ML20147A780
| ML20147A780 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| 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) | |
See also: IR 05000266/1996303
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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 Inc.
K. Parkinson, Examiner, Sonalysts Inc.
Approved by: M. Leach, Chief, Operator Licensing Branch
Division of Reactor Safety
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9701280303 970123
PDR ADOCK 05000266
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U.S. NUCLEAR REGULATORY COMMISSION
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i REGION lll
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Docket Nos: 50-266; 50-301 \
DPR-24; DPR-27 ^'
Licenses No:
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Reports No: 50-266/96303(OL); 50-301/96303(OL) i
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l Licensee: Wisconsin Electric Power Company I !
Facility: Point Beach Nuclear Plant Units 1 and 2
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Location: 6610 Nuclear Road
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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. 1
Approved by: M. Leach, Chief, Operator Licensing Branch
Division of Reactor Safety
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9701280303 970123
PDR
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ADOCK 05000266
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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 administration.
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 licenses.
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
administered.
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
limitations.
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)
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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 administration.
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 licenses.
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)
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e Two applicants failed the examination due to unsatisfactory performance while two
other applicants demonstrated weaknesses on the Administrative JPM tasks I
administered.
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
limitations.
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)
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Reoort Details
1. Operations j
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03 Operations Procedures and Documentation
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03.1 Ooerations Procedure Deficiencies I
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] a. Insoection Scooe
An inspection was conducted from November 20-27,1996, to assess procedure l
deficiencies identified during initial examination administration. I
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b. Observations and Findinas !
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The following procedure deficiencies were identified: i
1. 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.
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However, battery chargers D-107 and D-108 were designed such that they I
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could not function properly if connected to a DC bus without a battery also j
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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 4
chargers D-107 and D-108 outside design capabilities which could have
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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
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before the steps which restored power to bus DO3 and bus D04 in AOP 0.0,
i Attachment B.
2. 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
setpoint.
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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 performance.
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 ARB.
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c. 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 V.
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 minor.
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e The licensee had taken prompt corrective actions for the procedure
deficiencies which were considered adequate.
05 Operator Training and Qualification
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05.1 General Comments
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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) applicants. !
05.2 Examination Prenaration and Validation
The NRC prepared the examination in accordance with NUREG 1021 guidelines.
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 good.
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05.3 Examination Administration
a. 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 required.
e No common performanco deficiencies were observed.
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 with. 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
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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,
b. 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"
c. 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
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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 good. i
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05.4 Post Examination Activities
a. 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,
b. 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 weakness.
05.5 125 VDC Electrical Distribution System
a. 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 requirements.
b. 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).
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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
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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
OOS.
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Remainder of DC electrical distribution system was in a normal at power line
up. .
e Battery charger D-108 subsequently fails. 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
bus. 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 allowed.
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
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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-09.
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 reading.
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/S.
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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.
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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 specifics.
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c. Conclusions
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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
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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 rod.
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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.
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L Manaaement Meetinas
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- 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.
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PARTIAL LIST OF PERSONS CONTACTED
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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
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Enclosure 2
l 4 Post Examination Comrr'ents and Review 1 l A. Licensee Submitted Comments and NRC Resolution:
- The licensee's written examination post review resulted in the following two
- questions submitted for NRC review:
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1. QUESTION (RO EXAM 044/SRO EXAM 043):
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Given the following Unit 2 plant conditions: I
- Condenser vacuum is 25 inches Hg.
- Generator load is 250 MW.
Which one of the following is the MAXIMUM amount of time the turbine can l
be operated (Operational Back Pressure region curve is attached)?
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a. Operation is prohibited,
b.10 minutes.
c.1 hour.
d. Operation is unrestricted. .
ANSWER: b. %
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 question.
We recommend that both "a" and "b" be accepted for this question.
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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 response.
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2. 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?
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a. Permits operator override in the event of spurious actuation.
b. Prevents spurious actuation caused by S/G shrink and swell,
c. Allows time for reactor power to decrease below 40% before
actuation.
d. Allows time for the AFW pumps to recover S/G levels before
actuation.
ANSWER: a.
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 answer.
NRC RESOLUTION:
Concur with recommendation. This question was deleted from the
examination.
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 instrument.
RO 030; SRO 029 RVLIS (Reactor Vessel Level Indication System) readings that
would indicate the highest probability of core voiding.
RO 039; SRO 038 Symptoms of a stuck control rod requiring entry into AOP-6B,
" Stuck Rod or Malfunctioning Position Indication."
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RO 039; SRO 038 Required method to restore Reactor Coolant Pump (RCP) seal
cooling following a station blackout.
RO 045; SRO 044 Plant status concerns that the Operators would be confronted I
with following a station blackout in excess of seven hours. !
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 manually.
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 units.
RO 079 The equipment that is reset by the rod control step counter
reset switch.
RO 086 The FSAR design ratings for the station batteries.
RO 094 The pressurizer pressure control system response to a pressure
detector failure.
SRO 095 The conditions when EOP-0.0, "Rodiagnosis," could be used
based on operator judgement.
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1 l Enclosure 3 ,
SIMULATION FACILITY REPORT
! Facility Licensee: Point Beach Nuclear Plant
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Facility Licensee Docket Nos: 50-266; 50 301
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Operating Tests Administered On: October 8-11,1996
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- 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 3
observations,
i j While conducting the simulator portion of the operating tests, the following items were
- observed
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ITEM DESCRIPTION
! l 1. Plant Process Computer System A thermocouple tilt was incorrectly indicated ! during an individual rod position indication 4
malfunction for rod D-10. This was NRC
j identified. The licensee generated Simulator
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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.
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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 3
1875-2025 for the one hour run. However,
1 KVARs could not be raised greater than 1650.
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 96-0081. l
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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 troubleshoot.
6. 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 96-0080.
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U. S. NUCLEAR' REGULATORY COMMISSION
SITE SPECIFIC EXAMINATION
REACTOR OPERATOR LICENSE
REGION 3
1
,. , CANDIDATE'S NAME: MASTER EXAMINATION
FACILITY: Point Beach 1 & 2
REACTOR TYPE: PWR-WEC2
1
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 starts.
CANDIDATE'S
TEST VALUE SCORE %
44. 6e
hd=0 0 0. 0 0 - % TOTALS
FINAL GRADE
All work done on this examination is my own. I have neither given nor i received aid. I
Candidate's Signature
,
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PAGES 2, 3, AND 4, WERE USED AS ANSWER SHEETS
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AND ARE NOT INCLUDED IN THIS MASTER EXAMINATION COPY. , .
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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.
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 examination. This must be
done after you complete the examination.
3. 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 cheating.
,
4. Use black ink or dark pencil ONLY to facilitate legible reproductions.
5. Print your name in the blank provided in the upper right-hand corner of
the examination cover sheet and each answer sheet.
6. Mark your answers on the answer sheet provided. USE ONT.Y THE PAPER PROVIDED
AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
4
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 page.
8. Use abbreviations only if they are commonly used in facility literature.
Avoid using symbols such as < or > signs to avoid a simple transposition
error resulting in an incorrect answer. Write it out.
9. The point value for each question is indicated in parentheses after the
question.
10. Show all calculations, methods, or assumptions used to obtain an answer to
3
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 BLANK.
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 which is worth 0.25
) points, and you give five responses, each of your responses will be worth
0.20 points. 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 answers.
13. If the intent of a question is unclear, ask questions of the examiner only.
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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 revoked.
.
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QUESTION: 001 (1.00)
Given the following Unit 2 plant conditions:
- A large break LOCA has occurred.
- St t=0, an SI signal was generated.
'
- At t=30 seconds, a containment spray signal was generated.
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
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b. 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
c. 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,
b. Prevent backflow in a cooling unit in the event of fire in
Containment.
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 accident.
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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-tubes.
c. To reduce the mass inventory available to boil off in the event
of a steam break.
d. To prevent low power level o.scillation due to level dominant
control.
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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'
d. Emergency Diesel Generator Room
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QUESTION: 005 (1.00)
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Which ONE of the following would result if the reactor head to vessel inner "O" ring seal completely fails?
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a. Pressurizer Relief Tank level would increase.
! b. SI on high containment pressure.
c. SI on low PZR pressure.
d. Reactor Coolant Drain Tank level would increase.
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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 out.
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 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 pumps.
b. A CCW high radiation monitor alarm is received.
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c. CCW pump discharge low pressure alarm is received.
d. Unable to maintain CCW Surge Tank level.
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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).
d. 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. Shutdown.
- RCP 1A motor is uncoupled from the pump.
- RCS is filled and vented.
- Maintenance is inspecting RCP 1A seals.
Which ONE of the following minimizes leakage of reactor coolant upward
clong the RCP shaft?
a. The pump shaft mates with the thermal barrier casing. ,
b. Nozzle dam installation prevents RCS water from entering the RCP
shaft area.
c. Seal injection is maintained during this condition.
d. Seal leakoff collects any RCS leakage up the shaft and directs
it back to the VCT.
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
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b. B09-
c. 1B03
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 only.
b. Low pressurizer pressure and steamline low pressure only. :
c. Low pressurizer pressure, steamline low pressure, and
containment high pressure only.
d. Low pressurizer pressure, steamline low pressure, containment
high pressure, and manual.
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
detect.
b. Neutron flux is greater in the bottom half of the core during a
-startup.
c. To allow a spare detector to be installed in the upper half of
the instrument should the installed detector fail.
d. This position provides optimal cooling for the detector by
natural circulation of air in the detector well.
m A m. ,.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
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10 5
ELECTRICALLY ADJUSTED COMPENSATED ION CHAMBER
G
e 104
E
TYPICAL SHUTDOWN CURVES
$ 10 4
E
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m
m 108
Y
m
2
H
$ 104
- E
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U / IV
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$ 1010 l
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10 11 \ -
\ NEUTRON SOURCE LEVEL
ll N
N
30.u
O 2
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6 8
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10 12 14 16 18 20 22 24
TIME AFTER REACTOR SHUTDOWN (MINUTES)
TYPICAL GAMMA COMPENSATED CURVE FOR A COMPENSATED lON CHAM"5R
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REACTOR OPERATOR Page ~_3
QUESTION: 015 (1.00) j
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Giv:n the following Unit 2 plant conditions: ;
- Unit is at 90% power.
- Rod control in automatic.
- Power Range NI N41 fails HIGH (top of scale).
1
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
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
Tref.
d. Rods initially drive OUT and then STOP and remain at that
position.
QUESTION: 016 (1.00)
Givsn the following Unit 1 plant conditions:
- Reactor is shutdown.
- Reactor decay heat is being removed by natural circulation.
- RCS pressure is 1550 psig.
- Average core thermocouple temperature is 402 degrees F.
Which ONE of the following describes 'he approximate amount of
subcooling that exists in the RCS?
a. 50 deg F.
b. 100 deg F. j
c. 200 deg F.
d. 300 deg F.
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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 upper. i
' support plate.
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! b. They are arranged in strings extending from the upper support j
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plate to the core mid-plane. '
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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.
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, 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 head. ;
c. Catastrophic failure of the reactor vessel flange o-rings.
~
) d. Damage to the Core Exit Thermocouples's electrical circuitry 4
from overheating.
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QUESTION: 019 (1.00) !
, Which ONE of the following conditions will result in an automatic trip i
of 911 operating Condensate Pumps? l
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a., Steam generator feed pump suction pressure at 179 psig. )
b. Containment pressure at 5.2 psig.
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c. Steam generator level at 73%. l
d. Condenser hotwell level at 10 inches.
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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
cleared. l
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b. Locally reset the closing solenoid valves once the closure I
signal has been cleared and reset. j
c. Go to manual on the feedwater control valve controller and open
the valve. l
d. The condition which caused the isolation signal must be restored
to its pre-trip value.
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
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b. CVCS Holdup Tank
c. Pressurizer Relief Tank
d. Reactor Coolant Drain Tank
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QUESTION: 022 (1.00)
Tho'high pressure sensing line on the RCS loop B flow instrument FT-414
breaks.
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 indication.
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 indication. ]
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-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 power.
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b..The component is.in an abnormal alignment.
c. The component has lost DC control' power.
d. The component is running / energized.
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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 feet.
b. Pressure is 730 psig.
c. Boric acid concentration is 1910 ppm.
d. Outlet Isolation valve is OPEN and the control switch RED
position indicating light is NOT LIT.
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s, e l QUESTION: 025 (1.00) i j Given the following Unit 1 plant conditions: 1 !
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Pressurizer pressure defeat sw, itch is in its normal position.
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PT-449 (Yellow Channel), Pressurizer Pressure, has just failed
' LOW.
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2 ,
pressure control system to this failure?
e
- a. Only PORV 431C cannot operate automatically (PORV 430 operable).
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b. Only spray valve 431A closes (431B remains as is).
c. Both spray valves CLOSE fully.
d. Both PORVs are PREVENTED from automatic operation.
QUESTION: 026 (1.00)
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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 only.
b. Pressurizer PORV PCV-430 opens and all pressurizer heaters !
deenergize only. ,
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c. Pressurizer PORV PCV-430 opens and both pressurizer spray valves
open only.
d. Pressurizer PORV PCV-430 opens, all pressurizer heaters i
deenergize, and both pressurizer spray valves open. l
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REACTOR OPERATOR Page 19
QUESTION: 027 (1.00)
Giv:n the following Unit 2 plant conditions:
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- Unit is at 50% power. l
- Rod control is in MANUAL. l
- Lo6p B cold leg temperature ';ector TE-401B fails high.
- No operator action is taken. l
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Which ONE of the following will be the steady-state pressurizer level?
a. 20%
b. 33%
c. 46%
d. 100%
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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 .
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b. Pressurizer liquid temperature increases.
! c. The reference leg for LT-428 cools down due to a decrease in
containment temperature.
d. Containment pressure increases to 0.3 psig; containment
temperature remains constant.
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REACTOR OPERATOR Page 20
&ScjN h I'20
QUESTION: 029- (1.00)- [gt sM
Which ONE of the following is the reason for the 30.s d time delay
associated with the AMSAC actuation relay? l
a. Permits operator. override in the- nt of spurious actuation.
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b. Prevents spurious actuat caused by S/G shrink and swell. .l
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c. Allows time for tor power to decrease below 40% before
actuation.
d._ Allo ime for the AFW pumps to recover S/G levels before
uation.
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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 running.
b. Narrow Range reading 38 ft. with NO RCPs running.
c. Wide Range reading 65 ft. with ONE RCP running.
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d. Narrow Range reading 95 ft. With BOTH RCPs running.
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REACTOR OPERATOR Page 21 QUESTION: 031 (1.00)
Given the following Unit 1 plant conditions:
- Unit is operating at 75% steady state power.
- All systems are in automatic control.
-\The "A" S/G atmospheric valve fails open.
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
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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
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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?
1
a. 30 minutes.
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b. 1 hour.
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c. 24 hours.
d. 2000 hours. ;
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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
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REACTOR OPERATOR Page 23 l
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QUESTION: 036 (1.00) :
Givsn the following Unit 2 plant conditions:
- RHR is in service.
- RCS pressure is 320 psig and INCREASING.
- RCS temperature is 340 degrees F and INCREASING.
- ALL system lineups are in a normal shutdown configuration for
-solid-plant operation.
Which ONE of the following will act FIRST to prevent overpressurizing
tha'RHR System?
a. Pressurizer PORVs will open.
b. RHR return isolation valve will auto close.
c. RHR pump hot leg relief valve RH-PCV-861C will open.
d. RHR pumps discharge relief valve RH-PCV-861A will open. 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 seconds.
c. Going to TEST on the individual valves on the back of C03.
d. Reaching 90 psig in the crossover steam header.
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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 1
[ QUESTION: 039 (1.00)
- Given the following Unit 1 plant conditions
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- 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 1 ! c. Seal injection flow is initiated to cool the seals to less than
190 degrees F, then CCW flow to the RCP is established.
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d. CCW flow is initiated to cool the seals to less than 190 degrees
{ F, then seal injection flow is established. i .
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REACTOR OPERATOR Page 25 QUESTION: 040 (1.00)
Givsn the following Unit 1 plant conditions:
- A plant trip has just occurred.
- 2 control rods are stuck out of the core following the trip.
- 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?
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a. 600 gallons
b. 1,200 gallons !
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c. 2,400 gallons
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d. 3,000 gallons
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QUESTION: 041 (1.00)
Givan the following Unit 1 plant conditions:
- Steam generator A is faulted due to a feed line break outside of
-containment.
- The crew is performing actions of EOP-2, " Faulted Steam Generator
Isolation."
- The AFW system is in operation.
i
Which ONE of the following actions concerning the AFW pumps is required
by EOP-2?
a. Shutdown the AFW Pumps immediately. 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/Gs.
d. Isolate the AFW Pumps from S/G A (steam and AFW flow).
:
i
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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
1
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
1
.
QUESTION: 043 (1.00)
Given the following Unit 1 plant conditions:
- Unit is at 100% power.
- Condenser vacuum is decreasing.
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 Hg.
- Generator load is 250 MW.
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,
b. 10 minutes.
c. 1 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
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_ _ _ _ _ _ _ _ _ __ - . - _ __ _ . . _ _ .
___ _ . . _ _ .. _ _ REACTOR OPERATOR Page 28
f
QUESTION: 046 (1.00)
Given the following Unit 1 plant conditions:
- A loss of all AC power has occurred.
- ECA-0.0, " Loss of All AC Power" is in effect.
- 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
restart.
b. An uncontrolled use of water that may be needed for long term
cooldown.
c. An uncontrolled start of large loads on safeguards AC buses,
d. An uncontrolled depressurization of the RCS.
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? ,
1
a. May fail open; manual control will NOT be available to regain '
control. 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
j.
c. Service Water Pump area
d. Auxiliary Feed Pump room
.
QUESTION: 049 (1.00)
Technical Specifications require containment pressure to be between +3.0
psig and -2.0 psig during power operation.
~
1
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
d. 24-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
b. I, II and IV
_> I c. I, III and IV.
,
d. II, III, and IV.
J
QUESTION: 051 (1.00)
Given the following Unit 1 plant conditions:
- The Reactor tripped from 100% power.
- EOP-0, " Reactor Trip or Safety Injection" has been entered.
- The main turbine did NOT trip as expected.
- MANUAL turbine trip is unsuccessful.
Which ONE of the following is the NEXT action required per EOP-0?
a. Manually close MSIVs and Bypass valves.
b. Locally trip the main turbine.
c. Go to CSP-S.1, " Response to Nuclear Power Generation /ATWS."
d. Verify Safeguards Buses energized.
. - -
,
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) ! ! 'tubesheets. I 1 i b. To prevent excessive cooldown of the RCS. i i
c. To prevent S/G overpressurization.
l.
;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 occurred.
- Pressurizer PORV 1RC-431C lifted and is stuck open.
Which ONE of the following is the MAXIMUM pressure below which the PORV
isolation valve must be shut?
a. 2450 psig
b. 2335 psig
c. 2225 psig
d. 2010 psig
!
,
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. . ~ . - ~ . . - .- -. - .. -- . _ . __-.- . .. -. . _. - _ - . . - _ . _ . . -
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,
1
7
i'
-REACTOR OPERATOR Page 32
1 1 j QUESTION: 054 (1.00) !
<
-Given the following Unit 1 plant conditions:
, - Unit has tripped from 100% due to a small break LOCA. l
i '
- Conditions have stabilized and operators are evaluating the !
briteria for terminating SI.
'
'
--Adverse Containment conditions do NOT exist.
'
- 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% NR. !
1
a
- b. Pressurizer level indicates 114. I
1
j c. RCS subcooling'is 40 degrees F.
d. 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. 1
b. Reduce thermal shock on the Non-Regenerative Heat Exchanger.
c. VCT level will decrease causing possible damage to the charging
j. pumps. !-
d. Reduce thermal shock on the charging penetration into the RCS.
4. b
}
1
:
'
f ,
- - - c , - - --- _
. _ . . . . .. . _. _ _ _ _ . _ _ _ . _ _ _ . . . _ _ _ _ . . _ _ _ . . _ - , _ _ _ . _ _ _ _ _ . _. :
J
I ' REACTOR OPERATOR Page 33
J- f QUESTION: 056 (1.00) j _Given the following Unit 1 plant conditions:
1
'
i - Unit is in Cold Shutdown with RHR Cooling in progress. I - The RCS is solid. i
- kHR flow is lost and CANNOT be restored. l
- All other systems and components are available. 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 valve.
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 1 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 rupture.
- 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 instruments.
b. Depress both SR RESET pushbuttons.
c. Depress both IR BLOCK pushbuttons.
d. Deenergize N36.
_ _ . . . _ . _ _ _ . _ . _ _ . _ . _ _ - . _ _. .__ _- _ _ _ ____ _ _ _ _ _ _ .
i 1
REACTOR _ OPERATOR Page 34
1 .:
'
QUESTION: 058 (1.00)
1
- 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
chargingandletdownsystemoperations.
,
a. Pressurizer level is at 11% and decreasing.
3
'
b. Narrow range S/G level is 62% and increasing.
! I
c. VCT level is at 17% in AUTO control and decreasing.
i j. d. 3/G Blowdown Isolation valve closes due to high radiation.
,
! { 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 psig.
- 1B 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
4
c. 485 degrees F
-
d. 480 degrees F
U.
.
I i !
4 I , 9
e - >,m ., , - ,
_ . _ _ _ _ _ _ _ __ _ _ _ . _ _ _ _ _ . . _ . _ . _ . . _ . .._ _ _ . . . . - . . _ _ . _ _ _ . . ._ _. _
'
d ,
t
,
POINT REACH NUCLEAR PLANT EOP-3 Unit 1
'~
EMERGENCY OPERATING PROCEDURES MAJOR.
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 a. Using ruptured steam l generator pressure, detemine l
target"coretexit temperature.
1
.
1 ) Ruptured Steam Generator Core Exit
Pressure (psig) Temperature (*F)
i
1100
[460) 505
1000 [445] 490
- 900 [435) 480
g 800
[420] 465
4
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 N. .
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 ring.
b. To prevent reactor restart from an excessive cooldown.
c. To minimize thermal stresses t) S/G components,
d. To prevent exceeding reactor vessel cooldown rate limit.
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 injection.
b. It extends the time available to restore feed flow before bleed
and feed criteria is met.
c. It anticipates an RCS pressure decrease caused by spray valves
opening when air is restored to containment.
d. It anticipates an RCS pressure decrease caused by opening PZR
PORVs during bleed and feed.
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 events.
- Charging flow is reduced to minimum.
I - Pressurizer level decreases.
- Letdown. flow is secured and heaters turn off.
- 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
b. Pressurizer level channel 428 (Blue) failed high
c. Pressurizer level channel 427 (White) failed low.
d. Pressurizer level channel 427 (White) failed high.
__ _ _._. _ ._ _. _ _. _ . . _ . . . _ . . . . _ _ . . _ . _ . - _ . - _ . _ . _ - _ _ _ _ . . _ . . - _ _ _ . _
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 !
power.
- 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 F.
>
d. Core exit thermocouples at saturation temperature for steam
generator pressure.
,
QUESTION: 065 (1.00) j
Which ONE of the following " Procedure Usage Levels" allows performance
of all activities from memory?
. a. Infrequent Use !
b. 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 F.
$ .
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
used.
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 applicable.
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
b. 5 shifts per calendar year l
'
l
c. 7 shifts per calendar quarter
1
d. 7 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
c. 1300
d. 1500 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 operator.
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 room.
b. Immediately notify DCS by two-way radio and then leave the D/G
room.
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
1
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 Control.
\, :
b. Turbine manual using governor valve decrease.
c. Valve Position Limiter.
l
d. Turbine manual using Governor fast. 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 parameter.
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 ?
>
1
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 Operator.
b. Must be done by visually observing the operating link.
c. Can be accomplished using status lights,
d. Can be accomplished using remote indicators.
.- .. ._ _
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 present.
V :
b. Work is to be performed in a high radiation area.
c. Requires direct line of sight of the isolation.
d. Isolation boundaries are locked.
i
!
QUESTION: 076 (1.00) )
4
Giv@n the following Unit 1 conditions: ,
l
'
.l - Unit is at 100% power.
- VCT level is 23%.
- All controls are in automatic.
- LT-112, VCT level transmitter, fails high.
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 HUT.
b. Increases to 28% where auto-makeup stops.
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 operating.
- The South Condensate Storage Tank (CST) is isolated for repairs.
- The North CST is selected for Auxiliary Feedwater Pump suction at
both units.
Which ONE of the following is the MINIMUM volume required in the North
CST?
a. 13,000 gallons
b. 20,000 gallons
c. 26,000 gallons
d. 30,000 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 automatic.
The initial output for the step change at 80 % is:
a. larger due to the response of the Variable gain unit,
b. larger due to the response of the Power mismatch rate / lag unit.
c. smaller due to the response of the Variable gain unit.
d. smaller due to the response of the Power mismatch rate / lag unit.
.
_ . . . _ _ _ _ _ _ _ _ . _ . . _ _ . ._ _- . _ _ . _ _ . _ -. . .
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 recorder.
b. P/A converter, master cycler, multiplexer, bank overlap unit
trigger circuit. ,
4
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 alarm.
.
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 seal.
b. The pressure is too low and insufficient seal flott will be
obtained through the #3 seal.
c. The pressure is too high and will force excessive flow through
the #2 RCP seal. l
d. The pressure is too high and will force excessive flow through
the #3 RCP seal.
i
1
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 inspection.
b. Remove airborne radioactivity from containment immediately
following a LOCA.
c. Provide additional cooling to containment during hot shutdown
conditions.
d. Remove airborne radioactivity from containment during outages.
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 size.
b. Main Steam Flow Venturi.
c. Main Steam Isolation Valve.
d. Main Steam Flow Limiter.
- . . - - . . . _ _ .
1
REACTOR OPERATOR Page 47
1
l
1
l
QUESTION: 086 (1.00)
Giv:n the following Unit 2 plant conditions:
- The reactor was operating at 100% power when a reactor trip
occurred.
- A.1oss of all AC power occurred at the time of the reactor trip.
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
b. 1 hour
c. 8 hours l
d. 24 hours
1
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 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 PRT.
:
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 tank.
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 droplets.
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
accurate.
b. To ensure explosive mixtures will not exist in the Auxiliary
Building exhaust system.
c. To ensure hydrogen sampling system capacity is not exceeded.
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 cart.
b. RCCA change fixture carriage.
c. Manipulator crane gripper assembly.
d. Control rod drive shaft unlatching tool.
:
{
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
-
containment.
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 power. ;
I l l l QUESTION: 094 (1.00)
Giv:n the following Unit 1 plant conditions:
i
- The plant is at 100% power.
- Pressurizer pressure is in AUTOMATIC control.
- The Pressurizer Pressure Channel Defeat switch is in the normal
position.
- 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 setpoint.
I
b. Pressurizer heaters and spray valves operate normally to I
'
maintain pressurizer pressure.
I l c. Pressurizer pressure DECREASES, resulting in a LOW pressure ) l
reactor trip.
d. Pressurizer pressure INCREASES, resulting in a HIGH pressure l
reactor trip.
I
1
!
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 MANUAL.
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 block.
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 zero.
d. Resetting the Pulse-to-Analog (P-A) Converter removes the block.
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 legs.
b. Vapor bubbles formed in the core and condensing in the S/Gs to
flow back to the core through the cold legs.
c. Boiling in the core and steam condensing in the S/Gs to flow
back to the core through the hot legs. !
d. Vapor bubbles formed in the core and condensing in the head area
to flow back into the core.
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 identified.
I
b. Upon steam generator level reaching the high level alarm.
c. If RCS temperature decreases below 547 degrees F.
d. Actions to trip and runback the turbine are unsuccessful.
1
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 trip.
b. Loss of Circulating Water flow.
c. High level alarm on the Waste Holdup Tank.
.
d. An alarm condition on Radiation Monitor Channel RE-218.
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
b. 85 psig
c. 80 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 visitor.
l
b. Make a page announcement to direct the visitor to report to the
'
control room.
c. Notify security _to deenergize vital area door card readers,
d. Notify Central Alarm Station.
,
i
l
i
.
(********** END OF EXAMINATION **********)
- . . . , , . - - . - . . .-. - . - . . . . - . - . . . . . . . . . . . . . _ . . - . . . . . . . . . . . . . . . . - . -
~ REACTOR' OPERATOR < Page 54
1
. ANSWER: 001 ('1. 0 0 ) l
l
'
C.
I
i
REFERENCE:
*
. g
TRHB'10.12, Rev. 3, pg. 6; C1301AOT; Setpoint Document, Section 2.4; and
Logic Sheet 9. l
l
KA 026000A301 [4. 3 /4. 5] j
4
l
026000A301 ..(KA's)
.
ANSWER: 002 (1.00)
a.
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
C. !
l ,
REFERENCE:
l TRHB 13.7, Rev. 1, 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 )
9. ,
- REFERENCE
- ,
8 .
AOP-10A, Rev. 18, pg. 12.
[Similar to question used on 94 RO exam.] ;
KA 000068A101 [4.3/4.5]
000068A101 ..(KA's)
.
'
ANSWER: 005 (1.00)
d. 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) ;
C.
f
i
l
. .-. ~ .. . ... . - _ . ... .- . . - -. - - . - -. - .. - . - . - . -_ - .. . .... - .. .... . -..... ..
'
1
-
l
' REACTOR OPERATOR Page'56 I
f REFERENCE: 1
'
AOP-6C, Rev. 5, pg. 3 and LP 2441,.EO'2.17.
KA~0000017.205.[4.4/4.6]
.
(Similar question used on 95 SRO exam.]
.
v .
l 0000611205 ..(KA's)
? 1. i . ANSWER: 4
007 (1.00)
! d. j.
l-REFERENCE: 1. AOP-9B, Rev. 7, pg. 4.
i
- '
KA.000026G010 [3.6/3.5] 4
i
(Similar to question on 95 SRO exam.]
- 000026G010 ..(KA's)
1
!
I 008'
AMSWER: (1.00)
-
l
- a. '
i
l l
I
{ REFERENCE: { 10 CFR 20.1003, Definitions.
[Similar question used on.95 SRO exam.]
l
KA 194001K103 (2.8/3.4] l
.
194001K103 ..(KA's)
1
. 4
l i
i i
4 $ >
,
> , , - . _ -- . . .
!
REACTOR OPERATOR Page 57
l
*
ANSWER: 009 (1.00)-
5.
REFERENCE: .
5.- ? i
TRHB 10.2, Rev. I
5, pg. 16.
i
KA 003000K407 (3.2/3.41
1
003000K407 ..(KA's) ,
l
l
ANSWER: 010 (1.00) j
:
b. l
l
REFERENCE:
l
- TRHB 10.6, Rev. 3, pg. 71 and Master Data Book for B09. 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, Rev. 4, pg. 18 and C730 COT and Logic Sheet 7.
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 COT.
KA 015000A202 [3.1/3.5] '
4
015000A202 ..(KA's)
ANSWER: 014 (1.00)
i b.
REFERENCE:
TRHB 13.1, Rev. 3, pg. 11 and C1310 COT.
KA 015000K601 [2.9/3.2]
015000K601 ..(KA's)
-_ _ . _ _ _
. . .. . - _ _. .. . . __
REACTOR OPERATOR Page 59 ANSWER: 015 (1.00)
b.
REFERENCE: ,
TRHB 13.8, Rev. 1, pg. 6 and C0105 COT; and TRHB 13.8, pg. 19 and Logic Shtot 16. 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, Rev. 3, pg. 8 and TRHB Figure 10.4.1. KA 017020K102 [3.3/3.5] 017020K102 ..(KA's)
-. - - _ . ~ . - . . . . . . . . . . . - ~ . . - . . _ _ . . - . . . . . - . . - . - _ . - . . 1-
). ;
'
1
1
; REACTOR OPERATOR.
1
Page 60
1
. j. ANSWER: 018 .(1.00)
.b. l
a 1 REFERENCE: * -
s . i
1
,
BG EOP-0.2, Rev. 13, pg. 7 and C7033 COT. l ; KA 022000K404 [2.8/3.1]
., j hl
022000K404 ..(KA's)
i s
#
J $ ANSWER: 019 (1.00)
- l
I b. ,
1
,t , ?! j= REFERENCE: '
TREB 11.2, Rev. 5, pg. 11 and C3912 COT and Logic Sheet 24
. KA 056000A204 [2.6/2.8)
1
t ; j 056000A204 ..(KA's) i . \
l ANSWER: 020 (1.00) i t a. 1 i
,
- REFERENCE
4 TRHB 13.7, Rev. 1, pgs 28 and 29 and C4210 COT and Logic Sheet 10 l. i [EBQ 052-05-149A] 1 1
l KA'059000K419 [3.2/3.4]
{
; 059000K419 ..(KA's) 1 1
1 . k 4 3
-
.. .- ~ . - . . . , . . . , - . ,
. ._ _ . .- .. ._ ._. __. REACTOR. OPERATOR Page 61
!
ANSWER: 021 (1.00)
b.
REFERENCE:
,
TRHB 10.15, Rev. 5, pg. 24 and C4504AOT and P&ID 648J972 Sheet 1
KA 071000A405 [2.6/2.6]
071000A405 ..(KA's)
AMSWER: 022 (1.00)
d.
REFERENCE:
TRHB 10.2, Rev. 5, Figure 10.2.1.B and P&ID 541F091 Sheet 1
KA 002000K606 [2.5/2.8)
002000K606 ..(KA's)
ANSWER: 023 (1.00)
b.
REFERENCE:
TRHB 10.8, Rev. 4, 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, Rev. 4, pg. 18 and Technical Specification 15.3.3.
3
KA 006000G005 [3.5/4.2] I
006000G005 ..(KA's)
l l
9 ANSWER: 025 (1.00) a.
j '
REFERENCE: TRHB 10.3, Rev. 3, Figure 10.3.7, Logic Sheet 18 and ICP 10.2 KA 010000A403 [4.0/3.8)
. '
010000A403 ..(KA's)
ANSWER: 026 (1.00)
- a.
,
REFERENCE: 1
1
TRHB 13.5, Rev. 2, 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
-
2 s I i l ANSWER: 027 (1.00) ! i I '
c. l
l l REFERENCE: j i; j TRHB 13.6, Rev. 1, 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, Rev. 1, pg. 5.
'
KA 011000A101 (3.5/3.6)
011000A101 ..(KA's)
WER: 029 (1.00)
!
a. l
4
REFERENCE: g)ggppy I
TRHB 13.3, Rev. 4 g. 37. wwW
KA 012000K406 (3.2/3.5
012000K406 ..(KA's)
@
_
. _ . _ . . . . _ _ . _ _ . . - . . _ . . _ _ . . _ = _ _ _ _ . _ _ _ . - . _ . . _ . _ _ _ _ . _ . - . . _ _ _ REACTOR OPERATOR' Page 64 ANSWER: 030 (1.00)
l- 'b.
REFERENCE:
N-
l BG CSP-I.3, Rev. 7, pg. 1.
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, Rev. 2, Figure 12.5.1 and Master Data Book
KA 062000K201 [3.3/3.4)
i
062000K201 ..(KA's)
'
i I ! !
. . _ _ -- - , _ . _ . , . _ . ,
- _ - - _
. - . . ~ . . . - . . . . . . . . . . . - - . . - . . _ . . . . - . . - . . . . - _ . - . . . . . _ . - ..-- -....-. -... ..
1 [ i i REACTOR OPERATOR Page 65 I ! ' , l I j-. ANSWER: 033 (1.00) -
t
l l C.
-
, 4
REFERENCE: ,
,
. . ;
TRHB 12.2, Rev. 1, 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)
'
a.
r
I
,
REFERENCE:
TRHB 12.8, Rev. 4, pg. 74 and OP-11A, EDG G01/G02
KA 064000A206 [2.9/3.31
1
1
064000A206 ..(KA's)
ANSWER: 035 (1.00)
d.
REFERENCE:
TRHB 13.12, Rev. 3, pg. 14.
KA 073000A402 (3.7/3.7]
1
i
'
073000A402 ..(KA's)
i
I
i
l
!
I
1
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 11.2.
KA 005000K109 [3.6/3.9)
005000K109 ..(KA's)
.
ANSWER: 037- (1.00)
d.
REFERENCE:
.TRHB 13.10, Rev. 2, pg. 2; 62804 COT; Setpoint Decament 14.13; P&ID PBM- 241;.OP-1c, Step 7.3.5 KA 041020K603 [2.7/2.9) 041020K603 ..(KA's)
ANSWER: 038 (1.00)
a.
l
l
REFERENCE: '
AOP-6B, Rev. 6, pg. 1 and LP2441, EO 2.8.
)
KA 000005A201 [3.3/4.1]
000005A201 ..(KA's)
,
.m.. - - . _ . - ,
-
- . , - - , . ,, . _ - , _ , , , 4
. _.._.m _ _ _ . . - . . . . . . . . . _ . _ , _ . _ . . . . . = _ . _- _ .
, i e -
i
< , REACTOR OPERATOR Page 67 i ! i r
ANSWER: 039 (1.00)
b.
REFERENCE: ,
ECA 0.1, Rev. 10, pg. 12 j
KA 000015G007 [3.1/3.2]
000015G007 ..(KA's)
.
ANSWER: 040 (1.00)
c.
REFERENCE:
EOP-0.1, Rev. 14, pg. 7.
KA 000024A205 [3.3/3.9)
000024A205 ..(KA's)
ANSWER: 041 (1.00)
d.
l
I
REFERENCE:
1
EOP-2, Rev. 10, pg. 4. j
KA 000040A110 [4.1/4.11 !
:
i
000040A110 ..(KA's) l
l
1
!
!
.
.._....-.~u..
!
REACTOR OPERATOR Page 68
,
:
'
ANSWER: 042 (1.00) !
'
h b. .l . .
l
REFERENCE: )
s' . :
1 l
- EOP-1, Rev. 19, Foldout page. ;
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,.Rev. 4, pg. 1. ;
KA 000051A202 (3.9/4.1) I
000051A202 ..(KA's)
. _ _ _ _ . . _ . . . . . _ . . _ _ . . - . . . _ . _ . . _ _ _ . _ _ . . _ . . _ . _ . - _ _ _ . _ _ _ _ . _ . _ . . . _ _ _ . _ _ . _ _ . .
.
:
1 ! . ' REACTOR OPERATOR Page-69 ' ? 1 !
.,
2
i
- ANSWER
- 045 (1.00)
1 i a. i4 !~ l i j 1 REFERENCE:
, .
LP 0462, Rev. 6, pg. 7 and LO 2.3.
000055K302 [4.3/4.61
000055K302 ..(KA's)
AMSWER: 046 (1.00)
c.
. REFERENCE:
BG ECA-0.0, Rev. 14, pg. 14 and LP 0462, LO 2.2.
KA 000055A106 [4.1/4.5]
000055A106 ..(KA's)
ANSWER: 047 (1.00)
!
i
d.
REFERENCE:
TR11B 12. 9, Rev. 2, pg. 11.
000057A218 [3.1/3.1]
l
000057A218 ..(KA's)
I
I
l
REACTOR OPERATOR Page 70 j ANSWER: 048 (1.00)
d. I
REFERENCE:
3 . TRHB 11.14, Rev. 7, pg. 26. (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. 15.3.6-7. KA 000069G003 [3.3/3.9) 000069G003 ..(KA's)
ANSWER: 050 (1.00)
c.
,
REFERENCE:
CSP-C.1, Rev. 12, pg. 20. 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 2.16. 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 2.15. 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 2.3.
KA 000008A101 (4.2/4.0)
!
000008A101 ..(KA's)
;
REACTOR OPERATOR Page 72
-ANSWER: 054 ' ( 1. 0 0 )
9
REFERENCE:
s .
,
EOP-1.2, Rev. 13, Foldout Page-and LP 0435, LO 2.6.
1
KA 000009A234 [3.6/4.2]
000009A234 ..(KA's)
.
ANSWER: 055 (1.00)
d.
l
REFERENCE: !
OP-5A, Rev. 28, pg. 1 and THRB 10.6, Rev. 3, pg. 7.
KA 000022K307 [3.0/3.2]
!
~
000022K307 ..(KA's)
ANSWER: 056 (1.00) )
a. \
I
REFERENCE:
SEP-1.1, Rev.-0, pg. 3.
KA 000025K101 [3.9/4.3]
!
000025K101 ..(KA's)
REACTOR OPERATOR Page 73
i
i
!
i
' ANSWER: 057 (1.00) j
e.
.
REFERENCE:
'
1 i
EOP-3, Rev. 19, pg. 32. !
KA 000032A101 [3.1/3.4] 1
:
000032A101 ..(KA's) ;
ANSWER: 058 (1.00) '
i
a.
REFERENCE:
AOP-1A, Rev. 8, pg. 5.
KA.000037G011 [3.9/4.1)
;
000037G011 ..(KA's)
s
ANSWER: 059 (1.00)
b.
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)
c. 1
REFERENCE:
s . BG CSP-H.5, Rev. 5, pg. 4 and LP 1998, EO 2.6. ) KA 000054K102 [3.6/4.2] l 000054K102 ..(KA's) l
ANSWER: 061 (1.00)
b.
i
l
REFERENCE:
1
BG CSP-H.1, Rev. 10, pg. 6. i
KA 000054K304 (4.4/4.6]
000054K304 ..(KA's)
I
ANSWER: 062 (1.00)
a.
REFERENCE:
AOP-0.0, Rev. 8, pg. 2. 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 18.
KA 000028K202 [2.6/2.7) ;
000028K202 . . ( YA ' s )
i
ANSWER: b64 (1.00) j
c.
l
REFERENCE:
EOP-0.1, Rev. 14, pg. 14.
KA 000056K101 [3.7/4.2]
000056K101 ..(KA's)
I
1
ANSWER: 065 (1.00)
b.
REFERENCE:
NP 1.1.4, Rev. 1, pg. 3.
KA 194001A101 [3.3/3.4]
194001A101 ..(KA's) l
l
l
1
!
,__ _ . . . _ _ _ _ . . _ _ _ . _ ~ , . _ . _ _ . . _ . . . . . _ _ . ~ . _ _ . _ . ~ . . _ . - _ _ . _ _ . _ _ _ _ _ . .
\
!.
REACTOR OPERATOR Page 76
,
ANSWER: 066- (1.00)
d.
t
REFERENCE:
\ -
=EOP-0, Rev. 20, pg. 1 and SEP-2, Rev. 2, pg. 1.
KA 194001A102 [4.1/3.9]
194001A102 ..(XA's)
ANSWER: b67 (1.00)
C.
REFERENCE:
10 CFR 55.53 and OM 3.10, Rev. 5, pg. 18.
KA 194001A103 [2.5/3.4]
194001A103 ..(KA's)
ANSWER: 068 (1.00)
c.
REFERENCE: !
NP-1.7.5, Rev. 3, pg. 5. i
KA 194001A103 [2.5/3.4) j
194001A103 ..(KA's)
l
. _ _ _ . _ _ _ _ . _ _ . _ . _ . _ _ _ . . _ . _ _ . . . . _ _ . _ . _ _ . . _ . _ . _ . . . . . . _ _ . . - . - _ . . . . .
-
. . _ _ _ . . , -
,
,
REACTOR OPERATOR Page 77
i , 4
ANSWER: 069- (1.00) i
i
- c. .
- i
'
! REFERENCE: 4 3 . l NP 1.7.10, Rev. 1, pg. 1. 2
KA 194001A105 [3.6/3.8) >
i
.
1 , , 194001A105 ..(KA's)
'
i b i e i ANSWER: .
070 (1.00)
4 1 b. !
REFERENCE: '
l
4
Manager's Expectations, 93-1.
l
KA 194001A103 [2.5/3.4] l
1 i i , j 194001A103 ..(KA's) I
- l
- ,
i 4 l ANSWER: 071 (1.00) .
d.
>
REFERENCE:
l 'OM 3.10, Rev. 5, pg. 2.
KA 294001A110 [2.9/3.9]
194001A110 ..(KA's)
1
1
_ _ _ . . _ _ . . _ - _ . . _ . . . _ . _ . . . . _ _ _ _ _ _ _ . . _ . . . , _ . . _ . _ . _ _ _ . . _ _ . _ . _ . _ _ _ _ . . _ . _ _ _ _ _ _ . _ _ _ _ _
,
i. 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, Rev. 1, pg. 2. ! l j KA 194001A103 [2.5/3.4] l 4
l 194001A103 ..(KA's)
s 4
ANSWER: 074 (1.00)
b.
REFERENCE:
OM 3.17, Rev. O, pgs. 4 & 5.
KA 194001K101- [3.6/3.7)
l
194001K101 ..(KA's) ,
:
l
I
1
REACTOR OPERATOR Page 79
'
ANSWER: 075 (1.00)
C.
l
.
REFERENCE: )
-
%.
l
NP 1.9.15, Rev. 2, pg. 14.
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
5.5
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. 15.3.4.
KA 061000G005 (3.3/4.0)
061000G005 ..(KA's)
'i
REACTOR OPERATOR Page 80 ANSWER: 078 (1.00)
c.
REFERENCE:
s .
,
TRHB 13.8, Rev. 1, pg. 3,7 and C0105 COT.
KA 001000K403 [3.5/3.8)
001000K403 ..(KA's)
ANSWER: 079 (1.00) 'C. REFERENCE:
TRHB 10.5, Rev. 4, pg. 10.
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 COT.
KA 003000A205 [2.5/2.8)
.003000A205 ..(KA's)
_ _ .
l REACTOR OPERATOR Page 81
'
ANSWER: 081 (1.00)
d.
REFERENCE:
s .
OP-3A, Rev. 37, pg. 2.
KA 004000K601 [3.1/3.3]
004000K601 ..(KA's)
ANSWER: 082 (1.00)
a.
REFERENCE:
TRHB 13.12, Rev. 3, pg. 15 and C5005 COT.
KA 072000A202 (2.8/2.9]
t
072000A202 ..(KA's)
ANSWER: 083 (1.00)
'
a.
,
REFERENCE:
TRHB 13.13, Rev. 1, pg. 2.
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, Rev. 4, 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. 2.
KA 039000K101 [3.1/3.2]
039000K101 ..(KA's)
ANSWER: 086 (1.00)
b.
t ' REFERENCE:
l
TRHB:12.7,.Rev. 4, 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
; Q.
REFERENCE:
<
\ :
! TRHB 13.9, Rev. 2, pg. 4. . l KA 075000A203 [2.5/2.7) !
,
t 075000A203 ..(KA's) ! i'
:
1 l ANSWER: b88 (1.00) , i c. i a
- REFERENCE:
} 1 OP-4C, Rev. 13, pg. 1. 1
KA 007000G007 [2.9/3.1)
l 007000G007 ..(KA's) 3 i .
ANSWER: 089 (1.00)
d.
REFERENCE:
TRHB 10.9, Rev. 3, 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- s. . '
TRHB.10.1, Rev. 4, pg. 18.
! KA 027000K501 [3.1/3.4)
i j 027000K501 ..(KA's) 1.
i
! 4
ANSWER: 091 (1.00)
i. q
b. li
j. '
' REFERENCE:
.
[
J TRHB 10.11, Rev. 3, pg. 3. 4 (~ J
KA 028000K502 [3.4/3.9)
j ! 028000K502 ..(KA's) j
- 1
1
., 4 ,
ANSWER: 092 (1.00)
- b. i
i
REFERENCE:
TRHB 10.13, Rev. 1, pg. 6.
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, Rev. 3, pg. 2.
[EBQ 052-06-005A]
1 I
'
f
KA 078000K302 [3.4/3.6]
'
,
u
' '
078000K302 ..(KA's) I
1
,
i i i ANSWER: 0
094 (1.00) j
l
a.
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 Document. !
l
' i
'
l KA 000027A101 [4.0/3.9) i { 000027A101 ..(KA's)
- i
. .
l ANSWER: t
095 (1.00)
i a.
'
i { REFERENCE: ,
AOP-6A, Rev. 7, pg. 10; AOP-6A BG pg 16.
1 i
104 000003A102 [3.6/3.4)
3
000003A102 ..(KA's) , 4
i $ . - -
.-. _- ,
REACTOR OPERATOR. Page 86
ANSWER: 096 (1.00)
c
, REFERENCE:
- . '.. *
1 LP 0435, Rev. 7, pg. 11. I.
KA 000011K101 [4.1/4.4]
1 3 j. 000011K101 ..(KA's) i 4- Y 7
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)
d.
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, 1
. REACTOR OPERATOR Page 87 ,
'
l
4
4
l' ANSWER: e
099 (1.00)
I
b.
^ I' . ,
REFERENCE:
s .
AOP-5B,~Rev. 10, pg. 2.
l . l 4
KA 000065A201 [2.9/3.2] !
i
-
.
1
- 000065A201 . ,KA's)
ANSWER: kOO (1.00) *
4
'd.
! 1 4 4
REFERENCE:
] NP 1.7.1, Rev. 1, pg. 3. ( KA 194001K105 [3.1/3.4] ,
- - l
l
:
194001K105 ..(KA's)
1 i 4-
i
l
!
1 i
!
,
(********** END OF EXAMINATION **********)
- , - - - - _ . . _ . - .- . _ . -_ _ -
e j U. S. NUCLEAR REGULATORY COMMISSION
SITE SPECIFIC EXAMINATION
SENIOR OPERATOR LICENSE
i REGION 3 "
MASTER EXAMINATION.
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 sheets. 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
4
examination starts.
}i CANDIDATE'S l TEST VALUE SCORE %
i
1 l
99fa
)qht_100.u0 % TOTALS
1
FINAL GRADE
!. .1 j All work done on this examination is my own. I have neither given nor
i
,
received aid.
! I i Candidate's Signature
1
3
i i
k l' 1 3
.
l
.
PAGES 2, 3 AND 4 WERE USED AS ANSWER SHEETS
AND ARE NOT INCLUDED IN THIS MASTER EXAMINATION COPY.
!
,
.
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 examination.
3. Restroom trips are to be limited and only one applicant at a time may
leave. You must avoid all contacts with anyone outside the examination
room to avoid even the appearance or possibility of cheating.
4. Use black ink or dark pencil ONLY to facilitate legible reproductions.
5. Print your name in the blank provided in the upper right-hand corner of
the examination cover sheet and each answer sheet.
6. Mark your answers on the answer sheet provided. USE ONLY 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 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 answer. 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 questions. 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 points. 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.
3 t e
_ _ _ _ _ . _ _ . _ _ _ _ . _ _ _ _ _ _ ._. __
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.
i
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
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.
1 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. , b. 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 suction.
b. Pump the hotwell to the CST using a condensate pump
c. Align AFW pump suction to the fire system.
d. Connect a fire hose to the CST drain valve and backfill the CST,
! l
i
*
l. . I i . _ . - . -
. . . ,
1
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-tubes.
c. To reduce the mass inventory available to boil off in the event
of a steam break.
d. To prevent low power level oscillation due to level dominant
control.
QUESTION: 004 (1.00)
Given the following Unit 2 plant conditions:
- The plant has just been manually runback by 30% power within the
last minute.
- All systems are operable and in automatic.
- Actual Tavg is 9 degrees F higher than Tref.
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
d. 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 pressure.
c. SI on low PZR pressure.
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 pumps.
b. A CCW high radiation monitor alarm is received.
c. CCW pump discharge low pressure alarm is received.
d. Unable to maintain CCW Surge Tank level.
.
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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. 2
- Maintenance is inspecting RCP 1A seals.
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.
- c. Seal ~ injection is maintained during this condition.
'
d. Seal leakoff collects any RCS leakage up the shaft and directs {
it back to the VCT.
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QUESTION: 009~ (1.00)
Which ONE of the following is the ALTERNATE power supply for the 2 P2A
Charging Pump?
a. B08 i
b. B09
c. 1B03
d. 2B03
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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
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d. Trip of main feedwater pumps !
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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 only. I
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b. Low pressurizer pressure and steamline low pressure only. !
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c. Low pressurizer pressure, steamline low pressure, and !
containment high pressure only. l
l 1 l d. Low pressurizer pressure, steamline low pressure, containment ! j high pressure, and manual. ! ! !
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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
*
! !.
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_ . - - . _- -. -..-. . - - . . . . ~ . . ~ . _ . . . _ . . ~ - - . - . - _ - . _ . . _ . . . . . - - . _ . . ~ . . . - -
a
/!&
fY
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10 4 1
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10 5 Y
i 1 G ELECTRICALLY AOJUSTED COMPENSATED 13N CHAMBER -
$
w
104
I Q.
k TYPICAL SHUTDOWN CURVES
m
J
C 10 7
4
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o
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w
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i E i ' e 104
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0
.
.
2 4 .
.
6 8 10 12 14
s
16 18 20 22 24
, 1 TIME
* AFTER REACTOR SHUTDOWN (MINUTES) l
- l
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l 4 1 . i TYPICAL GAMMACOMPENSATED CURVE FOR A COMPENSATED lON CHAMBER ! 1
-
1
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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
startup.
c. To allow a spare detector to be installed in the upper half of
the instrument should the installed detector fail.
d. This position provides optimal cooling for the detector by
natural circulation of air in the detector well.
1
QUESTION: 014 (1.00)
Given the following Unit 2 plant conditions:
- Unit is at 90% power.
- Rod control in automatic.
- 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
Tref.
d. Rods initially drive OUT and then STOP and remain at that
position.
.
SENIOR REACTOR OPERATOR Page 13 QUESTION: .015 (1.00)
Given the following Unit 1 plant conditions:
'
- Reactor is shutdown.
, Reactor-decay heat is being removed by natural circulation.
- RCS pressure is 1550 psig.
- Average core thermocouple temperature is 402 degrees F.
Which ONE of the following describes the approximate amount of
subcooling.that exists in the RCS?
a. 50 deg F.
b. 100 deg F.
c. 200 deg F.
d. 300 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 plate.
b. They are arranged in strings extending from the upper support
plate to the core mid-plane.
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.
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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 head.
c. Catastrophic failure of the reactor vessel flange o-rings.
d. Damage to'the Core Exit Thermocouples's electrical circuitry
from overheating. ,
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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 psig. ;
b. Containment pressure at 5.2 psig.
c. Steam generator level at 73%.
d. Condenser hotwell level at 10 inches.
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i 1 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?
j. -a Momentarily depressing the feedwater control valve bypass reset i pushbutton (FWCV BYPASS) once the closure signal has been j cleared.
4
b. 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 valve. l
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l d. The condition which caused the isolation signal must be restored j to its pre-trip value.
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
c. Pressurizer Relief Tank
d. Reactor Coolant Drain Tank ;
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SENIOR REACTOR OPERATOR Page 16
QUESTION: 021 (1.00)
The high pressure sensing line on the RCS loop B flow instrument FT-414
breaks.
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 indication.
c. All RCS loop B flow instruments are affected (PT-414, FT-415,
FT-416) with HIGH flow indication.
d. All RCS loop B flow instruments are affected (PT-414, FT-415,
FT-416) with LOW flow indication.
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 power.
b. The component is in an abnormal alignment.
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c. The component has lost DC control power. I
d. The component is running / energized.
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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 feet.
b. Pressure is 730 psig.
c. Boric acid concentration is 1910 ppm.
1
d. Outlet Isolation valve is OPEN and the control switch RED
position indicating light is NOT LIT.
QUESTION: 024 (1.00)
Given the following Unit 1 plant conditions:
-
Pressurizer pressure defeat switch is in its normal position.
-
PT-449 (Yellow Channel), Pressurizer Pressure, has just failed
LOW.
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 operation.
.
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SENIOR REACTOR OPERATOR Page 18
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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 only.
b. Pressurizer PORV PCV-430 opens and all pressurizer heaters
deenergize only.
,
c. Pressurizer PORV PCV-430 opens and both pressurizer spray valves
open only.
d. 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% power.
'
- Rod control is in MANUAL.
- Loop B cold leg temperature detector TE-401B fails high.
, - No operator action is taken. 1 t l Mhich ONE of the following will be the steady-state pressurizer level? l
a. 20%
b. 33%
c. 46%
d. 100%
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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-428.
b. Pressurizer liquid temperature increases.
c. The reference leg for LT-428 cools down due to a decrease in
containment temperature.
d. Containment pressure increases to 0.3 psig; containment
temperature remains constant.
!
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 swell.
c. Allows ti for reactor power to decrease below 40% before
actua n.
llows time for the AFW pumps to recover S/G levels before
actuation.
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 running.
b. Narrow Range reading 38 ft. with NO RCPs running.
c. Wide Range reading 65 ft. with ONE RCP running.
d. Narrow Range reading 95 ft. with BOTH RCPs running.
.
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SENIOR. REACTOR OPERATOR Page 20
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QUESTION: 030 (1.00)
Given the following Unit 1 plant conditions * '
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- Unit is operating at 75% steady state power.
- All systems are in automatic control.
- The."A" S/G atmospheric valve fails open.
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
.
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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
c. 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 b. 1 hour. j c. 24 hours. , d. 2000 hours. 4
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QUESTION: 034 (1.00) '
Which ONE of the following RMS displays must be selected to access alert
and alarm setpoints? )
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a. Status Grid
4
b. Sector Display
! l c. HDSR
d. Trend
4
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SENIOR REACTOR OPERATOR Page 22- QUESTION: 035 .(1.00)
Given the following Unit 2 plant conditions:
- RHR is in service. !
- RCS pressure is 320 psig and INCREASING. l
- RCS temperature is 340 degrees F and INCREASING.
- ALL system lineups are in a normal shutdown configuration for
solid plant operation.
Which ONE of the following will act FIRST to prevent overpressurizing
the RHR System?
a. Pressurizer PORVs will open.
b. RHR return isolation valve will auto close,
c. RHR pump hot leg relief valve RH-PCV-861C will open.
d. RHR pumps discharge relief valve RH-PCV-861A will open.
1
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 seconds.
c. Going to TEST on the individual valves on the back of C03.
d. Reaching 90 psig in the crossover steam header. !
.
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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
location.
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 thermocouples. !
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 occurred.
- Power has been restored.
- The crew has transitioned to ECA 0.1.
- ECA 0.1 directs restoration of RCP seal cooling.
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 established.
b. CCW flow is initiated to cool the seals to less than 150 degrees l
F, then seal injection flow is established.
c. Seal injection flow is initiated to cool the seals to less than
190 degrees F, then CCW flow to the RCP is established.
d. CCW flow is initiated to cool the seals to less than 190 degrees
F, then seal injection flow is established. !
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SENIOR REACTOR OPERATOR Page 24 QUESTION: 039 (1.00)
Given the following Unit 1 plant conditions:
- A plant trip has just occurred.
- 2 control rods are stuck out of the core following the trip.
- 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
b. 1,200 gallons
c. 2,400 gallons
d. 3,000 gallons
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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
containment. ;
- The crew is performing actions of EOP-2, " Faulted Steam Generator '
Isolation."
- The AFW system is in operation.
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/Gs. i
d. Isolate the AFW Pumps from S/G A (steam and AFW flow).
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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% power.
- Condenser vacuum is decreasing.
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
b. 23" Hg vacuum
c. 21" Hg vacuum
I
d. 19" Hg vacuum
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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 unrestricted. r
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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
1
'
b. An unmonitored release of radioactivity
!
li c. Loss of containment integrity i
d. A Steam Generator overpressurization condition
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POINT BEACH NUCLEAR. PLANT AOP-5A
} ABNORMAL OPERATING PROCEDURES MAJOR i Revision 4 . LOSS OF CONDENSER-VACUUM August 21,1995 !. f .
%
g.
s
FIGURE 1
*
1 . ./ 4 ! i i
RECOMMENDED OPERATIONAL REGIONS FOR NUCLEA
a
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REFERENCE USE
. .
- SENIOR REACTOR OPERATOR. 'Page 27
QUESTION: 045 (1.00)
Given the following Unit 1 plant conditions:
- A loss of all AC power has occurred.
- ECA-0.0, " Loss of All AC Power" is in effect.
- Per ECA-0.0, certain Engineered Safeguards equipment control
switches are placed in Pull-out.
'Which ONE of the following events is prevented by this switch alignment?
a. An uncontrolled cooldown of the RCS and possible reactor
restart.
b. An uncontrolled use of water that may be needed for long term
cooldown.
c. Ar uncontrolled start of large loads on safeguards AC buses.
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
control.
b..May fail open; but manual control will be available to regain
control.
c. May fail closed; manual control will NOT be available to regain
control.
-d. May fail closed; but manual control will be available to regain
control.
.
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 +3.0
psig and -2.0 psig during power operation.
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
b. 1 hour
c. 6 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
a. I, II and III
b. I, II and IV
'
c. I, III and IV.
d. II, III, and IV.
QUESTION: 050 (1.00)
Given the following Unit 1 plant conditions:
- The Reactor tripped from 100% power.
- EOP-0, " Reactor Trip or Safety Injection" has been entered.
- 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 valves.
b. Locally trip the main turbine.
c. Go to CSP-S.1, " Response to Nuclear Power Generation /ATWS."
d. Verify Safeguards Buses energized.
l l 4 i
. - . . -
_ . . . ._ _ _ _ . = . _ _ . . . _ _ _ _ _ . _ _ _ _ , , _ _ _ - . . _ _ _ _ _ . _ . - _ _ . _ . . . . . . .
_
. . . _ . . . . _ _ _
} i j SENIOR REACTOR OPERATOR Page 30 1 ! !
- 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)
- tubesheets. ~
j b. To prevent excessive cooldown of the RCS.
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
1
- A reactor trip and safety injection has occurred.
l
- Pressurizer ~PORV 1RC-431C lifted and is stuck open. 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
1
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 LOCA.
- Conditions have stabilized and operators are evaluating the
criteria for terminating SI.
- Adverse Containment conditions do NOT exist.
- A maximum of 50 gpm feedwater flow is available to each S/G.
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% NR.
b. Pressurizer level indicates 12%.
c. RCS subcooling is 40 degrees F.
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 RWST. ;
1
b. Reduce thermal shock on the Non-Regenerative Heat Exchanger. l
i
'
c. VCT level will decrease causing possible damage to the charging
pumps.
d. Reduce thermal shock on the charging penetration into the RCS. ,
.
. . .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 progress.
- The RCS is solid.
- RHR flow is lost and CANNOT be restored. '
- All other systems and components are available.
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 valve.
b. Start a charging pump, with flow through an RHR heat exchanger,
and initiate Hot Leg Injection.
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.
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 rupture.
- 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.
- Source Range Instruments are NOT energized.
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 instruments.
b. Depress both SR RESET pushbuttons.
c. Depress both IR BLOCK pushbuttons.
d. Deenergize N36.
..
. , .
, . .- . . . 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 operations.
_
4 a. Pressurizer level is at 11% and decreasing.
l .
b. Narrow range S/G 1evel is 62% and increasing.
c. VCT level is at 17% in AUTO control and decreasing.
j d. S/G Blowdown Isolation valve closes due to high radiation. 4
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 psig.
- 1B S/G pressure is 1050 psig.
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.19.
~
STEAM GENERATOR TUBE * RUPTURE- j
/, October;26, 1995..
UNIT-1 ,
STEP
ACTION / EXPECTED RESPONSE -4.
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
.
20
Initiate RCS Cooldown
a. Using ruptured steam
generator pressure, detennine
target 2 core ~ exit temperature.
*
__
_ _ _ _ _
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 ,
'
.
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. Step 20 Continued on Next Page . . ,
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_
O
33
'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 ring.
b. To prevent reactor restart from an excessive cooldown.
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"?
a. It establishes natural circulation to enhance the bleed and feed )
capability of safety injection. I
i
b. It extends the time available to restore feed flow before bleed
and feed criteria is met.
I
c. It anticipates an RCS pressure decrease caused by spray valves !'
opening when air is restored to containment.
d. It anticipates an RCS pressure decrease caused by opening PZR
PORVs during bleed and feed.
.
. _ _ _ _ _ _ . . . . _ - .. . ._
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
_
,
b. D16
.
c. D18
d. D21
s a
QUESTION: 062 (1.00)
4
,
With the Pressurizer Level Control Selector Switch in the NORMAL
position, a pressurizer level instrument failure causes the following l
SEQUENTIAL plant events.
- Charging flow is reduced to minimum.
- Pressurizer level decreases.
- 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)
a. Pressurizer level channel 428 (Blue) failed low
b. Pressurizer level channel 428 (Blue) failed high
c. Pressurizer level channel 427 (White) failed low. I
d. Pressurizer level channel 427 (White) failed high.
..
- - - .
. _ _ ._._.. _ . . .__ _. . . - - - _ _ _ . . . . _ _
]
!
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
power.
- 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
b. Information Use
c. Reference Use
d. Continuous Use
l l l .
* ? 1
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 F.
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
used.
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 applicable.
d. 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
d. 7 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
1
QUESTION: 068 (1.00) j
1
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 operator.
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 room.
b. Immediately notify DCS by two-way radio and then leave the D/G
room.
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,
t
.
.
l
l
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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 Control.
b. Turbine manual using governor valve decrease,
c. Valve Position Limiter.
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
d. 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 parameter.
a. Channel Calibration
b. Channel Functional Test
c. Channel Check
d. Channel Verification
!
!
!
1
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
b. 4 hours
c. 8 hours
d. 12 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 Operator.
'
b. Must be done by visually observing the operating link.
c. Can be accomplished using status lights.
d. Can be accomplished using remote indicators.
,
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? 2
a. A QC certified inspector is present.
l
l
b. Work is1to be performed in a high radiation area. l
c. Requires direct line of sight of the isolation. l
;
d. Isolation boundaries are locked. j
!
! .
QUESTION: 075 (1.00)
;
Given the following Unit 1 conditions:
- Unit is at 100% power. l
- VCT level is 23%.
- All controls are in automatic.
- LT-112, VCT level transmitter, fails high.
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 HUT. I
b. Increases to 28% where auto-makeup stops,
c. Decreases to 17% where auto-makeup initiates,
d. Decreases to 0% (empty).
.
- - _ - _ ... _ . _ ._ . . - -. . _ . _ .
,
l
SENIOR REACTOR OPERATOR Page 42
,
QUESTION: 076 (1.00)
Given the following plant conditions:
- Both units are operating. 1
- The South condensate Storage Tank (CST) is isolated for repairs.
- The North CST is selected for Auxiliary Feedwater Pump suction at
both units.
Which ONE of the following is the MINIMUM volume required in the North
i CST?
a. 13,000 gallons
b. 20,000 gallons
c. 26,000 gallons
d. 30,000 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.
4 ) i
.
..-. - . . .
!
P
SENIOR REACTOR OPERATOR Page 43 QUESTION: 078 (1.00)
Given the following Unit 2 plant conditions: .
- Operating at 100% power.
- All systems are operable. 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 Exchanger.
b. A tube leak in the Seal Water Heat Exchanger.
c. A tube leak in the Non-Regenerative Heat Exchanger.
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 elevator.
i
b. The elevator will not rise if any Spent Fuel Pool area radiation
monitor alarms.
c. Operators verify the elevator basket is empty before raising the
elevator. l
d. The elevator will not rise if any fuel element is in it.
..
, , - -
.. _ _ __ _ _ _ _ _ . . _ _ _ .. _ ._
SENIOR REACTOR OPERATOR Page 44
i
l
QUESTION: 080 (1.00)'
,
l
Given the following Unit 2 plant conditions- l
1
- Reactor power is 1215 MWt
- RCS pressure is in the normal operating band.
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?
a. 1
b. 2
c. 3
d. 4
l
@
_ _ _ . . _ . . . __ _ . . . . _ _ . _. _ _ _ _ _ _ _ . _ _ _ . _ _ . _ . . _ _ _ _ _ . . . _
. i Figure 15.2.1-2
'
REACTOR CORE SAFETY LIMITS
- . .
POINT BEACH UNIT 2
s
: 670
1 -
660 -
. 4
- 650
l 2425 psia !
- MO -
1
2250 psia
. 630 - <
Q
h
3 '
3 620 -
u
,
y 2000 psia
! i
b610
o
-
i
j 600 -
1775 psia
j 590 - 1
- 580 -
1
,
i
i
- 570 -
? . !, ' ' ' ' ' ' ' ' '
560 ' ' '
'
.
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2
.
Core Power (fraction of 1518.5 MWt)
; s.
~
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% power.
- Control rods are in AUTOMATIC.
Which ONE of the following instrument malfunctions would result in a
CONTINUOUS rod withdrawal?
a. Loop "A" Thot RTD fails LOW.
.
b. Power range channel N-42 fails LOW.
c. PT-485 fails HIGH.
i
'
d. Loop "B" Tcold RTD fails HIGH.
> l
QUESTION: 083 (1.00)
Given the following Unit 1 plant conditions:
1
- A Control Bank D rod was dropped and recovered.
- The Pulse to Analog Converter was NOT reset as required by
- AOP-6A, Dropped Rod. 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.
3
c. If control rods are inserted, Bank C control rods will begin
insertion at a lower value of Control Bank D position.
d. If control rods are inserted, Bank C control rods will begin
insertion at a higher value of Control Bank D position.
.
1 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 step.
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
subcooling.
b. To prevent RCP damage from cavitation due to operation in a two
phase system.
c. To minimize cooldown rate if a main steam line break is in
progress.
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
b. 60 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 Superintendent.
b. Duty Operating Supervisor,
c. Duty and Call Superintendent.
d. Duty Technical Advisor.
,
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
4 i
Page 48
] SENIOR REACTOR OPERATOR 5
-
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. 4
. i ! QUESTION: 089 (1.00) l The following boration flowpaths are listed in AOP-6E, " Alternate >
Boration/ Loss of Shutdown Margin."
I. CVCS and RWST.
II. CVCS and Blender.
III. 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?
a. II, I, III
!
b. II, III, I
c. III, I, II
.d. III, II, I
___ . - . _ . _ . _ . . _ . _ . . - . .._-.--.___.._.______-,_..._.___m. . _ _ _ . _ . _
SENIOR REACTOR OPERATOR Page 49
QUESTION: 090 (1.00)
Given the following containment history with a small LOCA in progress:
Tima Cnmt Temp. 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
- -
b. 0830 )
c. 0845
i
d. 0900 '
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 rate.
1
, c. Prevent oxygen inleakage.
\
, 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
'*
.b. 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
minimized.
b. Boron dilution of the RCS is minimized,
c. Adverse secondary system chemistry conditions are eliminated.
d. RCS cooldown rate is maximized.
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?
a. 2
b. 3
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 detected. i
1
c. During the performance of EOP-1, " Loss of Reactor Or Secondary
Coolant," when a RED path is detected in Heat Sink.
d. After transition to EOP-0.1, " Reactor Trip Recovery", following
an inadvertent reactor trip.
1
l
1
1
QUESTION: 096 (1.00)
Given the following Unit 1 plant conditions:
- An event occurred at 0800 that was classified as a Site Area
Emergency.
- The plant evacuation alarm was actuated at the time of the
classification.
'Which ONE of the following is the MAXIMUM time by which accountability ;
must be completed? !
a. 0815
b. 0830
c. 0845
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. 3
- Both-loop cold leg temperatures are 295 degrees F.
- 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 plant. 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
;
.- -. - - -. ..._- . . _ ... ._ - . - - _. _ __ . . . . - -
. _
-
4
ST-4
INTEGRITY gey;3o$
$
02-24-92
, .
CSP-P.1
TEMPERATURES
IN BOTH
NO mmma GO TO
CSP P.1 ,
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-P.2
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 page.
1
Which one of the following conditions is identified by this tag?
,
l
,
a. The valve is out of its normal lineup position.
b. The valve is a boundary valve for a system containing caustic
fluid.
c. The valve is to be radiographed. !
d. The valve has a contact dose rate of 325 mrem /hr.
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
+ !
!
1 j
!
i.__. _ ._ _ _ _
_ . __ __ _ . _ _ _ . . _ _ _ . __
SENIOR REACTOR OPERATOR Page 54
I
l
l
ANSWER: 001 (1.00) l
1
a. l
l
l
REFERENCE:
,
TRHB 10.1, Rev. 4, pg. 15 and C1805 COT and TRHB Figure 10.1.24
KA 022000G007 [3.3/3.5]
i 1
'
022000G007 ..(KA's)
1
ANSWER: 002 (1.00)
a.
REFERENCE:
EOP-0, Rev. 20. Appendix A, pg. 2.
KA 000040A110 (4.1/4.1)
000040A110 ..(KA's)
ANSWER: 003 (1.00)
c.
REFERENCE:
TRHB 13.7, Rev. 1, pg. 2.
KA 059000A302 [2.9/3.1)
059000A302 ..(KA's)
. .__ _ _ . _ _ _ __ . . _ . . _ . . _ . . ._.
SENIOR REACTOR OPERATOR Page 55
ANSWER: 004 (1.00)
c.
REFERENCE,: .
Logic Sheet 17 and Setpoint Document Section 5.2
KA 041020K105 [3.5/3.6] ,
I
.
041020K105 ..tKA's)
.
ANSWER: 005 (1.00)
a.
l
REFERENCE:
AOP-10A, Rev. 18, pg.. 12.
[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, Rev. 5, 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
1
,
ANSWER: 007 (1.00)
- d.
l ' REFERENCE:
J t -
1 AOP-9B, Rev. 7, pg. 4. ; l
'
KA 000026G010 [3.6/3.5] l
1
[Similar to question on 95 SRO exam.] '
000026G010 ..(KA's)
, i 1 $ AMSMER: 008 (1.00)
a. l
1
-
,
REFERENCE: 1
TRHB 10.2, Rev. 5, pg. 16.
l
W KA 003000K407 (3.2/3.4] l
003000K407 ..(KA's)
,
ANSWER: 009 (1.00) ]
b.
REFERENCE:
!
TRHB 10.6, Rev. 3, pg. 71 and Master Data Book for B09.
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 7.
KA 013000A302 -[4.1/4.21
013000A302 ..(KA's)
.
ANSWER: 011 (1.00)
b.
REFERENCE:
TRHB 13.4, Rev. 4, pg. 18 and C730 COT and Logic Sheet 7. KA 013000A402 [4.3/4.4] 013000A402 ..(KA's)
ANSWER: 012 (1.00)
d.
REFERENCE:
TRHB 13.1, Rev. 3, Figure 13.1.19 and C1315 COT.
KA 015000A202 [3.1/3.5)
015000A202 ..(KA's)
1
i
.. - - . __ _ _ _ _ . __ _ _ . - -
'
SENIOR REACTOR OPERATOR Page 58
i
l ANSWER: 013 (1.00)
b.
l
REFERENCE:
,
TRHB 13.1, Rev. 3, pg. 11 and C1310 COT.
KA 015000K601 (2.9/3.2)
015000K601 ..(KA's)
i
l
-
ANSWER: 014 (1.00)
i
b. !
l
'
REFERENCE: l
TRHB 13.8, Rev. 1, pg. 6 and C0105 COT; and TRHB 13.8, pg. 19 and Logic
Sheet 16.
i
! KA 015000K302 (3.3/3.51
015000K302 ..(KA's)
i ANSWER: 015 (1.00)
c. REFERENCE: Steam Tables
1
KA 017020K502 (3.7/4.0]
017020K502 ..(KA's)
r !
. _ _ _ _ - - . . . _ _ _ _ _ _ _ _ . . _ . . . _ _ _ _ _ _ _ _ _ . _ . . __.._.-__.____....____._m. . _ _ _ _
i
- SENIOR REACTOR OPERATOR Page 59
' s l
ANSWER: 016 (1.00)
'
.c.
REFERENCE:
'
:
TRHB 13.2, Rev. 3., pg. 8 and TRHB Figure 10.4.1.
KA 017020K102 [3.3/3.5]
017020K102 ..(KA's)
. ANSWER: 017 (1.00) l
b. )
i
l
REFERENCE:
BG EOP-0.2, Rev. 13, pg. 7 and C7033 COT.
KA 022000K404 [2.8/3.1]
022000K404 ..(KA's)
ANSWER: 018 (1.00)
b.
REFERENCE:
TRHB 11.2, Rev. 5, pg. 11 and C3912 COT and Logic Sheet 24
KA 056000A204 [2.6/2.8)
056000A204 ..(KA's)
l
l
1
i
. -
r
SENIOR REACTOR OPERATOR Page 60
ANSWER: 019 (1.00)
l a. t .
!
l
. REFERENCE: I
l ~. :
TRHB 13.7, Rev. 1, 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
d.
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
-
1 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, Rev. 4, pg. 18 and Technical Specification 15.3.3.
KA 006000G005 [3.5/4.2]
006000G005 ..(KA's) i
I
i
ANSWER: 024 (1.00) )
I a.
REFERENCE:
.!
TRHB 10.3, Rev. 3, Figure 10.3.7, Logic Sheet 18 and ICP 10.2
l
KA 010000A403. [4. 0/3. 8]
010000A403 ..(KA's)
!
_
;
SENIOR REACTOR OPERATOR Page 62
AN.9WER: 025 (1.00)
'a. .
;
REFERENCE:
, .
TRHB 13.5, Rev. 2, 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)
c.
1
i
REFERENCE: I
TRHB 13.6, Rev. 1, pg. 7 and C1006 COT and Logic Sheet 18
KA 011000K604 [3.1/3.1]
011000K604 ..(KA's)
1
ANSWER: 027 (1.00)
a.
I
' REFERENCE: ,
TRHB 13.6, Rev. 1, pg. 5.
KA.011000A101 [3.5/3.6]
011000A101 ..(KA's)
i
, - - - -
1
SENIOR REACTOR OPERATOR Page 63
ANS : 028 (1.00)
"'
kUd317Y
REFERENCE: ,
[ggggg l
TRHB 13.3, Rev. 4, pg. 37.
KA 012000K406 (3.2/3.5)
IC)Nf
012000K406 ..(KA's)
,
-ANSWER: 029 (1.00)
b. l
,
REFERENCE:
l
BG CSP-I.3, Rev. 7, pg. 1. l
i
KA 016000G015 (3.6/3.8) l
;
i
016000G015 ..(KA's)
,
ANSWER: 030 (1.00)
d.
i'
REFERENCE:
TRHB 13.8, pg. 20.
KA 035010k501 [3.4/3.9)
035010K501 ..(KA's)
. _ _ . .. _ ... .. . . _ _ _... _ _ . _ .4 _ - _ _ , _ _ _ . _ . _ _ _ _ _ . . . _ _ . . _ _ _ _ . . . . _ _ . _ _
4 l' 1 3
- SEMIOR REACTOR
- OPERATOR Page 64
j' 4 . t
- ANSWER
- 031 (1.00)
l'
- a.
i
REFERENCE:
, .
TRHB 12.5, Rev. 2, Figure 12.5.1 and Master Data Book
KA 062000K201 [3.3/3.4]
062000K201 . .(KA's)
ANSWER: 032 (1.00) l
c.
REFERENCE:
TRHB 12.2, Rev. 1, pg. 38, C4105AOT, OI-110, Rev. 1, and PC-29, Rev. 29
Monthly Best
KA 064000K402 (3 9/4.2] .
064000K402 . .(KA's)
!
I
i
ANSWER: 033 (1.00) 1
a.
REFERENCE:
l
TRHB 12.8, Rev. 4, pg. 74 and OP-11A, EDG G01/G02
l
KA 064000A206 [2.9/3.3]
l
064000A206 . .(KA's)
,
-
. . _ . .._..m..~ . . _._.__.___-__....__.m . _ _ _ , _ _ . _ . . _ . . _ . _ . _ . . _ _ . _ _ _ _ _ . . _ _ _
!
Page 65
'
SENIOR REACTOR OPERATOR
I
i
{. ANSWER: 034 (1.00) 4
d.
!
:
F ! REFERENCE: i * 1. I 'TRHB 13.12, Rev. 3, 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,.Rev. 7, . pg . 17, C0506AOT.and Setpoint' Document, Sections 2.3,
i 8.2 and 11.2. f KA 005000K109 [3.6/3.9] ! i 4 $
005000K109 ..(KA's)
a
1 s
ANSWER: 036 (1.00)
d.
<
REFERENCE:
!
'
TRHB 13.10, Rev. 2, pg. 2; 62804 COT; Setpoint Document 14.13; P&ID PBM-
241; OP-1c, Step 7.3.5
5
KA 041020K603 (2.7/2.9)
041020K603 ..(KA's)
I
. i ;
4 $
.~ . . . ~ . - . - . . . . - . . . . . . . . - - . - . - . - - . . . - . - - - - - . . - . .
. . - . . . . . .
!
l
! SENIOR REACTOR OPERATOR Page 66' '
.
i
- I
j'ANSUER: 037 (1.00)
-
. l
a.
1 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
c.
REFERENCE:
EOP-0.1, Rev. 14, pg. 7.
!
KA 000024A205 [3.3/3.9)
000024A205 ..(KA's)
:
,
SENIOR REACTOR OPERATOR Page 67 ANSWER: 040 (1.00)
d.
,
REFERENCE: ,
s .
EOP-2, Rev. 10, pg. 4.
KA 000040A110 [4.1/4.1)
000040A110 ..(KA's)
.
ANSWER: 041 (1.00)
b.
REFERENCE: .
EOP-1, Rev. 19, Foldout page.
KA 000040A205 [4.1/4.5]
000040A205 ..(KA's)
i
ANSWER: 042 (1.00)
, c. l
REFERENCE:
TRHB 13.9, Rev. 2, pg. 4 and Setpoint Document, Section 14.2.
KA 000051K301 [2.8/3.1]
000051K301 ..(KA'c) !
l
I
--. -...- . - .. . . - . _ - . ~ ~ . . . - ~ . - . ~ . - . - - _ _ - . - . . . . - ~ . . . . . -... _. _ - -
l: :
SENIOR REACTOR OPERATOR Page 67
:
;
ANSWER: 040 -(1.00) i
d.
REFERENCE:
s . ;
EOP-2,-Rev. 10, pg. 4.
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 page.
KA 000040A205 [4.1/4.5]
000040A205 ..(KA's)
i
ANSWER: 042 (1.00)
.c.
REFERENCE:
TRHB 13.9, Rev. 2, pg. 4 and Setpoint Document, Section 14.2.
KA 000051K301 [2.8/3.1)
000051K301 ..(KA's)
!- ! . !e ! l.
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. 1.
KA 000051A202 [3.9/4.1) i
000051A202 ..(KA's)
l
1
ANSWER: 044 (1.00)
l
a.
!
REFERENCE: i
!
LP 0462, Rev. 6, pg. 7 and LO 2.3.
000055K302 [4.3/4.6]
000055K302 ..(KA's)
ANSWER: 045 (1.00)
c.
REFERENCE: e
BG ECA-0.0, Rev. 14, pg. 14 and LP 0462, LO 2.2.
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 2
d.
? i
- REFERENCE
j $..- *
.
i TRHB 12.9,'Rev. 2, pg. 11. i
000057A218 [3.1/3.1]
!, j 000057A218 ..(KA's) L . ! j ANSWER: b47 ( 1. 0 0 )- . i d. 4 i.. REFERENCE: I
*
l 'TRHB 11.14, Rev. 7, pg. 26. 1 t '
1
i a [Similar to EBQ 052-01-065A] ! '
KA 000067K102 [3.1/3.9]
000067K102 ..(KA's) i
i
ANSWER: 048 (1.00)
!
b.
!
REFERENCE:
Tech. Spec. 15.3.6, pg. 15.3.6-7.
KA 000069G003 [3.3/3.9)
000069G003 ..(KA's)
SENIOR REACTOR OPERATOR Page 70
!
ANSWER: 049 (1.00)
c.
REFERENCE.
, .
CSP-C.1, Rev. 12, pg. 20. i
KA 000074K201 (3.6/3.9) ,
i
000074K201 ..(KA's)
ANSWER: 050 (1.00)
d.
REFERENCE:
EOP-0, Rev. 20, pg. 2 and LP 0405 LO 2.16.
KA 000007A202 [4. 3/4. 6)
000007A202 ..(KA's)
- ANSWER
- 051 (1.00)
b.
REFERENCE:
BG EOP-0, Rev. 21, pg. 9 and LP 0405, LO 2.15.
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 2.3.
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 2.6.
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, pg. 7.
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. 3.
KA 000025K101 (3.9/4.3]
000025K101 ..(KA's)
l
ANSWER: 056 (1.00)
a.
REFERENCE:
EOP-3, Rev. 19, pg. 32.
KA 000032A101 [3.1/3.4)
000032A101 ..(KA's)
l
ANSWER: 057 (1.00)
a.
REFERENCE:
AOP-1A, Rev. 8, pg. 5. KA 000037G011 [3.9/4.1) 000037G011 ..(KA's)
-SENIOR REACTOR OPERATOR Page 73
- ANSWER: - '058 (1.00)
b.
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, Rev. 5, pg. 4 and LP 1998, EO 2.6.
.KA 000054K102 [3.6/4.2]
000054K102 ..(KA's)
;
ANSWER: 060 (1.00)
b. !
REFERENCE:
BG CSP-H.1, Rev. 10, pg. 6.
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, Rev. 8, pg. 2.
KA 000058A203 [3.5/3.9)
000058A203 ..(KA's)
1
ANSWER: 062 (1.00) l
l
b. I
\
l
REFERENCE:
TRHB 13.6, Rev. 1, pg. 6 and Figure 13.6.1 and Logic Sheet 18. ]
KA 000028K202 (2.6/2.7)
000028K202 ..(KA's)
ANSWER: 063 (1.00)
C. ,
REFERENCE:
EOP-0.1, Rev. 14, pg. 14.
KA 000056K101 [3.7/4.2]
000056K101 ..(KA's)
. - -. . . .. . .- . .-...- . .-- - . . . ~ . . . . _ - - - - - SENIOR REACTOR-OPERATOR Page 75
, i
'
h 1' l' ANSWER: 064 (1.00)-
b.
1
REFERENCE:
'
: .
i
NP 1.1.4, Rev. 1, pg. 3.
\ '
KA.194001A101 [3.3/3.4] l
l
1
194001A101 ..(KA's)
ANSWER: 065 (1.00)
d.
.J
REFERENCE: l
l
EOP-0, Rev. 20, pg. 1 and SEP-2, Rev. 2, pg. 1. l
KA 194001A102 [4.1/3.9]
194001A102 ..(KA's)
f l'
ANSWER: 066 (1.00)
C.
REFERENCE:
10 CFR 55.53 and OM 3.10, Rev. 5, pg. 18.
KA 194001A103 [2.5/3.4]
194001A103 ..(KA's)
! , i
l l
l .. .
-. .
-_. _ . .
SENIOR REACTOR OPERATOR Page 76
i
.
ANSWER:' 067 (1.00)
c.
REFERENCE:
, l
NP-l'. 7. 5, Rev. 3, pg. 5.
KA 194001A103 (2.5/3.4]
194001A103 ..(KA's)
ANSWER: 068- (1.00)
c.
REFERENCE:
NP 1.7.10, Rev. 1, pg. 1.
KA 194001A105 [3.6/3.8]
194001A105 ..(KA's)
ANSWER: 069 (1.00)
.b.
REFERENCE:
Manager's Expectations, 93-1.
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)
d.
L
i
REFERENCE: '
y, .
OM 3.10, Rev. 5, pg. 2.
KA 194001A110 [2.9/3.9]
194001A110 ..(KA's)
ANSWER: 071 (1.00)
c.
.
REFERENCE: ,
-
i
Tech. Spec. Definition 15.1.f.
KA 194001A113 [4.3/4.1] ,
194001A113 ..(KA's) 1
l
ANSWER: 072 (1.00)
b.
REFERENCE:
NP'1.6.6, Rev. 1, pg. 2.
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, Rev. O, pgs. 4 & 5.
KA'194001K101 [3.6/3.7)
194001K101 ..(KA's)
ANSWER: 074 (1.00)
c.
REFERENCE:
NP 1.9.15, Rev. 2, pg. 14.
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
5.5
KA 004010A105 [3.0/3.2)
004010A105 ..(KA's)
SENIOR REACTOR OPERATOR Page 79 ANSWER: 076 (1.00)
c.
REFERENCE:
1 -
TRHB 11.4, Rev. 5, pg. 15 and Tech. Spec. 15.3.4.
KA 061000G005 [3.3/4.0]
061000G005 ..(KA's)
ANSWER: 077 (1.00)
c.
REFERENCE:
Tcch. Spec. Bases 15.3.6.A.1.c.
KA 029000K104 [3.0/3.1]
029000K104 ..(KA's)
ANSWER: 078 (1.00)
b.
REFERENCE:
TRHB 10.9, Rev. 3, Figure 10.9.1 and C0801AOT.
KA 008010A302 [3.0/3.1]
008010A302 ..(KA's)
l
b
SENIOR REACTOR OPERATOR Page 80
i
l'
ANSWER: 079 (1.00)
l
c.
l
l
REFERENCE: l
t *
l
RP-2D, Rev. 10, .pg. 6.
l
l
KA 034000K402 [2.5/3.3)
034000K402 ..(KA's)
ANSUER: 080' (1.00)
d.
REFEREFCE:
Tech. Spec. 15.2.1.
KA 002000G005 [3.6/4.1)
002000G005 ..(KA's)
ANSWER: 081 (1.00)
b.
REFERENCE:
. Tech. Spec. 15.3.3 Basis, pg. 15.3.3-9.
KA 026000A102 [3.6/3.9)
026000A102 ..(KA's)
.. . - _ _ . . . . . - _ _ _ . _ _ . . . . _ ~ . _ _ _ _ _ _ -.-_.__..~._.m._ ...___ - . _ _ - - - . -_
SENIOR REACTOR OPERATOR Page 81
!- i L iIANSWER: 082 (1 00) i ,
C.
2
- l
1 REFERENCE- 4 ,
s. .
,
TRHB 13.8, Rev. 1, pg. 10 and LP 2441, EO 2.4 and Logic. Sheet 16
! KA'000001K105 [3.5/3.8] 1 4 1
000001K105 ..(KA's)
~ i j ANSWER: 083 (1.00) 1 4 a. 3 i } REFERENCE:
.AOP-6A, Rev. 7, pg. 13, and TRHB 13.13, Rev. 1, pg. 4.
KA 000003A101 [2.9/2.9)
'000003A101 ..(KA's)
ANSWER. 084 (1.00)
d.
REFERENCE:
BG EOP-1, Rev. 18, pg. 3.
KA 000011K314 [4.1/4. 2 ]
000011K314 ..(KA's) '
. -
_ _ .. _ _ _ _ _ _
SEMIOR REACTOR OPERATOR Page 82
, ANSWER: 085 (1.00)
b.
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
1
I
} ANSWER: 086 (1.00)
b.
h
, REFERENCE:
, NP 1.9.14, Rev. O, pg. 8. 4
l
KA 000067G012 [3.4/3.4]
000067G012
'
..(KA's)
ANSUER: 087 (1.00)
d.
j REFERENCE: 2
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
1
l
SENIOR REACTOR OPERATOR Page 83 ANSWER: 088 (1.00)
C.
REFERENCE:
'.. *
Tcch. Spec. Bases 15.3.1.C.
KA 000076K305 (2.9/3.6]
[Similar question used on 94 RO exam.]
000076K305 ..(KA's)
ANSWER: 089 (1.00)
b.
REFERENCE:
AOP-6E, Rev. 5, pg. 1.
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) C. REFERENCE:
.
s .
OP-9D, Rev. 12, pg. 3.
KA 000060G007 [3.1/3.4]
000060G007 ..(KA's)
ANSWER: 092 (1.00)
a.
REFERENCE:
TRHB 13.2, Rev. 3, pg. 50, RMS Alarm Setpoint and Response Book
KA 000061A101 (3.6/3.6)
4
000061A101 ..(KA's)
.
ANSWER: 093 (1.00)
J
a.
4
REFERENCE:
LP 0441, Rev. 9, pg. 24 and LO 2.17, EOP-3.1 Background, Rev. 11, pg. 1.
KA 000038K306 (4.2/4.5]
000038K306 ..(KA's)
- . _ _ _ . . _ _ _ __
SENIOR REACTOR OPERATOR Page 85 ANSWER: 094 (1.00)
C.
REFERENCE: I
, .
3, pg. 2.
'
PBF-5101, Rev.
KA 000036G007 (3.2/3.5] i
000036G007 ..(KA's) l
ANSWER: 095 (1.00)
!
a.
!
REFERENCE:
EOP-0.0, Rev. 5, pg. 2. i
KA 194001A102 [4.1/3.9]
194001A102 ..(KA's)
ANSWER: 096 (1.00)
b.
REFERENCE:
EP 6.0, Rev. 35, pg. 8.
KA 194001A116 [3.1/4.4]
194001A116 ..(KA's)
. SENIOR REACTOR OPERATOR Page 86
ANSWER: 097 (1.00)
c.
REFERENCE:
' *
.
ST-4, Rev. 3.
KA 194001A108 [2.6/3.1]
194001A108 ..(KA's)
ANSWER: 098 (1.00)
d.
REFERENCE:
EP 2.0, Rev. 30, pg. 5.
KA 194001A116 [3.1/4.4]
194001A116 ..(KA's)
ANSWER: 099 (1.00)
d.
REFERENCE:
HP 3.2.9, Rev. 2, pg. 3. i
'
KA 194001K104 [3.3/3.5]
194001K104 ..(KA's)
. - - ~ . - _. - - . . . . . . . _ _ SENIOR REACTOR OPERATOR Page 87
d .
ANSWER: 100 (1.00) l
b.
'
REFERENCE:
,
NP 4.2.1.18, Rev. O, pg. 4.
KA 194001K104 [3.3/3.5]
'
l
1
l
194001K104 ..(KA's)
i
;
,
I
l
l
l
i
.:
(********** END OF EXAMINATION **********)
}}