IR 05000334/1998300: Difference between revisions

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{{Adams
{{Adams
| number = ML20248L051
| number = ML20237B808
| issue date = 06/03/1998
| issue date = 08/17/1998
| title = Exam Rept 50-334/98-300OL Conducted on 980420-24 & 0518.Exam Results:Three SRO Instant Candidates Passed All Portions of Initial License Exam
| title = Forwards NRC Operator Licensing Exam Rept 50-334/98-300OL for Exam Administered on 980420-24 & 0518
| author name =  
| author name = Curley V
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =  
| addressee name =  
| addressee affiliation =  
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000334
| docket = 05000334
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-344-98-300OL, NUDOCS 9806100390
| document report number = 50-334-98-300OL, NUDOCS 9808190282
| package number = ML20248L042
| document type = INTERNAL OR EXTERNAL MEMORANDUM, MEMORANDUMS-CORRESPONDENCE
| document type = EXAMINATION REPORT, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 1
| page count = 105
}}
}}


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=Text=
=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.


U. S. NUCLEAR REGULATORY COMMISSION REGION 1 Docket No.:  50-334
Item #1 a) FACILITY SUBMITTED OUTLINE AND INITIAL EXAM SUBMITTAL DESIGNATED FOR DISTRIBUTION UNDER RIDS CODE A070.
         ,
 
Report No.: -  98-300 License No.:   DPR-66 -
b) AS GIVEN OPERATING EXAMINATION, DESIGNATED FOR DISTRIBUTION UNDER RIDS CODE A070.
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
: Item #2- EXAMINATION REPORT WITH THE'AS'GIVEN WRITTEN EXAMINATION ATTACHED, DESIGNATED"FOR DISTRIBUTION UNDER. RIDS CODE IE42 $
y.
 
,
  '
9808190282 990817 PDR ADOCK 05000334 >
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_ _ _ _ _ _ _ . _ - _ _ . _ _ _ _ _ _ _ _ _ _ - - _ _ . - - __
e gp ucoq M O UNITEo STATES
[ ,.,, ' gg  NUCLEAR REGULATORY COMMISSION
  '
G  C REGloN I t, -[  475 ALLENDALE RoAo
%- d''  KING oF PRUSSIA, PENNSYLVANIA 19406 1415
*****
June 3, 1998 Mr. J. 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.
 
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.
L 4h+oo*V 3/A      _ _ - - - _ - _ - - _ - _
J
 
_ _ _ _ - - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _-___________ _-_- _ - ._ _ -
Mr. J. ' 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
 
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.
 
l I
l ii a
 
_ - _ _ _ - _ _ - - _ _ _ _ _ _ _ - --    ,
        - -
        ;
I Report Details  l
        )
1. Operations  ]
 
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 l
i        )
 
- _ - _ _ _ _ _ _ _ . _ _ _ _ _ - _ - _ - _ . _ - - - _ _ _
.  .
 
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.  >
l
          !
l
 
1
 
I
 
r L
 
- - - - - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ .
c ,
 
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-        - . _ . __ _________ ________________ _
 
_ _ _ - _ _ - _ _ _ _ _ _ _ - - _ _ _ - _ _ _
. .
I l
 
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
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\
L________.--_______         ..a
 
    - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
,
        . .
:
 
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 i
;        . _ _ _ - - - 4
 
    - - . - _ _ - - - - _ - - _ _ _ - - - _ _ _
,  ..
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 l
        ;
 
- - - - _ - - - _ - - - -      I
 
__ - _ _ _ . _ _ _ _ _ _ - _ - _ - _ - _ - _ _ _ _ _ _ - _ . _ _ _ _ -
        . .
EXECUTIVE SUMMARY Beaver Valley Unit 1 Nuclear Power Plant Inspection Report No. 50-334/98 300 Operations
 
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.


9806100390 980603 POR V ADOCK 05000334 pg
ii
&_________:-__________ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . - _ _ J
...


h i-t EXECUTIVE SUMMARY Beaver Valley Unit 1 Nuclear Powar Plant inspection Report No. 50-334/98-300 Operations Three Unit 1 senior reactor operator instant (SROI) candidates passed all portions of the initial license examinatio 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, l
    '
succinct, and that all crew members were kept wellinformed. Crew briefings were routinely held during those instances in which time permitted. The written examination was developed at tha appropriate SRO knowledge level, as werc the job performance 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 examinatio A;l th aa credidates met or exceeded Regulatory Guide (REGUIDE) 1.8, Rev. 2 for a Senior Reactor Operator instant Licens 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, il t    _ _ _ - . . _ _ . - - - _ - - - _ _ _ - - - _ .
. >
        ...
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- _- _ _ _ -- -      )
 
      . .
 
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
 
j L
 
        - - - - - _ _ _ _ _ _ - - . - . - . - - - _ _ - - - _ - _ - _ - -
.  .
 
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 Report Details I. Ooerations 05 Operator Training and Qualificatiora 05.1 Senior Reactor Ooerator Initial Examinations 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 Revision 1 of NUREG-1122," Knowledge and Abilities Catalog for Nuclear Power
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.
;- Plant Operators: Pressurized Water Reactors." The NRC examiners administered I- initial operating licensing portion of the examination to three Unit 1 senior reactor operstor instant (SROI) candidates. The facility's training organization administered i
the written examination.


! Observations and Findinas The results of SRO examination for Unit 1 are summarized below:
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.
SRO Pass / Fail Written ' 3/0 Operatirig 3/O 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 revie 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 simulatos scenarios and JPM's, because little or no changes were required after the demonstration . 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 NalC 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 i
. -
      .I


question 85, reaction W rne 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)
The licensee requires that each candidate maintain a training note book to document the above requirements. The instructions and requirements for maintaining the   '
remediate the candidates on missed questions and (2) perform an analysis to determine if training outlines should be enhance 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 up with two system-related questions. All candidates were also examined using JPMs and/or questions to evaluate the administrative requirement portion of the examinatio Simulator and JPM performance by the candidates was very goo Com.munications 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 scenario 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 are 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 rio an excellent job in adhering to the examiner standards and in developing the examination materials needed to administer the examination l
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
_ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ ______          _b
 
__
    . _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _
_
          . .
 
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.


    ~3 05.2 Angk: ant Trainina and Experience a.. Scope i
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.
Regulatory Guide 1.8 (REGUIDE), Rev. 2 requires certain requirements and cer+.ain 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 requirement Observations and Findinas
  - REGUIDE 1.8 requires that:
1 .
        .
e Each candidate, for a senior license, have a high school diploma or equivalent. The inspectors verified that all candidates met or exceeded the requiremen '
e Each candidate, for a senior license, have four years of responsible power plant l  experience. The inspectors verified that all candidates met or exceeded the requiremen e 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 excee%d the requiremen o 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 requiremen 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 M the TAM reflected the requirements of Regulatory Guide 1.8, Rev. 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 requirernents. The insp9ctor found D  inconsistencies within the documents. The TS referenced, "section 5.5 of ANSI l  lN18.1-1971 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 versio The TAM referenced, "REGUIDE 1.8."


L
L_        _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _
,:;
i u_______i._ __.i_. _ _ _ _ _ . .
  -


The licensee was conducting their training of perspective operators in accordance with REGUIDE 1.8, Rev. 2. This is delincated in the TAM. Se licensee issued Condition Report (CR) 980734, on April 9,1998, that descri es the inconsistenc 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 actio 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 document i E8 Review of the FSA 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 license V. Manaaement Meetinas X1 Exit Meeting Summary 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 examinatio 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 meeting _ _ _ _ _ _ _ _ _____ _____._________________ _ _
  - - - _ - _ _ _ _ _ _ _ _ _ - - _ - _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _
        ,
.. .


(;
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 l    5 PARTIAL LIST OF PERSONS CONTACTED SfAVER 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 l-B. Tuite, General Manager, Nuclear Operations L. Shad, Simulator Supervisor NEG T. Kenny, Senior Operations Engineer, Chief Examiner T. Fish, Operations Engineer
L. Shad, Simulator Supervisor NBC T. Kenny, Senior Operations Engineer, Chief Examiner l T. Fish, Operations Engineer J. D' Antonio, Operations Engineer Attachments:
~ J. D' Antonio, Operations Engineer Attachments:
1. Beaver Valley Unit 1 SRO Written Examination w/ Answer Key 2. Simulation Facility Report
1. Beaver Valley Unit 1 SRO Written Examination w/ Answer Key 2. Simulation Facility Report
_ - _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _
<
l l
l l
l
        --- --- J
 
8 O Attachment 1 BV-1 SRO WRITTEN EXAMINATION W/ ANSWER KEY l
l
. . .
 
- _ _ _ _ - - _ _ _ _ _ _ - - _ . - _ - - - _ _ _   -- _ - _ _ _ _ _ _ _ _ _ _ _ _    ._  . - - _ _ .
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
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f I
      -
Attachment 1 BV-1 SRO WRITTEN EXAMINATION W/ ANSWER KEY
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        - - _ _ _ _ _ __-._________
 
_ _ _ _ _ _ _ _ _ - _ - _ _ - _ - - - - - -
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
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -


i
_ _ _ _ - _ - - - -
!-
,Questiop Topic: l TS SDM & Emergency Boration Given the following conditions:
Quest 6om Topic: l Temperature trending during cooldown A cooldown is in progress. The milestones listed on Figure 1 of 10M-51.4C,(see attached) were reached at the following times:
=
! RCS T,,, - 355 F
  *
  *
(1) 0800
! RCS pressure - 400 psig
+
(2) 0833-
  *
  *
  (3) 0857
RCS boron concentration 2000 ppm
  . Shutdown margin is below Technical Specifications allowable value a
Emergency Boration is initiated at 30 gpm boric acid
  *
  *
.(4) 0917 What action, if any, is required to be taken to comply with Technical Specifications? RCS cooldown is acceptable to this point. RCS cooldown rate will not be exceeded if Figure I time limits are complied with from this point on, 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.
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.


i
, RCS cooldown exceeded Technical Specifications. RCS temperature must remain constant until
'
'
092 d. RCS cooldown exceeded Technical Specifications. Cooldown rate must be restored to within .
Obj Emergency Boration  IOM-7.4.S IV.A S2  Iss 4 Rev
Technical Specification limits by 094 A .s: Ia l Exam Level: lS l Cognitive Level: l Application l Explanatio nef Answer KA: l2. 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:
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.
Knowledge of operator responsibilities during all modes of plant operatio Reference Reference Number Reference Section Page Number (s)  Revision Lear Obj Station Shutdown - 10M-51. IV.A.1 ss 4 Rev Cooldown From MODE 3 to        12  ,
 
MODE 4 Beaver Valley - Unit 1  3.4. /4 4-22,4-27    Amend Technical Specifications         No.179 OM 6,7 & 10 Operational LP-SQS-RX IV. Lecture Qrestion Source l New l Question Modification Method l QIestion Source Comments: l M terial Required for Figure 1 of OM-51.4.C - Blowup curve to max 81/2 x 11 Examination:
Examination:
Pagt,1
l l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _
!
l Page5 L_____-__.


I l
'
Questio2 Tepic: l Core Sifety Limit Curve eval At 20% power, the ruaximum allowable T,,,is limited by the Reactor Core Safety Limit. The basis for l limiting T,,, under these conditions ensures that:
        , ,
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
'
'
a. DNBR remains greater than or equal to the safety analysis DNBR limit and the average enthalpy at the vessel exit will not exceed saturatio b. DNBR remaint, greater than or equal to the safety analysis DNBR limit and the highest enthalpy anywhere in the core will not equal saturatio DNBR remains less than the safety analysis DNBR limit and the average entha;py at the vessel exit l will not exceed saturatio !
        ;
d. DNBR remains less than the safety analysis DNBR limit and the highest enthalpy anywhere in the !
l e. Two SROs are required to approve the manipulation. 1 b. Specific permission is required from the NSS.
core will not exceed saturatio Ais: la l Eram Level: lS l Cognitive Level: l Memory l Explanatio e cf Answer KA: l 2.1.10 l RO Value: l2.7 l SRO Value: l3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title:        ,


KA Conduct of Operations      i St:tement:      l Knowledge of conditions and limitations in the facility licens Rt.ference  Reference Number Reference Section Page Number (s) Revision Lear Obj Reactor Protection System LP-SQS- II. .c i
c. Either the NSS or ANSS has to approve the manipulation.
l Q estion Source l New  l Question Modification Method l Question Source Comments: l M:terial Required for TS Figure 2. Ex:mination:
 
        ,
d. The General Manager, Nuclear Operations.
        .
Page 2


_-
.
Questio] Topic: l TS 3. During power operation the Diesel Generator #1 is declared inoperable. Subsequently the 1B Quench Spray pump is determined to be inoperabl Assuming all required surveillance are completed satisfactorily, what is the required Technical Specification action?        l I Restore both the 1B Quench Spray and Diesel Generator #1 operable status within 72 hours or be in !
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:
Hot Standby within the following 6 hour j 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 hour Restore the IB Quench Spray pump to operable status within one hour or be in Hot Standby within the following 6 hour d. Restore the 1B Quench Spray pump or Diesel Generator # 1 to operable status within 2 hours or be in Hot Standby within the following 6 hour Ars: ld l Exam Level: lS l Cognitive Level: l Application l Explanatio ufAnswer 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:
KA Conduct of Operations Statement:
Ability to apply technical specifications for a syste Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Technical Specifications  TS 3.0.5,3.6.2.1, 3.8. Containment LP-SQS-13.01    5 12 Depressurization Systems Qrestion Source l Facility Exam Bank l Question Modification Method l QIestion Source Comments: l M:terial Required for Technical Specifications Examination:
Knowledge of how to conduct and verify valve lineups.
Page 3
 
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 Tcpic: l FFD requirements
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.


What are the fitness for-duty requirements, with respect to alcohol, for an unscheduled RO who has been  ]
l called out?
l
l
!
The on the spot change:
a.


a. The RO may report to work if he/she has consumed alcohol within the past FIVE hours, but will be required to pass a breath analysis tes I
can be approved by TWO members of management, ONE holding a valid SRO license on Unit 1.
        <
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 NS 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 tes d. The RO shall not report to work if he/she has consumed alcohol within the past FIVE hours.


Ars: la l Exam Level: lS l Cognitive Level: l Application l Explanatio c cf 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 SRO Group: l1 System / Evolution Title:
b. becomes effective 14 calendar days following review by the OSC and approval of the GMNO.
KA Conduct of Operations St:tement:
 
Knowledge of facility requirements for controlling vital / controlled access.
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        '
 
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.
 
.


R;ference  Reference Number Reference Section Page Number (s) Revision Lear Obj Fitness-For-Duty Program 1/2 NPDAP 2.14 IV.2 & 3  2 0 For Duquesne Light Employees        i Conduct Of Operations 1/2LP-SQS-4 Vil ,39 )
      --_----__ - _ - ___
Q:estion Source l New  l Question Modification Method l Qdestion Source Comments: l M;terial Required for 1/2 NPDAP 2.14 Examination:
        . .
l 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?
l Page 4
a. The ANSS blocks the non applicable portions.


Question Topic: l TS SDM & Emerg'ncy Boration Given the following conditions:
b. The STA blocks the non-applicable portions and the RO verifies they are correct.
a RCS T,,, - 355 F a RCS pressure - 400 psig
* RCS boron concentration - 1000 ppm o Shutdown margin is below Technical Specifications allowable value
* Emergency Boration is initiated at 30 gpm boric acid
* A 70 ppm RCS boron concentration change is required to restore the required SDM Of times listed below, which is the MINIMUM emergency boration time that will ensure the required boric acid has been added? minutes minutes
      . c. 21 minutes d. 24 minutes Ans: lc 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. The correction factor of a cf Answer 1.18 multiplied by 500 would result in 590 gallons of boric acid. 590/30gpm = 19 minutes 40 second KA: l 2.1.25 l RO Value: l2.8 l SRO Value: l3.1 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution        )
Title:
          )
KA Conduct of Operations Statement:
- Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance dat Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj ]
Emergency Boration  IOM-7. I S2  1ss 4 Rev i
i Beaver Valley Unit 1 -  3.1. /4 1 1 Amend Technical Specifications      No. 91 CVCS  LP-SQS- I Question Source l New  l Question Modification Method l    l
          '
Question Source Comments: l Material Required for IOM-7.5 Figures 7-7,7-8 & Table 7- Examination:        l


l Page 5 t          i 1        ---_- i
c. The system engineer blocks the non-applicable portions and the ANSS verifies they are correct.


Question Topic: l Parmission for deviation from NS In addition to normal requirements for manipulating components, which of the following describes who is required to approve placing component in other than its Normal System Alignment (NSA)? Two SROs are required to approve the manipulatio b. Specific permission is required from the NS Either the NSS or ANSS has to approve the manipulatio d. The General Manager, Nuclear Operation ,
d. The PO blocks the non-applicable portions and the RO verifies they are correct.
Ans: lc l Exam level: lS l Cognitive Level: l Memory l Explanatio o cf Answer KA: l 2.1.29 l RO Value: l3A l SRO Value: l3.3 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title:
KA Conduct of Operations Statement:
Knowledge of how to conduct and verify valve lineup Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Question Source l Facility Exam Bank l Question Modification Method l  .'
Qzestion Source Comments: l M;terial Required for Ex:mination:
        )
l I-Page 6


Questio2 Tcpic: l Procedure change rules for type of procedure While at 100% power, an OMCN is to be written to change IOM-7.4.L " Blender Boration Operation." This change adds a step that directs placing ONE bank of Pressurizer heaters in MANUAL prior to initiating a boration. An Operations Unit Non-Intent Reviewer has determined that this does NOT change the intent of the procedur The on the spot change: can be approved by TWO members of management, ONE holding a valid SRO license on Unit b. becomes effective 14 calendar days following review by the OSC and approval of the GMN cannot be made because use of the procedure is not expected in the next 30 day d. cannot be rnade because this is a safety related procedur ATs: Ia l Exam Level: lS l Cognitive Level: l Comprehension l Esplanatio
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:
  '
acf Answer
  -
KA: l2. l RO Value: l2.3 l SRO Value: l3.3 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title:
KA Equipment Control Statement:
KA Equipment Control Statement:
Knowledge of the process for making changes in procedures as described in the safety analysis repor Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Control Of Operating 1/2OM-48. C. I .a B 10  iss 4 Rev Procedures      13 Conduct Of Operations I/2LP-SQS-4 . , 9 QIestion Source l New  l Question Modification Method l QIestion Source Comments: l M:terial Required for Examination:
Knowledge of surveillance procedures.
,
 
Page 7
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
 
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.


,Q uestion Tcpic: l Omissions in OSTs A partial OST is to be performed. Which of the following is an acceptable method of blocking the portions of the OST that are NOT applicable? The ANSS blocks the non-applicable portion b. The STA blocks the non-applicable portions and the RO verifies they are correc c. The system engineer blocks the non-applicable portions and the ANSS verifies they are correc d. The PO blocks the non-applicable portions and the RO verifies they are correc Ars: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio o cf Answer KA: l 2.2.12 l RO Value: l3.0 l SRO Value: l3.4 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title:
b. Operation of the tagged component will be affected because a portion of the system is not in NSA.
KA Equipment Control Statement:
Knowledge of surveillance procedure Reference    Reference Section Page Number (s) Revision Lear l Reference Number Obj Adherence and  1/20M-48. VI.B.17  10 iss 3 Rev Familiarization to operating    18 Procedures Conduct Of Operations 1/2LP-SQS-4 i0 Q:estion Source l New  l Question Modification Method l Q estion Source Comments: l M:terial Required for Examination:
Page 8


Question Tepic: l Caution Tags Use of a Caution Tag is PROHIBITED for which of the following conditions? Special additional manual actions are required to operate the tagged componen b. Operation of the tagged component will be affected because a portion of the system is not in NS c. As a temporary replacement for a component label that has fallen of d. As a warning that operation of the component will cause erratic indication.
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.


Ars: lc l Exam Level: lS l Cognitive Level: l Memory l Esplanatio c of Answer KA: l 2.2.13 l RO Value: l3.6 l SRO Value: l l Section: l PWG l ROGroup: l 1 l SRO Group: l1 System / Evolution Title:
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
KA Equipment Control Statement:
l  Title:           '
!  KA Equipment Control St:tement:
Knowledge of tagging and clearance procedures.
Knowledge of tagging and clearance procedures.


Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Use of Caution Tags 1/20M-48. I ,3 iss 4 Rev
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


Conduct Of Operations 1/2LP-SQS-4 Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Ex:mination:
Conduct Of Operations I/2LP-SQS-48.1 V.P  7    10 15
Page 9
.
              ,
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
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _          _.


Question Tcpic: l SRO control Which of the following describes a responsibility of the Refueling SRO during fuel movement?
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
The Refueling SRO will: initial the Fuel Assembly Handling Deviation Report with NSS concurrenc b. be located on the manipulator crane structure during most fuel handling activitie maintain the DLC Master Copy of the Fuel Handling data Sheet d. continuously monitor source range count leve Ars: lb l Exam Level: lS l Cognitive Level: l Memory l Esplanatio c ef Aaswer KA: l2.2.31 l RO Value: l1.6 l SRO Value: l3.8 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title:
              *
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:
KA Equipment Control Statement:
Knowledge of SRO fuel handling responsibilitie Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj R: fueling Administrative Book 1 -lRP-12R- II.DA.b.15) 10 Iss 0 Rev Section     0 Fuel Handling Operations LP-SQS-6.13 Il .b Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for Ex::mination:
Knowledge of SRO fuel handling responsibilities.
   .
 
l Page 10 l
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 High Radiation Definition l Technical Specifications requires radiction areas to be isolated by locked doors if the radiation levels are greater than: mrem /hr b. 500 mrem /hr mrem /hr            i
Question Topic: l SRO action for gas releise Given the following conditions:
              '
  * Reactor power- 100%
l .
  * Discharge of Waste Gas Decay Tank [lGW-TK-1 Al is planned for 1000 on 4/22/98
i d. 5000 mrem /hr Ans: lc l Eman 12 vel: lS  l Cognitive Level: l Memory  l Explanatio eef Answer KA: l2. 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:
  * . The RWDA-G had been approved on 1500 on 4/20/98
KA Radiation Control Statement:
  + The meteorological information indicates Stability Class A for atmospheric conditions
Knowledge of 10 CFR: 20 and related facility radiation control requirement Reference  Reference Number  Reference Section Page Number (s)  Revision  Lear Obj Technical Specifications    6.12  6-23  188
  * "Ihe status of the Gaseous Effluent Monitors is as follows:
   .
  . Gaseous Waste / Process vent [RM-GW-108A] noble gas channel inoperable
Q'estion 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 Ex:mination:
   - 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?
l l
a. The release can be initiated without restriction.
Page11


Question Tople: l SRO action for gas release Given the following conditions:
* Reactor power- 100%
* Discharge of Waste Gas Decay Tank [lGW-TK-1 A] 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
.* The status of the Gaseous Efiluent 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 .
equipment status and other conditions do NOT change? The release can be initiated without restrictio .
b. The release can be initiated only if sampling of the release stream is analyzed at least one per every FOUR hours.
b. The release can be initiated only if sampling of the release stream is analyzed at least one per every FOUR hours.


, The release cannot be made because the 72-hour effective time limit for the RWDA-G has elapse d. The release cannot be made because the Stability Class for release is unacceptabl Ars: lc l Esam Level: lS l Cognitive level: l Comprehension l Explanatio a cf Answer KA: l2. 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:       -
c. The release cannot be made because the 72-hour effective time limit for the RWDA-G has elapsed.
KA Radiation Control Statement:
 
Knowledge of the requirements for reviewing and approving release permit Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Decay Tank Discharge IOM-19. step 7 NOTE E3 iss 3 Rev
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
 
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


Gaseous Waste Disposal LP-SQS-1 .0, ODCM 3.3.3.10 17 13 5 System Question Source lNew  l Question Modification Method l Question Source Comments: l Material Required for IOM-19. Examination:
  , _ _ _ .
,.
I Given the following conditions:
Page 12 c_  _
:
i The reactor has been shutdown for 2 days.


.- - - _ _ _ _ . _ _ _ - . - - - _ _ _ _ - _ _ _ - - _ _ _ _ _ _ - _ - _ - _ - _ - _ _ _ - _ - _    . _ -_ _ _ _ - _ . _
L * RCS temperature is 150 'F.       '
e Question Topic: l Time to Core Boiling Given the following conditions:
l
  * The reactor has been shutdown for 2 day * RCS temperature is 150 F.
'
* RCS pressure is atmospheric.


L  * RCS pressure is atmospheri .
* PZR is a normal level for shutdown cooling.
  -* PZR is a normal level for shutdov;n coolin Assume RHR is lost. Which of the following describes the time available until core boiling occurs?
 
  ( Using the attached references, AOP 1.10.1 attachments 1,2,3, & 4)
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.
a. Less than 10 minutes.


l  b. Il'to 20 minute to 30 minutes, d. 31 to 40 minute Ams: ld l Exam Level: lS     l Cognitive Level: l Application l Explanatio e of Answer KA: l2. l RO Value: l3.3 l SRO Value: l l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title:
. b. I1 to 20 minutes, c. 21 to 30 minutes.
KA Emergency Procedures / Plan Statement:
 
Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR mitigation strategies).
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.


Reference    Reference Number  Reference Section Page Number (s) Revision Lear Obj ResidualHeat Removal  AOP 1.1 , Attachment i Iss 3 A System Loss          Rev 5 Residual Heat Removal  LP-SQS-1 ,10 System Question Source l Previous 2 NRC Exams l Question Modification Method l Question Source Comments:  l Material Required for  AOP 1.10.1 Attachments I,2 3 & Examination:
* 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.


        -      l
* The STA informs' crew that B loop Two is 283 F and slowly dropping.
        .


i Page 13    i s
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".


f C : ion Topic: l Implementation of Orange Path l Given the following conditions:
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:
* An unisolable steam line break has occurred on SG "B"
+ SG "A" and "C" levels were overfe * A reactor trip and SI occu * Pressurizer pressure is 1180 psig
* Pressurizer level is 12%
!
* T.v,is 400 F and slowly dropping I
* E-0 " Reactor Trip Or Safety Injection", step 9 is being performe * The STA informs' crew that B loop T u is 283 F and slowly droppin What is the EOP flowpath that will be followed given the above conditions? Immediately transition to FR-P.1 " Response To imminent Pressurized Thermal Shock Condition", Perform actions of E-0 through diagnosis of steamline break, then transition to E-2 " Faulted Steam Generator Isolation"  . Perform actions of E-0 through diagnosis of steamline break, then transition to FR-P.1 " Response to Imminent Pressurized Thermal Shock Condition" Perform actions of E-0 through diagnosis of steamline break, then transition to FR-P.2 " Response to Anticipated Pressurized Thermal Shock Condition" Ass: lc l Exam tevel: IS l Cognitive Level: ! Application l Explanation ofAnswer KA: l 2.4.14 l RO Value: l3.0 l SRO Value: l l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title:
KA  Emergency Procedures / Plan Statement:
KA  Emergency Procedures / Plan Statement:
Knowledge of general guidelines for EOP flowchart us Reference  Reference Number Reference Section Page Number (s) Revision Lear Ob]
Knowledge of general guidelines for EOP flowchart use.
Subcriticality - Status Tree F- ORANGE PATH  IssIB Rev1 Reactor Trip Or Safety IOM-53B.4.E-0 1. Ist paragraph 1 IssIB Injection Background      Rev 5 EOP Introduction  LP-SQS-5 I
 
        '
Reference Section Page Number (s) Revision Learn.
Question Source l New l Question Modification Method l Question Source Comments: l Material Required for F-0.4 and Att 5-D Examination:
 
Page 14
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        _ ____ ________ _ _


Questici Topic: l EOP Usags During Critical Safety Function Status Tree monitoring it was determined that TWO functions had Orange P:ths. One of the Orange paths is FR-H.1, Response to Loss of Secondary Heat Sink.
. - __
  ,_ -
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?
Which Critical Safety Function, also Orange, would take precedence over FR-H.l?
a. FR-C.1, Response to Inadequate Core Cooling b. FR-Z.1, Response to High Containment Pressure FR-P.1, Response to Imminent Pressurized Thermal Shock Condition d. FR-1.1, Response to High Pressurizer Level A s: la l Exam Level: lS l Cognitive Level: l Comprehension 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:
Explanatio a cf Answer KA: l 2.4.16 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 Emergency Procedures / Plan Stat: ment:
KA Emergency Procedures / Plan Stitement:
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.
Knowledge of EOP implementation hierarchy and coordination with other support procedures.


Reference  Reference Number Reference Section Page Number (s)  Revision Lear Obj EOP Executive Volume - 1/20M-53 .B  9  IssIB User's Guide       Rev 3 EOP Introduction LP-SQS-5 Question Source l Facility Exam Bank l Question Modification Method l QIestion Source Comments: l M;terial Required for Examination:
Reference Section Page Number (s) Revision Learn.
Page 15
 
_______________-___-____ _ - _ -
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 l--    _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
 
__
_ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ - -_ _ _ - _ _ -
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.


Questio] Te pic: l Functionil Recovery Procedure usage During a loss of all Emergency 4KV AC Power, When are Functional Restoration Procedures implemented?
Obj Fire Protection  NPDAP 3.5  111.N  3 6 Conduct of Operations 1/2LP-SQS-48.1   10  1 l
a. Immediately upon electrical power restoration to I AE or ID Immediately upon exiting ECA-0.0 " Loss of all 4KV AC Emergency Power " When directed by ECA-0.1 " Loss of all Power Recovery Without SI Required" or ECA-0.2. " Loss t of all AC Power Recovery With SI Required"
Question Source l New l Question Modification Method l Question Source Comments: l~
  ' When ECA-0.1 " Loss of all Power Recovery Without SI Required" or ECA-0 2. " Loss of all AC Power Recovery With SI Required" is completed.
M;terial Required for Ermination:
!
!    Page 17
          ;


l l
______-____ _ _____ -__- _ _ _
l A:s: lc l Exam Level: lS l Cognitive Level: l Memory l
            ^
        .
            *
Explanatio a cf Answer
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: l 2.4.16 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 Ability to predict and/or monitor changes in parameters associated with operating the Control Rod Drive System Statement: controls including:
KA Emergency Procedures / Plan St:tement:
T-ref Reference  Reference Number Reference Section Page Number (s) Revision Learn.
Knowledge of EOP implementation hierarchy and coordination with other support procedure Reference  Reference Number Reference Section Page Number (s) Itevision Lear Obj EOP Executive Volume - 1/20M-53 V issIB )
 
User's Guide      Ret 3 EOP Introduction LP-SQS-5 IV. Q:estion Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l
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
Mrterial Required for Ezrmination:       )
Page 18 i
'
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _
 
. _ _ _ _ _ _ _
 
Giv:n the following conditions:
          :
* Reactor Power - 72%        '
* Control Rods are at step 210 on Control Bank D
 
* 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
i
        .
          !
Page 16
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 Question Source lNew  l Question Modification Method l Question Source Comments:  l Material Required for Examination:
l Page 19 I
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              *
Question Topic: l Subcooling Margin During a natur:.1 circulation cooldown the required number of CFDM fans cannot be started.


Questio2 Topic: l Fire Brigade Responsibility s During a plant fire, who is responsible for coordinating fire-fighting activities with the offsite fire department l chiefs?
During the cooldown, upper head voiding is prevented by:
l i The ANSS when acting as the Fire Brigade Chie b. The ANSS when acting as the Fire Brigade Captain.
a. venting the head via reactor vessel head vents, b. verifying incore thermocouple temperatures are within an allowable range ofloop temperatures.


1              h c. The affected Unit's NSS.
c. increasing the minimum subcooling margin during portions of the cooldown.


'
d. periodically injecting cold Safety injection water into the Hot legs.
d. The Nuclear Operator when he/she is acting as the Fire Brigade Captai Ans: la l Exam Level: lS  l Cognitive 12 vel: l Memory    l Explanatio o ef Answer KA: 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 System / Evolution Title:
 
KA Emergency Procedures / Plan Statement:
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:
Knowledge of fire in the plant precedur Reference     Reference Section   Page Number (s) Revision Lear l Reference Number        Obj Fire Protection  NPDAP Il Conduct of Operations 1/2LP-SQS-4 Question Source l New  l Question Modification Method l Qrestion Source Comments: l Material Required for Examination:
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.
Page 17
 
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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, M:terial Required for Ermination:
Page 20 L-            _ _ - __-_ - .
 
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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.


a    u
d. prevents exceeding RCS cooldown rates.
          ,4  .
  *
          .
Question Topic: l Rod motion control If a power mismatch signal is generated by the Rod Control System, which of the following parameters determines the magnitude of the gain imposed by the variable gain unit? Median Tave i Median delta T N44 Power Turbine Impulse pressure Ans: ld    l Exam Level: lS    l Cognitive Level: l Memory l l
Explanatio n of Answer i  KA: l 001 Al.02      l RO Value: l3.1 l SRO Value: l3.4 l Section: lSYS l RO Group: l 1 l SRO Group: l1 l
l  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 Lear Ob)
Reactor Control and      lOM-l.$.A.51  1 iss 4 Rev Protection            0 Re:ctor Control and      10M-l . i .D  13 iss 4 Rev 13 Protection            1 Full Length Rod Control      LP-SQS- l  Question Source l NRC Exam Bank      l Question Modification Method l l
Question Source Comments:      l M:terial Reouired for Examination:
I
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Page 18
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      *
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.
            .
            .
                .
I'
Question Topic: l Misoperation of Bank Selector Switch Given the following conditions:


    *  Reactor Power - 72%
Obj Reactor Coolant Pump IOM-6.4.A II.V  3 iss 4 Rev Startup
    . Control Rods are at step 210 on Control Bank D
    * 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 l    instead of MANUAL
!    * Rods are withdrawn 5 steps before this is discovered l
If the Control Rod Bank Sel Sw is placed in Manual at this point, which of the following will occur? Upon shutdown, all Control Bank D rods will remain 5 steps withdrawn from the cor b. Upon shutdown, the ROD BOTTOM / ROD DROP alarm will actuate 5 steps sooner than expected.


, While operating, the Rod Insertion Limit alarms (A4-116 and A4-134) for Control Bank D would l      actuate 5 steps lower than the actual alarm setpoint positions.
RCS - Reactor Coolant LP SQS-6.3  Ill.A  24 4 12.A Pumps


i l While operating, the Bank Demand Position Indict. tion will read 5 steps lower than the Analog Rod l       Position Indication.
Question Source l New  l Question Modification Method l  I Qrestion Source Comments: l Matrial Required for Ex;mination:
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l Page 21 u___.________        J


l
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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.
Ars: la    l Exam Level: lS    l Cognitive Level: l Comprehension l l  Explanatio a cf Answer              ,
KA: l 001 K4.02      l RO Value. l l SRO Value: l3.8 l Section: lSYS l RO Group: l 1 l SRO Group: l1 I
System / Evolution        Control Rod Drive System Title:
KA      Knowledge of Control Rod Drive System design feature (s) and or interlock (s) which provide for the following:
i  St:tement:
l        Control rod mode select control (movement control)
R:ference        Reference Number Reference Section Page Number (s) Revision Lear Obj reactor Control & Protection        I OM-1. Bank Overlap 15-16 iss 4 Rev-Instrumentation and             1 Controls Fulllength Rod Control        LP-SQS- Ill. .a QYestion Source l New          l Question Modification Method l Q estion Source Comments:        l Miterial Required for              l Examination:
,.
l Page 19
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Questio] Tepic: l Subcooling Margin      l During a natural circulation cooldown the required number of CRDM fans cannot be starte During the cooldown, upper head voiding is prevented by:
Which of the following lists all running RCPs?
e. venting the head via reactor vessel head vent b. verifying incore thermocouple temperatures are within an allowable range ofloop tensperature increasing the minimum subcooling margin during portions of the cooldow d. periodically injecting cold Safety Injection water intc the Hot leg Ars: lc l Exam Level: lS l Cognitive Level: l Comprehension i Espirnatio a cf Answer       ,
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: l 002 K5.15 l RO Value: l4.2 l SRO Value: l4.6 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Reactor Coolant System Title:
KA Knowledge of electrical power supplies to the following:
KA Knowledge of the operational implications of the following concepts as they apply to the Reactor Coolant System:
Statement:
Statement:
Reasons for maintaining subcooling margin during natural circulation Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj EOP Generic issues LP-SQS-5 Natural Circulation IOM 53B.4.ES- IssIB Cooldown Background      Rev 4 Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l M:t: rial Required for Ex:mination:
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 -
i Page 20
Remor Coolant Pumps Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for Examination:
f
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Question Topic: l SG temperature effect upon start of RCP Given the following conditions:
Given th2 following conditions:
  * Plant heatup in progress
  * Plant heatup in progress
  * RCS temperature - 175 F
  * RCS temperature - 175 F
  * RCS pressure -325 psig
  * RCS pressure - 325 psig
,
  * Charging pump [lCH-P-1B] is in service.
  * Pressurizer level- 28%
 
l * 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: is not applicable since this is the first RCP to be starte b. prevents an RCS overpressure even prevents exceeding RCS heatup rate =.
* Charging pump [1CH-P-1 A] is inoperable.
      ~ ~
 
d. prevents exceeding RCS cooldown rate Ans: lb l Exam Level: lS l Cognitive Level: l Memory  l Explanatio
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?
  ::cf Answer KA: l 003 Kl.10 l RO Value: l3.0 l SRO Value: l3.2 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution l Reactor Coolant Pump System Title: l KA Knowledge of the physical connections and/or cause-effect relationships between Reactor Coolant Pump System Statement: ard the following:
a.
RCS Reference  Reference Number Reference Section Page Number (s) Revision ' Lear Obj Reactor Coolant Pump IOM-6. I iss 4 Rev Stanup      7 RCS - Reactor Coolant LP-SQS- li .A Pumps Question Source l New  l Question Modification Method l Question Source Comments: l M:terial Required for Examination:        l Page 21
 
        ._ _ _ - - _ _ - _ _ _ _ _ -
(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:       {


Questio: Tcpie: l RCP power supplies The reactor is at 35% with the electrical busses in NSA. Unit Station Service Transformer ID develops a fault opening [4KV ACB 241D) USST ID Supply to 1C 4KV Bus and [4KV ACB 341D] USST 1D Supply to ID 4KV Bus The auto bus transfer fails to operate on C & D Bu Which of the following lists all running RCPs? RCP1A b. RCP 1 A.and 1B RCP IB and IC RCP IC Ats: lb l Esam Level: lS l Cognitive Level: l Memory l Esplanatio e af Answer KA: l 003 K2.01 l RO Value: l l SRO Value: l3.1 l Section: lSYS l RO Group: l 1 l SRO Group: l1 Syst:m/ Evolution Reactor Coolant Pump System Title:
1 KA Conduct Of Operations Statement:
KA Knowledge of electrical power supplies to the following:
Ability to apply technical specifications for a system.
Statement:      _
RCPS Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj 4KV Distribution System LP-SQS 3 !!1. Reactor Coolant System - LP-SQS- .C. I  i1 4 1 Reactor Coolant Pumps Qrestion Source l Nev/  l Question Modification Method l Qrestion Source Comments: l Material Required for Eemination:
l l
l


Page 22
Ref;rence  Reference Number Reference Section Page Number (s) Revision Learn.


Question Topic: l TS sval for charging pump l
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 QTestion Source Comments: l Mat: rial Required for Technical Specifications Ermi:stion:
Given the following conditions:
l Page 23 l
l . * - Plant heatup in progress
        ;
* RCS temperature - 175 F'
* RCS pressure - 325 psig -
-
i e Charging pump [1CH-P-1B] is in servic '
* Charging pump [lCH-P-1 A) is inoperabl Which of the following describes limitations, if any,if[1CH-P-1C] were to be placed in service on AE Bus, and [1CH P-1B] were to be removed from service?
a.- [1CH-P-1B] must be stopped and placed in PULL-TO-LOCK prior to taking [lCH P-1C] out of PULL-TO-LOC b. [lCH-P_-1B] must be stopped and placed in AUTO prior to taking [1CH-P-1C) out of PULL-TO-'
LOC c . [1CH-P-1B and 1C) r ?v be run simultaneously for up to 15 minutes, after which [1CH-P-1B] must be stopped and placed in PULL-TO-LOC ' d. Both Charging Pumps mg be run without restriction until [1CH-P-1B] is removed from servic Ass: lc l Exam Level: lS l Cognitive Level: l Comprehension l Explomatio i cefAnswer KA: l 2.1.12  l RO Value: l2.9 l SRO Value: l l Section: l SYS l RO Group: l 1 l SRO Group: l1 System /Evolutica Chemical and Volume Control System Title:
KA  Conduct Of Operations Statement:
Ability to apply technical specifications for a syste Reference  Reference Number Reference Section Page Number (s) Revision 12ar Obj Beaver Valley - Unit 1  3.4. /4 4-27a Amend Technical Specifications     No.193 Placing the Spare Charging l OM-7. I W 9-13 Iss 4 Rev 12 Pump into Operation      10 CVCS  LP-SQS- IV.A. B  28  12 Question Source l New   l Question Modincation Method l Question Source Comments: l
      -
Material Required for Technical Specifications Examtmation:
Page 23


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o
Question Topic: l Evd of1:ak in R gin Hx Given the following conditions:
    ~
  * ' Reactor power- 90%
Question Topic: l Ev:1cfleak in Regen Hx Given the following conditions:
  * Pressurizer level- 51% stable
  * Reactor power - 90%
* VCT level - 30% rising
  * Pressurizer level- 51% stable a VCT level- 30% rising o Letdown flow on [F1-CH-150]- 60 gpm
* Letdown flow on [FI-CH-150] - 60 gpm
  * Charging flow on [F1-CH-122] - 45 gpm j
  * Charging flow on [FI-CH-122] - 45 gpm
a Seal Injection flows - 8 gpm (A); 10 gpm (B); 7 gpm (C)     l
. Seal Injection flows - 8 gpm (A); 10 gpm (B); 7 gpm (C)
  * RCP #1 seal leakoff flows - 4 gpm (A); 4 gpm (B); 2 gpm (C)
  * 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 leak exists in the Seal Water Heat Exchange b. RCP #1 Seal Bypass Valve [MOV-CH-307] was inadvertently opene c. Letdown Pressure Control valve [PCV-CH-145] has failed ope d. A leak exists in the CVCS Non-Regenerative Heat Exchenge Ans: la l Exam Level: lS l Cognitive Level: l Comprehension     l Explanatio c cf Answer KA: 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:
Which of the following would result in the conditions above?
KA Knowledge of the of the efreu of a loss or malfunction on the following will have on the Chemical and Volume Statement: Control System:
a. A leak exists in the Seal Water Heat Exchanger.
Heat exchange 3 and condensers Ref;rence  Reference Number Reference Section     Page Number (s) Revision Lear Obj CVCS  LP-SQS- I , 9 Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for Examination:
 
!
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 w__
 
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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.


Question Topic: l RHR/RCS pressure response Given the following conditions:
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 Question Source l New   l Question Modification Method l Question Source Comments: l M:terial Required for Ex mliation:
* Plant cooldown is in progress at 20 F/hr
I l
* RCS temperature - 155 F
Page 25
*
- _ _ _ _ _
Pressurizer level [LI-l RC-462] Cold Calib - 100%
* RHR Pump 1 A is running with flow of 4000 gpm set on [MOV-RH-605] RHR Flowin AUTO
-*
[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%
  * 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? RHR flow will decrease and RCS pressure will decreas b. ' RHR flow will increase and RCS pressure will increas RHR flow will remain the same and RCS pressure will decreas d. RHR flow will remain the same and RCS pressure will increas I Ans: ld l Exam level: lS l Cognitive level: l Comprehension l Explanatio ocf Answer
          '
KA: l 005 K3.01 l RO Value: l3.9 l SRO Value: l l Section: lSYS l RO Group: l 3 l SRO Group: l3 System / Evolution Residual Heat Removal System Title:
KA Knowledge of the effect that a loss or mstfunction of the Residual Heat Removal System will have on the Statement: following:
RCS Reference Reference Number Reference Section Page Number (s) Revision Lear Obj Residual Heat Removal IOM-10. E, F  A 8-9 Iss 4 Rev System Startup(Plant     9 l
cooldown) And Operation RHRS LP-SQS 1 D.2.e, f 7-8 8 5.a. b, f; 10 Question Soutee l New l Question Modification Method j Question Source Comments: l Material Required for Examination:
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Question Topic: l Los ef ONE St Accum Given the following conditions:
'
  * Reactor power is 55%
Question Topic: l Loss cf ONE St Accum Given the following conditions:
  *
  * Reactor poweris 55%
Accumulator [ISI-TK-1 A) level is 85%
  * - Accumulator [1SI-TK-1 A] level is 85%
l
  * Accumulator [1SI-TK-1 A] pressure is 657 psig
  * Accumulator [1SI-TK-1 A] pressure is 657 psig
  * SI Accumulator Isolation Valve [MOV-1SI-865A] is closed
  * SI Accumulator Isolation Valve [MOV-ISI-865A] is closed
  * The lockoutjack is removed a Reactor shutdown was initiated due to the accumulator conditions
  * 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?
l Which of the following states the response of the SI Accumulators if a Design Basis LOCA occurs on the
a. THREE Accumulators will fully inject into the core.
'
 
Loop B Cold Leg? THREE Accumulators will fully inject into the cor b. THREE Accumulators will fully inject into the core, provided the operator manually opens [MOV-ISI-865A]. TWO Accumulators,1B and IC, will fully inject to the cor d. ONE Accumulator, IC, will fully inject to the cor Ans: ld l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio 'Ihe 1B Accumulator will discharge through the break o gf Answer KA: l 006 K6.02 l RO Value: l3.4 l SRO Value: l3.9 l Section: lSYS l RO Group: l 2 l SRO Group: l2 System / Evolution Emergency Core Cooling System Title:
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:
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 flor 4 tanks (accumulators)
Core flood tanks (accumulators)
Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj i
Reference  Reference Number Reference Section Page Number (s) Revision Learn.
SIS   LP-SQS-1 Vill.D.), XI. ,23 4 7.d,1 Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for Ermination:
 
j l
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.


Page 26 l
Obj Alann - Pressurizer Relief lOM-6.4.AAF PC No. 2 AAF 2-3 iss 4 Rev Tank levelliigh-Low
        - _ _ __-______ - -_ -


j.
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 Qrestion Source Comments: l M;terial Required for Examination.


j I
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Question Topic: l Source of PRT conditions Reactor is a 100% with all systems in NSA. The operator observes that PRT level has increase Which of the following can cause the level increase?
Page 27
a. . A relief valve on the CCR system inside containment has lifte b. RCP #2 Seal Leak off flow has increase A PORV is leakin d. RCP #1 Seal Leak off flow has increase Ars: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio n ef Answer KA: l 007 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:
 
Statement:
    ~ ~~
Components which discharge to the PRT Reference  Reference Number Reference Section Page Number (s) Revision Lear O,,bj Alarm - Pressurizer Relief lOM-6.4.AAF  PC No. 2  AAF 2-3 iss 4 Rev Tank Levelliigh-Low      3  i Pressurizer end Pressure LP-SQS- .B. R , lief Systems Reactor Coolant System- LP-SQS- Reactor Coolant Pumps Question Source l New   l Question Modification Method l Question Source Comments: l
    ~
        )
          < ,
Material Required for Examination:
Question hpic: l PORV opss: ion
        !
[MOV-RC-535] Pressurizer Power ReliefIso!ation Vcive is closed due to [PCV-RC-455C] PORV leaking.
l Page 27    l l
 
[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
 
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
 
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
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 Question Source Comments:    l Mat: rial Required for 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.


(        l Question Topic: l PORV operation
All Tm temperatures are equal.
[MOV-RC-535] Pressurizer Power ReliefIsolation Valve is closed due to [PCV-RC-455C] PORV leakin [PT-RC-445] Pressurize Pressure has failed downscal Select the available automatic overpressure protection, if an ) No PORVs will protect against overpressur b. Only PCV-RC-455D will protect against overpressur Only PCV-RC-456 will protect against overpressur i d. Both PCV-RC-456 and 455D will protect against overpressur ~
Ars: la l Eram Level: IS l Cognitive Level: l Application l Explanatio c(f 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  Reference Number Reference Section Page Number (s) Revision Lear Obj Instrument Failure Procedure AOM-6.4-IF  Figure  22  iss 4 Rev


Pressurizer & Pressure Relief LP-SQS- Syrtem Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for IOM-6.4-l P      I Ex:mination:
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.
Page 28


Questici Tcple: l Pressurizer Lev 1 Rx trip ,
b. closer to its OPdeltaT trip setpoint, but will be farther from its OTdeltaT trip setpoint.
Pressurizer Level Control Channel Selector is selected to LT 459 & 460. All plant conditions are stabl Which of the following will result in a reactor trip due to high pressurizer level? At 5% power LT-RC-461 fails lo b. At 5% power LT-RC-459 fails hig At 25% power LT-RC-460 fails low.


l l d. At 25% power LT-RC-461 fails lo Avs: lc l Exam Level: lS l Cognitive Level: l Comprehension l l Explanatio o cf Answer
c. farther from its OPdeltaT trip setpoint, but will be closer to its OTdeltaT trip setpoint.
'KA: l 01i K1.04 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 Statement: System and the following:
RPS Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj ~
RCS - Instrument failure IOM-6.4.IF  ll.a II.C. IF 8-9 iss 4 Rev


Pressurizer and Pressure LP-SQs- .D. Rtlief System Q:estion Source l New  l Question Modification Method l Q estion Source Comments: l M;terial Required for Ex:mination;
d. farther from both OPdelt f and OTdeltaT trip setpoints.


        !
Ans: ia l Eram Level: lS  l Cognitive Level: l Comprehension l  _
Page 29
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.


e Question T:pic: l Evil OTDT & OPDT setpoints on input failure During operation at 97% power one Tuo, instrument is reading 4 degrees higher than other Tuo, instrument All Ta temperatures are equa Which of the following describes the effect on OPdeltaT and OTdeltaT for the loop with the highest T,,,7 Loop deltaT will be closer to both OPdeltaT and OTdeltaT trip setpoint b. closer to its OPdeltaT trip setpoint, but will be farther from its OTdeltaT trip setpoin farther from its OPdeltaT trip setpoint, but will be closer to its OTdeltaT trip setpoin d. farther from both OPdeltaT and OTdeltaT trip setpoint A;s: la l Eram Level: lS  l Cognitive Level: l Comprehension l Explanatio e cf Answer KA: l 012 A2.05 l RO Value: l 3.l* l SRO Value: l 3.2* l Section: lSYS l RO Group: l 2 l SRO Group: l2 System / Evolution Reactor Protection System Title:
Obj RCS Instrument Failure IOM-6.4fF ll.B,111. IF 32-33,35 36 iss 4 Rev
KA Ability to (a) predict the impacts of the following on the Reactor Protection System and (b) based on those Statement: 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. l Obj RCS-Instrument Failure IOM-6.4.lF !!.B, II IF 32-33,35-36 iss 4 Rev


Reactor Protection System LP-SQS- V.C.16  25-26 6 8 Reactor Coolant System LP-SQS- I .a. b Question Source l Facility Exam Bank  l Question Modification Method l Question Source Comments: l MI.terial Required for       i Ex mination:
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 Question Source Comments: l Mr.terial Required for Eximiration:
        !
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RPS testing is in progress for RPS train B and the status of the breakers are as follows:
i
 
- * 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:
Question Topic: l Operation of BOTli Bypass Trip Br akers RPS testing is in progress for RPS train B and the status of the breakers are as follows:
l
* Reactor trip beakers (RTA and RTB) closed l * Reactor bypass breaker B (BYB) closed l
'
Bypassing both RPS trains simultaneously is prevented by:
l tripping only BYA ifit is racked in and its CLOSE pushbutton is depresse b. tripping only BYB if BYA is fully racked in, c. preventing closure of BYA ifit is racked i d. tripping all reactor trip and bypass breakers if BYA is racked in aad its CLOSE pushbutton is depresse A:s: ld l Exam Level: lS   l Cognitive Level: l Memory       l Esplanatio              '
c ef Answer KA: l 012 A3.07 l RO Value: l4.0 l SRO Value: l 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:
KA Ability to monitor automatic operations of the Reactor Protection System including:
Setement:
Statement:
Trip breakers Reference   Reference Number   Reference Section     Page Number (s) Revision Lear Obj Reactor Control and  lO M 1. RP,2nd paragraph    2  iss 4, Protection - Summary              Re Description Reactor Protection System  LP-SQS- .1      7  6 8, 9 Hardware QIestion Source l Facility Exam Bank    l Question Modification Method l Qrestion Source Comments:  l Material Required for Examination:              ,_
Trip breakers l
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Reference Reference Number Reference Section Page Number (s) Revision Learn.
.. _ _ _ .


l Question Topic: l Containment Pressura logics Containment pressure instrument PT-LM-100C has failed downscale. All appropriate actions of lOM-1.4.IF, Instrument Failure Procedure, have been completed.
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 Question Source Comments: l Material Required for Exami::stion:          .


l Subsequently PT-LM-100D fails upscal Which of the following lists all expected actions? CIA and SI b. CIA, SI and MSLI l CIB and MSLI d. CIA, CIB, SI and MSLI A's: lb l Exam Level: lS  l Cognitive Level: ! Comprehension l Explanatio o tf Answer KA: l 013 A2.06 l RO Value: l 3.7* l SRO Value: l4.0 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution Engineered Safety Features Actuation System Title:
          ,
KA Ability to (a) predict the impacts of the following on the Engineered Safety Features Actuation System and (b)
I I
Statement: based on those predictions, use procedures to correct, control, or mitigate the consequences of those abncrmal operation:
i
Inadvertent ESFAS actuation Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Instrument Failure Procedure 10M-1.4.lF  l Iss 4 Rev i
:
Reactor Protection Trip LP-SQ- Logics Q estion Source l Facility Exam Bank  l Question Modification Method l Q:estion Source Comments: l M;terial Required for Examination:
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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.


Questio2 Tepic: l Operation following Si sign 11 A steam break has occurred causing an SI on high containment 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 SI Reset Pushbuttons are depressed? Neither train of S1 will rese b. Only one train of SI will rese Both trains of SI will reset but one train will immediately reinitiat d. Only one train of S1 will reset. The reset train will immediately reinitiate.
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 l
l


A:s: lc l Exam Level: lS l Cognitive Level: l Application l Esplanatio a cf Answer KA: l 013 A3.02 l RO Value: l4.1 l SRO Value: l l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution Engineered Safety Features Actuation System Title:
KA Ability to monitor automatic operations of the Engineered Safety Features Actuation System including:
Statement:
Operation of actuated equipment Reference  Reference Number Reference Section Page Number (s) Revision    Lear Obj FS AR Logic Diagrams Figure 7.2-1 Sheet 8 Reactor Protection System LP-SQS- VI.E. j l
Qrestion Source l Facility Exam Bank l Question Modification Method j Question Source Comments: l Miterial Required for Figure 7.21 Sheet 8 Examination:            l l
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L-
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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.


I Question Topic: l ROD BOTTOM alarm f- During a reactor startup, when does the ROD BOTTOM / ROD DROP alarm (A4-126) become active for i each control bank?
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
The alarm will actuate for a dropped rod for: any Control Bank whenever Control Bank A RPI output is above 20 step b. each Control Bank whenever that Control Bank demand position is above 35 step 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 abov.: 20 step 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 step Ans: Id l Exam Level: lS l Cognitive level: l Memory l Explanatio o ef Answer KA: l2.4.31 l RO Value: l3.3 l SRO Value: l3.4 l Section: l SYS l RO Group: l 2 l SRO Group: l1 System / Evolution Rod Position Indication System i
            '
Title:
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:
KA Emergency Procedures / Plan Statement:
Knowledge of annunciators alarms and indications, and use of the respoE instruction Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Reactor Control & Protection IO M l. RPI, Ist & 2nd 16 Iss 4 Rev
Knowledge of annunciators alarms and indications, and use of the response instructions.
  - Summary Description   paragraphs   1 RPI and Insertion Limits LP-SQS- VI.B. C 5-6  5 2.b c Reactor Control and lOM-l . Protection Setpoints Question Source l Previous 2 NRC Exams l Question Modification Method l Question Source Comments: l Mrterial Required for Ex mination:
 
i l
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
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Question Source l Previous 2 NRC Exams ~ l Question Modification Method l Question Source Comments: l Mat; rial Required for Examination:           1
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[.
- - __ __- - - _ _ _ _ - _ _ _ _ _ _ - _ _ _ - - _ _ - - _ _ _ - _    _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ - - _ _ _ ___-_ -_  . _ - _ .
l l- Question Tcpic: l determination of NIS counts by IR/SR status Given the following conditions:
Question'Tep6c: l detennin: tion cf NIS counts by IR/SR status Giv;n the follswing conditions:
  * Reactor tripped from 100% power
  * Reactor tripped from 100% power
  * Following transition to ES-0.1 " Reactor Trip Response", Intermediate l Range NIS is reading 1E-7 amps
  * 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?
  * 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?
, minutes.
c. 4 minutes.


,
b. 8 minutes.
b. 8 minute minute minute A;s: la 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 n ef Answer rate). This SUR is used with IR activation setpoint ~ IE-10 gives time of 4.02 minute KA: l 015 K$.06 l RO Value: l3.4 l SRO Value: l3.7 l Section: lSYS l RO Group: l 1 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:
c. 10 minutes.
Subcritical multiplications and NIS indications Reference Reference Number Reference Section Page Number (s) Revision Lear Obj Reactor Excore Inst. System IOM-2. IR 2nd paragraph 9 iss 4 Rev
 
  - Major Components     1 Excore Instrumentation LP-SQS- IV. , 8 System Question Source l New l Question Modification Method l Qrestion Source Comments: l Material Required for Examination:
d. 13 minutes.
l
 
          !
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.
l Page 35
 
        - _ __-___-_-__-_-_-_-__a
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 Question Source Comments:   l Material Required for Examination:
i.
 
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Question Topic: l Leak in RVLIS A leak has occurred at the inlet to a RVLIS differential pressure transmitter.
- _ _ - - _ _  - _ _ _ _ _ _
            ' '
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?
Which of the following describes RVLIS system indication and how the leak will be isolated?
l RVLIS hydraulic isolator position will indicate a leak has occurred. The leak will automatically isolate.
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.


l b. RVLIS hydraulic isolator position will indicate a leak has occurred. The leak can only be isolated l by closing a manual isolation valv RVLIS high volume sensor position will indicate a leak has occurred. The leak will automatically isolat d. RVLIS high volume sensor position will indicate a leak has occurred. The leak can only by isolated by closing a manual isolation valv Ars: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio ccf Answer KA: l 016 K3.01 l RO Value: l3.4* l SRO Value: l3.6* l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Non-Nuclear Instrumentation System Title:
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:
KA   Knowledge of the effect that a loss or malfunction of the Non-Nuclear instrumentation System will have on the Statement:   following:
RCS Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj RVLIS Hydraulic isolator IOM-6.4.AG  IV.A.7, 8  AG2  Iss 4 Rev Malfunction       0 RVLSI & Core Cooling LP-SQS- II.B.e, f; ll.G.c; I ,16-17.,22- 1 6 Monitor     23 Question Source l New l Question Modification Method l Q:estion Source Comments: l M;terial Required for Ermination:
RCS Ref;rence    Reference Number Reference Section Page Number (s) Revision Learn.
i i
 
l l
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:
l Page 36
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.


i-l Question Topic: l Ev I of Natural Circulation for conditions
'
'
Given the following conditions:
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  ,
,
 
* A loss of offsite power occurred l * A natural circulation cooldown was initiated l
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:
a The five hottest T/Cs average temperature - 555 F
f Page 37
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          ~
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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 E
 
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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.
 
What would be the status of the Quench Spray (QS) system?
j
 
(Assume no operator action has been performed in the Quench Spray system.)
 
'
'
* RCS wide range pressure -1275 psig
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.
* All RCS Loop Tw - 552 F
* All RCS Loop T,oi - 544 F
* All SG pressures - 940 psig          j l
Adequate natural circulation flow: (Refer to Att. 6A & 2G)
            ' exists and the RCS is subcoole . b. does not exist and the RCS is subcoole exists and the RCS is at saturation.


b. BOTH QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle
,
,
d.' does not exist and the RCS is at saturation.
  . 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:            !
,
,
Ans: lb l Exam Level: lS l Cognitive Level: l Application l Explanatio o of Answer KA: l 017 A3.01 l RO Value: l 3.6* l SRO Value: l 3.8' l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution in-Core Temperature Monitor System Title:
! KA             1 Knowledge of the physical connections and/or cause-effect relationships between Containment Spray System and  j l- Statement:   the following:
KA Ability to monitor automatic operations of the in-Core Temperature Monitor System including:
            !
Statement:
!    ECCS Reference     Reference Number Reference Section   Page Number (s) Revision Learn.
Indications of normal, natural, and interrupted circulation of RCS Reference Reference Number Reference Section Page Number (s) Revision Lear Obj 0 F Plus Subcooling Based 10M-53A.I.6-A  I  issIB on Core Exit TCs        Rev 2 Natural Circulation  EOP Attachment 2-G I  2  issIB Verification        Rev 2 EOP Generic issues  LP-SOS-5 VII Question Source l NRC Exam Bank  l Question Modification Method l Question Source Comments: l Material Required for Steam tables, EOP att. 2-G and 6A Examination:
Page 37
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Question Topic: l Power supply following CIB The Containment Air Recirculation fans are in NSA prior to a transient which causes CI After CIB occurs, what will be the status of the Containment Air Recirculation fans? Running in fast speed
Obj
!
,
b. Running in slow speed l
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
! c. Tripped but the power supply is energized l
Recirculation Rev 4 l CNMT Depressurization    LP-SQS-13.1 V.D.1    17-18 5.b
l d. Tripped with the power supply deenergized Ans: ld l Exam Level: lS l Cognitive level: l Comprehension l Explanatio c cf Answer KA: l 022 K2.01 l RO Value: l 3.0* l SRO Value: l l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Containment Cooling System Title:
! System
KA Knowledge of electrical power supplies to the following:
,
Statement:
Question Source l New     l Question Modification Method l l Question Source Comments:   l Materi;I Required for Esimitation:
Containment cooling fans Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj CNMT Vent - Summary IOM-44C. CNMT Air 1 Iss 4 Rev Description   Recirculation 0 Containment Ventilation LP-SQS-44 !!. ,7 Systems Question Source l New l Question Modification Method l Q'estion Source Comments: l M;terial Required for Ex mination:
            ,
Page 38
i Page 39


Question T ple: l Quench Spray rIsponse to RWST lev 1 Given the following conditions:
_ - - - _ - _ _ _ _ _ _ - . _ _ _ - _. _..  - _ _ _ . -
* Reactor trip, Si and CIB occurred from 100% power due to a LOCA
* RWST levei har decreased to 3 feet 9 inches
* CIB has not been rese What would be the status of the Quench Spray (QS) system?
(Assume no operator action has been performed in the Quench Spray system.) BOTil QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle Bypass Isol Vivs closed, and TWO QS Chemical Injection pumps are runnin b. BOTH QS pumps are running with [MOV-lQS-103 A,103B] QSPP Cavitating Venturi Nozzle Bypass Isol Vivs closed, and FOUR QS Chemical Injection pumps are runnin 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 runnin 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 runnin Ans: la l Exam Level: lS  l Cognitive level: I Comprehension l Explanatio ocf Answer        l KA: l 026 Kl.01  l RO Value: l4.2 l SRO Value: l4.2 l Section: lSYS l RO Group: l 2 l SRO Group: l1 System / Evolution  Containment Spray System Title:
KA  Knowledge of the physical connections and/or cause-effect relationships between Containment Spray System and Statement: the following:
ECCS Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Loss of reactor Or Secondary  E-l  step 30 22 Iss IB Coolant        Rev 4 Transfer to Cold Leg  ES step 6  6  IssiB Recirculation      Rev 4 ,
CNMT Depressurization  LP SQS-1 V.D. I  17-18 System Question Source l New  l Question Modification Method l  l Question Source Comments:  l Material Required for Examination:
l l
l Page 39 l
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Question Topic: l Recombiner Ops
      - - - - _ _ _ _ _ _ _ -_ ____ _
        '
          * *
Given the following conditions:
Cries Topict l Recombiner Ops Given the f;llowing conditions:
l * A LOCA has occurred 24 hours ago
* A LOCA has occurred 24 hours ago
  * ONE Hydrogen recombiner is placed in service when hydrogen concentration reaches 0.5%
  * ONE Hydrogen recombiner is placed in service when hydrogen concentration reaches 0.5%
With a recombiner in operation, containment pressure: should be maintained at approximately 8.9 PSIA, to prevent excessive recombiner flo will be adequate for recombiner operation ifit is maintained between 8.9 PSIA and -3 PSIG should be maintained slightly above atmospheric, to ensure sufficient recombiner flo should be maintained at approximately -2PSIG, to ensure sufficient recombiner flo Ans: Ic l Exam level: IS l Cognitive Level: l Application l Explanation   ~
With a recombiner in operation, containment pressure:
of Answer KA: l 028 A1.01 l RO Value: l3.4 l SRO Value: l l Section: lSYS l RO Group: l 3 l SRO Group: l2 System / Evolution Hydrogen Recombiner and Purge Control System Title:
a. should be maintained at approximately 8.9 PSIA, to prevent excessive recombiner flow.
KA ANiity to predict and/or monitor changes in parameters associated with operating the Hydrogen Recombiner and Statement: Purge Control System controls including:
 
Hydrogen concentration Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Post DBA Hydrogen Control 10M-46. th paragraph 1  Iss 44:
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.
System - Summary     Re Description Post DBA H2 Control LP-SQS-4 .C. ,9 System System Question Source l New l Question Modification Method l Question Source Comments: l M:.terial Required for OM 46. Examination:
 
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
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.
.
 
Question Topic: l Ev:luation cf a leak Given the following conditions:
d the top of the fuel.
* Reactor power is 85%
 
* Spent Fuel Pool is aligned for cooling
Ans: lc  l Exam Level: lS l Cognitive Level: l Memory l Explanatio       ,
* - 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 of Answ:r        *
a. ~25 feet above the top of the fue b. ~23 feet above the top of the fue c. ~10 feet above the top of the fue d. the top of the fue Ans: Ic    l Eram Level: lS     l Cognitive Level: l Memory l Explanatio n of Answer KA: l 033 A2.03     l RO Value: l3.1 l SRO Value: l3.5 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution       Spent Fuel Pool Cooling System Title:
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:
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 Lear Obj Fuel Pool Cooling and       lOM-20. ss 4 Rev Purification             3 Fuel Pool Cooling and       LP-SQS-2 ,9b Purification
Abnormal spent fuel pool water level or loss of water level Reference   Reference Number Reference Section Page Number (s) Revision Learn.
              '
 
Question Source l New         l Question Modification Method l Question Source Comments:        l Material Required for Examination:
*
i
          [
!'
Obj t .
Page 4i L
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
          !
i l
I i
Page 41 i
          )
 
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.


Question Topic: l Transfer Cart Operation Which of following describes the interlock between the conveyor car drive and the upende s when l transferring the conveyor car from the transfer canal to the refueling cavity?
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:
l Both upenders must be in the down position before the conveyor car can be moved.
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.


t      .
Obj
i  b. Only the upender in the refueling cavity must be in the down position before the conveyor car can be move Only the upender in the transfer canal must be in the down position before the conveyor car can be move 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 upende A ns: la  l Exam Level: lS  l Cognitive Level: l Memory l Esplanatio s ef Answer KA: l 034 K4.02  l RO Value: l2.5 l SRO Value: l3.3 l Section: l SYS l RO Group: l 3 l SRO Group: l2 System / Evolution  Fuelliandling Equipment System Title:
~ 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
KA  Knowledge of Fuel Handling Equipment System design feature (s) and or interlock (s) which provide for the Statement:  following:
Fuel movement Reference    Reference Number Reference Section Page Number (s) Revision Lear Obj Fuel Handling Operations   LP-SQS-6.13 XI.11. .a 1 RP-12R- II. iss 0 Rev


Question Source l NRC Exam Bank   l Question Modification Method l Question Source Cominents:   l Material Required for Ex:mination:
Question Source l NRC Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Ermination:
Page 42
Page 42
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_______________
{; question Top 6es l SG level program Reactor power is 25% and all plant systems are in NSA.
 
. . . . - - _ . . -
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.


! Which failure would decrease feedwater flow to all SGs? ONE condenser steam dump fails ope b. Heater Drain receiver Level Control Valve [LCV-ISD-106B] fails open.
d. Combined Feedwater Header Pressure channel (PS-lFW-151] fails high.


I l c. Turbine First Stage Pressure channel [PT-lMS-446] fails low.
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


l l d. - Combined Feedwater Header Pressure channel [PS-1FW-151] fails hig Ans: lc l Exam Level: lS l Cognitive Level: l Comprehension l l Explanatio c of Answer KA: l 035 Kl.01 l RO Value: l4.2 l SRO Value: l4.5 l Section: lSYS l RO Group: l 2 l SRO Group:,l2 System / Evolution Steam Generator System Title:
r l
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 i Reference Section Page Number (s) Revision Lear Obj SG Feedwater System - IOM-24.lD  SGWLC  7-8 iss 4 Rev Instrumentation and Controls    2 SG Feedwater System - 1OM-24.4.lF  Attachment 5, II. IF 38 iss 4 Rcv
'ustrument Failure      2 Feedwater System  LP-SOS-2 Ill.E.1 .A
' Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for Examination:
i l
Page 43
Page 43
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Questio1 Topic: l Effect cf MS- PT-464 failing high Given the following conditions:
        * *
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
  * The unit is in MODE 3 preparing for normal plant cooldown
  * Condenser Steam Dump System is automatically controlling T,y at 547 F in Steam Pressure Mode
  * Condenser Steam Dump System is automatically controlling T, at 547 *F in Steam Pressure Mode
  * [PT-1 MS-464] Main Steam Header Pressure fails high Which one of the following describes the effect this will have on the Condenser Steam Dump system? Two banks of steam dumps will open and remain open until manually close b. Two banks of steam dumps will open but should reclose with no operator actio All banks of steam dumps will open and remain open until manually close d. All banks of steam dumps will open but should reclose with no operator actio Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio o e f Answer KA: l 041 K6.03 l RO Value: l2.7 l SRO Value: l2.9 l Section: l SYS l RO Group: l 3 l SRO Group: l3 System / Evolution Steam Dump System and Turbine Bypass Control Title:
  * [PT-1MS-464] Main Steam Header Pressure fails high l
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:     )
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.
Controller and positioners, including ICS, S/G, CRDS Ref;rence Reference Number Reference Section Page Number (s) Revision Lear Obj M:in Steam System IOM-21.5.A.24  I Iss 4 Rev
 
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
 
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:
 
Page 44 i
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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.


Main Steam System  LP SQS-2 Question Source lNew  l Question Modification Method l Qrestion Source Comments: l Material Required for        j Examination:        l l
I b. P-4 would have to be cleared and the FWI FW BYPASS VALVE RESET pushbuttons would have to be depressed.
l Page 44 L.__


Quesison Topic: l NPS11 for FW Given the following conditions:
c. SI would have to be reset and the FWI FW BYPASS VALVE RESET pushbuttons would have to be depressed.
'
* Reactor power - 100%
* A load rejection occua and the plant stabilizes at 45% power
'* ' Load rejection bistables " LOAD REJ 15-50%" and " LOAD REJ GREATER THAN 50%"
are lit How are the Steam Generator Feed Pumps [1FW-P-1 A,1B] protected from a loss of suction pressure during the load rejection?          j l
a. The Feedwater Heater Bypass Valve [TV-1CN-100] opened and closed FOUR minutes late b. The Heater Drain Receiver Level Control Valve [LCV-ISD-106B] was maintained fully open until LOW-LOW level was sensed in the Heater Drain Receiver, c. The Heater Bypass to Heater Drain Pump Suction Valve [TV-CN-125] opened and closed four minutes late d. The Condensate Pumps Recirculation Valve [FCV-lCN-101] closed on the 15-50% load rejection and reopened FIVE minutes late Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio e cf Answer KA: l 056 Al.08 l RO Value: l2.3 l SRO Value: l 2.6* l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Condensate System Title:
KA Ability to predict and/or monitor changes in parameters associated with operating the Condensate System controls Statement: including:
MFW pump suction pressure Reference  Reference Number Reference Section Page Number (s)  Revision Lear l
            !
Obj Load Rejection  AOP 1.3 step I1  7    Iss 3A Rev6 Figure 22-6 - Step Load 1OM 22.5. Iss 4 Rev  i Rejection Ckt        0 Extraction Steam and lleater LP-SQS 23  Ill. .E Drains Question Source l Other Facility  [ Question Modification Method l Question Source Comments: l        l M;terial Required for Examination:
'
l Page 45
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      -
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.
    ,  .
        ' '
        '
  's    , - ,
l Question Topie: l Restoration of FW capability An inadvertent Si signal occurred at 100% power. The condition causing the SI signal is no longer present.


j All systems function as designed and RCS conditions stabilize as expected following the inadvertent S Which of the following states the condition (s) that would have to be met to feed via IFCV-lFW-479(489)(499)], SG F d.4ypass FCVs? Only the FWI FW BYPASS VALVE RESET pushbuttons would have to be depresse P-4 would have to be cleared and the FWI FW BYPASS VALVE RESET pushbuttons would have to be depresse SI would have o be reset and the FW1 FW BYPASS VALVE RESET pushbuttons would have to be depresse S' would have to be reset, P-4 would have to be cleared and the FWI FW BYPASS VALVE RESET pushbuttons would have to be depresse Ans: Ia l Exam Level: lS l Cognitt/c Level: l Application l-Esplanatio u of Answer KA: l 059 A4.1l l RO Value: l3.1 l SRO Value: l3.3 l Section: lSYS l RO Group: l 1 l SRO Group: ll System / Evolution Main Feedwater System Title:
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:
KA Ability to manually operate and/or monitor in the control room:
Statement:
Statement:
Recovery from automatic feedwater isolation Reference Reference Numtwr Reference Section Page Number (s) Revision Lear Obj Feedwater System LP-SQS-2 lil.E. I .j, 7.A.(12)
Recovery from automatic feedwater isolation R:f rence Reference Number Reference Section Page Number (s) Revision  Learn.
Reactor Protection Systems LP-SQS- VI. Updated FSAR    Figure 7.2-1 sheet 1&l3 Question Source l Facility Exam llank l Question Modification Method l Question Source Comments: l Material Required for Figure 7.2-1 sheet 1 & 13 Examination:
Page 46


      .,
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 e  - -
l     Page 46 N--__-________.
        .
 
Qrestion Topic: l SGWLC inputs Given the following conditions:
. _ _ _ . _ _ _ _ _ _ _ _ _ ._ . _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _
  * Reactor power is 20%
l . .          i Questio2 Tcpic: l SGWLC inputs Given the following conditions:
  * Feedwater has been transferred to the Main Feed Regulating Valves
L
* All systems are NSA
  * . Reactor power is 20%
* Narrow Range SG 1C levelis 44%
          '
* [FCV-lFW-499] 1C SG FW Bypass Viv is manually opened 15%
  ( Feedwater has been transferred to the Main Feed Regulating Valves j * All systems are NSA l + f Narrow Range SG IC levelis 44%
Aftet plant conditions stabilize, which parameter (s) will be different from those prior to [FCV-1FW-499]
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?
opening?
l Only [FCV-lFW-498] IC Main FW Reg Viv position [FCV-lFW-498] IC Main FW Reg Viv position and Narrow Range SG IC Level Only Narrow Range SG IC Level d. Narrow Range SG 1C level and Stm Gen 1C Feed Flow indication A:s: l'n l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio n 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 l System / Evolution Main feedwater System l Title:
l c. Only [FCV-IFW-498] IC Main FW Reg Viv position
KA Knowledge of the physical connections and/or cause-efTect relationships between Main Feedwater System and the St:tement: following:
.
S/GS water level control system Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj SG Feedwater System - lOM 24. SGWLC 7-8 Iss 4 Rev Instrumentation and Controls     2 Feedwater System  LP-SQS-2 .E.1 .A
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 l Qrestion Source Comments: l i
Question Source l New  l Question Modification Method j Question Source Comments: l Material Required for Esanlaation:
M:terial Required for
          )
'
!    Page 47 L_____-_________.        .i
Examination:
.
Page 47
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l f
_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ - _ _
t Question Topic: l Relationship of AFW steam supply & feed suppli:s to SG Given the following conditions:
            - .
Question Tcpic: l R,1 tionship of AFW stram supply & feed supplies to SG Given the following conditions:
  * Reactor power - 100%
  * 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:


* A loss of all AC power occurs o Auxiliar" Feed Pump IFW-P-2 starts and runs
* The stean supply line from SG B to IFW-P-2 ruptures at the connection to the main steam lin * 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? All SGs will blowdown through the mpture, and NO auxiliary feed will be availabl SG A and SG B will blowdown through the rupture, but NO auxiliary feed will be availabl 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 . Only SG B will blowdown through the rupture, and auxiliary feed can be established from SG A"s: ld l Exam Level: lS l Cognitive Level: l Memory  l f.xplanatio c ef Answer KA: l 061 K3.02 l RO Value: l4.2 l SRO Value: l4.4 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution Auxiliary / Emergency Feedwater System Title:
KA Knowledge of the efTect that a loss or malfunction of the Auxiliary / Emergency Feedwater System will have on I Statement: the following:
S/G Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj SG Feedwater System  IOM-24. Auxiliary Feed Pumps 2-3 iss 4; Rev 2 Feedwater System  LP-SQS-2 til. .13 SG Feedwater System  LP-SQS-2 Ill.L. Q :estion Source lNew  l Question Modification Method l Question Source Comments: l M;terial Required for Enmination:
Page 48 l
Page 48 l
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)
Questici Tepic: l Overcurrent effect on breaker operation The Unit is at 85%. Which of the following conditions will result in bus I AE being maintained deenergize [ACB-1 A10] 1 AE Emergency Bus feeder breaker trips on overcurren b.- I AE Emergency Bus reverse phase PT blows a fus [ACB-41C) 1 A Normal 4KV Bus Feeder Breaker trips on overcurren d. [ACB-41C) 1 A Normal 4KV Bus Feeder Breaker trips on Unit Statioa Service Tranformer 1C Differential Tri )
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.
Ars: la l Exam Level: lS l Cognitive Level: l Comprehension     l Explanatio ucf Answer KA: 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      i System / Evolution A.C. Electrical Distribution Title:
 
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-
 
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:
KA Knowledge of A.C. Electrical Distribution design feature (s) and or interlock (s) which provide for the following:
St:tement:
Statement:
Bus lockouts Reference Reference Number Reference Section     Page Number (s) Revision Lear Obj 4160V Emergency Bus I AE lOM-36.4.ACZ        iss 3 Rev ACB 1 A10 Auto Trip          i Diesel Generators LP-SQS-3 Qrestion Source l Previous 2 NRC Exams l Question Modification Method l Question Source Comments: l
Bus lockouts Ref;rence Reference Number Reference Section   Page Number (s) Revision Learn.
                -
 
Material Required for Examination:
Obj 4KV Station Service System IO M 36.1.E    20-21 iss 4 Rev
Page 49
- 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      ,
Question Tcpic: l Br:aker interlock (s)
Reactor power is 25% during a startup. Electrical loads have been transferred to the Unit Station Service Transformer (USST).


In order for Bus I A to be setup for Auto Bus Transfer to the System Station Service Transformer, which of the following lists the required position of the Live Bus Transfer switch and the control switch for ACB 41A7 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 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 ATs: la    l Exam Level: lS    l Cognitive Level: l Memory l Explanatio a cf Answer KA: l 062 K4.01      l HO 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 Title:
125 V DC Control System IOM-39.1  3
KA      Knowledge of A.C. Electrical Distribution design feature (s) and or interlock (s) which provide for the following:
        {
St:tement:
iss 4 Rev '
Bus lockouts R:,fer ace        Reference Number Reference Section Page Number (s) Revision Lear Obj MV Station Service System        IOM-36. iss 4 Rev
- Specific Instrumentation            I and Controls 4KV Distribution        LP-SQS 3 .
            ,
QIestion Source l New        l Question Modification Method l Qrestion Source Comments:        l M:terial Required for i Enmination:
Page 50 l
l              --  -_-


Questio2 Topic: l Response to ground indication l
125 VDC LP-SOS-39.1 I
l DC Bus 1-2 ground voltmeter went from 0 volts to -105 volts. The DC Bus is in NSA for 100% power operation Which of the following describes the effect the ground will have on DC bus operations? The ground has caused actual voltage to the DC loads to decrease to 105 Volt b. The affected battery will discharge significantly faster than designe c. The bus will operate as required but the bus reliability has decrease d. Another ground on the same polarity of the bus will cause a short circui Ars: lc  l Exam Level: lS l Cognitive Level: l Memory l Explanatio o cf Answer KA: l 063 A2.01  l RO Value: l2.5 l SRO Value: l 3.2* l Section: lSYS l RO Group: l 2 l SRO Group: l1 System / Evolution  D.C. Electrical Distribution Title:
Q 'estion Source l New l Question Modification Method l Question Source Comments: l i
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:
M:terial Required for        l Ex:mination:       {
Grounds
i.
            ~
Reference    Reference Number Reference Section Page Number (s) Revision Lear j Obj 125 V DC Control System-  IOM 3 A.16  2  Iss 3 Rev Precautions & Setpoints        0 125 V DC Control System  IOM-3 iss 4 Rev


125 VDC    LP-SQS-3 Q:estion Source l New    l Question Modification Method l Question Source Comments:  l Material Required for Examination:
l l
            !
Page5I L.____._____ _
i
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l i
Page 51
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Question Topic: l Reverse power trip of DG Diesel Generator No.1 is paralleled to 4160V Bus I AE for testing. The operator is in the process of adjusting load and voltage wb: the Governor Control switch sticks in the LOWER position.
              . .
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?
If NO operator action is taken, what will be the Diesel Generator response to this condition?
DG frequency will: decrease and the diesel will trip on reverse powe b. decrease and the diesel will trip on overcurren c. remain constant but the diesel will trip on reverse powe d. remain constant but the diesel will trip on overcurrent.
DO frequency will:
c. decrease and the diesel will trip on reverse power.


A*s: lc l Exam Level: lS l Cognitive Level: l Comprehension l     ]
^
Explanatio c ef Answer KA: l 064 Al.08 l RO Value: l l SRO Value: l3.4 l Section: lSYS l RO Group: l 2 l SRO Group: l2
b. decrease and the diesel will trip on overcurrent.
. System / Evolution Emergency DieselGenerators Title:
 
KA Ability to predict and/or monitor changes in parameters associated with operating the Emergency Diesel Setement: Generators controls including:
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)
Maintaining minimum load on ED/G (to prevent reverse power)
Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Transferring Emergency IOM-36. IV.A.9 & CAUTION Q2  Iss 4 Rev Feed      3 Transferring Emergency Busses I AE And IDF From Emergency Feed To Normal Feed Alarm DIESEL  IOM-34.ADU  A8-127  ADUl  iss 3 Rev GENERATOR NO. I      1 REVERSE POWER Diesel Generators  LP-SQS-3 V Q'estion Source l New  l Question Modification Method l Question Source Comments: l Material Required for Examination:
Reference Section Page Number (s) Revision Learn.
Page 52
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Questiol Tepic: l Diesel Generator Trips A loss of off-site power occurred and the diesel generators are supplying the emergency buse Which of the following will trip a diesel generator? The governor control switch in the control room is held in the RAISE positio b. A governor failure causes engine speed to increase to 1050 RP Thejacket cooling water pump trip d. The coupling fails on the lube oil pum Ass: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio a c f Answer l 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 DieselGenerators Title:
Reference      Reference Number Ob}
,
IV.A.9 & CAUTION Q2  iss 4 Rev Transferring Emergency    lOM 36.4.Q
 
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
 
GENERATOR NO. I REVERSE POWER            6 VI.13  29 Diesel Generators      LP-SQS-36.2 Question Source l New      l Question Modification Method l Question Source Comments:     l Material Required for Eur.mination:
Page $2
_ . . -
        . . . _____  _ _ _ _ _ _  ___
 
_
_ _ _ _ _ _ _ _ _ _ - ,
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:
KA Knowledge of Emergency Diesel Generators design feature (s) and or interlock (s) which provide for the following:
Statement:
Stat: ment:
Trips for ED/G while operating (normal or emergency)
Trips for ED/G while operating (normal or emergency)
Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Local- Overspeed Trip IOM 36.4.AFN  1  Iss 3 Rev
Reference  Reference Number Reference Section Page Number (s) Revision Learn.
 
Obj Local. Overspeed Trip IOM-36.4.AFN  1  1ss 3 Rev


Diesel Generators LP-SQS-3 Technical Specifications  4.8.1.1.2. /4 8.4a Q;estion Source l Facility Exam Bank l Question Modification Method l Qrestion Source Comments: l M;terial Required for Examination:
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 Q:estion Source Comments: l M:t: rial Required for Ex:mination:
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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.


Question Topic: l Drain Tank Isol: tion Given the following conditions:
d. The Low Level Waste Drain pump tripping on low flow rate.
,
* Low Level Waste Drain Tank level is 110 inches I' * The discharge permit has been approved at discharge rate of 15 gpm a The discharge is in progress at 15 gpm What condition will automatically stop the release? Both [TV-LW-105] Liquid Waste Effluent Trip valve and [FCV-LW-104-2] High Range Liquki Waste Efiluent Flow Control Valve closing on high-high radiation signal from [RM-LW-104). [FCV-LW-104-2] High Range Liquid Waste Effluent Flow Control Valve closing on low flow rate, [FCV-LW-104-1] Low Range Liquid Waste Effluent Flow Control Valve closing on low Waste Drain Tank leve d. The Low Level Waste Drain pump tripping on low flow rat Ans: la l Esam 12 vel: lS l Cognitive Level: l Memory l Explanatio o c.f Answer KA: l 068 A4.04 l RO Value: l3.8 l SRO Value: l3.7 l Section: lSYS j RO Group: l 1 l SRO Group: l1 System / Evolution Liquid Radwaste System Title:
KA Naility to manually operate and/or monitor in the control room:
Statement:
Aut(,matic isolation Reference  Reference Number Reference Section Page Number ('.) Revision Lear Obj Liquid Waste Disposal LP-SQS-17,1 II.C.7. 8 & 10 11-13 3 System      _
Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for Examination:
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f Questici Tepic: l Annunciator Operation l Due to a Steam Generator Tube Leak a Condenser Air Ejector Vent Monitor [RM-1SV-100] High alarm
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:
'
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? Condenser Air Ejector Vent Monitor [RM-1SV-100] rising to the High-High alarm setpoin b. Sto.'m Generator Blowdown Samnle Monitor [RM-ISS-100] rising to the High alarm Setpoin Steam Generator N-16 Monitor [RM-1MS-102] rising to the High alarm Setpoin High Capacity Steam Generator Blowdown Monitor [RM-1BD-101] rising to the High alarm Setpoint.-
Ais: lb l Exam Level: lS l Cognitive Level: l Memory l Esplanatio a of Answer 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 Systert/ Evolution Process Radiation Monitoring System Title:
KA Ability to manually operate and/or monitor in the control room:
KA Ability to manually operate and/or monitor in the control room:
Stuement:
Stat: ment:
Radiation monitoring system control panel R:ference Reference Number Reference Section Page Number (s) Revision Lear Obj Rui Monitoring System - I OM-43. Iss 4 Rev Instrumentation and Controls    3 Radiation Monitoring System LP-SQS-4 Q:estit,n Source l New  l Question Modification Method l QIestion Source Comments: l Mrterial Required for Examination:
Automatic isolation Reference Reference Number Reference Section Page Number (s) Revision Learn.
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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:
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Qrestion Topic: l Evaluation of available tir sources A leak has occurred in the Station Air System in the Fuel Building. [PI-ISA-101] Station Air Main IIeader and [PI-ll A-106] Station Instrument Air 11eader pressure indications are both lowerin When Station Air pressure decreases to a specific setpoint, (TV-ISA-105] Station Air Header Trip Valve will: open to supply instrument air load b. open to supply contailunent air load close to ensure all station air will be supplied to the instrument air loads.
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.


I d. close to maintain air to all station load Ans: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio a of Answer KA: l 078 K4.02 l RO Value: l3.2 l SRO Value: l3.5 l Section: lSYS l RO Group: l 3 l SRO Group: l3 System / Evolution 3 instnament Air System Title:  l KA Knowledge of instrument Air System design feature (s) and or interlock (s) which provide for the following:
b. Steam Generator Blowdown Sample Monitor [RM-ISS-100] rising to the High alarm Setpoint.
St:tement:
Cross-over to other air systems Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Compressed Air Systems - IOM-34. Station Air Header Trip 5  iss 4 Rev instrumentation.: and Controls D  Valve  0 VOND  34-1 Compressed Air  LP-SQS-3 IV.A & D  15  5 Qrestion Source l New  l Question Modification Method l Qrestion Source Comments: l M terial Required for Ex:mination:
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Question Topic: l Containm nt Building Penetrations during refurling Which of the following is NOT part of the Technical Specification. definition of CONTAINMENT INTEGRITY a. .The containment leakage monitoring system is OPERABL b. All equipment hatches are closed and seale The sealing mechanism associated with each penetration is OPERABL d. The containment leakage rates are within their LCO limit '
c. Steam Generator N-16 Monitor [RM-1MS-102] rising to the High alarm Setpoint.
Ans: la l Esam Level: lS  l Cognitive Level: l Comprehension l Esplanatio e Cf Ahswer KA* l 103 Kl.02  l RO Value: l3.9 l SRO Value: l 4.l* l Section: ISYS l HO Group: l 3 l SRO Group: l2 System / Evolution  Containment System Title:        _
KA  Knowledge of the physical connections and/or cause-efTect relationships between Containment System and the Statement: following:
Containment isolation / containment integrity Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Technical Specification  3/4. /4 9-4 Containment System  LP-SQS-4 V .h Question Source l New  l Question Modification Method l QIestion Source Comments:  l Material Required for Examination:
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d. High Capacity Steam Generator Blowdown Monitor [RM-1BD-101] rising to the High alarm Setpoint.
l Question Topic: l Determination cf pow;r increae Given the following conditions:
  * EOL
  * Reactor power is 80% steady state
  * RCS T.,, i; on program
  * Control Rod position - 160 steps on Control Bank D e Control Rods begin to withdraw e When Control Bank D is at 170 steps the Control Rod Bank Sel Sw is placed in MANUAL stopping rod motion If N0 further operator action is taken, what would be the affect on actual power level and RCS T,,, after conditions stabilize? 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,,, would remain approximately 571 Reactor power would remain approximately 80% and RCS T,,, would rise by an amount equivalent to the reactivity additio d. Neither reactor power nor RCS T.,, would be significantly affecte Ans: lc l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio Reactivity addition by rod movement would add power to RCS. Since turbine load controls power level on NIS, acf Answer RCS would heat up. By using Power defect curves could determine the equivalent power level the reactivity would allow and the associated Tavg at that power 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 concern)
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: l 001 AKl.03 l RO Value: l3.9 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l1 System / Evolution Continuous Rod Withdrawal Title:
KA Ability to manually operate and/or monitor in the control room:
KA Knowledge of the operational implications of the following concepts as they apply to Continuous Rod St:tement: Withdrawal:
Statement:
Relationship of reactivity and reactor power to rod movement Reference Reference Number Reference Section Page Number (s) Revision Lear Obj Full Length Rod Control LP-SQS Question Source l New l Quesilon Modification Method l Question Source Comments: l htterial Required for Examination:
Radiation monitoring system control panel Ref;reIce - Reference Number Reference Section Page Number (s) Revision  Learn.
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Question Topic: l Operation cf Disconnect Switch Given the following conditions:
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:
  * Reactor power - 5%
  * Control rod F-6 in Control Bank D has fully droppe * Recovery of the dropped rod is in progress per AOP 1.1.5 " Dropped RCCA"
   * All Disconnect Switches in Control Bank D are in DISCONNECT except for F-6 l  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 b. An urgent failure will be received, however rod recovery can proceed with the Control Rod Bank Sel Sw in Manua A non-urgent failure will be received which will not affect control rod movemen d. An urgent failure will be received, however rod recoverj can proceed after depressing the Rod Control Alarm Reset pushbutto Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio      ]
o c.f Answer KA: l 003 AK2.05 l RO Value: l2.5 l SRO Value: l2.8 l Section: l EPE l 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:
Statemer.t:
Control rod drive power supplies and logic circuits Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Dropped RCCA  A O P 1. iss 3A Rev 7 1
        '
Alarm - ROD CONTROL IOM 1.4.AAR A4105 Corrective AARI Iss3 Rev SYSTEM URGENT    Action NOTE  2 FAILURE Full Length Rod Control LP-SQS !!.O.3 & IV. & 16  10;16 l
Question Source l New  l Question Modification Method l Question Source Comments: l Material 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 Q1estion Source Comments:  l 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.
Quest 6on Topic: l Operational limits & basis with given stuck rod Given the following condition,s:


  * Reactor power - 85%
d. The containment leakage rates are within their LCO limits.
  * Load increase is in progress
  -
a Control Bank D is 2 steps above the RIL e Control rod K-6 indicates 15 steps below the remaining rods in Control Bank D
  + Control rod trippability is confirmed
  * Shutdown Margin is verified to be satisfied If the NSS decides to continue power operation with the control rod misaligned, which of the following describes required power reduction and the associated reason?
Reactor power must be reduced to at least: % power within ONE hour to remain in compliance with Rod Insertion Limit restriction b. 75% power within ONE hour to provide assurance of fuel rod integrity during continued operation % power within FOUR hours to remain in compliance with Rod Insertion Limit restriction I % power within FOUR hours to provide assurance of fuel rod integrity during continued operation Ans: lb    l Exam Level: lS  l Cognitive Level: l Application l Explanatio e cf Answer KA: l 005 AKl.% l RG Value: l2.9 l SRO Value: l l Section: l EPE l RO Group: l 1 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 Lear Obj Beaver Valley - Unit 1      3.1.3.1 (ACTION C.3) 3/4 1-18-19 Amend Technical Specifications        No.154 Be;.ver Valley - Unit i      Bases 3/4. B 3/4 l-4 Amend Technical Specifications        No.141 Full Length Rod Control    LP-SQS . Question Source l NRC Exam Bank    l Question Modification Method l Question Source Comments:    l Maternal Required for    Technical Specifications Examination:


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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i 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.


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Ref;rence      Reference Number  Reference Section Ob]
Question Topic: l Steam Dump AiTects
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:
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Given the following conditions:
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l l .* , Reactor tripped from 100% power l * Reactor trip breaker (RTB), which provides P-4 input to Reactor Trip Controller, CANNOT be opened after the trip
 
  * Reactor trip breaker (RTA) opened l Which of the following identifies where the RCS temperature should stabilize prior to placing the Steam l  Pressure Mode Selector Switch in Steam Pressure Mode?
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a. 543 b. 547 d. 554 Ans: le l Exam Level: lS    l Cognitive Level: l Comprehension l Esplanatio c of Answer KA: l 007 EA2.03 l RO Value: l4.2 l SRO Value: l l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution  Reactor Trip Title:              !
Question Ttpic: l Operation cf Disconnect Switch Given the following conditions:
                )
  * Reactor power - 5%
KA  Ability to determine and interpret the following as they apply to Reactor Tn Statement:
  * Control rod F-6 in Control Bank D has fully dropped.
Reactor trip breaker position Reference      Reference Number Reference Section Page Number (s)  Revision Lear Obj M in Steam Systems    IOM 21.5.A.24  1    iss 4 Rev
 
* 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.


M:in Steam System -    IOM-21. various  3-6    iss 4 Rev instrumentation and Controls            1 Main Steam Supply / Steam    LP-SQS-2 Ill.D, Ill.E, V.C.5, 12-14,27 28,    i .e, Dump System      V. Question Source l New      l Question Modification Method l      j Question Source Comments:    l        1 M terialRequired for Examination:
c. A non-urgent failure will be received which will not affect control rod movement.
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d. An urgent failure will be received, however rod recovery can proceed after depressing the Rod Control Alarm Reset pushbutton.
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Questio1 Tr pic: l Operation of controt rods during an ATWS A manual reactor trip was inititted at 100%, however the reactor will not trip. Step 1 of FR-S.1 is being performed. Control rods are in AUTOMATI With the turbine tripped, which of the following describes required action concerning control rod insertion?
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:
Control rods should be inserted in:
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 Question Source Comments:    l Mat: rial Required for Ex:mination:     ,
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' MANUAL even if they are inserting in AUTOMATI b. AUTOMATIC provided rods are inserting in AUTOMATI AUTOMATIC until reactor power is less than 15% where the rods will stop, requiring MANUAL insertio d. AUTOMATIC until the Rod Insertion Limit is reached where the rods will stop, requiring MANUAL insertio Ais: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio acf Answer 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 System / Evolution Reactor Trip Title:
              .
KA Knowledge of the reasons for the following responses as they apply to Reactor Trip:     1 Setement:           )
 
Actions contained in EOP for reactor trip       I R ference Reference Number Reference Section Page Number (s) Revision   Lear Obj Response To Nuclear Power FR- step 1. RNO 2   issIB G;neration- ATWS       Rev 4 Response To Nuclear Power IOM-53.4.FR .1 Knowledp 57 1531B Generation- ATWS       Rev 4 Background EOPs  LP-SQS-5 ,3 Q estion Source l New  l Question Modification Method l QIestion Source Comments: l         j M:.terial Required for Examination:
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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
 
          )
* 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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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
 
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:
I Page 61 l
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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:
,
Page 62


Question Topic: l Eval cf vapor space leak -Tech Spec limit Given the following conditions:
_ _ _ _ _ - . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ -
- * '.'Ihe reactor is operating at 100% power
, Question Tcpic: l Eval c f vapor space leak -Tech Spec limit Given the following conditions:
* ' A 1.2 gpm valve packing leak has occurred on [PCV-RC-455B] PRZR Spray Viv
  *
* The Primary Drains Transfer Tank level is increasing -
The reactor is operating at 100% power
Which of the following describes what type ofleakage this is and based on the leak size what action is .
  * A 1.2 gpm valve packing leak has occurred on [PCV-RC-455B] PRZR Spray Viv a
required per Technical Specifications?
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:
This leak is considered:
a. Primary boundary LEAKAGE that requires Technical Specification entry, b.- Identified LEAKAGE that does not require Technical Specification entr Unidentified LEAKAGE that requires Technical Specification entr d. Unidentified LEAKAGE that does not require Technical Specification entr Ams: lb l Exna level: lS l Cognitive level: l Comprehension l Esplanatio n ef Answer
c. Primary boundary L EAKAGE that requires Technical Specification entry.
        ~
 
KA: l 2.2.22 l RO Valae: l3.4 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Pressurizer Vapor Space Accident Title:
b. Identified LEAKAGE that does not require Technical Specification entry.
KA Equipment Control Statement:
 
Knowledge of limiting conditions for operations and safety limit Reference  Reference Number Reference Section Page Number (s) Revision lear )
c. Unidentified LEAKAGE that requires Technical Specification entry.
Beaver Valley -Unit i   1.14,3.4, l 3,3/4 4-13 Technical Specifications RCS LP SQS- Vl .g Qrestion Source - l New l Question Modincation Method l QIestion Source Comments: l     __
 
M;terial Required for Examination:
d. Unidentified LEAKAGE that does not require Technical Specification entry.
l l
 
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:
l Page 63
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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.


l l
d. Continued use of adverse containment parameters is required due to both the containment pressure and radiation conditions.
Page 63


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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:
Question Tople: l Basis for use cf ADVERSE Cnmt v11uis Given the following conditions:
  + A LOCA has occurred
  + Containment pressure increased to 6.0 psig a : Containment radiation has increased to 1.5E+5 R/h Ninety minutes later containment pressure decreases to 3.0 psig and containment radiation has decreased to 4E+4 R/hr. Integrated CNMT tadiation dose is 2.3E+5 Rad Which of the following describes whether the use of adverse containment parameters can be discontinued? Use of adverse containment parameters can be discontinue b. Continued use of adverse containment parameters is required only due to the containment radiation reading . Continued use of adverse containment parameters is required only due to the contaimnent pressure
  - condition d. Continued use of adverse containment parameters is required due to both the containment pressure and radiation condition Ams: la l Exam Level: lS  l Cognitive Level: l Application   l Esplanatio c af Answer KA: l 009 EK3.16 l RO Value: l3.8 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Small Break LOCA Title:
              ~
KA Knowledge of the reasons for the following responses as they apply to Small Break LOCA:
KA Knowledge of the reasons for the following responses as they apply to Small Break LOCA:
Stat: ment:
Stat: ment:
Containment temperature, pressure, humidity and level limits Ref;rence  Reference Number Reference Section    Page Number (s)   Revision Lear Obj Generic Instrumentation IOM 538.5.Gi-2 I IssIB Rev 2
Containment temperature, pressure, humidity and level limits Page Number (s) Revision Learn.
_EOP Generic issues  LP-SQS-5 X.B.6. 8   22-23   1 15 i
 
Question Source l New   l Question Modification Method l Question Source Comments: l
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   _
;  Material Required for Subcooling Attachment 6-A Ex mination:
Question Source Comments: {
I i
M:terial Required for - Subcooling Attachment 6-A Ex:mination:
I Page 64 -
 
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Question Topic: l Ev:1 e f conditions for tripping RCPs Given the following conditions:
      - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
l * A LOCA has occurred
Question'Tcpic: l Evilcf conditions for tripping RCPs Given the following conditions:
* A LOCA has occurred
  * Containment pressure is 9.2 psig and lowering
  * Containment pressure is 9.2 psig and lowering
  * RCS pressure has stabilized at 325 psig
  * RCS pressure has stabilized at 325 psig
  * Steam generator pressures are 800 psig and lowering
  * Steam generator pressures are 800 psig and lowering
  * All ECCS equipment has responded as required Which of the following describes when the RCPs should be tripped?
  * fall ECCS equipment has responded as required Which of the following describes when the RCPs should be tripped?
; Immediately When the highest steam generator pressure reaches 700 psi When the highest steam generator pressure reaches 525 psi When the lowest steam generator pressure reaches 700 psi ~
a. Immediately b. When the highest steam generator pressure reaches 700 psig.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio c of Answer 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:
 
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:
KA Ability to operate and / or monitor the following as they apply to Large Break LOCA:
Statement:
Statement:
Securing of RCPs Reference Reference Number Reference Section Page Number (s) Revision Lear Obj Reactor Trip Or S1  IOM 53.A.E-0  Foldout IssIB Rev 5 EOP Generic issues LP-SQS-$ Terminal Ob Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for Examination:
Securing of RCPs         i i RefereIce Reference Number Reference Section Page Number (s)   Revision learn.
Page 65
 
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 Question Source Comments: l
,
M:.t: rial Required for Ex:mliation:
 
l l
<
Page 65 l
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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.


l
[FI-lRC-435] RCL IC Flow indicates 80%.
        !
d. [FI-lRC-414] RCL l A Flow indicates upscale.
; Question Topic: l Determination of RCP/r: actor trip      l i
i f Reactor power is 357o. Which of the following combinations ofloop flow conditions indicates that a reactor l trip should have occurred?
! [FI-lRC-414] RCL. l A Flow indicates 80%.
  [FI-lRC#'q RCL IB Flow indicates 80%.      l
        '
l l b. [FI-IRC-414] RCL 1 A Flow indicates 80%.
i (FI-lRC-415] RCL 1 A Flow indicates 80%. [FI-l RC-414] RCL 1 A Flow indicates downscale.


l  [FI-lRC-435] RCL 1C Flow indicates 80%.
[F1-1RC-415] RCL 1 A Flow indicates 80%.
d. [FI-lRC-414] RCL 1 A Flow indicates upscal [FI-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:
Ars: lb l Exam level: lS l Cognitive Level: l Memory  l Explanatio
KA   Ability to operate and / or monitor the following as they apply to Reactor Coolant Pump Malfunctions:
,
a cf Answer KA: l 015 AAl.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:
Statement:
Reactor trip alarms, switches, and indicators Reference Reference Number Reference Section Page Number (s) Revision Lear Obj R: actor Coolant System - 10M-6.4 IF ll Iss 4 Rev Instrument Failure Procedure     6 Reactor Coolant System LP-SQS- ,6 Question Source l New l Question Modification Method l Qrestion Source Comments: l M:t: rial Required for Enmination:
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
Page 66


_ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ - - _ _ _ _ - _ - _ _ _ - _ _ - _ _ - - . __ __ _ -
- - _ - _ _ - _ _ - _ - _ - - _ _ _ _ _ _ - _ - _ _   _ - - - _ _ _ _ . _- _ _ . _ _ _ _ _ _ _ _ _ .   .
Question Topic: l Failure cf makeup Giv:n the following conditions:
  ' Question Topic: l Fcilure cf makeup Given the following conditions:
- VOLUME CONTROL TANK LEVEL HIGH LOW (A3-53) has alarmed
  - VOLUME CONTROL TANK LEVEL HIGH-LOW (A3-53) has alarmed         ;
- [LI-lCH-115] Volume Control Tank Level (VB-A) failed offscale high Actual VCTlevel will:
  - [LI-1CH-115] Volume Control Tank Level (VB-A) failed offscale high Actual VCT level will:
l remain constan b. decrease until automatic makeup initiate decrease until the charging pump suction transfers to the RWS d. decrease until the VCT is empt Ars: ld l Esam level: lS     l Cognitive Level: l Application l Esplanatio o sf Answer KA: l 022 AA1.08 l RO Value: l3.4 l SRO Value: l3.3 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution  Loss of Reactor Coolant Makeup
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:
Title:
KA Ability to operate and / or monitor the following as they apply to Loss of Reactor Coolant Makeup:
KA Ability to operate and / or monitor the following as they apply to Loss of Reactor Coolant Makeup:
Statement:
Statement:
g VCTlevel Reference     Reference Number  Reference Section Page Number (s) Revision Lear Obj Alarm - A3 53 VCT Level   IOM-7.4.AAX   PC 4,5  A 2-3 Iss 4 Rev High Low           0 CVCS - Instrumentation and    IOM 7. Auto M/U, LCVs 1-2, 8-9 Iss 4 Rev Controls           2 CVCS     LP-SQS- Ill.D. .g. Question Source l New       l Question Modification Method l QIestion Source Comments:     l Material Required for   OM Figure 7-39 Examination:
VCT level Reference   Reference Number  Reference Section Page Number (s) Revision Learn.
        .
 
l        Page 67 L__----___-_____________-_--__-__-_____--___
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:
  ..
I Page 67
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  - _ _ _ _- - . - - _ - - - _ _ _ _ _ _ _ - - - _ - _ . _ - - . . - - _ - _ _ - . - - - _ _ . - _ _ - _ - - _ _  - - _ - - - _ _ - - - .
  - - - - _ _ _ _ _ _ _ - _ _
!
l          .
Question Topic: l Boration and SDM Tech Spec l i
Q:estion Tc'pic: l Boration and SDM Tcch Specs  _
Given the following conditions:
Given the following conditions:       ,
  * [1CH-P-2A] Boric Acid Transfer Pump is out of service
l
  * - [1CH-P-2A] Boric Acid Transfer Pump is out of service
  * RCS Temperature is 420 *F
  * RCS Temperature is 420 *F
  * SDM is 167 delta K/K l * S/D Banks are fully withdrawn if[lCH P-2B] Boric Acid Transfer Pump trips, HOW will required Technical Specification Shutdown Maryn be restored?.
  * SDM is 1.67 delta K/K
! BORATE, by gravity feeding the in-service Boric Acid tank to the blender.
*: 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.


1 Emergency borate through the Emergency Boration valve [MOV-CH-350]. Align the suction of the charging pump to the RWS d. Open the reactor trip breaker Ans: lc l Exam Level: lB l Cognitive Level: l Application     l Explanatio o of Answer KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l l Section: l EPE l RO Group: l 1 l SRO Group: l1
d. Open the reactor trip breakers.
~
 
System / Evolution Emergency Boration Title:
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             i Statement:               J Ability to apply technical specifications for a syste Reference Reference Number   Reference Section     Page Number (s) Revision Lear Obj
KA   Conduct Of Operations Statentent:
~
Ability to apply tcchnical specifications for a system.
Technical Specifications    3.1.1.1, 3.1.2.2, and 3.1. !
 
CVCS   LP-SQS- ,
Reference -  Reference Number Reference Section Page Number (s) Revision Learn.
Question Source l New   l Question Modification Method l
 
_
Obj Technical Specifications    3.1.1.1, 3.1.2.2, and 3.1.2.6 CVCS   LP-SQS-7.1    6 11
Question Source Comments: l
,
  ~
Qeestion Source l New   l Question Modification Method l Question Source Comments:   l
Material Required for Technical Specifications Examination:
  ''
Miterial Required for   Technical Specifications Ex::mination:
Page 68
Page 68


Question Topic: l Emerg;ncy Boration requirements Following a turbine load rejection, control rods are automatically inserted causing ROD CONTROL BANK D LOW LOW alarm (A4-124) to be receive Which of the following is the required action by procedure?
. _ _ - - - _ _ _ _ - - - _ - _ - _ _ - _ - _ - _ _ _ _ - - _  - - _ - - - _ _ _ _ - - _ _ _ - - -  _ - - -
a. Place the rods in manual and withdraw them until the alarm clear b. Place the rods in manual and allow temperature to stabiliz c. Emergency borat d. Borate via the normal flow path until the CONTROL BANK D LOW-LOW alarm clear A*s: lc l Esam Level: lS l Cognitive Level: l Memory l Explanatio
_ ,
,
* 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.
a af Answer
:
,
i Which of the following is the required action by procedure?
KA: l 2.4.31 l RO Value: l3.3 l SRO Value: l3.4 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Emergency Boration Title:
a.
KA Emergency Procedures / Plan
 
'
Place the rods in manual and withdraw them until the alann clears.
Statement:
 
Knowledge of annunciators alarms and indications, and use of the response instruction Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Emergency Boration 10M-7. I  1  iss 4 Rev
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.


Rod Control Bank D Low 10M-l.4.ABF   1 Iss 3 Rev Low       I CVCS LP-SQS- .p Q:estion Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Examination:
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 Qrestion Source Comments:   l
              !
M:t: rial Required for-            !
Ex:mination:
i
'
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   .
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l 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 e__________ __


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I I
Question Topic: l Evt! ofloss cf 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 starte Which of the following describes when the other RHR pump should be started and the basis for this decision?
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.
The second RHR pump should be started: immediately, to avoid any heatup of the RC b. only after investigating the cause of the running pump trip, to avoid losing the second pum c. only after observing an RCS heatup, to avoid unnecessary starts of the RHR pum d. within five minutes, which is the most limiting time until boiling will occu Ans: lb l Exam Level: lS    l Cognitive Level: l Memory      l
'
Esplanatio o cf Answer KA: l 025 AK1.01 l RO Value: l3.9 l SRO Value: l 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 RilRS during all modes of operation R:ference    Reference Number Reference Section    Page Number (s) Revision Lear Obj Residuallicat Removal    AOP 1.1 Caution      2 iss 3A System Loss              Rev5 OM 53C- AOPs    LP-SQS-5 Question Source l New    l Question Modification Method l Qrestion Source Comments:    l Miterial Required for Examination:
Page 70


I Question Te ple: l Loss of CCW during a loss of power 1B and 1C Component Cooling Water Pumps [lCC-P-1B & IC] are BOTH racked to the Connect position on the DF bu Which of the following control switch positions describes when BOTH [lCC-P-IC] and [lCC-P-1B] will fail to restart on a D/G load sequence signal, following a DF bus undervoltage condition?
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?
! [1CC-P-1B]- After START, [lCC-P-1C]- After START l [lCC-P-1B]- PULL-TO-LOCK, [1CC-P-1C]- After Start [1CC-P-1B]- After STOP, [1CC-P-1C]- PULL-TO-LOCK [1CC-P-1B]- After STOP, [1CC-P-1C]- After STOP ATs: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio o cf Answer 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: l1 System / Evolution Loss of Component Cooling Water Title:
l
KA Ability to determine and interpret the following as they apply to Loss of Component Cooling Water:
        !
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:
Statement:
The cause of possible CCW loss Reference Reference Number Reference Section Page Number (s) Revision Lear Obj Reactor Plant Component IOM-15. issue 4 and Neutron Tank Cooling      Rev i Water (CCRS)        l Itactor Plant Component LP-SQS-1 l and Neutron Tank Cooling Water (CCRS)       ,
        {
l Qrestion Source l New  l Question Modification Method l Q:estion Source Comments: l M:terial Required for       j Examination:
The cause of possible CCW loss     l l
l Page 71
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 I
l l
l l
l
,
Page 71 L _-__-


_ _ __ - -__ ___- -_   - _ _ _ _ - _ _ - . _ _ _ . - _ - - - _ - - _ _ _ - _ _ _ _ _ - _ _ - - _ _ _ _ _ _     . .
_ . - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _
Question Topic: l Effect cf r;ference leg br;ak Given the following conditions:
              * ,
  * Reactor power - 100% .
Question Tepic: l Effect cf reference leg break
  * A leak develops on the reference leg for the controlling Pressurizer level sensor l
    '
How will charging flow respond over next five minutes?
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:
Charging flow will:
                ) decrease to the minimum valu !
a. decrease to the minimum value.
b. decrease and then return to the initial value, increase to makeup for the loss through the lea d. increase to the maximum flow valu Ars: la l Exam Level: lS    l Cognitive Level: l Comprehension l Explanatio a rf Answer             ]
 
KA: l 028 AK1.01 l RO Value: l 2.8' 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:
b. decrease and then return to the initial value.
KA Knowledge of the operationa! implications of the following concepts as they apply to Pressurizer Le al Control St-tement: Malfunction:
 
PZR reference leak abnormalities Reference Reference Number     Reference Section Page Number (s) Revision Lear Obj Patssurizer and Pressure LP-SQS- .D. Rr, lief System Reactor Coolant System - lOM-6.4.lF        12 4 6 Instrument Failure Procedure Question Source l New       l Question Modification Method j Qrestion Source Comments: l Miterial Required for Ex:mination:
c. increase to makeup for the loss through the leak.
i Page 72 -
 
i
d. increase to the maximum flow value.
!'
 
!            - - _ _ .-______________________-________-__-________L
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:
I l
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.
l'
Question Topic: l AFW Actuation due to AMSAC


Given the following conditions:
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:
* Reactor power - 100%      l e Both feedwater pumps trip
KA   Ability to determine and interpret the following as they apply to Anticipated Transient Without Scram:
        {'
  * The reactor fails to trip
        )
Which of the following describes when AMSAC should trip the turbine?  ,
i Immediately after the feedwater pumps tri ,
I b. Immediately after feedwater flow decreases below 25% flo j seconds after the feedwater pumps tri d. 25 seconds after feedwater flow decreases below 25% flo I Ans: ld l Exam Level: lS l Cognitive Level: l Memory l
'Explanatio e of Answer KA: l 029 AA2.09 l RO Value: l4.4 l SRO Value: l4.5 l Section: l EPE j 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:
Statement:
Occerrence of a main turbine / reactor trip Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj ATWS Mitigation System 10M-458. ,2 1ss 4 Rev Actuation Circuitry      0 AMSAC  LP-SQS-4 II.D. Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for Examination:
Occurrence of a main turbine / reactor trip Refirence    Reference Number Reference Section - Page Number (s) Revision Learn.
Page 73


l f Question Topic: l Evaluation of SR NIS voltage failure What would be the plant response to the following conditions?
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:
o The plant is operating at 100% power f
;
call systems are NSA oThe "A" train Source range RESET /BT OCK switch is inadvertently turned to the BLOCK positio The reactor would trip, and N31 SR would energize b. The reactor would not trip, and N31 SR would not energize.
          ;
i 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]
 
UFSAR fig. 7.2  sheet 3
    !!!.E.3  7  Iss 4 Rev Reactor Excore Instrument 10M-2.1.c
 
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
,
 
- _ _ - _ - _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ - _ - _ - _ _ _ .  - ,_ - _ - . . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _  _
              ._ _ _ _ - _
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.


I The reactor would trip, and N31 SR would not energize d. The reactor would not trip, and N31 SR would energize Ars: lb l Exam Level: lS    l Cognitive Level: l Application l
'
'
Explanatio n cf Answer l
  + Intermediate channel N-36 fails HIGH.
l 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 IUmge St:tement: Nuclear Instrumentation:
-* Reactor power remains at 6%.
Effects of voltage changes on performance RIference      Reference Numher Reference Section Page Number (s) Revision Lear Obj UFSAR fig. sheet 3    4 Reactor Excore instrument    10M 2. lit. Iss 4 Rev System          I Question Source l New      l Question Modification Method l Question Source Comments:    l M:terial Required for    UFSAR fig. 7.2 sheet 3 Examination:
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.
I        Page 74 L - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ---  --- - _ - _ _ -_


i Question Tople: l Eval of failed IR channel on SU Given the following conditions:
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:
    * Plant startup is in progres * All power range channels indicate 6% reactor powe * Intermediate channel N-36 fails HIG * Reactor power remains at 6%.
KA   Conduct Of Operations Stateunent:
Which of the following describes required operator actions? Initiate a reactor trip, enter E-0, and FR- Immediately commence a controlled reactor shutdow Raise power to greater than PIO and block both intermediate range Continue power operation A~s: Ib l Exam Level: lS l Connitive Level: l Memory' l Explanation of Answer KA: l2. l RO Value: l3.7 l SRO Value: l3.8 l Section: lEPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss of Intermediate Range Nuclear Instrumentation Title:
Knowledge of conduct of operations requirements.
KA Conduct Of Operations St'tement:
Knowledge of conduct of operations requirement Reference  Reference Number  Reference Section Page Number (s) Revision Lear Obj Excore Instrumentation  LP-SQS- V.C.3.c & c 16-17 5 5,8,12 System Conduct of Operations  1/20M-48. VI. iss 3 Rev


Conduct of Operations  I/2LP SQS-4 l QTestion Source l New  l Question Modification Method l Question Source Comments: l       l
Ref rence    Reference Number Reference Section Page Number (s)  Revision Learn.
            !
 
Meterial Required for       l Et mination:
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
 
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:
 
I~
Page 75
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.
Question Tepic: l Fuel Handling accident syst:ms risponse A fuel assembly was suptured during movement in the fuel buildin Which of the following describes how the fuel building evacuation alarm is actuated?
a. The alarm must be manually initiated from the control roo b. [RM-1RM-206] and [RM-1RM-207] Fuel Pool Bridge Area Monitors will sound the evacuation alar [RM-IVS-103A, B] Fuel Building Ventilation Exhaust monitors will sound the evacuation alarm.


i d. The alarm must he manually initiated from either the fuel building or the control roo '
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:
A!s: lc l Exam Level: lS l Cognitive Level: l Memory    l Esplanatio o cf Answer            ,
KA-   Ability to determine and interpret the following as they apply to Fuel llandling incidents:
~KA: l 036 AA2.02 l RO Value: l3.4 l SRO Value: l l Section: l EPE l RO Group: l 3 l SRO Group: l3 System / Evolution Fuel Handling incidents Title:
Statement:
KA- Ability to determine and interpret the following as they apply to Fuel Handling Incidents:
Occurrence of a fuel handling incident Reference Section Page Number (s) l Revision Learn.
Statement:             )
Occunence of a fuel handling incident         1 R,ference  Reference Number Reference Section   Page Number (s) Revision Lear Obj Irradiated Fuel Damage AOP 1.4 ss 3A Rev 3 OM 53C- AOPs  LP-SQS-53 Q estion Source l Facility Exam Bank l Question Modification Method i QYestion Source Comments: l Material Required for Examination:


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
Page 76
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_ _ _ - - - _ _ _ _ _ _ _ _ _ -
              .


i Questio2 Tepic: l R:sponse of SG 1:ak detection monitors At what power level will the steam generator leakage N-16 Radiation Monitors [RM-MS-102A,B, & C]
_ _ _ . _ _
BEGIN to provide valid leak rates, in GPD7 a. 5%
_ _ - _ _ _ _ _ _ - - _ _ - - _ _ _ _ _ - _ _ _
b. 20%
            -
~
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%
c. 30%
d. 50 %
d. 50 %
Ars: lb l Exam Level: lS l Cognitive Level: l Memory  l Explanatio Def Answer KA: l 037 AAl.06 l RO Value: l 3.8* l SRO Value: l 3.9* l Section: l EPE l RO Group: l 2 l SRO Group: l2
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:
I J
Page 77 i
I L__._______________.
 
. _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
            * .
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.
System / Evolution Steam Generator Tube Leak Title:
 
KA Ability to operate and / or monitor the following as they apply to Steam Generator Tube Leak:
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:
Statement:
Main steam line rad monitor meters R'.ference Reference Number Reference Section Page Number (s) Revision Lear Obj Radiation Monitoring 10M-43. Iss 4 Rev Systems - Major components      2 OM $3C- AOPs  LP SQS-53 QIestion Source l Facility Exam Bank l Question Modification Method l Q estion Source Comments: l M:terial Required for Examination:
Eooldown of RCS to specified temperature Reference Reference Number Reference Section Page Number (s) Revision Learn.
Page 77


Question Topic: l Evaluation of coo:down temperature /cooldown
Obj Steam Generator Tube    82  issIB Rupture Background      Rev 5 EOPs  Ll"-SQS-53.3      3
'
_
Given the following conditions-
Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Examination:
* A Steam Generator Tube Rupture has occurred
,
  . E-3, Steam Generator Tube Rupture, is being performed i The RCS has been cooled down to the target temperatur In order to maintain RCS subcooling, intact steam generator pressure must be maintained: greater than the ruptured generato b. equal to the ruptured generator.
l Page 78
_ _ _ _ _ - - -
 
          ,
'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.


, greater than the saturation pressure of the RC d. less than the ruptured generato Ans: ld l Exam Level: lS l Cognitive Level: l Application l Explanatio ,
Obj SG Feedwater System -  10M 24.lD  Feedwater Isolation 2, 6 1ss 4, Instrumentation and Controls      Rev.2
c cf Answer KA:, l 038 EA1.36 l RO Value: l4.3 l SRO Value: l l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Steam Generator Tube Rupture Title:
,
KA Ability to operate and / or monitor the following as they apply to Steam Generator Tube Rupture:
Reactor Protection System LP-SQS-l .1  V.E.5 38 6 9
Statement:
''
Cooldown of RCS to specified temperature Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Steam Generator Tube    82  iss 1B Rupture Background      Rev 5 EOPs  LP-SQS 5 Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l M terial Required for Examination:
Question Source jNew  l Question Modification Method l Question Source Comments: ]
l l
Mat: rial Required for Enmination:
    .
!
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l
          ]
l Page 79
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Question Topic: l Ev;luation c.f FW condition Given the following conditions:
          ,
  * A steam break has occurred on SG "A" l * A reactor trip was manually initiated I
          - .,
'
Question Tepic: l Effect & mitigation techniques Given the following conditions:
o A Si has NOT been initiated o No operator actions have oeen performed on the feedwater syste * Only SG "A" narrow range level has decreased below 12%.
  * An uncontrolled depressurization of all steam generators has occurred
  * RCS T,., are (A) 542 F,(B) 550 'F,(C) 550 F i Which of the following is the expected status of feedwater?
  * 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?
l The feedwater regulating valves will be shut. The Turbine Driven AFW pump will be runnin b. The feedwater regulating valves will be shut. All AFW pumps will be runnin A complete FWI isolation will be initiated. All AFW pumps will be runnin d. The feedwater system will be in the same lineup as prior to the reactor trip, except the FRVs will be throttled close Ans: la l Exam Level: lS l Cognitive Level: l Application l Explanatio ocf Answer KA: l 040 AA1.02 l RO Value: l4.5 l SRO Value: l4.5 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Steam Line Rupture Title:
a. A minimum AFW flow to all steam generators is maintained.
KA Ability to operate and / or monitor the following as they apply to Steam Line Rupture:
Statement:
Feedwater isolation Reference  Reference Numher Reference Section Page Number (s) Revision Lear Obj SG Feedwater System - 10M-24.lD  Feedwater Isolation 2, 6 iss 4, instrumentation and Controls      Re Reactor Protection System LP-SQS- V. ~
Question Source l New  l Question Modification Method l Q'estion Source Comments: l Material Required for Examination:


i I
b. SGs are intermittently fed to assure that a wide range levels remain above 10%.
Page 79 l
c. Only reducing AFW flow as necessary to reduce the cooldown rate to less than 100 F.


Q::estio3 Topic: l Effect & mitigation techniques Given the following conditions:
d. AFW feed rate is limited to maintain constant level, provided the level is above 10% wide range.
*
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 minimum AFW flow to all steam generators is maintaine b. SGs are intermittently fed to assure that a wide range levels remain above 10%. Only reducing AFW flow as necessary to reduce the cooldown rate to less than 100 d. AFW feed rate is limited to maintain constant level, provided the level is above 10% wide range.


] Ans: Ia l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio n 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:
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:
KA Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture:
Statement:
Statement:
Effects of feedwater introduction on dry S/G Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Uncontrolled  ECA- STEP 6  5  issIB, Depressurization of all SGs      re ,
EfTects of feedwater introduction on dry S/G Reference  Reference Number . Reference Section Page Number (s) Revision  Learn.
Uncontrolled 10M-538.4.ECA- I issIB; Depressurization of all SGs      Re Background EOPs  LP-SQS-5 Question Source l New  l Question Modification Method l Question Source Comments: l Material Required for Examination:
Page 80


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Obj Uncontrolled  ECA-2.1  STEP 6  5  iss1B, Depressttrization of all SGs      rev.4
Question Topic: l Block of steam dumps on turbine trip A loss of condenser vacuum has occurred due a leak in the condenser. Main Condenser Steam Dumps are open following a turbine tri As vacuum decreases, at what condenser vacuum will Main Condenser Steam Dumps close?
,
a. 25" Hg Vacuum b. 20" Hg Vacuum c. 10" Hg Vacuum " lig Vacuum Ans: lb l Exam Level: lS l Cognitive Level: l Memory     l Explanatio n of Answer KA: l 051 AK3.01 l RO Value: l2.8* l SRO Value: l 3.I' l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution  Loss of Condenser Vacuum Title:
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 J
l l
Page 80 L__________
 
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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:
KA  Knowledge of the reasons for the following responses as they apply to Loss of Condenser Vacuum:
Statement:
Statement:
Loss of steam dump capability upon loss of condenser vacuum h Reference  Reference Nuinber Reference Section    Page Number (s) Revision Lear Obj Main Steam Supply / Steam  LP-SQS-2 i
I ms of steam dump capability upon loss of condenser vacuum R:.f;rence  Reference Number   Page Number (s) Revision   Learn.
              !
 
Dump System 10M-26. '
l Reference Section Obj
Question Source l New  l Question Modification Method l Question Source Comments:  l Material Required for
            '
_
25  4    ,
Examination:
M in Steam Supply / Steam  LP-SQS-21.1 Dump System           _
l l
1OM-26.2.B  IO 4    7
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Q:estion Source - l New  l Question Modification Method l Q estion Source Comments:  l Mit: rial Required for
.
Ex:mination:
Page 81
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.


I i
d. All SG pressures would be decreasing from the break via the auxiliary feedwater lines.
Question Tople: l D. termination cf Feedline break
 
'
Ans: la l Exam Level: lS   l Cognitive Level: l Comprehension l Explanatio
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 setpoin Following the reactor trip and SI, which of the following SG pressure indications would exist? Only SG "A" pressure would be decreasing from the brea b. All SG pressures would be decreasing from the break via the main steam line All SG pressures would be decreasing from the break via the main feedwater line d. All SG pressures would be decreasing fro:n the break via the auxiliary feedwater line A*s: la l Exam Level: lS l Cognitive Level: l Comprehension l Esplanatio
} u of Answer
  ,
  ,
e cf 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: 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:
KA Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater:
Stat ment:
Statement:
MFW line break depressurizes the S/G (similar to a steam line break)
MFW line break depressurizes the S/G (similar to a steam line break)
Reference  Reference Number Reference Section l Page Number (s) Revision  Lear Obj M in Steam Supply / Steam LP-SQS-2 ,4g Dump System Miin Steam System I OM-21. iss 4 Rev
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


VOND 24-1 Q estion Source l New l Question Modification Method l Q:estion Source Comments: l Mit: rial Required for Ex:mination:
VOND   24 1 Question Source l New   l Question Modification Method l Question Source Comments:   l
~ Material Required for Examination:
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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 i
i Page 83 f-
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                . . j 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:
Questi:a Tcpic: l Load required to be left in AUTO l
KA   Knowledge of the reasons for the following responses as they apply to Station Blackout:
A loss of all 4KV busses has occurred. ECA-0.0 has been implemented to the point of placing deenergized l equipment in PULL TO LOCK. The iDF emergency bus has been selected to cross tie to Unit Which of the following l AE Emergency Bus loads shall remain in the AUTO position and the basis for leaving that pump in AUTO? Reactor River Water Pump to assure that the diesel has cooling upon startu Charging Pump to restore seal flo Charging Pump to restore Pressurizer leve d. Component Cooling Water Pump to restore cooling to the thermal barrie Avs: la l Exam Level: lS   l Cognitive Level: l Memory     l Explanatio a cf Answer 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:
St:tement:
KA Emergency Procedures / Plan St:tement:
Actions contained m EOP for loss of olTsite and onsite power Reference Number  Reference Section   Page Number (s) Revision Learn.
Knowledge of operational implications of EOP warnings, cautions, and note Reference  Reference Number  Reference Section     Page Number (s) Revision Lear Obj Loss of All Emergency 4KV 10M-53 A. I.ECA- Caution Step 14    10 1ss1B AC Power            Rev 4 Emergency Operating LP-SQS-5 Procedures Qrestion Source l Facility Exam Bank  l Question Modification Method l QIestion Source Comments: l M terial Required for Examination:
Page 83
              . _ - _ .


Question Topic: l Purpose of Si R set If an SI actuation signal is received when performing ECA-0.0, " Loss of All Emergency 4KV Power", the SI signal should be:
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
a. ' reset to prevent lockout of the stub busse b. reset to permit manual loading of equipment of an Emergency bu allowed to remain active to ensure rapid injection of core cooling water when power is restore d allowed to remain active to ensure the load sequencer re-initiates when the DG starts.


Ass: lb l Exam Level: lS l Cognitive Level: l Memory  l Explanatio e cf Answer KA: l 055 EK3.02 l_ RO Value: l4.3 l SRO Value: l4.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Station Blackout Title:
EOPs     LP-SQS-53.3 Question Source l NRC Exam Bank       l Question Modification Method l Question Source Comments:     l Material Required for Examination:
KA  Knowledge of the reasons for the following responses as they apply to Station Blackout:
              .
St::tement:
Actions contained in EOP for loss of offsite and onsite power Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Loss of All Emergency 4KV ECA- steps 31 & 37 22 &25 iss 1B; AC Power      Rev 4 Loss of All Emergency 4KV 10M-53 B.4.ECA- Step 31, Basis 127 Iss IB; AC Power Background     Rev 4 EOPs  LP-SQS-5 Q estion Source l NRC Exam Bank l Question Modification Method l Question Source Comments: l M terial Required for Examination:
      .
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Page 84
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Question Teple: l RCS temperatures What is the expected response of RCS Ilot and Cold leg temperatures during the first few minutes following a reactor trip froml00% power COINCIDENT with a loss of offsite power? Ilot leg temperatures will rise, and Cold leg temperatures will remain relatively constant, until natural circulation flow is establishe b. Ilot leg temperatures and Cold leg temperatures will both rise, until natural circulation flow is establishe Ilot leg temperatures will remain relatively constant and Cold leg temperatures will drop, until natural circulation flow is establishe d. Ilot leg temperatures will rise and Cold leg temperatures will drop, until natural circulation flow is establishe ATs: la    l Exam Level: lS l Cognitive Level: l Memory        l Explanatio o cf Answer KAt 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 lystem/ Evolution    Loss of Off-Site Power Title:
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?
KA    Ability to determine and interpret the following as they apply to Loss of Off-Site Power:
a. Hot leg temperatures will rise, and Cold leg temperatures will remain relatively constant, until natural circulation flow is established.
St:tement:                  __
 
Reactor coolant temperature, pressure, and PZR level recorders Reference      Reference Number    Reference Section    Page Number (s) Revision Lear Obj R: actor Trip Response    ES- Note before step 3    3-4 Iss lil Rev 4 EOPs      LP-SQS-5 l Q:estion Source l Facility Exam llank    l Question Modification Method l Q:estion Source Comments:    l Miterial Required for Examination:
.
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.
Page 85 l
 
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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:
_
l l
 
            )
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I-Question Topic: l Effect of a loss of Vital AC on Feedw:ter Given the following conditions:
_ _. . _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _
l
,          . .
        *
I Question Tcpic: l Effect of a loss of V;tal AC on Feedwater Given the following conditions:         )
            )
I
  * Reactor power is 74%
  * Reactor power is 74%
  * Feedwater control is in automatic
  * 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? -
  * 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?
a. The FRVs willimmediately fail cpe b. The FRVs will immediately fail close The FRVs will drift shu d. ' The FRVs will transfer to either MANUAL or AUTO HOL A''s: ld l Exam Level: lS l Cognitive Level: l Memory l Explanatio ocf Answer KA: l 057 A A2.19 l RO Value: l4.0 l SRO Value: l4.3 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Loss of Vital AC Instrument Bus Title:
c. The FRVs will immediately fail open.
KA Ability to determine and interpret the following as they apply to Loss of Vital AC Instrument Bus:
 
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:
Statement:
The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus Reference Reference Number Reference Section Page Number (s) Revision Lear Obj Alarm VitalIlus I,II,111,IV 10M-38.4.AAA, AAC,  2 Trouble  AAE. AAG 120V AC Distribution LP SQS 3 System Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Examination:
The plant automatic actions that will occur o - [/of a vital ac electrical instrument bus     {
Page 86
,          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 Question Source j Facility Exam Bank l Question Modification Method l     j Question Source Comments: l Material Required for Er.mination:
I
 
l l
l l
l Page 86       l l
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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 Question Source l New      l Question Modification Method j    j Question Source Comments:     l       l M:.terial Required for      IOM 39.5.B.6(28 pages)
        -__
Ex mination:
Question Tcpic: l Effect of a loss of DC on RCPs l Which of the following is the effect that losing 125 VDC Bus I will have on the Reactor Coolant Pumps?
l 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 l c. Component cooling water will be lost to all RCP d.- Seal water flow to the RCPs will be isolate Ans: lc l Eram Level: lS l Cognitive Level: l Application l Esplanatio oaf Answer KAt . 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 SystIm/ Evolution Loss of DC Power Title:
              ,
KA Ability to determine and interpret the following as they apply to Loss of DC Power:
i i
St:tement:
j l
DC loads lost; impact on to operate and monitor plant systems Ref;rence Reference Number Reference Section Page Number (s) Revision Lear Obj OM 39  10M-39. 5. Table 39-6  all iss 4 Rev
l i
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              !
QIestion Source lNew  l Question Modification Method l Q:estion Source Comments: l M: trial Required for IOM-39.5.B.6(28 pages)
l Page 87
Enmination:
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Page 87
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Question Topic: l Evalef Tech Spec Given the following conditions:
Question Topic: l Evalef Tech Spec Given the following conditions:
e Unit 1 is in MODE 6
  *  Unit 1 is in MODE 6
  * Unit 2 is in MODE 1
  * Unit 2 is in MODE 1
  * Movement ofirradiated fuel is ongoing in the Unit 1 Containment only
  = Movement ofirradiated fuel is ongoing in the Unit 1 Containment only
* Monitor RM-1RM-218A Control Room Area - Unit I has failed low What action is required for the above conditions? 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 operabl Within ONE hour, verify that Control Room Area - Unit I monitor [RM-1RM-218B] is operabl Within ONE hour, suspend all operations involving movement ofirradiated fuel.
  * 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.


A s: lb l Exam Level: lS l Cognitive Level: l Application l Explanatio
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 Question Source lNew      l Question Modification Method l Question Source Comments:     l Mit: rial Required for     Tech Specs Exaination:
      '
o e f Answer KA: l 061 AA2.06 l RO Value: l3.2 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l SRO Group: ~l2 System / Evolution Area Radiation Monitoring System Title:
KA Ability to determine and interpret the following as they apply to Area Radiation Monitoring System:
Statement:
Required actions if alenn channel is out of service Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Be;ver Valley - Unit i  3.3.3.1, Table 3.3-6,1.c, 3/4 3-33-3-35 Amend Technical Specifications  Action 41   119 Radiation Monitoring System LP-SQS-4 V .a Qyestion Source l New  l Question Modification Method l Q:estion Source Comments: l Material Required for Tech Specs Examination:
Page 88
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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?
Question Topic: l EfYect cf restoring air using IIA-9 During a loss of containment air, which of the following is the possible effect of opening [llA-90]
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:
Instrument Air to Containment Air Isol Valve too quickly? Station Air compressor trips I
b. CVCS letdown isolation SG Main FW Feed Reg Vivs failing open d. Main Steam Line Trip Valve closure Ass: ld l Exam Level: lS l Cognitive Level: l Memory l Explanatio o cf 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:
KA Knowledge of the reasons for the following responses as they apply to Loss ofinstrument Air:
Statement:
Statement:
Actions contained in EOP for ' ass ofinstrument air Reference  Reference Numaer Reference Section Page Number (s) Revision Lear Obj Loss of Containment AOP 1.3 Caution before step 4 3   iss 3A Instrument Air       Rev 3 OM $3C- AOPs  LP-SQS-53 Question Source l New  l Question Modification Method l Qrestion Source Comments: l Miterial Required for Examination:
Actions contained in EOP for loss of instrument air Reference  Reference Number Reference Section Page Number (s) Revision Learn.
i
 
      .
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 Question Source l New  l Question Modification Method l Q:estion Source Comments: l Meterial Required for Ex mination:
i I
i, t
Page 89
Page 89
      . _ _ _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ ..
        . _ _ ___-________-__ ____ D


_ _ _ _ _ . _ - _ - = _ _ _ _ _ _ _ - _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ . _- __ _ ___ - - -. . _ _ _ _ - _ - _ _ . . _ _ _ _ _ _ _ _ _ _ _
        - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ ..
:
            . .
Question Tepic: l Type of detection / extinguishing eqpt for use
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.
Which of the following describes the fire protection afforded for the primary process rack area?         j Carbon Dioxide is released to the area by manual actuation onl Carbon Dioxide is released to the area by automatic actuation of smoke detection or by manual actuatio Halon is released to the area by manual actuation onl Halon is released to the area by automatic actuation of smoke detection or by manual actuation.
 
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.


l Ars: ld     l Exam Level: lS  l Cognitive Level: l Memory l Espinnatio cef Answer KA: l 067 AA1.08 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:
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:
K Ability to operate and / or monitor the following as they apply to Plant Fire on Site:
KA  Ability to operate and / or monitor the following as they apply to Plant Fire on Site:
Statement:
Statement:
Fire fighting equipment used on each class of fire Reference       Reference Number Reference Section Page Number (s)   Revision Lear Obj Fire Protection System -     IOM-33. flalon paragraphs 1 & 4 5    iss 4; Summary Description             Re Fire Protection System     LP-SQS-3 E. l .e Qrestion Source l New       l Question Modification Method l Qrestion Source Comments:     l M;terial Required for Ermination:
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:
Page 90
Page 90
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    .
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_ . - _ _ _ _
i 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.


Question Tepic: l Pressurizer level control A fire in the control room has resulted in control room evacuation. Plant control has been transferred to local control panels as requir.:d by lOM-56C.1, Altemate Safe Shutdown from Outside the Control Roo Until a cooldown is initiated from the BIP, pressurizer level is maintained by charging via: [MOV-RC-556A, B, C) Reactor Coolant Loop Fill Valves to the RCS loop b. the normal charging connectio the RCP seal d. the BI A7s: lc l Exam level: lS l Cognitive Level: l Memory l Explanatio n cf Answer KA: l 068 AA1.30 l RO Value: l3.4 l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1
d. the BIT.
; System / Evolution Control Room Evacuation Title:
 
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:
KA Ability to operate and / or monitor the following as they apply to Control Room Evacuation:
St:tement:
Statement:
Operation of the letdown system Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Alternate Safe Shutdown LP-STA-56 VI. .a from Outside the Control Room       _
Operation of the letdown system Reference  Reference Number Reference Section Page Number (s)     Revision Learn.
Question Source lNew  l Question Modification Method l Q estion Source Comments: l M terial Pequired for Examination:
 
Page 91
Obj Alternate Safe Shutdown . LP STA 56C.! VI.A.3  12      2 6.a from Outside the Control Room
              .
I Question Source l New  l Question Modification Method l Question Source Comments: l M;tirist Required for Ex:mination:
-
,
              )
Page 91 l
i (      _ _ _ . . _ _ . _ _ _ . . _ . _ _ _ _ - - . _ _ _ _ _ _ _ _ . . _ _ _ _ - - - - _ _ _ _ _ _  _d


Ouestion Tepic: l Controller locition Which of the following identifies the components used by the operator stationed at the BIP (Backup Indicating Panel) to lower pressurizer level? [SOV-1RC-102B] RCVS Reactor Vessel Vent Viv
__- _ _ _ _ - . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
          . .
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-103B] RCVS Pressurizer Vent Viv
  [SOV-1RC-105] RCVS Vent to Containment Isolation Viv b. [LCV-1CH-460A and B] Ltdn to Regen Hx Isol
  [SOV-1RC-105] RCVS Vent to Containment Isolation Viv b. [LCV-1Cli-460A and B] Ltdn to Regen lix Isol
  [TV-CH-200B] 60 GPM Ltdn Orifice Cnmt Isol Viv Letdown will flow to the degasifier via [LCV-115A], which has failed to the degasifier position.
  [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 Question Source l New  l Question Modification Method l Question Source Comments: l M;terial Required for Ex:mination:
l
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l    Page 92 I
\
L.
 
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-
a, 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 .
l Emergency Operating    LP-SQS-53.3    1 3 Procedures i
Question Source l Facility Exam Bank    l Question Modification Methcd l Question Source Comments:    l Material Required for Ex mination:


I [MOV-CII-201] Excess Ltdn HX Inlet Isolation Viv
!
!
[MOV-lCH-137] Excess Ltdn HX Flow Control Viv [PCV-1RC-455D] PZR PORV Relief Viv
Page 93 l
[PCV-IRC-456] PZR PORV Relief Viv
l L
      ~
 
,
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
Ans: la l Exam Level: lS l Cognitive level: l Memory l Explanatio ecf 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:
)::est6on Top 6c: l Actions to lower R/A lev Is Ziv:n the following conditions:
St:tement:
* Reactor power has just been raised from 20% to 100%
Auxiliary shutdown panellayout Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Misc. Safety-Related 10M-45. (BIP) Indications 7  1ss 4 Rev Systems Summary      1 Description Alternate Safe Shutdown LP-STA 56 Outside the Control Room R: actor Coolant System - lOM-6. Instrumentation and Controls Qyestion Source l New  l Question Modification Method l QIestion Source Comments: l M terial Required for Examination:
* Dose Equivalent Iodine has just been reported as 5.0 pci/ gram.
Page 92
 
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:
Question Topic: l Basis f r starting rn RCP            ;
KA Knowledge of the reasons for the following responses as they apply to liigh Reactor Coolant Activity:
An RCP is started in FR-C.1, " Response to Inadequate Core Cooling", in order to: allow using RVLIS Dynamic Range indication to determine core void conten temporarily improve core cooling until some form of makeup flow to the RCS can be establishe enhance the cooling caused by rapid depressurization of the steam generator establish pressurizer spray flow to reduce RCS pressure to cause low pressure systems to injec Ans: Ib l Esam Exvel: IS    l Cognitive Level: I Comprehension l Explanation i
Stateent:
of Answer l KA: l 074 EK2.01 l RO Value: l3.6 l SRO Value: l3.8 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution inadequate Core Cooling Title:
Corrective actions as a result of high fission-product radioactivity level in the RCS Reference  Reference Number Reference Section Page Number (s) Revision Learn.
KA Knowledge of the inter;lations between Inadequate Core Cooling and the following:
Statement:
RCP Reference  Reference Number   Reference Section l Page Number (s)   Revision Lear Obj Response to inadequate Core 10M 538.4.FR- I    Iss IB Cooling Background            Rev4 Emergency Operating LP-SQS-5 Procedures Question Source l Facility Exam Bank    l Question Modification Method l Question Source Comments: l M:terial Required for Examination:
l Page 93


l Question Topk: l Actions to lower R/A levels Given the following conditions:
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
* Reactor power hasjust been raised from 20% to 100%
; Question Source l NRC Exam Bank  l Question Modification Method l l Questic3 Source Comments: l M:terial Required for Ex:mirtion:
;
* Dose Equivalent Iodine hasjust been reported as 5.0 ci/ gra Which of the following explains why operation can continue with Dose Equivalent Iodine above the Technical Specification LCO limit? To allow for CVCS removal of the crud released by the power chang b. The Technical Specification LCO limit is conservative enough, to allow extended periods (> 7 days)
of exceeding the limi c. To accommodate the iodine that was released during the power chang d. The probability of a Large break LOCA occurring during the time period Iodine is above the limit, presents an acceptable ris Ans: ic l Exam Level: lS l Cognitive Level: l Memory  l Explanatio c ef Answer 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 System / Evolution liigh Reactor Coolant Activity Title:
KA Knowledge of the reasons for the following responses as they apply to High Reactor Coolant Activity:
Statement:
Corrective actions as a result of high fission-product radioactivity level in the RCS Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Technical Speci0 cations LP-SQS-TS    0 4 Bc:ver Valley Unit i  Bases 3/4 4-4 83/44-4 Amend No 102 Question Source l NRC Exam Bank  l Question Modification Method l Qrestion Source Comments: l M;terial Required for Extmins*lon:
Page 94
Page 94
      .- .
        .________--__---_--- - - - - -
I 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.


Question Topic: l Securing Si flow Which of the following describes the required subcooling requirements before ternt nating i SI in ES-1.1, S1 Termination?
b. is based on the expected pressure after Si is terminated.
The required subcooling: is based on saturation conditions plus instrument errors, b. is based on the expected pressure after SI is terminate is based on the expected temperatures 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 reinitiatio A's: la l Enam Level: lS l Cognitive Level: l Memory  l Esplanatio o ef Answer
 
_KA: l E02 EK 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 S1 Termination Title:
d. provides for a 50 *F margin to saturation to avoid reinitiation.
KA Knowledge of the reasons for the following responses as they apply to S1 Termination:
 
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:
Statement:
Normal, abnormal and emergency operating procedures associated with (S1 Termination).
,  Nonnat, abnormal and emergency operating procedures associated with (Si Termination).
 
Reference  Reference Number Reference Section Page Number (s) Revision Learn.


Reference  Reference Number Reference Section Page Number (s) Revision  Lear Obj SI Termination /Reinitiation IOM053B.5.Gl-11 II. issIB Rev1 EOP Generic issues LP-SQS-5 i   LO 3 Question Source iNew  l Question Modification Method l Q estion Source Comments: l Miterial Required for Examination:
Obj St Termination /Reinitiation IOM053B.5.GI l1 II.A.i  3  issiB RevI EOP Generic issues LP-SQS-53.2    i LO3 Question Source l New  l Question Modification Method l Q estion Source Comments: l M;terial Required for Examination:
J
          !
          (
i
i
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?
           )
Page 95
Page 95
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- _ _ _ _ _ _ _ _ . __-.        l
 
_ _ _ - . _ _ _ _ _ _ _ _ - . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _  __ _ _ _ _ _ __
            * .
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.
___
Question Topic: l Basis for required Pressmaer Level A reactor trip and SI have occurred, and the control room operators are responding to a small-break LOCA.


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].
d. adequate PZR steam space to absorb pressure fluctuations during RCP start.
In addition to ensuring that RCS conditions are under adequate operator control, the basis for this pressurizer level ensures: that a reduction in subcooling does not occur when SI flow is reduce sufficient inventory such that PZR level does not drop low when an RCP is starte pressurizer level indication is not due to a void in the vessel hea d. adequate PZR stearn space to absorb pressure fluctuations during RCP start.


Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio n of Answer KA: l E03 EK l RO Value: l3.7 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution LOCA Cooldown and Depressurization Title:
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:
KA Knowledge of the interrelations between LOCA Cooldown and Depressurization and the following:
Statement:
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 tL facility.
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 Question Source . l NRC Exam Bank  l Question Modification Method l Question Source Comments: l M'.t: rial Required for Extininstion:
,
Page 96
 
F 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.


Reference  Reference Number Reference Section Page Number (s) Revision Lear Obj Post LOCA Cooldown and ES- step 15  10  issIB Depressurization      Rev5 Post LOCA Cooldown and l OM-53 B.4.ES- issIB Depressurization      Rev 5 EOP Generic issues  I LP-SQS-5 II.B.1, Il ,10  3, 4 Question Source l NRC Exam Bank  l Question Modification Method l Question Source Comments: l Material Required for Examination:
d. ECA-1.1 " Loss of Emergency Coolant Recirculation", in order to deal with the loss of available inventory for core cooling.
l Page 96


Question Topic: l Purpose of ECA Given the following conditions:
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: ,
o A small break LOCA has occurred due to a break at some unknown location outside containmen o Performance of ECA - 1.2 "LOCA Outside Containment" did not isolate the brea o At the completion ECA - 1.2 "LOCA Outside Containment", RCS pressure is still dropping At the conclusion of ECA - 1.2 "LOCA Outside Containment" the operating crew should transition to E-0 "Rx Trip or SI" in order to reverify that all automatic actions have been complete b. ' E-3 "SGTR", since there are adequate steps within this procedure to deal with these condition ES-0.0 "Rediagnosis" in an attempt to diagnosis the break locatio d. ECA-l.1 " Loss of Emergency Coolant Recirculation", in order to deal with the loss of available inventory for core coolin A s: ld l Exam Level: lS l Cognitive Level: l Comprehension l Esplanatio o cf Answer
, ;
~
K- " 204 EK l RO Value: l3.8 l SRO Value: l4.0 l Section: lEPE l RO Group: l 2 l SRO Group: l1 System / Evolution LOCA Outside Containment Title:
KA Knowledge of the interrelations between LOCA Outside Containment and the following:
KA Knowledge of the interrelations between LOCA Outside Containment and the following:
Statement:
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 facilit Reference Reference Number Reference Section Page Number (s) Revision Lear Obj LOCA Outside Containment IOM-53B.ECA- IssIB Background      Rev 3 Emergency Operating LP-SQS-5 Procedures Q estion Source l New  l Question Modification Method j Qrestion Source Comments: l M;terial Required for Examination:
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.
s Page 97
 
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 i
Q estion Source l New  l Question Modification Method l QIestion Source Comments: l Mit: rial Required for         ,
Ex mination:         !
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l    Page 97
 
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      - . _ _ _ _ _ _ - - _ - _ - - _ _ _ _ _ - - _ _ _ _ _ - - _ _ _
s 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 :


.
l c. Requiring all CRDM fans to be runnung.
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _  _ __ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _    _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _
Questics Tcpic: l Apply procedural direction for cooldown              )
l
                  '
During a natural circulation cooldown with RVLIS unavailable, 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 volds in the RCS head ri.gion by : Requiring all CRDM fans to be runnung.


b. Limiting the allowable increase in pressurizer level.
b. Limiting the allowable increase in pressurizer level.


. Limiting the maximum temperature on Core Exit Thermocouple.
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.


l
Reference  Reference Number Reference Section  Page Number (s)  Revision Learn.
, Requiring a minimum of 200F subcooling.
 
Obj NaturalCirculation  ES-0.4  step 9  8    Iss1B Coold:wn With Steam YcM        Rev 4 in Vessel (Without RVLIS)


l Ans: lb  l Exam Level: lS    l Cognitive Level: l Comprehension l Explanatio a cf Answer KA: l E10 EA l RO Value: l3.4 l SRO Value: l3.9 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:
EOPs LP-SQS-53.3 Q:estion Source l Facility Exam Bank - l Question Modification Method l Q:estion Source Comments: l Mat: rial Required for Ex mination:
KA  Ability to determine and interpret the following as they apply to Natural Circulation with Steam Void in Vessel StItement:  with/without RVLIS:
Adherence to appropriate procedures and operation within the limitations in the facility's license and amendment Ref;rence    Reference Number    Reference Section Page Number (s)    Revision Lear Obj N:tural Circulation  ES- step 9  8    Iss1B Cooldown With Steam Void              Rev 4 in Vessel (Without RVLIS)
EOPs    LP-SQS-5 Qr.',stion Source l Facility Exam Bank       l Question Modification Method l Qxestion Source Comments:   l M;terial Required for Exr.mination:
Page 98
Page 98
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Question'Tople: l Condition retulting in loss cf recire      !
f l
Given the following conditions:
'
j
Question Topic: l Condition resulting in loss ef recirc Given the following conditions:
  * A LOCA has occurred a Due to low RWST level a transfer to Cold Leg Recirculation has occurred.
  * A LOCA has occurred
 
* Due to low RWST level a transfer to Cold Leg Recirculation has occurre * All automatic actions for the transfer to Cold Leg Recirculation are
*
;  complete.
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 EOPs  LP-SQS-53.3      6 Question Source l New  l Question Modification Method l Question Source Comments: l Mat: rial Required for          j Ex:minatlon:
i i
l i
l i
l Page 99
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_ _ _ _ . . _ _ _______ _ -___.__- _ _ _ _ _ _ _ _ _ _
  - * [ISI-P-1B] LHS1 pump is not available
          '
  .
s Question Topic: l CIB setpoints How long aft:r a CIB signalis received will the quench sprcy and containment spray pumps start?
  * Containment pressure - 12.4 psig Which of the following would result in a loss ofinjection flow? RCS pressure - 450 psig
t a. [QS-P-1 A,B] Quench Spray pumps - 5 seconds
  [MOV-ISI-862A] 1 A LHSI Pump RWST Suct Viv fails open b. RCS pressure -250 psig
  [lRS-P-2A, B) Outside Recirc Spray Pumps = 120 seconds
  [MOV-1SI-863A] 1 A LHS1 to Chg Pumps Sup Viv fails closed
[lRS-P-I A, B] Inside Recirc Spray Pumps = 22. seconds b [QS-P-1 A,B] Quench Spray pumps - 60 seconds
  . RCS pressure - 380 psig (CH-P-1 A] 1 A Charging /HHSI Pump trips
  [lRS-P-1 A] Inside Recirc Spray Pump, [lRS-P-2B) Outside Recirc Spray Pump = 120 seconds
  [MOV-ISI-863B] 1B LHSI To Chg Pumps Sup Valve fails close d. RCS pressure - 180 psig
[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
  [MOV-ISI-885A] 1 A LHSI PP Mini Flow Isol Valve fails open A s: lb  l Exam Level: lS      l Cognitive Level: l Comprehension l Explanatio n ef Answer                 ,
[lRS-P-1 A, B] Inside Recirc Spray Pumps = 210 seconds
KA: l ElI EA l RO Value: l3.4 l SRO Value: l4.2 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution     Loss of Emergency Coolant Recirculation Title:
[lRS-P-2A, B) Outside Recirc Spray Pumps = 225 seconds d. [QS-P-1 A,B] Quench Spray pumps - 5 seconds
KA   Ability to determine and interpret the following as they apply to Loss of Emergency Coolan: Recirculation:
[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:
Statement:
Facility conditions and selection of appropriate procedures during abnormal and emergency operation Reference       Reference Number Reference Section Page Number (s) Revision  Lear Obj Transfer To Cold Leg        ES step 4  3 issIB Recirculation            Rev 4 EOP Attachment 1-0        IOM-53 A.I .1-G  step 2 2 IssIB Rev 2 EOPs        LP-SQS 5 Question Source l New        l Question Modification Method l Question Source Comments:        l Material Required for Examination:
Annunciators and conditions indicating signals, and remedial actions associated with the (High Containment Pressure).
Page 99
 
_ _ _ _ _ _  ______ _ _ _ - _ -
Reference Section Page Number (s) Revision Learn.
 
Ref;rence Reference Number Ob]
IOM-13.2.B    2   Iss 4 Rev Containment


Question Topic: l CIB setpoints How long after a CIB signal is received will the quench spray and containment spray pumps start? [QS-P-1 A,B] Quench Spray pumps - 5 seconds
Depressurization System 27  5 5 Containment  LP-SQS-13.01 Depressurization System Question Source l New   I Question Modification Method l Cc. ion Source Comments: l Material Required for Examination:
  [lRS-P-2A, B] Outside Recirc Spray Pumps = 120 seconds
l l
  [lRS-P-I A, B] Inside Recirc Spray Pumps = 225 seconds [QS-P-1 A,B] Quench Spray pumps - 60 seconds
l Page 1(Mi
  [lRS-P-l A) Inside Recirc Spray Pump, [lRS-P-2B] Outside Recirc Spray Pump = 120 seconds
  - -   _ _ __ -
  [lRS-P-1B] Inside Recirc Spray Pump, [lRS-P-2A] Outside Recire Spray Pump = 210 seconds [QS-P-1 A,B] Quench Spray pumps - 60 seconds
   [1RS-P-1 A, B] Inside Recirc Spray Pumps = 210 seconds
  [lRS-P-2A, B] Outside Recirc Spray Pumps = 225 seconds [QS-P-1 A,B] Quench Spray pumps - 5 seconds
  [lRS-P-1 A] inside Recire Spray Pump, [lRS-P-2B] Outside Recirc Spray Pump = 210 seconds
  [lRS-P-1B] Inside Recirc Spray Pump, [lRS-P-2A] Outside Recire Spray Pump = 225 seconds Ans: Id  I Esam Level: IS  l Cognitive Level: l Memory l Explanation of Answer KA: l E14 EKl.3 l RO Value: l3.3 l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution liigh Contaimnent Pressure Title:
KA   Knowledge of the operational implications of the following concepts as they apply to High Contaimnent Pressure:
Setement:
Annunciators and conditions indicating signals, and remedial actions associated with the (High Containment Pressure).


Reference    Reference Number Reference Section Page Number (s) Revision Lear Obj Containment  IOM 13. Iss 4 Rev Depressurization System      3 Containment  LP-SQS-13.01  27  5 5 Depressurization System Question Source l New    l Question Modification Method l Question Source Comments:  l Material Required for        {
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Examination:        :
a  .>
Page itxt
S nior R rct:r Operztor Anzwsr Ksy
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.d        28. a 29. c  j o o i
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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
)
;. d-        43. c D. a        44. b  _
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!). c        45. a 9. b        46. a-   ,
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Senior R=ct:r Operet:r Answ:r Key 1. o      26. d 2. a      27. c 3. d-      28. a 4. o      29. c 5. c'
t o S:nisr R:rctor Opertter Answsr Ksy 76. c 51. c
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'52. c   77. b 53 b. 78. d 79, a 54. e -
32. b 8. c      33. c 9. c      M. d 10. b-      35. a   .
;55. b   80. a 156. c   81. b 57. a   82. a
11. d      36. a
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'12, c     37. b 13. d      38. d 14. ' c      39. a 15. a      40. g d [[J   Ty[0 666 o m M A ,4 41 c  b O [////(8
58. c   83. a 59. - a  84. b 60. b  85. a   .
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61. c  86. d
16. c 17. a      42. a 18. d     43. c 19. a     44. b 20. c-      45. a
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62 b   87. c
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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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95, a 70. b
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22. b      47. a-23. c      48. d l
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o  )
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.


Senior Reactor Operator Answer Key 51. c  76. c 52. c  77. b 53. b  78. d 54, a  79. a 55. b  80. a 56. c  81. b G7. a  82. a 58. c  83. a 59. a  84. b 60. b  85. a .
b. SR/O.D.
61. c  86. d 62. b  87. c 63. b  88. b 64. a  89. d 65. a  90 c'
66, b  9' -
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67. d  ./ . a
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68. c  93. b
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69. c  94. c 70. b  95. a
.71 b  96. b
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72. a  97. d 73. d  98. b 74. b  99. b 75. b  100 d Page 2
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_ - _____-_-_L
10M-46. Beaver Valley Power Stat;on    UnN1
,
      *
Issue 4 Revision 2 Post-DBA Hydrogen Control System      Page A 1 of 5 Operating Procedures
. Hydrogen Recombiner Startup
~ PURPOSE This procedure describes the startup of the Post DBA Hydrogen Recombiner following the unlikely occurrence of a loss of coolant accident. This is accomplished by first Jetting 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 EO '
l        -
l- I PRECAUTIONS AND LIMITATIONS L If hydrogen concentration is it 5%, consult TSC before placing Recombiners in operatio During accident conditions, radiation levels may be high in the Recombiner are Limit the time spent in this are In order for the Hydrogen Recombiners to operate with sufficient flow, Containment -
            ..-
pressure must be controlled as close as possible to -2 psig (13 psla). Howeve g Containment pressure must remain below -2 psig (13 psia) to ensure Containment i
l
'    remains substmospheri N#
li INITIAL CONDITIONS A. The EOPs require the Hydrogen Recombiners to be placed in servic '
          .
B. The NSS has approved the performance of this procedur C. The 480 VAC distribution system is operabl D. The following procedure is available: M-46.4.G. "Placir.g Wide Range Containment Hydrogen Monitoring System in Operation".


I INSTRUCTIONS Note: Valves for the A Recombiner are given in procedure, valves for the B Recombiner are in parenthesi * Place the Hydronen Rembiner in Service
_ - _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
,
4 o 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.
    . Contact Radcon to determine what type of protective apparel is to be worn and any shielding require . Obtain the following keys to unlock [1HY-101,~ 102,103,104,110, iii,196 and 197].
a. SRI b. SR/ .


- -. - _ . _ --_ ____ _ . _ _ _ _ _ _ _ _ - - _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ - _ _ _ - _
No simulator deficiencies, that affected the scenario examinations or JPMs, were identified during the execution of the examination.
Attachment 2 l      SIMULATION FACILITY REPORT Facility Licensee: Beaver Vallev Unit 1 Facility Docket No: 50-334 Operating Tests Administered from: April 20-24,1998
! 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 used in future evaluations. No licensee action is required in response to these observation No simulator deficiencies, that affected the scenario examinations or JPMs, were identified during the execution of the examination.


!
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Revision as of 07:33, 25 January 2022

Forwards NRC Operator Licensing Exam Rept 50-334/98-300OL for Exam Administered on 980420-24 & 0518
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)
v • d • e


Text

{{#Wiki_filter:. - _ _ - _ _ _ _ _ - _ _ _ _ _ - kugo sf 17, 1 g 7[( 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 $

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9808190282 990817 PDR ADOCK 05000334 > 

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_ _ _ _ _ _ _ . _ - _ _ . _ _ _ _ _ _ _ _ _ _ - - _ _ . - - __ e gp ucoq M O UNITEo STATES 

[ ,.,, ' gg  NUCLEAR REGULATORY COMMISSION
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G C REGloN I t, -[ 475 ALLENDALE RoAo 

%- d  KING oF PRUSSIA, PENNSYLVANIA 19406 1415
*****

June 3, 1998 Mr. J. 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 g;,A2, . L 4h+oo*V 3/A _ _ - - - _ - _ - - _ - _ J

_ _ _ _ - - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _-___________ _-_- _ - ._ _ - Mr. J. ' 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

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.

l I l ii a

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        - -
        ;

I Report Details l 

        )

1. Operations ]

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 l i ) 

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.  .

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. > l 

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c ,

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- - . _ . __ _________ ________________ _

_ _ _ - _ _ - _ _ _ _ _ _ _ - - _ _ _ - _ _ _ 

. .

I l

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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       . .

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 i

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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 l 

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EXECUTIVE SUMMARY Beaver Valley Unit 1 Nuclear Power Plant Inspection Report No. 50-334/98 300 Operations

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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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- _- _ _ _ -- - ) 

      . .

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.

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.  .

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 _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ ______ _b

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    . _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _

_ 

          . .

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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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 < l l l l l 

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8 O Attachment 1 BV-1 SRO WRITTEN EXAMINATION W/ ANSWER KEY l l 

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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 

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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  ; 

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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 ___- _-___ - - 

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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 - - _ _ _ _ _ __-._________

_ _ _ _ _ _ _ _ _ - _ - _ _ - _ - - - - - - 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) 

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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

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_____-__.

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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 ' 

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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 

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i l Page 6 

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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 * - , 

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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 '

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.

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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

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

Conduct Of Operations I/2LP-SQS-48.1 V.P 7 10 15 . 

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Question Source l Facility Exam Bank l Question Modification Method l  ; l Question Source Comments: l l Mat: rial Required for l Ex mination: 

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l Page 9 l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.

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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 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ 

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Technical Specific:tions requires radittion areas to be isolated by locked doors if the radittion levels are greater than: c. 100 mrem /hr 

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b. 500 mrem /hr c. 1000 mrem /hr d. 5000 mrem /hr Ans: lc l Exam Level: lS l Cognitive Level: l Memory I 

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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 _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _

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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

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 

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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: 

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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 _ ____ ________ _ _ 

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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 

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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 l-- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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

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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 

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Giv:n the following conditions: 

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* Reactor Power - 72%         '
* Control Rods are at step 210 on Control Bank D

* 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 

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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 

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KA Knowledge of Control Rod Drive System design feature (s) and or interlock (s) which provide for the following: i 

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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 Question Source lNew l Question Modification Method l Question Source Comments: l Material Required for Examination: l Page 19 I _ __ . _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ a

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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, M:terial Required for Ermination: Page 20 L- _ _ - __-_ - .

__ __ 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

RCS - Reactor Coolant LP SQS-6.3 Ill.A 24 4 12.A Pumps

Question Source l New l Question Modification Method l I Qrestion Source Comments: l Matrial Required for Ex;mination: 

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l Page 21 u___.________ J 

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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 Question Source Comments: l Material Required for Examination: f 

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Page 22 L _ - - . . . 

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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 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 QTestion Source Comments: l Mat: rial Required for Technical Specifications Ermi:stion: l Page 23 l 

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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 w__

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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 

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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 Question Source l New l Question Modification Method l Question Source Comments: l M:terial Required for Ex mliation: I l Page 25 - _ _ _ _ _

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t ' 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:

Page 26 l - - - - - - - - - _ _ _ _ _ _ _ _ _ 

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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

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 Qrestion Source Comments: l M;terial Required for Examination.

j I I l l l l l ! Page 27 

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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

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

Examleation: I Page 28 ) l _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ 

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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 

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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

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 Question Source Comments: l Mat: rial Required for Ex :miration: ! Page 29 L ) 

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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

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 Question Source Comments: l Mr.terial Required for Eximiration: l l l

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RPS testing is in progress for RPS train B and the status of the breakers are as follows: i 

- * 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 Question Source Comments: l Material Required for Exami::stion: . 

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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 l l

l l Page 32 l L-

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

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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: ! ! Page 33 l t __ .______-____a 

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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 Question Source Comments: l Mat; rial Required for Examination: 1 

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l l Page 34 i L i 

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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 Question Source Comments: l Material Required for Examination: i.

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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.

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b. RVLIS hydraulic isolator position will indicate a leak has occurred. The leak can only be isolated 

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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 ,

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 E 

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Question Tople: 1 C---h Spray E-,-sse 12 RWST level Given the f:llswing conditi:ns: i 

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* Reactor trip, Si and CIB occurred from 100% power due to a LOCA
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* RWST level has decreased to 3 feet 9 inches
* CIB has not been reset.

What would be the status of the Quench Spray (QS) system? j 

(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: 

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! 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: 

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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 - _ _ _ - _ _ _ _ -

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

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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 

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i l I i Page 41 i 

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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

Question Source l NRC Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Ermination: Page 42 _______________ 

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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: 

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r l Page 43 _ _ _ - _ - _ _ _ _ _ _ _ 

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Question Topic: l Effect cf MS- PT-464 fdling high Given the follswing conditions: l l 

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* 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

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:

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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%
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( 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 

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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 

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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: 

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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:

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) 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 

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Expla;atio j o of Answer 

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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-

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 ,

125 V DC Control System IOM-39.1 3 

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iss 4 Rev '

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.

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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

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

GENERATOR NO. I REVERSE POWER 6 VI.13 29 Diesel Generators LP-SQS-36.2 Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Eur.mination: Page $2 _ . . - 

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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

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 Q:estion Source Comments: l M:t: rial Required for Ex:mination: l l l Page 53 l 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

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: I l Page 55 

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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 Q1estion Source Comments: l Mit; rial Required for Examination: Page 56 - _ _ - - _ - _ _ .

_ _ - _ _ _ _ - _ - _ _ _ - - ,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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i 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 I l k________ _ _ _ _

- _ _ _ _ - _ _ - - - ___- _ ____-_-__ _ ____ __- ___ - __ - ____-_-- - _-__._ __ _ ___ _ -_____-__ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ 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 Question Source Comments: l Mat: rial Required for Ex:mination: , i 

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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

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* 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: I 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

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: I Page 61 l 

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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: l Page 63 - - _ - - - _ _ _ - _ _ _ _ _ 

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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:

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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 Question Source Comments: l , M:.t: rial Required for Ex:mliation:

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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%. 

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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 

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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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l . 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 

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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 Qrestion Source Comments: l 

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M:t: rial Required for-  ! Ex:mination: i ' 

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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 e__________ __

I I 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) 

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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 I l l l l l , Page 71 L _-__-

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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: I l Page 72 _ _ _ _ _ _ _ _ - - - .__ l

_ _ _ _ _ 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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i 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]

UFSAR fig. 7.2 sheet 3 

    !!!.E.3  7  Iss 4 Rev Reactor Excore Instrument 10M-2.1.c

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

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:

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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: I J Page 77 i I L__._______________.

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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 

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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: ! l 

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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 

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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 

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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 J l l Page 80 L__________ 

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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 

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Q:estion Source - l New l Question Modification Method l Q estion Source Comments: l Mit: rial Required for 

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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 

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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

VOND 24 1 Question Source l New l Question Modification Method l Question Source Comments: l 

~ Material Required for Examination:

Page 82 _ _ _ - __ _ _ _

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

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Qrestion Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l M:,terial Required for Examination: I i i Page 83 f- 

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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

EOPs LP-SQS-53.3 Question Source l NRC Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Examination: 

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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: _ l l 

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_ _. . _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ , . . I Question Tcpic: l Effect of a loss of V;tal AC on Feedwater Given the following conditions: ) 

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* 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 Question Source j Facility Exam Bank l Question Modification Method l      j Question Source Comments: l Material Required for Er.mination:

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l l l l l Page 86 l l l

_____ - -- - - _-___-______-- _ _ _ _ - - _ _ - _ - _ - _ - _ - - - - _ _ _ - _ _ _ _ - _ . _ _ _ _ _ - - _ -- _. - _ _ _ _ - - _ _ _ _ _ _ _ - , 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 Question Source l New l Question Modification Method j j Question Source Comments: l l M:.terial Required for IOM 39.5.B.6(28 pages) Ex mination: 

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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 Question Source lNew l Question Modification Method l Question Source Comments: l Mit: rial Required for Tech Specs Exaination: Page 88 _ _ _ _ - _ _ _ -

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 Question Source l New l Question Modification Method l Q:estion Source Comments: l Meterial Required for Ex mination: i I i, t Page 89 

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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: Page 90 

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_ . - _ _ _ _ i 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 

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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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I Question Source l New l Question Modification Method l Question Source Comments: l M;tirist Required for Ex:mination: 

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Question Topic: l Controller lo' cation 

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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 Question Source l New l Question Modification Method l Question Source Comments: l M;terial Required for Ex:mination: l , l Page 92 I \ L.

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a, 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 ( , 

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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 . l Emergency Operating LP-SQS-53.3 1 3 Procedures i Question Source l Facility Exam Bank l Question Modification Methcd l Question Source Comments: l Material Required for Ex mination:

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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

Question Source l NRC Exam Bank l Question Modification Method l l Questic3 Source Comments
l M:terial Required for Ex:mirtion:

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I 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 Question Source l New l Question Modification Method l Q estion Source Comments: l M;terial Required for Examination: J 

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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 Question Source . l NRC Exam Bank l Question Modification Method l Question Source Comments: l M'.t: rial Required for Extininstion: , Page 96

F 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 i Q estion Source l New l Question Modification Method l QIestion Source Comments: l Mit: rial Required for , Ex mination:  ! 

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s 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 :

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)

EOPs LP-SQS-53.3 Q:estion Source l Facility Exam Bank - l Question Modification Method l Q:estion Source Comments: l Mat: rial Required for Ex mination: Page 98 

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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 EOPs LP-SQS-53.3 6 Question Source l New l Question Modification Method l Question Source Comments: l Mat: rial Required for j Ex:minatlon: i i l i l i l Page 99 

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s 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

Depressurization System 27 5 5 Containment LP-SQS-13.01 Depressurization System Question Source l New I Question Modification Method l Cc. ion Source Comments: l Material Required for Examination: l l l Page 1(Mi 

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, c         M. a
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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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55. b 80. a 156. c 81. b 57. a 82. a
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58. c 83. a 59. - a 84. b 60. b 85. a . 61. c 86. d 

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- _ _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ o ) Beaver Valley Power Station Unit i 10M-46.4. A Post-DBA Hydrogen Control System 

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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.

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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 

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lit. INITIAL CONDITIONS 1 

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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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_ - _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ 4 o 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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