% %
!
t.
 
,
:
I i
  .
!
,
    .-.
 
  . . - - - - . . . - . . . - . . . - - - . . - . ~ _ . _ . - - . . , ~ . - - . . _
' SENIOR REACTOR OPERATOh      Page 19 QUESTION: 027 (1.00)
Which ONE of the'following conditions on LT-428 would result in an    I increase in indicated pressurizer level?
a. A leak in the reference leg of the pressurizer level transmitter, LT-42 Pressurizer liquid temperature increase c. The reference leg for LT-428 cools down due to a decrease in containment temperatur Containment pressure increases to 0.3 psig; containment temperature remains constan !
QUESTION: 028 (1.00) .  [#  gj Which ONE of the following is the reason for t  second time delay associated with the AMSAC actuation relay?
a. Permits operator override  the event of spurious actuation, b. Prevents spurious tuation caused by S/G shrink and swel c. Allows ti for reactor power to decrease below 40% before actua llows time for the AFW pumps to recover S/G levels before actuatio QUESTION: 029 (1.00)
Which ONE of the following RVLIS readings indicates the highest probability of core voiding?
a. Wide Range reading 98 ft. with NO RCPs runnin b. Narrow Range reading 38 ft. with NO RCPs runnin c. Wide Range reading 65 ft. with ONE RCP runnin d. Narrow Range reading 95 ft. with BOTH RCPs runnin .
 
      'l
      .
SENIOR. REACTOR OPERATOR    Page 20
      ;
l QUESTION: 030 (1.00)
Given the following Unit 1 plant conditions *    '
I
- Unit is operating at 75% steady state powe All systems are in automatic contro The."A" S/G atmospheric valve fails ope Which ONE of the following-describes the plant response to this condition? (Assume'no operator action is taken.)
 
a. Turbine load decreases by 5%, reactor power remains stable at 75%.
b. Turbine governor valves open in response to lower steam header-pressure to increase turbine load to 80%.
c. Control rods insert to maintain reactor power at 75%.
d. Control rods withdraw and raise reactor power to 80% where it stabilizes.
 
QUESTION: 031 (1.00)
Which ONE of the following describes the normal, emergency, and alternate emergency sources of power to Safeguards 4.16 KV bus 1A06?
Normal Emergency  Alternate Emergency
______ _________  _________--________
a. Load Bus 1A04 EDG G03  EDG G04 b. Load Bus 1A04 EDG G01  EDG G02 c. Load Bus 1A03 EDG G01  EDG G04 d. Load Bus 1A03 EDG G03  EDG G02
.
. - , - -  - - - , , , - , ,  - - _- .c -
      .% , rey
 
.~. -. - ... - ..
SENIOR REACTOR OPERATOR    Page 21 i
;
QUESTION: 032 (1.00)
a Which ONE of the following will provide an automatic shutdown of the gas turbine while operating in local control?
'
a. Low axial compressor suction pressure b. Low lube oil level f Low lube oil pressure
,
d. Low control air supply pressure
, QUESTION: 033 (1.00)
:
Which ONE of the following is the maximum rated run time for Emergency
-
Diesel Generator G01 loaded at 3050 KW?
a. 30 minutes, j hour.
 
j c. 24 hours.
 
, hours.
 
l
      !
;
.
QUESTION: 034 (1.00)    '
Which ONE of the following RMS displays must be selected to access alert and alarm setpoints?    )
l
'
a. Status Grid
 
;
b. Sector Display
!
l c. HDSR d. Trend
 
'
,
I J
i
 
      .
SENIOR REACTOR OPERATOR    Page 22-QUESTION: 035 .(1.00)
Given the following Unit 2 plant conditions:
- RHR is in servic !
- RCS pressure is 320 psig and INCREASIN l
- RCS temperature is 340 degrees F and INCREASIN ALL system lineups are in a normal shutdown configuration for solid plant operatio Which ONE of the following will act FIRST to prevent overpressurizing the RHR System?
a. Pressurizer PORVs will ope b. RHR return isolation valve will auto close, c. RHR pump hot leg relief valve RH-PCV-861C will ope d. RHR pumps discharge relief valve RH-PCV-861A will ope QUESTION: 036 (1.00)    )
-Which ONE of the following conditions will ARM the Turbine Crossover ;
. Steam Dump System?    !
a. Selecting MANUAL on the mode selector switch, b. AUTO selected on the mode selector switch and a 10% lord drop in less than 120 second c. Going to TEST on the individual valves on the back of C0 d. Reaching 90 psig in the crossover steam heade !
.
    , __
  ,, . - , ,
 
      !
SENIOR REACTOR OPERATOR    Page 23 QUESTION: 037 (1.00)
Which ONE of the following is a symptom of a stuck control rod that would require entry into AOP-6B, " Stuck Rod Or Malfunctioning Position Indication", following a transient?
a. Im individual RPI in 8 step disagreement with the bank demand locatio b. A variation in NIS instrumentation resulting in a quadrant tilt of 1.2%.
c. A variation in core outlet thermocouples of 8% relative to i symmetric thermocouple !
d. A variation in axial flux of 1.2% of axial peak at any location relative to symmetrical trace.
 
QUESTION: 038 (1.00)
Given the following Unit 1 plant conditions:
- A loss of all AC power has occurre Power has been restore The crew has transitioned to ECA ECA 0.1 directs restoration of RCP seal coolin Which ONE of the following is the method required to restore RCP seal cooling?
a. Seal injection flow is initiated to cool the seals to less than 150 degrees F, then CCW flow to the RCP is establishe b. CCW flow is initiated to cool the seals to less than 150 degrees l F, then seal injection flow is establishe c. Seal injection flow is initiated to cool the seals to less than 190 degrees F, then CCW flow to the RCP is establishe d. CCW flow is initiated to cool the seals to less than 190 degrees F, then seal injection flow is establishe !
l l
 
      !
I i
l
 
i
 
      !
i SENIOR REACTOR OPERATOR    Page 24 QUESTION: 039 (1.00)
Given the following Unit 1 plant conditions:
- A plant trip has just occurre control rods are stuck out of the core following the tri An emergency boration has been initiated by the resctor operator in accordance with EOP-0.1, " Reactor Trip Response."
 
'Which ONE of the following lists the MINIMUM injected volume of boric acid necessary to satisfy the required amount of boration?
a. 600 gallons ,200 gallons c. 2,400 gallons ,000 gallons J
QUESTION: 040 (1.00)
l Given the following Unit 1 plant conditions:
- Steam generator A is faulted due to a feed line break outside of containmen ;
- The crew is performing actions of EOP-2, " Faulted Steam Generator '
Isolation."
 
- The AFW system is in operatio Which ONE of the following actions concerning the AFW pumps is required by EOP-2?
i a. Shutdown the AFW Pumps immediately, b. Maintain at least 50 gpm AFW flow to each S/G with narrow range levels less than 8%.
c. Run the AFW Pumps only if less than 200 gpm is available to the S/G i Isolate the AFW Pumps from S/G A (steam and AFW flow).
 
..
  -c,e. - - -yr
 
SENIOR REACTOR' OPERATOR    Page 25 QUESTION: 041 (1.00)
'In accordance with EOP-1, " Loss of Reactor or Secondary Coolant", which ONE of.the following groups of parameters is required to be verified, in addition to Pressurizer level, prior to terminating SI flow?
a. RCS subcooling,_ secondary heat sink, and containment pressure b. Secondary heat sink, RCS pressure, and RCS subcooling c. RCS subcooling, RCS pressure, and containment radiation levels d. Secondary heat sink, containment pressure, and RCS pressure QUESTION: 042 (1.00)
Given the following Unit 1 plant conditions:
'
- Unit is at 100% powe Condenser vacuum is decreasin Which.ONE of the following condenser vacuum conditions will FIRST result ;
in the' loss of condenser steam dump availability?  j a. 25" Hg vacuum    i " Hg vacuum c. 21" Hg vacuum I
d. 19" Hg vacuum I
l
.-    l
 
_ _ , _ . _ m _ _ _ . _ . ___ . _ _ _ - _ _ _ _ . _ _ _ _ _ - ._ _- -
i.
 
l SENIOR REACTOR OPERATOR    Page 26
 
QUESTION: 043 (1.00)
'
Given the following Unit 2 plant conditions:
; - Condenser-vacuum is 25 inches Hg.
 
;
- Generator load is 250 MW.
 
I i Which ONE of the following is the MAXIMUM amount of time the turbine can
; be operated (Operational Back Pressure region curve is' attached.)?
j a. Operation is prohibited.
 
$- b.'10 minutes, i
c. 1 hour.
 
;
,
d. Operation is unrestricte r
'
l i
 
<
j ' QUESTION: 044 (1.00)
i Which ONE of the following concerns would an operator most likely be
.
confronted with during a total loss of AC power in excess of 7 hours?
l
'
a. Loss of secondary heat sink condition
 
'
b. An unmonitored release of radioactivity
      !
li c. Loss of containment integrity    i
*
d. A Steam Generator overpressurization condition i
s
?-
t
,      l
-
l
!
;
3      ,
*
,
i l
*
I i
!
s j
i
!
    ..
 
. ~ . _ _ _ _ . _ _ _ . . . _ _ . . _ _ _ _ . . _          . _ _ _ _ _ _ _ . .        . _ . _ . . _ -
POINT BEACH NUCLEAR. PLANT            AOP-5A
}    ABNORMAL OPERATING PROCEDURES            MAJOR i                Revision 4
. LOSS OF CONDENSER-VACUUM            August 21,1995
!.
f .
  % s FIGURE 1
          *
 
.
./
 
!
i i
RECOMMENDED OPERATIONAL REGIONS FOR NUCLEA a
e WITH 2 DOUBLE FLOW LOW PRESSURE ENDS. ,    ,
,
,    U ;m.._.._.._.._..........__ .. n%:::
                      ,
    :: m = ==n ; =- r: g ; . . n i n : n ''" ; a : ;: = :r _ . _ _ . . . . . _. . __ . . _. . . .. . . _ .. . . . .=n...n  un:n:==
                ---_ .=. . .: _~;;;.-
                  . . .:;._;4;;;g:::::,;.;;;nM;-*::::::;
                  .. _ . . . .    =;;nn
    =en      .n            . .
    :: = :== ::== =n :=:: r:=::=n: = = ==. . . =.  . = :- ; == =====::  ;;==:==:=n:n h' :a::n =;;n;;c;
              ====n  - = L n;;. =mgr :n:y;;-
n =. n..gn; : . . . ~. .._. _. .. .. . .N .:;L; ..
    =_. =. . . . :. r.. : ; .
    . n=m-        n = r=. m. .=rr. ; a_=.
                    -
w!HE!.c -
                      ...n= ::: = m un:.n.=4
                        =
n u n . r. . . . . . . _ . m. _=. . =. . . . n.==. . .
            . . .. .. ..... =. ...m
            ; :=. . . .u. . . . .E..m
                . . . . . =.
                := =m
                  . ;,_=.
                  .
m: n= : =, --- -  r .:.=. . . .=. ;.g;.:.-- ._._.=. ne: . .=. .. .n n :
                    ..-..m....=...
      .
          .g  m;,!=. ;;;, g;c;
                .
                  .;,=
Fi!P651~r ~        ;
                      . =i ai E g n; hie:EPf:E iE          g=li~i[
5 " ..,,, .1= :f:n:
    :n::n:=: : :_= . . . . . . . . n.n H n .n :::=3  = m : . .iE_njE...H..i!!m...Eii!ii: J E.ii.H. P' .
                    .
                      . --
                      .
aE
                      =.u . . m. F~rir .=:. .: = . .n a;;
    ::::= = ::; .
          . .:::.: n =::n:n::
r.; . . . . . ....
mum:m:-
t . . . . . . .
                      .
    ::=;.:=r:====.._.................  =::nnum  :m:.. n.:_ =n =n= n n:n n::M:.?= m ;===:::n n:n:n:nin:::=!r-
            .
                      === = im : = r = =
                ::=.
    :
4.;**Z:: ;' U.,.: ;...: :....: _U.C_
        . . . . . : : .: . n : o r = n n : =. = n n:=:;;:  m= =n::. n ::.r == :n=;:p!!;rn. =. m: i::E E :n ni!i m m = = n.==m=: ; ~ -- u n::
L.; : '' . J.'  . ". ':
      .. . G.Jn ::r.n:n n;;; m. . , .: n :;.: " ' . ' y . '; ' . . '. .'. O_,* . ;.* ".:* .* ' .' _'. :; : "n -
        . . . .
_ . . _ :.: :::..
            . . : : . : n :L.": n:' : : ' :2.::: . , ..;.. .n..: .;;;",, . , f_; ,.::::
              : ::;..:. .::. :.2  ;;;;;; ;, g :;;;; . *
                    - ..- . ''::. I..-:::.:'.:!n;:
                      -
      = : : =.Z ~=''. -' '. :=n:n:n-.
              .
              ::;
                  , . . . ,  -
T.-  .
    $n :z: n! =
      ":  !! -" *,I n -        .--
          .n k:' . ''.n: r
                  <.-
3  ~ . - H: :: : ::- n;, ;: ? =+ n = I E =. . : . : m 2. n' ; mi -'- :' .-.: n - _J --.- n - ;  . .- . . _ ._
      : ';; ;C'.nM:''.~ '';;*
    '      *.      -
4 == :: =O!  : *.'n":'. ; ' :n ' :=:20:
1: l'  .
            '. n' M ;';-~ ;' ::;' n; . . :; a b ; u : n LM  :n:r-: % .: ; *:: :  :
              , .
                  : an!!!CEUij :m 0Z=: M ={l::::!!m mm*n= ." ::: = ''" -=- = x =-'': -
g  u n a :: n n : r_ ==n n: n : = =. r.'..u..=..u. =..y;::;::;C:n"";L ;;f;;;::n: :::;;;n;;n::;;;
          '      :g:::
    :n.::: :::n: = n: m=m =r.-.  :::.3-
          .
m m m =n=:ne      i=n -. . . ' : = : : ' 2 : " . . = n" :.'-    mum " : n=::n:n  !!!!!:.=:.:!!iE-uniirEE!!.      : .: = :::= :==:=:=.
                        :-
          .
                        -
J hhi!!! iik'3
          ..:
            ==..:m := n :.cm.:n  n :: n =:;d
                .
                  =n
                    -
                    = = m = u:p==u  m=; ;.
O  n :m ==f r b kr:= 53E=Nib h. .-.  '            aiii7c =_----
m    C= 7 ~  ' k=/Ii . 2u m i'I! ;'li
          " m= := rt
          ''.'.I:[N!Ih!UiI
          ; ="m~ 2: - =  !.U:I=i'i"!EiiNiiiE!
u= = : en - m ; n h =*r : ''  A[i;UUj{,IiE~'m['.=''
                  =". . - .. . . : m' ' M =m=    ~
                      ; n ": ' '' n n =bn
                      - 'Ei: i -b O
r. =-": -::n.=.,...=...=...r
    . . _ . . . . .=. U    :
          ::m  ==:=~;--~~~~~        =  r n i =.
kNbkb.N lE2.)Eh @
      . .
        .=....:..n..~..-.=..l==.n.m_.:.m..t.=..,..=....:rc.'.t
            . .-        ----
n 4      J5 5 BEE. [555N 53!I'i;T;dt[i-
              .
                        'g'.
        . .is[EEii. 7 ! ' ~T'r ~' r M u" 2 =:= = =iiiUil.Gii!{ {iiI ~- -*3 ' n g:;'':"4. n ': n    . , . . ,;;
                        .- g-
                ..
3.- IE:fi!E  ::Ei2 ,g g  ::*=::r4 * *- -
    .
                    - -
Ig'-''.  """
0  c.=:= E
        .
      - t .u:
        . n.:.:./ n= : . : ' ~ b : ::: : .      M2= =2mui!!hE E': mEE    E E ~.i" h;E M.E"
                    ; m n n2 :::M" :: =
I  :=:.:M::==. = :=_h.* - . z: = a=w  ==. =_. n: m:            ''m n="
                    -
    =r : = = ~    =nm:            m m m g. = : n= m =:- m n= ==r=
g  *: ::::n;;::'
    ,
12Mm . * ,, .:'. a O=.
_-==.c ==: . . . . _ _ .
        :'"::=:-~r=.r=: =r=t:.Z'J:::- :- .;J.;t-n
            = : :;;.= :.: _ . . . . _ . . . . _
                    : = ; m nw n u M 2 2: = := :== -
E  **4
    ;;r"':;*nr
    . ;  J ".': =::2/2 a : Mm * R
          :
L**,-'j::'':*n ;- ::=
            .
                ;; =.u =: ::M==i:imiin nm  tia 1_1:: u ;;;im.a=;;ii.:di =:E:E    =
y *:r  . p. : -"-
      . g"" = ! ::u ", = u n= r m r- r' . :.r n: n. :::::: 3- :: = ---*"m*:  Jtunn=
              : :;--':.:.=
                : n:nm;:r:::;g;;;;- u n:n;; ;m :gg n n-. rg  .
  =  g p
      :; .
_n _:.': _.. J : ~~" n "* 1    ::*- u ': m :?:-  n --- ;:n **- :" .n- ::
            ::2:::*=':-':=n;rm::;a;;:;;::m    M:; - ;;:,n,..;;;;:;:E
                    -  : : m ;.- ;,; r .;-. ---- -
                        ;;;Jn;,=;..
                        .
n g;g=.g M p= =:<=r ::.* = MuMun =.n ..= =:':': =-"::; { = : = :._:
        ~ . .
            ::::  -
n g:;;n ;:; n :::-- . ~,, ~. ---  ; n ;;:;g;1 7:-; -;,f- ;,;;. ;; 7; nr;gr r ;. g; .;;- ;; =,
..
kl e p:rM
    .
M: ::
      =. ;.:n n mM um ::  =n::==
          .
          .:.::n t.:In::; :n:M
        :: n jM *':: M:' = *. = ::.c.M;a u = :: = :; = m =n m m = t- ;:n:ra r    == y= : =_2: .
                =M:M=M=-;
                = ::= ==: ==E.;.: :==.=2  g :  w-===.--
9 .:.:n M::M gn:;- :;---&g- M ;:n :: r n,:: =
      : . . : = :M = M =:=::=m=m n:=i  ::. m mm'.:::;r: . m :x: = n :: ::;:=%:.. nn a MU:C ;t n M .m := :u=::: =-  = ===_ == == u
  { ,::RM:n.=gC
    ..
M:M:**M:.  : C p: =::C::::L.T::::.L'n
      :' : : : : :: . ::: n : ':--" ::: ;
          : ^ '
Z ^ =::'
n:::M=_ . :'.y :=';-~ ' ,=-=: u u=:r = mi=!:==:==:
                %  m::= n a = :w =- a ::::=d-= cm x
g
    : == : M = : :
      :::::. :
      = = = : u u = =5 :. -- ;'.;L :' 21: g' :: C :::::;: ':;.:* :C W . . : ;:U n .:::::::. ::;;4 -.;:: -C n:mj:n::; :;;;.:t ; ;; a g r--
            -
              .
              .
                  : g..;:.;m :m : :;=---- .
m:=
                      ; -- qC =
N : 2M' :::n:::=== u="M::m:=::: 22 ;;.;. . g_ =. ..=_.M.:n =.sM:=na.M::n = M=::n=M:::=m!=: n ::::::::::;::::::;:-- . :mn Ed
            . . , . . . _
CD l em ==**"C'""
M'=-' :'r4.:::nC.R ~2---'n :::===n:M:  L' -- ;" ':n=:M  u r u n _..: =2; ;:m:::: =:=n::t 2::M..................................,.=.._..u=....=...2.==_=a=  .,;
          :
                =.= :.c = a: = r    . . _ . . . t==
u = =; c:m. : : ~~. = = '""' R = "=- .-^ m=n :' =Ly :;.;; n=M.=;; :;- I=;
          .
Mm M% *Mn!'' M :=:=2:
U;~,*,: MM:MMM;;;::~;n u. aW: '=.=,* . '.; :=: : :::':2 M:::::2'J " M = : : = :==            ,.- - r == ;=nf : _=; -=n : =-- =
u n:=M =
            . '*M'".m;:u' n:'-=. .=:::  = n = = ::: n M M = n = :-
M:.' Mn' ~  ::...            =__=_..._: t. M:::M=M:  m =...u .
        ::: = =;=4M= = m"'* u''
        .
          *1n:::::n%"-' 'n M!. nt:==:~M::n n:M C :ZM r.; M.' 2 =::'::n::
                . . ' : .i=::: 44M: :n  ;;;;;;: U
                    ;;
                    * C;:::n ;;"; :=,**'T*1:, =f **l"
                      . :;C n J' ': M ::::* M N  jt:.M=:n:  MU=:*MM::M::M:*:M:M:=":M
    . . . . _ ~ . . .... M:::nn:M=:n;:      ::r' = = : f...=MM: mM::;;= = = = =
                  .
                        . .:U M*M:::U:;n::22:    Z * :-1:= =.un  MMC;;"':  .~="%:M:::=:::::=M::',--  = ' .n = = =n'"= a n ==:::=-
m=:=: u;n =
    :M M::!M. l*M:: ':. : M"'M:M:F:Z:::n.:: M: .:Mn:M ;2'*! !M:::M:::U *n :;;;;n!!:U=j:::
          ;;"::6:n n'::f:::  V "-':* MM=*
              !::.
J :'mnM=::::::::::      nc* *- * . _ _ _ _
    , $555555*5!5-{:m::      ;.;;'MMMCM;;"Ma'.Z.:: lMnn:Mn :.:.10::!': '-M:nJ.::::;;:MM::nn::;;;;n::=;;;;M ::n;:n:=n :=_U a  M ::::::M::M  :M : : n ::M M M : : :: :: : M : :: = ::::::: n :: m M ;;;;.u n;m:::==M M M:M M MM;:nn        CJ==='; Mn"::M M : n M M
 
e        M;;;:mn:;n;;;;;;;M;;.~.;. u M.=:;L;.:.a 0  10 0  200  300 i
e  t    i  i 400  500  600  70c GROSS LOAD-MW 393A421
  .                T
          .pngcT6i7 REFERENCE USE
        .    .
: SENIOR REACTOR OPERATO 'Page 27 QUESTION: 045 (1.00)
Given the following Unit 1 plant conditions:
- A loss of all AC power has occurre ECA-0.0, " Loss of All AC Power" is in effec Per ECA-0.0, certain Engineered Safeguards equipment control switches are placed in Pull-ou 'Which ONE of the following events is prevented by this switch alignment?
a. An uncontrolled cooldown of the RCS and possible reactor restar b. An uncontrolled use of water that may be needed for long term cooldow c. Ar uncontrolled start of large loads on safeguards AC buse d. An uncontrolled depressurization of the RCS.
 
-QUESTION: 046 .(1.00)
Which ONE of the following describes the response of Feedwater Regulating Valves following loss of a RED or BLUE instrument bus?
a. May fail open; manual control will NOT be available to regain contro b..May fail open; but manual control will be available to regain contro c. May fail closed; manual control will NOT be available to regain contro May fail closed; but manual control will be available to regain contro .
 
SENIOR REACTOR OPERATOR    Page 28 QUESTION: 047 (1.00)
Which ONE of the following locations / equipment is protected by a Halon System?
a. Diesel Generator rooms b. Hydrogen Seal Oil package c. Service Water Pump area d. Auxiliary Feed Pump room QUESTION: 048 (1.00)
Technical Specifications require containment pressure to be between + psig and -2.0 psig during power operatio Which'ONE of the following is the MAXIMUM time containment pressure is allowed to be outside this, band per Technical Specifications before action must be initiated to shut down the plant?
a. 30 minutes hour hours d. 24 hours
.
 
. - . - . . . . . .~ .. -~-.-~-- - - . .
    - - . . . . . - - . . . . . . ..- - - .. .
SENIOR REACTOR OPERATOR      Page 29 QUESTION: 049 (1.00)
A LOCA has occurred and both RCPs have been started per CSP-C.1,
" Response to Inadequate Core Cooling."
 
Given the below list of criteria:
I. Core cooling provided by low or high head SI II. Narrow range reactor-vessel level greater than 25 feet III. Any RCS hot leg loop less then 350 degrees F IV. Core exit thermocouples less than 1200 degrees F Which ONE of the following gives the three conditions from the above list that allow securing of the RCPs per CSP-C.17 I, II and III I, II and IV
' I, III and I II, III, and I QUESTION: 050 (1.00)
Given the following Unit 1 plant conditions:
  - The Reactor tripped from 100% powe EOP-0, " Reactor Trip or Safety Injection" has been entere The main turbine did NOT trip as expected.
 
l  - MANUAL turbine trip is unsuccessful.
 
!
Which ONE of the following is the NEXT action required per EOP-0?
a. Manually close MSIVs and Bypass valve b. Locally trip the main turbin c. Go to CSP-S.1, " Response to Nuclear Power Generation /ATWS."
 
d. Verify Safeguards Buses energized.
 
l l
 
i
  . -  . . -
 
_ . . . ._ _ _ _ . = . _ _ . . . _ _ _ _ _ . _ _ _ _ , , _ _ _ - . . _ _ _ _ _ . _ . - _ _ . _ . . . . . . .
_
        . . . _ . . . . _ _ _
}
i j SENIOR REACTOR OPERATOR      Page 30
 
!
!
:
; QUESTION: 051 (1.00)
.
'
Which ONE of the following is the reason for the main feedwater i: isolation following a reactor' trip?
a. To prevent thermal shocking of the steam generator (S/G)
;  tubesheet ~
j  b. To prevent excessive cooldown of the RC c. To. prevent S/G overpressurization.
:  d. To preserve secondary water for a subsequent RCS cooldown.
 
.
i I
I i
QUESTION: 052 (1.00)
l Given the following Unit 1 plant conditions:      l
 
  - A reactor trip and safety injection has occurre l
  - Pressurizer ~PORV 1RC-431C lifted and is stuck ope l l
Which ONE of the following is the MAXIMUM pressure below which the PORV    i isolation valve must be shut?      !
a. 2450 psig b. 2335 psig c. 2225 psig d. 2010 psig
 
I
          .
  @
 
~ . . _ _ . . _ . _ . ._. . . .. . . _ . _ __ _.._ ._ _ _ . . __. ._ _ . _ _ . . . .
I SENZOR REACTOR OPERATOR    Page 31 QUESTION: 053 (1.00)
Given the following Unit 1 plant conditions:
- Unit has tripped'from 100% due to a small break LOC Conditions have stabilized and operators are evaluating the criteria for terminating S Adverse Containment conditions do NOT exis A maximum of 50 gpm feedwater flow is available to each S/ Which ONE of the following conditions would PREVENT SI termination per EOP-1.2, "Small Break LOCA Cooldown and Depressurization"?
a. Both steam generator levels indicate 5% N b. Pressurizer level indicates 12%.
c. RCS subcooling is 40 degrees d. Pressurizer pressure is 2050 psig.
 
QUESTION: 054 (1.00)
In procedure OP-5A, " Reactor Coolant Volume Control" there is a PRECAUTION that states:  "Do not secure letdown flow without also securing charging flow ..."
Which ONE of the felic  ng statements describes why charging flow should  !
also be isolated? (Assume all systems are in a normal at power lineup.)
 
a. VCT level will decrease until charging pump suction shifts to .
i the RWS ;
 
b. Reduce thermal shock on the Non-Regenerative Heat Exchange l i
        '
c. VCT level will decrease causing possible damage to the charging pump d. Reduce thermal shock on the charging penetration into the RC ,
.
  .  . .w =+
 
SENIOR REACTOR OPERATOR    Page 32 QUESTION: 055 (1.00)
Given the following Unit 1 plant conditions:
- Unit is in Cold Shutdown with RHR Cooling in progres The RCS is soli RHR flow is lost and CANNOT be restore '
- All other systems and components are availabl Which ONE of the following methods of cooling will be utilized to remove the core decay heat?
a. Feed a S/G using an AFW pump, and bleed steam through the respective S/G atmospheric relief valv b. Start a charging pump, with flow through an RHR heat exchanger, and initiate Hot Leg Injectio c. Feed the RCS with Safety Injection, use letdown to remove decay hea d. Start a charging pump,.with flow through an RHR heat exchanger, and initiate Cold Leg Injection.
 
QUESTION: 056 (1.00)    l Given the following Unit 1 plant conditions:
- A reactor trip occurred 30 minutes ago due to a Steam Generator tube ruptur The crew is taking the actions of EOP-3, " Steam Generator Tube Rupture".
 
- N-35 indicates 1.4E-10 amps and N-36 indicates 1.6E-10 amp Source Range Instruments are NOT energize Which ONE of the following actions is required by the Reactor Operator? !
a. Allow N36 to decay to less than 1.5E-10 amps which will automatically energize the SR instrument b. Depress both SR RESET pushbutton c. Depress both IR BLOCK pushbutton d. Deenergize N3 ..
    . , .
 
  , . .- . .  .
SENIOR REACTOR OPERATOR    Page 33
.,
QUESTION: 057 (1.00)
'
Which ONE of the following conditions requires tripping the reactor per AOP-1A, " Reactor Coolant Leak", for a confirmed steam generator tube i
'
leak? Assume appropriate actions have been taken per AOP-1A regarding charging and letdown system operation _
4 a. Pressurizer level is at 11% and decreasin l    .
b. Narrow range S/G 1evel is 62% and increasin c. VCT level is at 17% in AUTO control and decreasing.
 
j d. S/G Blowdown Isolation valve closes due to high radiation.
 
QUESTION: 058 (1.00)
The crew is responding to a ruptured tube in 1B steam generator (S/G)
using EOP-3, " Steam Generator Tube Rupture."
 
Given the following plant conditions:
- 1A S/G pressure is 950 psi B S/G pressure is 1050 psi Which ONE of the following is the required core exit temperature that the RCS must be cooled down to prior to depressurization? (Page 15 of EOP-3 attached).
 
a. 505 degrees F b. 490 degrees F j c. 485 degrees F
,
d. 480 degrees F
,
i .
j
 
  -      i
.      l
      \
' POINT BEACH NUCLEAICPLANT
  "-EMERGENCY OPERATING PROCEDURES EOP-3 Unit 1 MAJOR Revision.1 ~
STEAM GENERATOR TUBE * RUPTURE-    j
/,      October;26, 199 UNIT-1      ,
STEP ACTION / EXPECTED RESPONSE    - RESPONSE NOT OBTAINED CAUTION:
(Step) THE FOLLOWING STEP MAY THAN 100 F IN ONE HOUR. REQUIRE A COOLDOWN NOTE:
(Procedure) controlled cooldown has been initiatedRCP trip criteria
    .
 
Initiate RCS Cooldown Using ruptured steam generator pressure, detennine target 2 core ~ exit temperatur *
__
_ _ _ _ _
Ruptured Steam Generator Pressure (psig)  Core Exit Temperature (*F)
1100 1000  (460) 505 900  (445] 490
~gg  800  (435] 480 700  (420] 465 600  [405] 450 500  (390) 435 400  (370) 415 340  (350) 395 (335) 380  ,
      '
  .
l
      ,
l i
I
  . .
  . Step 20 Continued on Next Page . . ,
l
_
O
 
    'p=-c >,oc a
    <x
    ,_
  >C- 's = e c2_
    -
_ , _
 
  . -. ._ .- - - - .
    - - - .
      ,
l1 SENIOR REACTOR OPERATOR    Page 34 j l
i
*
l QUESTION: 059 (1.00)
Which ONE of the following statements explains why AFW flowrate is procedurally restricted to 100 gpm when recovering steam generator (S/G) !
level if the level has fallen below 55 inches on the wide range q indication?
l
, a. To minimize water hammer to the S/G feed rin b. To prevent reactor restart from an excessive cooldow c. To minimize thermal stresses to S/G. components.
 
#
d. To prevent exceeding reactor vessel cooldown rate limit.
 
d QUESTION: 060 (1.00)
Which ONE of the following is the basis for stopping all ' cps in CSP-H.1, " Response to Loss of Secondary Heat Sink"? It establishes natural circulation to enhance the bleed and feed )
capability of safety injectio I i
b. It extends the time available to restore feed flow before bleed and feed criteria is me I It anticipates an RCS pressure decrease caused by spray valves !'
opening when air is restored to containmen It anticipates an RCS pressure decrease caused by opening PZR PORVs during bleed and fee .
 
  . _ _ _ _ _ _ . . . . _ - .. . ._
i
,
SENIOR REACTOR OPERATOR    Page 35
: QUESTION: 061 (1.00)
I Loss of which ONE of the following distribution panels will result in a i
<
DUAL PLANT TRIP?
a. D13    l
_
, D16
.
c. D18 d. D21 s
a QUESTION: 062 (1.00)
 
,
With the Pressurizer Level Control Selector Switch in the NORMAL position, a pressurizer level instrument failure causes the following l SEQUENTIAL plant event Charging flow is reduced to minimu Pressurizer level decrease Letdown flow is secured and heaters turn off.
 
,
- Pressurizer level increases until high level trip occurs.
 
,
Which ONE of the following instrument failures could have occurred?
(Assume NO operator action) Pressurizer level channel 428 (Blue) failed low Pressurizer level channel 428 (Blue) failed high Pressurizer level channel 427 (White) failed lo I Pressurizer level channel 427 (White) failed hig ..
  - - - .
 
. _  _ ._._.. _ . . .__ _. . . - - - _ _ _ . . . . _ _
      ]
      !
SENIOR REACTOR OPERATOR    Page 36 QUESTION: 063 (1.00)
Given the following Unit 2 plant conditions:
- The plant has tripped from 100% power due to a loss of off-site powe The crew is required to verify natural circulation ~^in EOP-0.1,
  " Reactor Trip Response."
 
Which ONE of the following parameters satisfies one of the criteria for  i
      '
indication of natural circulation?
a. Steam generator pressures slowly trending upward.
 
l b. RCS Cold leg temperatures are increasing, c. RCS subcooling is 36 degrees F.
 
.
d. Core exit thermocouples at saturation temperature for steam
!  generator pressure.
 
I
!
l QUESTION: 064 (1.00)
Which ONE of the following " Procedure Usage Levels" allows performance og all activities from memory?
a. Infrequent Use Information Use c. Reference Use d. Continuous Use l
l l
.
*
?
 
J
 
SENIOR REACTOR OPERATOR    Page 37 QUESTION: 065 (1.00)
The applicability statement for Procedure EOP-0, " Reactor Trip Or Safety Injection", indicates that this procedure is used for initiating events occurring where RCS hot leg temperature is greater than or equal to 350 degrees Which ONE of the following describes the applicability of EOP-0 when RCS temperature is below 350 degrees F and a LOCA occurs?
a. EOP-0 cannot be used, procedure EOP-0.0, "Rediagnosis" is to be use b. EOP-0 cannot be used unless directed by the Critical Safety Function Status Tree c. EOP-0 cannot be used unless a step-by-step evaluation is made to determine if each action is still applicabl EOP-0 cannot be used, Shutdown Emergency Procedures (SEPs) are to be used.
 
QUESTION: 066 (1.00)
Which ONE of the following is the MINIMUM number of shifts, per 10 CFR 55, on which you must actively perform operator or senior operator duties to maintain your license in an ACTIVE status? (ASSUME 8 hour shifts.)
 
a. 5 shifts per calendar quarter b. 5 shifts per calendar year c. 7 shifts per calendar quarter shifts per calendar year
.
 
SENIOR REACTOR OPERATOR    Page 38 QUESTION: 067 (1.00)    i
      !
Which ONE of the following is the LATEST time that alcohol consumption 1 would be allowed prior to assuming the watch at 1800 per NP-1.7.5, l
" Fitness for Duty Policy and Procedure"?
I a. 0600 b. 1000 c. 1300 d. 1500 l
 
QUESTION: 068 (1.00)    j
 
A bomb threat has been received and the Duty Shift Superintendent (DSS)
has initiated a plant evacuation. A suspicious device has been discovered in an emergency diesel generator (D/G) room by the extra reactor operato Which ONE of the following states the actions to be taken by the individual discovering the device, in accordance with NP 1.7.10?
a. Immediately notify the DSS by two-way radio and then leave the D/G roo Immediately notify DCS by two-way radio and then leave the D/G roo c. Leave the D/G room, and then notify the DSS by phon d. Leave the D/G room, and then notify DCS by phone, t
      .
    .
l l
l
 
SENIOR REACTOR OPERATOR    Page 39 QUESTION: 069 (1.00)
Which ONE of the following is the preferred method to rapidly reduce turbine load in response to a casualty per the Managers Expectations?
a. Operator Automatic Load Rate Contro b. Turbine manual using governor valve decrease, c. Valve Position Limite d. Turbine manual using Governor fast.
 
QUESTION: 070 (1.00)    i Which ONE of the following is an individual who is authorized to be a member of the fire brigade and to perform operations health physics functions?
a. Unit 1 Control Operator b. Unit 2 Control Operator c. Duty Technical Advisor Extra Operator QUESTION: 071 (1.00)
Which ONE of the following describes the process of determining an instrument's accuracy by comparing the indication to other independent instrument channels measuring the same paramete a. Channel Calibration b. Channel Functional Test c. Channel Check d. Channel Verification
      !
      !
      !
 
i i
      '
SENIOR REACTOR OPERATOR    Page 40 l
QUESTION: 072 (1.00)
I An oncoming Reactor Operator has worked the following schedule:
Day: 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Hrs: 12 8 12 8 8 8 8 12 8 12 12 12 _12 ?
Which ONE of the following is the MAXIMUM number of hours the individual may work on day 14 without obtaining special authorization assuming the operator has a minimum of 8 hours off between each shift?
a. 2 hours hours hours hours    j
      !
I QUESTION: 073 (1.00)
Which ONE of the following is correct regarding INDEPENDENT verification of control board sliders?    l a. Must be done by the Unit Control Operato '
b. Must be done by visually observing the operating lin c. Can be accomplished using status light d. Can be accomplished using remote indicator ,
p
 
  .
__. ,
h i
I l SENIOR REACTOR OPERATOR    Page 41
!
l QUESTION: 074 (1.00)
'Which ONE of the following conditions is required in order to perform j
. work on a system using " Positive Control" in lieu of danger tags?
 
a. A QC certified inspector is presen l l
b. Work is1to be performed in a high radiation are l c. Requires direct line of sight of the isolatio l
      ;
d. Isolation boundaries are locke j
      !
    !  .
QUESTION: 075 (1.00)
      ;
Given the following Unit 1 conditions:
- Unit is at 100% powe l
- VCT level is 23%.
- All controls are in automati LT-112, VCT level transmitter, fails hig Which ONE of the following describes the FINAL ACTUAL VCT level? (Assume no. operator action.)
 
l a. Increases to 78% and then diverts to the HU I Increases to 28% where auto-makeup stops, c. Decreases to 17% where auto-makeup initiates, Decreases to 0% (empty).
 
.
 
  - - _ - _ ... _ . _ ._ . . - -. . _ . _ .
      ,
l SENIOR REACTOR OPERATOR    Page 42
      ,
QUESTION: 076 (1.00)
Given the following plant conditions:
- Both units are operatin The South condensate Storage Tank (CST) is isolated for repair The North CST is selected for Auxiliary Feedwater Pump suction at both unit Which ONE of the following is the MINIMUM volume required in the North i CST?
;
a. 13,000 gallons ,000 gallons c. 26,000 gallons ,000 gallons
,
QUESTION: 077 (1.00)
*
i Which ONE of the following is the reason the Containment Purge Air i Supply and Exhaust Valves are required to be locked closed during l
operations at power?
I a. The valves are NOT seismically qualified to operate during a
,
design basis earthquake.
 
.
b. The valve actuators do NOT have class 1E penetration conductor
; overcurrent protection devices.
 
t
'
c. The valves capability to close during a design basis loss-of-coolant accident has NOT been demonstrated, d. The related piping systems outside containment are NOT seismically qualified.
 
)
i
.
 
      ..-. - . . .
      !
P SENIOR REACTOR OPERATOR    Page 43 QUESTION: 078 (1.00)
Given the following Unit 2 plant conditions:    .
- Operating at 100% powe All systems are operabl l
- While in AUTO rod control, Control Bank "D" starts ~ stepping in  !
slowly, but at a noticeable rate.
 
-Which ONE of the following events will cause this response?
a. A tube leak in the Regenerative Heat Exchange b. A tube leak in the Seal Water Heat Exchange c. A tube leak in the Non-Regenerative Heat Exchange d. A tube leak in the Excess Letdown Heat Exchanger.
 
QUESTION: 079 (1.00)
Which ONE of the following prevents inadvertently raising an irradiated fuel assembly in the New Fuel Elevator?
a. Manipulator crane is interlocked to prevent loading irradiated fuel assembly into the elevato i b. The elevator will not rise if any Spent Fuel Pool area radiation monitor alarm c. Operators verify the elevator basket is empty before raising the elevato l d. The elevator will not rise if any fuel element is in i ..
  , , - -
 
.. _ _  __ _ _ _ _ _ . . _ _ _ .. _ ._
SENIOR REACTOR OPERATOR    Page 44 i
l QUESTION: 080 (1.00)'
      ,
l Given the following Unit 2 plant conditions-  l
 
- Reactor power is 1215 MWt
- RCS pressure is in the normal operating ban Which ONE of the following RCS average temperatures would FIRST result in exceeding a safety limit? (Tech Spec Figure 15.2.1-2 attached).
 
a. 590 degrees F b. 600 degrees F c. 610 degrees F d. 620 degrees F QUESTION: 081 (1.00)
Which ONE of the following states the MINIMUM number of operable 4 containment fan cooler units required to ensure containment integrity !
following a Limiting Design Basis accident if ONE Containment Spray Pump is inoperable? d. 4 l
@
 
_ _ _ . . _ . . . __ _ . .  . . _ _ . _. _ _ _ _ _ _ _ . _ _ _ . _ _ . _ . . _ _ _ _ _ . . . _
.
i    Figure 15.2.1-2
;
'
REACTOR CORE SAFETY LIMITS
- . .
POINT BEACH UNIT 2 s
:  670
 
-
660 -
.
:  650 l      2425 psia
!
;  MO -
;
 
2250 psia
. 630 -
<
Q h
 
'
3 620 -
u
  ,
y    2000 psia
!
i b610 o
  -
i j  600 -
1775 psia j  590 -
 
;  580 -
 
,
i i
;  570 -
?
.
!,    ' ' ' ' ' ' ' '
'
560      '  ' '
'
.
0 0.1 0.2 0.3 0.4 .6 .8 .1 1.2
:
.
Core Power (fraction of 1518.5 MWt)
; ~
nG-l j
Unit 2 - Amendment No. 169
-
November 17, 1995 a
 
  . . - . - - - . . . . . _ _ - .
SENIOR REACTOR OPERATOR    Page 45 t
QUESTION: 082 (1.00)
Given the following Unit 2 plant conditions:
- Unit is at 50% powe Control rods are in AUTOMATI Which ONE of the following instrument malfunctions would result in a CONTINUOUS rod withdrawal?
a. Loop "A" Thot RTD fails LO . Power range channel N-42 fails LO c. PT-485 fails HIG i
      '
d. Loop "B" Tcold RTD fails HIGH.
 
>      l
*
QUESTION: 083 (1.00)
Given the following Unit 1 plant conditions:
 
- A Control Bank D rod was dropped and recovere The Pulse to Analog Converter was NOT reset as required by
; AOP-6A, Dropped Ro l I
Which ONE of the following will occur on the next rod movement?
a. If control rods are inserted, the Rod Insertion Limit Alarm will be received at a lower rod position than required, b. If control rods are withdrawn, Overtemperature Delta T will NOT stop Control Bank D withdrawal when required.
 
c. If control rods are inserted, Bank C control rods will begin insertion at a lower value of Control Bank D positio If control rods are inserted, Bank C control rods will begin insertion at a higher value of Control Bank D positio .
 
4
 
i SENIOR REACTOR OPERATOR    Page 46 QUESTION: 084 (1.00)
EOP-1, " Loss of Reactor or Secondary Coolant", Step 1: " Check if RCPs should be tripped", is a continuous action ste Which ONE of the following is the basis for continuously monitoring for the criteria to perform this step in response to a LOCA?
a. To minimize RCP run time with less than the required amount of subcoolin b. To prevent RCP damage from cavitation due to operation in a two phase syste c. To minimize cooldown rate if a main steam line break is in progres d. To prevent excessive RCS inventory loss and a potential inadequate core cooling condition.
 
QUESTION: 085 (1.00)
Which ONE of the following is the MAXIMUM allowable time to actuate AFW following an ATWS with a loss of feedwater?
a. 30 seconds seconds c. 120 seconds d. 180 seconds
 
SENIOR REACTOR OPERATOR    Page 47 QUESTION: 086 (1.00)
Which ONE of the_following Control Room personnel should proceed to the scene of a fire?
a., Duty Shift Superintenden b. Duty Operating Supervisor, c. Duty and Call Superintenden d. Duty Technical Adviso ,
QUESTION: 087 (1.00)
Which ONE of the following is the proper sequence of major actions for removing decay heat from the core per CSP-C.1, " Response to Inadequate Core Cooling"?
a. Reinitiation of high pressure safety injection; open PORVs; RCP restart; rapid secondary depressurization b. Rapid secondary depressurization; reinitiation of high pressure safety injection; open PORVs' RCP restart c. RCP restart; reinitiation of high pressure safety injection; rapid secondary depressurization; open PORVs d. Reinitiation of high pressure safety injection; rapid secondary
'depressurization; RCP restart; open PORVs _ . . . . _ . _ _ _ _ _ . _ _ . _ . . _ . - _ _ _. . _ _ m._ .._.m - _ . - . . , _ _ . _ _ _ _ . . _ _
i
 
i Page 48
] SENIOR REACTOR OPERATOR
 
;
        -
J
' QUESTION: 088 (1.00)
j Which ONE of the following is the reason'that Technical Specifications
' - ' requires Tavg to be decreased below 500 degrees F after the reactor is
-
chut down for excessive Reactor Coolant Activity?
a.' Reduce the severity of a possible pressurized thermal shock i  condition by limiting the amount of cooldown that can occur.
 
i
: b. Minimize potential for containment contamination from
:  inadvertent PORV operation.
 
.
l c. Prevent uncontrolled release of radioactivity if a steam i  generator tube ruptures.
 
i
! d. Ensure reactor stays shut down to minimize release of additional i  ' fission products into the coolant.
 
;
.
i
! QUESTION: 089 (1.00)
l The following boration flowpaths are listed in AOP-6E, " Alternate
>
Boration/ Loss of Shutdown Margin."
: CVCS and RWS I CVCS and Blende II CVCS and CV-350 (Emergency Boration Valve).
 
Which ONE of the following sequences is the AOP-6E order of preference for the above flowpaths (first choice, second choice, etc.) to establish alternate boration when required? II, I, III
        ! II, III, I III, I, II
. III, II, I
 
___ . - . _ . _ . _ . . _ . _ . . - . .._-.--.___.._.______-,_..._.___ . _ _ _ . _ . _
SENIOR REACTOR OPERATOR      Page 49 QUESTION: 090  (1.00)
Given the following containment history with a small LOCA in progress:
Tima Cnmt Tem Cnmt Press  Cnmt Humidity Cnmt Radiation 0815 '178.Deg. F
  '
2 psig  90% 9.0 x 10E2 R/Hr 0830 180 Deg. F  4 psig  .100% 7.3 x 10E3 R/Hr  1 0845 183 Deg. F  6 psig  100% 9.5 x 10E4 R/Hr 0900 185 Deg. F  8 psig  100% 2.0 x 10E5 R/Hr l
'
Which ONE of the following describes the EARLIEST time at which adverse containment existed?
i a. 0815        1
  -        - ) i '
QUESTION: 091  (1.00)
While discharging a Gas Decay Tank (GDT), operators are cautioned to
" Maintain gas decay tank pressure at or above 5 psig at all times."
 
i Which ONE of the following is the reason for this caution?    I a. Prevent condensation in the GDT.
 
!
b. Prevent exceeding the required discharge flow rat , Prevent oxygen inleakag \
, d. Prevent overheating Waste Gas Compressors.
 
!
!
(
l l
I e
i
)
i
-      -
 
SENIOR REACTOR OPERATOR    Page 50 QUESTION: 092  (1.00)
Which ONE of the following AREA radiation monitors has a control function?
a. RE-101, Control Room Monitor
  '*
  . RE-105, SFP Low Range Monitor
      :
      '
c. RE-108, Drumming Area Monitor d. RE-114, CVCS Holdup Tank Monitor
  .
QUESTION: 093  (1.00)
Which ONE of-the_following is the reason that EOP-3.1, " Post-SGTR  .
      '
Cooldown Using Feedwater" is the preferred procedure for Post-SGTR Cooldown?
a. Radiological releases and contamination of secondary systems are minimize b. Boron dilution of the RCS is minimized, c. Adverse secondary system chemistry conditions are eliminate d. RCS cooldown rate is maximize QUESTION: 094  (1.00)
During a core shuffle, which ONE of the following is correct regarding the MAXIMUM number of RCCAs that can be removed from the core each refueling and st111' ensure 5% shutdown margin is maintained? c. The number is determined by a Reactor Engineering Calculation d-. The number is determined by the Core Loading Supervisor
_ _ _ _ _ _ _ _ _ - _ _ _ _ _ .  . . - - .- - . - __
 
' SENIOR REACTOR OPERATOR    Page 51 QUESTION: 095 (1.00)
      ;
During which ONE of the following conditions is EOP-0.0, "Rediagnosis", I
. approved for use based on operator judgement?  l a. After transition to EOP-3, " Steam Generator Tube Rupture."
 
,
b. During the performance of EOP-0.2, " Natural Circulation Cooldown," due to loss of off-site power when a twenty (20) gpm Steam Generator tube leak is detecte i
 
c. During the performance of EOP-1, " Loss of Reactor Or Secondary Coolant," when a RED path is detected in Heat Sin d. After transition to EOP-0.1, " Reactor Trip Recovery", following an inadvertent reactor tri l
 
1 QUESTION: 096 (1.00)
Given the following Unit 1 plant conditions:
- An event occurred at 0800 that was classified as a Site Area Emergenc The plant evacuation alarm was actuated at the time of the classificatio 'Which ONE of the following is the MAXIMUM time by which accountability ;
must be completed?    !
a. 0815 d. 0900 i
 
_ . . _ . . _ m _ .. _ m. _ _ _ _. _ _ _ _ . _ ___ _ _ .m. _ _ _ _ _ _ _ _ _ . _ _ _ _ ._ _ _
.
!
j- SENIOR REACTOR' OPERATOR      Page 52 i-i!
!
;
QUESTION: 097  (1.00)
'Given the following Unit 1 plant conditions:
3  - A reactor trip from full power occurred 60 minutes'ago.
 
i  - RCS' pressure is 620 psig.
 
  - Both-loop cold leg temperatures are 295 degrees CSF Status Tree ST-4 is provided.
 
!
Which ONE of the following states the-appropriate color code for the J Integrity CSF?
i-
;  a.-Green i
b. Yellow c. Orange d. Red QUESTION: 098  (1.00)
The two emergency planning zones are the    at a 10 mile radius and the  at a 50 mile radius from the plan i Which ONE of the following completes the above statement?
a. Low Population Zone, Exclusion Area Boundary b. Exclusion Area Boundary, Low Population Zone c. Ingestion Exposure Pathway, Plume Exposure Pathway l
d. Plume Exposure Pathway, Ingestion Exposure Pathway    ,
        !
l
        ;
 
.- -. - - -. ..._- . . _ ... ._ - . - - _. _ __ . . . . - -
      . _
        -
 
ST-4 INTEGRITY    gey;3o$
        $
02-24-92
, .
CSP- TEMPERATURES IN BOTH NO  mmma  GO TO CSP ,
I
  - COLD LEGS GREATER THAN  E    l I
!
;
285'F YES l    TEMPERATURES NO  GOTO IN BOTH
    { COLD LEGS
      -  CSP P.2 i
GREATERTHAN 315 F
'
YES TEMPERATURES NO  l IN BOTH  l
      -
COLD LEGS
'      GREATERTHAN 345 F  YES TEMPERATURE
'9 DECREASEIN BOTH COLD
+- LEGS LESS NO      CSF SAT THAN 100*F IN THE LAST YES      GOTO 60 MINUTES      CSP PA TEMPERATURES NO IN BOTH F--- COLD LEGS GREATERTHAN 315 F  YES RCS PRESSURE
    -
LESS THAN    GO TO 425 PSIG    CSP- YES I
  ---
        , CSF SAT NO !
TEMPERATURES
  -
IN BOTH COLD LEGS GREATER THAN 360 F YES I    CSF l
SAT l
l l
,
 
    . - . _ . .
SENIOR REACTOR OPERATOR    Page 53
.
l QUESTION: 099 (1.00)
While conducting a plant tour inside the Radiation Control Area you see
,
a velve tagged with a tag similar to the one shown on the attached pag Which one of the following conditions is identified by this tag?
,
l
,
a. The valve is out of its normal lineup positio b. The valve is a boundary valve for a system containing caustic flui c. The valve is to be radiographe !
d. The valve has a contact dose rate of 325 mrem /h i QUESTION: 100 (1.00)
Which ONE of the following is the MAXIMUM annual TEDE dose limit an individual can receive from Planned Special Exposure?
a. 2 rem
, b. 5 rem
;
c. 10 rem d. 25 rem
,
-
  (********** END OF EXAMINATION **********)
 
,____ - _ _ . _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _  __
hy Fh i
l
!
!
I I
I I
!
      . - - . .
,.    /  N I    POINT BEACH NUCLEAR PLANT I
:
l i
IDENTIMCATION #
t
!
!    "
i
        !
l l
l
;
i i
?
l.
 
I a
PBF - 4081A  Revision 0 08/92
,
i
+
!
!
 
j
!
i._ _ ._  _ _ _
 
_ . __ __ _ . _ _ _ . . _ _ _ . __
SENIOR REACTOR OPERATOR    Page 54 I
l l
ANSWER: 001 (1.00)    l 1 l l
l REFERENCE:
,
TRHB 10.1, Re , pg. 15 and C1805 COT and TRHB Figure 10.1.24 KA 022000G007 [3.3/3.5]
i      1
      '
022000G007 ..(KA's)
 
ANSWER: 002 (1.00) REFERENCE:
EOP-0, Rev. 20. Appendix A, pg. KA 000040A110 (4.1/4.1)
000040A110 ..(KA's)
ANSWER: 003 (1.00) REFERENCE:
TRHB 13.7, Re , pg. KA 059000A302 [2.9/3.1)
059000A302 ..(KA's)
 
  . .__ _ _ . _ _ _ __ . . _ . . _ . . _ . . . SENIOR REACTOR OPERATOR    Page 55 ANSWER: 004 (1.00) REFERENCE,: .
Logic Sheet 17 and Setpoint Document Section KA 041020K105 [3.5/3.6]    ,
I
      .
041020K105 ..tKA's)
.
ANSWER: 005 (1.00) l REFERENCE:
AOP-10A, Rev. 18, pg.. 1 [Similar to question used on 94 RO exam.]
KA 000068A101 [4.3/4.5]
000068A101 ..(KA's)
ANSWER: 006 (1.00)
d.
 
l REFERENCE:
TRHB 10.15, Re , pg. 7 and C4313AOT and P&ID Sb9ets 541F091 Sheet 1 I
and 9645971 Sheet 1A KA 068000K107 [2.7/2.9)
068000K107 ..(KA's)
I
!
l l
 
  . .~ _ . . -
:
i
: SENIOR REACTOR OPERATOR  Page 56 ,
'
l
 
,
ANSWER: 007 (1.00)
; l
'
REFERENCE:
J t -
 
AOP-9B, Re , p .
;    l
'
KA 000026G010 [3.6/3.5]    l
 
[Similar to question on 95 SRO exam.]  '
000026G010 ..(KA's)
,
i
 
$ AMSMER: 008 (1.00) l
 
-
,
REFERENCE:    1 TRHB 10.2, Re , pg. 1 l W KA 003000K407 (3.2/3.4]    l 003000K407 ..(KA's)
    ,
ANSWER: 009 (1.00)    ] REFERENCE:
    !
TRHB 10.6, Re , pg. 71 and Master Data Book for B0 KA 004000K203 [3.3/3.5)
004000K203 ..(KA's)
 
SENIOR REACTOR OPERATOR  Page 57 ANSWER: 010 (1.00)
b.
 
REFERENCE:
'.:
-
TRHB 10.16, Rev. 3, pg. 12 and C7304 COT and Logic Sheet KA 013000A302 -[4.1/4.21 013000A302 ..(KA's)
.
ANSWER: 011 (1.00)
b.
 
REFERENCE:
TRHB 13.4, Re , pg. 18 and C730 COT and Logic Sheet KA 013000A402 [4.3/4.4]
013000A402 ..(KA's)
ANSWER: 012 (1.00)
d.
 
REFERENCE:
TRHB 13.1, Re , Figure 13.1.19 and C1315 CO KA 015000A202 [3.1/3.5)
015000A202 ..(KA's)
 
i
 
  .. - - . __ _ _ _ _ . __ _ _ . - -
'
SENIOR REACTOR OPERATOR    Page 58 i
l ANSWER: 013 (1.00)
b.
 
l REFERENCE:
,
TRHB 13.1, Re , pg. 11 and C1310 CO KA 015000K601 (2.9/3.2)
015000K601 ..(KA's)
i l
-
ANSWER: 014 (1.00)
i !
l
'
REFERENCE:      l TRHB 13.8, Re , pg. 6 and C0105 COT; and TRHB 13.8, pg. 19 and Logic Sheet 1 i
! KA 015000K302 (3.3/3.51 015000K302 ..(KA's)
i ANSWER: 015 (1.00) REFERENCE:
Steam Tables
:
 
KA 017020K502 (3.7/4.0]
017020K502 ..(KA's)
r
!
 
. _ _ _ _ - - . . . _ _ _ _ _ _ _ _ . . _ . . . _ _ _ _ _ _ _ _ _ . _ . . __.._.-__.____....____._ . _ _ _ _
i
: SENIOR REACTOR OPERATOR      Page 59
'
s l
ANSWER:  016 (1.00)
'
. REFERENCE:
  '
  :
TRHB 13.2, Re ., pg. 8 and TRHB Figure 10. KA 017020K102 [3.3/3.5]
017020K102  ..(KA's)
. ANSWER:  017 (1.00)      l )
i l
REFERENCE:
BG EOP-0.2, Rev. 13, pg. 7 and C7033 CO KA 022000K404 [2.8/3.1]
022000K404  ..(KA's)
ANSWER:  018 (1.00) REFERENCE:
TRHB 11.2, Re , pg. 11 and C3912 COT and Logic Sheet 24 KA 056000A204 [2.6/2.8)
056000A204  ..(KA's)
l l
 
i
        . -
 
r SENIOR REACTOR OPERATOR    Page 60 ANSWER: 019 (1.00)
l a.
 
t      .
      !
l
. REFERENCE:      I l  ~. :
TRHB 13.7, Re , pgs 28 and 29 and C4210 COT and Logic Sheet 10 l
[EBQ 052-05-149A]
l
,
KA 059000K419 [3.2/3.4]    !
      '
l I
059000K419  ..(KA's)    !
!
I L ANSWER: 020 (1.00)
i b.
 
l REFERENCE:
.TRHB 10.15, Rev. 5, pg. 24 and C4504AOT and P&ID 648J972 Sheet 1 l KA 071000A405 [2.6/2.6]    ~
071000A405  ..(KA's)
i
      !
ANSWER: .021 (1.00)    l REFERENCE:      i
- .TRHB 10.2, Rev. 5, Figure 10.2.1.B and P&ID 541F091 Sheet 1 KA 002000K606 [2.5/2.8)
I j 002000K606  ..(KA's)
!
I i
-
 
l ., . - . . . _ , . _ . . -, , . , ,
 
SENIOR REACTOR OPERATOR  Page 61 ANSWER: 022 (1.00)
b.
 
REFERENCE:
, .
TRHB .10. 8 ', Rev. 4, pg. 15 and Control Panel CO2 KA 006030A401 [4.4/4.4]
006030A401 ..(KA's)
'
      !
ANSWER: 023 (1.00)    !
c.
 
REFERENCE:
TRHB 10.8, Re , pg. 18 and Technical Specification 15. KA 006000G005 [3.5/4.2]
006000G005 ..(KA's)  i I
i ANSWER: 024 (1.00)    )
I a.
 
REFERENCE:
      .!
TRHB 10.3, Re , Figure 10.3.7, Logic Sheet 18 and ICP 1 l KA 010000A403. [4. 0/3. 8]
010000A403 ..(KA's)
      !
 
_
      ;
SENIOR REACTOR OPERATOR    Page 62 AN.9WER: 025 (1.00)
' .
      ;
REFERENCE:
, .
TRHB 13.5, Re , pg. 21, C0905C01, Logic Sheet 18 and P&ID 541F091 Sheet 1 KA 010000K301 [3.8/3.9)
010000K301 ..(KA's)
ANSWER: 026 (1.00) i REFERENCE:    I TRHB 13.6, Re , pg. 7 and C1006 COT and Logic Sheet 18 KA 011000K604 [3.1/3.1]
011000K604 ..(KA's)
 
ANSWER: 027 (1.00) I
' REFERENCE:    ,
TRHB 13.6, Re , p .
KA.011000A101 [3.5/3.6]
011000A101 ..(KA's)
i
  , - - - -
 
SENIOR REACTOR OPERATOR  Page 63 ANS : 028 (1.00)
"'
kUd317Y REFERENCE: ,
  [ggggg l TRHB 13.3, Rev. 4, pg. 3 KA 012000K406 (3.2/3.5)
IC)Nf 012000K406 ..(KA's)
,
-ANSWER: 029 (1.00) l
    ,
REFERENCE:
l BG CSP-I.3, Rev. 7, pg. l i
KA 016000G015 (3.6/3.8)  l
    ;
i 016000G015 ..(KA's)
    ,
ANSWER: 030 (1.00) i'
REFERENCE:
TRHB 13.8, pg. 2 KA 035010k501 [3.4/3.9)
035010K501 ..(KA's)
 
. _ _ . .. _ ... .. . . _ _ _... _ _ . _ .4 _ - _ _ , _ _ _ . _ . _ _ _ _ _ . . . _ _ . . _ _ _ _ . . . . _ _ . _  _
 
l'
 
3
; SEMIOR REACTOR: OPERATOR        Page 64 j'
 
.
t
; ANSWER:  031 (1.00)
l'
; a.
 
i REFERENCE:
  , .
TRHB 12.5, Re , Figure 12.5.1 and Master Data Book KA 062000K201 [3.3/3.4]
062000K201  . .(KA's)
ANSWER:  032 (1.00)        l REFERENCE:
TRHB 12.2, Re , pg. 38, C4105AOT, OI-110, Re , and PC-29, Rev. 29 Monthly Best KA 064000K402 (3 9/4.2] .
064000K402  . .(KA's)
          !
I i
ANSWER:  033 (1.00)        1 REFERENCE:
l TRHB 12.8, Re , pg. 74 and OP-11A, EDG G01/G02 l
KA 064000A206 [2.9/3.3]
l 064000A206  . .(KA's)
          ,
          -
 
. . _ . .._..m..~ . . _._.__.___-__...._ . _ _ _ , _ _ . _ . . _ . . _ . _ . _ . . _ _ . _ _ _ _ _ . . _ _ _
          !
Page 65
          '
SENIOR REACTOR OPERATOR
;
I i
;
{. ANSWER:  034 (1.00)
 
;
d.
 
!
:
F
! REFERENCE: i *
1.
 
I 'TRHB 13.12, Re , pg. 14.
 
!
KA 073000A402 [3.7/3.71      6 1          ;
j 073000A402  ..(KA's)      '
;
:
l ' ANSWER:  035 (1.00)
i 3 a.
 
j i,
l
.i REFERENCE:
TRHB 10.7,.Re , . pg . 17, C0506AOT.and Setpoint' Document, Sections 2.3, i 8.2 and 11.2.
 
f KA 005000K109 [3.6/3.9]
!
i
 
$
005000K109  ..(KA's)
a
:
 
s ANSWER: 036 (1.00)
;
d.
 
<
REFERENCE:
!
'
TRHB 13.10, Re , pg. 2; 62804 COT; Setpoint Document 14.13; P&ID PBM-241; OP-1c, Step 7.3.5
 
KA 041020K603 (2.7/2.9)
:
;
041020K603  ..(KA's)
;
I
.
i
;
 
$
 
.~ . . . ~ . - . - . . . . - . . . . . . . . - - . - . - . - - . . . - . - - - - - . . - . .
        . . - . . . . . .
        !
l
:
! SENIOR REACTOR OPERATOR      Page 66'
'
        .
i
:        I j'ANSUER:  037 (1.00)
        -
.
l a.
 
i
,' REFERENCE:
'AOP-6B, Rev. 6, pg. 1 and LP2441, EO 2.8'.
i I, KA'000005A201 [3.3/4.1]      l
        '
!        ,
i i
1        '
#
000005A201  ..(KA's)
:
!
  '
I Ai?SWER: 038 .(1.00)      i l
b.:
REFERENCE:
ECA'O.1, Rev. 10, pg. 12 KA 000015G007 [3.1/3.2]
000015G007  ..(KA's)
ANSWER: 039 (1.00)      I REFERENCE:
EOP-0.1, Rev. 14, pg. !
KA 000024A205 [3.3/3.9)
000024A205  ..(KA's)
        :
        ,
 
SENIOR REACTOR OPERATOR  Page 67 ANSWER: 040 (1.00) ,
REFERENCE:    ,
s .
EOP-2, Rev. 10, pg. KA 000040A110 [4.1/4.1)
000040A110 ..(KA's)
    .
ANSWER: 041 (1.00)
b.
 
REFERENCE:    .
EOP-1, Rev. 19, Foldout pag KA 000040A205 [4.1/4.5]
000040A205 ..(KA's)
i ANSWER: 042 (1.00)
    , l REFERENCE:
TRHB 13.9, Re , pg. 4 and Setpoint Document, Section 1 KA 000051K301 [2.8/3.1]
000051K301 ..(KA'c)    !
l I
 
--. -...- . - .. . . - . _ - . ~ ~ . . . - ~ . - . ~ . - . - - _ _ - . - . . . . - ~ . . . . . -... _. _ - -
l:        :
SENIOR REACTOR OPERATOR      Page 67
        :
        ;
ANSWER: 040 -(1.00)      i REFERENCE:
s .      ;
EOP-2,-Rev. 10, pg. KA 000040A110 [4.1/4.1)
i
.000040A110 ..(KA's)    !
      .
ANSWER: b41 (1.00)      '
l i
l ~b.
 
l l
l-l REFERENCE:
EOP-1, Rev. 19, Foldout pag KA 000040A205 [4.1/4.5]
000040A205 ..(KA's)
i ANSWER: 042 (1.00)
. REFERENCE:
TRHB 13.9, Rev. 2, pg. 4 and Setpoint Document, Section 1 KA 000051K301 [2.8/3.1)
000051K301 ..(KA's)
!-
!
.
!e
! I
    ;
l-SENIOR REACTOR OPERATOR  Page 68 l
l ANSWER: 043 -(1.00)
  $f)w n C//4DW ;
l  dSr CD && P l ,
RLF ENCE:  ggg gggge7
.
'AOP-5A, Rev. 4, pg. KA 000051A202 [3.9/4.1)  i 000051A202 ..(KA's)
l
 
ANSWER: 044 (1.00)
l !
REFERENCE:    i
    !
LP 0462, Re , pg. 7 and LO K302 [4.3/4.6]
000055K302 ..(KA's)
ANSWER: 045 (1.00) REFERENCE:  e BG ECA-0.0, Rev. 14, pg. 14 and LP 0462, LO KA 000055A106 [4.1/4. 5 ]
000055A106 ..(KA's)
 
.-_..._.--.__.__...__...._._..__._.._..._..___...m._._.-._. _ . _ ___. ....._... _... _ _ __ ...._ _.. _ ..._... _ __
'
        .
j.
 
~
        :
l t SENIOR REACTOR OPERATOR      Page 69 1
:
-
l
        )
i,        ,
l-ANSWER:  046 ( l'. 0 0 )
i
 
d.
:
?
i
; REFERENCE:
j  $..- *
  .
i TRHB 12.9,'Re , pg. 11.
 
i
*
000057A218 [3.1/3.1]
:
!,
j  000057A218  ..(KA's)
L
.
!
j ANSWER:  b47 ( 1. 0 0 )-
.
i d.
 
i.. REFERENCE:
I
        *
l 'TRHB 11.14, Re , pg. 2 t        '
 
i a [Similar to EBQ 052-01-065A]
!
'
KA 000067K102 [3.1/3.9]
000067K102  ..(KA's)    i i
ANSWER:  048 (1.00)
        ! !
REFERENCE:
Tech. Spec. 15.3.6, pg. 15.3.6- KA 000069G003 [3.3/3.9)
000069G003  ..(KA's)
 
SENIOR REACTOR OPERATOR  Page 70
    !
ANSWER: 049 (1.00) REFERENC , .
CSP-C.1, Rev. 12, pg. 2 i KA 000074K201 (3.6/3.9)  ,
i 000074K201 ..(KA's)
ANSWER: 050 (1.00) REFERENCE:
EOP-0, Rev. 20, pg. 2 and LP 0405 LO 2.1 KA 000007A202 [4. 3/4. 6)
000007A202 ..(KA's)
; ANSWER: 051 (1.00) REFERENCE:
BG EOP-0, Rev. 21, pg. 9 and LP 0405, LO 2.1 KA 000007K301 [4.0/4.6]
000007K301 ..(KA's)
l
 
SENIOR REACTOR OPERATOR  Page 71 ANSWER: 052 (1.00)
b.
 
REFERENCE:
, .
EOP-0, Rev. 20, pg. 18 and LP 0405, LO KA 000008A101 (4.2/4.0]
~000008A101 ..(KA's)
.
ANSWER: 053 (1.00)
e.
 
REFERENCE:
EOP-1.2, Rev. 13, Foldout Page and LP 0435, LO KA 000009A234 [3.6/4.2]
000009A234 ..(KA's)
ANSWER: 054 (1.00)
.d.
 
REFERENCE:
OP-5A, Rev. 28, pg. 1 and THRB 10.6, Rev. 3, p .
KA 000022K307 [3.0/3.2]
000022K307 ..(KA's)
    ;
 
  ~1 SENIOR REACTOR OPERATOR Page 72 l
ANSWER: 055 (1.00)
a.
 
REFERENCE:
3 .
SEP-1.1, Rev. O, pg. KA 000025K101 (3.9/4.3]
000025K101 ..(KA's)
l ANSWER: 056 (1.00)
a.
 
REFERENCE:
EOP-3, Rev. 19, pg. 3 KA 000032A101 [3.1/3.4)
000032A101 ..(KA's)
l ANSWER: 057 (1.00)
a.
 
REFERENCE:
AOP-1A, Re , p .
KA 000037G011 [3.9/4.1)
000037G011 ..(KA's)
 
-SENIOR REACTOR OPERATOR  Page 73
: ANSWER: - '058 (1.00) REFERENCE:
, .
EOP-3, Rev. 19, pg. 15 and BG EOP-3, Step 20 (PROVIDE: EOP-3, pg. 15 of 36]
KA 000038A136 -[4.3/4.5]
000038A136 ..(KA's)
  ,
ANSWER: 059 (1.00)
c.
 
' REFERENCE:    R BG CSP-H.5, Re , pg. 4 and LP 1998, EO .KA 000054K102 [3.6/4.2]
000054K102 ..(KA's)
    ;
ANSWER: 060 (1.00) !
REFERENCE:
BG CSP-H.1, Rev. 10, pg. KA 000054K304 [4.4/4.6]
000054K304 ..(KA's)
 
    . . . .
SENIOR REACTOR OPERATOR    Page 74 ANSWER: 061 (1.00)
a.
 
REFERENCE:    i y .    !
AOP-0.0, Re , p .
KA 000058A203 [3.5/3.9)
000058A203 ..(KA's)
 
ANSWER: 062 (1.00)    l l I
      \
l REFERENCE:
TRHB 13.6, Re , pg. 6 and Figure 13.6.1 and Logic Sheet 1 ]
KA 000028K202 (2.6/2.7)
000028K202 ..(KA's)
ANSWER: 063 (1.00) ,
REFERENCE:
EOP-0.1, Rev. 14, pg. 1 KA 000056K101 [3.7/4.2]
000056K101 ..(KA's)
 
. - -. . . .. . .- . .-...- . .-- - . . . ~ . . . . _ - - - - -
SENIOR REACTOR-OPERATOR    Page 75
,        i
        '
h 1'
l' ANSWER: 064 (1.00)- REFERENCE:
  '
  : .
i NP 1.1.4, Re , pg. \ '
KA.194001A101 [3.3/3.4]      l l
 
194001A101 ..(KA's)
ANSWER: 065 (1.00) .J REFERENCE:      l l
EOP-0, Rev. 20, pg. 1 and SEP-2, Re , p . l KA 194001A102 [4.1/3.9]
194001A102 ..(KA's)
f l'
ANSWER: 066 (1.00) REFERENCE:
10 CFR 55.53 and OM 3.10, Re , pg. 1 KA 194001A103 [2.5/3.4]
194001A103 ..(KA's)
;
!
,
i
:
l l
;
l ..  .
  -. .
    - _ . .
 
SENIOR REACTOR OPERATOR  Page 76 i
    .
ANSWER:' 067 (1.00) REFERENCE:
  ,  l NP-l'. 7. 5, Rev. 3, pg. KA 194001A103 (2.5/3.4]
194001A103 ..(KA's)
ANSWER: 068- (1.00) REFERENCE:
NP 1.7.10, Rev. 1, pg. KA 194001A105 [3.6/3.8]
194001A105 ..(KA's)
ANSWER: 069 (1.00)
. REFERENCE:
Manager's Expectations, 93- KA 194001A103 [2.5/3.4]
l 194001A103 ..(KA's)  l l
L i
l t    l i    I i
- - - - - __ _ - .
 
SENIOR-REACTOR' OPERATOR Page 77 I
ANSWER: 070 (1.00) L i
REFERENCE:  '
y, .
OM 3.10, Re , pg. KA 194001A110 [2.9/3.9]
194001A110 ..(KA's)
ANSWER: 071 (1.00) .
REFERENCE:  ,
  -
i Tech. Spec. Definition 15. KA 194001A113 [4.3/4.1]  ,
194001A113 ..(KA's) 1 l
ANSWER: 072 (1.00)
b.
 
REFERENCE:
NP'1.6.6, Re , p .
KA 194001A103 [2.5/3.4]
  !
194001A103 .(KA's)
  ,
  <
l l
 
    . - . - _ - - - -
~
SENIOR REACTOR OPERATOR    Page 78 ANSWER': - 073 (1.00)
b.
 
REFERENCE:
  *
'..
-OM 3.17, Re O, pgs. 4 & KA'194001K101 [3.6/3.7)
194001K101 ..(KA's)
ANSWER: 074 (1.00)
c.
 
REFERENCE:
NP 1.9.15, Rev. 2, pg. 1 KA 194001K102 [3.7/4.1)
194001K102 ..(KA's)
ANSWER: 075 (1.00)
d.
 
REFERENCE:      I TRHB 10.6, Rev. 3, pg. 35 and C0413 COT and Setpoint Document, Section KA 004010A105 [3.0/3.2)
004010A105 ..(KA's)
 
SENIOR REACTOR OPERATOR  Page 79 ANSWER: 076 (1.00)
c.
 
REFERENCE:
1 -
TRHB 11.4, Re , pg. 15 and Tech. Spec. 15. KA 061000G005 [3.3/4.0]
061000G005 ..(KA's)
ANSWER: 077 (1.00)
c.
 
REFERENCE:
Tcch. Spec. Bases 15.3.6.A. KA 029000K104 [3.0/3.1]
029000K104 ..(KA's)
ANSWER: 078 (1.00)
b.
 
REFERENCE:
TRHB 10.9, Re , Figure 10.9.1 and C0801AO KA 008010A302 [3.0/3.1]
008010A302 ..(KA's)
 
l b
SENIOR REACTOR OPERATOR  Page 80 i
l'
ANSWER: 079 (1.00)
l l l
REFERENCE:  l t *
l RP-2D, Rev. 10, .pg. l l
KA 034000K402 [2.5/3.3)
034000K402 ..(KA's)
ANSUER: 080' (1.00)
d.
 
REFEREFCE:
Tech. Spec. 15. KA 002000G005 [3.6/4.1)
002000G005 ..(KA's)
ANSWER: 081 (1.00)
b.
 
REFERENCE:
. Tech. Spec. 15.3.3 Basis, pg. 15.3.3- KA 026000A102 [3.6/3.9)
026000A102 ..(KA's)
 
.. . - _ _ . . . . . - _ _ _ . _ _ . . . . _ ~ . _ _ _ _ _ _  -.-_.__..~._.m._ ...___ - . _ _ - - - . -_
SENIOR REACTOR OPERATOR      Page 81
!-
i L
iIANSWER:  082 (1 00)
i
,
C.
 
;        l 1 REFERENCE-4  , .
,
TRHB 13.8, Re , pg. 10 and LP 2441, EO 2.4 and Logic. Sheet 16
! KA'000001K105 [3.5/3.8]
 
4
 
000001K105  ..(KA's)
~
i j ANSWER:  083 (1.00)
 
4 a.
 
i
} REFERENCE:
.AOP-6A, Rev. 7, pg. 13, and TRHB 13.13, Rev. 1, pg. KA 000003A101 [2.9/2.9)
'000003A101  ..(KA's)
ANSWE (1.00) REFERENCE:
BG EOP-1, Rev. 18, pg. KA 000011K314 [4.1/4. 2 ]
000011K314  ..(KA's)    '
 
  . -
_ _  .. _ _ _ _ _ _
SEMIOR REACTOR OPERATOR    Page 82
, ANSWER: 085 (1.00) i REFERENCE:
, .
LG CSP-S.1, Rev. 13, pg. 6 and LP 1996, LO 2.6.
 
,
>
KA 000029A115 [4.1/3.9]
l 000029A115 ..(KA's)    I
 
I
} ANSWER: 086 (1.00)
b.
 
h
, REFERENCE:
, NP 1.9.14, Rev. O, pg. l KA 000067G012 [3.4/3.4]
000067G012
'
  ..(KA's)
ANSUER: 087 (1.00)
d.
 
j REFERENCE:
 
LP 1997, Rev. 2, pgs. 22-24, LO 2.6, CSP-C.1 Background (page 1), BG EOP pg. 13 KA 000074K103 [4.5/4.9)    ;
      ;
      !
000074K103 ..(KA's)
;
l l
 
l SENIOR REACTOR OPERATOR  Page 83 ANSWER: 088 (1.00)
C.
 
REFERENCE:
'.. *
Tcch. Spec. Bases 15.3. KA 000076K305 (2.9/3.6]
[Similar question used on 94 RO exam.]
000076K305 ..(KA's)
ANSWER: 089 (1.00)
b.
 
REFERENCE:
AOP-6E, Re , p .
KA 000024A202 [3.9/4.4]
000024A202 ..(KA's)
ANSUER: 090 (1.00)
d.
 
REFERENCE:
LP 0405, Rev. 11, pg. 16 and LO 2.13 and EOP-0, Rev. 20, pg. 2 KA 000009A211 (3.8/4.1]
000009A211 ..(KA's)
 
  .- .- - . .- -
SENIOR REACTOR OPERATOR    Page 84
-
ANSWER: 091 (1.00) REFERENCE:
.
s .
OP-9D, Rev. 12, pg. KA 000060G007 [3.1/3.4]
000060G007 ..(KA's)
ANSWER: 092 (1.00) REFERENCE:
TRHB 13.2, Re , pg. 50, RMS Alarm Setpoint and Response Book KA 000061A101 (3.6/3.6)
;
 
000061A101 ..(KA's)
.
ANSWER: 093 (1.00)
J a.
 
REFERENCE:
LP 0441, Rev. 9, pg. 24 and LO 2.17, EOP-3.1 Background, Rev. 11, p .
KA 000038K306 (4.2/4.5]
000038K306 ..(KA's)
 
- . _ _ _ . . _ _ _ __
SENIOR REACTOR OPERATOR  Page 85 ANSWER: 094 (1.00)
C.
 
REFERENCE:    I
, .
3, pg. '
PBF-5101, Re KA 000036G007 (3.2/3.5]  i 000036G007 ..(KA's)  l ANSWER: 095 (1.00)
    ! !
REFERENCE:
EOP-0.0, Re , pg. i KA 194001A102 [4.1/3.9]
194001A102 ..(KA's)
ANSWER: 096 (1.00)
b.
 
REFERENCE:
EP 6.0, Rev. 35, pg. KA 194001A116 [3.1/4.4]
194001A116 ..(KA's)
 
. SENIOR REACTOR OPERATOR Page 86 ANSWER: 097 (1.00) REFERENCE:
' *
.
ST-4, Rev. KA 194001A108 [2.6/3.1]
194001A108 ..(KA's)
ANSWER: 098 (1.00)
d.
 
REFERENCE:
EP 2.0, Rev. 30, pg. KA 194001A116 [3.1/4.4]
194001A116 ..(KA's)
ANSWER: 099 (1.00)
d.
 
REFERENCE:
HP 3.2.9, Re , p . i
  '
KA 194001K104 [3.3/3.5]
194001K104 ..(KA's)
 
  . - - ~ . - _. - - . . . . . . . _ _
SENIOR REACTOR OPERATOR    Page 87 d
.
ANSWER: 100 (1.00)    l '
REFERENCE:
,
NP 4.2.1.18, Rev. O, p .
KA 194001K104 [3.3/3.5]
'
l
 
l 194001K104 ..(KA's)
i
;
,
I l
l l
i
    .:
  (********** END OF EXAMINATION **********)
}}
}}

Revision as of 23:04, 21 July 2020

NRC Operator Licensing Exam Repts 50-266/96-303OL & 50-301/96-303OL Administered During Wk of 961007 & Insp Conducted on 961120-27.Violation Noted.Exam Results:Three Out of Five ROs & Two Out of Two SROs Passed Exam
ML20147A780
Person / Time
Site: Point Beach  NextEra Energy icon.png
Issue date: 01/23/1997
From:
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20147A759 List:
References
50-266-96-303OL, 50-301-96-303OL, NUDOCS 9701280303
Download: ML20147A780 (194)
v • d • e


Text

{{#Wiki_filter:.- - .-. . . _ . -. ._ .

U.S. NUCLEAR REGULATORY COMMISSION REGION lll Docket Nos: 50-266; 50-301 Licenses No: DPR-24; DPR-27 Reports No: 50-266/96303(OL); 50-l'31/96303(OL) Licensee: Wisconsin Electric Power Company Facility: Point Beach Nuclear Plant Units 1 and 2 Location: 6610 Nuclear Road Two Rivers, WI 54241 Dates: October 7-11,1996; November 20-27,1996 Inspectors: J. Lennartz, Chief Examiner, Rlli P. Cataldo, Examiner, Rll! T. Guilfoil, Examiner, Sonalysts In K. Parkinson, Examiner, Sonalysts In Approved by: M. Leach, Chief, Operator Licensing Branch Division of Reactor Safety i 

      !

9701280303 970123 PDR ADOCK 05000266 V PDR 

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

U.S. NUCLEAR REGULATORY COMMISSION 

           /

i REGION lll 

      '

Docket Nos: 50-266; 50-301 \ DPR-24; DPR-27 ^' Licenses No: 

          *
          ,

l ' Reports No: 50-266/96303(OL); 50-301/96303(OL) i 

           ,
         .. -

l Licensee: Wisconsin Electric Power Company I ! Facility: Point Beach Nuclear Plant Units 1 and 2 Location: 6610 Nuclear Road ! Two Rivers, WI 54241 i l Dates: October 7-11,1996; November 20-27,1996 Inspectors: J. Lennartz, Chief Examiner, Rlll l P. Cataldo, Examiner, Rlll , T. Guilfoil, Examiner, Sonalysts Inc.

l K. Parkinson, Examiner, Sonalysts Inc.

Approved by: M. Leach, Chief, Operator Licensing Branch Division of Reactor Safety 

      .

l l 9701280303 970123 PDR V ADOCK 05000266 PDR _ - - - - . - , , ___ . - . ___ 

.

. EXECUTIVE SUMMARY Point Beach Nuclear Plant, Units 1 and 2 Examination Report 50-266-96303; 50-301-96303 Operator initiallicensing examinations were administered to five Reactor Operator (RO) applicants and two Senior Reactor Operator (SRO) applicants during the week of October 7,1996. Additionally, an inspection was conducted from November 20-27, 1996, to assess procedure deficiencies and apparent operator training weaknesses identified during initial examination administratio Examination Results: 

* Three RO license applicants and both SRO license applicants passed the examinations and will be issued licenses. Two RO license applicants failed the operating examination due to unsatisfactory performance on Administrative wb Performance Measure (JPM) tasks and were denied license Ooerations_;
* An identified error in Abnormal Operating Procedure 0.0, " Vital DC System j Malfunction," would have aligned 125 VDC Electrical Distribution System Battery 4 Chargers D-107 and D-108 outside design capabilities which could have resulted in damage to required DC loads. This was considered a violation of 10 CFR 50, Appendix B, Criterion V. (Section 03.1)
* Two applicants failed the examination due to unsatisfactory performance while two i other applicants demonstrated weaknesses on the Administrative JPM tasks I administere Additionally, unsatisfactory performance was demonstrated by at least one applicant on five of the six different Administrative JPM tasks administered which included: completing portions of the criticality checklist; proper method to independently verify valve position; and determination of Main Turbine loading limitation The demonstrated weaknesses illustrated an apparent lack of attention to detail, lack of procedural knowledge, and a failure to follow procedures for the Administrative JPM tasks administered. The applicants' inability to perform the Administrative JPM tasks administered on this examination was considered a weakness. (Section 05.3)
* Operator training regarding the operating limitations and differences between the 125 VDC system battery chargers was considered a weakness. (Section 05.5)

.

. EXECUTIVE SUMMARY Point Beach Nuclear Plant, Units 1 and 2 Examination Report 50-266-96303; 50-301-96303 Operator initial licensing examinations were administered to five Reactor Operator (RO) applicants and two Senior Reactor Operator (SRO) applicants during the week of October 7,1996. Additionally, an inspection was conducted from November 20-27, 1996, to assess procedure deficiencies and apparent operator training weaknesses identified during initial examination administratio Examination Results: e Three RO license applicants and both SRO license applicants passed the examinations and will be issued licenses. Two RO license applicants failed the operating examination due to unsatisfactory performance on Administrative Job Performance Measure (JPM) tasks and were denied license Ooerations: e An identified error in Abnormal Operating Procedure 0.0, " Vital DC System Malfunction," would have aligned 125 VDC Electrical Distribution System Battery Chargers D-107 and D-108 outside design capabilities which could have resulted in , damage to required DC loads. This was considered a violation of 10 CFR 50, l Appendix B, Criterion V. (Section 03.1) l e Two applicants failed the examination due to unsatisfactory performance while two other applicants demonstrated weaknesses on the Administrative JPM tasks I administere Additionally, unsatisfactory performance was demonstrated by at least one applicant on five of the six different Administrative JPM tasks administered which included: completing portions of the criticality checklist; proper method to independently verify valve position; and determination of Main Turbine loading limitation The demonstrated weaknesses illustrated an apparent lack of attention to detail, lack of procedural knowledge, and a failure to follow procedures for the Administrative JPM tasks administered. The applicants' inability to perform the Administrative JPM tasks administered on this examination was considered a weakness. (Section 05.3) e Operator training regarding the operating limitations and differences between the 125 VDC system battery chargers was considered a weakness. (Section 05.5) 

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

Reoort Details 1. Operations j i 03 Operations Procedures and Documentation l 03.1 Ooerations Procedure Deficiencies I 

       !

] Insoection Scooe An inspection was conducted from November 20-27,1996, to assess procedure l deficiencies identified during initial examination administratio I l i Observations and Findinas  ! - I The following procedure deficiencies were identified: i The examiners identified that Abnormal Operating Procedure (AOP) 0.0, 

 " Vital DC System Malfunction," Revision 9, Attachment B, "DC Distribution
Panel Power Supplies," would have aligned battery chargers D-107 and D-j 108 to DC busses D03 and D04 respectively without the related battery also connected.

I However, battery chargers D-107 and D-108 were designed such that they I . could not function properly if connected to a DC bus without a battery also j

connected. Additionally, if battery chargers D-107 and D-108 were aligned f to a bus without the related battery, damage could occur to the DC loads connected to the bus. (This issue is discussed further in Section 05.5 of l

this report).

i l The identified error in AOP-0.0 would have aligned 125 VDC system battery

chargers D-107 and D-108 outside design capabilities which could have , resulted in damage to related DC loads. This was considered a violation of j 10 CFR 50, Appendix B, Criterion V.

i The licensee took prompt corrective actions and issued a temporary procedure charge to AOP 0.0 the same day that the NRC identified the issue. A caution which indicated that battery chargers D-107, D-108 and D- ' 4 109 could not be aligned to a DC bus independent of a battery was added i before the steps which restored power to bus DO3 and bus D04 in AOP 0.0, i Attachment B.

The examiners identified that Annunciator Response Books (ARB) 1CO41 A 3-5, Revision 3, dated August 4,1995, Section 7.0, and 2C04 2A 3-7, Revision 3, dated March 6,1995, Section 7.0, incorrectly identified the power range overpower rod stop logic as two out of four (2/4). The actual logic was one out of four (1/4) power range instruments reading greater than setpoin .
-
      \

The incorrectly stated logic confused one RO license applicant while performing a JPM to remove a failed power range nuclear instrument from j service. The applicant reviewed the annunciators that were lit for the task conditions prior to performing the task and then indicated to the examiner that annunciator 2CO4 2A 3-7, " Power Range Overpower Rod Stop" should not be lit since only one power range instrument was failed and the rod stop was a two out four logic. The applicant then referenced ARB 2C04 2A 3-7 to confirm that the rod stop was a two out of four logic. The examiner acknowledged the applicant's concern regarding an incorrect annunciator being lit and then prompted the applicant to continue with task performanc After the examiner identified the procedure error to the licensee, prompt I corrective action was taken to issue a procedure change request (ARB 96-003) to revise the AR l Conclusions l The examiners concluded the following regarding the identifie i procedure deficiencies , e The identified error in AOP-0.0 was considered a violation of 10 CFR 50, Appendix B, Criterion e The identified error in the ARB could result in some operator confusion, as I observed during examination administration, but would not result in any I inappropriate actions being taken and therefore, the safety consequences were mino l l e The licensee had taken prompt corrective actions for the procedure deficiencies which were considered adequat Operator Training and Qualification I 05.1 General Comments 

,

Operator initial licensing examinations were administered at the Point Beach Nuclear Plant during the week of October 7,1996, in accordance with NUREG-1021, Revision 7, " Operator Licensing Examiner Standards," to five Reactor Operator (RO) applicants and two Senior Reactor Operator (SRO) applicant ! 05.2 Examination Prenaration and Validation The NRC prepared the examination in accordance with NUREG 1021 guideline The examination material was validated at the Point Beach Nuclear Plant during the week of September 23,1996, with licensee training and operations personnel - assigned to the examination team. Licensee support during examination development and validation was considered goo . 05.3 Examination Administration Dynamic Simulator Examination The examiners observed the following regarding the license applicants' performance during the dynamic simulator examination: e The RO applicants routinely identified Technical Specification (T/S) entry conditions and were able to quickly reference and correctly apply applicable T/Ss when require e No common performanco deficiencies were observe The examiners observed the following regarding dynamic simulator examination administration: e A simulator operatcr inserted the incorrect IC (initial conditions) set at the start of one dynamic examination. An experimental IC set was inserted that established the same general plant conditions, including reactor power level, as the IC set validated for the examination. However, the experimental IC set had a positive moderator temperature coefficient vice a negative l coefficient which the scenario was validated wit l This examination had been in progress for approximately 15 minutes when the simulator operator identified the error to the Chief Examiner. The examination was temporarily delayed at that point so the IC set that had l been previously validated for use on the examination was established. This l delayed examination administration approximately 30 minutes. The I applicants were required to exit the simulator and wait for the correct setup to be established which added unnecessary stress to the applicants, JPM Walkthrouah Examination Unsatisfactor/ performance was demonstrated by at least one applicant on the following five Administrative JPM tasks: e " Complete Selected Portions of the Criticality Checklist CL-1 A" e " Determine Main Turbine Loading Limitations for a Turbine Hot Plant Startup" i e " Conduct a Valve Lineup Verification Check of Various Valves" ' e " Conduct Personnel Monitoring For Radioactive Contamination" e " Perform Activation of the Communications Area in the Operations Support Center" Conclusions The examiners concluded that the applicants were not adequately prepared for the Administrative JPM tasks. The demonstrated weaknesses on the Administrative JPM tasks illustrated an apparent lack of attention to detail, lack of procedural knowledge, and a failure to follow procedures. The applicants difficulty in 

.

l 

.

performing the Administrative JPM tasks administered for this examination was ; considered a weakness. Additionally, the examiners concluded that the license ! applicants were well prepared for the dynamic simulator examination and that the l ROs' knowledge of and ability to use technical specifications was goo i i 05.4 Post Examination Activities Written Examination The licensee submitted two written examination post review comments for l consideration and the NRC's post examination review identified 13 written I examination questions where a majority of the applicants failed to provide the correct response. The licensee's submitted comments with the NRC's resolution as

well as the NRC's post review results were documented in Enclosure 2, " Post i Examination Comments and Review."

The examiners identified an additional generic knowledge weakness regarding the ) Rod Control system response to a failed high Power Range Nuclear trotrument ' 

(PRNI) (RO question 015; SRO question 014). Knowledge regarding the " Power j Range Overpower Rod Stop" logic of one out of four (1/4) PRNis reading 105% '

was needed to answer this question correctly. Five out of the seven applicants answered this question incorrectly. This knowledge weakness appeared related to the identified procedure error (Section 03.1.b.2) where the ARB incorrectly identified the power range overpower rod stop logic as two out of four (2/4).

The examiners reviewed the training material regarding the power range overpower : rod stop logic in Training Handbook (TRHB),13.1, " Nuclear Instrumentation System," Revision 3, January 3,1992, and determined that the correct logic (1/4) was referenced, Conclusions The examiners concluded, based on training material review, that the correct logic for the power range overpower rod stop was being taught. However, the ARB procedure error may have contributed to the knowledge weaknes .5 125 VDC Electrical Distribution System Insoection Scooe An inspection was conducted from November 20-27,1996, to assess an apparent operator training weakness identified during NRC post examination review of a JPM question regarding the 125 VDC station battery operability requirement Observations The safety-related 125 VDC electrical distribution system included four main DC distribution busses (D01, D02, D03, and D04), five station batteries (D-05, D-06, D-105, D-106 and D-305) and six battery chargers (D-07, D-08, D-09, D-107, D-108, and D-109).

_ __ 

.

l i 

     ,

A JPM question required the applicants to determine the applicable T/S actions for the following set of conditions regarding the 125 VDC electrical distribution j system: ' e Both units at 100% power l e Safety-related station battery D-105 had been out of service (OOS) for 4 ' hours and safety-related station battery D-305 (swing battery) was also OO e l Remainder of DC electrical distribution system was in a normal at power line u . e Battery charger D-108 subsequently fail l Battery D-105 would normally be connected to DC distribution bus D03. Swing battery D-305, if operable, could also have been aligned to DC distribution Bus D03. However, the above conditions resulted in only battery charger D-107 aligned to DC distribution Bus D03 and only battery D-106 on bus D04 while all other DC l distribution busses had a related station battery and battery charger aligned to the l bu I Five applicants stated that battery charger D-109 had to be connected to DC bus D-04 within 2 hours per T/S 15.3.7.B.1.1 which stated "one of the four connected battery chargers may be inoperable for a period not to exceed 2 hours."

Additionally, the five applicants indicated that after battery charger D-109 was , connected to bus D-04 the plant could remain at power for 20 hours without having to take further actions per T/S 15.3.7.B.1.i which stated "one of the four connected safety-related station batteries may be inoperable for a period not exceeding 24 hours provided four battery chargers remain operable with one charger carrying the DC loads of each main DC distribution bus."

Those five applicants provided the expected response as determined by the licensee members assigned to the examination validation team. Additionally, the T/Ss implied that this condition was allowe However, two applicants stated that battery charger D-107 could not function if connected to bus D03 without a battery also connected, due to the charger's o.ssign. Therefore bus D03 did not have any power which would require a plant shutdown per T/S 15.3.0.B which stated "in the event a limiting condition for operation (LCO) cannot be satisfied because of equipment failures or limitations beyond those specified in the permissible conditions of the LCO, action shall be initiated within one hour to place the affected unit in hot shutdown within seven hours of entering this LCO."

The examiners determined that battery chargers D-107, D-108, and D-109 could not function properly if connected to their related DC distribution bus without a battery also connected due to the chargers' design. Battery chargers D-107, D-108, and D-109 could supply respective DC loads while maintaining the batteries at full charge as well as recharge a partially discharged battery while carrying normal loads which was in accordance with requirements described in the Final Safety Analysis Report (FSAR). However, if D-107, D-108, or D-109 were aligned to a

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

.

d 1 

.

bus without a battery, damage could occur to DC loads connected to the bus. This limitation did not exist for battery chargers D-07, D-08, and D-0 The licensee wrote a condition report (CR 96 1246) regarding this issue and i summarized the battery charger differences and limitations in the Operations l Notebook which was used to inform the operating crews via required readin Further, the licensee stated that a T/S amendment was not necessary; however, I T/S 15.3.7 basis would be amended by May 1,1997, to clarify the T/ ' The examiners reviewed training material pertaining to the 125 VDC Electrical Distribution System which included the System Description (TRHB 12.7, Revision i 4), Electrical Distribution Integrated Operations Lesson Plans (LP2440, Revision 0; j LPO121, Revision 10), and the DC System Operation Lesson Plan (LP2449, Revision O). The differences between battery chargers D-107, D-108, D-109 and

D-07, D-08, D-09, regarding their ability to function as designed if connected to a l

. DC bus without the related battery also connected was not identified in the training . material.

' The examiners queried a few licensed operators regarding the battery charger limitations and differences. The operators questioned had varying degrees of knowledge regarding this issue. Some operators knew that battery chargers D-107, l D-108, and D-109 could not function if connected to a DC bus without a connected ' battery while other operators only indicated that they knew the chargers were designed differently and could not provide any specific I Conclusions , The licensee took prompt action to identify this issue to the operating crews via required reading. However, based on responses provided by operators questioned,

the level of knowledge regarding this issue varied. Clarification of T/S 15.3.7 basis i

was necessary to remove the implication that T/Ss allcwed aligning the 125 VDC system outside design capabilities which the licensee stated would be completed by May 1,1997. Additionally, the examiners concluded that operator training regarding the operating limitations and differences between the 125 VDC system battery chargers was a weakness.

- 05.6 Simulator Fidelity The Examiners observed some simulator modeling deficiencies during examination administration which are documented in Enclosure 3, " Simulation Facility Report."

Two deficiencies identified had minor impact on examination administration (the inability to increase Emergency Diesel Generator KVARs, Enclosure 3, Item #3; and the inability to manually control "A" feedwater regulating valve, Enclosure 3, item 

 #5). Additionally, one license applicant was confused by an indicated thermocouple tilt on the plant computer during a rod position indication malfunction (Enclosure 3, item #1) and incorrectly diagnosed the event as a dropped ro ,

.._ _ __ -- . .______ ___. .__ _ _- _.. . . . _ . . . .
'

i , 

.
'

The examincts concluded that the identified simulator deficiencies did not preclude } completion of valid evaluations of license applicant performance and that the impact on simulator training would be minimal.

L Manaaement Meetinas i

X1 Exit Meeting Summary The examiners conducted an exit meeting with members of licensee management i on October 11,1996, and on November 27,1996. The licensee acknowledged the j findings presented and indicated that the materials reviewed were not considered
proprietary.

d 

       !
I i

i l 1  ! + l I h.

$ i ! 1 1 e 1 I

I d b I . e

j i

a

___ 

.- _ _ . . . . _ _ _ _ . . - - _ _ . _ _ _ . _ _ _ _ . _ _ _ _ . _ _ __ _  _ . _ . _ _ _ . . _ . . _ _ _ . _ _ _ . . __ .__ _
.
:
!'
-

i, PARTIAL LIST OF PERSONS CONTACTED i Licensee j S. Patulski, Site Vice President j C. Cerovac, Assistant Operations Manager, Programs .' C. Gray, Assistarn Operations Manager K. Grote, initial Program Administrator

T. Guay, Regulatory Services Manager
R. Harper, Training Performance Shift Superintendent
R. Seizert, Training Manager l P. Smith, Acting Operations Training Coordinator j T. Staskal, Operations Manager

! l NRC A. McMurtray, Senior Resident inspector

l

i i

s q
.

g f4 i

4 i ! i a

h $

4 < i <

l 

!

i

t 

!

i s ! 10 

     .,. ,     .

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

.

Enclosure 2 l 4 Post Examination Comrr'ents and Review

l Licensee Submitted Comments and NRC Resolution:

The licensee's written examination post review resulted in the following two
questions submitted for NRC review:

i ' QUESTION (RO EXAM 044/SRO EXAM 043): i Given the following Unit 2 plant conditions: I 

 - Condenser vacuum is 25 inches H Generator load is 250 M Which one of the following is the MAXIMUM amount of time the turbine can l be operated (Operational Back Pressure region curve is attached)?
       '

a. Operation is prohibited, b.10 minute c.1 hou d. Operation is unrestricte . ANSWER: % REFERENCE: AOP-5A, Rev,4, pg.1; K/A 000051 A202 (3.9/4.1) LICENSEE COMMENT: The question was based on the appropriate use of the figure (provided as an attachment) from AOP-5A and the recognition that operation in the " avoid operation" region should be limited to no more than 10 minutes (discussion item 2.6 on page 1 of AOP-5A).

The given condenser information of 25 inches Hg (or 5 inches Hg absolute) is also addressed in step 6.1 of AOP-SA which required reactor and turbine trip if condenser pressure lowers to 5 inches Hg absolute. This corresponds to answer "a" for this questio We recommend that both "a" and "b" be accepted for this questio l NRC RESOLUTION: Licensee comment partially accepted. Based on the technical information provided there was only one correct answer for the question. A condenser vacuum of 25 inches Hg (5 inches absolute) required a reactor trip and turbine trip per AOP-5A and therefore, operation in the " avoid operation" region was prohibited. The answer key was changed to accept choice "a" as the correct respons > 

.
. QUESTION (RO EXAM 029/SRO EXAM 028):

Which one of the following is the reason for the 30 second time delay associated with the AMSAC actuation relay? i a. Permits operator override in the event of spurious actuatio b. Prevents spurious actuation caused by S/G shrink and swell, c. Allows time for reactor power to decrease below 40% before actuatio d. Allows time for the AFW pumps to recover S/G levels before actuatio ANSWER: REFERENCE: TRHB 13.3, Rev,4, pg. 37; K/A 012000K406 (3.2/3.5) LICENSEE COMMENT: The word " override" used in the correct answer implies a blocking or defeating of the AMSAC circuitry which would not be an appropriate response to an AMSAC actuation signal. The TRHB used as a reference for this question states that the 30 second delay is to " allow the operator time to correct the problem" such as starting the other feed pump, manually opening a feed reg valve, etc. None of the remaining distractors provide a correct response. We recommend that this question be thrown out due to there being no correct answe NRC RESOLUTION: Concur with recommendation. This question was deleted from the examinatio B. NRC Post Written Examination Review: The NRC's post examination review identified the following questions where at least 50% of the applicants answered the question incorrectly. These were considered generic weaknesses and are being provided to the Point Beach training staff for consideration and implementation into their SAT based program:

Question # Descriotion of Knowledae Weakness RO 015; SRO 014 Control rod response in automatic following failure of a power range nuclear instrumen RO 030; SRO 029 RVLIS (Reactor Vessel Level Indication System) readings that would indicate the highest probability of core voidin RO 039; SRO 038 Symptoms of a stuck control rod requiring entry into AOP-6B, 

  " Stuck Rod or Malfunctioning Position Indication."

      /

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

. RO 039; SRO 038 Required method to restore Reactor Coolant Pump (RCP) seal cooling following a station blackou RO 045; SRO 044 Plant status concerns that the Operators would be confronted I with following a station blackout in excess of seven hour ! RO 050; SRO 049 Conditions that allow securing RCPs per CSP-C.1, " Response To inadequate Core Cooling."

RO 051; SRO 050 The immediate actions of EOP-0, " Reactor Trip or Safety injection," following the Main Turbine's failure to trip automatically and manuall RO 070; SRO 069 The preferred method to rapidly reduce turbine load in response to a casualty per the " Managers Expectations."

RO 077; SRO 076 The minimum Condensate Storage Tank (CST) volume required by T/Ss when only one CST is supplying both unit RO 079 The equipment that is reset by the rod control step counter reset switc RO 086 The FSAR design ratings for the station batterie RO 094 The pressurizer pressure control system response to a pressure detector failur SRO 095 The conditions when EOP-0.0, "Rodiagnosis," could be used based on operator judgemen I

3 _ - - 

       ]

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

. O

l Enclosure 3 , SIMULATION FACILITY REPORT ! Facility Licensee: Point Beach Nuclear Plant i Facility Licensee Docket Nos: 50-266; 50 301 i Operating Tests Administered On: October 8-11,1996

This form is to be used only to report observations. These observations do not constitute

, audit or inspection findings and are not, without further verification and review, indicative ' of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC i certification or approval of the simulation facility other than to provide information that l may be used in future evaluations. No licensee act.. .., required in response to these

observations, i j While conducting the simulator portion of the operating tests, the following items were

observed

i , ITEM DESCRIPTION ! l 1. Plant Process Computer System A thermocouple tilt was incorrectly indicated ! during an individual rod position indication

malfunction for rod D-10. This was NRC j identified. The licensee generated Simulator

' Discrepancy Report (SDR) 96-0111 to track and troubleshoot.

! 2. Rod Control Master Counter The rod control master counter reading in the simulator booth indicated 556 while actual l reading should have been 543 following 4 simulator reset. This was NRC identified. The licensee generated SDR 96-0112 to track and 4 troubleshoot.

. 3. Emergency Diesel Generator Technical Specification Test 81, " Emergency Diesel Generator G-01 Monthly," step 4.24 i required G01 load at 2500 KW and KVARs at

1875-2025 for the one hour run. However, 1 KVARs could not be raised greater than 165 This was NRC identified. The licensee generated 'l SDR 96-0113 to track and troubleshoot.

1 4. Plant Process Computer The Plant Process Computer did not reinitialize i several times following simulator reset. The 

:    licensee had previously identified this problem

] and troubleshooting efforts were being tracked j by SDRs 96-0063 and 96-008 l

l l 

,    ,    __

     . . - . . - - . -

e e 5. "A" Feedwater Regulating Valve Manual control of "A" Feedwater Regulating Valve should have been functional during an "A" steam generator controlling feedflow channel malfunction; however,it was not. This resulted in an unplanned reactor trip. The licensee had previously identified this but could not get the problem to repeat. The licensee reopened a j previously closed Simulator Fidelity Report (SFR) l 

  #1121 to track and troubleshoo . Plasma Display  The Plasma Display tracked contrcl rod  ;

movement but incorrectly indicated rod position  ! 6 steps different than actual. The licensee had previously identified this problem and tre 2bleshooting efforts were being tracked by SDR 96-008 . . . - . . . . . . - -.. . . - - . - .. ...~ . - -. ~. . - . - . - - - - - . . . - - . - . . - .

1 U. S. NUCLEAR' REGULATORY COMMISSION SITE SPECIFIC EXAMINATION REACTOR OPERATOR LICENSE REGION 3 

,. ,  CANDIDATE'S NAME:   MASTER EXAMINATION FACILITY:    Point Beach 1 & 2 REACTOR TYPE:   PWR-WEC2

DATE ADMINISTERED: 96/10/07 I INSTRUCTIONS TO CANDIDATE: Usa.the answer sheets provided to document your answers. Staple this cover shsat on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at lecst 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE  % 44. 6e hd=0 0 0. 0 0 -  % TOTALS FINAL GRADE All work done on this examination is my ow I have neither given nor i received ai I Candidate's Signature 

        ,

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

, PAGES 2, 3, AND 4, WERE USED AS ANSWER SHEETS , AND ARE NOT INCLUDED IN THIS MASTER EXAMINATION COP , .

. l l l l l

_ _ Page 5 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examinatio This must be done after you complete the examinatio . RSstroom trips are to be limited and only one applicant at a time may leave. You must avoid all contacts with anyone outside the examination i room to avoid even the appearance or possibility of cheatin , 4. Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provide USE ONT.Y THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

7. Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question pag . Use abbreviations only if they are commonly used in facility literatur Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it ou . The point value for each question is indicated in parentheses after the questio . Show all calculations, methods, or assumptions used to obtain an answer to

any short answer questions.

, 11. Partial credit may be given except on multiple choice questions. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . Proportional grading will be applied. Any additional wrong information that is provided may count against yo For example, if a question is worth one point and asks for four responses, each of which is worth 0.25 ) points, and you give five responses, each of your responses will be worth 0.20 point If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answer . If the intent of a question is unclear, ask questions of the examiner onl . - . - - _ _ _ Page 6 14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition, turn in all scrap paper.

15. Ensure all information you wish to have evaluated as part of your answer is on your answer sheet. Scrap paper will be disposed of immediately following the examination.

~16. To pass #the examination. you must achieve a grade of 80% or greater.

17. There is a time limit of four (4) hours for completion of the examination.

18. When you are done and have turned in your examination, leave the examination i Grea (EXAMINER.WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoke . 

     !

i l 

     ;

i l

J 

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

e i . REACTOR OPERATOR Page 7 . QUESTION: 001 (1.00) Given the following Unit 2 plant conditions: 

- A large break LOCA has occurre St t=0, an SI signal was generate '
- At t=30 seconds, a containment spray signal was generate Which ONE of the following describes the times at which containment 4 spray components will operate?

a. t=0 sec: spray pumps A and B start ~; t=30 sec: spray discharge valves 860A, B, C, D open t=1 min and 30 sec: NaOH addition valves 836A and B open i t*0 sec: spray discharge valves 860A, B, C, D open t=30 sec: spray pumps A and B start t=1 min and 30 sec: NaOH addition valves 836A and B open t=30 sec: spray discharge valves 860A, B, C, D open t=40 sec: spray pumps A and B start t=2 min and 30 sec: NaOH addition valves 836A and B open d. t=30 sec: spray pumps A and B start t=40 sec: spray discharge valves 860A, B, C, D open t=2 min and 30 sec: NaOH addition valves 836A and B open QUESTION: 002 (1.00) Which ONE of the following describes the purpose of the back draft dampers installed in the Containment Air Recirculation System? a. Prevent unit air backflow when the accident fan is running and the cooling fan is not, Prevent backflow in a cooling unit in the event of fire in Containmen c. Serve as a system air backflow damper in idle cooling units (both accident and cooling fans secured).

c. Serve as explosion dampers preventing duct work collapse during an acciden .- - =. _ - __ .- . - . .. . REACTOR OPERATOR Page 8 QUESTION: 003 (1.00) Which ONE of the following explains why the steam generator level program is reduced at low power? a. To minimize time delays in plant transient response due to 

'" thermal lag",
-

b. To prevent thermal stratification above the U-tube c. To reduce the mass inventory available to boil off in the event of a steam brea d. To prevent low power level o.scillation due to level dominant control.

. , QUESTION: 004 (1.00) Which ONE of the following is the final location for the Unit 1 Control Operator after leaving the control room during AOP-10A, " Control Room Inaccessibility", when the control room is inaccessible due to a fire? a. Auxiliary Feed Pump Room l b. 4160 Vital Switchgear Room c. PAB Elevation 8' Emergency Diesel Generator Room T QUESTION: 005 (1.00) i Which ONE of the following would result if the reactor head to vessel inner "O" ring seal completely fails? , Pressurizer Relief Tank level would increase.

! SI on high containment pressur c. SI on low PZR pressur Reactor Coolant Drain Tank level would increase.

. v 

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

4 REACTOR OPERATOR Page 9 i i j QUESTION: 006 (1.00) l Givsn the following Unit 1 plant conditions: l . l - Tavg is 552 deg F. 

- Tref is 550 deg F.

! - Rod control is in' Automatic.

- Control bank D is stepping OUT.

1 i Which ONE of the following describes the required IMMEDIATE ACTION (S)? I a. Manually trip the reactor and perform EOP-0, Reactor Trip or i Safety Injection immediate actions.

l' b. Select Bank D rods and verify rod control is operable by driving Bank D in and ou c. Place the control bank selector switch to Manual.

. ! d. Maintain rods in automatic because the rod motion is in response to the Tavg/ Tref mismatch.

.

1 ' QUESTION: 007 (1.00) Which ONE of the following conditions would require a reactor trip during a Loss of Component Cooling Water (CCW) ., according to AOP-9B, 

" Loss of Component Cooling"?

a. CCW low flow alarms on the SI pump b. A CCW high radiation monitor alarm is receive ] l c. CCW pump discharge low pressure alarm is receive d. Unable to maintain CCW Surge Tank leve .~ ,.

JREACTOR OPERATOR Page 10 QUESTION: 008 (1.00) According to 10 CFR 20, which ONE of the following is the definition of Total Effective Dose Equivalent (TEDE)? a. It is the sum of the Deep Dose Equivalent (DDE) and the iCommitted Effective Dose Equivalent (CEDE).

b. It is the sum of the Deep Dose Equivalent (DDE) and the Committed Dose Equivalent (CDE).

c. It is the sum of the Shallow Dose Equivalent, Whole Body (SDE, WB) and the Committed Effective Dose Equivalent (CEDE). It is the sum of the Shallow Dose Equivalent, Max Extremity (SDE, ME) and the Deep Dose Equivalent (DDE).

QUESTION: 009 (1.00) Given the following Unit 1 plant conditions: 

- Unit 1 is in Cold. Shutdow RCP 1A motor is uncoupled from the pum RCS is filled and vente Maintenance is inspecting RCP 1A seal Which ONE of the following minimizes leakage of reactor coolant upward clong the RCP shaft?

a. The pump shaft mates with the thermal barrier casin , b. Nozzle dam installation prevents RCS water from entering the RCP shaft are c. Seal injection is maintained during this conditio d. Seal leakoff collects any RCS leakage up the shaft and directs it back to the VC REACTOR OPERATOR Page 11 QUESTION: 010 (1.00) Which'ONE of the following is the ALTERNATE power supply for the 2 P2A Ch9rging Pump? a. B08 

;

b. B09- B03 d. 2B03 QUESTION: 011- (1.00) Which ONE of the following actuations WILL be caused by an AUTOMATIC SI, but WILL NOT be caused by a manual-SI? a. Containment Spray b. Containment Isolation c. Closure of emergency diesel generator output breakers onto safeguards busses d. Trip of main feedwater pumps QUESTION: 012 (1.00) Which ONE of the following safety injection actuation signals would be automatically unblocked if pressurizer pressure increased to 1800 psig Gfter safety injection had been manually blocked? I' a. Low pressurizer pressure onl b. Low pressurizer pressure and steamline low pressure onl : c. Low pressurizer pressure, steamline low pressure, and containment high pressure onl d. Low pressurizer pressure, steamline low pressure, containment high pressure, and manua REACTOR OPERATOR Page 12 QUESTION: 013 (1.00) Which ONE of the following curves (shown on the following page) is indicative of an under compensated Intermediate Range. Nuclear Instrument? a.1 Curve I b.. Curve II c. Curve III d. Curve IV QUESTION: 014' (1.00) Which ONE of the following describes why the Source Range detectors are located at the lower quarter of the core rather than at core centerline? a. The instrument tubes are angled in toward_the core, so this location places the detectors closer to the flux they are to detec b. Neutron flux is greater in the bottom half of the core during a-startu c. To allow a spare detector to be installed in the upper half of the instrument should the installed detector fai d. This position provides optimal cooling for the detector by natural circulation of air in the detector wel m A ,.4 a A.-,.J *..a d..1--_.,J A ah...J@_J-_J5 ._a u __4_4 ..___L., _m . A J - .e.- mJ m_2 ,a 4 l . l , 1 i i 10 4 1 

'

i l 10 5 ELECTRICALLY ADJUSTED COMPENSATED ION CHAMBER G e 104 E TYPICAL SHUTDOWN CURVES 

$ 10 4 E

o W m m 108 Y m

H 

$ 104
E

' o U / IV l 

$ 1010          l lil i N 10 11      \  -
      \  NEUTRON SOURCE LEVEL ll N N

3 O 2 

  .

  . .

6 8 

    .  . . \ . . .

10 12 14 16 18 20 22 24 TIME AFTER REACTOR SHUTDOWN (MINUTES) TYPICAL GAMMA COMPENSATED CURVE FOR A COMPENSATED lON CHAM"5R 

 .

l l l ' . I { 

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

! REACTOR OPERATOR Page ~_3 QUESTION: 015 (1.00) j l i Giv:n the following Unit 2 plant conditions:  ; 

- Unit is at 90% powe Rod control in automati Power Range NI N41 fails HIGH (top of scale).

Which ONE of the following describes the response of the control rods? 

(A cume no operator action).

a. Rods initially drive IN and then drive OUT to match Tavg with Tre b. Rods initially drive IN and then STOP and remain at that position, c. Rods initially drive OUT and then Drive IN to match Tavg with Tre d. Rods initially drive OUT and then STOP and remain at that positio QUESTION: 016 (1.00) Givsn the following Unit 1 plant conditions: 

- Reactor is shutdow Reactor decay heat is being removed by natural circulatio RCS pressure is 1550 psi Average core thermocouple temperature is 402 degrees Which ONE of the following describes 'he approximate amount of subcooling that exists in the RCS? deg deg j deg deg l
     ,

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

REACTOR OPERATOR Page 14' i l i QUESTION: 017 (1.00) Which ONE of the following describes the location of the core exit 1 thermocouples? , 1 1 a. They are-arranged in a horizontal plane just above the uppe i 

' support plat ...
       ]

! b. They are arranged in strings extending from the upper support j 

'

plate to the core mid-plan ' i c. They are arranged in a horizontal plane just above the upper ! core plate.

, d. They are arranged in strings extending from the upper core plate to the core mid-plane.

. , 

*

, QUESTION: 018 (1.00) . Which ONE of the following conditions is the reactor vessel cooling fans j j (cavity cooling and control rod shroud fans) used to mitigate during a l' l natural circulation cooldown in accordance with EOP-0.2, " Natural l Circulation Cooldown"? i a. Creep of the reactor vessel head flange bolts.

I b. Formation of a steam bubble in the reactor vessel hea ; c. Catastrophic failure of the reactor vessel flange o-ring ~ ) d. Damage to the Core Exit Thermocouples's electrical circuitry

from overheating.

k 

   .
     --
      - .-

_ _ REACTOR OPERATOR Page 15 

     ,

     )

QUESTION: 019 (1.00)  ! , Which ONE of the following conditions will result in an automatic trip i of 911 operating Condensate Pumps? l ' a., Steam generator feed pump suction pressure at 179 psi ) b. Containment pressure at 5.2 psi ! c. Steam generator level at 73%. l d. Condenser hotwell level at 10 inche l QUESTION: 020 (1.00) hhich ONE of the following conditions must be met in order to reopen a bypass feed regulating valve from the control room after a valid automatic isolation from Hi-Hi S/G level has occurred? a. Momentarily depressing the feedwater control valve bypass reset pushbutton (FWCV BYPASS) once the closure signal has been l cleare l l b. Locally reset the closing solenoid valves once the closure I signal has been cleared and rese j c. Go to manual on the feedwater control valve controller and open the valv l d. The condition which caused the isolation signal must be restored to its pre-trip valu REACTOR OPERATOR- Page 16 QUESTION: 021 (1.00) Which.ONE of the following components is the largest supply of gas to ths Waste Gas System during normal full power operations? a. Volume Control Tank 

\. -

b. CVCS Holdup Tank c. Pressurizer Relief Tank d. Reactor Coolant Drain Tank I QUESTION: 022 (1.00) Tho'high pressure sensing line on the RCS loop B flow instrument FT-414 break Which ONE of the following describes the resulting RCS flow indication? 1 a. Only one (1) RCS loop B flow instrument is affected (PT-414) with HIGH flow indication, b. Only one (1) RCS loop B flow instrument is affected (PT-414) with LOW flow indicatio c. All RCS loop B flow instruments are affected (PT-414, FT-415, FT-416) with HIGH flow indication, j d. All RCS loop B flow instruments are affected (PT-414, FT-415, FT-416) with LOW flow indicatio ]

-REACTOR OPERATOR Page 17 QUESTION: 023 (1.00) Which ONE of the following is indicated if a light is ILLUMINATED on the SI/ Spray Ready Status Panel? . a. The component has lost AC powe , b..The component is.in an abnormal alignmen c. The component has lost DC control' powe d. The component is running / energize . QUESTION: 024 (1.00) Which ONE of'the following "A" SI Accumulator parameters needs to be corrected for the "A" SI Accumulator to be OPERABLE while the reactor is operating at 100% power? a. Water Volume is 1130 cubic fee b. Pressure is 730 psi c. Boric acid concentration is 1910 pp d. Outlet Isolation valve is OPEN and the control switch RED position indicating light is NOT LI . - . . - . . - . . - . , . . - . . _ . - - . ~ . - . . . ... - - . . - - - ~ . - . . - , . . - . - - . REACTOR OPERATOR Page 18 ' s, e l QUESTION: 025 (1.00) i j Given the following Unit 1 plant conditions:

! 

 -

Pressurizer pressure defeat sw, itch is in its normal position.

+ - PT-449 (Yellow Channel), Pressurizer Pressure, has just failed 

 ' LO li
. Which ONE of the following describes the response of the pressurizer

, pressure control system to this failure? e

a. Only PORV 431C cannot operate automatically (PORV 430 operable).

l b. Only spray valve 431A closes (431B remains as is).

c. Both spray valves CLOSE full d. Both PORVs are PREVENTED from automatic operatio QUESTION: 026 (1.00) 

     . .

i Which ONE of the following statements describes how plant operations are , affected if Loop A RCS Wide Range Pressure instrument, PT-420, fails ' HIGH during Low "emperature Overpressure Mitigation System operation? a. Pressurizer PORV PCV-430 opens onl b. Pressurizer PORV PCV-430 opens and all pressurizer heaters  ! deenergize onl , 

        .

c. Pressurizer PORV PCV-430 opens and both pressurizer spray valves open onl d. Pressurizer PORV PCV-430 opens, all pressurizer heaters i deenergize, and both pressurizer spray valves ope l 

 . . _ _ . _ .

REACTOR OPERATOR Page 19 QUESTION: 027 (1.00) Giv:n the following Unit 2 plant conditions: t 4 

- Unit is at 50% powe l
- Rod control is in MANUA l
- Lo6p B cold leg temperature ';ector TE-401B fails hig No operator action is take l
     '

Which ONE of the following will be the steady-state pressurizer level? a. 20% b. 33% % % 

,

QUESTION: 028 (1.00) 

; Which ONE of the following conditions on LT-428 would result in an
increase in indicated pressurizer level?

. a. A leak in the reference leg of the pressurizer level transmitter, LT / 2 8 .

b. Pressurizer liquid temperature increases.

! c. The reference leg for LT-428 cools down due to a decrease in containment temperatur d. Containment pressure increases to 0.3 psig; containment temperature remains constant.

,. d f I

l l REACTOR OPERATOR Page 20 

   &ScjN  h I'20 QUESTION: 029- (1.00)-  [gt sM Which ONE of the following is the reason for the 3 d time delay associated with the AMSAC actuation relay?    l a. Permits operator. override in the- nt of spurious actuatio '- l      l b. Prevents spurious actuat caused by S/G shrink and swel .l l

c. Allows time for tor power to decrease below 40% before actuatio d._ Allo ime for the AFW pumps to recover S/G levels before uatio , QUESTION: 030 (1.00) Which ONE of the following RVLIS readings indicates the highest probability of core voiding? a. Wide Range reading 98 ft, with NO RCPs runnin b. Narrow Range reading 38 ft. with NO RCPs runnin c. Wide Range reading 65 ft. with ONE RCP runnin l d. Narrow Range reading 95 ft. With BOTH RCPs runnin l i l l l l I 

   . - - _ , , - . . -

REACTOR OPERATOR Page 21 QUESTION: 031 (1.00) Given the following Unit 1 plant conditions: 

- Unit is operating at 75% steady state powe All systems are in automatic contro \The "A" S/G atmospheric valve fails ope Which ONE of the following describes the plant response to this condition? (Assume no operator action is taken.)

a. Turbine load decreases by 5%, reactor power remains stable at 75%. b. Turbine governor valves open in response to lower steam header pressure to increase turbine load to 80%. c. Control rods insert to maintain reactor power at 75%. d. Control rods withdraw and raise reactor power to 80% where it stabilizes.

QUESTION: 032 (1.00) Which ONE of the following describes the normal, emergency, and altornate emergency sources of power to Safeguards 4.16 KV bus 1A06? Normal Emergency Alternate Emergency ______ _________ ___________________ a. Load Bus 1A04 EDG G03 EDG G04 b. Load Bus 1A04 EDG G01 EDG G02 c. Load Bus 1A03 EDG G01 EDG G04 d. Load Bus 1A03 EDG G03 EDG G02

l 

-._m . _ - _ . _ . . _ . _ . _ . > _ _ . . _ _ . . _ . . . . ~ . _ . . . _ . . . _ _ _ . - _ . . _ .
    -    . . . . - . . . _ . - _ _ . _ . . _ . _ . . _ . . -

REACTOR' OPERATOR Page 22

, , QUESTION: 033 (1.00) ! l Which ONE of the following will provide an automatic shutdown of the gas - turbine while operating in local' control? l a. Low axial compressor suction pressure 

 -

j t j b. Low lube oil level ! c. Low lube oil pressure d. Low control air supply pressure QUESTION: b34 (1.00)- Which ONE of the following is the maximum rated run time for Emergency Diesel Generator G01 loaded at 3050 KW?

a. 30 minute ' l i' hou ; c. 24 hour hour ; l QUESTION: 035 (1.00) Which ONE of the following RMS displays must be selected to access alert and alarm setpoints? a. Status Grid b. Sector Display c. HDSR d. Trend 

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

REACTOR OPERATOR Page 23 l

1 QUESTION: 036 (1.00)  : Givsn the following Unit 2 plant conditions: 

 - RHR is in servic RCS pressure is 320 psig and INCREASIN RCS temperature is 340 degrees F and INCREASIN ALL system lineups are in a normal shutdown configuration for-solid-plant operatio Which ONE of the following will act FIRST to prevent overpressurizing tha'RHR System?

a. Pressurizer PORVs will ope RHR return isolation valve will auto clos c. RHR pump hot leg relief valve RH-PCV-861C will ope d. RHR pumps discharge relief valve RH-PCV-861A will ope l QUESTION: 037 (1.00) Which ONE of the following conditions will ARM the Turbine Crossover Steam Dump System? a. Selecting MANUAL on the mode selector switch, b. AUTO selected on the mode selector switch and a 10% load drop in less than 120 second c. Going to TEST on the individual valves on the back of C0 d. Reaching 90 psig in the crossover steam header.

I I l l l

o i

! 

      -

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

! l REACTOR OPERATOR Page 24

4 . QUESTION: 038 (1.00) i j Which ONE of the following is a symptom of a stuck control rod that would require entry into AOP-6B, " Stuck Rod Or Malfunctioning Position

Indication", following a transient?

i i a.'An individual RPI in 8 step disagreement with the bank demand i location.

I b. A variation in NIS instrumentation resulting in a quadrant tilt of 1.2%. j -c. A variation in core outlet thermocouples of 8% relative to l symmetric thermocouples.

! ! d. A> variation in axial flux of 1.2% of axial peak at any location j relative to symmetrical trace.

i

[ QUESTION: 039 (1.00)

Given the following Unit 1 plant conditions
- A loss of all AC power has occurred.

' 

- Power has been restored.

i - The crew has transitioned to ECA 0.1.

! - ECA 0.1 directs restoration of RCP seal cooling.

i a j Which ONE of the following is the method required to restore RCP seal j cooling? i 1 a. Seal injection flow is initiated to cool the seals to less than l 150 degrees F, then CCW flow to the RCP is established.

l l b. CCW flow is initiated to cool the seals to less than 150 degrees F, then seal injection flow is established.

l

! c. Seal injection flow is initiated to cool the seals to less than 190 degrees F, then CCW flow to the RCP is established.

l d. CCW flow is initiated to cool the seals to less than 190 degrees { F, then seal injection flow is established.

i . 

:

f 

.

! l

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

REACTOR OPERATOR Page 25 QUESTION: 040 (1.00) Givsn the following Unit 1 plant conditions: 

- A plant trip has just occurre control rods are stuck out of the core following the tri Anremergency boration has been initiated by the reactor operator in accordance with EOP-0.1, " Reactor Trip Response."

Which ONE of the following lists the MINIMUM injected volume of boric ccid necessary to satisfy the required amount of.boration? l a. 600 gallons ,200 gallons  ! 

-

I 

     :

c. 2,400 gallons I d. 3,000 gallons I QUESTION: 041 (1.00) Givan the following Unit 1 plant conditions: 

- Steam generator A is faulted due to a feed line break outside of-containmen The crew is performing actions of EOP-2, " Faulted Steam Generator Isolation."

- The AFW system is in operatio i Which ONE of the following actions concerning the AFW pumps is required by EOP-2? a. Shutdown the AFW Pumps immediatel l b. Maintain at least 50 gpm AFW flow to each S/G with narrow range levels less than 8%. l c. Run the AFW Pumps only if less than 200 gpm is available to the ~ S/G Isolate the AFW Pumps from S/G A (steam and AFW flow).

i 

     -

REACTOR OPERATOR Page 26 I l 

     '

, QUESTION: 042 (1.00) In cccordance with EOP-1, " Loss of Reactor or Secondary Coolant", which ONE of the following groups of parameters is required to be verified, in cddition to Pressurizer level, prior to terminating SI flow? a. ' RCS subcooling, secondary heat sink, and containment pressure

b. Secondary heat sink, RCS pressure, and RCS subcooling c. RCS subcooling, RCS pressure, and containment radiation levels d. Secondary heat sink, containment pressure, and RCS pressure 

.

QUESTION: 043 (1.00) Given the following Unit 1 plant conditions: 

- Unit is at 100% powe Condenser vacuum is decreasin Which ONE of the following condenser vacuum conditions will FIRST result in the loss of condenser steam dump availability?

a. 25" Hg vacuum b. 23" Hg vacuum c. 21" Hg vacuum d. 19" Hg vacuum i , I i

REACTOR OPERATOR Page 27 QUESTION: 044 (1.00) Given the following Unit 2 plant conditions: 

- Condenser vacuum is 25 inches H Generator load is 250 M L:

Which ONE of the following is the MAXIMUM amount of time the turbine can be operated (Operational Back Pressure region curve is attached.)? a. Operation is prohibited, minute hour, d. Operation is unrestricted.

QUESTION: 045 (1.00) Which ONE of the following concerns would an operator most likely be confronted with during a total loss of AC power in excess of 7 hours? a. Loss of secondary heat sink condition b. An unmonitored release of radioactivity c. Loss of containment integrity d. A Steam Generator overpressurization condition 

 . - . . . - - -.  - .

_ - . _ . . - - . . - . . - _ - . . - . . - . - . - . _ - - - . - .- l POINT BEACH NUCLEAR. PLANT AOP-5A

ABNORMAL OPERATING PROCEDURES MAJOR l Revision 4 l LOSS OF CONDENSER-VACUUM August 21,1995 ~ ~ 

                      >

l l @V FIGURE 1 

         .

i !, . j RECOMMENDED OPERATIONAL REGIONS FOR NUCLEAR UNITS 

  , ' ,     WITH 2 DOUBLE FLOW LOW PRESSURE ENDS. ,   ,
                   ,

! s - - 

    .i
        - - - -   ~ - - -
           :__  __ =-~.
           :.__ -
    .. . . . - _
         -- : :::_ r-         -

c. : f_;: : g; 

             .._~=..__=.+i. _
     ._
   :_ -   4 -
         -

_ 

          - _ ,  _.=  - --

_ 

                  -
                    .
                    : , ._f_ --__
    . . _  _ -   - _, -
          - -    - --
                 . _  ,  .
           . ,  _---   , ; , __ .; __ _ ; ,.   , _ . _ ,
   $-   ,. _ _ ', -  . : _    ' -

_ 

                  '
                     n

. . _ _ . - . , . - r i 

         .. '            .-

} 1 - 

              -
              -
    ._ 1 - f

_ _- _ _ r _ C__ -_ _ s ,= c 

               -

T _E : ' -_ r_ ::. -... . . . 'r 1 _ iz . : -- 

       -
              - i__..--_    . _ ,     --.  - _
            -
             <

ie t ,- - _ 

                 -- -__ .1 _ .-=-._  ___:_- _
W 4 =
- 7 __ - _ .. - -

_5- 1 _- 

            - '
            ---._,_'-'____-
               ,-
                 : ~ 21 __ __-
                     ---
                    ;; 3 :=
   >=
           . . - :-_'_ _ ._ i- , , , ,- , .    - ~_ ;_ -
     -
     - ._  __ ,  . . - .2         -] _ c1 F ri - : ,= -

1 3 -- 

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

r-.J: :"r.;_ - -T ' 

             -

! _J " ; - . :' _ :_ _ _ - 2 - 

                 , '. . -. -

O -_ _--c >

y . .. -=-~-r- E ~ ._ /

a 01 ,-_ - . _=_,__._:-_=,:-_- . . .; - - 

             -- ..
                - -
                 - - - -

l C3 _ 

    :: .-,:_. < --   sy *
           .
                  - [_ -  - - -
                     = :r- ,
                - -: ~ _
              ~
         ,1
          ==--g - . i v ---
,   4 3 _: r: __  --

V _._.. -- i r" .. ;= - 27 -- i -:- 

             ., ~' r- -
              "-
              - - _ :   .r_ -I ~.a ~;j";r;, --~

_2 

                     "[J

2 W 

    '
      --= ::_ ~ 1- =1
                 -

_ 

                   :- b
   ; _h

_.- Eg- -_E : 4 - 

         -- -~-- 1 '.;i r e-: 3__ .

1 a- ..-.,-- 

     .

_ _ _r -2-- - - _______e t _.=r, _. : .. .. _ _ : iE _-- _ =i _ _ r -- r_-'~r:- -- _~: ; - . ~ ' _ I D ,Z'--~._~ ~ - 212 - - :- L 

                    ~~~

2,,, .E.:. : i - 

 ,.. y *E - Jfi: r-  - " ; _ ~ - 1 _- - ~' Z r-    EE --  --  #. -~ T' "  -

E 2_ j . . Eg 

                    =r_E_"~.- ,r:r :. ,: -: c_
                . .
             ,

lj gw, w 12 -: - = e ;_ , 

          = ; =__.,s g y- r:._ .- .,.'  _,  ;,, .  : .;._ : =-;.1 35Er 35;;gggg E     _: .__.._:
              .,
   ' i _ i : ,-._   _:-           =
'

_ 

        ~--      _

s:__ E;_s=_resE 

           - 1 :[. ,_:#~- = ~:
                   ~

_ "'d3 

                     -

   < Ei_- _:E=r ,E 1:J;    i_ ~_~ _T~          _ _- jgEE3

-

a) 

     =- - _ 2 2 g ,,, :_ _ ::1: -- : . , - ,__-  - _=  -_: =_ 1 ..r;..

_;------  : _ _u:-:7 _ _1 s. . _s .=- 4 : -.-.. _ . 1 - ... 

              -:    :_  c =- . ;=-
   ' : , '-
             -2

_ ar : . ,- : =  : : , - .2: ==_=_ _ = 

         :
          .       _

j 

     . :-_:[ _r _;.--

L ::~ . . -  : '. _2 

          . _r ; ; ..;, r _ - -.:,
              -
               ' : - _.- ~-i '
                 . J - _' 'E_.5:EE
                   -
   - - - :.-
    : _":    , ; _ , _,   ,  _ . . _  _: - - :  : _ .~: , .=; ==> __

] 

   .

_ ..:._ ; :- .:; . . _ - ::_ .._ = .:, . . : :- - _ :: . ;_. . . .r s; _ -.:: u;_._;;_:g._ - _- _ 3 . :,: . 

     - gt _ . .. _ :.~.   :.ry ...: .. = ;:g; _ :_ _: :. j;: _. :. ...           . . . . . .- ,... :_: ;i = p: ;_: .:;;
    ,
     :
       .: : . : .- -- -
                ~

3 ~ r_ * g.- , 

           ,_ . . '- r + .; . . . T. ! . . . . .......t..:. : . i. : . r . ., .:.._..... . .. _ . ;;. . .:: _:. :.' Eg

_ 

                     = ==

_ _ ::__

I 0 10 0 200 300 400 500 600 70c , GROSS LOAD-MW i ' 393A421 . l i i a d 

!

> T I

i

i i

i 

!
 @
!

l j 

          % dk J'or 6ui6        REFERENCE USE
,

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

___ _ . . _ _ .. _ _ REACTOR OPERATOR Page 28 f QUESTION: 046 (1.00) Given the following Unit 1 plant conditions: 

- A loss of all AC power has occurre ECA-0.0, " Loss of All AC Power" is in effec Det ECA-0.0, certain Engineered Safeguards equipment control switches are placed in Pull-out.

J Which ONE of the following events is prevented by this switch alignment?

a. An uncontrolled cooldown of the RCS and possible reactor restar b. An uncontrolled use of water that may be needed for long term cooldow c. An uncontrolled start of large loads on safeguards AC buses, d. An uncontrolled depressurization of the RC QUESTION
047 (1.00)

l Which ONE of the following describes the response of Feedwater R0gulating Valves following loss of a RED or BLUE instrument bus? ,

a. May fail open; manual control will NOT be available to regain ' contro I b. May fail open; but manual contrcl will be available to regain control, l c. May fail closed; manual control will NOT be available to regain I control, d. May fail closed; but manual control will be available to regain j control.

i i

, m__ . _ . _ . = . . . . _ . . _ _ _ . _ . _ _ _ . . _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ . _ . . . _ . 

          '

! ! REACTOR OPERATOR Page 29 l 

          '

QUESTION: 048 (1.00) 

'Which ONE of the following locations / equipment is protected by a Halon
. System?

a. Diesel' Generator rooms I 

 ,;
.b. Hydrogen Seal Oil package c. Service Water Pump area d. Auxiliary Feed Pump room
 .

QUESTION: 049 (1.00) Technical Specifications require containment pressure to be between + psig and -2.0 psig during power operatio ~

Which ONE of the following is the MAXIMUM time containment pressure is' l allowed to be outside this band per Technical Specifications before  ; action must be initiated to shut down the plant? i-a, 30 minutes l b. 1 hour i i c. 6 hours hours l t i ! i !

!

i ! 

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

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

REACTOR OPERATOR Page 30 QUESTION: 050 (1.00) A LOCA has occurred and both RCPs have been started per CSP-C.1, i 

"Rasponse to Inadequate Core Cooling."

' Given the,below list of criteria:

I. Core cooling provided by low or high head SI

- II. Narrow range reactor vessel level greater tl'an 25 feet j III. Any RCS hot leg loop less then 350 degrees 1 , 

.IV. Core exit thermocouples less than 1200 degrees F
Which ONE of the following gives the three conditions from the above

list that allow securing of the RCPs per CSP-C.17 a. I; II and III I, II and IV _> I I, III and I , II, III, and IV.

J QUESTION: 051 (1.00) Given the following Unit 1 plant conditions: 

- The Reactor tripped from 100% powe EOP-0, " Reactor Trip or Safety Injection" has been entere The main turbine did NOT trip as expecte MANUAL turbine trip is unsuccessfu Which ONE of the following is the NEXT action required per EOP-0?

a. Manually close MSIVs and Bypass valve b. Locally trip the main turbin c. Go to CSP-S.1, " Response to Nuclear Power Generation /ATWS."

d. Verify Safeguards Buses energize . - - 

     ,

t 1 l ' 

- REACTOR OPERATOR    Page 31 i

i . l ! QUESTION: 052- (1.00) l Which ONE of the following is the reason for the main feedwater l l icolation following a reactor. trip? 1' a. To prevent thermal shocking of the' steam generator (S/G) ! ! 'tubesheet I

i b. To prevent excessive cooldown of the RCS.

i i c. To prevent S/G overpressurizatio ;d. To preserve secondary water.for a subsequent RCS cooldown. 

(

i' . l ! QUESTION: 053- (1.00) 

-

Given the following Unit 1 plant conditions:

i - A reactor trip and safety. injection has occurre Pressurizer PORV 1RC-431C lifted and is stuck ope Which ONE of the following is the MAXIMUM pressure below which the PORV isolation valve must be shut? a. 2450 psig psig c. 2225 psig psig 

     !
     ,
 . , ~ . - - - , - - . - - - . - -- w

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

_ 

,

7 i' 

-REACTOR OPERATOR      Page 32

1 j QUESTION: 054 (1.00) ! 

<
-Given the following Unit 1 plant conditions:
, - Unit has tripped from 100% due to a small break LOC l i

' 

- Conditions have stabilized and operators are evaluating the    !

briteria for terminating SI.

' 

'
--Adverse Containment conditions do NOT exis '
- A maximum of 50 gpm feedwater-flow is available to each S/G.
. Which ONE of the following conditions.would PREVENT SI termination per j EOP-1.2, "Small Break LOCA Cooldown and Depressurization"?

a. Both steam generator levels ir^1cate .5% N !

a

b. Pressurizer level indicates 11 I

j c. RCS subcooling'is 40 degrees Pressurizer pressure is 2050 psig.

i

j QUESTION: 055 (1.00) l In procedure OP-5A, " Reactor Coolant Volume Control" there is a i PRECAUTION that states: "Do not secure letdown flow without also I securing charging flow ..." i i- Which ONE of the following statements describes why charging flow should j also be isolated?.(Assume all systems are in a normal at power lineup.)

!. i a a. VCT level will decrease until charging pump suction shifts to i the RWST.

b. Reduce thermal shock on the Non-Regenerative Heat Exchange c. VCT level will decrease causing possible damage to the charging pumps.

!-

d. Reduce thermal shock on the charging penetration into the RCS.

4.

b 

}

:
'

f , 

   -  - - c , - - --- _

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

J I ' REACTOR OPERATOR Page 33 J-f QUESTION: 056 (1.00) j _Given the following Unit 1 plant conditions: 

        '

i - Unit is in Cold Shutdown with RHR Cooling in progress.

I - The RCS is solid.

i 

 - kHR flow is lost and CANNOT be restore l
 - All other systems and components are availabl l l

I Which ONE of the following methods of cooling will be utilized to remove ! the core decay heat?

a. Feed a S/G using an AFW pump, and bleed steam through the j respective S/G atmospheric relief valv i i b. Start a charging pump, with flow through an RHR heat exchanger,

and initiate Hot Leg Injection.

1 c. Feed the RCS with Safety Injection, use letdown to remove decay 

,  heat.
d. Start a charging pump, with flow through an RHR heat exchanger,

> and initiate Cold Leg Injection.

i

i QUESTION: 057 (1.00) i Given the following Unit 1 plant conditions: ] ! - A reactor trip occurred 30 minutes ago due to a Steam Generator ] tube ruptur The crew is taking the actions of EOP-3, " Steam Generator Tube ' Rupture".

- N-35_ indicates 1.4E-10 amps and N-36 indicates 1.6E-10 amps.

j - Source Range Instruments are NOT energized.

! Which ONE of the following actions is required by the Reactor Operator?  ; l . l a. Allow N36 to decay to less than 1.5E-10 amps which will automatically energize the SR instrument b. Depress both SR RESET pushbutton c. Depress both IR BLOCK pushbutton d. Deenergize N3 _ _ . . . _ . _ _ _ . _ . _ _ . _ . _ _ - . _ _. .__ _- _ _ _ ____ _ _ _ _ _ _ . i

REACTOR _ OPERATOR Page 34

.: 

'

QUESTION: 058 (1.00)

Which ONE of the following conditions requires tripping the reactor per l AOP-1A, " Reactor Coolant Leak", for a confirmed steam generator tube i lenk? Assume appropriate actions have been taken per AOP-1A regarding chargingandletdownsystemoperation ,

a. Pressurizer level is at 11% and decreasing. 

'

b. Narrow range S/G level is 62% and increasing.

! I c. VCT level is at 17% in AUTO control and decreasing.

i d. 3/G Blowdown Isolation valve closes due to high radiatio , ! { QUESTION: 059 (1.00) i

  • The crew is responding to a ruptured tube in 1B steam generator-(S/G)

using EOP-3, " Steam Generator Tube Rupture."

i i Given the following plant conditions: 

- 1A S/G pressure is 950 psi B S/G pressure is 1050 psig.

l Which ONE of the following is the required core exit temperature that j the RCS must be cooled down to prior to depressurization? (Page 15 of , EOP-3 attached).

a. 505 degrees F

b. 490 degrees F

c. 485 degrees F 

-

d. 480 degrees F U.

. I i 

!

I ,

e - >,m ., , - ,

_ . _ _ _ _ _ _ _ __ _ _ _ . _ _ _ _ _ . . _ . _ . _ . . _ . .._ _ _ . . . . - . . _ _ . _ _ _ . . ._ _ ' d , t , POINT REACH NUCLEAR PLANT EOP-3 Unit 1 

 '~

EMERGENCY OPERATING PROCEDURES MAJO Revision;19.

, ! STEAM GENERATOR TUBE ~ RUPTURE ^ October;26, 1995..

)

! UNIT-1 STEP 

        .
         - -
         ==.
ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED i

CAUTION:

iStep) THE FOLLOWING STEP MAY REQUIRE A C00LDOWN OF MORE THAN 100 F IN DNE HOUR.

i NOTE: RCP trip criteria do HQI apply af ter a RCS (Procedure) controlled cooldown has been initiated.

.

20 Initiate RCS Cooldown i f Using ruptured steam l generator pressure, detemine l target"coretexit temperature. 

.

) Ruptured Steam Generator Core Exit Pressure (psig) Temperature (*F) i

1100 

    [460) 505 1000  [445] 490
900 [435) 480 g 800
    [420] 465

700 

    [405) 450 600
    [390) 435 (    500
    [370] 415 400
    [350) 395 340
    [335) 380
          ,
   . . .

Step 20 Continued on Next Page . . . 

.     /

r=. 

    %

1: m: :7 3 .

REACTOR OPERATOR Page 35 QUESTION: 060 (1.00) Which ONE of the following statements explains why AFW flowrate is procedurally restricted to 100 gpm when recovering steam generator (S/G) Icvol if the level has fallen below 55 inches on the wide range indication? 

,

a. To minimize water hammer to the S/G feed rin b. To prevent reactor restart from an excessive cooldow c. To minimize thermal stresses t) S/G components, d. To prevent exceeding reactor vessel cooldown rate limi l QUESTION: 061 (1.00) l 

     !

Which ONE'of the following is the basis for stopping all RCPs in CSP- j H.1, " Response to Loss of Secondary Heat Sink"? a. It establishes natural circulation to enhance the blead and feed capability of safety injectio b. It extends the time available to restore feed flow before bleed and feed criteria is me c. It anticipates an RCS pressure decrease caused by spray valves opening when air is restored to containmen d. It anticipates an RCS pressure decrease caused by opening PZR PORVs during bleed and fee l REACTOR OPERATOR Page 36 l-l QUESTION: 062 (1.00)' l Loca of which ONE of the following distribution panels will result in a l DUAL PLANT TRIP? a.,D13, b. D16 l c. D18 i l d. D21 

,

l QUESTION: 063 (1.00) l I With the Pressurizer Level Control Selector Switch in the NORMAL l position, a pressurizer level instrument failure causes the following SEQUENTIAL plant event Charging flow is reduced to minimum.

I - Pressurizer level decrease Letdown. flow is secured and heaters turn of Pressurizer level increases until high level trip occurs.

l ! Which ONE of the following instrument failures could have occurred? 

(Assume NO operator action)

a. Pressurizer level channel 428 (Blue) failed low Pressurizer level channel 428 (Blue) failed high c. Pressurizer level channel 427 (White) failed lo Pressurizer level channel 427 (White) failed hig __ _ _._. _ ._ _. _ _. _ . . _ . . . _ . . . . _ _ . . _ . _ . - _ . - _ . _ . _ - _ _ _ _ . . _ . . - _ _ _ . _ l REACTOR OPERATOR Page 37 ! ! QUESTION: 064 (1.00)  ;

' Givsn the following Unit 2 plant conditions: - i 

- The plant has tripped from 100% power due to a loss of off-site    !

powe The crew is required to verify natural circulation in EOP-0.1, , 

" Reactor Trip Response."

I i Which ONE of the following parameters satisfies one of the criteria for i indication of natural circulation? l , a. Steam generator pressures slowly trending upward.

, b. RCS Cold leg temperatures are increasing.

! l c. RCS subcooling is 36 degrees > d. Core exit thermocouples at saturation temperature for steam generator pressur , QUESTION: 065 (1.00) j Which ONE of the following " Procedure Usage Levels" allows performance of all activities from memory? . Infrequent Use ! Information Use t ' c. Refelence Use d. Continuous Use I t , i I , l , l ' 

      - -_

_ , _ _ -. 

     )

REACTOR OPERATOR Page 38 ] l I QUESTION: 066 (1.00) l l Th3 applicability statement for Procedure EOP-0, " Reactor Trip Or Safety i Injection", indicates that this procedure is used for initiating events ; occurring where RCS hot leg temperature is greater than or equal to 350 ' d*grees $ . Which ONE of the following describes the applicability of EOP-0 when RCS temperature is below 350 degrees F and a LOCA occurs? a. EOP-0 cannot be used, procedure EOP-0.0, "Rediagnosis" is to be use b. EOP-0 cannot be used unless directed by the Critical Safety Function Status Trees, c. EOP-0 cannot be used unless a step-by-step evaluation is made to determine if each action is still applicabl d. EOP-0 cannot be used, Shutdown Emergency Procedures (SEPs) are to be used.

QUESTION: 067 (1.00) Which ONE of the following is the MINIMUM number of shifts, per 10 CFR 55, on which you must actively perform operator or senior operator duties to maintain your license in an ACTIVE status? (ASSUME 8 hour shifts.)

a. 5 shifts per calendar quarter I shifts per calendar year l 

     '

l c. 7 shifts per calendar quarter 1 shifts per calendar year l

REACTOR OPERATOR Page 39 i QUESTION: 068 (1.00) Which ONE of the following is the LATEST time that alcohol consumption ; would be allowed prior to assuming the watch at 1800 per NP-1.7.5, ; 

     '
" Fitness for Duty Policy and Procedure"?

a.'0600 j b. 1000 l 

     !

QUESTION: 069 (1.00) A bomb threat has been received and the Duty shift Superintendent (DSS) has initiated a plant evacuation. A suspicious device has been diccovered in an emergency diesel generator (D/G) room by the extra reactor operato Which ONE of the following states the actions to be taken by the individual discovering the1 device, in accordance with NP 1.7.10? a. Immediately notify the DSS by two-way radio and then leave the D/G roo Immediately notify DCS by two-way radio and then leave the D/G roo c. Leave the D/G room, and then notify the DSS by phone, d. Leave the D/G room, and then notify DCS by phone, j

l REACTOR OPERATOR- Page 40 i i QUESTION: 070 (1.00) Which ONE of the'following is the preferred method to rapidly reduce turbine load in response to a casualty per the Managers Expectations? 

     !

l a. Operator Automatic Load Rate Contro \, : b. Turbine manual using governor valve decreas c. Valve Position Limite l d. Turbine manual using Governor fas l 

     \

QUESTION: 071 (1.00) Which ONE of the following is an individual who is authorized to be a member-of the fire brigade and to perform operations health physics functions?' a. Unit 1 Control Operator b. Unit 2 Control Operator c. Duty; Technical Advisor d. Extra Operator QUESTION: 072 (1.00) Which ONE of the following describes the process of determining an instrument's accuracy by comparing the indication to other independent instrument channels measuring the same paramete a. Channel Calibration b. Channel Functional Test c. Channel Check d. Channel Verification l 

     :

i REACTOR OPERATOR Page 41 

     :

i QUESTION: 073 (1.00) , l 

     '

An oncoming Reactor Operator has worked the following schedule: Day: 1 2 3 4 5 6 7 8 9 10 11 12 13 14 , l Hrs: 12 8 12 8 8 8 8 12 8 12 12 12 12 ? 

     >

Which ONE of the following is the MAXIMUM number of hours the individual i may work on day 14 without obtaining special authorization assuming the l operator has a minimum of 8 hours off between each shift? a. 2 hours , b. 4 hours c. 8 hours d. 12 hours QUESTION: 074 (1.00) Which ONE of the following is correct regarding INDEPENDENT verification of control board sliders? a. Must be done by the Unit Control Operato b. Must be done by visually observing the operating lin c. Can be accomplished using status lights, d. Can be accomplished using remote indicator .- .. ._ _ REACTOR OPERATOR- Page 42 

     !

. , QUESTION: 075 (1.00) Which ONE of the following conditions is required in order to perform work on a system using " Positive Control" in lieu of danger tags? a. A QC certified inspector is presen V : b. Work is to be performed in a high radiation are c. Requires direct line of sight of the isolatio Isolation boundaries are locke i 

     !

QUESTION: 076 (1.00) )

Giv@n the following Unit 1 conditions: , l 

     '
.l - Unit is at 100% powe VCT level is 23%.
- All controls are in automati LT-112, VCT level transmitter, fails hig Which ONE of the following describes the FINAL ACTUAL VCT level? (Assume l no operator action.)    1

, a. Increases to 78% and then diverts to the HU Increases to 28% where auto-makeup stop I c. Decreases to 17% where auto-makeup initiates, d. Decreases to 0% (empty).

'l

REACTOR OPERATOR Page 43 QUESTION: 077 (1.00) Givsn the following plant conditions: 

- Both units are operatin The South Condensate Storage Tank (CST) is isolated for repair The North CST is selected for Auxiliary Feedwater Pump suction at both unit Which ONE of the following is the MINIMUM volume required in the North CST?

a. 13,000 gallons ,000 gallons ,000 gallons ,000 gallons QUESTION: 078 (1.00) WHICH ONE of the following describes the initial output signal from the rod control system for a 12% step change in turbine load with power at 80% as compared to a 12% step change in turbine load with power at 40%? Assume rods in automati The initial output for the step change at 80 % is: larger due to the response of the Variable gain unit, larger due to the response of the Power mismatch rate / lag uni smaller due to the response of the Variable gain uni smaller due to the response of the Power mismatch rate / lag uni .

_ . . . _ _ _ _ _ _ _ _ . _ . . _ _ . ._ _- . _ _ . _ _ . _ -. . . l ' REACTOR OPERATOR Page 44 i QUESTION: 079 (1.00) Which ONE of the following groups of equipment will be reset by the rod control step counter reset switch? a. Slave cycler, master cycler, urgent alarm, RPI, NR-45 recorde b. P/A converter, master cycler, multiplexer, bank overlap unit trigger circui ,

c. Alarm circuits, master cycler, slave cycler, step counter, bank overlap unit, and P/A converter, d. Master cycler, slave cycler, step counter, P/A converter and urgent alar . QUESTION: 080 (1.00) Which ONE of the following describes the effect of VCT pressure being at 13 psig on the operation of the Reactor Coolant Pump (RCP) seals? a. The pressure is too low and insufficient seal flow will be obtained from the #2 sea b. The pressure is too low and insufficient seal flott will be obtained through the #3 sea c. The pressure is too high and will force excessive flow through the #2 RCP sea l d. The pressure is too high and will force excessive flow through the #3 RCP sea i

l 

  -. .

_ _ _ _ . _ _ _ . . _ . _ . _ . . _ _ . . . _ _ . _ . . - . REACTOR OPERATOR Page 45 QUESTION: 081 (1.00) During a normal reduction in power using boration, which ONE of the following is the reason that additional pressurizer heaters should be ennrgized? a.'Al* low an increased ramp rate for the power change b. Ensure positive pressurizer pressure control is estaIblished prior to starting the power change c. Maintain PZR pressure in normal operating range during the power change d. Equalize the reactor coolant system and Pressurizer boron concentrations QUESTION: 082 (1.00) Which ONE of the following indications on the RMS Station Grid would be indicative of an RMS Area monitor that has failed LOW 7 > Tha detector unit and channel number would be colored ... a. light blue b. yellow c. violet d. white I 

  . . . _ _ - _

REACTOR OPERATOR Page 46 4 QUESTION: 083 (1.00) Which ONE of the following ds used as the reactor power input to the Rod i Insortion Limit (RIL) compacer? j l a. Average Delta-T b. Auctioneered High Tavg ) c. Turbine Impulse Pressure d. Auctioneered High Power Range NI 

.

QUESTION: 084 (1.00) Which ONE of the following describes the purpose of the Containment Purge System? a. Reduce containment atmosphere activity levels prior to personnel entry for the monthly containment inspectio b. Remove airborne radioactivity from containment immediately following a LOC c. Provide additional cooling to containment during hot shutdown condition d. Remove airborne radioactivity from containment during outage QUESTION: 085 (1.00) Which ONE of the following components is designed to limit the rate of affected S/G blowdown during a steam line break inside containment on Unit 27 a. Main Steam Line siz b. Main Steam Flow Ventur c. Main Steam Isolation Valv d. Main Steam Flow Limite . . - - . . . _ _ .

REACTOR OPERATOR Page 47

l

l QUESTION: 086 (1.00) Giv:n the following Unit 2 plant conditions: 

- The reactor was operating at 100% power when a reactor trip occurre A.1oss of all AC power occurred at the time of the reactor tri I Which ONE of the following is the length of time, based on FSAR
 '

rsquirements, the DOS station battery is DESIGNED to carry its normal chutdown loads? l a. 30 minutes i hour c. 8 hours l d. 24 hours

QUESTION: 087 (1.00) i Tha operation of which ONE of the following systems / components ir ! interlocked with a SPECIFIED NUMBER [ minimum or maximum] of Circulating Water Pumps to be in operation? . I a. Steam dump system b. Control Room ventilation condenser c. Air ejectors , d. Main turbine above 50% rated load ,

1

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

       .
       .

REACTOR OPERATOR Page 48 l l 

' QUESTION: 088   (1.00)

Which ONE of the following is the reason a nitrogen overpressure is mtintained in the Pressurizer Relief Tank (PRT)? a. Reduce the amount of water in the PRT necessary to quench 

 ' pressurizer relief valve discharge.

L b. Reduce the potential for overpressurizing the PRT and prevent deformation of the rupture disks.

I c. Reduce the potential for an explosive mixture of hydrogen and l oxygen in the PRT.

! d. Reduce the potential for corrosion in the PR : i l l l QUESTION: 089 (1.00) Which ONE of=the following conditions would result in AUTOMATIC closure of the CCW surge tank vent valve, RCV-017, if open? I a. High level in the CCW surge tank.

! b. Low level in the CCW surge tan c..High radiation level at CCW surge tank vent.

l d. High radiation level at CCW pump suction.

l QUESTION: 090 (1.00) Which ONE of the following describes the primary purpose of the activated charcoal filters in the Containment Cleanup System? a. Remove radioactive iodine, b. Remove moisture and water droplet c. Remove particulate matter.

. d. Remove explosive gasses.

I i

. 

  - . ,

I REACTOR OPERATOR Page 49 QUESTION: 091 (1.00) Which ONE of the following is the reason the Post Accident Containment Ventilation System (PACVS) is placed in service when hydrocan concentration is about 3%? a.'To ensure containment hydrogen analyzer indications remain accurat b. To ensure explosive mixtures will not exist in the Auxiliary Building exhaust syste c. To ensure hydrogen sampling system capacity is not exceede I d. To ensure oxygen concentration is sufficient to support hydrogen recombiner operation.

QUESTION: 092 (1.00) Which ONE of the following fuel handling components does NOT require l air / nitrogen'for operation? l a. Fuel transfer car b. RCCA change fixture carriag c. Manipulator crane gripper assembl d. Control rod drive shaft unlatching too :

{ REACTOR OPERATOR Page 50 QUESTION: 093 (1.00) Which ONE of the following explains why the pressurizer spray valves rcmain operable on a loss of instrument air to Unit 1 containment? a. A backup regulator from service air to Unit 1 containment 

' instrument air supply will open to supply air to Unit 1
-

containmen b. A backup regulator from the installed nitrogen twelve pack is lined up when at power.

, j c. A backup regulator from the Unit 2 containment instrument air ' supply will open to supply air to Unit 1 containment, d. A backup regulator from an installed nitrogen bottle is lined up when at powe ; I l l l QUESTION: 094 (1.00) Giv:n the following Unit 1 plant conditions: i 

- The plant is at 100% powe Pressurizer pressure is in AUTOMATIC contro The Pressurizer Pressure Channel Defeat switch is in the normal positio No operator action is taken.

I Which ONE of the following actions occurs when pressurizer pressure detector PT-431 fails LOW? a. Pressurizer PORV (PCV-430) cycles to maintain pressure below the reactor trip setpoin I b. Pressurizer heaters and spray valves operate normally to I 

     '

maintain pressurizer pressure.

I l Pressurizer pressure DECREASES, resulting in a LOW pressure ) l reactor tri Pressurizer pressure INCREASES, resulting in a HIGH pressure l reactor trip.

I 

     !

t 

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

_ REACTOR OPERATOR Page 51 QUESTION: 095 (1.00) Wh n recovering a dropped rod in control bank C, an " Urgent Failure" alarm is received. This condition normally blocks rod motion for the affscted bank in AUTO and MANUA Which ONElof the following reasons explains why the recovery of the dropped rod is possible? a. Selecting the specific bank with the Bank Selector Switch overrides the bloc b. Rod motion is allowed since only 3 of 4 disconnect switches for the affected power cabinet are opened, c. Rod motion is allowed when the Demand Step Counter for the affected banks is reset to zer d. Resetting the Pulse-to-Analog (P-A) Converter removes the bloc QUESTION: 096 (1.00) Which ONE of the following describes the flow path that allows REFLUX cooling to remove heat from the core? a. Boiling in the core with steam flowing out the break to the sump for recirculation and then back to the hot leg b. Vapor bubbles formed in the core and condensing in the S/Gs to flow back to the core through the cold leg c. Boiling in the core and steam condensing in the S/Gs to flow back to the core through the hot leg ! d. Vapor bubbles formed in the core and condensing in the head area to flow back into the cor l

i REACTOR OPERATOR Page 52 I l 

     ;
     ,

l QUESTION: 097 (1.00)  ! l Which ONE of the following conditions requires the main steam isolation l valves to be shut during the performance of CSP-S.1, " Response to Nuclear Power Generation /ATWS"? ' a.1If a steam generator tube rupture is identifie I b. Upon steam generator level reaching the high level alar If RCS temperature decreases below 547 degrees d. Actions to trip and runback the turbine are unsuccessfu l l QUESTION: 098 (1.00) l l Uhich ONE of the following conditions will result in the AUTOMATIC l closure of Waste Condensate Discharge valve WL-018? l a. Waste Condensate Pumps tri Loss of Circulating Water flo c. High level alarm on the Waste Holdup Tan . d. An alarm condition on Radiation Monitor Channel RE-21 QUESTION: 099 (1.00) Which ONE of the following is the pressure at which IA-3014, " Service Air to Instrument Air Backup Valve" will BEGIN to open on a loss of 3 Instrument Air? a. 90 psig i i psig psig d. 75 psig l

l REACTOR OPERATOR Page 53 l QUESTION: 100 (1.00) Which ONE of the following actions is required immediately by the escort-if a visitor becomes separated from the escort? - a. Make a page announcement to direct roving security personnel to Treport to last known locatlon of visito l b. Make a page announcement to direct the visitor to report to the 

     '

control roo c. Notify security _to deenergize vital area door card readers, d. Notify Central Alarm Statio , i l i 

.
 (********** END OF EXAMINATION **********)

- . . . , , . - - . - . . .-. - . - . . . . - . - . . . . . . . . . . . . . _ . . - . . . . . . . . . . . . . . . . - . -
~ REACTOR' OPERATOR  <       Page 54

. ANSWER:  001 ('1. 0 0 )        l l
          ' I i

REFERENCE: 

  *
 . g TRHB'10.12, Re , pg. 6; C1301AOT; Setpoint Document, Section 2.4; and Logic Sheet l l

KA 026000A301 [4. 3 /4. 5] j

l 026000A301 ..(KA's) 

  .

ANSWER: 002 (1.00) REFERENCE: l TRHB 10.1, Rev. 4, pg. 15 and C1805 COT and TRHB Figure 10.-l.24- 

-KA 022000G007 [3.3/3.5)
          !

022000G007 ..(KA's) i ANSWER: 003 (1.00) j i ! l , REFERENCE: l TRHB 13.7, Re , pg. 2.

l l KA 059000A302 [2.9/3.1) 

          ,

059000A302 . . ( FA ' s ) r l !. l ! 

,
. , -  - - -  ,.m-..
     -

e REi.CTOR OPERATOR Page 55 , 

     !
     !
     !

' ANSWER: 004 ( l '. 0 0 ) ,

REFERENCE
,

8 . AOP-10A, Rev. 18, pg. 1 [Similar to question used on 94 RO exam.]  ; KA 000068A101 [4.3/4.5] 000068A101 ..(KA's) 

.
     '

ANSWER: 005 (1.00) t 

     :

REFERENCE: TRHB 10.15, Rev. 5, pg. 7 and C4313AOT and P&ID Sheets 541F091 Sheet 1 ! and.9645971 Sheet 1A KA 068000K107 [2.7/2.9] 

     ,

068000K107 ..(KA's) 

     .

ANSWER: 006 '(1.00)  ; f i l 

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

'

- l 

' REACTOR OPERATOR      Page'56  I f REFERENCE:

'

AOP-6C, Re , pg. 3 and LP 2441,.EO'2.1 KA~0000017.205.[4.4/4.6] 

    .
(Similar question used on 95 SRO exam.]

. v . l 0000611205 ..(KA's)

? 1.

i . ANSWER:

007 (1.00) ! d.

j.

l-REFERENCE: AOP-9B, Re , p .

i

'

KA.000026G010 [3.6/3.5] 4 i 

(Similar to question on 95 SRO exam.]
000026G010 ..(KA's)
         !

I 008' AMSWER: (1.00) - l

'

i l l I { REFERENCE: { 10 CFR 20.1003, Definition [Similar question used on.95 SRO exam.] l KA 194001K103 (2.8/3.4] l .

194001K103 ..(KA's) 

.

l i i i

$ > 

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

    !

REACTOR OPERATOR Page 57 l 

   *

ANSWER: 009 (1.00)- 5.

REFERENCE: . 5.- ? i TRHB 10.2, Re I 5, pg. 1 i KA 003000K407 (3.2/3.41

003000K407 ..(KA's) , l l ANSWER: 010 (1.00) j 

    : l l

REFERENCE: l

TRHB 10.6, Re , pg. 71 and Master Data Book for B0 I KA-004000K203 [3.3/3.5]

004000K203 ..(KA's) ANSWER: 011 (1.00) b.

REFERENCE: TRHB 10.16, Rev. 3, pg. 12 and C7304 COT and Logic Sheet 7 KA 013000A302 (4.1/4.2] , I 013000A302 ..(KA's) l 

   .

REACTOR OPERATOR -Page 58-ANSWER: 012 (1.00) b.

REFERENCE: t - TRHB 13.4, Re , pg. 18 and C730 COT and Logic Sheet KA 013000A402 [4.3/4.4] 013000A402 ..(KA's) 

.

ANSWER: 013 (1.00) d.

REFERENCE: TRHB 13.1, Rev. 3, Figure 13.1.19 and C1315 CO KA 015000A202 [3.1/3.5] '

015000A202 ..(KA's) ANSWER: 014 (1.00) i b.

REFERENCE: TRHB 13.1, Re , pg. 11 and C1310 CO KA 015000K601 [2.9/3.2] 015000K601 ..(KA's) 

    -_ _ . _ _ _

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

REACTOR OPERATOR Page 59 ANSWER: 015 (1.00) b.

REFERENCE: , TRHB 13.8, Re , pg. 6 and C0105 COT; and TRHB 13.8, pg. 19 and Logic Shtot 1 KA 015000K302 [3.3/3.5] 015000K302 ..(KA's) ANSWER: 016 (1.00) c.

REFERENCE: Steam Tables KA 017020K502 [3.7/4.0] 017020K502 ..(KA's) ANSWER: 017 (1.00) C.

REFERENCE: TRHB 13.2, Re , pg. 8 and TRHB Figure 10. KA 017020K102 [3.3/3.5] 017020K102 ..(KA's) 

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

1- ).  ; 

'

1 

; REACTOR OPERATO Page 60

. j. ANSWER: 018 .(1.00) 

. l a

1 REFERENCE: * - s . i

, BG EOP-0.2, Rev. 13, pg. 7 and C7033 CO l 

; KA 022000K404 [2.8/3.1]

., j hl

022000K404 ..(KA's) i s 

 #

J $ ANSWER: 019 (1.00)

l I ,

,t , ?! j= REFERENCE: ' TREB 11.2, Re , pg. 11 and C3912 COT and Logic Sheet 24 . KA 056000A204 [2.6/2.8)

t  ; j 056000A204 ..(KA's) i . \

l ANSWER: 020 (1.00) i t a.

i 

,
REFERENCE

4 TRHB 13.7, Re , pgs 28 and 29 and C4210 COT and Logic Sheet 10 l.

i [EBQ 052-05-149A]

1 l KA'059000K419 [3.2/3.4] { 

; 059000K419  ..(KA's)

1

. k

3 

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

 . ._ _ . .- .. ._ . _ REACTOR. OPERATOR    Page 61

! ANSWER: 021 (1.00) REFERENCE: 

,

TRHB 10.15, Re , pg. 24 and C4504AOT and P&ID 648J972 Sheet 1 KA 071000A405 [2.6/2.6] 071000A405 ..(KA's) AMSWER: 022 (1.00) REFERENCE: TRHB 10.2, Re , Figure 10.2.1.B and P&ID 541F091 Sheet 1 KA 002000K606 [2.5/2.8) 002000K606 ..(KA's) ANSWER: 023 (1.00) REFERENCE: TRHB 10.8, Re , pg. 15 and Control Panel CO2 KA 006030A401 [4.4/4.4] 006030A401 ..(KA's) 

  -. .- .. - - . .

. REACTOR OPERATOR Page 62 ANSWER: 024 (1.00) . c.

! REFERENC,E: , TRHB 10.8, Re , pg. 18 and Technical Specification 15.3.3.

KA 006000G005 [3.5/4.2] I 006000G005 ..(KA's) l l

ANSWER: 025 (1.00) a.

j ' REFERENCE: TRHB 10.3, Re , Figure 10.3.7, Logic Sheet 18 and ICP 1 KA 010000A403 [4.0/3.8) . ' 010000A403 ..(KA's)

ANSWER: 026 (1.00)

a.

, REFERENCE: 1

TRHB 13.5, Re , pg. 21, C0905C01, Logic Sheet 18 and P&ID 541F091 Sheet 1 KA 010000K301 [3.8/3.9) 010000x301 ..(KA's) i 

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

s . ] REACTOR OPERATOR Page 63 

         -

s I i l ANSWER: 027 (1.00)  ! i I ' l l l REFERENCE: j i; j TRHB 13.6, Re , pg. 7 and C1006 COT and Logic Sheet 18 KA 011000K604 (3.1/3.11 ,

2 011000K604 ..(KA's) ! 

 .
ANSWER:

028 (1.00)

a.

l l REFERENCE: TRHB 13.6, Re , p . ' KA 011000A101 (3.5/3.6) 011000A101 ..(KA's) WER: 029 (1.00) 

         ! l

REFERENCE: g)ggppy I TRHB 13.3, Rev. 4 . wwW KA 012000K406 (3.2/ K406 ..(KA's) 

    @

_ 

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

REACTOR OPERATOR' Page 64 ANSWER: 030 (1.00) l- ' REFERENCE: N-l BG CSP-I.3, Re , pg. KA 016000G015 [3.6/3.8) 016000G015 ..(KA's)- i ! ANSWER: 031 (1.00) d.

! i. REFERENCE: ! TRHB 13.8, pg. 20.

l KA 035010k501 [3.4/3.9) l 035010K501 

         '
  ..(KA's)

l ANSWER: 032 (1.00) a.

l REFERENCE: 

         *

TRHB 12.5, Re , Figure 12.5.1 and Master Data Book KA 062000K201 [3.3/3.4) i 062000K201 ..(KA's) 

         '

i I ! ! 

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

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

[ i i REACTOR OPERATOR Page 65 I ! ' , l I j-. ANSWER: 033 (1.00) - t l l ,

REFERENCE: , 

 ,
 . .        ;

TRHB 12.2, Re , pg. 38, C4105AOT, OI-110, Rev. 1, and PC-29, Rev. 29 Monthly Best KA 064000K402 (3.9/4.2] 064000K402 ..(KA's) 

         ,
 '
         .
         (

ANSWER: 034 (1.00) 

         ' r I
         ,

REFERENCE: TRHB 12.8, Re , pg. 74 and OP-11A, EDG G01/G02 KA 064000A206 [2.9/3.31

1 064000A206 ..(KA's) ANSWER: 035 (1.00) REFERENCE: TRHB 13.12, Rev. 3, pg. 1 KA 073000A402 (3.7/3.7]

i 

         '

073000A402 ..(KA's) i I i l 

         !

I

REACTOR OPERATOR Page 66 ANSWER: 036 (1.00) a.

REFERENCE: 

'-

TRHB 10.7, Rev. 7, pg.'17, C0506AOT and Setpoint Document,. Sections 2.3, 8.2 and 1 KA 005000K109 [3.6/3.9) 005000K109 ..(KA's) 

.

ANSWER: 037- (1.00) d.

REFERENCE: 

.TRHB 13.10, Re , pg. 2; 62804 COT; Setpoint Decament 14.13; P&ID PBM-241;.OP-1c, Step 7. KA 041020K603 [2.7/2.9)

041020K603 ..(KA's) ANSWER: 038 (1.00) l l REFERENCE: ' AOP-6B, Re , pg. 1 and LP2441, EO ) KA 000005A201 [3.3/4.1] 000005A201 ..(KA's) 

      ,
 . . _ . - ,
  -
   - . , - - , . ,, . _ - , _ , , , 4

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

, i e - i < , REACTOR OPERATOR Page 67 i ! i r ANSWER: 039 (1.00) REFERENCE: , ECA 0.1, Rev. 10, pg. 12 j KA 000015G007 [3.1/3.2] 000015G007 ..(KA's) 

  .

ANSWER: 040 (1.00) REFERENCE: EOP-0.1, Rev. 14, p . KA 000024A205 [3.3/3.9) 000024A205 ..(KA's) ANSWER: 041 (1.00) l I REFERENCE:

EOP-2, Rev. 10, pg. j KA 000040A110 [4.1/4.11  ! 

      :

i 000040A110 ..(KA's) l l 

      !
      !

.
 .._....-.~ !

REACTOR OPERATOR Page 68 

     ,
     :

' ANSWER: 042 (1.00)  ! 

  '

h b.

.l . . l REFERENCE: ) s' .  : 1 l

EOP-1, Rev. 19, Foldout pag ;

KA 000040A205 (4.1/4. 5 ) 000040A205 ..(KA's) i I i l a 

 *

! ANS#ER: '043 (1.00) i C.

l

} REFERENCE: . 1.

j TRHB 13.9, Rev. 2, pg. 4 and Setpoint Document, Section 14.2.

1 i i KA 000051K301 (2.8/3.1) ) i - i 000051K301 ..(KA's) 

     'I ANSWER: 044 (1.00)
.L /fvt/Sves &//#Cw$ h45 CE/ />P P4M t-iTY
*

FERENCE: ls 5 r BM~v A amw csmm car- : AOP-5A,.Re , p .  ; KA 000051A202 (3.9/4.1) I 000051A202 ..(KA's) 

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

. 

           :

1  ! . ' REACTOR OPERATOR Page-69 ' ?

! 

           .,

i

ANSWER
045 (1.00)

i i4 !~ l i j 1 REFERENCE: 

 , .

LP 0462, Re , pg. 7 and LO K302 [4.3/4.61 000055K302 ..(KA's) AMSWER: 046 (1.00) . REFERENCE: BG ECA-0.0, Rev. 14, pg. 14 and LP 0462, LO KA 000055A106 [4.1/4.5] 000055A106 ..(KA's) ANSWER: 047 (1.00) 

           !

i REFERENCE: TR11B 12. 9, Re , pg. 1 A218 [3.1/3.1] l 000057A218 ..(KA's)

I I l REACTOR OPERATOR Page 70 j ANSWER: 048 (1.00) I REFERENCE: 3 . TRHB 11.14, Re , pg. 2 (Similar to EBQ 052-01-065A] I KA 000067K102 [3.1/3.9) 000067K102 ..(KA's) ANSWER: 049 (1.00) b.

REFERENCE: Tech. Spec. 15.3.6, pg. 15.3.6- KA 000069G003 [3.3/3.9) 000069G003 ..(KA's) ANSWER: 050 (1.00) , REFERENCE: CSP-C.1, Rev. 12, pg. 2 KA 000074K201 (3.6/3.9) 000074K201 ..(KA's)

REACTOR OPERATOR Page 71 ANSWER: 051 (1.00) d.

REFERENCE: 

,

EOP-0, Rev. 20, pg. 2 and LP 0405 LO 2.1 KA 000007A202 (4.3/4.6] 000007A202 ..(KA's) ANSWER: 052 (1.00) b.

REFERENCE: BG EOP-0, Rev. 21, pg. 9 and LP 0405, LO 2.1 KA 000007K301 (4.0/4.6] 000007K301 ..(KA's) ANSWER: 053 (1.00) b.

REFERENCE: EOP-0, Rev. 20, pg. 18 and LP 0405, LO KA 000008A101 (4.2/4.0) 

   !

000008A101 ..(KA's) 

   ;

REACTOR OPERATOR Page 72-ANSWER: 054 ' ( 1. 0 0 )

REFERENCE: s . 

,

EOP-1.2, Rev. 13, Foldout Page-and LP 0435, LO KA 000009A234 [3.6/4.2] 000009A234 ..(KA's) 

.

ANSWER: 055 (1.00) l REFERENCE:  ! OP-5A, Rev. 28, pg. 1 and THRB 10.6, Rev. 3, pg. KA 000022K307 [3.0/3.2] 

    !
    ~

000022K307 ..(KA's) ANSWER: 056 (1.00) ) \ I REFERENCE: SEP-1.1, Rev.-0, pg. KA 000025K101 [3.9/4.3] 

    !

000025K101 ..(KA's)

REACTOR OPERATOR Page 73 i i 

   !

i ' ANSWER: 057 (1.00) j . REFERENCE: 

'

1 i EOP-3, Rev. 19, pg. 3 ! KA 000032A101 [3.1/3.4] 1 

   :

000032A101 ..(KA's)  ; ANSWER: 058 (1.00) ' i REFERENCE: AOP-1A, Rev. 8, pg. KA.000037G011 [3.9/4.1) 

   ;

000037G011 ..(KA's) s ANSWER: 059 (1.00) REFERENCE: EOP-3, Rev. 19, pg. 15 and BG EOP-3, Step 20 (PROVIDE: EOP-3, pg. 15 of-36) 

'KA 000038A136 (4.3/4.5)

000038A136 ..(KA's)

_ -_ - __ . - . _ - . REACTOR OPERATOR Page 74 i 

    !

l l ANSWER: 060 (1.00) REFERENCE: s . BG CSP-H.5, Re , pg. 4 and LP 1998, EO ) KA 000054K102 [3.6/4.2] l 000054K102 ..(KA's) l ANSWER: 061 (1.00) i l REFERENCE:

BG CSP-H.1, Rev. 10, pg. i KA 000054K304 (4.4/4.6] 000054K304 ..(KA's) I ANSWER: 062 (1.00) a.

REFERENCE: AOP-0.0, Re , pg. KA 000058A203 [3.5/3.9) 000058A203 ..(KA's)

_ _. _..- _.. _ __ _._ _ . .. _ _- _ _ _ _ _ _ _ _ . - _ . . . . _ _ . _ . _ _ , _ - . . REACTOR OPERATOR Page 75 ANSWER: 063 (1.00) b.

REFERENCE: , 

'
.:

TRHB 13.6, Rev. 1, pg. 6 and Figure 13.6.1 and Logic Sheet 1 KA 000028K202 [2.6/2.7)  ; 000028K202 . . ( YA ' s ) i ANSWER: b64 (1.00) j l REFERENCE: EOP-0.1, Rev. 14, pg. 1 KA 000056K101 [3.7/4.2] 000056K101 ..(KA's) I

ANSWER: 065 (1.00) b.

REFERENCE: NP 1.1.4, Re , pg. KA 194001A101 [3.3/3.4] 194001A101 ..(KA's) l l l 

        !

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

\

!. REACTOR OPERATOR Page 76 , ANSWER: 066- (1.00) t REFERENCE: 

 \ -
=EOP-0, Rev. 20, pg. 1 and SEP-2, Rev. 2, pg. KA 194001A102 [4.1/3.9]

194001A102 ..(XA's) ANSWER: b67 (1.00) REFERENCE: 10 CFR 55.53 and OM 3.10, Rev. 5, pg. 1 KA 194001A103 [2.5/3.4] 194001A103 ..(KA's) ANSWER: 068 (1.00) REFERENCE:  ! NP-1.7.5, Re , p . i KA 194001A103 [2.5/3.4) j 194001A103 ..(KA's) l 

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

, REACTOR OPERATOR Page 77 i ,

ANSWER: 069- (1.00) i i

.
i
           '

! REFERENCE: 4 3 . l NP 1.7.10, Re , pg. 1.

KA 194001A105 [3.6/3.8) > i 

  .

1 , , 194001A105 ..(KA's) 

           '

i b i e i ANSWER: . 070 (1.00)

1 b.

! REFERENCE: ' l

Manager's Expectations, 93-1.

l KA 194001A103 [2.5/3.4] l 1 i i , j 194001A103 ..(KA's) I

l
,

i

l ANSWER: 071 (1.00) . d.

> REFERENCE: l 'OM 3.10, Rev. 5, pg. KA 294001A110 [2.9/3.9] 194001A110 ..(KA's)

1

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

           , i 1            l REACTOR OPERATOR          Page 78 i

i ! ANSWER: 072 (1.00) i C.

i REFERENCE: 

 * .
.

Tech. Spec. Definition 15.1.f.

a ! KA 194001A113 [4.3/4.1] a !- ' 194001A113 ..(KA's) i ANSWER: b73 (1.00) . b.

I ! REFERENCE: l NP 1.6.6, Re , p . ! l j KA 194001A103 [2.5/3.4] l

l 194001A103 ..(KA's)

s

ANSWER: 074 (1.00) REFERENCE: OM 3.17, Rev. O, pgs. 4 & KA 194001K101- [3.6/3.7) l 194001K101 ..(KA's) , 

           :

l I

REACTOR OPERATOR Page 79 

     '

ANSWER: 075 (1.00) l 

     .

REFERENCE: ) 

-
%.

l NP 1.9.15, Rev. 2, pg. 1 l KA 194001K102 [3.7/4.1) 194001K102 ..(KA's) ANSWER: 076 (1.00) d.

REFERENCE: TRHB 10.6, Rev. 3, pg. 35 and C0413 COT and Setpoint Document,-Section KA 004010A105 [3.0/3.2] i 004010A105 ..(KA's) ANSWER: 077 (1.00) C.

REFERENCE: TRHB 11.4, Rev. 5, pg. 15 and Tech. Spec. 15. KA 061000G005 (3.3/4.0) 061000G005 ..(KA's) 

   'i REACTOR OPERATOR  Page 80 ANSWER: 078 (1.00)

c.

REFERENCE: s . 

,

TRHB 13.8, Re , pg. 3,7 and C0105 CO KA 001000K403 [3.5/3.8) 001000K403 ..(KA's) ANSWER: 079 (1.00) 'C.

REFERENCE: TRHB 10.5, Rev. 4, pg. 1 KA 001010K604 [2.9/3.2) 001010K604 ..(KA's) ANSWER: 080 (1.00) a.

REFERENCE: i TRHB 10.6, Rev. 3, pg. 34 and C5702 CO KA 003000A205 [2.5/2.8) 

.003000A205 ..(KA's)

_ _ .

l REACTOR OPERATOR Page 81 ' ANSWER: 081 (1.00) REFERENCE: s . OP-3A, Rev. 37, pg. KA 004000K601 [3.1/3.3] 004000K601 ..(KA's) ANSWER: 082 (1.00) REFERENCE: TRHB 13.12, Re , pg. 15 and C5005 CO KA 072000A202 (2.8/2.9] t 072000A202 ..(KA's) ANSWER: 083 (1.00) ' a.

, REFERENCE: TRHB 13.13, Re , p . KA 014000A103 [3.6/3.8) 014000A103 ..(KA's) 

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

, REACTOR OPERATOR Page 82 ANSWER: 084 (1.00) o l 

'd.

! l . L REFERENCE: 

  , .

TRHB 10.01, Re , pg. 17.

l KA 029000G004 [2.9/3.0) L i l 029000G004 ..(KA's) , l , 

  -

l ANSWER: 085 (1.00) b.

i-l REFERENCE: l l TRHB 11.1, Rev. 6, pg. KA 039000K101 [3.1/3.2] 039000K101 ..(KA's) ANSWER: 086 (1.00) b.

t ' REFERENCE: l TRHB:12.7,.Re , pg. 7; and FSAR Section 8.2, pg. 8.2-19 l KA 063000A101 [2.5/3.3] l i 063000A101 ..(KA's) L i  ;

i n , I I 

. - -  ., , __

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

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

I 

,

REACTOR OPERATOR Page 83 

!
         '

! l ANSWER: 087 (1.00) i 

; REFERENCE:
<
 \ :

! TRHB 13.9, Re , pg. 4.

. l KA 075000A203 [2.5/2.7) ! 

         ,

t 075000A203 ..(KA's)  ! i' 

         :

l ANSWER: b88 (1.00) , i c.

i a

REFERENCE:

} 1 OP-4C, Rev. 13, pg. 1.

KA 007000G007 [2.9/3.1) l 007000G007 ..(KA's)

i . ANSWER: 089 (1.00) REFERENCE: TRHB 10.9, Re , pg. 5 and C0802AOT and RMS Alarm Setpoint and Rosponse Book 

[Similar to EBQ 051-06-009A)

KA 008000A204 [3.3/3.5) 008000A204 ..(KA's) 

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

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

) ! REACTOR OPERATOR Page 84 T d ] ANSWER: 090 (1.00)

9..

i j 

; REFERENCE:

j- . ' TRHB.10.1, Re , pg. 18.

! KA 027000K501 [3.1/3.4)

i j 027000K501 ..(KA's) i !

ANSWER: 091 (1.00) i.

q li j.

' 

' REFERENCE:

. 

[

J TRHB 10.11, Rev. 3, pg. 3.

(~ J KA 028000K502 [3.4/3.9) j ! 028000K502 ..(KA's) j

1

.,

, ANSWER: 092 (1.00)

i i

REFERENCE: TRHB 10.13, Re , pg. KA.034000G009 [3.0/3.0] 034000G009 ..(KA's) i l 

          . . .

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

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

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

i 

:
 ~

REACTOR OPERATOR Page 85 

:

i i ?

ANSWER: .093 (1.00) l > 

         !

i d.

. REFERENCE: i TRHB'11.18, Re , pg. [EBQ 052-06-005A] 1 I 

'

f KA 078000K302 [3.4/3.6] 

         '
,

u ' '

078000K302 ..(KA's) I 

 ,

i i i ANSWER:

094 (1.00) j l i l' REFERENCE: 

.TIUiB 10. 3, Rev. 3, Figure 10.3.7 and LP 2438 LO 2.1, Logic Sheet 18 and    -l

. the Setpoint Documen ! l ' i 

         '

l KA 000027A101 [4.0/3.9) i { 000027A101 ..(KA's)

i
.          .

l ANSWER: t 095 (1.00) i ' i { REFERENCE: , AOP-6A, Re , pg. 10; AOP-6A BG pg 16.

i 104 000003A102 [3.6/3.4)

000003A102 ..(KA's) 

,

i $ . - - 

    .-. _-   ,

REACTOR OPERATO Page 86 ANSWER: 096 (1.00) c , REFERENCE:

. '.. *

1 LP 0435, Re , pg. 11.

I.

KA 000011K101 [4.1/4.4]

3 K101 ..(KA's) i 4-Y

ANSWER: 097 (1.00) l d.

J ! REFERENCE: . ! CSP-S.1, Rev. 13, pg. 2.

i.

i KA 000029K308 [3.6/3.8] ] , ' 000029K308 ..(KA's) 

' ANSWER: 098 (1.00) REFERENCE:

TRHB 13.12, Rev. 3, pg. 49; RMS Alarm Setpoint and Response Book KA 000059K201 [2.7/2.8) 

.000059K201 ..(KA's)

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

i l 2, 

. REACTOR OPERATOR        Page 87
,

' l

4

l' ANSWER: e 099 (1.00) I b.

^ I' . , REFERENCE: s . AOP-5B,~Rev. 10, pg. 2.

l . l

KA 000065A201 [2.9/3.2]  ! i 

-
.
000065A201 . ,KA's)

ANSWER: kOO (1.00) * 

'd.

!

4

REFERENCE: ] NP 1.7.1, Rev. 1, pg. 3.

( KA 194001K105 [3.1/3.4] ,

- l l
:

194001K105 ..(KA's)

i 4-

i l !

i 

          !
          ,
   (**********  END OF EXAMINATION **********)
      - , - - - - _ . . _ . - .- . _ . -_ _ -

e j U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION

SENIOR OPERATOR LICENSE i REGION 3 " MASTER EXAMINATIO CANDIDATE'S NAME: ~ 4 FACILITY: Point Beach 1 & 2 REACTOR TYPE: PWR-WEC2 l DATE ADMINISTERED: 96/10/07

! INSTRUCTIONS TO CANDIDATE: Use the answer sheets provided to document your answers. Staple this cover i sheet on top of the answer sheet Points for each question are indicated in {- parentheses after the question. The passing grade reqaires a final grade of

at least 80%. Examination papers will be picked up four (4) hours after the

examination starts.

}i CANDIDATE'S l TEST VALUE SCORE % i

l 99fa 

)qht_100.u0   % TOTALS

FINAL GRADE !. .1 j All work done on this examination is my ow I have neither given nor i , received aid.

! I i Candidate's Signature

3 i i

k l'

3 

.

l

. PAGES 2, 3 AND 4 WERE USED AS ANSWER SHEETS AND ARE NOT INCLUDED IN THIS MASTER EXAMINATION COP ! 

   ,
   .

l l 

   ,
@'

_ _ _ . . - _ . _ _ _ - ... ~ _ _ > Page 5 l NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

l 2. After the examination has been completed, you must sign the statement on ! the cover sheet indicating that the work is your own and you have not ! received or given assistance in completing the examinat' ion. This must be done after you complete the examinatio . Restroom trips are to be limited and only one applicant at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG . Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question page.

l 8. Use abbreviations only if they are commonly used in facility literature, l Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answe Write it out.

! 9 The point value for each question is indicated in parentheses after the queFtion.

l 10. Show all calculations, methods, or assumptions used to obtain an answer to j any short answer questions.

l 11. Partial credit may be given except on multiple choice question Therefore, j ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

l 12. Proportional grading Will be applied. Any additional wrong information that is provided may count against you. For example: if a question is worth one point and asks for four responses, each of wnich is worth 0.25 points, and you give five responses, each of your responses will be worth 0.20 point If one of your five responses is incorrect. 0.20 will be deducted and your total credit for that question will be 0.80 instead of i 1.00 even though you got the four correct answers, l 13. If the intent of a question is unclear, ask questions of the examiner only.

t e

_ _ _ _ _ . _ _ . _ _ _ _ . _ _ _ _ _ _ . __ Page 6 14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheet In addition, turn in all scrap paper.

i 15. Ensure all information you wish to have evaluated as part of your answer is on your answer shee Scrap paper will be disposed of immediately following the examination.

> 16. To pass the examination, you must achieve a grade of 80% or greate . There is a time limit of four (4) hours for completion of the examinatio . When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoked.

t l l l 

,

e 

     '

t

I SENIOR REACTOR OPERATOR Page 7 ! l l l 

     .

! QUESTION: 001 (1.00) f l Which.ONE of the following describes the purpose of the back draft 

   -

!. dampers installed in the Containment Air Recirculation System? ! l a. Prevent unit air backflow when the accident fan is running and j the cooling fan is not.

, Prevent backflow in a cooling unit in the event of fire in l Containment.

i , c.. Serve as a system air backflow damper in idle cooling units l (both accident and cooling fans secured).

L d. Serve as explosion dampers preventing duct work collapse during ! an accident.

l l QUESTION: 002 (1.00) Which ONE of the following is the PREFERRED alternate AFW supply which must be aligned during a loss of. secondary coolant when the CST level l ' decreases to.less than 8 feet, according to EOP-1, " Loss of Reactor or i Secondary Coolant"? I l a. Align. service water to AFW pump suctio b. Pump the hotwell to the CST using a condensate pump c. Align AFW pump suction to the fire syste d. Connect a fire hose to the CST drain valve and backfill the CST, ! l

i 

*

l. . I i . _ . - . - 

    . . . ,

SENIOR REACTOR OPERATOR Page 8 . QUESTION: 003 (1.00) Which ONE of the following explains why the steam generator level program is reduced at low power? a. To minimize time delays in plant transient response due to 

" thermal lag".

, b. To prevent thermal stratification above the U-tube c. To reduce the mass inventory available to boil off in the event of a steam brea d. To prevent low power level oscillation due to level dominant contro QUESTION: 004 (1.00) Given the following Unit 2 plant conditions: 

- The plant has just been manually runback by 30% power within the last minut All systems are operable and in automati Actual Tavg is 9 degrees F higher than Tre Which ONE of the following describes the response of the steam dump valves?

a. Two dump valves full open and two modulating b. Four dump valves full open 

     )

c. Four dump valves full open and two modulating Six dump valves full open and two modulating 

.

     ;

SENIOR REACTOR OPERATOR Page 9 ] l QUESTION: 005 (1.00)  ; Which ONE of the following is the final location for the Unit 1 Control Operator after leaving the control room during AOP-10A, " Control Room Inaccessibility", when the control room is inaccessible due to a fire? a. Auxiliary Feed Pump Room b. 4160 Vital Switchgear Room c. PAB Elevation 8' d. Emergency Diesel Generator Room i QUESTION: 006 (1.00) Which ONE of the following would result if the reactor head to vessel inner "O" ring seal completely fails? a. Pressurizer Relief Tank level would increase, b. SI on high containment pressur c. SI on low PZR pressur d. Reactor Coolant Drain Tank level would increase.

QUESTION: 007 (1.00) Which ONE of the following conditions would require a reactor trip during a Loss of Component Cooling Water (CCW), according to AOP-9B, 

" Loss of Component Cooling"?

a. CCW low flow alarms on the SI pump b. A CCW high radiation monitor alarm is receive c. CCW pump discharge low pressure alarm is receive d. Unable to maintain CCW Surge Tank leve . 

.. .__ -_ . _ . . . _ . _ _ . _ . _ _ . ... - . - _ . ___ _ -. __ _ _ .___._~. . . ___ _

e I j 1 SEMIOR' REACTOR OPERATOR Page 10 l

i i-QUESTION: 008 (1.00) Given'the following Unit 1 plant conditions: 

 '
- Unit 1 is in Cold Shutdown.

i - RCP 1A motor is uncoupled from the pump.

{ - RCS is filled and vented. 

 - Maintenance is inspecting RCP 1A seal Wh1ch ONE of the following minimizes leakage of reactor coolant upward
.

along the RCP shaft? I a. The pump shaft mates with the thermal barrier casing.

. b. Nozzle dam installation prevents RCS water from entering the RCP shaft area.

Seal ~ injection is maintained during this condition.

' Seal leakoff collects any RCS leakage up the shaft and directs { it back to the VC ! i 

       !

QUESTION: 009~ (1.00) Which ONE of the following is the ALTERNATE power supply for the 2 P2A Charging Pump? a. B08 i B09 c. 1B03 B03 t 

       ,
 ..

. _ _ . _ . _ __ _ _ _ . . _  _ . _ . . _ . _ . . . ~ _ ~ . _ _ _ _ _ _ . _ _ . _ _ _ _ . . _ _ . _ _ _ . _ _ _
         !

J l SENIOR REACTOR OPERATOR- Page 11 ! l QUESTION: 010 (1.00) l l Which ONE of the-following actuations WILL be. caused by an AUTOMATIC SI, but WILL NOT be caused by a manual SI? i ! a.' Containment Spray l l b. Containment Isolation l c. Closure of emergency diesel generator output breakers onto l . safeguards busses

d. Trip of main feedwater pumps  ! I I l QUESTION: 011 (1.00) l Which ONE of the following safety injection actuation signals would be automatically unblocked if pressurizer pressure increased to 1800 psig after safety injection had been manually blocked? a. Low pressurizer pressure onl I l \ b. Low pressurizer pressure and steamline low pressure onl !

c. Low pressurizer pressure, steamline low pressure, and  ! containment high pressure onl l l 1 l d. Low pressurizer pressure, steamline low pressure, containment  ! j high pressure, and manua ! !  ! l 

         ;

QUESTION: 012 (1.00) l l- Which ONE of the following curves (shown on the following page) is , ' indicative of an under compensated Intermediate Range Nuclear  ! Instrument? . a. Curve I l b. Curve II c. Curve III {

d. Curve IV
 *

! !. 

,._ . . . , - -

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

          /!&

fY i 5 %/ , e i ! l i a 10 4 1 i i . 10 5 Y i 1 G ELECTRICALLY AOJUSTED COMPENSATED 13N CHAMBER - 

 $

w 104 I Q.

k TYPICAL SHUTDOWN CURVES m J C 10 7

E o a - w 

 <

i E i ' e 104 W w z i b w 104 E a

U IV l $ 1010 i i ll1 ~ l N i } 10 11 - 

           )
      %  NEUTRON SOURCE LEVEL

! j ll g l 

  .

-

10 12 . . . . \- 

       .
         .

2 4 . 

         .

6 8 10 12 14 s 16 18 20 22 24 , 1 TIME 

   * AFTER REACTOR SHUTDOWN (MINUTES)     l

- l 

           '

l

1 . i TYPICAL GAMMACOMPENSATED CURVE FOR A COMPENSATED lON CHAMBER  ! 

  -

i e 

..

.

< ,

i I SENIOR REACTOR OPERATOR Page 12 l l i i I l QUESTION: 013 (1.00) Which ONE of the following describes why the Source Range detectors are located at the lower quarter of the core rather than at core centerline? a. The instrument tubes are angled in toward the core, so this location places the detectors closer to the flux they are to detect, b. Neutron flux is greater in the bottom half of the core during a startu c. To allow a spare detector to be installed in the upper half of the instrument should the installed detector fai d. This position provides optimal cooling for the detector by natural circulation of air in the detector wel QUESTION: 014 (1.00) Given the following Unit 2 plant conditions: 

- Unit is at 90% powe Rod control in automati Power Range NI N41 fails HIGH (top of scale).

Which ONE of the following describes the response of the control rods? 

(Assume no operator action).

a. Rods initially drive IN and then drive OUT to match Tavg with Tref, b. Rods initially drive IN and then STOP and remain at that position, c. Rods initially drive OUT and then Drive IN to match Tavg with Tre d. Rods initially drive OUT and then STOP and remain at that positio .

SENIOR REACTOR OPERATOR Page 13 QUESTION: .015 (1.00) Given the following Unit 1 plant conditions: 

     '
- Reactor is shutdow , Reactor-decay heat is being removed by natural circulatio RCS pressure is 1550 psi Average core thermocouple temperature is 402 degrees Which ONE of the following describes the approximate amount of subcooling.that exists in the RCS?

a. 50 deg b. 100 deg c. 200 deg deg F.

QUESTION: 016 (1.00) 

     '

Which ONE of the following describes the location of the core exit thermocouples? a. They are arranged in a horizontal plane just above the upper support plat b. They are arranged in strings extending from the upper support plate to the core mid-plan c. They are arranged in a horizontal plane just above the upper core plat d. They are arranged in strings extending from the upper core plate , to the core mid-plan l 

..
     ;

_ - _ ._ . . _ . - __

SENIOR REACTOR. OPERATOR Page 14 QUESTION: 017- .( 1. 0 0 ) Which ONE of the following conditions is the reactor vessel cooling fans (cavity cooling and control rod shroud fans) used to mitigate during a natural circulation cooldown in accordance with EOP-0.2, " Natural Circulation Cooldown"? a. Creep of the reactor vessel head flange bolts, b. Formation of a steam bubble in the reactor vessel hea c. Catastrophic failure of the reactor vessel flange o-ring d. Damage to'the Core Exit Thermocouples's electrical circuitry from overheatin , i i

QUESTION: 018 (1.00) , I Which ONE of the following conditions will result in an automatic trip ' of all operating Condensate Pumps? 

     ]

a. Steam generator feed pump suction pressure at 179 psi ; b. Containment pressure at 5.2 psi c. Steam generator level at 73%. d. Condenser hotwell level at 10 inche i 

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

b $ SENIOR REACTOR OPERATOR Page 15 i i

i QUESTION: 019 (1.00) i Which ONE of the following conditions must be met in order to reopen a i bypass feed regulating valve from the control room after a valid

, automatic isolation from Hi-Hi S/G level has occurred? a Momentarily depressing the feedwater control valve bypass reset i pushbutton (FWCV BYPASS) once the closure signal has been j cleare Locally reset the closing solenoid valves once the closureL . signal has been cleared and reset.

i' c. Go to manual on the feedwater control valve controller and open i the valv l

l 

        '

l d. The condition which caused the isolation signal must be restored j to its pre-trip valu QUESTION; .020 (1.00) Which ONE of the following components is the largest supply of gas to j the Waste Gas System during normal full power operations? a. Volume Control Tank b. CVCS Holdup Tank Pressurizer Relief Tank d. Reactor Coolant Drain Tank  ; i O 

   . .  .- - .

SENIOR REACTOR OPERATOR Page 16 QUESTION: 021 (1.00) The high pressure sensing line on the RCS loop B flow instrument FT-414 break Which ONE of the following describes the resulting RCS flow indication? l a. Only one (1) RCS loop B flow instrument is affected (PT-414) ! with HIGH flow indication, b. Only one (1) RCS loop B flow instrument is affected (FT-414) with LOW flow indicatio c. All RCS loop B flow instruments are affected (PT-414, FT-415, FT-416) with HIGH flow indicatio d. All RCS loop B flow instruments are affected (PT-414, FT-415, FT-416) with LOW flow indicatio QUESTION: 022 (1.00) Which ONE of the following is indicated if a light is ILLUMINATED on the SI/ Spray Ready Status Panel? a. The component has lost AC powe b. The component is in an abnormal alignmen l c. The component has lost DC control powe I d. The component is running / energize ! l ) e i l 

   . SENIOR REACTOR OPERATOR    Page 17 I

QUESTION: 023 (1.00) Which ONE of the following "A" SI Accumulator parameters needs to be corrected for the "A" SI Accumulator to be OPERABLE while the reactor is operating at 100% power? a. Water Volume is 1130 cubic fee b. Pressure is 730 psi c. Boric acid concentration is 1910 pp d. Outlet Isolation valve is OPEN and the control switch RED position indicating light is NOT LI QUESTION: 024 (1.00) Given the following Unit 1 plant conditions: 

-

Pressurizer pressure defeat switch is in its normal positio PT-449 (Yellow Channel), Pressurizer Pressure, has just failed LO Which ONE of the following describes the response of the pressurizer pressure control system to this failure? a. Only PORV 431C cannot operate automatically (PORV 430 operable).

b. Only spray valve 431A closes (431B remains as is).

c. Both spray valves CLOSE fully, d. Both PORVs are PREVENTED from automatic operatio . % 

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

SENIOR REACTOR OPERATOR Page 18 I QUESTION: 025 (1.00) 

       ,

Which ONE of the following statements describes how plant operations are affected if Loop A RCS Wide Range Pressure instrument, PT-420, fails HIGH during Low Temperature Overpressure Mitigation System operation? I 

       '

a. Pressurizer PORV PCV-430 opens onl Pressurizer PORV PCV-430 opens and all pressurizer heaters deenergize onl , c. Pressurizer PORV PCV-430 opens and both pressurizer spray valves open onl Pressurizer PORV PCV-430 opens, all pressurizer heaters l deenergize, and both pressurizer spray valves open.

i QUESTION: 026 (1.00) 

-Given the following Unit 2 p1~ ant conditions:

! 

 - Unit is at 50% powe '
 - Rod control is in MANUA Loop B cold leg temperature detector TE-401B fails high.

, - No operator action is taken.

t l Mhich ONE of the following will be the steady-state pressurizer level? l % % c. 46% % ! t.

,

I i 

 .

! , 

   .-.

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

' SENIOR REACTOR OPERATOh Page 19 QUESTION: 027 (1.00) Which ONE of the'following conditions on LT-428 would result in an I increase in indicated pressurizer level? a. A leak in the reference leg of the pressurizer level transmitter, LT-42 Pressurizer liquid temperature increase c. The reference leg for LT-428 cools down due to a decrease in containment temperatur Containment pressure increases to 0.3 psig; containment temperature remains constan ! QUESTION: 028 (1.00) . [# gj Which ONE of the following is the reason for t second time delay associated with the AMSAC actuation relay? a. Permits operator override the event of spurious actuation, b. Prevents spurious tuation caused by S/G shrink and swel c. Allows ti for reactor power to decrease below 40% before actua llows time for the AFW pumps to recover S/G levels before actuatio QUESTION: 029 (1.00) Which ONE of the following RVLIS readings indicates the highest probability of core voiding? a. Wide Range reading 98 ft. with NO RCPs runnin b. Narrow Range reading 38 ft. with NO RCPs runnin c. Wide Range reading 65 ft. with ONE RCP runnin d. Narrow Range reading 95 ft. with BOTH RCPs runnin . 

      'l
      .

SENIOR. REACTOR OPERATOR Page 20 

      ;

l QUESTION: 030 (1.00) Given the following Unit 1 plant conditions * ' I 

- Unit is operating at 75% steady state powe All systems are in automatic contro The."A" S/G atmospheric valve fails ope Which ONE of the following-describes the plant response to this condition? (Assume'no operator action is taken.)

a. Turbine load decreases by 5%, reactor power remains stable at 75%. b. Turbine governor valves open in response to lower steam header-pressure to increase turbine load to 80%. c. Control rods insert to maintain reactor power at 75%. d. Control rods withdraw and raise reactor power to 80% where it stabilizes.

QUESTION: 031 (1.00) Which ONE of the following describes the normal, emergency, and alternate emergency sources of power to Safeguards 4.16 KV bus 1A06? Normal Emergency Alternate Emergency ______ _________ _________--________ a. Load Bus 1A04 EDG G03 EDG G04 b. Load Bus 1A04 EDG G01 EDG G02 c. Load Bus 1A03 EDG G01 EDG G04 d. Load Bus 1A03 EDG G03 EDG G02 

.
. - , - -  - - - , , , - , ,  - - _- .c -
      .% , rey

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

SENIOR REACTOR OPERATOR Page 21 i

QUESTION: 032 (1.00) a Which ONE of the following will provide an automatic shutdown of the gas turbine while operating in local control? ' a. Low axial compressor suction pressure b. Low lube oil level f Low lube oil pressure , d. Low control air supply pressure , QUESTION: 033 (1.00)

Which ONE of the following is the maximum rated run time for Emergency - Diesel Generator G01 loaded at 3050 KW? a. 30 minutes, j hour.

j c. 24 hours.

, hours.

l 

     !

. QUESTION: 034 (1.00) ' Which ONE of the following RMS displays must be selected to access alert and alarm setpoints? ) l ' a. Status Grid

b. Sector Display ! l c. HDSR d. Trend 

'
,

I J i 

     .

SENIOR REACTOR OPERATOR Page 22-QUESTION: 035 .(1.00) Given the following Unit 2 plant conditions: 

- RHR is in servic !
- RCS pressure is 320 psig and INCREASIN l
- RCS temperature is 340 degrees F and INCREASIN ALL system lineups are in a normal shutdown configuration for solid plant operatio Which ONE of the following will act FIRST to prevent overpressurizing the RHR System?

a. Pressurizer PORVs will ope b. RHR return isolation valve will auto close, c. RHR pump hot leg relief valve RH-PCV-861C will ope d. RHR pumps discharge relief valve RH-PCV-861A will ope QUESTION: 036 (1.00) ) -Which ONE of the following conditions will ARM the Turbine Crossover ; . Steam Dump System?  ! a. Selecting MANUAL on the mode selector switch, b. AUTO selected on the mode selector switch and a 10% lord drop in less than 120 second c. Going to TEST on the individual valves on the back of C0 d. Reaching 90 psig in the crossover steam heade ! 

.
    , __
  ,, . - , ,

     !

SENIOR REACTOR OPERATOR Page 23 QUESTION: 037 (1.00) Which ONE of the following is a symptom of a stuck control rod that would require entry into AOP-6B, " Stuck Rod Or Malfunctioning Position Indication", following a transient? a. Im individual RPI in 8 step disagreement with the bank demand locatio b. A variation in NIS instrumentation resulting in a quadrant tilt of 1.2%. c. A variation in core outlet thermocouples of 8% relative to i symmetric thermocouple ! d. A variation in axial flux of 1.2% of axial peak at any location relative to symmetrical trace.

QUESTION: 038 (1.00) Given the following Unit 1 plant conditions: 

- A loss of all AC power has occurre Power has been restore The crew has transitioned to ECA ECA 0.1 directs restoration of RCP seal coolin Which ONE of the following is the method required to restore RCP seal cooling?

a. Seal injection flow is initiated to cool the seals to less than 150 degrees F, then CCW flow to the RCP is establishe b. CCW flow is initiated to cool the seals to less than 150 degrees l F, then seal injection flow is establishe c. Seal injection flow is initiated to cool the seals to less than 190 degrees F, then CCW flow to the RCP is establishe d. CCW flow is initiated to cool the seals to less than 190 degrees F, then seal injection flow is establishe ! l l 

     !

I i l

i 

     !

i SENIOR REACTOR OPERATOR Page 24 QUESTION: 039 (1.00) Given the following Unit 1 plant conditions: 

- A plant trip has just occurre control rods are stuck out of the core following the tri An emergency boration has been initiated by the resctor operator in accordance with EOP-0.1, " Reactor Trip Response."

'Which ONE of the following lists the MINIMUM injected volume of boric acid necessary to satisfy the required amount of boration? a. 600 gallons ,200 gallons c. 2,400 gallons ,000 gallons J QUESTION: 040 (1.00) l Given the following Unit 1 plant conditions: 

- Steam generator A is faulted due to a feed line break outside of containmen ;
- The crew is performing actions of EOP-2, " Faulted Steam Generator '

Isolation."

- The AFW system is in operatio Which ONE of the following actions concerning the AFW pumps is required by EOP-2? i a. Shutdown the AFW Pumps immediately, b. Maintain at least 50 gpm AFW flow to each S/G with narrow range levels less than 8%. c. Run the AFW Pumps only if less than 200 gpm is available to the S/G i Isolate the AFW Pumps from S/G A (steam and AFW flow).

.. 

  -c,e. - - -yr

SENIOR REACTOR' OPERATOR Page 25 QUESTION: 041 (1.00) 'In accordance with EOP-1, " Loss of Reactor or Secondary Coolant", which ONE of.the following groups of parameters is required to be verified, in addition to Pressurizer level, prior to terminating SI flow? a. RCS subcooling,_ secondary heat sink, and containment pressure b. Secondary heat sink, RCS pressure, and RCS subcooling c. RCS subcooling, RCS pressure, and containment radiation levels d. Secondary heat sink, containment pressure, and RCS pressure QUESTION: 042 (1.00) Given the following Unit 1 plant conditions: 

'
- Unit is at 100% powe Condenser vacuum is decreasin Which.ONE of the following condenser vacuum conditions will FIRST result ;

in the' loss of condenser steam dump availability? j a. 25" Hg vacuum i " Hg vacuum c. 21" Hg vacuum I d. 19" Hg vacuum I l 

.-     l

_ _ , _ . _ m _ _ _ . _ . ___ . _ _ _ - _ _ _ _ . _ _ _ _ _ - ._ _- - i.

l SENIOR REACTOR OPERATOR Page 26

QUESTION: 043 (1.00) 

'

Given the following Unit 2 plant conditions: 

; - Condenser-vacuum is 25 inches Hg.

- Generator load is 250 MW.

I i Which ONE of the following is the MAXIMUM amount of time the turbine can

be operated (Operational Back Pressure region curve is' attached.)?

j a. Operation is prohibited.

$- b.'10 minutes, i c. 1 hour.

, d. Operation is unrestricte r ' l i 

<

j ' QUESTION: 044 (1.00) i Which ONE of the following concerns would an operator most likely be 

.

confronted with during a total loss of AC power in excess of 7 hours? l ' a. Loss of secondary heat sink condition 

'

b. An unmonitored release of radioactivity 

      !

li c. Loss of containment integrity i

d. A Steam Generator overpressurization condition i s ?- t , l 

-

l !

3 ,

, i l 

*

I i 

!

s j i 

!
   ..

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

POINT BEACH NUCLEAR. PLANT AOP-5A } ABNORMAL OPERATING PROCEDURES MAJOR i Revision 4 . LOSS OF CONDENSER-VACUUM August 21,1995 !. f . 

 % s FIGURE 1
         *

. ./

! i i RECOMMENDED OPERATIONAL REGIONS FOR NUCLEA a e WITH 2 DOUBLE FLOW LOW PRESSURE ENDS. , , , , U ;m.._.._.._.._..........__ .. n%::: 

                     ,
    :: m = ==n ; =- r: g ; . . n i n : n " ; a : ;: = :r _ . _ _ . . . . . _. . __ . . _. . . .. . . _ .. . . . .=n...n   un:n:==
                ---_ .=. . .: _~;;;.-
                  . . .:;._;4;;;g:::::,;.;;;nM;-*::::::;
                  .. _ . . . .    =;;nn
    =en      .n             . .
    :: = :== ::== =n :=:: r:=::=n: = = ==. . . =.  . = :- ; == =====::  ;;==:==:=n:n h' :a::n =;;n;;c;
              ====n   - = L n;;. =mgr :n:y;;-

n =. n..gn; : . . . ~. .._. _. .. .. . .N .:;L; .. 

    =_. =. . . . :. r.. : ; .
    . n=m-        n = r=. m. .=rr. ; a_=.
                    -

w!HE!.c - 

                     ...n= ::: = m un:.n.=4
                       =

n u n . r. . . . . . . _ . m. _=. . =. . . . n.==. . . 

           . . .. .. ..... =. ...m
            ; :=. . . .u. . . . .E..m
               . . . . . =.
                := =m
                 . ;,_=.
                  .

m: n= : =, --- - r .:.=. . . .=. ;.g;.:.-- ._._.=. ne: . .=. .. .n n : 

                    ..-..m....=...
     .
          .g  m;,!=. ;;;, g;c;
               .
                 .;,=

Fi!P651~r ~  ; 

                      . =i ai E g n; hie:EPf:E iE           g=li~i[

5 " ..,,, .1= :f:n: 

    :n::n:=: : :_= . . . . . . . . n.n H n .n :::=3  = m : . .iE_njE...H..i!!m...Eii!ii: J E.ii.H. P' .
                    .
                     . --
                     .

aE 

                     =.u . . m. F~rir .=:. .: = . .n a;;
    ::::= = ::; .
          . .:::.: n =::n:n::

r.; . . . . . .... mum:m:- t . . . . . . . 

                     .
    ::=;.:=r:====.._.................  =::nnum   :m:.. n.:_ =n =n= n n:n n::M:.?= m ;===:::n n:n:n:nin:::=!r-
           .
                      === = im : = r = =
               ::=.
    :

4.;**Z:: ;' U.,.: ;...: :....: _U.C_ 

       . . . . . : : .: . n : o r = n n : =. = n n:=:;;:  m= =n::. n ::.r == :n=;:p!!;rn. =. m: i::E E :n ni!i m m = = n.==m=: ; ~ -- u n::

L.; : . J.' . ". ': 

     .. . G.Jn ::r.n:n n;;; m. . , .: n :;.: " ' . ' y . '; ' . . '. .'. O_,* . ;.* ".:* .* ' .' _'. :; : "n -
        . . . .

_ . . _ :.: :::.. 

           . . : : . : n :L.": n:' : : ' :2.::: . , ..;.. .n..: .;;;",, . , f_; ,.::::
              : ::;..:. .::. :.2   ;;;;;; ;, g :;;;; . *
                    - ..- . ::. I..-:::.:'.:!n;:
                      -
     = : : =.Z ~=. -' '. :=n:n:n-.
             .
              ::;
                 , . . . ,   -

T.- . 

    $n :z: n! =
     ":   !! -" *,I n -         .--
          .n k:' . .n: r
                 <.-

3 ~ . - H: :: : ::- n;, ;: ? =+ n = I E =. . : . : m 2. n' ; mi -'- :' .-.: n - _J --.- n - ; . .- . . _ ._ 

     : ';; ;C'.nM:.~ ;;*
    '      *.      -

4 == :: =O!  : *.'n":'. ; ' :n ' :=:20: 1: l' . 

           '. n' M ;';-~ ;' ::;' n; . . :; a b ; u : n LM  :n:r-: % .: ; *:: :   :
             , .
                  : an!!!CEUij :m 0Z=: M ={l::::!!m mm*n= ." ::: = " -=- = x =-: -

g u n a :: n n : r_ ==n n: n : = =. r.'..u..=..u. =..y;::;::;C:n"";L ;;f;;;::n: :::;;;n;;n::;;; 

         '       :g:::
    :n.::: :::n: = n: m=m =r.-.   :::.3-
         .

m m m =n=:ne i=n -. . . ' : = : : ' 2 : " . . = n" :.'- mum " : n=::n:n  !!!!!:.=:.:!!iE-uniirEE!!.  : .: = :::= :==:=:=. 

                        :-
          .
                       -

J hhi!!! iik'3 

         ..:
           ==..:m := n :.cm.:n  n :: n =:;d
                .
                 =n
                   -
                    = = m = u:p==u  m=; ;.

O n :m ==f r b kr:= 53E=Nib h. .-. ' aiii7c =_---- m C= 7 ~ ' k=/Ii . 2u m i'I! ;'li 

         " m= := rt
          .'.I:[N!Ih!UiI
          ; ="m~ 2: - =   !.U:I=i'i"!EiiNiiiE!

u= = : en - m ; n h =*r : A[i;UUj{,IiE~'m['.= 

                 =". . - .. . . : m' ' M =m=    ~
                     ; n ": '  n n =bn
                     - 'Ei: i -b O

r. =-": -::n.=.,...=...=...r 

    . . _ . . . . .=. U    :
         ::m   ==:=~;--~~~~~        =   r n i =.

kNbkb.N lE2.)Eh @ 

      . .
        .=....:..n..~..-.=..l==.n.m_.:.m..t.=..,..=....:rc.'.t
           . .-         ----

n 4 J5 5 BEE. [555N 53!I'i;T;dt[i- 

              .
                        'g'.
       . .is[EEii. 7 ! ' ~T'r ~' r M u" 2 =:= = =iiiUil.Gii!{ {iiI ~- -*3 ' n g:;:"4. n ': n    . , . . ,;;
                       .- g-
                ..

3.- IE:fi!E  ::Ei2 ,g g  ::*=::r4 * *- - 

    .
                   - -

Ig'-. """ 0 c.=:= E 

       .
      - t .u:
       . n.:.:./ n= : . : ' ~ b : ::: : .      M2= =2mui!!hE E': mEE    E E ~.i" h;E M.E"
                    ; m n n2 :::M" :: =

I  :=:.:M::==. = :=_h.* - . z: = a=w ==. =_. n: m: m n=" 

                   -
    =r : = = ~    =nm:            m m m g. = : n= m =:- m n= ==r=

g *: ::::n;;::' 

    ,

12Mm . * ,, .:'. a O=. _-==.c ==: . . . . _ _ . 

       :'"::=:-~r=.r=: =r=t:.Z'J:::- :- .;J.;t-n
           = : :;;.= :.: _ . . . . _ . . . . _
                   : = ; m nw n u M 2 2: = := :== -

E **4 

    ;;r"':;*nr
    . ;   J ".': =::2/2 a : Mm * R
         :

L**,-'j:::*n ;- ::= 

            .
               ;; =.u =: ::M==i:imiin nm   tia 1_1:: u ;;;im.a=;;ii.:di =:E:E    =

y *:r . p. : -"- 

      . g"" = ! ::u ", = u n= r m r- r' . :.r n: n. :::::: 3- :: = ---*"m*:  Jtunn=
              : :;--':.:.=
               : n:nm;:r:::;g;;;;- u n:n;; ;m :gg n n-. rg   .
 =  g p
     :; .

_n _:.': _.. J : ~~" n "* 1  ::*- u ': m :?:- n --- ;:n **- :" .n- :: 

            ::2:::*=':-':=n;rm::;a;;:;;::m    M:; - ;;:,n,..;;;;:;:E
                   -   : : m ;.- ;,; r .;-. ---- -
                       ;;;Jn;,=;..
                       .

n g;g=.g M p= =:<=r ::.* = MuMun =.n ..= =:':': =-"::; { = : = :._: 

       ~ . .
           ::::  -

n g:;;n ;:; n :::-- . ~,, ~. ---  ; n ;;:;g;1 7:-; -;,f- ;,;;. ;; 7; nr;gr r ;. g; .;;- ;; =, 

..

kl e p:rM 

    .

M: :: 

     =. ;.:n n mM um ::  =n::==
         .
          .:.::n t.:In::; :n:M
        :: n jM *':: M:' = *. = ::.c.M;a u = :: = :; = m =n m m = t- ;:n:ra r    == y= : =_2: .
                =M:M=M=-;
                = ::= ==: ==E.;.: :==.=2  g :  w-===.--

9 .:.:n M::M gn:;- :;---&g- M ;:n :: r n,:: = 

     : . . : = :M = M =:=::=m=m n:=i   ::. m mm'.:::;r: . m :x: = n :: ::;:=%:.. nn a MU:C ;t n M .m := :u=::: =-  = ===_ == == u
  { ,::RM:n.=gC
    ..

M:M:**M:.  : C p: =::C::::L.T::::.L'n 

      :' : : : : :: . ::: n : ':--" ::: ;
          : ^ '

Z ^ =::' n:::M=_ . :'.y :=';-~ ' ,=-=: u u=:r = mi=!:==:==: 

                %   m::= n a = :w =- a ::::=d-= cm x

g 

    : == : M = : :
     :::::. :
      = = = : u u = =5 :. -- ;'.;L :' 21: g' :: C :::::;: ':;.:* :C W . . : ;:U n .:::::::. ::;;4 -.;:: -C n:mj:n::; :;;;.:t ; ;; a g r--
           -
             .
             .
                  : g..;:.;m :m : :;=---- .

m:= 

                      ; -- qC =

N : 2M' :::n:::=== u="M::m:=::: 22 ;;.;. . g_ =. ..=_.M.:n =.sM:=na.M::n = M=::n=M:::=m!=: n ::::::::::;::::::;:-- . :mn Ed 

           . . , . . . _

CD l em ==**"C'"" M'=-' :'r4.:::nC.R ~2---'n :::===n:M: L' -- ;" ':n=:M u r u n _..: =2; ;:m:::: =:=n::t 2::M..................................,.=.._..u=....=...2.==_=a= .,; 

         :
               =.= :.c = a: = r     . . _ . . . t==

u = =; c:m. : : ~~. = = '""' R = "=- .-^ m=n :' =Ly :;.;; n=M.=;; :;- I=; 

         .

Mm M% *Mn! M :=:=2: U;~,*,: MM:MMM;;;::~;n u. aW: '=.=,* . '.; :=: : :::':2 M:::::2'J " M = : : = :== ,.- - r == ;=nf : _=; -=n : =-- = u n:=M = 

            . '*M'".m;:u' n:'-=. .=:::  = n = = ::: n M M = n = :-

M:.' Mn' ~  ::... =__=_..._: t. M:::M=M: m =...u . 

       ::: = =;=4M= = m"'* u
        .
          *1n:::::n%"-' 'n M!. nt:==:~M::n n:M C :ZM r.; M.' 2 =::'::n::
                . . ' : .i=::: 44M: :n   ;;;;;;: U
                    ;;
                   * C;:::n ;;"; :=,**'T*1:, =f **l"
                      . :;C n J' ': M ::::* M N  jt:.M=:n:  MU=:*MM::M::M:*:M:M:=":M
    . . . . _ ~ . . .... M:::nn:M=:n;:      ::r' = = : f...=MM: mM::;;= = = = =
                 .
                       . .:U M*M:::U:;n::22:     Z * :-1:= =.un  MMC;;"':  .~="%:M:::=:::::=M::',--   = ' .n = = =n'"= a n ==:::=-

m=:=: u;n = 

   :M M::!M. l*M:: ':. : M"'M:M:F:Z:::n.:: M: .:Mn:M ;2'*! !M:::M:::U *n :;;;;n!!:U=j:::
         ;;"::6:n n'::f:::  V "-':* MM=*
             !::.

J :'mnM=:::::::::: nc* *- * . _ _ _ _ 

   , $555555*5!5-{:m::       ;.;;'MMMCM;;"Ma'.Z.:: lMnn:Mn :.:.10::!': '-M:nJ.::::;;:MM::nn::;;;;n::=;;;;M ::n;:n:=n :=_U a   M ::::::M::M  :M : : n ::M M M : : :: :: : M : :: = ::::::: n :: m M ;;;;.u n;m:::==M M M:M M MM;:nn         CJ==='; Mn"::M M : n M M

e M;;;:mn:;n;;;;;;;M;;.~.;. u M.=:;L;.:.a 0 10 0 200 300 i e t i i 400 500 600 70c GROSS LOAD-MW 393A421 

 .                T
          .pngcT6i7 REFERENCE USE
       .    .
SENIOR REACTOR OPERATO 'Page 27 QUESTION: 045 (1.00)

Given the following Unit 1 plant conditions: 

- A loss of all AC power has occurre ECA-0.0, " Loss of All AC Power" is in effec Per ECA-0.0, certain Engineered Safeguards equipment control switches are placed in Pull-ou 'Which ONE of the following events is prevented by this switch alignment?

a. An uncontrolled cooldown of the RCS and possible reactor restar b. An uncontrolled use of water that may be needed for long term cooldow c. Ar uncontrolled start of large loads on safeguards AC buse d. An uncontrolled depressurization of the RCS.

-QUESTION: 046 .(1.00) Which ONE of the following describes the response of Feedwater Regulating Valves following loss of a RED or BLUE instrument bus? a. May fail open; manual control will NOT be available to regain contro b..May fail open; but manual control will be available to regain contro c. May fail closed; manual control will NOT be available to regain contro May fail closed; but manual control will be available to regain contro .

SENIOR REACTOR OPERATOR Page 28 QUESTION: 047 (1.00) Which ONE of the following locations / equipment is protected by a Halon System? a. Diesel Generator rooms b. Hydrogen Seal Oil package c. Service Water Pump area d. Auxiliary Feed Pump room QUESTION: 048 (1.00) Technical Specifications require containment pressure to be between + psig and -2.0 psig during power operatio Which'ONE of the following is the MAXIMUM time containment pressure is allowed to be outside this, band per Technical Specifications before action must be initiated to shut down the plant? a. 30 minutes hour hours d. 24 hours 

.

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

SENIOR REACTOR OPERATOR Page 29 QUESTION: 049 (1.00) A LOCA has occurred and both RCPs have been started per CSP-C.1, 

" Response to Inadequate Core Cooling."

Given the below list of criteria: I. Core cooling provided by low or high head SI II. Narrow range reactor-vessel level greater than 25 feet III. Any RCS hot leg loop less then 350 degrees F IV. Core exit thermocouples less than 1200 degrees F Which ONE of the following gives the three conditions from the above list that allow securing of the RCPs per CSP-C.17 I, II and III I, II and IV ' I, III and I II, III, and I QUESTION: 050 (1.00) Given the following Unit 1 plant conditions: 

 - The Reactor tripped from 100% powe EOP-0, " Reactor Trip or Safety Injection" has been entere The main turbine did NOT trip as expected.

l - MANUAL turbine trip is unsuccessful.

! Which ONE of the following is the NEXT action required per EOP-0? a. Manually close MSIVs and Bypass valve b. Locally trip the main turbin c. Go to CSP-S.1, " Response to Nuclear Power Generation /ATWS."

d. Verify Safeguards Buses energized.

l l

i 

  . -   . . -

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

        . . . _ . . . . _ _ _

} i j SENIOR REACTOR OPERATOR Page 30

! !

QUESTION
051 (1.00)

. ' Which ONE of the following is the reason for the main feedwater i: isolation following a reactor' trip? a. To prevent thermal shocking of the steam generator (S/G)

tubesheet ~

j b. To prevent excessive cooldown of the RC c. To. prevent S/G overpressurization.

d. To preserve secondary water for a subsequent RCS cooldown.

. i I I i QUESTION: 052 (1.00) l Given the following Unit 1 plant conditions: l 

 - A reactor trip and safety injection has occurre l
 - Pressurizer ~PORV 1RC-431C lifted and is stuck ope l l

Which ONE of the following is the MAXIMUM pressure below which the PORV i isolation valve must be shut?  ! a. 2450 psig b. 2335 psig c. 2225 psig d. 2010 psig

I 

         .
 @

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

I SENZOR REACTOR OPERATOR Page 31 QUESTION: 053 (1.00) Given the following Unit 1 plant conditions: 

- Unit has tripped'from 100% due to a small break LOC Conditions have stabilized and operators are evaluating the criteria for terminating S Adverse Containment conditions do NOT exis A maximum of 50 gpm feedwater flow is available to each S/ Which ONE of the following conditions would PREVENT SI termination per EOP-1.2, "Small Break LOCA Cooldown and Depressurization"?

a. Both steam generator levels indicate 5% N b. Pressurizer level indicates 12%. c. RCS subcooling is 40 degrees d. Pressurizer pressure is 2050 psig.

QUESTION: 054 (1.00) In procedure OP-5A, " Reactor Coolant Volume Control" there is a PRECAUTION that states: "Do not secure letdown flow without also securing charging flow ..." Which ONE of the felic ng statements describes why charging flow should  ! also be isolated? (Assume all systems are in a normal at power lineup.)

a. VCT level will decrease until charging pump suction shifts to . i the RWS ;

b. Reduce thermal shock on the Non-Regenerative Heat Exchange l i 

       '

c. VCT level will decrease causing possible damage to the charging pump d. Reduce thermal shock on the charging penetration into the RC , 

.
  .  . .w =+

SENIOR REACTOR OPERATOR Page 32 QUESTION: 055 (1.00) Given the following Unit 1 plant conditions: 

- Unit is in Cold Shutdown with RHR Cooling in progres The RCS is soli RHR flow is lost and CANNOT be restore '
- All other systems and components are availabl Which ONE of the following methods of cooling will be utilized to remove the core decay heat?

a. Feed a S/G using an AFW pump, and bleed steam through the respective S/G atmospheric relief valv b. Start a charging pump, with flow through an RHR heat exchanger, and initiate Hot Leg Injectio c. Feed the RCS with Safety Injection, use letdown to remove decay hea d. Start a charging pump,.with flow through an RHR heat exchanger, and initiate Cold Leg Injection.

QUESTION: 056 (1.00) l Given the following Unit 1 plant conditions: 

- A reactor trip occurred 30 minutes ago due to a Steam Generator tube ruptur The crew is taking the actions of EOP-3, " Steam Generator Tube Rupture".

- N-35 indicates 1.4E-10 amps and N-36 indicates 1.6E-10 amp Source Range Instruments are NOT energize Which ONE of the following actions is required by the Reactor Operator? ! a. Allow N36 to decay to less than 1.5E-10 amps which will automatically energize the SR instrument b. Depress both SR RESET pushbutton c. Depress both IR BLOCK pushbutton d. Deenergize N3 .. 

   . , .

 , . .- . .  .

SENIOR REACTOR OPERATOR Page 33 ., QUESTION: 057 (1.00) ' Which ONE of the following conditions requires tripping the reactor per AOP-1A, " Reactor Coolant Leak", for a confirmed steam generator tube i ' leak? Assume appropriate actions have been taken per AOP-1A regarding charging and letdown system operation _ 4 a. Pressurizer level is at 11% and decreasin l . b. Narrow range S/G 1evel is 62% and increasin c. VCT level is at 17% in AUTO control and decreasing.

j d. S/G Blowdown Isolation valve closes due to high radiation.

QUESTION: 058 (1.00) The crew is responding to a ruptured tube in 1B steam generator (S/G) using EOP-3, " Steam Generator Tube Rupture."

Given the following plant conditions: 

- 1A S/G pressure is 950 psi B S/G pressure is 1050 psi Which ONE of the following is the required core exit temperature that the RCS must be cooled down to prior to depressurization? (Page 15 of EOP-3 attached).

a. 505 degrees F b. 490 degrees F j c. 485 degrees F , d. 480 degrees F , i . j 

 -      i
.       l
      \
' POINT BEACH NUCLEAICPLANT
 "-EMERGENCY OPERATING PROCEDURES EOP-3 Unit 1 MAJOR Revision.1 ~

STEAM GENERATOR TUBE * RUPTURE- j /, October;26, 199 UNIT-1 , STEP ACTION / EXPECTED RESPONSE - RESPONSE NOT OBTAINED CAUTION: 

(Step) THE FOLLOWING STEP MAY THAN 100 F IN ONE HOUR. REQUIRE A COOLDOWN NOTE:
(Procedure) controlled cooldown has been initiatedRCP trip criteria
    .

Initiate RCS Cooldown Using ruptured steam generator pressure, detennine target 2 core ~ exit temperatur * __ _ _ _ _ _ Ruptured Steam Generator Pressure (psig) Core Exit Temperature (*F) 1100 1000 (460) 505 900 (445] 490 ~gg 800 (435] 480 700 (420] 465 600 [405] 450 500 (390) 435 400 (370) 415 340 (350) 395 (335) 380 , 

      '
  .

l 

      ,

l i I 

  . .
  . Step 20 Continued on Next Page . . ,

l _ O 

   'p=-c >,oc a
    <x
    ,_
  >C- 's = e c2_
   -

_ , _ 

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

l1 SENIOR REACTOR OPERATOR Page 34 j l i

l QUESTION: 059 (1.00) Which ONE of the following statements explains why AFW flowrate is procedurally restricted to 100 gpm when recovering steam generator (S/G) ! level if the level has fallen below 55 inches on the wide range q indication? l , a. To minimize water hammer to the S/G feed rin b. To prevent reactor restart from an excessive cooldow c. To minimize thermal stresses to S/G. components.

d. To prevent exceeding reactor vessel cooldown rate limit.

d QUESTION: 060 (1.00) Which ONE of the following is the basis for stopping all ' cps in CSP-H.1, " Response to Loss of Secondary Heat Sink"? It establishes natural circulation to enhance the bleed and feed ) capability of safety injectio I i b. It extends the time available to restore feed flow before bleed and feed criteria is me I It anticipates an RCS pressure decrease caused by spray valves !' opening when air is restored to containmen It anticipates an RCS pressure decrease caused by opening PZR PORVs during bleed and fee . 

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

i , SENIOR REACTOR OPERATOR Page 35

QUESTION: 061 (1.00)

I Loss of which ONE of the following distribution panels will result in a i 

<

DUAL PLANT TRIP? a. D13 l _ , D16 . c. D18 d. D21 s a QUESTION: 062 (1.00) 

,

With the Pressurizer Level Control Selector Switch in the NORMAL position, a pressurizer level instrument failure causes the following l SEQUENTIAL plant event Charging flow is reduced to minimu Pressurizer level decrease Letdown flow is secured and heaters turn off.

, 

- Pressurizer level increases until high level trip occurs.

, Which ONE of the following instrument failures could have occurred? 

(Assume NO operator action) Pressurizer level channel 428 (Blue) failed low Pressurizer level channel 428 (Blue) failed high Pressurizer level channel 427 (White) failed lo I Pressurizer level channel 427 (White) failed hig ..
  - - - .

. _  _ ._._.. _ . . .__ _. . . - - - _ _ _ . . . . _ _
      ]
      !

SENIOR REACTOR OPERATOR Page 36 QUESTION: 063 (1.00) Given the following Unit 2 plant conditions: 

- The plant has tripped from 100% power due to a loss of off-site powe The crew is required to verify natural circulation ~^in EOP-0.1,
 " Reactor Trip Response."

Which ONE of the following parameters satisfies one of the criteria for i 

      '

indication of natural circulation? a. Steam generator pressures slowly trending upward.

l b. RCS Cold leg temperatures are increasing, c. RCS subcooling is 36 degrees F.

. d. Core exit thermocouples at saturation temperature for steam ! generator pressure.

I ! l QUESTION: 064 (1.00) Which ONE of the following " Procedure Usage Levels" allows performance og all activities from memory? a. Infrequent Use Information Use c. Reference Use d. Continuous Use l l l . 

*
?

J

SENIOR REACTOR OPERATOR Page 37 QUESTION: 065 (1.00) The applicability statement for Procedure EOP-0, " Reactor Trip Or Safety Injection", indicates that this procedure is used for initiating events occurring where RCS hot leg temperature is greater than or equal to 350 degrees Which ONE of the following describes the applicability of EOP-0 when RCS temperature is below 350 degrees F and a LOCA occurs? a. EOP-0 cannot be used, procedure EOP-0.0, "Rediagnosis" is to be use b. EOP-0 cannot be used unless directed by the Critical Safety Function Status Tree c. EOP-0 cannot be used unless a step-by-step evaluation is made to determine if each action is still applicabl EOP-0 cannot be used, Shutdown Emergency Procedures (SEPs) are to be used.

QUESTION: 066 (1.00) Which ONE of the following is the MINIMUM number of shifts, per 10 CFR 55, on which you must actively perform operator or senior operator duties to maintain your license in an ACTIVE status? (ASSUME 8 hour shifts.)

a. 5 shifts per calendar quarter b. 5 shifts per calendar year c. 7 shifts per calendar quarter shifts per calendar year 

.

SENIOR REACTOR OPERATOR Page 38 QUESTION: 067 (1.00) i 

     !

Which ONE of the following is the LATEST time that alcohol consumption 1 would be allowed prior to assuming the watch at 1800 per NP-1.7.5, l 

" Fitness for Duty Policy and Procedure"?

I a. 0600 b. 1000 c. 1300 d. 1500 l

QUESTION: 068 (1.00) j

A bomb threat has been received and the Duty Shift Superintendent (DSS) has initiated a plant evacuation. A suspicious device has been discovered in an emergency diesel generator (D/G) room by the extra reactor operato Which ONE of the following states the actions to be taken by the individual discovering the device, in accordance with NP 1.7.10? a. Immediately notify the DSS by two-way radio and then leave the D/G roo Immediately notify DCS by two-way radio and then leave the D/G roo c. Leave the D/G room, and then notify the DSS by phon d. Leave the D/G room, and then notify DCS by phone, t 

     .
    .

l l l

SENIOR REACTOR OPERATOR Page 39 QUESTION: 069 (1.00) Which ONE of the following is the preferred method to rapidly reduce turbine load in response to a casualty per the Managers Expectations? a. Operator Automatic Load Rate Contro b. Turbine manual using governor valve decrease, c. Valve Position Limite d. Turbine manual using Governor fast.

QUESTION: 070 (1.00) i Which ONE of the following is an individual who is authorized to be a member of the fire brigade and to perform operations health physics functions? a. Unit 1 Control Operator b. Unit 2 Control Operator c. Duty Technical Advisor Extra Operator QUESTION: 071 (1.00) Which ONE of the following describes the process of determining an instrument's accuracy by comparing the indication to other independent instrument channels measuring the same paramete a. Channel Calibration b. Channel Functional Test c. Channel Check d. Channel Verification 

     !
     !
     !

i i 

     '

SENIOR REACTOR OPERATOR Page 40 l QUESTION: 072 (1.00) I An oncoming Reactor Operator has worked the following schedule: Day: 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Hrs: 12 8 12 8 8 8 8 12 8 12 12 12 _12 ? Which ONE of the following is the MAXIMUM number of hours the individual may work on day 14 without obtaining special authorization assuming the operator has a minimum of 8 hours off between each shift? a. 2 hours hours hours hours j 

     !

I QUESTION: 073 (1.00) Which ONE of the following is correct regarding INDEPENDENT verification of control board sliders? l a. Must be done by the Unit Control Operato ' b. Must be done by visually observing the operating lin c. Can be accomplished using status light d. Can be accomplished using remote indicator , p 

  .

__. , h i I l SENIOR REACTOR OPERATOR Page 41 ! l QUESTION: 074 (1.00) 

'Which ONE of the following conditions is required in order to perform j

. work on a system using " Positive Control" in lieu of danger tags?

a. A QC certified inspector is presen l l b. Work is1to be performed in a high radiation are l c. Requires direct line of sight of the isolatio l 

     ;

d. Isolation boundaries are locke j 

     !
   !  .

QUESTION: 075 (1.00) 

     ;

Given the following Unit 1 conditions: 

- Unit is at 100% powe l
- VCT level is 23%.
- All controls are in automati LT-112, VCT level transmitter, fails hig Which ONE of the following describes the FINAL ACTUAL VCT level? (Assume no. operator action.)

l a. Increases to 78% and then diverts to the HU I Increases to 28% where auto-makeup stops, c. Decreases to 17% where auto-makeup initiates, Decreases to 0% (empty).

. 

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

l SENIOR REACTOR OPERATOR Page 42 

     ,

QUESTION: 076 (1.00) Given the following plant conditions: 

- Both units are operatin The South condensate Storage Tank (CST) is isolated for repair The North CST is selected for Auxiliary Feedwater Pump suction at both unit Which ONE of the following is the MINIMUM volume required in the North i CST?

a. 13,000 gallons ,000 gallons c. 26,000 gallons ,000 gallons , QUESTION: 077 (1.00)

i Which ONE of the following is the reason the Containment Purge Air i Supply and Exhaust Valves are required to be locked closed during l operations at power? I a. The valves are NOT seismically qualified to operate during a , design basis earthquake.

. b. The valve actuators do NOT have class 1E penetration conductor

overcurrent protection devices.

t ' c. The valves capability to close during a design basis loss-of-coolant accident has NOT been demonstrated, d. The related piping systems outside containment are NOT seismically qualified.

) i 

.

     ..-. - . . .
      !

P SENIOR REACTOR OPERATOR Page 43 QUESTION: 078 (1.00) Given the following Unit 2 plant conditions: . 

- Operating at 100% powe All systems are operabl l
- While in AUTO rod control, Control Bank "D" starts ~ stepping in  !

slowly, but at a noticeable rate.

-Which ONE of the following events will cause this response? a. A tube leak in the Regenerative Heat Exchange b. A tube leak in the Seal Water Heat Exchange c. A tube leak in the Non-Regenerative Heat Exchange d. A tube leak in the Excess Letdown Heat Exchanger.

QUESTION: 079 (1.00) Which ONE of the following prevents inadvertently raising an irradiated fuel assembly in the New Fuel Elevator? a. Manipulator crane is interlocked to prevent loading irradiated fuel assembly into the elevato i b. The elevator will not rise if any Spent Fuel Pool area radiation monitor alarm c. Operators verify the elevator basket is empty before raising the elevato l d. The elevator will not rise if any fuel element is in i .. 

  , , - -

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

SENIOR REACTOR OPERATOR Page 44 i l QUESTION: 080 (1.00)' 

     ,

l Given the following Unit 2 plant conditions- l 

- Reactor power is 1215 MWt
- RCS pressure is in the normal operating ban Which ONE of the following RCS average temperatures would FIRST result in exceeding a safety limit? (Tech Spec Figure 15.2.1-2 attached).

a. 590 degrees F b. 600 degrees F c. 610 degrees F d. 620 degrees F QUESTION: 081 (1.00) Which ONE of the following states the MINIMUM number of operable 4 containment fan cooler units required to ensure containment integrity ! following a Limiting Design Basis accident if ONE Containment Spray Pump is inoperable? d. 4 l 

@

_ _ _ . . _ . . . __ _ . . . . _ _ . _. _ _ _ _ _ _ _ . _ _ _ . _ _ . _ . . _ _ _ _ _ . . . _ . i Figure 15.2.1-2 

'

REACTOR CORE SAFETY LIMITS 

- . .

POINT BEACH UNIT 2 s 

:   670

- 660 - .

650 l 2425 psia

!

MO -

2250 psia . 630 - < Q h

' 3 620 - u 

 ,

y 2000 psia ! i b610 o 

  -

i j 600 - 1775 psia j 590 -

580 -

, i i

570 -

? . !, ' ' ' ' ' ' ' ' ' 560 ' ' ' ' 

.

0 0.1 0.2 0.3 0.4 .6 .8 .1 1.2 

.

Core Power (fraction of 1518.5 MWt) 

; ~

nG-l j Unit 2 - Amendment No. 169 - November 17, 1995 a 

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

SENIOR REACTOR OPERATOR Page 45 t QUESTION: 082 (1.00) Given the following Unit 2 plant conditions: 

- Unit is at 50% powe Control rods are in AUTOMATI Which ONE of the following instrument malfunctions would result in a CONTINUOUS rod withdrawal?

a. Loop "A" Thot RTD fails LO . Power range channel N-42 fails LO c. PT-485 fails HIG i 

     '

d. Loop "B" Tcold RTD fails HIGH.

> l

QUESTION: 083 (1.00) Given the following Unit 1 plant conditions: 

- A Control Bank D rod was dropped and recovere The Pulse to Analog Converter was NOT reset as required by
AOP-6A, Dropped Ro l I

Which ONE of the following will occur on the next rod movement? a. If control rods are inserted, the Rod Insertion Limit Alarm will be received at a lower rod position than required, b. If control rods are withdrawn, Overtemperature Delta T will NOT stop Control Bank D withdrawal when required.

c. If control rods are inserted, Bank C control rods will begin insertion at a lower value of Control Bank D positio If control rods are inserted, Bank C control rods will begin insertion at a higher value of Control Bank D positio .

4

i SENIOR REACTOR OPERATOR Page 46 QUESTION: 084 (1.00) EOP-1, " Loss of Reactor or Secondary Coolant", Step 1: " Check if RCPs should be tripped", is a continuous action ste Which ONE of the following is the basis for continuously monitoring for the criteria to perform this step in response to a LOCA? a. To minimize RCP run time with less than the required amount of subcoolin b. To prevent RCP damage from cavitation due to operation in a two phase syste c. To minimize cooldown rate if a main steam line break is in progres d. To prevent excessive RCS inventory loss and a potential inadequate core cooling condition.

QUESTION: 085 (1.00) Which ONE of the following is the MAXIMUM allowable time to actuate AFW following an ATWS with a loss of feedwater? a. 30 seconds seconds c. 120 seconds d. 180 seconds

SENIOR REACTOR OPERATOR Page 47 QUESTION: 086 (1.00) Which ONE of the_following Control Room personnel should proceed to the scene of a fire? a., Duty Shift Superintenden b. Duty Operating Supervisor, c. Duty and Call Superintenden d. Duty Technical Adviso , QUESTION: 087 (1.00) Which ONE of the following is the proper sequence of major actions for removing decay heat from the core per CSP-C.1, " Response to Inadequate Core Cooling"? a. Reinitiation of high pressure safety injection; open PORVs; RCP restart; rapid secondary depressurization b. Rapid secondary depressurization; reinitiation of high pressure safety injection; open PORVs' RCP restart c. RCP restart; reinitiation of high pressure safety injection; rapid secondary depressurization; open PORVs d. Reinitiation of high pressure safety injection; rapid secondary 

'depressurization; RCP restart; open PORVs _ . . . . _ . _ _ _ _ _ . _ _ . _ . . _ . - _ _ _. . _ _ m._ .._.m - _ . - . . , _ _ . _ _ _ _ . . _ _

i

i Page 48 ] SENIOR REACTOR OPERATOR 

        -

J 

' QUESTION: 088 (1.00)

j Which ONE of the following is the reason'that Technical Specifications ' - ' requires Tavg to be decreased below 500 degrees F after the reactor is - chut down for excessive Reactor Coolant Activity? a.' Reduce the severity of a possible pressurized thermal shock i condition by limiting the amount of cooldown that can occur.

i

b. Minimize potential for containment contamination from
inadvertent PORV operation.

. l c. Prevent uncontrolled release of radioactivity if a steam i generator tube ruptures.

i ! d. Ensure reactor stays shut down to minimize release of additional i ' fission products into the coolant.

. i ! QUESTION: 089 (1.00) l The following boration flowpaths are listed in AOP-6E, " Alternate > Boration/ Loss of Shutdown Margin."

CVCS and RWS I CVCS and Blende II CVCS and CV-350 (Emergency Boration Valve).

Which ONE of the following sequences is the AOP-6E order of preference for the above flowpaths (first choice, second choice, etc.) to establish alternate boration when required? II, I, III 

        ! II, III, I III, I, II
. III, II, I

___ . - . _ . _ . _ . . _ . _ . . - . .._-.--.___.._.______-,_..._.___ . _ _ _ . _ . _ SENIOR REACTOR OPERATOR Page 49 QUESTION: 090 (1.00) Given the following containment history with a small LOCA in progress: Tima Cnmt Tem Cnmt Press Cnmt Humidity Cnmt Radiation 0815 '178.Deg. F 

 '

2 psig 90% 9.0 x 10E2 R/Hr 0830 180 Deg. F 4 psig .100% 7.3 x 10E3 R/Hr 1 0845 183 Deg. F 6 psig 100% 9.5 x 10E4 R/Hr 0900 185 Deg. F 8 psig 100% 2.0 x 10E5 R/Hr l ' Which ONE of the following describes the EARLIEST time at which adverse containment existed? i a. 0815 1 

 -        - ) i '

QUESTION: 091 (1.00) While discharging a Gas Decay Tank (GDT), operators are cautioned to 

" Maintain gas decay tank pressure at or above 5 psig at all times."

i Which ONE of the following is the reason for this caution? I a. Prevent condensation in the GDT.

! b. Prevent exceeding the required discharge flow rat , Prevent oxygen inleakag \ , d. Prevent overheating Waste Gas Compressors.

! ! ( l l I e i ) i 

-      -

SENIOR REACTOR OPERATOR Page 50 QUESTION: 092 (1.00) Which ONE of the following AREA radiation monitors has a control function? a. RE-101, Control Room Monitor 

  '*
  . RE-105, SFP Low Range Monitor
      :
      '

c. RE-108, Drumming Area Monitor d. RE-114, CVCS Holdup Tank Monitor 

  .

QUESTION: 093 (1.00) Which ONE of-the_following is the reason that EOP-3.1, " Post-SGTR . 

      '

Cooldown Using Feedwater" is the preferred procedure for Post-SGTR Cooldown? a. Radiological releases and contamination of secondary systems are minimize b. Boron dilution of the RCS is minimized, c. Adverse secondary system chemistry conditions are eliminate d. RCS cooldown rate is maximize QUESTION: 094 (1.00) During a core shuffle, which ONE of the following is correct regarding the MAXIMUM number of RCCAs that can be removed from the core each refueling and st111' ensure 5% shutdown margin is maintained? c. The number is determined by a Reactor Engineering Calculation d-. The number is determined by the Core Loading Supervisor _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . . . - - .- - . - __

' SENIOR REACTOR OPERATOR Page 51 QUESTION: 095 (1.00) 

     ;

During which ONE of the following conditions is EOP-0.0, "Rediagnosis", I 

. approved for use based on operator judgement?   l a. After transition to EOP-3, " Steam Generator Tube Rupture."

, b. During the performance of EOP-0.2, " Natural Circulation Cooldown," due to loss of off-site power when a twenty (20) gpm Steam Generator tube leak is detecte i

c. During the performance of EOP-1, " Loss of Reactor Or Secondary Coolant," when a RED path is detected in Heat Sin d. After transition to EOP-0.1, " Reactor Trip Recovery", following an inadvertent reactor tri l

1 QUESTION: 096 (1.00) Given the following Unit 1 plant conditions: 

- An event occurred at 0800 that was classified as a Site Area Emergenc The plant evacuation alarm was actuated at the time of the classificatio 'Which ONE of the following is the MAXIMUM time by which accountability ;

must be completed?  ! a. 0815 d. 0900 i

_ . . _ . . _ m _ .. _ m. _ _ _ _. _ _ _ _ . _ ___ _ _ .m. _ _ _ _ _ _ _ _ _ . _ _ _ _ ._ _ _ . ! j- SENIOR REACTOR' OPERATOR Page 52 i-i! !

QUESTION: 097 (1.00) 

'Given the following Unit 1 plant conditions:

3 - A reactor trip from full power occurred 60 minutes'ago.

i - RCS' pressure is 620 psig. 

 - Both-loop cold leg temperatures are 295 degrees CSF Status Tree ST-4 is provided.

! Which ONE of the following states the-appropriate color code for the J Integrity CSF? i-

a.-Green i

b. Yellow c. Orange d. Red QUESTION: 098 (1.00) The two emergency planning zones are the at a 10 mile radius and the at a 50 mile radius from the plan i Which ONE of the following completes the above statement? a. Low Population Zone, Exclusion Area Boundary b. Exclusion Area Boundary, Low Population Zone c. Ingestion Exposure Pathway, Plume Exposure Pathway l d. Plume Exposure Pathway, Ingestion Exposure Pathway , 

        !

l 

        ;

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

ST-4 INTEGRITY gey;3o$ 

       $

02-24-92 

, .

CSP- TEMPERATURES IN BOTH NO mmma GO TO CSP , I 

 - COLD LEGS GREATER THAN  E    l I

!

285'F YES l TEMPERATURES NO GOTO IN BOTH 

   { COLD LEGS
     -   CSP P.2 i

GREATERTHAN 315 F ' YES TEMPERATURES NO l IN BOTH l 

     -

COLD LEGS ' GREATERTHAN 345 F YES TEMPERATURE '9 DECREASEIN BOTH COLD 

+- LEGS LESS NO       CSF SAT THAN 100*F IN THE LAST YES       GOTO 60 MINUTES       CSP PA TEMPERATURES NO IN BOTH F--- COLD LEGS GREATERTHAN 315 F  YES RCS PRESSURE
   -

LESS THAN GO TO 425 PSIG CSP- YES I 

 ---
       , CSF SAT NO !

TEMPERATURES 

 -

IN BOTH COLD LEGS GREATER THAN 360 F YES I CSF l SAT l l l , 

   . - . _ . .

SENIOR REACTOR OPERATOR Page 53 . l QUESTION: 099 (1.00) While conducting a plant tour inside the Radiation Control Area you see , a velve tagged with a tag similar to the one shown on the attached pag Which one of the following conditions is identified by this tag? 

,

l , a. The valve is out of its normal lineup positio b. The valve is a boundary valve for a system containing caustic flui c. The valve is to be radiographe ! d. The valve has a contact dose rate of 325 mrem /h i QUESTION: 100 (1.00) Which ONE of the following is the MAXIMUM annual TEDE dose limit an individual can receive from Planned Special Exposure? a. 2 rem 

, b. 5 rem

c. 10 rem d. 25 rem , - 

 (********** END OF EXAMINATION **********)

,____ - _ _ . _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _ __ hy Fh i l ! ! I I I I ! 

     . - - . .
,.    /   N I    POINT BEACH NUCLEAR PLANT I

l i IDENTIMCATION # t ! ! " i 

       !

l l l

i i ? l.

I a PBF - 4081A Revision 0 08/92 , i + ! 

!

j 

!

i._ _ ._ _ _ _

_ . __ __ _ . _ _ _ . . _ _ _ . __ SENIOR REACTOR OPERATOR Page 54 I l l ANSWER: 001 (1.00) l 1 l l l REFERENCE: 

,

TRHB 10.1, Re , pg. 15 and C1805 COT and TRHB Figure 10.1.24 KA 022000G007 [3.3/3.5] i 1 

      '

022000G007 ..(KA's)

ANSWER: 002 (1.00) REFERENCE: EOP-0, Rev. 20. Appendix A, pg. KA 000040A110 (4.1/4.1) 000040A110 ..(KA's) ANSWER: 003 (1.00) REFERENCE: TRHB 13.7, Re , pg. KA 059000A302 [2.9/3.1) 059000A302 ..(KA's) 

 . .__ _ _ . _ _ _ __ . . _ . . _ . . _ . . . SENIOR REACTOR OPERATOR    Page 55 ANSWER: 004 (1.00) REFERENCE,: .

Logic Sheet 17 and Setpoint Document Section KA 041020K105 [3.5/3.6] , I 

     .

041020K105 ..tKA's) 

.

ANSWER: 005 (1.00) l REFERENCE: AOP-10A, Rev. 18, pg.. 1 [Similar to question used on 94 RO exam.] KA 000068A101 [4.3/4.5] 000068A101 ..(KA's) ANSWER: 006 (1.00) d.

l REFERENCE: TRHB 10.15, Re , pg. 7 and C4313AOT and P&ID Sb9ets 541F091 Sheet 1 I and 9645971 Sheet 1A KA 068000K107 [2.7/2.9) 068000K107 ..(KA's) I ! l l 

  . .~ _ . . -

i

SENIOR REACTOR OPERATOR Page 56 ,

' l

, ANSWER: 007 (1.00)

l
'

REFERENCE: J t -

AOP-9B, Re , p . 

;     l

' KA 000026G010 [3.6/3.5] l 

[Similar to question on 95 SRO exam.]   '

000026G010 ..(KA's) , i

$ AMSMER: 008 (1.00) l

- 

,

REFERENCE: 1 TRHB 10.2, Re , pg. 1 l W KA 003000K407 (3.2/3.4] l 003000K407 ..(KA's) 

    ,

ANSWER: 009 (1.00) ] REFERENCE: 

    !

TRHB 10.6, Re , pg. 71 and Master Data Book for B0 KA 004000K203 [3.3/3.5) 004000K203 ..(KA's)

SENIOR REACTOR OPERATOR Page 57 ANSWER: 010 (1.00) b.

REFERENCE: 

'.:
-

TRHB 10.16, Rev. 3, pg. 12 and C7304 COT and Logic Sheet KA 013000A302 -[4.1/4.21 013000A302 ..(KA's) 

.

ANSWER: 011 (1.00) b.

REFERENCE: TRHB 13.4, Re , pg. 18 and C730 COT and Logic Sheet KA 013000A402 [4.3/4.4] 013000A402 ..(KA's) ANSWER: 012 (1.00) d.

REFERENCE: TRHB 13.1, Re , Figure 13.1.19 and C1315 CO KA 015000A202 [3.1/3.5) 015000A202 ..(KA's)

i 

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

' SENIOR REACTOR OPERATOR Page 58 i l ANSWER: 013 (1.00) b.

l REFERENCE: 

,

TRHB 13.1, Re , pg. 11 and C1310 CO KA 015000K601 (2.9/3.2) 015000K601 ..(KA's) i l 

-

ANSWER: 014 (1.00) i ! l ' REFERENCE: l TRHB 13.8, Re , pg. 6 and C0105 COT; and TRHB 13.8, pg. 19 and Logic Sheet 1 i ! KA 015000K302 (3.3/3.51 015000K302 ..(KA's) i ANSWER: 015 (1.00) REFERENCE: Steam Tables

KA 017020K502 (3.7/4.0] 017020K502 ..(KA's) r ! 

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

i

SENIOR REACTOR OPERATOR Page 59

' s l ANSWER: 016 (1.00) ' 

. REFERENCE:
  '
  :

TRHB 13.2, Re ., pg. 8 and TRHB Figure 10. KA 017020K102 [3.3/3.5] 017020K102 ..(KA's) 

. ANSWER:   017 (1.00)      l )

i l REFERENCE: BG EOP-0.2, Rev. 13, pg. 7 and C7033 CO KA 022000K404 [2.8/3.1] 022000K404 ..(KA's) ANSWER: 018 (1.00) REFERENCE: TRHB 11.2, Re , pg. 11 and C3912 COT and Logic Sheet 24 KA 056000A204 [2.6/2.8) 056000A204 ..(KA's) l l

i 

       . -

r SENIOR REACTOR OPERATOR Page 60 ANSWER: 019 (1.00) l a.

t . 

      !

l 

. REFERENCE:      I l  ~. :

TRHB 13.7, Re , pgs 28 and 29 and C4210 COT and Logic Sheet 10 l 

[EBQ 052-05-149A]

l , KA 059000K419 [3.2/3.4]  ! 

      '

l I 059000K419 ..(KA's)  ! ! I L ANSWER: 020 (1.00) i b.

l REFERENCE: 

.TRHB 10.15, Rev. 5, pg. 24 and C4504AOT and P&ID 648J972 Sheet 1 l KA 071000A405 [2.6/2.6]    ~

071000A405 ..(KA's) i 

      !

ANSWER: .021 (1.00) l REFERENCE: i 

- .TRHB 10.2, Rev. 5, Figure 10.2.1.B and P&ID 541F091 Sheet 1 KA 002000K606 [2.5/2.8)

I j 002000K606 ..(KA's) ! I i 

-

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

SENIOR REACTOR OPERATOR Page 61 ANSWER: 022 (1.00) b.

REFERENCE: 

, .

TRHB .10. 8 ', Rev. 4, pg. 15 and Control Panel CO2 KA 006030A401 [4.4/4.4] 006030A401 ..(KA's) 

'
     !

ANSWER: 023 (1.00)  ! c.

REFERENCE: TRHB 10.8, Re , pg. 18 and Technical Specification 15. KA 006000G005 [3.5/4.2] 006000G005 ..(KA's) i I i ANSWER: 024 (1.00) ) I a.

REFERENCE: 

     .!

TRHB 10.3, Re , Figure 10.3.7, Logic Sheet 18 and ICP 1 l KA 010000A403. [4. 0/3. 8] 010000A403 ..(KA's) 

     !

_ 

     ;

SENIOR REACTOR OPERATOR Page 62 AN.9WER: 025 (1.00) 

' .
     ;

REFERENCE: 

, .

TRHB 13.5, Re , pg. 21, C0905C01, Logic Sheet 18 and P&ID 541F091 Sheet 1 KA 010000K301 [3.8/3.9) 010000K301 ..(KA's) ANSWER: 026 (1.00) i REFERENCE: I TRHB 13.6, Re , pg. 7 and C1006 COT and Logic Sheet 18 KA 011000K604 [3.1/3.1] 011000K604 ..(KA's)

ANSWER: 027 (1.00) I ' REFERENCE: , TRHB 13.6, Re , p . KA.011000A101 [3.5/3.6] 011000A101 ..(KA's) i 

 , - - - -

SENIOR REACTOR OPERATOR Page 63 ANS : 028 (1.00) 

"'

kUd317Y REFERENCE: , 

  [ggggg l TRHB 13.3, Rev. 4, pg. 3 KA 012000K406 (3.2/3.5)

IC)Nf 012000K406 ..(KA's) 

,

-ANSWER: 029 (1.00) l 

   ,

REFERENCE: l BG CSP-I.3, Rev. 7, pg. l i KA 016000G015 (3.6/3.8) l 

   ;

i 016000G015 ..(KA's) 

   ,

ANSWER: 030 (1.00) i' REFERENCE: TRHB 13.8, pg. 2 KA 035010k501 [3.4/3.9) 035010K501 ..(KA's) 

. _ _ . .. _ ... .. . . _ _ _... _ _ . _ .4 _ - _ _ , _ _ _ . _ . _ _ _ _ _ . . . _ _ . . _ _ _ _ . . . . _ _ . _  _

l'

3

SEMIOR REACTOR
OPERATOR Page 64 j'

. t

ANSWER
031 (1.00)

l'

a.

i REFERENCE: 

 , .

TRHB 12.5, Re , Figure 12.5.1 and Master Data Book KA 062000K201 [3.3/3.4] 062000K201 . .(KA's) ANSWER: 032 (1.00) l REFERENCE: TRHB 12.2, Re , pg. 38, C4105AOT, OI-110, Re , and PC-29, Rev. 29 Monthly Best KA 064000K402 (3 9/4.2] . 064000K402 . .(KA's) 

          !

I i ANSWER: 033 (1.00) 1 REFERENCE: l TRHB 12.8, Re , pg. 74 and OP-11A, EDG G01/G02 l KA 064000A206 [2.9/3.3] l 064000A206 . .(KA's) 

          ,
         -

. . _ . .._..m..~ . . _._.__.___-__...._ . _ _ _ , _ _ . _ . . _ . . _ . _ . _ . . _ _ . _ _ _ _ _ . . _ _ _
         !

Page 65 

         '

SENIOR REACTOR OPERATOR

I i

{. ANSWER: 034 (1.00)

d.

! 

:

F ! REFERENCE: i * 1.

I 'TRHB 13.12, Re , pg. 14.

! KA 073000A402 [3.7/3.71 6 1  ; j 073000A402 ..(KA's) ' 

:

l ' ANSWER: 035 (1.00) i 3 a.

j i, l .i REFERENCE: TRHB 10.7,.Re , . pg . 17, C0506AOT.and Setpoint' Document, Sections 2.3, i 8.2 and 11.2.

f KA 005000K109 [3.6/3.9] ! i

$ 005000K109 ..(KA's) a

s ANSWER: 036 (1.00)

d.

< REFERENCE: ! 

'

TRHB 13.10, Re , pg. 2; 62804 COT; Setpoint Document 14.13; P&ID PBM-241; OP-1c, Step 7.3.5

KA 041020K603 (2.7/2.9)

041020K603 ..(KA's)

I 

.

i 

;

$ 

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

l

! SENIOR REACTOR OPERATOR Page 66' ' 

        .

i

I j'ANSUER: 037 (1.00)
        -

. l a.

i ,' REFERENCE: 

'AOP-6B, Rev. 6, pg. 1 and LP2441, EO 2.8'.

i I, KA'000005A201 [3.3/4.1] l 

        '

! , i i 1 '

000005A201 ..(KA's)

! 

 '

I Ai?SWER: 038 .(1.00) i l b.: REFERENCE: ECA'O.1, Rev. 10, pg. 12 KA 000015G007 [3.1/3.2] 000015G007 ..(KA's) ANSWER: 039 (1.00) I REFERENCE: EOP-0.1, Rev. 14, pg. ! KA 000024A205 [3.3/3.9) 000024A205 ..(KA's) 

        :
        ,

SENIOR REACTOR OPERATOR Page 67 ANSWER: 040 (1.00) , REFERENCE: , s . EOP-2, Rev. 10, pg. KA 000040A110 [4.1/4.1) 000040A110 ..(KA's) 

   .

ANSWER: 041 (1.00) b.

REFERENCE: . EOP-1, Rev. 19, Foldout pag KA 000040A205 [4.1/4.5] 000040A205 ..(KA's) i ANSWER: 042 (1.00) 

    , l REFERENCE:

TRHB 13.9, Re , pg. 4 and Setpoint Document, Section 1 KA 000051K301 [2.8/3.1] 000051K301 ..(KA'c)  ! l I 

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

l:  : SENIOR REACTOR OPERATOR Page 67 

        :
        ;

ANSWER: 040 -(1.00) i REFERENCE: s .  ; EOP-2,-Rev. 10, pg. KA 000040A110 [4.1/4.1) i 

.000040A110 ..(KA's)     !
      .

ANSWER: b41 (1.00) ' l i l ~b.

l l l-l REFERENCE: EOP-1, Rev. 19, Foldout pag KA 000040A205 [4.1/4.5] 000040A205 ..(KA's) i ANSWER: 042 (1.00) 

. REFERENCE:

TRHB 13.9, Rev. 2, pg. 4 and Setpoint Document, Section 1 KA 000051K301 [2.8/3.1) 000051K301 ..(KA's) !- ! . !e ! I 

    ;

l-SENIOR REACTOR OPERATOR Page 68 l l ANSWER: 043 -(1.00) 

  $f)w n C//4DW ;

l dSr CD && P l , RLF ENCE: ggg gggge7 

.
'AOP-5A, Rev. 4, pg. KA 000051A202 [3.9/4.1)   i 000051A202 ..(KA's)

l

ANSWER: 044 (1.00) l ! REFERENCE: i 

    !

LP 0462, Re , pg. 7 and LO K302 [4.3/4.6] 000055K302 ..(KA's) ANSWER: 045 (1.00) REFERENCE: e BG ECA-0.0, Rev. 14, pg. 14 and LP 0462, LO KA 000055A106 [4.1/4. 5 ] 000055A106 ..(KA's) 

.-_..._.--.__.__...__...._._..__._.._..._..___...m._._.-._. _ . _ ___. ....._... _... _ _ __ ...._ _.. _ ..._... _ __

' 

        .

j.

~ 

        :

l t SENIOR REACTOR OPERATOR Page 69 1 

-

l 

        )

i, , l-ANSWER: 046 ( l'. 0 0 ) i

d.

? i

REFERENCE

j $..- * 

  .

i TRHB 12.9,'Re , pg. 11.

i

000057A218 [3.1/3.1]

!, j 000057A218 ..(KA's) L . ! j ANSWER: b47 ( 1. 0 0 )- . i d.

i.. REFERENCE: I 

        *

l 'TRHB 11.14, Re , pg. 2 t '

i a [Similar to EBQ 052-01-065A] ! ' KA 000067K102 [3.1/3.9] 000067K102 ..(KA's) i i ANSWER: 048 (1.00) 

        ! !

REFERENCE: Tech. Spec. 15.3.6, pg. 15.3.6- KA 000069G003 [3.3/3.9) 000069G003 ..(KA's)

SENIOR REACTOR OPERATOR Page 70 

   !

ANSWER: 049 (1.00) REFERENC , . CSP-C.1, Rev. 12, pg. 2 i KA 000074K201 (3.6/3.9) , i 000074K201 ..(KA's) ANSWER: 050 (1.00) REFERENCE: EOP-0, Rev. 20, pg. 2 and LP 0405 LO 2.1 KA 000007A202 [4. 3/4. 6) 000007A202 ..(KA's)

ANSWER
051 (1.00) REFERENCE:

BG EOP-0, Rev. 21, pg. 9 and LP 0405, LO 2.1 KA 000007K301 [4.0/4.6] 000007K301 ..(KA's) l

SENIOR REACTOR OPERATOR Page 71 ANSWER: 052 (1.00) b.

REFERENCE: 

, .

EOP-0, Rev. 20, pg. 18 and LP 0405, LO KA 000008A101 (4.2/4.0] 

~000008A101 ..(KA's)
.

ANSWER: 053 (1.00) e.

REFERENCE: EOP-1.2, Rev. 13, Foldout Page and LP 0435, LO KA 000009A234 [3.6/4.2] 000009A234 ..(KA's) ANSWER: 054 (1.00) 

.d.

REFERENCE: OP-5A, Rev. 28, pg. 1 and THRB 10.6, Rev. 3, p . KA 000022K307 [3.0/3.2] 000022K307 ..(KA's) 

    ;

  ~1 SENIOR REACTOR OPERATOR Page 72 l

ANSWER: 055 (1.00) a.

REFERENCE: 3 . SEP-1.1, Rev. O, pg. KA 000025K101 (3.9/4.3] 000025K101 ..(KA's) l ANSWER: 056 (1.00) a.

REFERENCE: EOP-3, Rev. 19, pg. 3 KA 000032A101 [3.1/3.4) 000032A101 ..(KA's) l ANSWER: 057 (1.00) a.

REFERENCE: AOP-1A, Re , p . KA 000037G011 [3.9/4.1) 000037G011 ..(KA's)

-SENIOR REACTOR OPERATOR Page 73

ANSWER: - '058 (1.00) REFERENCE:
, .

EOP-3, Rev. 19, pg. 15 and BG EOP-3, Step 20 (PROVIDE: EOP-3, pg. 15 of 36] KA 000038A136 -[4.3/4.5] 000038A136 ..(KA's) 

 ,

ANSWER: 059 (1.00) c.

' REFERENCE: R BG CSP-H.5, Re , pg. 4 and LP 1998, EO .KA 000054K102 [3.6/4.2] 000054K102 ..(KA's) 

    ;

ANSWER: 060 (1.00) ! REFERENCE: BG CSP-H.1, Rev. 10, pg. KA 000054K304 [4.4/4.6] 000054K304 ..(KA's) 

    . . . .

SENIOR REACTOR OPERATOR Page 74 ANSWER: 061 (1.00) a.

REFERENCE: i y .  ! AOP-0.0, Re , p . KA 000058A203 [3.5/3.9) 000058A203 ..(KA's)

ANSWER: 062 (1.00) l l I 

     \

l REFERENCE: TRHB 13.6, Re , pg. 6 and Figure 13.6.1 and Logic Sheet 1 ] KA 000028K202 (2.6/2.7) 000028K202 ..(KA's) ANSWER: 063 (1.00) , REFERENCE: EOP-0.1, Rev. 14, pg. 1 KA 000056K101 [3.7/4.2] 000056K101 ..(KA's) 

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

SENIOR REACTOR-OPERATOR Page 75 , i 

       '

h 1' l' ANSWER: 064 (1.00)- REFERENCE: 

 '
 : .

i NP 1.1.4, Re , pg. \ ' KA.194001A101 [3.3/3.4] l l

194001A101 ..(KA's) ANSWER: 065 (1.00) .J REFERENCE: l l EOP-0, Rev. 20, pg. 1 and SEP-2, Re , p . l KA 194001A102 [4.1/3.9] 194001A102 ..(KA's) f l' ANSWER: 066 (1.00) REFERENCE: 10 CFR 55.53 and OM 3.10, Re , pg. 1 KA 194001A103 [2.5/3.4] 194001A103 ..(KA's)

! , i

l l

l .. . 

 -. .
    - _ . .

SENIOR REACTOR OPERATOR Page 76 i 

   .

ANSWER:' 067 (1.00) REFERENCE: 

 ,   l NP-l'. 7. 5, Rev. 3, pg. KA 194001A103 (2.5/3.4]

194001A103 ..(KA's) ANSWER: 068- (1.00) REFERENCE: NP 1.7.10, Rev. 1, pg. KA 194001A105 [3.6/3.8] 194001A105 ..(KA's) ANSWER: 069 (1.00) 

. REFERENCE:

Manager's Expectations, 93- KA 194001A103 [2.5/3.4] l 194001A103 ..(KA's) l l L i l t l i I i 

- - - - - __ _ - .

SENIOR-REACTOR' OPERATOR Page 77 I ANSWER: 070 (1.00) L i REFERENCE: ' y, . OM 3.10, Re , pg. KA 194001A110 [2.9/3.9] 194001A110 ..(KA's) ANSWER: 071 (1.00) . REFERENCE: , 

  -

i Tech. Spec. Definition 15. KA 194001A113 [4.3/4.1] , 194001A113 ..(KA's) 1 l ANSWER: 072 (1.00) b.

REFERENCE: NP'1.6.6, Re , p . KA 194001A103 [2.5/3.4] 

  !

194001A103 .(KA's) 

  ,
  <

l l 

    . - . - _ - - - -
~

SENIOR REACTOR OPERATOR Page 78 ANSWER': - 073 (1.00) b.

REFERENCE: 

 *
'..

-OM 3.17, Re O, pgs. 4 & KA'194001K101 [3.6/3.7) 194001K101 ..(KA's) ANSWER: 074 (1.00) c.

REFERENCE: NP 1.9.15, Rev. 2, pg. 1 KA 194001K102 [3.7/4.1) 194001K102 ..(KA's) ANSWER: 075 (1.00) d.

REFERENCE: I TRHB 10.6, Rev. 3, pg. 35 and C0413 COT and Setpoint Document, Section KA 004010A105 [3.0/3.2) 004010A105 ..(KA's)

SENIOR REACTOR OPERATOR Page 79 ANSWER: 076 (1.00) c.

REFERENCE: 1 - TRHB 11.4, Re , pg. 15 and Tech. Spec. 15. KA 061000G005 [3.3/4.0] 061000G005 ..(KA's) ANSWER: 077 (1.00) c.

REFERENCE: Tcch. Spec. Bases 15.3.6.A. KA 029000K104 [3.0/3.1] 029000K104 ..(KA's) ANSWER: 078 (1.00) b.

REFERENCE: TRHB 10.9, Re , Figure 10.9.1 and C0801AO KA 008010A302 [3.0/3.1] 008010A302 ..(KA's)

l b SENIOR REACTOR OPERATOR Page 80 i l' ANSWER: 079 (1.00) l l l REFERENCE: l t * l RP-2D, Rev. 10, .pg. l l KA 034000K402 [2.5/3.3) 034000K402 ..(KA's) ANSUER: 080' (1.00) d.

REFEREFCE: Tech. Spec. 15. KA 002000G005 [3.6/4.1) 002000G005 ..(KA's) ANSWER: 081 (1.00) b.

REFERENCE: 

. Tech. Spec. 15.3.3 Basis, pg. 15.3.3- KA 026000A102 [3.6/3.9)

026000A102 ..(KA's) 

.. . - _ _ . . . . . - _ _ _ . _ _ . . . . _ ~ . _ _ _ _ _ _  -.-_.__..~._.m._ ...___ - . _ _ - - - . -_

SENIOR REACTOR OPERATOR Page 81 !- i L iIANSWER: 082 (1 00) i , C.

l 1 REFERENCE-4 , .

, TRHB 13.8, Re , pg. 10 and LP 2441, EO 2.4 and Logic. Sheet 16 ! KA'000001K105 [3.5/3.8]

4

000001K105 ..(KA's) ~ i j ANSWER: 083 (1.00)

4 a.

i } REFERENCE: 

.AOP-6A, Rev. 7, pg. 13, and TRHB 13.13, Rev. 1, pg. KA 000003A101 [2.9/2.9)
'000003A101   ..(KA's)

ANSWE (1.00) REFERENCE: BG EOP-1, Rev. 18, pg. KA 000011K314 [4.1/4. 2 ] 000011K314 ..(KA's) ' 

 . -

_ _ .. _ _ _ _ _ _ SEMIOR REACTOR OPERATOR Page 82 

, ANSWER: 085 (1.00) i REFERENCE:
, .

LG CSP-S.1, Rev. 13, pg. 6 and LP 1996, LO 2.6.

, > KA 000029A115 [4.1/3.9] l 000029A115 ..(KA's) I

I } ANSWER: 086 (1.00) b.

h 

, REFERENCE:

, NP 1.9.14, Rev. O, pg. l KA 000067G012 [3.4/3.4] 000067G012 

'
 ..(KA's)

ANSUER: 087 (1.00) d.

j REFERENCE:

LP 1997, Rev. 2, pgs. 22-24, LO 2.6, CSP-C.1 Background (page 1), BG EOP pg. 13 KA 000074K103 [4.5/4.9)  ; 

      ;
      !

000074K103 ..(KA's) 

;

l l

l SENIOR REACTOR OPERATOR Page 83 ANSWER: 088 (1.00) C.

REFERENCE: 

'.. *

Tcch. Spec. Bases 15.3. KA 000076K305 (2.9/3.6] 

[Similar question used on 94 RO exam.]

000076K305 ..(KA's) ANSWER: 089 (1.00) b.

REFERENCE: AOP-6E, Re , p . KA 000024A202 [3.9/4.4] 000024A202 ..(KA's) ANSUER: 090 (1.00) d.

REFERENCE: LP 0405, Rev. 11, pg. 16 and LO 2.13 and EOP-0, Rev. 20, pg. 2 KA 000009A211 (3.8/4.1] 000009A211 ..(KA's) 

  .- .- - . .- -

SENIOR REACTOR OPERATOR Page 84 - ANSWER: 091 (1.00) REFERENCE: . s . OP-9D, Rev. 12, pg. KA 000060G007 [3.1/3.4] 000060G007 ..(KA's) ANSWER: 092 (1.00) REFERENCE: TRHB 13.2, Re , pg. 50, RMS Alarm Setpoint and Response Book KA 000061A101 (3.6/3.6)

000061A101 ..(KA's) . ANSWER: 093 (1.00) J a.

REFERENCE: LP 0441, Rev. 9, pg. 24 and LO 2.17, EOP-3.1 Background, Rev. 11, p . KA 000038K306 (4.2/4.5] 000038K306 ..(KA's) 

- . _ _ _ . . _ _ _ __

SENIOR REACTOR OPERATOR Page 85 ANSWER: 094 (1.00) C.

REFERENCE: I 

, .

3, pg. ' PBF-5101, Re KA 000036G007 (3.2/3.5] i 000036G007 ..(KA's) l ANSWER: 095 (1.00) 

   ! !

REFERENCE: EOP-0.0, Re , pg. i KA 194001A102 [4.1/3.9] 194001A102 ..(KA's) ANSWER: 096 (1.00) b.

REFERENCE: EP 6.0, Rev. 35, pg. KA 194001A116 [3.1/4.4] 194001A116 ..(KA's)

. SENIOR REACTOR OPERATOR Page 86 ANSWER: 097 (1.00) REFERENCE: 

' *
.

ST-4, Rev. KA 194001A108 [2.6/3.1] 194001A108 ..(KA's) ANSWER: 098 (1.00) d.

REFERENCE: EP 2.0, Rev. 30, pg. KA 194001A116 [3.1/4.4] 194001A116 ..(KA's) ANSWER: 099 (1.00) d.

REFERENCE: HP 3.2.9, Re , p . i 

  '

KA 194001K104 [3.3/3.5] 194001K104 ..(KA's) 

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

SENIOR REACTOR OPERATOR Page 87 d . ANSWER: 100 (1.00) l ' REFERENCE: 

,

NP 4.2.1.18, Rev. O, p . KA 194001K104 [3.3/3.5] ' l

l 194001K104 ..(KA's) i 

;
,

I l l l i 

   .:
 (********** END OF EXAMINATION **********)

}}

Retrieved from "https://kanterella.com/w/index.php?title=IR_05000266/1996303&oldid=2914394"