ML20237B808
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| ML20237B808 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/17/1998 |
| From: | Curley V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-334-98-300OL, NUDOCS 9808190282 | |
| Download: ML20237B808 (1) | |
See also: IR 05000334/1998300
Text
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NOTE T0: NRC DOCUMENT CONTROL DESK'
MAIL STOP 0-5-D-24
FROM: Y8kS81 b* b"Al* 7 , LICENSING ASSISTANT
OPERATING LICENSING' BRANCH _ REGION I l
SUBJECT: OPERATOR LICENSING EXAMINATION ADMINISTERED GR-
Spail aa-a 4; 1999 an d Nau ifr se,9y, AT kkavea da//,y 7
U Ri l7 ] 'M7C UC
DOCKET NO. 5,!I3g i
fif tts L. h-24, If tt and
ON Mau ir s99r OPERATOR LICENSING EXAMINATIONS WERE ADMINISTERED
ATMEhEFkRENCEDFACILITY. ATTACHED YOU WILL FIND THE FOLLOWING
INFORMATION FOR PROCESSING THROUGH NUDOCS AND DISTRIBUTION TO THE
NRC STAFF, INCLUDING THE NRC PDR.
Item #1 a) FACILITY SUBMITTED OUTLINE AND INITIAL EXAM SUBMITTAL
DESIGNATED FOR DISTRIBUTION UNDER RIDS CODE A070.
b) AS GIVEN OPERATING EXAMINATION, DESIGNATED FOR DISTRIBUTION
UNDER RIDS CODE A070.
: Item #2- EXAMINATION REPORT WITH THE'AS'GIVEN WRITTEN EXAMINATION
ATTACHED, DESIGNATED"FOR DISTRIBUTION UNDER. RIDS CODE IE42 $
y.
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9808190282 990817
PDR ADOCK 05000334 >
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UNITEo STATES
[ ,.,, ' gg NUCLEAR REGULATORY COMMISSION
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REGloN I
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%- d KING oF PRUSSIA, PENNSYLVANIA 19406 1415
*****
June 3, 1998
Mr. J. E. Cross
President
Generation Group
Duquesne Light Company
Post Office Box 4
Shippingport, Pennsylvania 15077
SUBJECT:
BEAVER VALLEY UNIT 1 SENIOR REACTOR OPERATOR INITIAL
EXAMINATION REPORT NO. 50-334/98 300(OL)
Dear Mr. Cross:
This report transmits the findings of the senior reactor operator (SRO) licensing operating
examination, conducted by NRC examiners, during the week of April 20-24,1998 at the
Beaver Valley Unit 1 Nuclear Power Plant. The report also transmits the results of the
written portion of the examination, that was delayed until May 18,1998, as per your
request of March 13,1998. Based on the results, all three SRO applicants passed all
portions of the examination. At the conclusion, Mr. T. Kenny discussed the preliminary
findings with members of your staff,
The examination addressed areas important to public health and safety and was developed
and administered under interim Revision 8 to the Examiner Standards (NUREG-1021). All
portions of the examination were developed by Beaver Valley Power Station (BVPS) and
contractor personnel, while the NRC provided oversight and final approval prior to it's
administration. BVPS training personnel subsequently administered the, NRC-approved,
written portion of the examination, while the operating portion was administered by the
NRC.
In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter
and its enclosures will be placed in the NRC Public Document Room.
I
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4h+oo*V 3/A _ _ - - - _ - _ - - _ - _
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_ _ _ _ - - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _-___________ _-_- _ - ._ _ -
Mr. J. E. Cross - 2
' No reply to tbb ;ener i:, :Muired, however if any questions occur, regarding the
examination, please contact me at 610-337-5183,or by E-mail at RJC@NRC. GOV.
Sincerely,
.
LL *
_ /)
Richard J. Conte, Clpef
.A'
Operator Licensing and
Human Performance Branch
Division of Reactor Safety
Docket No. 50-334
Enclosure: Initial Examination Report No. 50-334/98-300(OL)
w/ Attachments 1 and 2
cc w/enci and Attachments 1-2:
K.Beatty, General Manager, Nuclear Support
cc w/ encl; w/o Attachments 1-2:
Susnil C. Jain, Vice President, Nuclear Services
R. Brandt, Vice President, Nuclear Operations Group and Plant Manager
R. LeGrand, Division Vice President, Nuclear Operations Group & Plant Manager
W. Kline, Manager, Nuclear Engineering Department
B. Tuite,~ General Manager, Nuclear Operations Unit
M. Pergar, Acting Manager, Quality Services Unit
J. Arias, Director, Safety & Licensing Department
J. MacDonald, Manager, System and Performance Engineering
M. Clancy, Mayor
Commonwealth of Pennsylvania
State of Ohio
State of West Virginia
L-____________.
_ - _ _ _ _ _ _ - _ _ _ _ - _ - _
d 9
Mr. J. E. Cross 3
Distribution w/enci and Attachments 1-2:
y DRS Master Examination File (V. Curley)
PUBLIC
Nuclear Safety Information Center (NSIC)
Distribution w/encI: w/o Attachments 1-2:
Region i Docket Room (with concurrences)
J. Wiggins, DRS
L. Nicholson, DRS
T. Kenny, Chief Examiner, DRS
N. Perry, DRP
D. Haverkamp, DRP
W. Axelson, DRA
NRC Resident inspector
DRS OL Facility File
DRS File
Distribution w/ encl: w/o Attachments 1-2 (VIA E-MAIL):
B. McCabe, OEDO ,
R. Capra, PD1-2, NRR
D. Brinkman, PDI-2, NRR
V. Nerses, PDI-2, NRR
R. Correia, NRR
F. Talbot, NRR
DOCDESK
Inspection Program Branch, NRR (IPAS) !
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. .
U. S. NUCLEAR REGULATORY COMMISSION
REGION 1
Docket No.: 50-334
Report No.: 98-300
License No.: DPR-66
Licensee: Duquesne Light Company
Facility: Beaver Valley Unit 1 Nuclear Power Plant
Location: Shippingport, Pennsylvania
Dates: April 20-24 and May 18,1998
Chief Examiner: T. Kenny, Senior Operations Engineer / Examiner
Examiners: J. D' Antonio, Operations Engineer / Examiner
T. Fish, Operations Engineer / Examiner
Approved By: Richard J. Conte, Chief, Operator Licensing and
Human Performance Branch
Division of Reactor Safety
i <8cw e m nero,
_ _ _ _ _ _ _ _ _ _ _ _ _
.. .
EXECUTIVE SUMMARY
Beaver Valley Unit 1 Nuclear Power Plant
inspection Report No. 50-334/98-300
Operations
Three Unit 1 senior reactor operator instant (SROI) candidates passed all portions of the
initiallicense examination.
Generic strengths were noted during the Unit 1 examination in the area of crew
communications, control board awareness, and crew briefings during the simulator portion
of the operating examination. The NRC examiners observed communications to be direct,
succinct, and that all crew members were kept wellinformed. Crew briefings were
routinely held during those instances in which time permitted. The written examination
j
was developed at the appropriate SRO knowledge level, as were the job performance '
measures and follow-up questions. Several JPMa, in lieu of questions, were appropriately
developed to test the knowledge level of the applicants in the administrative area of the
examination.
All three candidates met or exceeded Regulatory Guide (REGUIDE) 1.8, Rev. 2 for a Senior
Reactor Operator instant License.
Some of the site documents did not accurately delineate the current training requirements,
however, the licensee was in the process of correcting the errors. This was deemed a
minor violation.
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Report Details l
)
1. Operations ]
1
06 Operator Training and Qualifications
05.1 Senior Reactor Ooerator Initial Examinations
a. Scone
The NRC examiners reviewed on-site and in-office the examination as prepared by
Beaver Valley Power Station (BVPS) and contractor personnel in accordance with
the guidelines in interim Revision 8, of NUREG-1021," Examiner Standards," and
Revision 1 of NUREG-1122," Knowledge and Abilities Catalog for Nuclear Power
Plant Operators: Pressurized Water Reactors." The NRC examiners administered
initial operating licensing portion of the examination to three Unit 1 senior reactor
operator instant (SROI) candidates. The facility's training organization administered
the written examination.
b. Observations and Findinos
The results of SRO examination for Unit 1 are summarized below:
SRO Pass / Fail
Written 3/0
Operating 3/0
Overall 3/0
Overall the entire examination was well written and validated by the licensee prior
to the NRC reviewing it. This was evidenced by the few changes that were
required by the NRC after their review.
The written portion, job performance measures (JPMs) and simulator scenarios were
developed by Beaver Valley Power Station (BVPS) and their contractors in
accordance with NUREG-1021. The examination development team was comprised
of BVPS training and operation's representatives and a contractor. Allindividuals
involved signed a security agreement once the development of the examination
commenced. BVPS personnel validated the operation portions of the examination
prior to their submitting it to the NRC. The NRC subsequently reviewed and
observed the validation of all portions of the proposed examination. During the
examination preparation week, the NRC examiners noted that the facility staff had
performed good validation of the new simulator scenarios and JPM's, because little
or no changes were required after the demonstrations.
The written portion of the examination was administered on May 18,1998,and
consisted of 100 multiple choice questions. There were minor comments by the
NRC concerning the adequacy of four questions on the written examination, ;
however, the licensee promptly corrected them. The results of the written portion )
of the examination showed that question 51, regarding de bus ground faults and )
l
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. .
2
question 85, reaction of the reactor coolant system hot and cold leg temperatures,
during the first few minutes following a reactor trip coincident with a loss of offsite
power, were missed by all of the applicants. Discussions with the licensee showed
that they were aware of the problem and were taking appropriate actions to: (1)
remediate the candidates on missed questions and (2) perform an analysis to
determine if training outlines should be enhanced.
The operating portion of the examination was conducted from April 20-23,1998,
and consisted of thiee simulator scenarios and ten JPMs. All JPMs were followed
up with two system-related questions. All candidates were also examined using
JPMs and/or questions to evaluate the administrative requirement portion of the
examination.
Simulator and JPM performance by the candidates was very good.
Communications was also good, including the use of repeat backs. The examiners
noted that crew briefings were routinely performed by the SROs. Control board
awareness by all of the candidates was evident throughout each of the three
scenarios.
For the administrative segment of the operating portion of the examination,
administrative job performance measures (JPMs) were used in a number of
instances in lieu of administrative topic questions. The examiners determined that
candidate performance was good as evaluated in this area.
BVPS also exhibited good judgement in asking Region I for a one-month delay in
administering the written portion of the examination, because a quality product {
could not be produced in time to be administered with the operation portion in '
April 1998.
c. Conclusions
;
The candidates performed well on both the written and operating portions of the
examination, and thus were issued licenses. The candidates were well prepared for
the examination, indicating that the facility thoroughly evaluater' the knowledge and
ability of each candidate in an effort to determine their readiness to sit for an initial
NRC, SROI examination. Crew communications, control board awareness, and
crew briefings were very good. The training department continued to do an
excellent job in adhering to the examiner standards and in developing the !
examination materials needed to administer the examinations. >
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05.2 Acolicant Trainina an_d Exoerience
a. Scope
Regulatory Guide 1.8 (REGUIDE), Rev. 2 requires certain requirements and certain
obligations in the area of training and experience be satisfied by a license candidate
prior to taking the examination for a hot Senior Reactor Operators license. The
inspectors reviewed the three candidates' training records and NRC records to verify
compliance with these requirements.
b. Observations and Findinas
REGUIDE 1.8 requires that:
* Each candidate, for a senior license, have a high school diploma or equivalent. The
inspectors verified that all candidates met or exceeded the requirement.
* Each candidate, for a senior license, have four years of responsible power plant
experience. The inspectors verified that all candidates met or exceeded the
requirement.
* Each candidate, for a senior license, serve three months as an extra person on shift
in training for that position. Three months is the equivalent of 520 hours for a 40
hour work week. The inspectors verified that all candidates met or exceeded the
requirement.
* Each candidate, for a hot license, should manipulate controls of the facility during a
minimum of five reactivity changes. The inspectors verified that all candidates met
or exceeded the requirement.
The licensee requires that each candidate maintain a training note book to document
the above requirements. The instructions and requirements for maintaining the
forms are delineated in the Training Administrative Manual (TAM). The inspectors
verified that the inspected portion of the TAM reflected the requirements of
Regulatory Guide 1.8, Rev. 2.
Also, the inspectors reviewed the Technical Specifications (TS), The Quality
Assurance Manual (QAM) and The FSAR to determine if these documents
defineated the proper references to the training requirements. The inspector found
inconsistencies within the documents. The TS referenced, "section 5.5 of ANSI
N18.1-1971 and CFR Part 55." The QAM referenced, "10 CFR 50 and 10 CFR 55
and be in agreement with ANSI N18.1-1971." The FSAR referenced, "REGUIDE
1.8, Rev.1-R, September 1975" and had been updated since the original version.
The TAM referenced, "REGUIDE 1.8."
I- - . _ . __ _________ ________________ _
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. .
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The licensee was conducting their training of perspective operators in accordance
with REGUIDE .8, Rev. 2. This is delineated in the TAM. The licensee issued
Condition Report (CR) 980734, on April 9,1998, that describes the inconsistency.
After discussions and a review of the CR the inspector determined that the licensee
was taking corrective actions, and were expected to resolve the issue by
l '
June 12,1998, with the exception of the TS change, which may take longer. This
failure constitutes a violation of minor significance and is not subject to formal
enforcement action.
c. Conclusions
Current operator license training is being conducted in accordance with REGUIDE
- 1.8, however, site documents were not consistent with the proper reference to the
current NRC required training deaument, REGUIDE 1.8. The licensee was in the
process of changing the documents.
E8 Review of the FSAR.
While performing the preexamination activities discussed in this report, the
inspectors reviewed applicable portions of the FSAR, that related to the selected
examination questions or topic areas. One discrepancy discussed in the previous
paragraphs was identified and was being corrected by the licensee.
V. Manaaement Meetinas
X1 Exit Meeting Summary
On April 23 and May 20, the NRC examiners discussed their observations regarding the
examination with 8eaver Valley Unit 1 operations and training management
representatives. The examiners discussed candidate performance, including
communications and briefings among themselves, both were very good. The licensee did
not see the need to make comments following the administration of the written
examination.
The examiners a!so expressed their appreciation for the cooperation and assistance that
was provided during both the preparation end examination week by licensed operator
i training personnel and operations personnel. The following participated in the exit !
rneetings.
l l l l \ L________.--_______ ..a
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. .
5
PARTIAL LIST OF PERSONS CONTACTED
BEAVER VALLEY
K. Beatty, General Manager, Nuclear Support
R. Brooks, Sr. Nuclear Operations Instructor
W. Lindsey, Director, Operator Training
S. C-Jain, Vice President, Nuclear Services
B. Tuite, General Manager, Nuclear Operations
L. Shad, Simulator Supervisor
li&G
T. Kenny, Senior Operations Engineer, Chief Examiner
T. Fish, Operations Engineer
J. D' Antonio, Operations Engineer
Attachments:
1. Beaver Valley Unit ? SRO Written Examination w/ Answer Key
2. Simulation Facility Report
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U. S. NUCLEAR REGULATORY COMMISSION
REGION 1
Docket No.: 50 334
Report No.: 98-300
License No.: DPR-66
Licensee: Duquesne Light Company
Facility: Beaver Valley Unit 1 Nuclear Power Plant
Location: Shippingport, Pennsylvania
Dates: April 20 24 and May 18,1998
Chief Examiner: T. Kenny, Senior Operations Engineer / Examiner
j
Examiners: J. D' Antonio, Operations Engineer / Examiner j
T. Fish, Operations Engineer / Examiner :
Approved By: Richard J. Conte, Chief, Operator Licensing and
Human Performance Branch
Division of Reactor Safety
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. .
EXECUTIVE SUMMARY
Beaver Valley Unit 1 Nuclear Power Plant
Inspection Report No. 50-334/98 300
Operations
1
Three Unit 1 senior reactor operator instant (SR0 ) candidates passed all portions of the )
initial license examination.
Generic strengths were noted during the Unit 1 examination in the area of crew
communications, control board awareness, and crew briefings during the simulator portion
of the operating examination. The NRC examiners observed communications to be direct,
succinct, and that all crew members were kept wellinformed. Crew briefings were
routinely held during those instances in which time permitted. The written examinabon
was developed at the appropriate SRO knowledge level, as were the job per formance
measures and follow-up questions. Several JPMs, in lieu of questions, were appropriately
developed to test the knowledge level of the applicants in the administrative area of the
examination.
All three candidates met or exceeded Regulatory Guide (REGUIDE) 1.8, Rev. 2 for a Senior
Reactor Operator Instant License.
Some of the site documents did not accurately delineate the current training requirements,
however, the licensee was in the process of correcting the errors. This was deemed a
minor violation.
ii
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'
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Rooort Details i
L.,Qoerotions
05 Operator Training and Qualifications
05.1 Senior Reactor Ooerator Initial Examinations
a. Scope
The NRC examiners reviewed on-site and in-office the examination as prepared by
Beaver Valley Power Station (BVPS) and contractor personnel in accordance with
the guidelines in interim Revision 8, of NUREG-1021," Examiner Standards," and l
Revision 1 of NUREG-1122," Knowledge and Abilities Catalog for Nuclear Power l
Plant Operators: Pressurized Water Reactors." The NRC examiners administered l
initial operating licensing portion of the examination to three Unit 1 senior reactor !
operator instant (SROI) canJidates. The facility's training organization administered l
the written examination.
b. Observations and Findinos
The results of SRO examination for Unit 1 are summarized below:
SRO Pass / Fail )
Written 3/0
Operating 3/0
Overall 3/0 ;
Overall the entire examination was well written and validated by the licensee prior
to the NRC reviewing it. This was evidenced by the few changes that were
required by the NRC after their review.
The written portion, job performance measures (JPMs) and simulator scenarios were
developed by Beaver Valley Power Station (BVPS) and their contractors in
accordance with NUREG-1021. The examination development team was comprised
of BVPS training and operation's representatives and a contractor. Allindividuals
involved signed a security agreement once the development of the examination
commenced. BVPS personnel validated the operation portions of the examination
prior to their submitting it to the NRC. The NRC subsequently reviewed and
observed the validation of all portions of the proposed examination. During the
examination preparation week, the NRC examiners noted that the facility staff had
performed good validation of the new simulator scenctios and JPM's, because little
or no changes were required after the demonstrations.
The written portion of the examination was administered on May 18,1998,and
consisted of.100 multiple choice questions. There were minor comments by tise
NRC concerning the adequacy of four questions on the written examination,
however, the licensee promptly corrected them. The results of the written portion
of the examination showed that question 51, regarding de bus ground faults and
l E- _- _ _ _ -- - )
. .
2
question 85, reaction of the reactor coolant system hot and cold leg temperatures,
during the first few minutes following a reactor trip coincident with a loss of offsite
power, were missed by all of the applicants. Discussions with the licensee showed
that they were aware of the problem and were taking appropriate actions to: (1)
remediate the candidates on missed questions and (2) perform an analysis to
determine if training outlines should be enhanced.
The operating portion of the examination was conducted from April 20-23,1998, !
and consisted of three simulator scenarios and ten JPMs. All JPMs were followed l
up with two system-related questions. All candidates were also examined using
JPMs and/or questions to evaluate the administrative requirement portion of the l
examination.
Simulator and JPM performance, by the candidates was very good.
Communications was also good, including the use of repeat backs. The examiners
noted that crew briefings were routinely performed by the SROs. Control board
awareness by all of the candidates was evident throughout each of the three
scenarios.
For the administrative segment of the operating portion of the examination,
administrative job performance measures (JPMs) were used in a number of
instances in lieu of administrative topic questions. The exarniners determined that
candidate performance was good as evaluated in this area.
BVPS also exhibited good judgement in asking Region I for a one-month delay in
administering the written portion of the examination, because a quality product
could not be produced in time to be administered with the operation portion in
April 1998.
c. Conclusions
The candidates performed well on both the written and operating portions of the
examination, and thus were issued licenses. The candidates were well prepared for
the examination, indicating that the facility thoroughly evaluated the knowledge and
ability of each candidate in an effort to determine their readiness to sit for an initial
NRC, SROI examination. Crew communications, control board awareness, and
crew briefings were very good. The training department continued to do an
excellent job in adhering to the examiner standards and in developing the
examination materials needed to administer the examinations.
I
1
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3
05.2 Acolicant Trainina and Experience
a. Scope
Regulatory Guide 1.8 (REGUIDE), Rev. 2 requires certain requirements and certain
obligations in the area of training and experience be satisfied by a license candidate
prior to taking the examination for a hot Senior Reactor Operators license. The
inspectors reviewed the three candidates' training records and NRC records to verify
compliance with these requirements.
b. Observations and Findinas
REGUIDE 1.8 requires that:
o Each candidate, for a senior license, have a high school diploma or equivalent. The I
inspectors verified that all candidates met or exceeded the requirement.
e Each candidate, for a senior license, have four years of responsib!e power plant
experience. The inspectors verified that all candidates met or exceeded the
requirement.
o Each candidate, for a senior license, serve three months as an extra person on shift
in training for that position. Three months is the equivalent of 520 hours for a 40
hour work week. The inspectors verified that all candidates met or exceeded the
requirement,
e Each candidate, for a hot license, should manipulate controls of the facility during a
minimum of five reactivity changes. The inspectors verified that all candidates met
or exceeded the requirement.
The licensee requires that each candidate maintain a training note book to document
the above requirements. The instructions and requirements for maintaining the '
forms are delineated in the Training Administrative Manual (TAM). The inspectors
verified that the inspected portion of the TAM reflected the requirements of
Regulatory Guide 1.8, Rev. 2.
Also, the inspectors reviewed the Technical Specifications (TS), The Quality
Assurance Manual (QAM) and The FSAR to determine if these documents
delineated the proper references to the training requirements. The inspector found
inconsistencies within the documents. The TS referenced, "section 5.5 of ANSI
N18.1-1971 and CFR Part 55." The QAM referenced, "10 CFR 50 and 10 CFR 55
and be in agreement with ANSI N18.1-1971." The FSAR referenced, "REGUIDE
1.8, Rev.1-R, September 1975" and had been updated since the original version.
The TAM referenced, "REGUIDE 1.8."
I I
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_
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4
The licensee was conducting their training of perspective operators in accordance
with REGUIDE 1.8, Rev. 2. This is delineated in the TAM. The licensee issued
Condition Report (CR) 980734, on April 9,1998, that describes the inconsistency.
After discussions and a review of the CR the inspector determined that the licensee
was taking corrective actions, and were expected to resolve the issue by
June 12,1998, with the exception of the TS change, which may take longer. This
failure constitutes a violation of minor significance and is not subject to formal
enforcement action.
c. Conclusions
Current operator license training is being conducted in accordance with REGUIDE
1.8, however, site documents were not consistent with the proper reference to the
current NRC required training document, REGUIDE 1.8. The licensee was in the
process of changing the documents.
E8 Review of the FSAR.
While performing the preexamination activities discussed in this report, the
inspectors reviewed applicable portions of the FSAR, that related to the selected
examination questions or topic areas. One discrepancy discussed in the previous
paragraphs was identified and was being corrected by the licensee.
V. Manaaement Meetinas
X1 Exit Meeting Surnmary
On April 23 and May 20, the NRC examiners discussed their observations regarding the
examination with Beaver Valley Unit 1 operations and training management
representatives. The examiners discussed candidate performance, including
communications and briefings among themselves, both were very good. The licensee did
not see the need to make comments following the administration of the written
examination.
The examiners also expressed their appreciation for the cooperation and assistance that
was provided during both the preparation and examination week by licensed operator
training personnel and operations personnel. The following participated in the exit
meetings.
L_ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _
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5
PARTIAL LIST OF PERSONS CONTACTED
BEAVER VALLEY
K. Beatty, General Manager, Nuclear Support
R. Brooks, Sr. Nuclear Operations instructor
W. Lindsey, Director, Operator Training
S. C-Jain, Vice President, Nuclear Services
B. Tuite, General Manager, Nuclear Operations
'
L. Shad, Simulator Supervisor
NBC
T. Kenny, Senior Operations Engineer, Chief Examiner
l T. Fish, Operations Engineer
J. D' Antonio, Operations Engineer
Attachments:
1. Beaver Valley Unit 1 SRO Written Examination w/ Answer Key
2. Simulation Facility Report
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Attachment 1
BV-1 SRO WRITTEN EXAMINATION W/ ANSWER KEY
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t)sestion Topee: l Temperature trending during cooldown
A cooldown is in pr:gress. The milestones listed on Figure 1 of 10M-51.4C, (see attached) were reached at
the following times:
*-(1) 0800
* - (2) 0833-
*-(3) 0857 '
* (4) 0917
l What action, if any, is required to be taken to comply with Technical Specifications 7 l' l c. RCS cooldown is acceptable to this point. RCS cooldown rate will not be exceeded if Figure I time v l limits are complied with from this point on. ,
l
b. RCS cooldown is acceptable to this point. RCS cooldown rate may be exceeded even if Figure 1
times are complied with from this point on.
l c. RCS cooldown exceeded Technical Specifications. RCS temperature must remain constant until
0927. I
I
d. RCS cooldown exceeded Technical Specifications. Cooldown rate must be restored to within
'
l
Technical Specification limits by 0947. ;
- - Ams
- la l Eram Level: lS l Cognitive Level: l Application l
Explomatio I
o ef Answer
l
KA: l2.1.2 l RO Value: l3.0 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1
System / Evolution .
Title:
KA: Conduct of Operations
Statement:
Knowledge of operator responsibihties during all modes of plant operation.
Reference Reference Number Reference Section Page Nunsber(s) Revision Learn.
Obj ;
'
Station Shutdown - lOM 51.4.C IV.A.13.b C 1011 iss 4 Rev
Cooldown From MODE 3 to 12
MODE 4
l Beaver Valley - Unit 1 3.4.9.2 3/4422,427 Amend l Technical Specifications No.179 '
OM 6,7 & 10 Operational LP SQS-RX IV.D.4 20 6
Lecture
Question Source l New l Question Modification Method l
l Question Source Comments: l
.
M tirial Required for Figure 1 of OM-51.4.C - Blowup curve to max 81/2 x 11
l Ex:mination: !
Page1
L
___- _-___ - -
-
f. ~
*
Question Tepic: l Core Safety Limit Curve eval
At 20% power, the maximum dlawable T , is limited by the Reactor Core Safety Limit. The basis for
limiting T,y under these conditions ensures that:
a. DNBR remains greater than or equal to the safety analysis DNBR limit and the average enthalpy at
the vessel exit will not exceed saturation.
b. DNBR remains greater than or equal to the safety analysis DNBR limit and the highest enthalpy
anywhere in the core will not equal saturation.
c. DNBR remains less than the safety analysis DNBR limit and the average enthalpy at the vessel exit
will not exceed saturation.
d. DNBR remains less than the safety analysis DNBR limit and the highest enthalpy anywhere in the
core will not exceed saturation.
A ns: la l Exam Level: lS l Cognitive level: l Memory l
Explanatio
c ef Answer
KA: l 2.1.10 l RO Value: l2.7 l SRO Value: l 3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1
Syst:m/ Evolution
Title:
KA Conduct of Operations
Statement:
Knowledge of conditions and limitations in the facility license.
Reference L ference Number Reference Section Page Number (s) Revision Learn.
Obj
Reac3or Protection System LP-SQS-1.1 II.C.3 7 6 4.c
Question Source l New l Question Modification Method l
Question Source Comments: l
M trial Required for TS Figure 2.1.1
paination:
Page 2
L - - _ _ _ _ _ __-._________
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Question Tepic: l TS 3.0.5
During power oper tion the Diesel Generator #1 is declared inoperable. Subsequently the 1B Quench Spray
pump is determined to be inoperable.
Assuming all required surveillance are completed satisfactorily, what is the required Technical Specification
action?
I
s. - Restore both the 1B Quench Spray and Diesel Generator #1 operable status within 72 hours or be in
Hot Standby within the following 6 hours.
b. Restore either the IB Quench Spray pump or Diesel Generator #1 to operable status within 24 hours
or be in Hot Standby.within the following 6 hours.
c. Restore the 1B Quench Spray pump to operable status within one hour or be in Hot Standby within
the following 6 hours.
d. Restore the IB Quench Spray pump or Diesel Generator # 1 to operable status within 2 hours or be
in Hot Standby within the following 6 hours.
Ans: ld l Exam level: lS l Cognitive Level: l Application l
Explanatio
e ef Answer
{
KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1
System / Evolution
Title:
KA Conduct of Operations
Statement: )
Ability to apply technical speci0 cations for a system.
Ref;rence Reference Number Reference Section Page Number (s)
~
Revision Learn.
Obj
Technical Specifications TS 3.0.5,3.6.2.1,
3.8.1.1
Containment LP-SQS-13.01 5 12
Depressurization Systems
Q:estion Source l Facility Exam Bank l Question Modification Method l
Qrestion Source Comments: l
Material Required for Technical Specifications
Ex:mination:
l l Page 3 l l b-- _. ._ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- - - - _ - - - - _ _ _ _ _ _ - - _ _ _ _ - _ _ . _ _
e .
Question T pic: l FFD requirements
What are the fitness-for-duty requirements, with respect to alcohol, for an unscheduled RO who has be
called out?
c. The RO may report to work if he/she has consumed alcohol within the past FIVE hours, but will b
required to pass a breath analysis test,
b. The RO may report to work if he/she has consumed alcohol within the past FIVE hours, but will be
subject to a breath analysis test only if deemed necessary by the NSS.
c. The RO must report to work even if he/she has consumed alcohol within the past FIVE hours, but
will be required to pass a breath analysis test.
d. The RO shall not report to work if he/she has consumed alcohol within the past FIVE hours.
l Cognitive Level: l Application l
Ans: la l Exam Level: lS
Explanatio
e of Answer
KA: l 2.1.13
l RO Value: l2.0 l SRO Value: l2.9 l Section: l PWG l RO Group: l 1 l SHO Group: l1
Syst:. m/Evolut10n
Title:
KA Conduct of Operations
Statement:
Knowledge of facility requirements for controlling vital / controlled access.
Page Number (s) Revision Learn.
Reference Number Reference Section
Reference Ob)
2 0
Fitness-For-Duty Program 1/2 NPDAP 2.14 IV.2 & 3
For Duquesne Light
Employees 10 3.39
Vlli, 18
Conduct Of Operations I/2LP SQS-48.1
Question Source l New l Question Modification Method l
Q: estion Source Comments: l
Material Required for 1/2 NPDAP 2.14
Examination:
l ,
Page 4
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_ _ _ _ - _ - - - -
,Questiop Topic: l TS SDM & Emergency Boration
Given the following conditions:
=
! RCS T,,, - 355 F
*
! RCS pressure - 400 psig
*
RCS boron concentration 2000 ppm
. Shutdown margin is below Technical Specifications allowable value
a
Emergency Boration is initiated at 30 gpm boric acid
*
A 70 ppm RCS boron concentration change is required to restore the
required SDM
I
Of times listed below, which is the MINIMUM emergency boration time that will ensure the required boric
acid has been added?
a. 15 minutes
b. 17 minutes
_
c. 21 minutes
d. 24 minutes
Ans: {c l Exam Level: lS l Cognitive Level: l Application l
Explanatio
A 70 ppm change at Normal Operating Conditions would require 500 gallons boric acid. 'Ihe correction factor of
a of A'swer 1.18 multiplied by 500 would result in 590 gallons of boric acid. 590/30gpm = 19 minutes 40 seconds.
KAt l 2.1.25 l RO Value: l2.8 l SRO Value: l 3.1 l Section: l PWG l RO Group: l 1 l SRO Group: l1
System / Evolution
Title:
KA Conduct of Operations
Stat: ment:
-
Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain
perfonnance data.
Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.
'
Obj
Emergency Boration IOM-7.4.S IV.A S2 Iss 4 Rev
1
r Beaver Valley Unit 1 - 3.1.1.1 3/4 1-1 Amend
Technical Specifications
No. 91
CVCS LP-SQS-7.1 IV.E 28 12
Question Source l New l Question Modification Method l
Question Source Comments: l
Material Reouired for IOM-7.5 Figures 7-7,7-8 & Table 7-1.
Examination:
l
l
!
l
Page5
L_____-__.
'
, ,
Question Topic: l Permission for deviation from NSA.
in addition to normal requirements for manipulating components, which of the fcilowing describes who is
c
required to approve placing component in other than its Normal System Alignment (NSA)?
l '
;
l
e. Two SROs are required to approve the manipulation. 1
b. Specific permission is required from the NSS.
c. Either the NSS or ANSS has to approve the manipulation.
d. The General Manager, Nuclear Operations.
.
l Cognitive Level: l Memory l
Ans: lc l Exam Levet: lS
Explanatio
e of Answer
l Section: l PWG l RO Group: l 1 l SRO Group: l1
KA: l 2.1.29 l RO Value: l3.4 l SRO Value: l 3.3
System / Evolution
Title:
KA Conduct of Operations
Statement:
Knowledge of how to conduct and verify valve lineups.
Page Number (s) Revision Learn.
Ref;rence Reference Number Reference Section
Ob)
Question Source l Facility Exam Bank l Question Modification Method l
Question Source Comments: l
M:terial Required for
Ex:mination:
i
l
,
i
l
Page 6
--------____J
Question Tepic: l Procedure change rules for type of procedure
l
While at 100% power, an OMCN is to be written to change !OM-7.4.L " Blender Boration Operation." Thi !
change cdds a step that directs placing ONE bank of Pressurizer heaters in MANUAL prior to initia
boration. An Operations Unit Non-Intent Reviewer has determined that this does NOT change the intent
,
the procedure.
l !
The on the spot change:
a.
can be approved by TWO members of management, ONE holding a valid SRO license on Unit 1.
b. becomes effective 14 calendar days following review by the OSC and approval of the GMNO.
c.
cannot be made because use of the procedure is not expected in the next 30 days. I
d. cannot be made because this is a safety related procedure.
Ans: {a l Exam Level: lS 1 Cognitive Level: l Comprehension l
j
Explanatio
a ef Answer * - ,
'
KA: l2.2.6 l RO Value: l2.3 } SRO Value: l3.3 l Section: l PWG l RO Group: ll l SRO Group: l1
System / Evolution
Title:
KA Equipment Control
Stat; ment:
Knowledge of the process for making changes in procedures as described in the safety analysis report. l
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
ControlOf Operating 1/20M-48.2.B C.I.a B 10 1ss 4 Rev I
Procedures '
33
Conduct Of Operations 1/2LP-SQS-48.1 1.H.2 I
4 to g, 9
)
Q:estion Source l New l Question Modification Method l
QIestion Source Comments: l
Mmrial Required for
Examination:
Page 7
l.
.
--_----__ - _ - ___
. .
Question Tcpic: l Omissions in OSTs
A parti:1 OST is to be perforrned. Which of the following is an acceptable method of blocking the portio
of the OST that are NOT applicable?
a. The ANSS blocks the non applicable portions.
b. The STA blocks the non-applicable portions and the RO verifies they are correct.
c. The system engineer blocks the non-applicable portions and the ANSS verifies they are correct.
d. The PO blocks the non-applicable portions and the RO verifies they are correct.
l Cognitive Level: l Memory l
Ass: lc l Exam tevel: lS
Explanatio
aef Answer
l Section: l PWG l RO Group: l 1 l SRO Group: l1
KA: l 2.2.12 l RO Value: l3.0 l SRO Value: l 3.4
System / Evolution
Title:
KA Equipment Control
Statement:
Knowledge of surveillance procedures.
Reference Section Page Number (s) Revision Learn.
Ref.rence Reference Number
Ob)
VI.13.17 10 iss 3 Rev
Adherence and 1/20M-48.2.C
18
Familiarization to Operating
Procedures 10 10
Conduct Of Operations 1/2LP-SQS-48.1
Q7estion Sous ce l New l Question Modification Method l
Question Source Comments: l
Mit; rial Required for
Ex:mination:
Page 8
i
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
Questiod Topic: l Caution Tags
Use of a Caution Tag is PROHIBITED for which of the following conditions?
a. Special additional manual actions are required to operate the tagged component.
b. Operation of the tagged component will be affected because a portion of the system is not in NSA.
c. As a temporary replacement for a component label that has fallen off,
d. As a warning that operation of the component will cause erratic indication.
A s: lc l Exam Level: lS l Cognitive Level: l Memory l
Expiaratio
j acf Arswer
KA: l2.2.13 l RO Value: l3.6 l SRO Value: l3.8 l Section: l PWG l RO Group: l 1 l SRO Group: l1
Syst;m/ Evolution l
j
l Title: ' ! KA Equipment Control
St:tement:
Knowledge of tagging and clearance procedures.
R:f;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj j
Use cf Caution Tags 1/2OM-48.3.L IV.A 1-2,3 iss 4 Rev
l
4
Conduct Of Operations I/2LP-SQS-48.1 V.P 7 10 15
.
,
Question Source l Facility Exam Bank l Question Modification Method l ;
l
Question Source Comments: l l
Mat: rial Required for
l
Ex mination:
;
l Page 9 l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
*
Qrestion Topic: l SRO control
Which cf the following describes t. responsibility of the Refueling SRO during fuel movement?
The Refueling SRO will:
e. initial the Fuel Assembly Handling Deviation Report with NSS concurrence.
b. be located on the manipulator crane structure during most fuel handling activities.
! c. maintain the DLC Master Copy of the Fuel Handling data Sheets.
d.- continuously monitor source range count level.
Ans: lb l Exam level: lS l Cognitive Level: l Memory l
Explanatio
e cf Answer
KA: l2.2.31 l RO Value: l1.6 l SRO Value: l 3.8 l Section: l PWG l RO Group: l 1 l SRO Group: l1
System / Evolution
Title:
KA Equipment Control
Statement:
Knowledge of SRO fuel handling responsibilities.
Reference Reference Number Reference Sect. ion Page Number (s) Revision Learn.
Obj
Refueling Administrative Book 1 -1RP-12R-1.1 II.D.4.b.15) 10 iss 0 Rev
Section 0
Fuel Handling Onerations LP-SQS-6.13 Ill.B 5 5 2.b
Question Source l New l Question Modification Method l
Question Source Comments: l
Miterial Required for
Examination:
Page 10
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
.. , , _ _
- - _ _ _ _ _ -
Technical Specific:tions requires radittion areas to be isolated by locked doors if the radittion levels are
greater than:
c. 100 mrem /hr
,
b. 500 mrem /hr
c. 1000 mrem /hr
d. 5000 mrem /hr
Ans: lc l Exam Level: lS l Cognitive Level: l Memory I
!
Explanatio
a of Answer
KA: l2.3.1 l RO Value: l2.6 l SRO Value: l3.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1
System / Evolution
Title:
KA Radiation Control
Statement: I
Knowledge of 10 CFR: 20 and related facility radiation control requirements.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Technical Specifications 6.12 6-23 188
s
Question Source l New l Question Modification Method l
, Question Source Comments: l
Material Required for Verify Section 6 of the Technical Specification is not included in materials
Examination:
l
i
Page11 l
_ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _
- - _ _ _ - _ - _ - _ - - _ _ - _ - - _ - - _ _ _ _ _ - - _ - - _ - - ______ --________________ ____ ___ - _ _ _ - - _ _
* *
Question Topic: l SRO action for gas releise
Given the following conditions:
* Reactor power- 100%
* Discharge of Waste Gas Decay Tank [lGW-TK-1 Al is planned for 1000 on 4/22/98
* . The RWDA-G had been approved on 1500 on 4/20/98
+ The meteorological information indicates Stability Class A for atmospheric conditions
* "Ihe status of the Gaseous Effluent Monitors is as follows:
. Gaseous Waste / Process vent [RM-GW-108A] noble gas channel inoperable
- Gaseous Waste / Process vent (RM-GW-108B] noble gas channel inoperable
Preparation for the release was then delayed until .2300 on 4/23/98
Which of the following describes the status at the new planned time for release (2300 on 4/23/98), assuming
eq9ipment status and other conditions do NOT change?
a. The release can be initiated without restriction.
b. The release can be initiated only if sampling of the release stream is analyzed at least one per every
FOUR hours.
c. The release cannot be made because the 72-hour effective time limit for the RWDA-G has elapsed.
d. The release cannot be made because the Stability Class for release is unacceptable.
Ans: ic l Exam level: lS l Cognitive Level: l Comprehension l
Esplanatio
n of Amsyser
KA: l2.3.6 l RO Value: l2.1 l SRO Value: l3.1 l Section: l PWG l RO Group: l 1 l SRO Group: l1
System / Evolution
Title:
KA Radiation Control
Statement:
Knowledge of the requirements for reviewing and approving release permits.
Reference Number Reference Section Page Number (s) Revision Learn.
Reference
Ob]
Decay Tank Discharge IOM-19.4.E step 7 NOTE E3 iss 3 Rev
2
Gaseous Waste Disposal II.G, ODCM 3.3.3.10 17 18 5 9.e
LP-SQS-19.1
System
Question Source l New l Question Modification Method l
Question Source Comments: l
Mat; rial Required for IOM-19.4.E
Examination:
Page 12
_ _ _ - - __ _ __ _ __-_ _ _ -_-_ _ _ -_ - -_ - __ A
, _ _ _ .
I Given the following conditions:
i The reactor has been shutdown for 2 days.
L * RCS temperature is 150 'F. ' l '
* RCS pressure is atmospheric. * PZR is a normal level for shutdown cooling.
l
Assume RHR is lost. Which of the following describes the time available until core boiling occurs?
l
( Using the attached references, AOP 1.10.1 attachments 1,2,3, & 4)
a. Less than 10 minutes.
. b. I1 to 20 minutes,
c. 21 to 30 minutes.
d. 31 to 40 minutes.
Ans: ld l Exam level: lS l Cognitive Level: l Application l
Explanatio
o of Answer
KA: l2.4.9 l RO Value: l3.3 l SRO Value: l3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1
System / Evolution
Title:
KA Emergency Procedures / Plan
Statement:
I
Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR mitigation strategies).
-
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
i Residual Heat Removal AOP 1.10.1 11, Attachment I iss 3A
System Loss Rev 5
Residual Heat Removal . LP-SQS-10.1 8 9,10
System
Q~estion Source l Previous 2 NRC Exams l Question Modification Method l
Question Source Comments: l
Material Required for AOP 1.10.1 Attachments 1,2 3 & 4.
Ex mination:
.
I
Page 13
__.______-_____-_-______D
. .
Question Topic: l Implementation of Orange Path
Given the following conditions;
a An unisolable steam line break has occurred on SG "B"
+ , SG " A" and "C" levels were overfed.
* A reactor trip and SI occur.
* Pressurizer pressure is 1180 psig
* Pressurizer levelis 12%
= T.,is 400 'F and slowly dropping
* E-0 " Reactor Trip Or Safety Injection", step 9 is being performed.
* The STA informs' crew that B loop Two is 283 F and slowly dropping.
What is the EOP flowpath that will be followed given the above conditions?
a. Immediately transition to FR-P.1 " Response To Imminent Pressurized Thermal Shock Condition"
b. Perform actions of E-0 through diagnosis of steamline break, then transition to E-2 " Faulted Steam
Generator Isolation"
c. Perform actions of E-0 through diagnosis of steamline break, then transition to FR-P.1 " Response to
Imminent Pressurized Thermal Shock Condition"
d. Perform actions of E-0 through diagnosis of steamline break, then transition to FR-P.'2 " Response to
Anticipated Pressurized Thermal Shock Condition".
I Exam Ixvel: IS l Cornitive Ixvel: l Application l
Ass- lc
Explanation
ofAnswer
KAt l 2.4.14 l RO Value: l3.0 l SRO Value: l3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1
System / Evolution
Title:
KA Emergency Procedures / Plan
Statement:
Knowledge of general guidelines for EOP flowchart use.
Reference Section Page Number (s) Revision Learn.
Reference Reference Number Obj
ORANGE PATH iss IB
Suberiticality - Status Tree F-0.4
Rev1
1. Ist paragraph 1 IssIB
Reactor Trip Or Safety IOM-538.4.E-0
Rev5
Injection Background 1
LP-SQS-53.1 B.I 2
EOP Introduction '
Question Source lNew l Question Modification Method l
Question Source Comments: l
Material Required for F 0.4 and Att 5-D
Examination:
Page 14
E _ ____ ________ _ _
. - __
,_ -
During Critical Safety Function Status Tree monitoring it was determined that TWO functions had Orange
Pcths. One of the Orange paths is FR-H.1, Response to Loss of Secondary Heat Sink.
Which Critical Safety Function, also Orange, would take precedence over FR-H.l?
e. FR-C.1, Response to Inadequate Core Cooling
b. FR-Z.1, Response to High Containment Pressure
c. FR-P.1, Response to Imminent Pressurized Thermal Shock Condition
d. FR-I.1, Response to High Pressurizer Level
Ars: la l Eram Level: lS l Cognitive Level: l Comprehension l
Explanatio
o of Answer
RA: l 2.4.16 l RO Value: l3.0 l SRO Value: l 4.0 l Section: l PWG l RO Group: ll l SRO Group: l1
, Syst m/ Evolution
Title:
KA Emergency Procedures / Plan
Stat: ment:
Knowledge of EOP i;nplementation hierarchy and coordir.aion with other support procedures.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
EOP Executive Volume - 1/2OM-53B.2 III.B 9 iss1B
Users Guide Rev 3
EOP Introduction LP SQS-53.1 2
Question Source l Facility Exam Bank l Question Modification Method l
Qrestion Source Comments: l
M:terial Required for
Ex:mination:
i
l
l
l
Page 15
j
. .
Question Topic: l Functionil Recovery Procedure usage
During a loss of til Emergency 4KV AC Power, When c.re Functional Restoration Procedures implemented? l
c. Immediately upon electrical power restoration to I AE or IDF.
b. Immediately upon exiting ECA-0.0 " Loss of all 4KV AC Emergency Power "
c. When directed by ECA-0.1 " Loss of all Power Recovery Without SI Required" or ECA-0.2. " Loss
of all AC Power Recovery With SI Required"
d. When ECA-0.i " Loss of all Power Recovery Without SI Required" or ECA-0.2. " Loss of all AC
Power Recovery With SI Required" is completed.
l Cognitive Level: l Memory l
Ans: lc l Exam level: lS
Expleastic
o of Answer
KA: l 2.4.16 { RO Value: l3.0 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1
System / Evolution
Title:
KA Emergency Procedures / Plan
Statement:
Knowledge of EOP implementation hierarchy and coordination with other support procedures.
Reference Section Page Number (s) Revision Learn.
Ref;rence Reference Number
Obj
VI.D 15 iss 1B
EOP Executive Volume - 1/20M 53B.2
Rev 3
User's Guide I
IV.C.4 20 1
EOP Introduction LP SQS-53.1
Question Source l Facility Exam Bank l Question Modification Method l
Question Source Comments: l
Miterial Required for j
Ermination: 1
Page 16
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__
_ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ - -_ _ _ - _ _ -
I
Q7estion Topic: l Fire Brigade Responsibilities
'
During a plant fire, who is responsible for coordinating fire-fighting activities with the offsite fire department
chiefs?
a. The ANSS when acting as the Fire Brigade Chief.
b. The ANSS when acting as the Fire Brigade Captain.
c. The affected Unit's NSS.
- d. The Nuclear Operator when he/she is acting as the Fire Brigade Captain.
Ans: la- l Exam Level: lS l Cognitive Level: l Memory l Explanatio
l c of Answer i KAt l 2.4.27 l RO Value: l3.0 l SRO Value: l3.5 l Section: l PWG l RO Group: l 1 l SRO Group: l1
- Systesa/ Evolution
Title:
l- KA Emergency Procedures / Plan l Statessent:
Knowledge of fire in the plant procedure.
i Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Fire Protection NPDAP 3.5 111.N 3 6 Conduct of Operations 1/2LP-SQS-48.1 10 1
l
Question Source l New l Question Modification Method l Question Source Comments: l~ M;terial Required for Ermination:
! ! Page 17
;
______-____ _ _____ -__- _ _ _
^
*
Question Topic: l Rod motion control
If a power mismatch signal is g nerated by the Rod Control System, which of the following parameters
determines the magnitude of the gain imposed by the variable gain unit?
c. Median Tave
b. Median delta T
c. N44 Power
d. Turbine Impulse pressure
Assi - l d l Exam level: lS l Cognitive Level: l Memory l
Explanatio
o of Answer
KAt l 001 Al.02 l RO Value: l3.1 l SRO Value: l 3.4 l Section: l SYS l RO Group: l 1 l SRO Group: l1
System / Evolution Control Rod Drive System
Title:
KA Ability to predict and/or monitor changes in parameters associated with operating the Control Rod Drive System
Statement: controls including:
T-ref
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Reactor Control and lOM-1.5.A.51 1 Iss 4 Rev
Protection 0
Reactor Control and 10M 1.1.D 13 iss 4 Rev 13
Protection 1
Full Length Rod Control LP-SQS-1.3 l 7
Question Source l NRC Exam Bank l Question Modification Method l
Question Source Comments: l
Material Required for
Ex:mination:
l I
Page 18
i
'
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. _ _ _ _ _ _ _
5
Giv:n the following conditions:
:
* Reactor Power - 72% '
* Control Rods are at step 210 on Control Bank D
4
* AOP 1.1.1, Failure of RCCA Control Bank to Move, is implemented due to
rod control problems
* The RO incorrectly places the Control Rod Bank Sel Sw in CONTROL BANK D
i
instead of MANUAL
* Rods are withdrawn 5 steps before this is discovered ;
if the Control Rod Bank Sel Sw is placed in Manual at this point, which of the following will occur?
a. Upon shutdown, all Control Bank D rods will remain 5 steps withdrawn from the core.
b'.
Upon shutdown, the ROD BOTFOM/ ROD DROP alarm will actuate 5 steps sooner than expected.
c. While operating, the Rod Insertion L imit alarms (A4-116 and A4-134) for Control Bank D would
actuate 5 steps lower than the actual alarm setpoint positions. !
d. While operating, the Bank Demand Position Indication will read 5 steps lower than the Analog Rod
Position Indication.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
e ef Answer
!
KA: l 001 K4.02 l RO Value: l3.8 l SRO Value: l3.8 l Section: l SYS l RO Group: l 1 l SRO Group: l1
System / Evolution Control Rod Drive System
Title: l
!
KA Knowledge of Control Rod Drive System design feature (s) and or interlock (s) which provide for the following:
i
!
St:t: ment:
Control rod mode r: lect control (movement control) l
Rif;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj I
reactor Control & Protection IOM 1.1.D Bank Overlap 15-16 lss 4 Rev
-Instrumentation and 1
Controls
Full Length Rod Control LP-SQS-1.3 Ill.F.1 13 4 6.a
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Material Required for
Examination:
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*
Question Topic: l Subcooling Margin
During a natur:.1 circulation cooldown the required number of CFDM fans cannot be started.
During the cooldown, upper head voiding is prevented by:
a. venting the head via reactor vessel head vents,
b. verifying incore thermocouple temperatures are within an allowable range ofloop temperatures.
c. increasing the minimum subcooling margin during portions of the cooldown.
d. periodically injecting cold Safety injection water into the Hot legs.
Ans: {c l Exam Level: iS l Cognitive Level: l Comprehension l
Expla:stio
a cf A .swer
KA: l 002 K5.15 l RO Value: l4.2 l SRO Value: l 4.6 l See:lon: }SYS l RO Group: l 2 l SRO Group: l2
System / Evolution Reactor Coolant System
Title:
KA Knowledge of the operational implications of the following concepts as they apply to the Reactor Coolant Sys5mT
Stat: ment:
Reasons for maintaining subcooling margin during natural circulation
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
EOP Generic issues LP-SQS-53.2 1 13
Natural Circulation IOM-53 B.4.ES-0.2 1 23 iss1B
Cooldown Background Rev 4
Question Source l Facility Exam Bank l Question Modification Method l
Qrestion Source Comments: l,
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Ermination:
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Given the fol!owing conditions:
* Plant heatup in progress
* RCS temperature - 175 *F
* RCS pressure - 325 psig
* Pressurizerlevel-28%
* Preparations are underway for the start of the first RCP, RCP 1 A
The requirement of having less than 25 *F difference between SG temperature and the primary system
temperatures:
c. is not applicable since this is the first RCP to be started.
b. prevents an RCS overpressure event.
c. prevents exceeding RCS heatup rates.
d. prevents exceeding RCS cooldown rates.
Ans: lb l Exam Level: lS l Cognitive Level: l Memory l
Expiaratio
o cf Answer
KA: l 003 Kl.10 l RO Value: l3.0 l SRO Value: l 3.2 l Section: lSYS l RO Group: l 1 l SRO Group: l1
System / Evolution Reactor Coolant Pump System
Title:
KA Knowledge of the flysical connections and/or cause-effect relationships between Reactor Coolant Pump System
Statement: and the following:
RCS
Refirence Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Reactor Coolant Pump IOM-6.4.A II.V 3 iss 4 Rev
Startup
7
RCS - Reactor Coolant LP SQS-6.3 Ill.A 24 4 12.A
Pumps
1
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juestion Topic: l RCP power supplies
The reactor is ct 35% with the clectric:1 busses in NSA. Unit Station Service Transformer ID develops a
fault opening [4KV ACB 241D] USST 1D Supply to 1C 4KV Bus and [4KV ACB 341D) USST 1D S
to ID 4KV Bus. The auto bus transfer fails to operate on C & D Bus.
Which of the following lists all running RCPs?
c. RCP 1 A
b. RCP 1 A and IB
c. .RCP IB and 1C
d. RCP IC
l Cognitive Level: l Memory l
Ans: lb l Exam Level: lS
Explanatio
e of Answer
KAt l 003 K2.01
l RO Value: l3.1 l SRO Value: l3.1 l Section: l SYS l RO Group: l 1 l SRO Group: l1
System / Evolution Reactor Coolant Pump System
Title:
KA Knowledge of electrical power supplies to the following:
Statement:
RCPS
Page Number (s) Revision Learn.
Reference Number Reference Section
Reference Obj
4KV Distribution System LP-SQS 36.1 III.B.2 3
1I 4 i
LP-SQS-6.3 1.C.1
Reactor Coolant System -
Remor Coolant Pumps
Question Source l New l Question Modification Method l
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Material Required for
Examination:
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Page 22
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Given th2 following conditions:
* Plant heatup in progress
* RCS temperature - 175 F
* RCS pressure - 325 psig
* Charging pump [lCH-P-1B] is in service.
* Charging pump [1CH-P-1 A] is inoperable.
Which of the following describes limitations, if any, if[1CH-P-1C] were to be placed in service on AE Bus,
and {lCH-P-1B] were to be removed from service?
a.
(ICH-P-1B] must be stopped and placed in PULL-TO-LOCK prior to taking [lCH-P-lC] out of
PULL-TO-LOCK.
b. [1CH-P-1B] must be stopped and placed in AUTO prior to taking [1CH-P-lC) out of PULL-TO-
LOCK.
c. [lCH-P-1B and 1C] may be run simultaneously for up to 15 minutes, after which [1CH-P-1B] must
be stopped and placed in PULL-TO-LOCK.
d. Both Charging Pumps may be run without restriction until [1CH-P-1B]is removed from service.
Ans: lc l Exam Level: lS l Cognitive Level: l Comprehension l
Explaratio
xfAnswer
,KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l4.0 l Section: l SYS l RO Group: l 1 l SRO Group: l1
System / Evolution Chemical and Volume Control System
Title: {
1
1
KA Conduct Of Operations
Statement:
Ability to apply technical specifications for a system.
Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj l
Beav;r Valley - Unit 1 3.4.9.3 3/4 4-27a Amend i
Technical Specifications
No.193
Placing the Spare Charging IOM-7.4.W IV.C W 9-13 iss 4 Rev 12
Pump into Operation 10
CVCS LP-SQS-7.1 IV.A, B 28 12
Question Source jNew l Question Modification Method l
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Mat: rial Required for Technical Specifications
Ermi:stion:
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Question Topic: l Ev:1cfleak in Regen Hx
Given the following conditions:
* Reactor power - 90%
* Pressurizer level- 51% stable
a VCT level- 30% rising
o Letdown flow on [F1-CH-150]- 60 gpm
* Charging flow on [F1-CH-122] - 45 gpm
j
a Seal Injection flows - 8 gpm (A); 10 gpm (B); 7 gpm (C) l
* RCP #1 seal leakoff flows - 4 gpm (A); 4 gpm (B); 2 gpm (C)
Which of the following would result in the conditions above?
a. A leak exists in the Seal Water Heat Exchanger.
b. RCP #1 Seal Bypass Valve [MOV-CH-307] was inadvertently opened.
c. Letdown Pressure Control valve [PCV-CH-145] has failed open.
d. A leak exists in the CVCS Non-Regenerative Heat Exchanger.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
ocf Answer
KAt l 004 K6.07 l RO Value: l2.7 l SRO Value: l2.8 l Section: lSYS l RO Group: l 1 l SRO Group: l1
System / Evolution Chemical and Volume Control System
Title:
KA Knowledge of the of the effect of a loss or malfunction on the following will have on the Chemical and Volume
Statement: Control System:
Heat exchangers and condensers
Reference Section Page Number (s) Revision Learn.
Reference Reference Number
Obj
II.S 13 6 2, 9
CVCS LP-SQS-7.1
Question Source l New l Question Modification Method l
Question Source Comments: l
Material Required for
Examination:
Page 24
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Given the following conditions:
o Plant cooldown is in progress et 20 F/hr
i
a 3
RCS temperature - 155 F
a
!
Pressurizer level [LI-l RC-462] Cold Calib - 100% i
e
RHR Pump 1 A is running with flow of 4000 gpm set on [MOV-RH-605] RHR
Flowin AUTO
a
[MOV-RH-758] Residual Heat Removal Hx FCV demand is set at 40%
*
[MOV-CH-142] RH LTDN to Non Regen Hx Inle Flow Control Viv demand is
set to 75%
* 1
Controller for [PCV-CH-145] LP LTDN Back Press Reg Viv is set in MANUAL at the position that is
maintaining 50 psig with charging flow balanced
If[ HIC-RH-758] controller causes [MOV-RH-758] to close with NO operator action, which of the following
are the results for the first 10 minutes?
i
e. RHR flow will decrease and RCS pressure will decrease.
b. RHR flow will increase and RCS pressure will increase.
c. RHR flow will remain the same and RCS pressure will decrease,
d. RHR flow will remain the same and RCS pressure will increase.
l
Ans: ld l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
n cf Answer
KA: l 005 K3.01 l RO Value: l3.9 l SRO Value: l4.0 l Section: l SYS l RO Group: l 3 l SRO Group: l3
Syst:m/ Evolution Residual Heat Removal System
Title:
KA Knowledge of the effect that a loss or malfunction of the Residual Heat Removal System wi!! have on the
Stat ment: following:
RCS
Ref:rence Reference Nuinber Reference Section Page Number (s) Revision Learn.
Obj
Residualliot Removal lOM 10.4.A E, F A 8-9 iss 4 Rev
System Startup (Plant 9
cooldcwn) And Operation
RHRS LP-SQS-10.1 D.2.e, f 7-8 8 5.a. b, f; 10
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Question Topic: l Loss cf ONE St Accum
Given the following conditions:
* Reactor poweris 55%
* - Accumulator [1SI-TK-1 A] level is 85%
* Accumulator [1SI-TK-1 A] pressure is 657 psig
* SI Accumulator Isolation Valve [MOV-ISI-865A] is closed
* The lockoutjack is removed
* Reactor shutdown was initiated due to the accumulator conditions
Which of the following states the response of the SI Accumulators if a Design Basis LOCA occurs on the
Loop B Cold Leg?
a. THREE Accumulators will fully inject into the core.
b. THREE Accumulators will fully inject into the core, provided the operator manually opens [MOV.
ISI-865A).
c. TWO Accumulators,1B and IC, will fully inject to the core,
d. ONE Accumulator, IC, will fully inject to the core.
Ans: ld l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio The IB Accumulator will discharge through the break
a of Answer
KA: l 006 K6.02 l RO Value: l3A l SRO Value: l3.9 l Section: lSYS l RO Group: l 2 l SRO Group: l2
System / Evolution Emergency Core Cooling System
Title:
KA Knowledge of the of the effect of a loss or malfunction on the following will have on the Emergency Core Cooling
Statement: System:
Core flood tanks (accumulators)
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
SIS LP-SQS-I I .1 Vill.D.7, XI.C.2 18,23 4 7.d,12.a
Question Source l New l Question Modification Method l
_Question Source Comments: l
~5taterial Required for
Ex*miention:
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.__ Reactor is a 100% with di systems in NSA. The operator observes that PRT level has increased.
.
Which of the f;11owing can cause the level increase?
I
a. A relief valve on the CCR system inside containment has lifted.
b. RCP #2 Seal Leak off flow has increased.
c. A PORV is leaking.
d. RCP #1 Seal Leak off flow has increased.
Ans: lc l Exam Level: lS l Cognitive Level: l Memory l
Explanat8.,
a of Armr
KA ( A3.01
l RO Value: l 2.7' l SRO Value: l2.9 l Section: lSYS l RO Group: l 3 l SRO Group: l3
System / Evolution Pressurizer Relief Tank / Quench Tank System
Title:
KA
Ability to monitor automatic operations of the Pressurizer Relief Tank / Quench Tank System including:
Stat: ment:
Components which discharge to the PRT
Reference l Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Alann - Pressurizer Relief lOM-6.4.AAF PC No. 2 AAF 2-3 iss 4 Rev
Tank levelliigh-Low
3
Pressurizer and Pressure LP-SQS-6.4 1.B.2.c 4-5 4 7
ReliefSystems
Reactor Coolant System- LP-SQS-6.3
keactor Coolant Pumps
Question Source l New l Question Modification Method l
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Examination.
j
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Question hpic: l PORV opss: ion [MOV-RC-535] Pressurizer Power ReliefIso!ation Vcive is closed due to [PCV-RC-455C] PORV leaking. [PT-RC-445) Pressurize Pressure has ftiled downscale. Select the available automatic overpressure protection, if any.
s. No PORVs will protect against overpressure.
b. Only PCV-RC-455D will protect against overpressure.
c. Only PCV-RC-456 will protect against overpressure.
. d. - Both PCV-RC-456 and 455D will protect against overpressure.
Aas: Ia l Exam Level: lS l Cognitive Level: l Application l
Espinsatio
e of Answer
KA: l 010 K4.03 l RO Value: l3.8 l SRO Value: l4.1 l Section: l SYS l RO Group: l 2 l SRO Group:_ l2
System / Evolution Pressurizer Pressure Control System
Title:
KA Knowledge of Pressurizer Pressure Control System design feature (s) and or interlock (s) which provide for the
Statement: following:
Over pressure control
Reference Section - Page Number (s) Revision Learn.
Reference Reference Number
06)
Figure 22 iss 4 Rev
lastrument Failure Procedure IOM-6.4-IF
6
4 11
Pressunzer & Pressure Relief LP-SQS-6.4
g . _ .
Question Source - l New l Question Modification Method l
Question Source Comments: l
M;terial Required for IOM-6.4-I F
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Examleation:
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QuestionTopic: l Pressurizer 14 vel Rx trip
Pressurizer Level Control Channel selector is selected to LT 459 & 460. All plant conditions are str.ble.
Which of the following will result in a reactor trip due to high pressurizer level?
e. At 5% power LT-RC-461 fails low,
b. At 5% power LT-RC-459 fails high.
c. At 25'd power LT-RC-460 fails low.
d. At 25% power LT-RC-461 fails low.
A s: lc l Exam Level: lS l Cognitive Level: l Comprehension l
~
Explanatio
c of Answer
KA: l 011 Kl.04 l RO Value: l3.8 l SRO Value: l3.9 l Section: lSYS l RO Group: l 2 l SRO Group: l2
System / Evolution Pressurizer Level Control System
Title:
KA Knowledge of the physical connections and/or cause-effect relationships between Pressurizer Level Control
St:tement: System and the following:
RPS
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
RCS -Instrument failure IOM-6.4.lF ll.a. II.C.I.a IF8-9 Iss 4 Rev
6
Pressurizer and Pressure LP-SQs-6.4 1.D.I.f 9 10 4 12
RelicfSystem
Question Source l New l Question Modification Method l
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Ex :miration:
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Qucstion Topic: l Evil OTDT a OPDT setpoints on input f ilure
-During oper; tion ct 97% pcwer one T,, instrument is re ding 4 degrees higher than other T,, instruments.
All Tm temperatures are equal.
Which of the following describes the effect on OPdeltaT and OTdeltaT for the loop with the highest T,,7
Loop deltaT will be
c. closer to both OPdeltaT and OTdeltaT trip setpoints.
b. closer to its OPdeltaT trip setpoint, but will be farther from its OTdeltaT trip setpoint.
c. farther from its OPdeltaT trip setpoint, but will be closer to its OTdeltaT trip setpoint.
d. farther from both OPdelt f and OTdeltaT trip setpoints.
Ans: ia l Eram Level: lS l Cognitive Level: l Comprehension l _
Explanatio
o of Answer
KA: l 012 A2.05 l RO Value: l 3.l* l SRO Value: l 3.2* l Section: l SYS l RO Group: l 2 l SRO Group: l2
System / Evolution Reactor Protection System
Title:
KA Ability to (a) piedict the impacts of the following on the Reactor Protection System and (b) based on those
Statenient: predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Faulty or erratic operation of detectors and function generators
Reference Reference Number- Reference Section Page Number (s) Revision Learn.
Obj
RCS Instrument Failure IOM-6.4fF ll.B,111. IF 32-33,35 36 iss 4 Rev
6
Reactor Protection System LP SQS-1.1 V.C.16 25-26 6 8
Reactor Coolant System i LP-SQS-6.5 IV.A 17-20 5.a. b
Question Source l Facility Exam Bank l Question Modification Method l
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Eximiration:
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RPS testing is in progress for RPS train B and the status of the breakers are as follows:
i
3
- * Reactor trip beak:rs (RTA and RTB) closed i
* Reactor bypass breaker B (BYB) closed
Bypassing both RPS trains simultaneously is prevented by:
c. tripping only BYA ifit is racked in and its CLOSE pushbutton is depressed.
b. tripping only BYB if BYA is fully racked in.
c. preve iting closure of BYA ifit is racked in,
d. tripping all reactor trip and bypass breakers if BYA is racked in and its CLOSE pushbutton is
depressed.
Ans: ld l Exam level: lS l Cognitive level: l Memory l
Explanatio
o of Answer
KAt l 012 A3.07 l RO Value: l4.0 l SRO Value: l4.0 l Section: l SYS l RO Group: l 2 l SRO Group: l2
System / Evolution Reactor Protection System
Title:
KA Ability to monitor automatic operations of the Reactor Protection System including:
Statement:
Trip breakers
l
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Reactor Control and lO M l.l.B RP,2nd paragraph 2 iss 4,
Protection - Summary Rev.0
Description
Reactor Protection System LP-SQS-1.2 11.1 7 6 8, 9
Hardware
Question Source l Facility Exam Bank l Question Modification Method l
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Material Required for
Exami::stion: .
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Page 31
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Question Topic: l Containment Pressure logics
Containm:nt pressure instrument PT-LM-100C has fiiled downscale. All rpproprir,te tctions of lOM-
1.4.IF, Instrument Failure Procedure, have been completed.
Subsequently PT-LM-100D fails upscale.
Which of the following lists all expected actions?
c. CIA and Si
b. CIA, SI and MSLI
c. CIB and MSLI
d. CIA, CIB, SI and MSLI
A~s: lb l Exam level: lS l Cognitive level: l Comprehension l
Explanatio
e of Answer
KA: l 013 A2.06 l RO Value: l 3.7' l SRO Value: l4.0 l Section: lSYS l RO Group: l 1 l SRO Group: l1
System / Evolution Engineered Safety Features Actuation System
Title:
l KA Ability to (a) predict the impacts of the following on the Engineered Safety Features Actuation System and (b) '
Statement: based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal
operation:
,
inadvertent ESFAS actuation
R;ference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
instrument Failure Procedure 10M 1.4.IF !!.C 4 iss 4 Ra~
l
Reactor Protection Trip LP SQ-1.1 6 9
Logics
l
Question Source l Facility Exam Bank l Question Modification Method j
Question Source Comments: l
M;terial Required for
Ex:mination:
1
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Page 32
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Question Topic: l Operation f;11owing Si signal l
A steam break has occurred causing an SI on high containm:nt pressure. Reactor Trip Breaker BYA will
NOT open. The crew has transitioned to ES-1.1, SI Termination. If containment pressure remains above the
SI setpoint, which of the following will occur if both S1 Reset Pushbuttons are depressed?
c. Neither train of S1 will reset.
b. Only one train of SI will reset, j
c. Both trains of S1 will reset but one train will immediately reinitiate.
d. Only one train of SI will reset. The reset train will immediately reinitiate.
l
Ass: lc l Exam Level: lS l Cognitive Level: l Application l
Expla::stic
o of Answer
KAt l 013 A3.02 l RO Value: l 4.1 l SRO Value: l4.2 l Section: l SYS l RO Group: l 1 l SRO Group: l1
System / Evolution Engineered Safety Features Actuation System j
'
Title:
KA Ability to monitor automatic operations of the Engineered Safety Features Actuation System including: j
Statement:
Operation of actuated equipment
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obi j
FSAR Logic Diagrams Figure 7.2-1 Sheet 8 l
Reactor Protection System LP-SQS-1.1 VI.E.1.f 34-351 6 9
Question Source l Facility Exam Bank l Question Modification Method l
Question Source Comments: l
M:terial Required for Figure 7.21 Sheet 8
Examination:
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Question Tcpic: l ROD BOTTOM clarm
During a reactor startup, when does the ROD BOTTOM / ROD DROP tiarm (A4-126) become cctive for
each control bank?
The alarm will actuate for a dropped rod for:
e. any Control Bank whenever Control Bank A RPI output is above 20 steps.
b. each Control Bank whenever that Control Bank demand position is above 35 steps.
c. Control Banks A, B and C whenever their Control Bank demand position is above 35 steps, and for
Control Bank D whenever Control Bank D demand position is above 20 steps.
d. Control bank A whenever Control Bank A RPI output is above 20 steps, and for Control Banks B, C
and D whenever their Control Bank RPI output is above 35 steps.
Aus: ld l Exam level: lS l Cognitive level: l Memory l
Explanatio
c ef Answer
KAt l 2.4.31 l RO Value: l3.3 l SRO Value: l3.4 l Section: lSYS l R3 Group: l 2 l SRO Group: l1
System / Evolution Rod Position Indication System
Title:
KA Emergency Procedures / Plan
Statement:
Knowledge of annunciators alarms and indications, and use of the response instructions.
Reference Reference Number Reference Section Page Number (s) Revision Ixarn. ;
Obj
Reactor Control & Protection IOM-1.1.B RPI, Ist & 2nd 16 iss 4 Rev
- Summary Description paragraphs 1
RPI and Insertion Limits LP-SQS 1.4 VI.B. C 5-6 5 2.b, c
Reactor Control and lOM-1.2.B 1
Protection Setpoints l
l
Question Source l Previous 2 NRC Exams ~ l Question Modification Method l
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Examination: 1
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Question'Tep6c: l detennin: tion cf NIS counts by IR/SR status
Giv;n the follswing conditions:
* Reactor tripped from 100% power
* Following transition to ES-0.1 " Reactor Trip Response", Intermediate
Range NIS is reading IE-7 amps
* Five minutes later Intermediate range NIS is reading 2.2E-9 amps
How soon following the last reading will Source Range NIS provide correct readings?
c. 4 minutes.
b. 8 minutes.
c. 10 minutes.
d. 13 minutes.
Ams: Ia l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio P=Pi 10((T)(SUR)) Determine SUR form IRNIS readings over 5 minutes which gives SUR = -1/3 dpm (constant
a of Answer rate). This SUR is used with IR activation setpoint - IE-10 gives time of 4.02 minutes.
KA: l 015 K5.06 l RO Value: l3.4 l SRO Value: l 3.7 l Section: lSYS l RO Group: \ . l SRO Group: l1
System / Evolution Nuclear instrumentation System
Title:
KA Knowledge of the operational implications of the following concepts as they apply to the Nuclear Instrumentation
Statement: System:
Suberitical multiplications and NIS indications
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Reactor Excore Inst. System IOM-2,1.C 1R 2nd paragraph 9 iss 4 Rev
. Major Components 1
Excore Instrumentation IV.C.8 10 5 5, 8
LP-SQS-2.1
System
I
Question Source l New l Question Modification Method l
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Examination:
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Question Topic: l Leak in RVOS
A leak has occurred ct the inlet to a RVLIS differential pressure transmitter.
Which of the following describes RVLIS system indication and how the leak will be isolated?
n. RVLIS hydraulic isolator position will indicate a leak has occurred. The leak will automatically
isolate.
I
'
b. RVLIS hydraulic isolator position will indicate a leak has occurred. The leak can only be isolated
;
by closing a manualisolation valve.
c. RVLIS high volume sensor position will indicate a leak has occurred. The leak will automatically
isolate.
d. RVLIS high volume sensor position will indicate :. leak has occurred. The leak can only by isolated
by closing a manual isolation valve.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
e cf Answer
KA: l 016 K3.01 l RO Value: l 3.4* l SRO Value: l 3.6' l Section: l SYS l RO Group: l 2 l SRO Group: l2
System / Evolution Non-Nuclear Instrumentation System
Title:
KA Knowledge of the effect that a loss or malfunction of the Non-Nuclear instrumentation System will have on the
Statement: following:
RCS
Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj
RVLIS Hydraulic isolator IOM-6.4.AG IV.A.7, 8 AG2 iss 4 Rev
Malfunction 0
RVLSI & Core Cooling LP-SQS-6.7 li.B.e, f; !!.G.c; !!.H 4-5,15-17.,22- 1 6
Monitor 23
Question Source l New l Questior. Modification Method l
Question Source Comments: l
M::terial Required for
Ex"mination:
Page 36
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Questioni Topic: l Eval e f Natural Circulation for conditions
Given the foll:; wing conditi:ns:
* A loss of offsite power occurred j
+ A natural circulation cooldown was initiated l
* The five hottest T/Cs average temperature - 555 F
* RCS wide range pressure -1275 psig l
.* All RCS Loop Tu - 552 F
* All RCS Loop Ta - 544 F ,
* All SG pressures - 940 psig
l
Adequate natural circulation flow: (Refer to Att. 6A & 2G)
c. exists and the RCS is subcooled.
b. does not exist and the RCS is subcooled.
c. exists and the RCS is at saturation.
d. does not exist and the RCS is at saturation.
A ns: ib l Exam Level: lS l Cognitive Level: l Application l
Explanatio
a of Answer
KAt l 017 A3.01 l RO Value: l 3.6' l SRO Value: l 3.8' l Section: lSYS l RO Group: l 1 l SRO Group: l1
System /Evt ution in-Core Temperature Monitor System
Title:
KA Ability to monitor automatic operations of the in-Core Temperature Monitor System including:
Statement:
Indications of normal, natural, and interrupted circulation of RCS
R:,f;rence Reference Number Reference Section Page Number (s) Revision Learn.
'
Obj
0 F Plus Subcooling Based 10M-53A.I.6-A I Iss 1B
on Core Exit TCs Rev 2
Natural Circulation EOP Attachment 2-0 1 2 1ssIB
Verification Rev 2
EOP Generic issues LP-SQS-53.2 Vill.C 19 ,
12
Question Source l NRC Exam Bank l Question Modification Method l
Question Source Comments: l
Mr.terial Required for Steam tables, EOP att. 2-G and 6A
Examination:
f
Page 37
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Questies Topic: l Power supply following CIB
The Containment Air Recircul:ti:n fans are in NSA prior to o transient which causes CIB.
After CIB occurs, what will b'.: the status of the Containment Air Recirculation fans?
c. Running in fast speed
b. Running in slow speed
c. Tripped but the power supply is energized
d. Tripped with the power supply deenergized
Ass: ld l Exam level: lS l Cognitive Level: l Comprehension l
Explanatio
o of Answer
KA: l 022 K2.01 l RO Value: l 3.0* l SRO Value: l3.1 l Section: l SYS l RO Group: l 1 l SRO Group: l1
System / Evolution Containment Cooling System
Title:
KA- Knowledge of electrical power supplies to the following:
Statement:
Containment cooling fans
Reference Reference Number Reference Section Page Number (s) Revision learn.
Obj
__
CNMT Vent - Summary lOM-44C.I .B CNMT Air 1 Iss 4 Rev
Description Recirculation 0
Containment Ventilation LP-SQS-44C.I II.A.'l 1-2 4 5,7
Systems
Question Source l New l Question Modification Method l
Question Source Comments: l
M;terial Required for
Ex:mination:
Page 38
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Question Tople: 1 C---h Spray E-,-sse 12 RWST level
Given the f:llswing conditi:ns: i
.'i
* Reactor trip, Si and CIB occurred from 100% power due to a LOCA
>
* RWST level has decreased to 3 feet 9 inches
* CIB has not been reset.
1
What would be the status of the Quench Spray (QS) system?
j
1
(Assume no operator action has been performed in the Quench Spray system.)
'
c. BOTH QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle
Bypass Isol Vivs closed, and TWO QS Chemical Injection pumps are running.
b. BOTH QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle
,
. Bypass Isol Vivs closed, and FOUR QS Chemical Injection pumps are running.
- c.
BOTH QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle
Bypass Isol Vivs open, and TWO QS Chemical Injection pumps are running.
d. BOTH QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle
.
Bypass isol Vivs open, and FOUR QS Chemical Injection pumps are running.
Ans: Ia l Esam level: lS l Cognitive hvel: l Comprehension l
! Esplanatio
e of Answer KA: l 026 Kl.01 l RO Value: l4.2 l SRO Value: l 4.2 l Section: l SYS l RO Group: l 2 l SRO Group: l1 System / Evolution Containment Spray System Title: !
, ! KA 1
Knowledge of the physical connections and/or cause-effect relationships between Containment Spray System and j
l- Statement: the following:
!
! ECCS
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
,
Les cfreactor Or Secondary El step 30 22 issIB
l Coolant
Rev 4
( Transfer to Cold Leg ES-1.3 step 6 6 issIB i
Recirculation
Rev 4
l CNMT Depressurization LP-SQS-13.1 V.D.1 17-18 5.b ! System ,
Question Source l New l Question Modification Method l
l Question Source Comments: l
Materi;I Required for
Esimitation:
,
i
Page 39
1
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- - - - _ _ _ _ _ _ _ -_ ____ _
* *
Cries Topict l Recombiner Ops
Given the f;llowing conditions:
* A LOCA has occurred 24 hours ago
* ONE Hydrogen recombiner is placed in service when hydrogen concentration reaches 0.5%
With a recombiner in operation, containment pressure:
a. should be maintained at approximately 8.9 PSIA, to prevent excessive recombiner flow.
b. will be adequate for recombiner operation ifit is maintained between 8.9 PSIA and -3 PSIG
c. should be maintained slightly above atmospheric, to ensure sufficient recombiner flow.
d. should be maintained at approximately -2PSIG, to ensure sufficient recombiner flow.
l Esam level: IS l Coenitive Ixvel: l Application l
Ass: Ie
Explanation
of Assmr
KAt l U,A Al.01 l RO Value: l3.4 l SRO Value: l 3.8 l Section: l SYS l RO Group: l 3 l SRO Group: l2
System / Evolution Hydrogen Recombiner and Purge Control System
Title:
KA Ability to predict and/or monitor changes in parameters associated with operating the Hydrogen Recombiner and
Statement: Purge Control System controls including:
Hydrogen concentration
Reference Number Reference Section Page Number (s) Revision Learn.
Reference
Obi
. Post DBA Hydrogen Control 10M-46.1.B 4th paragraph 1 Iss 44:
Rev.0
System - Summary
Description
11.C.2.d 7 3 8,9
Post DBA H2 Control LP SQS-46.1
System
System
Question Source l New l Question Modification Method l
Carlon Source Comments: l
Material Required for OM 46.4.A
Examination:
Page 40
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- _ - _ _ _ _ _ _ _ - _ _ _ _ IQcestion Topic: l Evnluation c f a leak ) Given the f;112 wing conditions: I )o Reactor power is 85% l c Spent Fuel Pool is aligned for cooling
o
A leak has occurred in the suction of[FC-P-1 A] Fuel Pool Cooling Pump
If the leak remains unisolated, Spent Fuel Pool level should stabilize at:
!
c. ~25 feet above the top of the fuel.
b. ~23 feet above the top of the fuel.
c. ~10 feet above the top of the fuel.
d the top of the fuel.
Ans: lc l Exam Level: lS l Cognitive Level: l Memory l
Explanatio ,
c of Answ:r *
KA: l 033 A2.03 l RO Value: l3.1 l SRO Value: l3.5 l Section: lSYS l RO Group: l 2 l SRO Group: l2
System / Evolution Spent Fuel Pool Cooling System
Title:
KA Ability to (a) predict the impacts of the following on the Spent Fuel Pool Cooling System and (b) based on those
Statement: predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Abnormal spent fuel pool water level or loss of water level
Reference Reference Number Reference Section Page Number (s) Revision Learn.
*
[
Obj t .
Fuel Pool Cooling and IOM-20.1.B 3 iss 4 Rev I
Purification 3
Fuel Pool Cooling and LP-SQS-20.1 9 6,9b
Purification
I
Question Source l New l Question Modification Method l
Questici Source Comments: '
f
l
Material Required for
Ermination:
I
,
l
!
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l
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Page 41
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l
*
Question Topic: l Transfer Cart Operation Which cf f:ll: wing describes the interlock between the conveyot car drive and the upenders wh:n i transferring the conveyor car from the transfer canal to the refueling cavity? j
a. Both upenders must be in the down position before the conveyor car can be moved.
b. Only the upender in the refueling cavity must be in the down position before the conveyor car can be
moved.
c. Only the upender in the transfer canal must be in the down position before the conveyor car can be
moved.
d. If upender in the refueling cavity is not in the down position, movement of the conveyor car can be
initiated, however the conveyor car will stop before reaching the upender.
Ans: la l Exam Level: lS l Cognitive Level: l Memory l
Explanatio
o of Answer
KA: l 034 K4.02 l RO Value: l2.5 l SRO Value: l3.3 l Section: lSYS l RO Group: l 3 l SRO Group: l2 System / Evolution Fuel Handling Equipment System
Title:
KA Knowledge of Fuel Handling Equipment System design feature (s) and or interlock (s) which provide for the
Statement: following:
Fuelmovement
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
~ Fuel Handling Operations LP-SQS-6.13 XI.H.9.e 32 4 8.a
1 RP-12R-3.2 11.6.6 2 Iss 0 Rev
0
Question Source l NRC Exam Bank l Question Modification Method l
Question Source Comments: l
Material Required for
Ermination:
Page 42
_______________
. . . . - - _ . . -
Reactor power is 25% and all plant systems are in NSA.
Which failure would decrease feedw:t:r flow to all SGs?
c. ONE condenser steam dump fails open.
b. Heater Drain receiver Level Control Valve [LCV-ISD-106B] fails open.
c. Turbine First Stage Pressure channel [PT-1MS-446] fails low.
d. Combined Feedwater Header Pressure channel (PS-lFW-151] fails high.
Ans: lc l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
e of Answer
KAt l 035 Kl.01 l RO Value: l4.2 l SRO Value: l4.5 l Section: l SYS l RO Group; l 2 l SRO Group: l2
System / Evolution Steam Generator System
Title:
KA Knowledge of the physical connections and/or cause-effect relationships between Steam Generator System and the
Statement: following:
MFW/AFW systems
Reference Reference Number Reference Section Page Number (s) Revision Learn. ]
Obj
SO Feedwater System - IOM-24.l D l
SGWLC 7-8 Iss 4 Rev l
Instrumentation and Controls 2
SG Feedwater System - IOM-24.4.lF Attachment 5, ll.A.2 IF 38 iss 4 Rev
Instrument Failure 2
Feedwater System LP-SQS-24.1 Ill.E.10.d 14 1.A
Question Source l New l Question Modification Method l
Question Source Comments: l
Mat; rial Required for
Examination:
)
l
l
l
l
l
l
4
r l
Page 43 _ _ _ - _ - _ _ _ _ _ _ _
1
* *
Question Topic: l Effect cf MS- PT-464 fdling high
Given the follswing conditions: l
l
'
* The unit is in MODE 3 preparing for normal plant cooldown
* Condenser Steam Dump System is automatically controlling T, at 547 *F in Steam Pressure Mode
* [PT-1MS-464] Main Steam Header Pressure fails high
l
Which one of the following describes the effect this will have on the Condenser Steam Dump system?
c. Two banks of steam dumps will open and remain open until manually closed.
b. Two banks of steam dumps will open but should reclose with no operator action.
l
c. All banks of steam dumps will open and remain open until manually closed.
d. All banks of steam dumps will open but should reclose with no operator action.
Ams: lb l Exam level: lS l Cognitive level: l Comprehension l
Explanatio
a of Answer
KA: l 041 K6.03 l RO Value: l2.7 j SRO Value: l 2.9 l Section: lSYS l RO Group: l 3 l SRO Group: l3
System / Evolution Steam Dump System and Turbine Bypass Control
Title:
KA Knowledge of the of the effect of a loss or malfunction on the following will have on the Steam Dump System and
Statement: Turbine Bypass Control:
Controller and positioners, including ICS, S/G, CRDS
Refensee Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Main Steam System IOM-21.5.A.24 1 iss 4 Rev
0
M in Steam System LP-SQS-21.1 4 3
Question Source l New l Question Modification Method l
Question Source Comments: l
Material Required for
Ex:mination:
1
Page 44
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Question Tc pic: l NPSH for FW
Given the following conditions:
* Reactor power - 100%
*
A load rejection occurs and the plant stabilizes at 45% power
*
Load rejection bistables " LOAD REJ 15-50%" and "LOAO REJ GREATER THAN 50%"
are lit
How are the Steam Generator Feed Pumps [lFW-P-1 A,1 B] protected from a loss of suction pressure during
the load rejection?
j
e. The Feedwater Heater Bypass Valve [TV-1CN-100] opened and closed FOUR minutes later.
b. The Heater Drain Receiver Level Control Valve [LCV-1 SD-106B] was maintained fully open until
j
LOW-LOW level was sensed in the Heater Drain Receiver. j
c. The Heater Bypass to Heater Drain Pump Suction Valve [TV-CN-125] opened and closed four
minutes later,
d. The Condensate Pumps Recirculation Valve [FCV-lCN-101] closed on the 15-50% load rejection l
and reopened FIVE minutes later.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
o cf Answer
KA: l 056 Al.08 l RO Value: l2.3 l SRO Value: l2.6' l Section: jSYS l RO Group: l 1 l SRO Group: lI
Syst:m/ Evolution Condensate System
Title:
KA Ability to predict and/or monitor changes in parameters associated with operating the Condensate System controls
Stat: ment: including:
MFW pump suction pressure
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Load Rejection AOP 1.35.2 step 11 7 iss 3A
Rev 6
Figure 22-6 - Step Load lOM-22.5.A.6 1 Iss 4 Rev
Rejection Ckt 0
Extraction Steam and Heater LP-SQS-23 lil.C 7 8-9 12.E
Dra'ms
Question Source l Other Facility l Question Modification Method l
Q:estion Source Comments: l
Mat: rial Required for
Ex:mination:
i l
Page 45 _ _ _ _ _ _ _ - _ _ _ _ _ __
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. .
Q:estio2 Tepic: l Restor: tion of FW capability
i
An inadvertent SI signal occurred at 100% power. The condition causing the Si signal is no longer present. I
I
All systems function as designed and RCS conditions stabilize as expected following the inadvertent SI.
Which of the following states the condition (s) that would have to be met to feed via [FCV-lFW- l
479(489)(499)], SG FW Bypass FCVs? l
a. Only the FWI FW BYPASS VALVE RESET pushbuttons would have to be depressed.
I
b. P-4 would have to be cleared and the FWI FW BYPASS VALVE RESET pushbuttons would have
to be depressed.
c. SI would have to be reset and the FWI FW BYPASS VALVE RESET pushbuttons would have to be
depressed.
d. SI would have to be reset, P-4 would have to be cleared and the FWI FW BYPASS VALVE RESET
pushbuttons would have to be depressed.
Ars: la l Exam Level: lS l Cognitive Level: l Application l-
Explanatio
a e f Answer
KA: l 059 A4.11 l RO Value: l3.1 l SRO Value: l3.3 l Section: l SYS l RO Group: l 1 l SRO Group: l1
System / Evolution Main Feedwater System
Title:
KA Ability to manually operate and/or monitor in the control room:
Statement:
Recovery from automatic feedwater isolation
R:f rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Feedwater System LP-SQS-24.1 III.E.I1. 15-16 7 5.j,
7.A.(12)
Reactor Protection Systems LP-SQS-1.1 V1.E.5 38-39 6 9
Updated FSAR Figure 7.2-1 sheet
1 & 13
Question Source l Facility Exam Bank l Question Modification Method l
Q:estion Source Comments: l
M:.t: rial Required for Figure 7.2-1 sheet 1 & 13
Ex:mination:
l Page 46 N--__-________.
. _ _ _ . _ _ _ _ _ _ _ _ _ ._ . _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _
l . . i
Questio2 Tcpic: l SGWLC inputs
Given the following conditions:
L
* . Reactor power is 20%
'
( Feedwater has been transferred to the Main Feed Regulating Valves
j * All systems are NSA l + f Narrow Range SG IC levelis 44% j' . + [FCV-IFW-499] 1C SG FW Bypass Viv is manually opened 15%
AAer plant conditions stabilize, which parameter (s) will be different from those prior to [FCV-1 FW-499]
opening?
l c. Only [FCV-IFW-498] IC Main FW Reg Viv position .
b [FCV-lFW-498] IC Main FW Reg Viv position and Narrow Range SG IC Level
'
c. Only Narrow Range SG 1C Level
! d. Narrow Range SG 1C level and Stm Gen 1C Feed Flow indication
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension - l
Explanatio .
e of Answer
- KA
- l 059 Kl.04 l RO Value: l3.4 l SRO Value: l3.4 l Section: l SYS l RO Group: l 1 l SRO Group: l1
System / Evolution Main Feedwater System
Title:
l
KA Knowledge of the physical connections and/or cause-effect relationships between Main Feedwater System and the i
Statement: following:
S/GS water level control system
.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj ,
SG Feedwater System - IOM 24.1.D SGWLC 7-8 Iss 4 Rev
.
l
listrumentation and Controls 2
Feedwater System - LP-SQS-24.1 Ill.E.10. 14 15 7 1.A
,
Question Source l New l Question Modification Method j
Question Source Comments: l
Material Required for
Esanlaation:
)
! Page 47 L_____-_________. .i
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- .
Question Tcpic: l R,1 tionship of AFW stram supply & feed supplies to SG
Given the following conditions:
* Reactor power - 100%
* A loss of all AC power occurs )
. Auxiliary Fced Pump IFW-P-2 starts and runs !
* The steam supply line from SG B to 1FW-P-2 ruptures at the connection
to the main steam line. ;
,
* The steam break prevents access to the Main Steam Valve Room
Which of the following describes how the Auxiliary Feed System is affected by the above conditions?
c. All SGs will blowdown through the rupture, and NO auxiliary feed will be available.
b. SG A and SG B will blowdown through the rupture, but NO auxiliary feed will be available.
c. SG A and SG B will blowdown through the rupture, but auxiliary feed can be established by
opening the manual steam supply isolation valve from SG C.
d. Only SG B will blowdown through the rupture, and auxiliary feed can be established from SG A.
Ans: ld l Exam Level: lS l Cognitive Level: l Memory l
Explanatio
e of Answer
KAt l 061 K3.02 l RO Value: l4.2 l SRO Value: l 4.4 l Section: l SYS l RO Group: l 1 l SRO Group: l1
System / Evolution Auxiliary / Emergency Feedwater System
Title:
KA Knowledge of the effect that a loss or malfunction of the Auxiliary / Emergency Feedwater System will have on
Statement: the following:
S/G
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
SG Feedwater System lOM-24.1.C Auxiliary Feed Pumps 2-3 iss 4;
Rev 2
Feedwater System I,P-SQS-21.1 Ill.J.9 20 7 1.B
SG Feedwater System LP-SQS-21.1 Ill.L.3.a 22
Question Source lNew l Question Modification Method l
Question Source Comments: l
Material Required for
Ex:mination:
1
Page 48
l u__.
)
Question Tcpie: l Overcurrent eff:ct on br;aker operation
The Unit is ct 85M. Which of the following conditions will result in bus I AE being maintained deenergized.
l
c. [ACB-1 A10] 1 AE Emergency Bus feeder breaker trips on overcurrent.
b. l AE Ernergency Bus reverse phase PT blows a fuse.
c. [ACB-41C] 1 A Normal 4KV Bus Feeder Breaker trips on overcurrent.
d. [ACB-41C) l A Normal 4KV Bus Feeder Breaker trips on Unit Station Service Tranformer 1C
Differential Trip.
]
Ans: la l Eram Level: lS l Cognitive Level: l Comprehension l
'
Expla;atio j
o of Answer
'
KAt l 062 K4.01 l RO Value: l2.6 l SRO Value: l3.2 l Section: l SYS l RO Group: l 2 l SRO Group: l2
System / Evolution A.C. Electrical Distribution l
l
Title-
1
I
KA Knowledge of A.C. Electrical Distribution design feature (s) and or interlock (s) which provide for the following:
Statement:
Bus lockouts
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
4160V Emergency Bus I AE IOM-36.4.ACZ lss 3 Rev
ACB-1 A10 Auto Trip 1 I
Diesel Generators LP-SQS-36.2 8 6 I
Question Source { Previous 2 NRC Exams l Question Modification Method l
Q estion Source Comments: l
M:.t: rial Required for
Ex:mination:
i
Page 49
_. _ _ _ _ _ _ _ _ _ . _ _ -
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. .
Q estio2 Tepic: l Breiker interlock (s)
React:r power is 25% during a startup. Electrical loads have been transferred to the Unit Station Service
Transformer (USST).
In crder for Bus I A to be setup for Auto Bus Transfer to the System Station Service Transformer, which of
the f:llowing lists the required position of the Live Bus Transfer switch and the control switch for ACB
41A7
l
c. Live Bus Transfer Switch - OFF
ACB 41 A Control Switch- After Close
b. Live Bus Transfer Switch - OFF
ACB 41 A Control Switch - After Trip
c. Live Bus Transfer Switch - ON
ACB 41 A Control Switch - After Close
d. Live Bus Transfer Switch - ON
ACB 41 A Control Switch - After Trip
Ars: la l Exam Level: lS l Cognitive Level: l Memory l
Expla:atio
a ef Aiswer
KA: l 062 K4.01 l RO Value: l2.6 l SRO Value: l3.2 l Section: lSYS l RO Group: l 2 l SRO Group: l2
System / Evolution A.C. Electrical Distribution
Title:
KA Knowledge of A.C. Electrical Distribution design feature (s) and or interlock (s) which provide for the following:
Statement:
Bus lockouts
Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj
4KV Station Service System IO M 36.1.E 20-21 iss 4 Rev
- Specific Instrumentation I
and Controls
4KV Distribution LP-SQS-36.1 45 7 3.
Question Source { New l Question Modification Method {
Q estion Source Comments: l
Material Required for
Ex mination:
Page 50
DC Bus 1-2
oper:tions. ground voltmeter went from 0 volts to -165 volts. The DC Bus is in NSA for 1004
Which of the following describes the effect the ground will have on DC bus operations?
a.
The ground has caused actual voltage to the DC loads to decrease to 105 Volts.
b. The affected battery will discharge significantly faster than designed.
c. The bus will operate as required but the bus reliability has decreased.
d. Another ground on the same polarity of the bus will cause a short circuit.
Ans: lc l Exam Level: lS l Cognitive Level: l Memory l
Explanatio
e of Answer
KAt l 063 A2.01
l RO Value: l2.5 l SRO Value: l 3.2* j Section: l SYS l RO Group: l 2 l SRO Group: l1
System / Evolution D.C. Electrical Distribution
Title:
KA
Ability to (a) predict the impacts of the following on the D.C. Electrical Distribution and (b) based on those
Statement:
predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Grounds j
Reference Reference Number Reference Section Page Number (s) Revision 1.*a rn.
125 V DC Control System-
Obj
IOM-39.2 A.16 2 iss 3 Rev
Precautions & Setpoints ,
0
125 V DC Control System IOM-39.1 3
{
iss 4 Rev '
0
125 VDC LP-SOS-39.1
I
Q 'estion Source l New l Question Modification Method l
Question Source Comments: l
i
M:terial Required for l
Ex:mination: {
i.
l
l
Page5I
L.____._____ _
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. .
Question Tepic: l Rev;rse pow r trip of DG Diesel Generator No.1 is paralleled to 4160V Bus 1 AE for testing. The operator is in the process of adjusting load and voltage when the Governor Control switch sticks in the LOWER position. If NO operator action is taken, what will be the Diesel Generator response to this condition? DO frequency will:
c. decrease and the diesel will trip on reverse power.
^
b. decrease and the diesel will trip on overcurrent.
c. remain constant but the diesel will trip on reverse power.
d. remain constant but the diesel will trip on overcurrent.
l Cognitive Level: l Comprehension l
Aas: lc l Exam Level: lS
Exploratio
o of Answer
l RO Value: l3.1 l SRO Value: l3.4 l Section: l SYS l RO Group: l 2 l SRO Group: l2
_KAt l 064 Al.08
System / Evolution Emergency DieselGenerators
Title:
KA Ability to predct and/or monitor changes in parameters associated with operating the Emergency Diesel
Statement: Generators controls including:
Maintaining minimum load on ED/G (to prevent reverse power)
Reference Section Page Number (s) Revision Learn.
Reference Reference Number
Ob}
IV.A.9 & CAUTION Q2 iss 4 Rev
Transferring Emergency lOM 36.4.Q
3
Feed
Transferring Emergency
Busses 1 AE And IDF From
Emergency Feed To Normal
Feed
A8-127 ADU1 Iss 3 Rev
Alarm DIESEL- lOM-34.ADU
1
GENERATOR NO. I
REVERSE POWER 6
VI.13 29
Diesel Generators LP-SQS-36.2
Question Source l New l Question Modification Method l
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Q:estion Topic: l DieselGeneratorTrips
A loss cf off site power occurred and the diesel generators are supplying the emergency buses.
Which of the following will trip a diesel generator?
c. The govemor control switch in the control roorn is held in the RAISE position,
b. A governor failure causes engine speed to increase to 1050 RPM.
c. Thejacket cooling water pump trips.
d. The coupling fails on the lube oil pump.
Ans: lb l Exam Level: lS l Cognitive Level: l Memory l
Expla:atto
oe f A swer
KA: l 064 K4.02 l RO Value: l3.9 l SRO Value: l4.2 l Section: l SYS l RO Group: l 2 l SRO Group: l2
System / Evolution Emergency Diesel Generators
Title:
KA Knowledge of Emergency Diesel Generators design feature (s) and or interlock (s) which provide for the following:
Stat: ment:
Trips for ED/G while operating (normal or emergency)
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Local. Overspeed Trip IOM-36.4.AFN 1 1ss 3 Rev
1
Diesel Generators LP-SQS-36.2 8 6
Technical Specifications 4.8.1.1.2.b.4 3/4 8.4a
Question Source l Facility Exam Bank l Question Modification Method l
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Question Ttpic: l Drain Tank Isolation
Given the following conditions:
. Low Level Waste Drain Tank level is 110 inches
* The di charge permit has been approved at discharge rate of 15 gpm
. The discharge is in progress at 15 gpm
What condition will automatically stop the release?
c. Both (TV-LW-105] Liquid Waste Emuent Trip valve and [FCV-LW-104-2] High Range Liquid
Waste Emuent Flow Control Valve closing on high-high radiation signal from [RM-LW-104].
b. [FCV-LW-104-2] High Range Liquid Waste Emuent Flow Control Valve closing on low flow rate,
c. [FCV-LW-104-1] Low Range Liquid Waste Emuent Flow Control Valve closing on low Waste
' Drain Tank level.
d. The Low Level Waste Drain pump tripping on low flow rate.
Ans: la l Enam Level: lS l Cognitive Level: l Memory l
Explanatio
e af Answer
KAt l 068 A4.04 l RO Value: l3.8 l SRO Value: l3.7 l Section: l SYS l RO Group: l 1 l SRO Group: l1
Systent/ Evolution Liquid Radwaste System
Title:
KA Ability to manually operate and/or monitor in the control room:
Stat: ment:
Automatic isolation
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Liquid Waste Disposal LP-SQS-17.1 II.C.7,8 & 10 11-13 3 2.b
System
Question Source l New l Question Modification Method l
Question Source Comments: l
M tirial Required for
Ex mination:
L 1 i
Page 54
t.
___ ___ ____ _ __-_ _ - _ _ Question Tepic: l AnnunciItorOperation Due to a Steam Generator Tube Leak a Condenser Air Ejector Vent Monitor [RM-ISV-100] High r.larm i occurs causing Annuciator " Radiation Monitoring High"(A4-71) alarm to be received. Annuciator (A4-71)
, is acknowledged. Which of the following will cause Annuciator" Radiation Monitoring High"(A4-71) to . ! reflash? I
a. Condenser Air Ejector Vent Monitor [RM-ISV-100] rising to the High-High alarm setpoint.
b. Steam Generator Blowdown Sample Monitor [RM-ISS-100] rising to the High alarm Setpoint.
c. Steam Generator N-16 Monitor [RM-1MS-102] rising to the High alarm Setpoint.
d. High Capacity Steam Generator Blowdown Monitor [RM-1BD-101] rising to the High alarm
Setpoint.
Ass: lb l Exam Level: lS l Cognitive IAvel: l Memory l
Explaxtio
o of Arswer
KA: l 073 A4.02 l RO Value: l3.7 l SRO Value: l3.7 l Section: lSYS l RO Group: l 2 l SRO Group: l2
System / Evolution Process Radiation Monitoring System
Title:
KA Ability to manually operate and/or monitor in the control room:
Statement:
Radiation monitoring system control panel
Ref;reIce - Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Rad Monitoring System - lOM-43.1.D 10 iss 4 Rev
Instrumentation and Contro!s 3
RadiItion Monitoring System LP SQS-43.1 1
Qyestion Source l New l Question Modification Method j
Qrestion Source Comments: l
Miterial Required for
Examination:
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Question Tepic: l Evaluati:n of avail ble air sources
A leak has occurred in the Station Air System in the Fuel Building. [PI-lSA-101] Station Air Main Header
and [PI-llA-106) Station Instrument Air Header pressure indications are both lowering.
! When Station Air pressure decreases to a specific setpoint, [TV-lSA-105] Station Air Header Trip Valve l will:
a. open to supply instrument air loads.
b. open to supply containment air loads.
c, close to ensure all station air will be supplied to the instrument air loads.
d. close to maintain air to all station loads.
A s: lc l Exam Level: lS l Cognitive Level: l Memory l
Explanatio
e cf Answer
KA: l 078 K4.02 l RO Value: l3.2 l SRO Value: l3.5 l Section: l SYS l RO Group: l 3 l SRO Group: l3
System / Evolution Instrument Air System
Title:
KA Knowledge ofinstrument Air System design feature (s) and or interlock (s) which provide for the following:
Statement:
Cross-over to other air systems
Ref rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Compressed Air Systems - IOM 34.1.D Station Air Header Trip 5 iss 4 Rev
Instmmentation and Controls D Valve 0
VOND 34-1
Compressed Air LP SQS-34.1 IV.A & D 15 5
Question Source l New l Question Modification Method l
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Mit; rial Required for
Examination:
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,Questiga Topic: l Containm:nt Building Pznett:tions during r2 fueling
Which of the following is NOT part of the Technicr_1 Specification defmition of CONTAINMENT
INTEGRITY ~
a. ' The containment leakage monitoring system is OPERABLE.
b. All equipment hatches are closed and sealed,
c. . The sealing mechanism associated with each penetration is OPERABLE.
d. The containment leakage rates are within their LCO limits.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
o ef Answer
KA: l 103 KI.02 l RO Value: l3.9 l SRO Value: l 4.I' l Section: lSYS l RO Group: l 3 l SRO Group: l2
System / Evolution Containment System
Title:
KA -
Knowledge of the physical connections and/or cause-effect relationships between Containment System and the
Statement: following:
Containment isolation / containment integrity
Ref;rence Reference Number Reference Section Page Number (s) Revision Learn. j
Obj
7echnical Specification 3/4.9.4 3/49-4 l
Containment System LP-SQS-47.1 VI.B 20 4 8.h
Question Source l New l Question Modification Method l
Qrestion Source Comments: l
M;terial Required for
Examination:
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Page $7
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Question Tcpic: l Determination c f power increase
Given the following conditions:
* EOL
* Reactor poweris 80% steady state
a RCS T,,,is on program l
* Control Rod position - 160 steps on Control Bank D
* Control Rods begin to withdraw
* When Control Bank D is at 170 steps the Control Rod Bank Sei Sw is
placed in MANUAL stopping rod motion
If NO further operator action is taken, what would be the afrect on actual power level and RCS T,,, af
conditions stabilize?
a. Reactor power and RCS T,,, would both rise equally by an amount equivalent to the reactivity
addition.
b. Reactor power would rise by an amount equivalent to the reactivity addition and RCS T,,, wou
remain approximately 571 F.
c. Reactor power would remain approximately 80% and RCS T,,, would rise by an amount equiv
to the reactivity addition.
d. Neither r.: actor power nor RCS T,,, would be significantly affected.
l Cognitive Level: l Comprehension l
Ans: lc l Exam level: lS
Explanatio Reactivity addition by rod movement would add power to RCS. Since turbine load c~ntrols power leve
a c f An:wer
RCS would heat up. By using Power defect curves could detennine the equivalent power level the reac
would allow and the associated Tavg at that pawer will approximate the temperature of the RCS (Use of Power
Defect Curves provides an approximation because it includes Fuel temp / Doppler coefficient, but impact is
relatively small compared to moderator temp coefficient over area of concem)
KA: l 001 AKl.03 l RO Value: l3.9 l SRO Value: l4.0 l Section: lEPE l RO Group: l 2 l SRO Group: l1
Syst:m/ Evolution Continuous Rod Withdrawal
Title:
KA Knowledge of the operational implications of the following concepts as they apply to Continuous Rod
Statement: Withdrawal:
Relationship of reactivity and reactor power to rod movement
Page Number (s) Revision Learn.
Ref;rence Reference Number Reference Section
Ob]
4 16
Full Length Rod Control LP-SQS-1.3
Question Source l New l Question Modification Method l
I QIestion Source Comments: l
Miterial Required for
Ex;mination:
l
Page 58
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Question Ttpic: l Operation cf Disconnect Switch
Given the following conditions:
* Reactor power - 5%
* Control rod F-6 in Control Bank D has fully dropped.
* Recovery of the dropped rod ' in progress per AOP 1.1.5 " Dropped RCCA"
= All Disconnect Switches S antrol Bank D are in DISCONNECT except for F-6
Which of the following describes alarms that will be received and their effect on recovering the dropped
control rod?
a. An urgent failure will be received, however rod recovery can proceed with the Control Rod Bank
Sel Sw in Control Bank D.
b. An urgent failure will be received, however rod recovery can proceed with the Control Rod Bank
Sel Sw m Manual.
c. A non-urgent failure will be received which will not affect control rod movement.
d. An urgent failure will be received, however rod recovery can proceed after depressing the Rod
Control Alarm Reset pushbutton.
A*s: la l Exam level: lS l Cognitive Level: l Comprehension l
Explanatio
ocf Answer
KA: l 003 AK2.05 l RO Value: l2.5 l SRO Value: l2.8 l Section: l EPE j RO Group: l 2 l SRO Group: l1
System / Evolution Dropped Control Rod
Title:
KA Knowledge of the interrelations between Dropped Control Rod and the following:
Statement: )
Control rod drive power supplies and logic circuits
Ref;rence Reference Number Reference Section Page Number (s) Revision Learn. l
Obj
Dropped RCCA AOP 1.1.5 11 5 iss 3A
Rev 7
Alarm - ROD CONTROL lOM l.4.AAR A4-105 Conective AARI 1ss3 Rev
SYS'EM URGENT Action NOTE 2
FAILURE
Full Length Rod Control LP-SQS-1.3 II.G.3 & IV.A.3 14 & 16 10;16
Quest 6n Source l New l Question Modif; cation Method l
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Ex:mination: ,
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* Q r.stion Tcpic: l Operational limits & basis with given stuck rod Given the following conditions:
L l- '
* Reactor power- 85%
sa Load increase is in progress '
* Control Bank D is 2 steps above the RIL
* Control rod K-6 indicates 15 steps below the remaining rods in Control Bank D
* Control rod trippability is confirmed
1
)
* Shutdown Margin is verific : ~ be satisfied
If the NSS decides to continue power operation with the control rod misaligned, which of the following
describes required power reduzan and the associated reason?
Reactor power must be reduced to at least:
a. 75% power within ONE hour to remain in compliance with Rod Insertion Limit restrictions.
b. 75% power within ONE hour to provide assurance of fuel rod integrity during continued operations.
c. 50% power within FOUR hours to remain in compliance with Rod Insertion Limit restrictions.
di 50% power within FOUR hours to provide assurance of fuel rod integrity during continued
operations.
Ans: lb l Exam level: lS l Cognitive Level: l Application l
Explanatio
o cf Answer
KA: l 005 AKl.06 l RO Value: l2.9 l SRO Value: l 3.8 l Section: l EPE l RO Group: ll l SRO Group: l1
System / Evolution inoperable / Stuck Control Rod
Title:
KA Knowledge of the operational implications of the following concepts as they apply to inoperable / Stuck Control
Statement: Rod:
Bases for power limit, for rod misalignment
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Be:ver Valley - Unit 1 3.1.3.1 (ACTION C.3) 3/4118-19 Amend
Technical Specifications ~
No.154
Beaver Valley - Unit i Bar,= 55 ' 'i B 3/4 1-4 Amend
Technical Specifications No.141
Full Length Rod Control LP-SQS-1.3 111.1.1 15 15
Qrestion Source l NRC Exam Bank l Question Modification Method l
Question Source Comments: l
M terial Required for Technical Specifications
Ex mination:
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Page 60
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QuestioA Topic: l Steam Dump Affects
Given the following conditions:
* Reactor tripped from 100% power
*
Reactor trip breaker (RTB), which provides P-4 input to Reactor Trip
Controller, CANNOT be opened after the trip
* R actor trip breaker (RTA) opened
Which of the fo!!owing identifies where the RCS temperar tre should stabilize prior to placing the Steam
Pressure Mode Selector Switch in Steam Pressure Mode"
e. 543 F.
b. 547 F,
c. 549 F.
d. 554 F.
A':s: {c l Exam Level: lS l Cognitive Level: l Comprehension l
Explaratio
c of A swer
KAt l 007 EA2.03 l RO Value: l4.2 l SRO Value: l 4.4 l Section: l EPE l RO Group: l 2 l SRO Group: l2
System / Evolution Reactor Trip
Title:
KA Ability to determine and interpret the folicwing as they apply to Reactor Trip:
Statement:
Reactor trip breaker position
Refmace . Reference Number Reference Section Page Number (s) Revision Learn.
Obj
M in Steam Systems IOM 21.5.A.24 1 iss 4 Rev
0
Main Steam System - lOM 21.1.D various 3-6 iss 4 Rev j
Instrumentation and Controls 1
Main Steam Supply / Steam LP-SQS-21.1 lil.D, Ill.E, V.C.5, 12-14,27-28, i .e, 3.a
Dump System V.E.1 30-31
Question Source l New l Question Modification Method l
Question Source Comments: l
M:t; rial Required for
Ex mination:
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Page 61
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' *
Question Topic: l Operation cf control rods during an ATWS
A manual reactor trip was initiated at 100%, however the reactor will not trip. Step 1 of FR-S.1 is being
performed. Control rods are in AUTOMATIC.
With the turbine tripped, which of the following describes required action concerning control rod insenion?
Contml rods should be inserted in:
l' c. MANUAL even if they are insening in AUTOMATIC.
b. AUTOMATIC provided rods are inserting in AUTOMATIC.
c. ! AUTOMATIC until reactor power is less than 15% where the rods will '
stop, requiring MANUAL
insertion.
d. AUTOMATIC until the Rod Insertion Limit is reached where the rods will stop, requiring
MANUAL insenion.
Ans: Ib l Exam Level: lS l Cog itive t4<el: l Comprehension l-
Explanatio
o ef Asswer
KA: l 007 EK3.01 l RO Value: l4.0 l SRO Value: l4.6 l Section: l EPE - l RO Group: l 2 l SRO Group: l2
Systems / Evolution Reactor Trip
Tith:
KA Knowledge of the reasons for the following responses as they apply to Reactor Trip:
Statessent:
Actions contained in EOP for reactor trip
Reference Reference Number Reference Section Page Number (s) Revision learn.
Obj
Response To Nuclear Power FR-S.1 step 1, RNO 2 Iss iB
Generation- ATWS Rev 4
Response To Nuclear Power IOM-53.4.FR S.I 111.1 Knowledge 57 iss iB
Generation - ATWS Rev 4
h %round
EOPs LP-SQS-53.3 1. 3
Question Source j New l Question Modification Method l
Questlos Source Comments: l
M te'7Nauired for
Esar y' n:
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, Question Tcpic: l Eval c f vapor space leak -Tech Spec limit
Given the following conditions:
*
The reactor is operating at 100% power
* A 1.2 gpm valve packing leak has occurred on [PCV-RC-455B] PRZR Spray Viv
a
The Primary Drains Transfer Tank level is increasing
Which of the following describes what type ofleakage this is and based on the leak size what action is
required per Technical Specifications?
This leak is considered:
c. Primary boundary L EAKAGE that requires Technical Specification entry.
b. Identified LEAKAGE that does not require Technical Specification entry.
c. Unidentified LEAKAGE that requires Technical Specification entry.
d. Unidentified LEAKAGE that does not require Technical Specification entry.
Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
a cf Answer
KA: l 2.2.22 l RO Value: l3.4 l SRO Value: l 4.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2
Syst;m/ Evolution Pressurizer Vapor Space Accident
Title:
KA Equipment Control
Knowledge oflimiting conditions for operations and safety limits.
Rirence Reference Number Reference Section Page Number (s) Revision Learn. t
Obj
Beaver Valley -Unit i 1.14, 3.4,6.2 1-3,3/4 4-13
Technical Specifications
RCS LP-SQS-6.5 Vll.A 24 4 8.g
l
Q estion Source l New l Question Modification Method l I
Question Source Comments: l
M t; rial Required for
Examination:
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Question Topic: l Basis for use of ADVERSE Cnmt vElues
Given the following conditions:
'
* A LOCA has occurred -
* Containment pressure increased to 6.0 psig
* Containment radiation has increased to 1.5E+5 R/hr.
t '
Ninety minutes later containment pressure decreases to 3.0 psig and containment radiation has decreas
!
4E+4 R/hr. Integrated CNMT radiation dose is 2.3E+5 Rads.
Which of the following describes whether the use of adverse containment parameters can be discontinued?
a. Use of adverse containment parameters can be discontinued.
b. Continued use of adverse containment parameters is required only due to the containment radiation
readings.
c. Continued use of adverse containment parameters is required only due to the containment pressure
conditions.
d. Continued use of adverse containment parameters is required due to both the containment pressure
and radiation conditions.
Ans: ja l Exam Level: IS l Cognitive Level: l Application l
Espiaratio
o of Answer
KA: l 009 EK3.16 l RC Value: l3.8 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2
Systim/ Evolution Small Break LOCA
Thie:
KA Knowledge of the reasons for the following responses as they apply to Small Break LOCA:
Stat: ment:
Containment temperature, pressure, humidity and level limits
Page Number (s) Revision Learn.
Reference Number Reference Section
Reference Obj
li.D 12 13 issIB
Generic Instrumentation IOM-53B.5.GI 2
Rev 2
X.B.6, 8 22-23 1 15
EOP Generic issues LP-SQS-53.2
Question Source l New l Question Modification Method l _
Question Source Comments: {
M:terial Required for - Subcooling Attachment 6-A
Ex:mination:
1
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Question'Tcpic: l Evilcf conditions for tripping RCPs
Given the following conditions:
* A LOCA has occurred
* Containment pressure is 9.2 psig and lowering
* RCS pressure has stabilized at 325 psig
* Steam generator pressures are 800 psig and lowering
* fall ECCS equipment has responded as required
Which of the following describes when the RCPs should be tripped?
a. Immediately
b. When the highest steam generator pressure reaches 700 psig.
c. When the highest steam generator pressure reaches 525 psig,
d. When the lowest steam generator pressure reaches 700 psig.
^
Ans: la l Exam level: lS l Cognitive Level: l Comprehension l
Explanatio
o cf Asswer
KA: l 011 EA1.03 l RO Value: l4.0 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l1
System / Evolution Large Break LOCA
Title:
KA Ability to operate and / or monitor the following as they apply to Large Break LOCA:
Statement:
Securing of RCPs i
i RefereIce Reference Number Reference Section Page Number (s) Revision learn.
Obj
Reactor Trip Or S1 IOM-53.A.E-0 Foldout IssiB
Rev 5
- EOP Generic Issues LP-SQS-53.2 Terminal ;
Obj.
, Question Source l New l Question Modification Method l
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M:.t: rial Required for
Ex:mliation:
1
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Question Tc pic: l Determination of RCP/ reactor trip
Reactor power is 35%. Which of the following cornbinations ofloop flow conditions indicates that a reactor
trip should have occurred?
a. [F1-1RC-414] RCL 1 A Flow indicates 80%.
[FI-IRC-424] RCL 1B Flow indicates 80%.
b. [FI-1RC-414] RCL l A Flow indicates 80%.
[F1-IRC-415] RCL 1 A Flow indicates 80%.
c. [FI-1RC-414] RCL 1 A Flow indicates downscale.
[FI-lRC-435] RCL IC Flow indicates 80%.
d. [FI-lRC-414] RCL l A Flow indicates upscale.
[F1-1RC-415] RCL 1 A Flow indicates 80%.
~
Ans: lb l Eram level: lS l Cognitive Level: l Memory l
Explanatio
e ef Answer
KA: l 015 AA1.03 l RO Value: l 3.7* l SRO Value: l3.8 l Section: l EPE l RO Group: l 1 l SRO Group: l1
System / Evolution Reactor Coolant Pump Malfunctions
Title:
KA Ability to operate and / or monitor the following as they apply to Reactor Coolant Pump Malfunctions:
Statement:
Reactor trip alarms, switches, and indicators
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Reactor Coolant System - 10M-6.4-lF lil.A 27 iss 4 Rev
Instrument Failure Procedure 6
Reactor Coolant System LP-SQS-6.5 4 5,6
Question Source l New l Question Modification Method l
Questlen Source Comments: l
Material Required for
Exami:stion:
Page 66
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' Question Topic: l Fcilure cf makeup
Given the following conditions:
- VOLUME CONTROL TANK LEVEL HIGH-LOW (A3-53) has alarmed ;
- [LI-1CH-115] Volume Control Tank Level (VB-A) failed offscale high
Actual VCT level will:
a. remain constant.
b. decrease until automatic makeup initiates.
c. decrease until the charging pump suction transfers to the RWST.
d. decrease until the VCT is empty.
Ass: l d- l Eram level: lS l Cognitive Level: l Application l
Esple:.stic
c of Answer
KA: l 022 AAl.08 l RO Value: l3.4 l SRO Value: l3.3 l Section: l EPE l RO Group: l 2 l SRO Group: l2
Syst;m/ Evolution Loss of Reactor Coolant Makeup
,
Title:
KA Ability to operate and / or monitor the following as they apply to Loss of Reactor Coolant Makeup:
Statement:
VCT level
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Alarm- A3 53 VCT Level IOM 7.4.AAX PC 4,5 A 2-3 iss 4 Rev
High Low 0
CVCS - Instrumentation end lOM-7.1.D Auto M/U, LCVs 1-2, 8-9 Iss 4 Rev
Controls 2
CVCS LP SQS 7.1 lil.D.2.b 21 2.g, 6.a
Question Source l New l Question Modification Method l
Qrestion Source Comments: l
Material Required for OM Figure 7 39
Ex:mination:
..
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Q:estion Tc'pic: l Boration and SDM Tcch Specs _
Given the following conditions: ,
l
* - [1CH-P-2A] Boric Acid Transfer Pump is out of service
* RCS Temperature is 420 *F
* SDM is 1.67 delta K/K
*: S/D Banks are fully withdrawn
If[lCH-P-2B] Boric Acid Transfer Pump trips, HOW will required Technical Specification Shutdown
Margin be restored? -
. a. BORATE, by gravity feeding the in-service Boric Acid tank to the blender,
b. Emergency borate through the Emergency Boration valve [MOV-CH-350].
c. Align the suction of the charging pump to the RWST.
d. Open the reactor trip breakers.
Ans: lc l Exam level: lB l Cognitive Level: l Application l
Esplanatio
aefAnswer
KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l 4.0 l Section: l EPE l RO Group: l 1 l SRO Group: l1
Systems / Evolution Emergency Boration
Title:
KA Conduct Of Operations
Statentent:
Ability to apply tcchnical specifications for a system.
Reference - Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Technical Specifications 3.1.1.1, 3.1.2.2, and
3.1.2.6
CVCS LP-SQS-7.1 6 11
,
Qeestion Source l New l Question Modification Method l
Question Source Comments: l
Miterial Required for Technical Specifications
Ex::mination:
Page 68
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* Questian Trpic: l Emirgency Bor-tion requir;ments
Fellowing a turbine load rejection, control rods are automatically inserted causing ROD CONTROL BANK
D LOW-LOW alann (A4-124) to be received.
:
i
Which of the following is the required action by procedure?
a.
Place the rods in manual and withdraw them until the alann clears.
b. Place the rods in manual and allow temperature to stabilize.
c. Emergency borate. !
d. Borate via the normal flow path until the CONTROL BANK D LOW-LOW alarm clears.
Ars: lc l Exam Level: lS l Cogaltive level: l Memory l
Explanatio
a tf Answer
i
KA: l2.4.31 l RO Value: l3.3 l SRO Value: l 3.4 l Section: l EPE l RO Group: l 1 l SRO Group: l1 l
Systm/ Evolution Emergency Boration
Title: ;
I
KA Emergency Procedures / Plan
Statement: .
Knowledge of annunciators alanns and indications, and use of the response instructions.
Ref.rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj j
Emergency Boration 10M 7.4.S I 1 lss 4 Rev l
1 i
Rod Ccntrol Bank D Low 10M-1.4.ABF 1 iss 3 Rev
Low l
I :
CVCS LP-SQS-7.1 10.p I
Question Source l Facility Exam Bank l Question Modification Method l j
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M:t: rial Required for- !
Ex:mination:
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Page 69
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Qrestion Tepic: l Eval cfloss of RHR condition
While operating at 175 *F and the RCS depressurized, the running RHR pump trips. The other RHR pump
is available to be immediately started.
Which of the following describes when the other RHR pump should be started and the basis for this
decision'?
l
The second RHR pump should be started:
e. immediately, to avoid any heatup of the RCS.
b. only after investigating the cause of the running pump trip, to avoid losing the second pump.
c. only after observing an RCS heatup, to avoid unnecessary starts of the RHR pump.
d. within five minutes, which is the most limiting time until boiling will occur.
Ars: lb l Exam level: lS l Cognitive Level: l Memory l
Explanatio
e rf Answer
KA: l 025 AKl.01 l RO Value: l3.9 l SRO Value: l4.3 l Section: l EPE_ l RO Group: l 2 l SRO Group: l2
System / Evolution Loss of Residual Heat Removal System
Title:
KA Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat
Statement: Removal System:
Loss of RHRS during all modes of operation
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Residual Heat Removal AOP 1.10.1 Caution 2 1ss 3A
System Loss Rev 5
OM 53C- AOPs LP.SQS-53.C 5 4
Question Source l New l Question Modification Method l
Question Source Comments: l
Mat; rial Required for
Ex:mination:
Page 70
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Question Tz pic: l Loss of CCW during a loss of power
IB and 1C Component Cooling Water Pumps [lCC-P-1B & IC] are BOTH racked to the Connect position
on the DF bus.
Which of the following control switch positions describes when BOTH [lCC-P-1C) and [1CC-P-1B] will
fail to restart on a D/G load sequence signal, following a DF bus undervoltage condition?
l
!
e. [1CC-P-1B]- After START, [lCC-P-1C]- After START i
b. [1CC-P-1B]- PULL-TO-LOCK, [lCC-P-1C]- After Start
{
c. [lCC-P-1B]- After STOP, [1CC-P-1C]- PULL-TO-LOCK
d. [lCC-P-1B]- Aner STOP, [1CC-P-lC]- After STOP !
Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio !
ocf A swer
KA: l 026 AA2.02 l RO Value: l2.9 l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: j1 l
I
System / Evolution Loss of Component Cooling Water i
Title:
KA Ability to determine and interpret the following as they apply to Loss of Component Cooling Water: )
Statement:
{
The cause of possible CCW loss l
l
Ref;rence Reference Number Reference Section Page Number (s) Revision Learn. j
Obj j
Reactor Plant Component IOM 15.1.D 3 issue 4 l
and Neutron Tank Cooling Revi
W;ter (CCRS)
~
Reactor Plant Component LP-SQS-15.1 4 5
and Neutron Tank Cooling
Water (CCRS)
Qrestion Source l New l Question Modification Method l
Question Source Comments: l
l
Material Requit ed for
Ex mination.
l
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l
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Page 71
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Question Tepic: l Effect cf reference leg break
'
Given the following conditions:
* - Reactor power - 100% * A leak develops on the reference leg for the controlling Pressurizer level sensor
How will charging flow respond over next five minutes? Charging flow will:
a. decrease to the minimum value.
b. decrease and then return to the initial value.
c. increase to makeup for the loss through the leak.
d. increase to the maximum flow value.
Aas: la l Eram Level: }S l Cognitive Level: l Comprehension l Explanatio o cf Answer KA: l 028 AKl.01 l RO Value: l2.8* l SRO Value: l3.1 l Section: l EPE l RO Group: l 3 l SRO Group: l3 System / Evolution Pressurizer Level Control Malfunction Title: KA Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Statement: Malfunction:
PZR reference leak abnormalities
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Pressurizer and Pressure LP-SQS-6.4 1.D. I .f 9 10 4 14 ReliefSystem Reactor Coolant System - lOM-6.4.lF 12 4 6 Instrument Failure Procedure Question Source l New l Question Modification Method l Q:estion Source Comments: l
M t: rial Required for
Ex mination:
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Page 72
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Gtven the following conditions:
* Reactor power - 100%
= Both feedwaterpumps trip
= The reactor fails to trip
Which of the following describes when AMSAC s' ..,ald trip the turbine?
e. Immediately after the feedwater pumps trip.
b. Immediate!y after feedwater flow decreases below 25% flow.
c. 150 seconds after the feedwater pumps trip.
d. 25 seconds after feedwater flow decreases below 25% flow.
Ars: ld l Exam Level: lS l Cognitive Level: l Memory l
Explanatio
o s.f Answer
KA: l 029 AA2.09 l RO Value: l4.4 l SRO Value: l4.5 l Section: l EPE l RO Group: l 2 l SRO Group: l1
System / Evolution . Anticipated Transient Without Scram
Title:
KA Ability to determine and interpret the following as they apply to Anticipated Transient Without Scram:
Statement:
Occurrence of a main turbine / reactor trip
Refirence Reference Number Reference Section - Page Number (s) Revision Learn.
ATWS Mitigation System 10M-45B. I .B 1,2 iss 4 Rev
Actu: tion Circuitry 0
AMSAC LP-SQS-45.2 II.D.2.e 4 1 3
Q1estion Source l New l Question Modification Method l
Question Source Comments: l
Mat: rial Required for l
Ex miration:
;
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Page 73
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Qrestion Tc pic: l Evalu', tion of SR NIS voltage fiilure
What would be the plant response to the following conditions?
o The plant is operating at 100% power
call systems are NSA
oThe "A" train Source range RESET / BLOCK switch is inadvertently tumed to the BLOCK position.
c. The reactor would trip, and N31 SR would energize
b.' The reactor would not trip, and N31 SR would not energize.
c. The reactor would trip, and N31 SR would not energize
d. 'Ihe reactor would not trip, and N31 SR would energize
Ass: lb l Exam Level: lE l Cognitive level: l Application l
Explanatio
o of Answer
KA: l 032 AKl.01 l RO value: l2.5 l SRO Value: l3.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2
System / Evolution Loss of Source Range Nuclear Instrumentation
Title:
KA Knowledge of the operational implications of the following concepts as they apply to Loss of Source Range
Statement: Nuclear Instrumentation:
Effects of voltage changes on performance
Reference Section Page Number (s) Revision Learn.
Reference Reference Number
Ob]
4
UFSAR fig. 7.2 sheet 3
!!!.E.3 7 Iss 4 Rev
Reactor Excore Instrument 10M-2.1.c
1
System
Q:estion Source lNew l Question Modincation Method l
Qxestion Source Comments: l
M'.terial Required for UFSAR fig. 7.2 sheet 3
Ermination:
Page 74
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._ _ _ _ - _
Question Topic: l Ev*l of feiled IR channel on SU
Giv:n the following conditions:
* - Plant startup is in progress.
= All power range channels indicate 6% reactor power.
'
+ Intermediate channel N-36 fails HIGH.
-* Reactor power remains at 6%.
Which of the following describes required operator actions?
a. Initiate a reactor trip, enter E-0, and FR-S.I.
b. Immediately commence a controlled reactor shutdown.
c.L Raise power to greater than P10 and block both intermediate ranges.
d. Continue power oper tions.
Ass: Ib- l Exam level: IS I Cognitive level: l Memory ~ l
Expl: ration
of Answer
KA: l2.1.1 l RO Value: l3.7 l SRO Value: l3.8 l Section: l EPE l RO Group: l 2 l SRO Group: l2
Systean/ Evolution - Loss ofIntermediate Range Nuclear Instrumentation
Title:
KA Conduct Of Operations
Stateunent:
Knowledge of conduct of operations requirements.
Ref rence Reference Number Reference Section Page Number (s) Revision Learn.
Ob] l
Excare Instrumentation LP-SQS-2.1 V.C.3.c & e 16 17 5 5,8,12 '
System
Conduct of Operations - 1/20M-48.1.B VI.H.S 9 iss 3 Rev
17
Cenduct of Operations 1/2LP-SQS-48.1 6 l
Question Sourre lNew 'l Question Modification Method I f
Question Source Comments: l
Material Required for
Examination:
1
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Page 75
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Questkm Te pic: l Fu:1 llandling accident systems response A fuel assembly was ruptured during movement in the fuel building. Which of the following describes how the fuel building evacuation alarm is actuated?
c. The alarm must be manually initiated from the control room.
b. [RM-lRM-206] and [RM-1RM-207) Fuel Pool Bridge Area Monitors will sound the evacuation
alarm.
c. [RM-IVS-103A, B) Fuel Building Ventilation Exhaust monitors will sound the evacuation alarm.
d. ' Die alarm must be manually initiated from either the fuel building or the contml room.
l Cognitive Level: l Memory l
A s: lc l Exam Level: IS
Explanatio
oef Answer
l Section: l EPE l RO Group: l 3 l SRO Group: l3
KA: l 036 AA2.02 l RO Value: l3.4 l SRO Value: l 4.1
System / Evolution Fuel Handling incidents
Title:
KA- Ability to determine and interpret the following as they apply to Fuel llandling incidents:
Statement:
Occurrence of a fuel handling incident
Reference Section Page Number (s) l Revision Learn.
Reference Reference Number
Ob)
C.2 2 iss 3A
Irradiated fuel Damage AOP 1.49.1
Rev 3
5 6
OM 53C- AOPs LP-SQS-53C.I
Question Source l Facility Exarn Bank l Question Modification Method l
Q:estion Source Comments: l
M;t: rial Required for
Ex:mination:
Page 76
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TJuestion Tipic: l R:sponse cf SG leak detection monitors
At what power level will the steam generator leak:ge N-16 Radiation Monitors [RM-MS-102A,B, & C]
BEGIN to provide valid leak rates,in GPD?
c. 5%
b. 20 %
c. 30%
d. 50 %
Ans: lb l Exam Level: lS l Cognitive Level: l Memory l
Explanatio
c ef Answer
KA: l 037 AA1.06 l RO Value: l3.8' l SRO Value: l 3.9' l Section: l EPE l RO Group: l 2 l SRO Group: l2
System / Evolution Steam Generator Tube Leak
Title:
KA Ability to operate and / or monitor the following as they apply to Steam Generator Tube Leak:
St:tement:
Main steam line rad monitor meters
Ref:rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Radiation Monitoring 10M-43.1.C 8 Iss 4 Rev
Systems - Major components ?.
OM $3C- AOPs LP SOS-53C.I 6
Question Source l Facility Exam Bank l Question Modification Method l 3
Qrestion Source Comments: l
M:t: rial Required for
i
Ex:mination:
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Page 77
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Question Tcpie: l Evalu: tion of cooldown temperature /cooldown
Giv:n the following conditions:
* A Steam Generator Tube Rupture has occurred
* E-3, Steam Generator Tube Rupture, is being performed
* The RCS has been cooled down to the target temperature.
In order to maintain RCS subcooling, intact steam generator pressure must be maintained:
c. greater than the ruptured generator.
b. equal to the ruptured generator.
c. greater than the saturation pressure of the RCS.
d. less than the ruptured generator.
Ans: ld l Exam Level: lS l Cognitive Level: l Application l
Explanatio
c of Answer
KA: l 038 EA1.36 l RO Value: l4.3 l SRO Value: l 4.5 l Section: l EPE l RO Group: l 2 l SRO Group: l2
System / Evolution Steam Generator Tube Rupture
Title:
KA Abiltty to operate and / or monitor the following as they apply to Steam Generator Tube Rupture:
Statement:
Eooldown of RCS to specified temperature
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Steam Generator Tube 82 issIB
Rupture Background Rev 5
EOPs Ll"-SQS-53.3 3
_
Question Source l Facility Exam Bank l Question Modification Method l
Question Source Comments: l
Material Required for
Examination:
, l
Page 78 _ _ _ _ _ - - -
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'T)sestion Topic: l Eviluation cf FW condition
Given the following conditions:
*
A steam break has occurred on SG "A"
* A reactor trip was manually initiated
* A SI has NOT been initiated
.
,
-* No operator actions have been performed on the feedwater system.
* Only SG "A" narrow range level has decreased below 12%.
* RCS T. , are (A) f'.2 *F,(B) 550 F,(C) 550 F
Which of the following is the expected status of fe:dwater?
c. The feedwater regulating valves will be shut. The Turbine Driven AFW pump will be running.
b. The feedwater regulating valves will be shut. All AFW pumps will be running.
c. A complete FWI isolation will be initiated. All AFW ptunps will be running.
d. The feedwater system will be in the same lineup as prior to the reactor trip, except the FRVs will be
throttled closed.
Ans: la l Exam Level: lS l Cognitive Level: l Application l
Explanatio
o cf Answer
KA: l 040 AA1.02 l RO Vals?: l4.5 l SRO Value: l 4.5 l Section: l EPE l RO Group: l 1 l SRO Group: ll
System / Evolution Steam Line Rupture
Title:
KA Ability to operate and / or monitor the following as they apply to Steam Line Rupture:
Stat mcnt:
.,
Feedwater isolation
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
SG Feedwater System - 10M 24.lD Feedwater Isolation 2, 6 1ss 4,
Instrumentation and Controls Rev.2
,
Reactor Protection System LP-SQS-l .1 V.E.5 38 6 9
Question Source jNew l Question Modification Method l
Question Source Comments: ]
Mat: rial Required for
Enmination:
!
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Page 79
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Question Tepic: l Effect & mitigation techniques
Given the following conditions:
* An uncontrolled depressurization of all steam generators has occurred
* Current RCS cooldown rate is 125 *F/hr
Which of the following describes how drying out, of the steam generators, is avoided while trying to limit
cooldown rate?
a. A minimum AFW flow to all steam generators is maintained.
b. SGs are intermittently fed to assure that a wide range levels remain above 10%.
c. Only reducing AFW flow as necessary to reduce the cooldown rate to less than 100 F.
d. AFW feed rate is limited to maintain constant level, provided the level is above 10% wide range.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
c of Answer
KA: l 040 AKl.07 l RO Value: l3.4 l SRO Value: l4.2 l Section: l EPE l RO Group: l 1 l SRO Group: l1
^
System / Evolution Steam Line Rupture
,
Title:
KA Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture:
Statement:
EfTects of feedwater introduction on dry S/G
Reference Reference Number . Reference Section Page Number (s) Revision Learn.
Obj
Uncontrolled ECA-2.1 STEP 6 5 iss1B,
Depressttrization of all SGs rev.4
,
Uncontrolled 10M-53B.4.ECA 2.1 IV.6 25 iss 1B;
Depressurization of all SGs Rev.4
Background
EOPs LP-SQS-53.3 3
Question Source l New l Question Modification Method l
Question Source Comments: l
M:terial Required for
Ex:mination:
I
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Page 80
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Question Topic: l Block of steam dumps on turbine trip
A loss cf condenser vacuum has occurred due a leak in the condenser. Main Condenser Steam Dumps are
cpen following a turbine trip.
As vacuum decreases, at what condenser vacuum will Main Condenser Steam Dumps close?
z. 25" Hg Vacuum >
b. 20" Hg Vacuum
c. 10" Hg Vacuum
d. 5" Hg Vacuum
A"s: Ib l Exam Level: lS l Cognitive Level: l Memory l
Expla:stio
o cf Answer
KA: l 051 AK3.01 l RO Value: l 2.8' l SRO Value: l 3.l* l Section: l EPE l RO Group: l 1 l SRO Group: l1
System / Evolution Loss of Condenser Vacuum
Title:
KA Knowledge of the reasons for the following responses as they apply to Loss of Condenser Vacuum:
Statement:
I ms of steam dump capability upon loss of condenser vacuum
R:.f;rence Reference Number Page Number (s) Revision Learn.
l Reference Section
Obj
'
25 4 ,
M in Steam Supply / Steam LP-SQS-21.1
Dump System _
1OM-26.2.B IO 4 7
'
Q:estion Source - l New l Question Modification Method l
Q estion Source Comments: l
Mit: rial Required for
.
Ex:mination:
Page 81
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Question Topic: l Determination of Feedline break
A break has occurred on the feedwater line to SG "A" downstream of [MOV-FW-156A], Main Feed Line
Containment Isolation valve. Containment pressure increases to the Si setpoint.
Following the reactor trip and SI, which of the following SG pressure indications would exist?
a. Oniy SG "A" pressure would be decreasing from the break.
b. All SG pressures would be decreasing from the break via the main steam lines.
c. All SG pressures would be decreasing from the break via the main feedwater lines.
d. All SG pressures would be decreasing from the break via the auxiliary feedwater lines.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
} u of Answer
,
KA: l 054 AKl.01 l RO Value: l4.1 l SRO Value: l4.3 l Section: l EPE l RO Group: l 2 l SRO Group: l2
System / Evolution Loss of Main Feedwater
Title:
KA Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater:
Statement:
MFW line break depressurizes the S/G (similar to a steam line break)
Reference Section Page Number (s) Revision Learn.
Reference Reference Number
Ob]
4 1,4g
Main Steam Supply / Steam LP-SQS-21.1
Dump System
IOM-21.1.C 5 iss 4 Rev
Main Steam System
1
VOND 24 1
Question Source l New l Question Modification Method l
Question Source Comments: l
~ Material Required for
Examination:
Page 82
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Question Tepic: l Load required to be left in AUTO
l
A loss cf cli 4KV busses has occurred. ECA-0.0 has been implennnted to the point of placing deenergized
equipment in PULL TO LOCK. The IDF emergency bus has been selected to cross tie to Unit 2.
Which of the following l AE Emergency Bus loads shall remain in the AUTO position and the basis for
leaving that pump in AUTO?
e. Reactor River Water Pump to assure that the diesel has cooling upon startup.
b. Charging Pump to restore seal flow,
c. Charging Pump to restore Pressurizer level,
d. Component Cooling Water Pump to restore cooling to the thermal barrier.
Ans: la l Exam Level: lS l Cognitive Level: l Memory l
Explanatio
o tf Answer
j
KA: l 2.4.20 l RO Value: l3.3 l SRO Value: l4.0 l Section: l EPE l RO Group: l 1 l SRO Group: l1
System / Evolution Station Blackout
Title:
KA Emergency Procedures / Plan
Statement:
l
Knowledge of operational implications of EOP warnings, cautions, and notes.
Ref;r: nee Reference Number Reference Section Page Number (s) Revision Learn. J
Obj
Loss ef All Emergency 4KV 10M-53 A.I.ECA-0,0 Caution Step 14 10 issIB
AC P:wer Rev 4
Emergency Operating LP-SQS-53.3 1 3
Procedures
]
Qrestion Source l Facility Exam Bank l Question Modification Method l
Question Source Comments: l
M:,terial Required for
Examination:
I
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Page 83
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l
Questio] Tcpic: l Purpose of SI Reset l If an SI actuation signal is received when performing ECA-0.0," Loss of All Emergency 4KV Power", the S1 l signd should be:
j
a. reset to prevent lockout of the stub busses.
l
b. reset to permit manual loading of equipment of an Emergency bus. j
t
c. allowed to remain active to ensure rapid injection of core cooling water when power is restored.
d. allowed to remain active to ensure the load sequencer re-initiates when the DG starts.
Ans: lb l Exam Level: lS l Cognitive level: l Memory l
Explanatio
o of Answer
KA: l 055 EK3.02 l RO Value: l4.3 l SRO Value: l 4.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1
System / Evolution Station Blackout
Title:
KA Knowledge of the reasons for the following responses as they apply to Station Blackout:
St:tement:
Actions contained m EOP for loss of olTsite and onsite power
Reference Number Reference Section Page Number (s) Revision Learn.
Reference
Obj
Loss cf All Emergency 4KV ECA-0.0 steps 31 & 37 22 &25 issIB;
Rev 4
AC Power
Loss cf All Emergency 4KV 10M-53B.4.ECA-0.0 Step 31, Basis 127 iss IB;
AC Power Background Rev 4
3
EOPs LP-SQS-53.3
Question Source l NRC Exam Bank l Question Modification Method l
Question Source Comments: l
Material Required for
Examination:
.
Page 84
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Question Topic: l RCS temper-tures j
What is the expected response of RCS Hot and Cold leg temperatures during the first few minutes following
a reactor trip froml00% power COINCIDENT with a loss of offsite power?
a. Hot leg temperatures will rise, and Cold leg temperatures will remain relatively constant, until
natural circulation flow is established.
.
b.' Hot leg temperatures and Cold leg temperatures will both rise, until natural circulation flow is
established.
c. Hot leg temperatures will remain relatively constant and Cold leg temperatures will drop, until
natural circulation flow is established.
d. Hot leg temperatures will rise and Cold leg temperatures will drop, until natural circulation flow is
established.
Ans: la l Exam level: lS- - l Cognitive Level: l Memory l
Explanatio
o af Answer
KA: l 056 AA2.18 l RO Value: l3.8 l SRO Value: l4.0 l Section: l EPE l RO Group: l 3 l SRO Group: l3
Systim/ Evolution Loss of Off-Site Power
Thie:
KA Ability to determine and interpret the following as they apply to Loss of Off Site Power:
Statement:
Reactor coolant temperature, pressure, and PZR level recorders
R:f;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Reactor Trip Response ES-0.1 Note before step 3 3-4 issIB
Rev 4
EOPs LP-SQS-53.3 1 6
QIestion Source l Facility Exam Bank l Question Modification Method l
Qyestion Source Comments: l
M;t: rial Required for
Ex:mination:
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Page 85
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, . . I Question Tcpic: l Effect of a loss of V;tal AC on Feedwater
Given the following conditions: )
)
I
* Reactor power is 74%
* Feedwater controlis in automatic
* Loss of a single 120 VAC Vital bus has occurred
Which of the following describes the expected response of Main Feedwater Regulating Valves which do
NOT remain in AUTO?
c. The FRVs will immediately fail open.
b. The FRVs will ' immediately fait closed.
c. The FRVs will drift shut.
d. The FRVs will transfer to either MANUAL or AUTO HOLD. l
f
Aus: ld l Exam Level: lS l Cognitive Level: l Memory l
Explanatio
ocf Answer
KA: l 057 AA2.19 l RO Value: l4.0 l SRO Value: l4.3 l Section: l EPE l RO Group: l 1 l SRO Group: l1
Syst:m/ Evolution
Title:
l, Loss of Vital AC ln'strument Bus
KA Ability to determine and interpret the following u *Ney apply to Loss of Vital AC Instrument Bus:
Statement:
The plant automatic actions that will occur o - [/of a vital ac electrical instrument bus {
, j
Ref;rence Reference Numttr Reference Section Page Number (s) Revision Learn. l
Obj i
Alarm-Vital Bus I,11,111, IV 10M-38.4.AAA, AAC, 2
Trouble AAE. AAG
120V AC Distribution LP-SQS-38.1 32 6 6
System
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Question Tepic: l Effect of a loss of DC on RCPs
Which of the following is the effect that losing 125 VDC Bus I will have on the Reactor Coolant Pumps?
a. 'One or two RCPs will trip on undervoltage.
b. One or two RCP breakers will ONLY be able to be tripped using the mechanical trip at breaker. l
c. Component cooling water will be lost to all RCPs.
d. Seal water flow to the RCPs will be isolated.
Ass: lc l Eram tevel: lS l Cognitive Level: l Application l
Emploratio
l c of Answer j KA: l 058 AA2.03 l RO Value: l3.5 l SRO Value: l3.9 l Section: l EPE l RO Group: l 2 l SRO Group: l2
System / Evolution Loss of DC Power
Title:
KA Ability to determine and interpret the following as they apply to Loss af DC Power:
Statessent:
DC loads lost; impact on to operate and monitor plant systems
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
OM 39 10M-39. 5.B.6 Table 39-6 all iss 4 Rev
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Question Topic: l Evalef Tech Spec
Given the following conditions:
* Unit 1 is in MODE 6
* Unit 2 is in MODE 1
= Movement ofirradiated fuel is ongoing in the Unit 1 Containment only
* Monitor RM-lRM 218A Control Room Area - Unit 1 has failed low
What action is required for the above conditions?
o. No action is required because the monitor is not required to be operable.
b. Within ONE hour the respective Unit 2 control room monitor train shall be verified operable.
c. Within ONE hour, verify that Control Room Area - Unit I monitor [RM-1RM-218B] is operable.
d. Within ONE hour, suspend all operations involving movement ofirradiated fuel.
A"s: lb l Exam Level: 1S l Cognitive Level: l Application l
Esplanatio
oefAnswer
KA: l 061 AA2.06 l RO Value: l3.2 l SRO Value: l4.1 l Secti<;n: l EPE l RO Group: l 2 l SRO Group: l2
Syst;m/ Evolution Area Radiation Monitoring System
Title:
KA Ability to determine and interpret the following as they apply to Area Radiation Monitoring System:
Staternent:
Required actions if alarm channel is out of service
Reference Number Reference Section Page Number (s) Revision Learn.
Reference
Obj
Beaver Valley - Unit 1 3.3.3.1, Table 3.3-6,1.c, 3/4 3-33-3-35 Amend
Action 41 119
Technical Specifications
VI.A 31 4 7.a
Radiation Monitoring System LP-SQS-43.1
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Quest 6on Tcpie: l Effect of restoring cir using IIA 90.
During a loss of containment cir, which of the following is the possible effect of opening [IlA-90]
Instrument Air to Containment Air Iso! Valve too quickly?
c. Station Air compressor trips
b. CVCS letdown isolation
c. SG Main FW Feed Reg Vivs failing open
d. Main Steam Line Trip Valve closure
Ans- ld l Exam Level: lS l Cognitive Level: l Memory l
Explanatio
a of Answer
KA: l 065 AK3.08 l RO Value: l3.7 l SRO Value: l3.9 l Section: l EPE l RO Group: l 3 l SRO Group: l2
System / Evolution Loss ofinstrument Air
Title: 1
KA Knowledge of the reasons for the following responses as they apply to Loss ofInstrument Air:
Statement:
Actions contained in EOP for loss of instrument air
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Loss cf Containment AOP 1.34.2 Caution before step 4 3 iss 3 A
Instrument Air Rev 3
OM 53C. AOPs LP-SQS-53C.1 5 4
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Question Te pic: l Type of detection /extinguishmg eqpt for use Which cf the following describes the fire protection afforded for the primary process rack area?
a. Carbon Dioxide is released to the area by manual actuation only.
b. Carbon Dioxide is released to the area by automatic actuation of smoke detection or by manual
actuation.
c. Halon is released to the area by manual actuation only.
d. Halon is released to the area by automatic actuation of smoke detection or by manual actuation.
Ass: ld l Exam Level: lS l Cognitive level: l Memory l Esplanatio o of Answer KA: l 067 AAl.08 l RO Value: l3.4 l SRO Value: l3.7 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Plant Fire on Site Title: KA Ability to operate and / or monitor the following as they apply to Plant Fire on Site: Statement:
Fire fighting equipment used on each class of fire
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Fire Protection System - IOM.33.1.B Halon paragraphs 1 & 4 5 iss 4; Summary Description Rev.3 Fire Protection System LP-SQS-33.1 E.1 b.4 29 5 3.e Question Source l New l Question Modification Method l Question Source Comments: l M:tirial Required for Ex:mination:
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A fire in the control room has resulted in control room evacuation. Plant control has been transferred to local
control panels as required by IOM-56C.1, Alternate Safe Shutdown from Outside the Control Room.
Until a cooldown is initiated from the BIP, pressurber level is maintained by charging via:
c. [MOV-RC-556A, B, C] Reactor Coolant Loop Fill Valves to the RCS loops.
.
b. the normal charging connection.
c. the RCP seats.
d. the BIT.
Ans: lc l Exam level: lS l Cognitive tevel: l Memory l
'
Espiaratio
o of A:swer
KA: l 068 AA130 l RO Value: l3A l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1
System / Evolution Control Room Evacuation
Title:
KA Ability to operate and / or monitor the following as they apply to Control Room Evacuation:
Statement:
Operation of the letdown system
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Alternate Safe Shutdown . LP STA 56C.! VI.A.3 12 2 6.a
from Outside the Control
Room
.
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Question Topic: l Controller lo' cation
~
Which cf the following identifies the components used by the operator stationed at the BIP (Backup
Indicating Panel) to lower pressurizer level?
c. [SOV-1RC-102B] RCVS Reactor Vessel Vent Viv
[SOV-1RC-103B] RCVS Pressurizer Vent Viv
[SOV-1RC-105] RCVS Vent to Containment Isolation Viv
b. [LCV-1Cli-460A and B] Ltdn to Regen lix Isol
[TV-Cli-200B] 60 GPM Ltdn Orifice Cnmt Isol Viv
Letdown will flow to the degasifier via [LCV-115A], which has failed to the degasifier position.
c. [MOV-CH-201] Excess Ltdn HX Inlet Isolation Viv
[MOV-ICII-137] Excess Ltdn HX Flow Control Viv
d. [PCV-lRC-455D] PZR PORV Relief Viv
[PCV-1RC-456] PZR PORV Relief Viv
Ans: la l Exam Level: lS l Cognitive Level: l Memory {
Explanatio
e of Answer
KA: l 068 AK2.01 l RO Value: l3.9 l SRO Value: l4.0 l Section: l EPE l RO Group: l 1 l SRO Group: l1
System / Evolution Control Room Evacuation
Title:
KA Knowledge of the interrelations between Control Room Evacuation and the following:
Statement:
Auxiliary shutdown panellayout
Ref>rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Misc. Safety Related 10M-45.1.B (BIP) Indications 7 iss 4 Rev
Systems Summary 1
Description
Alternate Safe Shutdown LP-STA-56C.1 12 2 6
Outside the Control Room
Reactor Coolant System - IOM-6.1.D 5-6
Instrumentation and Controls
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Question Topic: l Basis for starting en RCP
An RCP is started in FR-C.1, " Response to Inadequate Core Cooling", in order to:
a. allow using RVLIS Dynamic Range indication to determine core void content.
b. temporarily improve core cooling until some form of makeup flow to the RCS can be established.
c. enhance the cooling caused by rapid depressurization of the steam generators.
d. establish pressurizer spray flow to reduce RCS pressure to cause low pressure systems to inject.
Ans: Ib l Exam level: lS l Cognitive level: 1 Comprehension ( ,
,
Explanation
of Answer
KA: l 074 EK2.01 l RO Value: l3.6 l SRO Value: l3.8 l Section: lEPE l RO Group: l 1 l SRO Group: l1
Systeam/ Evolution inadequate Core Cooling ,
Title: i
KA Knowledge of the interrelations between inadequate Core Cooling and the following:
St-tement:
RCP
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Ob]
Response to inadequate Core 10M-53B.4.FR-C.1 1 IssIB
Cooling Background Rev 4 .
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Emergency Operating LP-SQS-53.3 1 3
Procedures
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)::est6on Top 6c: l Actions to lower R/A lev Is Ziv:n the following conditions:
* Reactor power has just been raised from 20% to 100% * Dose Equivalent Iodine has just been reported as 5.0 pci/ gram.
Which of the following explains why operation can continue with Dose Equivalent lodine above the Technical Specification LCO limit?
c. To allow for CVCS removal of the crud released by the power change. .
b. The Technical Specification LCO limit is conservative enough, to allow extended periods (> 7 days)
of exceeding the limit.
c. To accommodate the iodine that was released during the power change.
d. The probability of a Large break LOCA occurring during the time period lodine is above the limit,
presents an acceptable risk.
Cas: Ic l Exam Level: lS l Cognitive tevel: l Memory l Guplanatio ecf A:swer
KA: l 076 AK3.05 l RO Value: l2.9 l SRO value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1
Cyst:m/Evtlution High Reactor Coolant Activity filtle:
KA Knowledge of the reasons for the following responses as they apply to liigh Reactor Coolant Activity:
Stateent:
Corrective actions as a result of high fission-product radioactivity level in the RCS
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
Technical Specifications LP-SQS-TS 0 4
Beaver Vtiley - Unit i Bases 3/4 4-4 B 3/4 4-4 Amend
No 102
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Question Topic: l Securing Si flow
L
Which cf the following describes the required subcooling requirements before terminating Si in ES-1.1, Si
! Termination? !
The required subcooling:
c. is based on saturation conditions plus instrument errors.
b. is based on the expected pressure after Si is terminated.
c. is based on the expected temperatures after SI is tuminated.
d. provides for a 50 *F margin to saturation to avoid reinitiation.
A ms: la l Exam Level: lS l Cognitive Level: l Memory l
Expla atio
r:of Answer
KA l E02 EK3.2 l RO Value: l3.3 l SRO Value: l3.8 l Section: l EPE l RO Group: l 1 l SRO Group: l1
System / Evolution Si Termination
Title:
KA Knowledge of tile reasons for the fellowing responses as they apply to St Termination:
Statement:
, Nonnat, abnormal and emergency operating procedures associated with (Si Termination).
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
St Termination /Reinitiation IOM053B.5.GI l1 II.A.i 3 issiB
RevI
EOP Generic issues LP-SQS-53.2 i LO3
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Question Tepic: l Basis for required Pressurizer Level
A reactor trip and SI have occurred, and the control room operators are responding to a small-break LOCA. l
All RCPs are tripped. The operators have proceeded to the recovery stage in ES-1.2, " Post-LOCA
Cooldown and Depressurization". A PZR PORV is used to depressurize the RCS until PZR level is greater
than 18% [50% ADVERSE CONTAINMENT).
In addition to ensuring that RCS conditions are under adequate operator control, the basis for this pressurizer
levIl ensures:
a. that a reduction in subcooling does not occur when SI flow is reduced.
b. sufficient inventory such that PZR level does not drop low when an RCP is started.
c. pressurizer level indication is not due to a void in the vessel head.
d. adequate PZR steam space to absorb pressure fluctuations during RCP start.
Ans: lb l Exam level: lS I Cognitive Level: l Comprehension l
Explanatio
a cf Answer
KA: l E03 EK2.2 l RO Value: l3.7 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l2
Systess/ Evolution LOCA CoolJown and Depressurization
Title:
KA Knowledge of the interrelations between LOCA Cooldown and Depressurization and the following:
Statunent:
Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems,
and relations between the proper operation of these systems to the operation of the facility.
Reference Reference Number Referen:c Section Page Number (s) Revision Learn.
Obj
Post LOCA Cooldown and ES l.2 step 15 10 issIB
Depressurintion Rev 5
Post LOCA Cooldown and lOM-53B.4.ES-1.2 25 iss 1B
Depressurization Rev 5
EOP Generic issues 1LP-SQS-53.2 [iLB.1, Ill.A 5,10 3,4
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Questici Tepic: l Purpose of ECA 1.2 Given the following conditions:
! o A small break LOCA ha occurred due to a break at some unknown location outside containment.
o Perfonnance of ECA - 1.2 "LOCA Outside Containment" did not isolate the break. o At the completion ECA - 1.2 "LOCA Outside Containment", RCS pressure is still dropping
I
At the conclusion of ECA - 1.2 "LOCA Outside Containment" the operating crew should transition to
l
a. E-0 "Rx Trip or SI" in order to reverify that all automatic actions have been completed.
b. E-3 "SGTR , since there are adequate steps within this procedure to deal with these conditions.
c. ES-0.0 "Rediagnosis" in an attempt to diagnosis the break location.
d. ECA-1.1 " Loss of Emergency Coolant Recirculation", in order to deal with the loss of available
inventory for core cooling.
A:s: ld l Esam Level: lS l Cognitive Level: l Comprehension l
Explantio
a cf Answer
KA: l E04 EK2.2 l RO Value: l3.8 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l1
SystIm/ Evolution LOCA Outside Containment
Title: ,
, ;
KA Knowledge of the interrelations between LOCA Outside Containment and the following:
St:tement:
Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems,
and relations between the proper operation of these systems to the operation of the facility.
R1fer:nce Reference Number Reference Section 1
Page Number (s) Revision Learn.
- Obj
LOCA Outside Containment IOM-538.ECA 1.2 1 issIB ; Background Rev 3 l Emergency Operating LP-SQS-53.3 1 1 Procedures
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Question Tepic: l Apply procedural direction for cooldown
During a natural circulation cooldown with RVLIS unav:ilable, it is likely that voids will form in the upper
head region. ES-0.4 " Natural Circulation Cooldown With Steam Void in the Vessel (Without RVLIS)",
limits the size of these voids in the RCS head region by :
1 l c. Requiring all CRDM fans to be runnung.
b. Limiting the allowable increase in pressurizer level.
c. Limiting the maximum temperature on Core Exit Thennocouples.
d. Requiring a minimum of 200F subcooling.
Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l
Explanatio
o ef Answer
KA: l E10 EA2.2 l RO Value: l3.4 l SRO Value: l 3.9 l Section: l EPE l RO Group: l 1 l SRO Group: l1
System / Evolution Natural Circulation with Steam Void in Vessel with/without RVLIS
Title:
KA Ability to determine and interpret the following as they apply to Natural Circulation with Steam Void in Vessel -
Stat: ment: with/without RVLIS:
Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj
NaturalCirculation ES-0.4 step 9 8 Iss1B
Coold:wn With Steam YcM Rev 4
in Vessel (Without RVLIS)
3
EOPs LP-SQS-53.3
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Question'Tople: l Condition retulting in loss cf recire ! Given the following conditions:
j
* A LOCA has occurred
a Due to low RWST level a transfer to Cold Leg Recirculation has occurred.
*
All automatic actions for the transfer to Cold Leg Recirculation are
complete.
a [1SI-P-1B] LHSI pump is not available
e Containment pressure - 12.4 psig
Which of the following would result in a loss ofinjection flow?
c. RCS pressure - 450 psig
[MOV-ISI-862A] 1 A LHS1 Pump RWST Suct Viv fails open
b. RCS pressure -250 psig
[MOV-1SI-863A] 1 A LHS! to Chg Pumps Sup Viv fails closed
c. RCS pressure - 380 psig
[CH-P-1 A] 1 A Charging /HHSI Pump trips ,
[MOV-lSI-863B] 1B LHS1 To Chg Pumps Sup Valve fails closed, i
d. RCS pressure - 180 psig
[MOV-ISI-885A] 1 A LHSI PP Mini Flow Isol Valve fails open
Anst lb l Exam Level: lS l Cognitive Level: l Comprehension l Expla:atio o cf Answer KA: l ElI EA2.1 l RO Value: l3.4 l SRO Value: l4.2 l Section: l EPE l RO Group: l 2 l SRO Group: l2 Syst:m/ Evolution Loss of Emergency Coolant Recirculation Title: KA Ability to determine and inte!pret the following as they apply to Loss of Emergency Coolant Recirculation: l Strt: ment:
Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
R:,f;re:ce Reference Number Reference Section Page Number (s) Revision Learn.
Obj l
Transfer To Cold Leg ES-1.3 step 4 3 issIB l Recirculation Rev 4 EOP Attachment 1-G IOM 53A.I.1-G step 2 2 1ssIB
Rev 2
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Question Topic: l CIB setpoints
How long aft:r a CIB signalis received will the quench sprcy and containment spray pumps start?
t
a. [QS-P-1 A,B] Quench Spray pumps - 5 seconds
[lRS-P-2A, B) Outside Recirc Spray Pumps = 120 seconds
[lRS-P-I A, B] Inside Recirc Spray Pumps = 22. seconds
b [QS-P-1 A,B] Quench Spray pumps - 60 seconds
[lRS-P-1 A] Inside Recirc Spray Pump, [lRS-P-2B) Outside Recirc Spray Pump = 120 seconds
[1RS-P-1B] Inside Recirc Spray Pump, [1RS-P-2A] Outside Recirc Spray Pump = 210 seconds
c. [QS-P-1 A,B) Quench Spray pumps - 60 seconds
[lRS-P-1 A, B] Inside Recirc Spray Pumps = 210 seconds
[lRS-P-2A, B) Outside Recirc Spray Pumps = 225 seconds
d. [QS-P-1 A,B] Quench Spray pumps - 5 seconds
[lRS-P-1 A)Inside Recirc Spray Pump,[lRS-P-2B] Outside Recirc Spray Pump = 210 seconds
[lRS-P-1B]Inside Recirc Spray Pump,[lRS-P-2A] Outside Recirc Spray Pump = 225 seconds
l Exam inel: IS l Cognitive Lescl: l Memorv l
Ae - Id
Explanation
of Answer
l Section: l EPE l RO Group: l 1 l SRO Group: l1
KA: l E14 EKl.3 l RO Value: l3.3 l SRO Value: l 3.6
Systema / Evolution High Containment Pressure
Title:
KA Knowledge of the operational implications of the following concepts as they apply to High Containment Pressure:
Statement:
Annunciators and conditions indicating signals, and remedial actions associated with the (High Containment
Pressure).
Reference Section Page Number (s) Revision Learn.
Ref;rence Reference Number
Ob]
IOM-13.2.B 2 Iss 4 Rev
Containment
3
Depressurization System
27 5 5
Containment LP-SQS-13.01
Depressurization System
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S nior R rct:r Operztor Anzwsr Ksy
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29. c j
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34, d
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(..d. 36. a 2. c 37. b ). d 38. d ), c ' 39 a B. ' a 40. g'd [[J Ty[0 666 O nt 4 6.4. 4 b. c 41. c f l 0 Sl/f/ff f.- c 42. a l )
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S:nisr R:rctor Opertter Answsr Ksy
76. c
51. c
'52. c 77. b
53 b. 78. d
79, a
54. e -
- 55. b 80. a
156. c 81. b
57. a 82. a
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59. - a 84. b
60. b 85. a .
61. c 86. d
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63. b 88. b
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64. a 89. d
65. a 90. d i
66. b 91. c
67. d 92. a
68. c 93. b
69. c 94. c
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70. b
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73. d 98. b
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Beaver Valley Power Station Unit i 10M-46.4. A
Post-DBA Hydrogen Control System
'
Issue 4 Revision 2
Op rating Procedures Page A 1 of 5
Hydrogen Recombiner Startup
1. PURPOSE
This procedure describes the startup of the Post OBA Hydrogen Recombiner following the
unlikely occurrence of a loss of coolant accident. This is accomplished by first setting up
the Hydrogen Analyzer and monitoring Containment hydrogen concentration. When the
concentration level reaches a preset value, the Hydrogen Recombiner is aligned and
started. This procedure is entered from an EOP.
'
II. PRECAUTIONS AND LIMITATIONS
A. If hydrogen concentration is 2: 5%, consult TSC before placing Recombiners in
operation.
B. During accident conditions, radiation levels may be high in the Recombiner area.
Limit the time spent in this area.
. .- - 1
-._ .
C. In order for the Hydrogen Recombiners to operate with sufficient flow, Containment '
pressure must be controlled as close as possible to -2 psig (13 psia). However, ..
Containment pressure must remain below -2 psig (13 psia) to ensure Containment g
remains subatmospheric. I
"
:
lit. INITIAL CONDITIONS 1
-
A. The EOPs require the Hydrog.en Recombiners to be placed in service.
B. The NSS has approved the performance of this procedure. .
C. The 480 VAC distribution system is operable.
D. The following procedure is available:
1. 10M-46.4.G," Placing Wide Range Containment Hydrogen Monitoring System in ,
Operation".
IV.- INSTRUCTIONS l
Note: Valves for the A Recombiner are given in procedure, valves for the B Recombiner
are in parenthesis.
I
A. Place the Hydroaen Recombiner in Service
1. Contact Radeon to determine what type of protective apparel is to be worn and
j
any shielding required.
2. Obtain the following keys to unlock [1HY-101,102.103,104,110, iii,196 and
f l 197]. l
a. SR/O.C.
b. SR/O.D.
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Attachment 2
SIMULATION FACILITY REPORT
Facility Licensee: Beaver Vallev Unit 1 Facility Docket No: 50 334 Operating Tests Administered from: April 20-24,1993 This form is used only to report simulator 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 certification or approval of the simulation facility other than to provide information that May be useJ in future evaluations. No licensee action is required in response to these observations. No simulator deficiencies, that affected the scenario examinations or JPMs, were identified during the execution of the examination.
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