GO2-96-176, Informs That Pilot Initial License Exam Outline to Be Administered by NRC on 961007,completed
| ML20132F549 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 09/06/1996 |
| From: | Albers J WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Brockman K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML20132F547 | List: |
| References | |
| GO2-96-176, NUDOCS 9612260011 | |
| Download: ML20132F549 (600) | |
Text
{{#Wiki_filter:_ __ ._. -._ __. . . -- _ _.- _ _- I l WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.o. Box 968
- 3000 George Washington Way
- Richland, Washington 99352-0968 = (509) 372 5000 September 6,1996 GO2-96-176 i
Docket No. 50-397 I l Mr. K. E. Brockman, Acting Director ) i Division of Reactor Safety ) U.S. NRC, Region IV i i 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 ,
Dear Mr. Brockman:
Subject:
WNP-2 OPERATING LICENSE NPF-21 PROPOSED PILOT INITIAL LICENSE EXAMINATION ; The proposed pilot initial license examination to be administered by the Nuclear Regulatory Commission on October 7,1996, has been completed. The examination is being sent via i overnight delivery to Mr. Howard Bundy at Region IV on September 6,1996, for evaluation and approval. Per Examiner Standards-201, the proposed examination is being sent to Mr. Bundy in a double envelope marked "FOR OFFICIAL USE ONLY" and "TO BE OPENED BY ADDRESSEE ONLY." WNP-2 requests that these materials be withheld from public disclosure until after the examination has been completed. Should you have any questions or desire additional information please contact W. D. Shaeffer, Superintendent, Operations Training at (509) 377-8266. Respectfully, J. . Albers (Mail Drop 1027) Nuclear Training Manager cc: T. O. McKernon, NRC/RIV J. L. Pellet, NRC/RIV L. J. Callan, NRC/RIV Document Control Desk - NRC T. G. Colburn, NRR N. S. Reynolds, Winston & Strawm NRC Sr. Resident Inspector, 927N D. L. Williams, BPA/399 9612260011 961219 PDR ADOCK 05000397 V PDR
l 1 W Facility: WNP2 ADMINISTRATIVE TOPICS - OUTLINE ES-301-1 1 Examination Level: SRO(l) (Circle, one) I 1 l Administrative Describe method of evaluation: Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Shift Manning 294001 A1.03 (3.7) Admin. requirements to read and initial the Night Order Book. 294001 A1.03 (3.7) Responsibilities of the " Reactivity Manager". Main Generator 294001 A1.08 (3.6) Simulator JPM. Excitation Curve Usage A.2 Troubleshooting 294001 Al.03 (3.7) liigh Risk troubleshooting activities. Plan / Authorization 294001 A1.03 (3.7) Authorization of troubleshooting activities. A.3 Radiation Control 29400lKl.03 (3.8) liigh Radiation Key locations. 294001 A1.16 (4.7) Life Saving/ Protect the Public Dose limits. A.4 Emergency Plan 294001 Al.16 (4.7) Classify the Event in Simulator scenario #3. 294001 A1.16 (4.7) Determine reportability requirements of Simulator l scenario #1. Examiner: Chief Examiner: JPM Cbeekrast per ES-301 j
- 1. 10 sRo(D/Ro applicants JPMs w/ 7 Contml raan and 3 int ant. i 2. 5 SRo(U) JPMs w/ 2 or 3 Control room and 2 or 3 int lant.
- 3. At least 7 different safety functions for sRo(D/Ro*s. 4. At least 5 different safety functions for SRo(U) applicants. ;
- 10. 1 Control raun JPM must be an ESP. 6. For each system selected, select I existing OR develop 1 new JPM.
- 7. At least 1 JPM related to shutdown or km power condition. 8. I or 2 JPMs require 'altemate paths".
- 9. At least 1 *in plant" JPM requires EoP or Abnormal actions. 10. At least 1 *in plant" JPM requirca escort into rad. controlled area.
- 11.
- Diversify" the prescripted questions among the Ka, As. and Os, 12. less than 30% overlap from last NRC Exam. (
- 13. At least 2 NEW or significantly altered JPMs for sRo(D/Ro's. 14. At least i NEW or sigmficantly altered JPM for SRo64
- 15. Admmistrative topics should be evaluated in JPMs whenever possible.
rather than prescripted questions. 41
ADMINISTRATIVE OUEST10NS/ ANSWER GUIDE - SRO(1) NRC Exam: October,1996 QUESTION 1: TOPIC /
SUBJECT:
A.1 Shift Manning (Open reference) NEW - 9/1/96 QUESTION: A review of the Night Order Book shows that it has been read and initialed by the Shift Manager and Control Room Supervisor. What other individuals, if any, are required to read and initial the Night Order Book? ANSWER: The Shift Support Supervisor. KA: 294001 A1.03 (3.7)
Reference:
PPM 1.3.1 Rev.26 Page 41 of 86 QUESTION 2: TOPIC /
SUBJECT:
A.1 Shift Manning (Open reference) Sienificant/v Altered 8/4/M QUESTION: During a reactor startup, who by title assumes the responsibility of Reactivity Manager? Describe the responsibilities included in assuming this position. ANSWER: The Control Room Supervisor. Responsibilities include ensuring a conservative approach to all operations involving core reactivity changes. KA: 294001 A1.03 (3.7)
Reference:
PPM 1.3.1 Rev.26 Page 18 of 86 l i QUESTION 3: TOPIC /
SUBJECT:
A.2 Troubleshooting Plan / Authorization (Open reference) Sienificantiv Altered - 8/4/% l QUESTION: The plant is operating at = 40% power. An I&C Technician requests permission to troubleshoot DEH computer inputs. Because the activity has been determined to have a significant chance of producing a Main Turbine trip, a written troubleshooting plan has been developed. What authorization is required for the written troubleshooting plan? ANSWER: The Shift Manager is responsible for obtaining concurrence from the Operations Manager to approve the activity. K/A: 294001 A1.03 (3.7)
Reference:
PPM 1.3.42 Sect. 5.5 Rev.13 Page 5 of 11
.- - - _ - - _ . . ~ . . , ~ . - - _ . . - . . . - . . . - . - - . . . - _ . . _ . - _ - - - . - - . . . - . -
P ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(1) NRC Exam: October,1996 QUESTION 4: TOPIC /
SUBJECT:
A.2 Troubleshooting Plan / Authorization (Open reference) QUESTION: What individual, by title, is responsible for authorizing ALL troubleshooting activities? ANSWER: The CRS/ Shift Manager. K/A: 294001 A1.03 (3.7)
Reference:
PPM 1.3.42 Sect. 5.6, Rev.13 Page 5 of 11 QUESTION 5: TOPIC /
SUBJECT:
A.2 (Open reference) NEW - 9///96 l l QUESTION: During an emergency condition, the Shift Manager has authorized release of keys for vital and protected areas. Where would you go (location) to get these keys? ANSWER: Central Alarm Station (CAS) and/or Secondary Alarm Station (SAS). l K/A: 294001 A1.03 (3.8)
Reference:
PPM 1.7.1, Rev.15, Section 2.3.3, page 6 of 13 QUESTION 6: TOPIC /
SUBJECT:
A.2 (Open reference) NEW - 9/1/96 l QUESTION: An operator,27 years of age, has received a total lifetime exposure of 10 rem total effective dose equivalent (TEDE). 'Due to an emergency for a life-saving situation, how much additional exposure could this individual receive without exceeding the Emergency Exposure Guides? ANSWER: 25 rem TEDE. l K/A: 294001 A1.16 (4.7) i
Reference:
PPM 1.I1.16, Rev. 2, Section 7.3, page 10 of 14
- i l
ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(1) NRC Exam: October,1996 l QUESTION 7: TOPIC /
SUBJECT:
A.4 EMERGENCY PLAN Follow-up Question - Simulator Scenario #3. QUESTION: What emergency classification would be required for this event and what is its basis? ANSWER: Site Area. 2.2.S.1 KA: 294001 A1.16 (4.7) f
Reference:
PPM 13.1.1 Attachment 5.2, Rev. 23 i I i QUESTION 8: TOPIC /
SUBJECT:
A.4 EMERGENCY PLAN Follow-up Question - Simulator Scenario #1 f I QUESTION: What event, if any, occurred during this scenario that would be " reportable"? . i If so, what agency (ies) are required to be notified and what , if any, time [ requirements apply? ; i (If multiple reportable events occurred, which event requires the earliest notification?)
]
l ANSWER: Declaration of Emergency classification (Site Area Emergency)
- State & Local agencies. - 15 min.
- Nuclear Regulatory Commission - I hr.
KA: 294001 A1.16 (4.7)
Reference:
PPM 1.10.1 rev 16, pg 16 i .2 r-- - 4 m i_,. * - , i
ADMINISTRA TIVE OUESTIONS/ ANSWER GUIDE - SRO(1) NRC Exam: October,1996 (Open reference question) QUESTION 1: A review of the Night Order Book shows that it has been read and initialed by the Shift Manager and Control Room Supervisor. What other individuals, if any, are required to read and initial the Night Order Book? i l a r ----- m -- . - -- - 7
- - - .-.. - - -. . _ . - _ .- . ~ - . - -
f ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(D 1 1 NRC Exam: October,1996 (Open reference question) l i j QUESTION 2: During a reactor startup, who by title assumes the responsibility of Reactivity l l Manager? I t l Describe the responsibilities included in assuming this position. , 1 i l l i 1 1 1
., g - ,- . - . - . . . . - -
l ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(I) l l NRC Exam: October,1996 (Open reference question) i QUESTION 3: - The plant is operating at = 40% power. An I&C Technician requests permission to troubleshoot DEH computer inputs. Because the activity has been determined to have a significant chance of producing a Main Turbine trip, a written troubleshooting plan has been developed. What authorization is required for the written troubleshooting plan? i i i
~ ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(l) NRC Exam: October,1996 (Open reference question) QUESTION 4: What individual, by title, is responsible for authorizing ALL troubleshooting activities?
_ - _ _ _. - - . .. . _ - . . . . - . - . . - . - ~ ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(1) NRC Exam: October,1996 (Open reference question) QUESTION 5: During an emergency condition, the Shift Manager has authorized release of keys for vital and protected areas. Where would you go (location) to get these keys? i i l l l l l i i l j l 1 l i
j ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(I) ] NRC Exam: October,1996 ! (Open reference question) d
- QUESTION 6: An operator,27 years of age, has received a total lifetime exposure of 10 rem total effective dose equivalent (TEDE). Due to an emergency for a life-saving situation, j how much additional exposure could this individual receive without exceeding the Emergency Exposure Guides?
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ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(D i NRC Exam: October,1996 ! (Simulator Scenario #3 follow-up question) ! r i t l QUESTION 7: What emergency classification would be required for this event and what is its basis? ! i i l i i i I r 9 t I I i I t t t
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_ . . . _ _ . _ _ . _ . _ _ . _ . . . . _ _ _ _ . _ . _ . _ _ . _ . _ . _ . __..____..._m ___ . _ . _ . . . _ . . _ t ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(1) NRC Exam: October,1996 (Simulator Scenario #1 follow-up question) QUESTION 8: What event, if any, occurred during this scenario that would be " reportable"? If so, what agency (ies) are required to be notified and what , if any, time l requirements apply? (If multiple reportable events occurred, which event requires the earliest notification?)
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i l , Ibility: WNP2 ADMINISTRATIVE TOPICS - OUTLINE ES-3011 Examination Level: SRO(U) (Circle one) Administrative Describe method of evaluation: 1 Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Shift Manning 294001 A1.03 (3.7) Admin. minimum staffing requirements during Refueling.
- 294001 A1.03 (3.7) Admin. Requirement for Shift Managers presence in the Control Room.
Power / Flow Map Usage 294001 A1.08 (3.6) Simulator JPM. l A.2 Troubleshooting 294001 A1.03 (3.7) High Risk troubleshooting activities. I Plan / Authorization 4 294001 A1.03 (3.7) Authorization of troubleshooting activities. j A.3 Radiation Control 294001 A1.16 (4.7) Admin. exposure " Hold Points". ' 294001 A1.16 (4.7) Life Saving/ Protect the Public Dose limits. A.4 Emergency Plan 294001 Al.16 (4.7) Classify the Event in Simulator scenario #3. 294001 A1.16 (4.7) Determine reportability requirements of Simulator scenario #1. 1 I Examiner: Chief Examiner: i I l i JPM Checklist per ES-301
- 1. 10 SRO(D/Ro applicants IPMs w/ 7 Cmtrol room and 3 in-plant. 2. 5 SRo(U) JPMs w/ 2 or 3 Control room and 2 or 3 in-plant.
- 3. At least 7 different safety functions for SRo(!)/Ro's. 4. At kast 5 different safety functions for SRO(U) applicants.
- 9. 1 Control room JPM must be an ESF. 6. For each system selected, select I exis6ng OR develop i new JPM,
- 7. At least 1 JPM related to shutdown or low power condition. 8. I or 2 JPMs require
- alternate paths *.
- 9. At least i "in plant
- JPM requires EOP or Abnormal acnons. 10. At least 1 *in plant JPM requires escort into rad. controlled area.
- 11.
- Diversify" the prescripted questens among the Ks. As, and Gs. 12. less than 30% overlap from last NRC Exam.
- 13. At least 2 NEW or significantly at.cred JPMs for SRO(D/Ro's. 14. At least i NEW or significantly ahered JPM for SRo(U).
- 15. Administrauve topics should be evaluated in JPM: whenever possible, rather than prescripted ques 6oiu.
40
i i
-ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U)
NRC Exam: October,1996 I QUESTION 1: TOPIC /
SUBJECT:
A.1 Shift Manning (Open reference) NEW - 9/l/M d i QUESTION: The plant is in OPERATIONAL CONDITION 5, refueling operations are in progress. j What is the Administrative minimum staffing requirement for CRO's in this condition? ANSWER: Two (2). KA: 294001 A1.03 (3.7) l
Reference:
PPM 1.3.1 Rev.26 Page 37 of 86 QUESTION 2: TOPIC /
SUBJECT:
A.1 Shift Manning (Open reference) Sienincantly Altered 8/4/M i i
- QUESTION
- The Plant is operating at 100% power, steady state, when a localized problem in the
- Plant requires the Shift Manager's presence.
1 Concerning the Control Room Command Function of the Shift Managers' job, what restriction, if any, is there on the Shift Manager leaving the Control Room? j ANSWER: None, the CRS is normally designated / delegated as having the Control Room i Command Function. i i KA: 294001 A1.03 (3.7)
Reference:
PPM 1.3.1 Rev.26 Page 38 of 86 { i } ! QUESTION 3: TOPIC /
SUBJECT:
A.2 Troubleshooting Plan / Authorization (Open reference) L Sienincantiv Altered - 8/4/96 i 4 i QUESTION: Who is responsible for approving "High Risk" troubleshooting activities? ! ANSWER: CRS/ Shift Manager after obtaining concurrence from the Operations Manager. i l K/A: 294001 A1.03 (3.7)
Reference:
PPM 1.3.42 Sect. 5.5 Rev.13 Page 5 of 11 {. i l- l l 1 l 4 i
-- - - . ~ - - - . . - . - - - - _ . - - . - _ - - . . - - - _ ~ _
ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U) NRC Exam: October,1996 QUESTION 4: TOPIC /
SUBJECT:
A.2 Troubleshooting Plan / Authorization (Open reference) QUESTION: The Plant is operating at 100% power, steady state. An I&C Tech. has made a request to troubleshoot DEH computer inputs. Who is responsible to ensure that the troubleshooting activities are described in the written plan and that the activities stay within this plan? ANSWER: The Work Supervisor. K/A: 294001 A1.03 (3.7)
Reference:
PPM 1.3.42 Sect. 5.3 Rev.13 Page 5 of 11 QUESTION 5: TOPIC /
SUBJECT:
A.2 (Open reference) NEW - 9///96 QUESTION: What two (2) conditions must be met prior to an individual exceeding a WNP-2 administrative exposure " hold point"? ANSWER: 1) An Increased Exposure Request shall be initiated by the worker's supervisor and completed in accordance with PPM 1.11.16. l i 2) Ensure at this time that the individual has either not received or has accounted ! for non WNP-2 occupational radiation exposure. l K/A: 294001 A1.16 (4.7)
Reference:
PPM 1.11.16, Rev. 2, Section 7.2.4, page 10 of 14 i QUESTION 6: TOPIC /
SUBJECT:
A.2 (Open reference) NEW - 9///M l QUESTION: An operator, 24 years of age, has received a total lifetime exposure of 6 rem total i effective dose equivalent (TEDE). Due to an emergency to protect the public health, l how much additional exposure could this individual receive without exceeding the Emergency Exposure Guides? 1 ANSWER: 10 rem TEDE. ) l ; ! 1 i 1 ! K/A: 294001 A1.16 (4.7) l
Reference:
PPM 1.11.16, Rev. 2, Section 7.3, page 10 of 14
i l ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U) l t NRC Exam: October,1996 : i QUESTION 7: TOPIC /
SUBJECT:
A.4 EMERGENCY PLAN Follow-up Question - Simulator Scenario #3. l QUESTION: What emergency classification would be required for this event and what is its basis? l ANSWER: Site Area. ' 2.2.S.1 l KA: 294001 A1.16 (4.7)
Reference:
PPM 13.1.1 Attachment 5.2, Rev. 23 , i l I l QUESTION 8: TOPIC /
SUBJECT:
A.4 EMERGENCY PLAN Follow-up Question - Simulator Scenario #1 QUESTION: What event, if any, occurred during this scenario that would be " reportable"? i If so, what agency (ies) are required to be notified and .what , if any, time requirements apply? (If multiple reportable events occurred, which event requires the earliest notification?) ! 1 l ANSWER: Declaration of Emergency classification (Site Area Emergency) l
- State & Local agencies. - 15 min.
- Nuclear Regulatory Commission -I hr.
KA: 294001 A1.16 (4.7)
Reference:
PPM 1.10.1 rev 16, pg 16 i i
l,
. ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U) l j NRC Exam: October,1996 1 (Open reference question) 1 i
QUESTION 1: The plant is in OPERATIONAL CONDITION 5, refueling operations are in progress.
. What is the Administrative minimum staffing requirement for CRO's in this condition?
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- ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U) 1 l l'
NRC Exam: October,1996 (Open reference question) QUESTION 2: The Plant is operating at 100% power, steady state, when a localized problem in the ; Plant requires the Shift Manager's presence. Concerning the Control Room Command Function of the Shift Managers' job, what restriction, if any, is there on the Shift Manager leaving the Control Room? ! l l i i I f 1 ' i I 1 l l l l
.. . _ . . . _ . . _ . . - _ . . . . _ , . _.. , _ , _ m. .- _ ,. , ,
I ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U) NRC Exam: October,1996 (Open reference question) QUESTION 3: Who is responsible for approving "High Risk" troubleshooting activities? l I
ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U) NRC Exam: October,1996 (Open reference question) QUESTION 4: The Plant is operating at 100% power, steady state. An I&C Tech. has inade a request to troubleshoot DEH computer inputs. Who is responsible to ensure that the troubleshooting activities are described in the written plan and that the activities stay within this plan? l i l
ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U) NRC Exam: October,1996 (Open reference question) QUESTION 5: What two (2) conditions must be met prior to an individual exceeding a WNP-2 administrative exposure " hold point"? l l
. _ . _ ._..__.....__.._____.._.___.__.._.._______.._._m. __.___
i ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U) j NRC Exam: October,1996 1
- 1 (Open reference question) j QUESTION 6
- An operator, 24 years of age, has received a total lifetime exposure of 6 rem total j effective dose equivalent (TEDE). Due to an emergency to protect the public health,
; how much additional exposure could this individual receive without exceeding the j Emergency Exposure Guides?
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&DMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U)
NRC Exam: October,1996 i i (Simulator Scenario #3 follow-up question) QUESTION 7: What emergency classification would be required for this event and what is its basis?
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. ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - SRO(U)
NRC Exam: October,1996 (Simulator Scenario #1 follow-up question) QUESTION 8: What event, if any, occurred during this scenario that would be " reportable"? If so, what agency (ies) are required to be notified and what , if any, time requirements apply? (If multiple reportable events occurred, which event requires the earliest notification?) I I l l
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.I f$cility: WNP2 ADMINISTRATIVE TOPICS - OUTLINE ES-301-1 l l I i
a Examination Level: RO (Circle one) I 1 Administrative Describe method of evaluation: Topic / Subject 1. ONE Administrative JPM OR Description 2. TWO Administrative Questions A.1 C.R. Indications 294001 A1.13 (4.5) Responsibility during Control Board walkdowns. Fire Bigade Composition 29400lKl.16 (3.5) Requirements for Fire Brigade members. Main Generator 294001 Al.08 (3.6) Simulator JPM. Excitation Curve Usage A.2 Clearance Order 29400lKl.02 (3.9) Admin. requirements of Independent Verification. Verification 294001Kl.02 (3.9) Requirements for Second Verification. A.3 Radiation Control 29400lKl.03 (3.8) High Radiation Key locations. 294001 A1.16 (4.7) Life Saving/ Protect the Public Dose limits. A.4 Emergency Plan 294001 A l.16 (2.9) Who may assume the duties of the Emergency Director. 294001 A1.16 (2.9) Responsibility of the Emergency Director. Examiner: Chief Examiner: Jl'M CberLiist per 13-301
- 8. 10 SRo(I)/Ro applicants JPMs w/ 7 cmtrol rawn and 3 m-plant. 2. 5 SRo(U) JPMs w/ 2 or 3 Cetrol room and 2 or 3 in-plant.
- 3. At least 7 different safety functions for SRo(D/Ro's 4. At least 5 different safety func6ms for SRo(U) applicants.
II. 1 Cmtrol romn JPM must be an ESP. 6. For each system selected. scicct i existmg oR develop i new JPM.
- 7. At least i JPM related to shutdown or low power condition. 8. I or 2 JPMs require *ahernate paths".
9, At least 1 *in plant
- JPM requires !!or or Atmormal actums. 10. At least I *in plant" JPM requires escort into rad. cetrolled area.
II. " Diversify
- the prescripted questions among the Ks. As. and os. 12. 12ss than 30% overlap from last NRC Exam.
- 13. At least 2 NEW or significantly altered JPMs for $Ro(!)/Ro's. 14. At least 1 NEW or sigmfwantly altered JPM for SRo(U) l$. Admmistrative topics should be evaluated in JPMt whenever possible.
rather Gum presenpted questims. l l l 42 J
I ADMINISTRATIVE OUEST10NS/ ANSWER GUIDE - RO NRC Exam: October,1996 QUESTION 1: TOPIC /
SUBJECT:
A.1 Shift Manning (Open reference) NEW - 9/1/96 1 QUESTION: During shift turnover and control panel walkdown, the Lead CRO informs you that l I&C is in the middle of performing an RPS surveillance that has been producing half- l scrams. I&C has been put on hold until shift turnover has been complete. What are your responsibilities during this turnover given the above situation? ANSWER: Review and understand the implications of the surveillance prior to assuming the shift. KA: 294001 A1.13 (4.5)
Reference:
PPM 1.3.1 Rev.26, Section 4.20, Page 57 of 86 l l l QUESTION 2: TOPIC /
SUBJECT:
A.1 Shift Manning (Open reference) Sieni/icant/v Altered 8/4/M QUESTION: Per the administrative limits, what requirements exist for Fire Brigade Team individuals? ANSWER: In addition to the Fire Brigade Leader, at least one (1) of these individuals will be an EO and two (2) of these individuals must be First Aid qualified. KA: 294001K1.16 (3.5)
Reference:
PPM 1.3.1, Section 4.13, Rev.26 Page 37 of 86 QUESTION 3: TOPIC /
SUBJECT:
A.2 Troubleshooting Plan / Authorization (Open reference) Sinnificantiv Altered - 8/4/96 QUESTION: Briefly explain the administrative requirements for completion of the " independently verified by" blocks in any PPM. ANSWER: This verification must be completed by a second qualified operator who does not accompany the first operator . He must independently identify the correct component and verify the position of the component as left by operator one. K/A: 294001KA1.02 (3.9)
Reference:
PPM 1.3.8
[ 6DMINISTRATIVE OUESTIONS/ ANSWER GUIDE - RO ) i NRC Exam: October,1996 l t l TOPIC /
SUBJECT:
A.2 Troubleshooting Plan / Authorization (Open reference) QUESTION 4: QUESTION: List three (3) of the alternate methods of performing a second verification, as applied , to independent verification. l l ANSWER: - Use of remote position indicators. i j - Use of process parameters (flow, pressure, current, voltage, etc.) j ! - Observation of the valve stem to aid in the determination of valve position. ; Authorized scribe marks on valve stems. ! j Functional mechanical position indicators. I t , K/A: 294001K1.02 (3.9) l' l
Reference:
PPM 1.3.12 Sect. 4.11.4, Rev.26 Page 34 of 86 I i QUESTION 5: TOPIC /
SUBJECT:
A.2 (Open reference) NEW - 9/1/96 l t I i QUESTION: During an emergency condition, the Shift Manager has authorized release of keys for vital and protected areas. Where would you go (location) to get these keys? f ANSWER: Central Alarm Station (CAS) and/or Secondary Alarm Station (SAS). K/A: 294001 A1.03 (3.8)
Reference:
PPM 1.7.1, Rev.15, Section 2.3.3, page 6 of 13 QUESTION 6: TOPIC /
SUBJECT:
A.2 (Open reference) NEW - 9/1/M QUESTION: An operator, 27 years of age, has received a total lifetime exposure of 10 rem total effective dose equivalent (TEDE). Due to an emergency for a life-saving situation, how much additional exposure could this individual receive without exceeding the Emergency Exposure Guides? ANSWER: 25 rem TEDE. ,
)
i I l K/A: 294001 A1.16 (4.7) ! i
Reference:
PPM 1.11.16, Rev. 2, Section 7.3, page 10 of 14
e
- . ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - RO i
NRC Exam: October,1996 QUESTION 7: TOPIC /
SUBJECT:
A.4 EMERGENCY PLAN QUESTION: When would the Control Room Supervisor (CRS) assume the duties of the Emergency l Director (ED)? ANSWER: When the Shift Manager (SM) is not in the control room or is incapable of performing , the duties of the Emergency Director. l l KA: 294001 A1.16 (2.9)
Reference:
PPM 13.1.1 , i QUESTION 8: TOPIC /
SUBJECT:
A.4 EMERGENCY PLAN ) l QUESTION: Who initially assumes the responsibility of the Emergency Director? ! l l ANSWER: The Shift Manager l KA: 294001 A1.16 (2.9)
Reference:
PPM 13.1.1 i
r
. ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - RO l
! NRC Exam: October,1996 (Open reference question) 1 h QUESTION 1: During shift turnover and control panel walkdown, the Lead CRO informs you that ! I&C is in the middle of performing an RPS surveillance that has been producing half-
- scrams. I&C has been put on hold until shift turnover has been complete. .
l What are your responsibilities during this turnover given the above situation? I, I a l
- 1 i l l
l I i l l 4 i 4
ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - RO NRC Exam: October,1996 (Open reference question) QUESTION 2: Per the administrative limits, what requirements exist for Fire Brigade Team individuals?
. ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - RO l NRC Exam: October,1996 I l (Open reference question) QUESTION 3: Briefly explain the administrative requirements for completion of the " independently verified by" blocks in any PPM. j i 1 i l I
i ? i . ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - RO i NRC Exam: October,1996 ; (Open reference question) ; ,4 i ; QUESTION 4: List three (3) of the alternate methods of performing a second verification, as applied
- to independent verification.
s e i t i i i 4 i k i a I 1 1 j k i i 2 1 4 i l i i a i i i 1
i l ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - RO NRC Exam: October,1996 : (Open reference question) QUESTION 5: During an emergency condition, the Shift Manager has authorized release of keys for vital and protected areas. : 3 Where would you go (location) to get these keys? l t f i s l I l 6 I i l f i l I i i
. ADMINISTRA71VE OUESTIONS/ ANSWER GUIDE - RO l
NRC Exam: October,1996 l (Open reference question) QUESTION 6: An operator,27 years of age, has received a total lifetime exposure of 10 rem total [ effective dose equivalent (TEDE). Due to an emergency for a life-saving situation, how much additional esposure could this individual receive without exceeding the ! Emergency Exposure Cuides? 6 i t B l i i i d k i k
. ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - RO NRC Exam: October,1996 (Open reference question) QUESTION 7: When would the Control Room Supervisor (CRS) assume the duties of the Emergency Director (ED)?
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I l s ADMINISTRATIVE OUESTIONS/ ANSWER GUIDE - RO l NRC Exam: October,1996 l (Open reference question) l I t QUESTION 8: Who initially assumes the responsibility of the Emergency Director 7 t i i I I h 1 , l l I , t t i I r i i t s 5 l 1 1 l l 1 l
ES-301 SCENARIO EVENTS Form ES-301-3 Simulator Facility: WNP-2 Scenario No. 1 File # Drill _961HLC.pgp Examiners: Candidates: Initial Conditions: Initialize IC 14, Plant operating at = 100% power. Turnover: The Plant has been operating at = 100% power for three months. No major evolutions are planned for this shift and no major equipment is out of service at this time. The Plant is currently in full compliance with all Technical Specifications and Regulatory requirements. Event No. Malf. No. Type
- Description 1 CLF - RLY I Blown Fuse in RC-1, 27AX relay. Simulated by de-NSF25 energizing relay NSF 25, 2 NIS5E I APRM "E" fails low.
3 RRS3B M Small steam leak in the Drywell. 4 RRS9B M Steam Leak triggered from Mode Switch out of "Run". (This elevates the steam leak to accelerate entry into PPM 5.2.1 and challenge PSP.) 5 N/A N Initiate Wet Well/Drywell Sprays as required. 6 N/A C/M MSIV Isolation on high temperature in the Steam Tunnel.
- Normal (N), Reactivity manipulation (R), Instrument malfunction (I), Component malfunction (C), Major transient (M)
Review Complete:
ES-301 OPERATOR ACTIONS ES-301-4 Scenario No.1 Page 2 Event No.1 l Brief Event
Description:
Blown Fuse in RC-1,27AX relay. Time Position Candidates Actions / Behavior CRO3 Receives, acknowledges and announces "RC-1 IAss of Power". l (Annunciator Panel P800-C1, 6-2) l 1 CRS Directs CRO3 to pull ARP, check RC-1 cabinet relays and report back ! what is found. l CRO3 Pull ARP, check RC-1 cabinet and finds Relay 27AX de-energized. Reports findings to the CRS. CRS Wiifies no Tech. Spec. LCO conditions with this problem. CRO3 Uses Electrical Drawing to verify operability of RC-1. CROI Continues to monitor Plant conditions. ; 1 Event No. 2 Brief Event
Description:
APRM "E" fails low. I l Time Position Candidates Actions / Behavior CROI Receives, acknowledges and announces "APRM DOWNSCALE". (Annunciator Panel P603. A8, 4-6) CROI Determine which channel is alarming by noting which power range channel has failed downscale, or by checking the alarm lights at H13-P608 or H13-P603. CRS Refer to Tech. Specs. 3.3.1 and 3.3.6 and consider bypassing the affected channel. C"5 Direct APRM "E" be bypassed. CRO1 Bypass APRM "E". CRO3 Continues to monitor Plant conditions.
E'S-301 OPERATOR ACTIONS ES-301-4 j Scenario No. _L Page 3 : 1 Event No. 3_ 1 Brief Event
Description:
Small steam leak in the Drywell. Time Position Candidates Actions / Behavior l CRO3 Possible recognition of Drywell pressure increase. CROI Receives, acknowledges and announces "DRYWELL PRESS. HI/ LOW ] ALERT". (Annunciator Panel P603. A7,5-3)
]
CRO3 Determine if Drywell pressure is high or low on CMS-PR-1 (H13-P601) and CMS-PR-2 (H13-P601). (Assuming Drywellpressure increase was not caught until the Alann was received) CRS Direct CRO1 to pull ARP. Refers to Tech. Spec. 3.6.1.6 when time permits. (Rate ofDrywell pressure increase will be critical) CRO1/ Continue to monitor Plant conditions. CRO3 Event No. _4. Brief Event
Description:
Steam leak in the Drywell gets larger when the Mode Switch is placed in the "Run" position. Time Position Candidates Actions / Behavior CRO3 Reports Drywell pressure increasing level and rate and monitors GDS. (PSP will be challenged by this scenario) CRS Enters PPM 5.1.1 and PPM 5.2.1 concurrently. CROI Continues to monitor Plaat conditions.
I!S-301 ' OPERATOR ACTIONS ES-301-4 Scenario No. _1_ Page 4 Event No. _5. Brief Event
Description:
Initiate Wet Well/Drywell Sprays when required. Time Position Candidates Actions / Behavior CRS Using PPM 5.2.1, briefs Crew on requirements and restrictions for the use of Drywell and Wetwell Sprays.
~
1 CRO3 Initiate Drywell and/or Wetwell Sprays as directed. CROI Continues to monitor Plant conditions. Event No. _61 Brief Event
Description:
MSIV Isolation caused by high temperature in the Steam Tunnel. Time Position Candidates Actions / Behavior . l CRO3 Receives, acknowledges and announces " LEAK DET. MSL TUNNEL AT HIGH". (Annunciator Panel P601. A3,3-8). (NOTE: This is the high alarm setpoint, which is set 15*F LT the Hi-HI (Isolation) setpoint). CRS Directs CRO3 to follow instructions in ARP. Directs CRO3 and CROI to inform him if or as soon as the MSIV's close. CRO3 Receives, acknowledges and announces " LEAK DET. MSL TUNNEL HIGH-HIGH" (Annunciator Panel P601. A3,1-7) and " LEAK DET. MSL TUNNEL HIGH-HIGH" (Annunciator Panel P601.A2,3-1). CRO3 Verifies MSIV's close, establish RPV/P control using the SRV's. CROI Establish RPV/L control with the Condensate Booster Pumps and/or RCIC. CRS Direct and/or verify RPV/P and RPV/L control being maintained. ALL Continue to monitor Plant conditions. IMPORTANT SCENARIO NOTES: 1.
i ESo301 SCENARIO EVENTS Form ES-301-3 i i Simulator Facility: WNP-2 Scenario No. __2_ File # Drill 962HLC.pgp ] 5 l 1 Examiners: Candidates: < l l Initial Conditions: Initialize IC 118, Plant operating at = 65% power. i l Tumover: The Plant is operating at = 65% power. BPA has requested WNP-2 to reduce power to 35% for the next 36 hours for load following reasons. RCIC has been out of service for 12 hours for repair of a small steam leak and is not expected to be return to service this shift. I&C has requested removal of RFP "B" from service as soon as possible for governor testing. This shift will continue the downpower per PPM 3.2.1' step 4.1.13, rod step 76-5. Event No. Malf. No. Type
- Description
'l N/A N/R Continue Plant Shutdown per PPM 3.2.1 step 4.1.3, Rod step 76-5 per BPA request.
2 N/A N Remove the "B" RFP from service per PPM 2.2.4 Sect. 5.10. 3 PMP pSS4 C LPCS-P-2 Shaft shears. 4 MAL 'R1 C RHR Loop "A" suction line break, unisolable. 5 MOVt. 6 I Due to failed Torque switches HPCS-V-23 fails to open when required and cannot be opened from the Control Room switch. 6 MAL CRD7A M Manual scram with Hydraulic ATWS active. , l MAL CRD7B
- Normal (N), Reactivity manipulation (R), Instrument malfunction (I), Component malfunction (C), Major transient (M)
Review Complete:
l i
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ES 301 OPERATOR ACTIONS ES-301-4 Scenario No.1 Page 2 4 i Event No.1 Brief Event
Description:
Continue Plant Shutdown per PPM 3.2.1 step 4.1.3, Rod step 76-5 per BPA request. Time Position Candidates Actions / Behavior CRS Directs CROI to verify Control Rod pull step and current Plant condition and then continue with the Shutdown per PPM 3.2.1 step ! I j 4.1.13, rod step 76-5. ! CROI Verifies Plant Status as directed, then continues with the Shutdown per PPM 3.2.1 step 4.1.13, rod step 76-5.
- i
} ALL Continue to monitor Plant conditions. 3 i i Event No.1 1 Brief Event
Description:
Remove the "B" RFP from service per PPM 2.2.4 Section 5.10. Time Position Candidates Actions / Behavior CRS Directs RFP "B" to be removed per PPM 2.2.4 Section 5.10 and directs
- check of Plant conditions prior to removal.
CRO3 Verifies Plant conditions and Procedural requirements, then removes i RFP "B" per PPM 2.2.4 Section 5.10. ALL Continue to monitor Plant conditions and perform Plant Shutdown. Event No.1 Brief Event
Description:
LPCS-P-2 Shaft shears. l Time Position Candidates Actions / Behavior CRO3 Receives, acknowledges and announces "LPCS PUMP DISCH. PRESS. HIGH/ LOW". (Annunciator Panel P601.A3,5-3) CRS Direct CROI to pull ARP. Refers to Tech. Specs. 3.4.3.2, 3.5.1, and 3.6.3. CRO3 Inhibits LPCS-P-1 by holding its control switch in "STOP" until the control power fuses are removed, then checks the status of LPCS-P-2. CRO3 Direct start of RHR-A-1 in full flow test to maintain availability. ALL Continue to monitor Plant conditions.
ES-301 OPERATOR ACTIONS ES-301-4 Scenario No. 2 Page 3 Event No. _4. Brief Event
Description:
RHR loop "A" suction line break, unisolable. Time Position Candidates Actions / Behavior CRO3 Receives, acknowledges and announces " REACTOR BLDG. FLOOR-SUMP R2 HI-HI". (Annunciator Panel P602.A13,1-1)
-CRS Enters PPM 5.3.1 and directs CRO3 to pull the ARP and PPM 4.12.4.10, Reactor Bldg. Area Flooding.
CRS Enter PPM 5.2.1 due to lowering wetwell level. Direct action to raise wetwell level per PPM 5.5.27 CROI Continues to monitor Plant conditions. Event No. .L Brief Event
Description:
LPCS-V-5 fails to open. Time Position Candidates Actions / Behavior CRO3 Recognize and report failure of HPCS-V-23 to open. CRS Directs OPS-2 to manually valve HPCS-V-23 locally. ALL Continue to monitor Plant conditions.
ES-301 OPERATOR ACTIONS ES-301-4 Scenario No. _2_ Page 4 Event No. 4_ Brief Event
Description:
Manual scram with Hydraulic ATWS active. Time Position Candidates Actions / Behavior CRS Using PPM 5.2.1, briefs Crew on requirements to maintain HCLL. When it is determined that wetwell level cannot be maintained above HCLL, enter PPM 5.1.1 and direct a manual scram. CROI Initiates manual Scram as directed, recognizes and reports failure of the , l scram to complete Rod insertion. l I CRS Exits PPM 5.1.1 and enters PPM 5.1.2. When conditions are met for emergency depressurization directs CRO to open seven SRV's. CRO3 Opens seven ADS SRV's. ALL Continues to control plant with PPM 5.1.2 and 5.2.1. ( l IMPORTANT SCENARIO NOTES:
- 1. Scenario is completed when RPV level is stabilized after E/D and control rods are being l inserted as directed by PPM 5.5.10 and PPM 5.5.11.
t l I t t l i l l I i s 1 l
. 6S-301 SCENARIO EVENTS Form ES-301-3 Simulator Facility: WNP-2 Scenario No. 3 File # Drill 963HLC.pgp Examiners: Candidates:
Initial Conditions: Initialize IC 115, Plant operating at = 24% power. Turnover: The Plant is continuing a Start-up at = 24% power. PPM 3.1.2 is completed through section 4.8, control rod sequence "A2", page 34, RWM group / step 34-5, control rod 18-51. Maintenance is currently repairing SLC-P-1B, which is out of service, and is not expected to be returned to service this shift. Event No. Malf. No. Type
- Description 1 N/A N/R Continue Plant Startup per PPM 3.1.2 section 4.9, control rod sequence "A2", page 34, RWM group / step 34-5, control rod 18-51.
2 XMT RRS106 1 RFW-LI-606A, RPV/L narrow range monitor, fails low. 3 ANN EPS35 C Reactor recirculation pump "A" trips. BKR EPS47 1 4 PLP CRD1 C Loss of Control Air header pressure. PMP CAS2 PMP CAS3 MAL CAS2C 5 MAL RMC4A M Control rods drift in and MSIV's drift closed. Reactor MAL RMC4B scram and entry into PPM 5.1.1. SWI RRS6C SWI RRS17C SWI RRS27C SWI RRS36C 6 RLY RPS25 M Electric ATWS, entry into PPM 5.1.2. THROUGH RLY RPS32 7 SNV CRDill5 I ARI fail to initiate rod insertion. l THROUGH . SNV CRDll22 8 AOV CFW29 C/l RFW-V-10A & 10B fail closed. ! AOV CFW28 l 1
- h:rmal (N), Reactivity m niput: tion (R), lastrumint m-Jfunction (I), Component m:lfunction (C). M-jor transi:nt (M) l >
l Review Complete: l l 1 l i l l l i
-. _ . . - . _ _ . _ . - - _ . - - - - - . - _ . . . - . . _ - . ~ _ _ - . . - __. , ES-301 OPERATOR ACTIONS ES-301-4 Scenario No.1 Page 3 Event No.1 Brief Event
Description:
Continue Plant Startup per PPM 3.1.2 section 4.9, control rod sequence "A2", page 34, RWM group / step 34-5, control rod 18-51. l Time Position Candidates Actions / Behavior CRS Directs CROI to verify control rod, pull step, and current Plant condition and then continue with the startup per PPM 3.1.2 section 4.9, control rod step 34-5 CROI Verifies Plant Status as directed, then continues with the startup per PPM 3.1.2, section 4.9, control rod step 34-5. ALL Continue to monitor Plant conditions. Event No.1 Brief Event
Description:
RFW-LI-606A, RPV/L narrow range monitor, fails low. Time Position Candidates Actions / Behavior CROI Receives and acknowledges annunciator P603-A8 3-7, "RPV LEVEL HIGH/ LOW ALERT. Confirms alarm by observing RPV level normal except RFW-LI-606A reading low. Reports findings to CRS. CRS Acknowledges reports and directs CRO1 to continue monitoring RPV level closely. Refers to Tech. Specs. CRO3 Verifies level with, and monitors P601 level instruments for changes. ALL Continue to monitor Plant conditions and performance of plant startup. ; i 1 i i l l
_ . _ _ _ _ _ _ _ _ _ . _ . . _ _ . - _ _ . _ _ - ._. .~ ___ . ._ _ ~ ES-301 OPERATOR ACTIONS ES-301-4
- Scenario No. 3_ Page 4 ;
- Event No. _J._
1 i Brief Event
Description:
- Reactor recirculation pump "A" trips.
i ' i' t i Time Position Candidates Actions / Behavior ! CRO3 Acknowledges annunciators and verifies Recirc pump "A" has tripped, report all information to the CRS. . j CRS Directs CRO3 to follow instructions in ARP. Refers to and has CRO3 l
- perform PPM 2.2.1, section 5.4, Single loop Operation. Check for -
] compliance with Tech. Spec. flow requirements for Single Loop j i Operations. ! CRO3 Performs all applicable portions of PPM 2.2.1, section 5.4. j CRO1 Verifies single loop flow LT 41,725 gpm per Tech. Spec. requirements i and ensures limits of the Power to Flow Map. I i j ALL Continue to monitor Plant conditions. i i i ^ Event No. __4_ i i Brief Event
Description:
Loss of Control Air header pressure. Time Position Candidates Actions / Behavior j CRO3 Acknowledges annunciators and verifies CAS pressure going down, report all information to the CRS. Dispatch operators to investigate and isolate if possible. Monitor for system affects due to the loss of CAS pressure.
- CRS Directs CRO3 to follow instructions in ARP. Briefs crew on potential I control rod drifts and MSIV closure should CAS pressure continue j down. Inform crew if and when actions should be taken at his direction.
1 i ALL Continues to monitor Plant conditions. l ! l l I I i i i t
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l ; i US-301. OPERATOR ACTIONS ES-301-4 Scenario No.1 Page 5 { i Event No. J_ Brief Event
Description:
Control rods drift in and MSIV's drift closed. Reactor scram and entry into PPM 5.1.1. Time Position Candidates Actions / Behavior Direct reactor scram, after second rod drift with full IN indication ! l CRS ' l and/or impending MSIV closure. l CROI Scrams and refers to PPM 3.3.1 CRS Enters PPM 5.1.1 as required. Directs CRO3 to take the MSIV switches ; to closed as required. CRO3 If/when the MSIV's have isolated, place control switches for MSIV's in the closed position. , ALL Continues to monitor Plant conditions. Event No. 6_ ; Brief Event
Description:
Electric ATWS, entry into PPM 5.1.2. i Time Position Candidates Actions / Behavior CROI Recognize and report Electric ATWS. CRS Exits PPM 5.1.1 and enters PPM 5.1.2 ALL Continue to monitor Plant conditions. i Event No.1 Brief Event
Description:
ARI fail to initiate rod insertion. Time Position Candidates Actions / Behavior CROl Report failure of ARI to insert control rods. CRS Continue to direct from PPM 5.1.2. ALL Continue to monitor Plant conditions.
1 ES-301 OPERATOR ACTIONS ES-301-4 Scenario No. _3_ Page 6 Event No. _6_ : i Brief Event
Description:
RFW-V-10A & 10B fail closed. I Time Position Candidates Actions / Behavior CRO3 Report RFW-V-10A/B failed closed, attempt to control level with l RFW-LIC-620. ' CRS Direct CRO3 to attempt to regain level control. ALL Continue to monitor Plant conditions. IMPORTANT SCENARIO NOTES: I 1. 1 i
s -301 SCENARIO EVENTS Form ES-301-3 Simulator Facility: WNP-2 Scenario No. 4 File # Drill _964HLC.pgp Examiners: Candidates: Initial Conditions: Initialize IC 119, Plant Startup at = 4% power. Turnover: The Plant is continuing a Startup at = 4% power. APRM "C" is out of service, bypassed, and I&C is working to repair. RCIC Quarterly Operability Test, PPM 7.4.7.3.3B, is due to be performed this shift. Event No. Malf. No. Type
- Description 1 N/A R Continue Plant Startup per PPM 3.1.2.
2 N/A N RCIC Quarterly Operability Test, PPM 7.4.7.3.3B 3 MAL RCI4 C/M RCIC steam leak upstream of RCIC-V-45. 4 MOV RCIl6 C RCIC steam line isolation, RCIC-V-8 fails to isolate due to being mechanically bound. 5 LOA EPS214 I Blown fuse on RCIC steam line isolation, RCIC-V-63. 6 MAL RRS4A M Small break LOCA, Reactor Recire. "A" suction line. 7 MOV CSS 7 I HPCS injection valve, HPCS-V-4 fails to auto open.
- Normal (N), Reactivity manipulation (R). Instrument malfunction (I), Component malfunction (C), Major transient (M)
Review Complete: l
s RS-301 OPERATOR ACTIONS ES-301-4 Scenario No. _4_ Page 2 Event No.1 Brief Event
Description:
Continue Plant Startup per PPM 3.1.2. Time Position Candidates Actions / Behavior CRS Directs CROI to verify control rod, pull step, and current plant conditions and then continue with the startup per PPM 3.1.2. CROI Verifies plant status as directed, then continues with the startup per PPM 3.1.2. ALL Continue to monitor Plant conditions. Event No. _2_ Brief Event
Description:
RCIC full flow surveillance, PPM Time Position Candidates Actions / Behavior CRS Review RCIC Quarterly Operability Test, PPM 7.4.7.3.3B. CRO3 Review RCIC Quarterly Operability Test, PPM 7.4.7.3.3B. ALL Continue to monitor Plant conditions and perform Plant Shutdown. Event No. _3. < l Brief Event
Description:
RCIC steam leak upstream of RCIC-V-45. Time Position Candidates Actions / Behavior CRO3 Receives and acknowledges annunciator " LEAK DET RCIC EQUIP AREA TEMP HI-HI" P601. CRS Directs CRO3 to follow instructions in ARP. Check for compliance with l Technical Specifications. CRO3 Pull ARP P601-A41-8 and verify isolation. ALL Continue to monitor Plant conditions. ! l
, ES-301 OPERATOR ACTIONS ES-301-4 Scenario No. _4_ Page 3 Event No. J. Brief Event
Description:
RCIC steam line isolation, RCIC-V-8 fails to isolate due to being mechanically bound. Time Position Candidates Actions / Behavior CRO3 Recognize RCIC-V-8 has failed to close during verification of RCIC isolation on annunciator " LEAK DET RCIC EQUIP AREA TEMP HI-HI". Report failure to isolate to CRS. CRS Direct CRO3 to isolate RCIC steam line using RCIC-V-63 and RCIC-V-76. CRO3 Attempts to close RCIC-V-63 and valve position is lost. Investigation reveals a blown motor fuse and the isolation valve is still open. Reports unisolable steam leak to CRS. CRS Acknowledges, enters PPM 5.3.1 when high area temperatures are received. CROI Continues to monitor Plant conditions. NOTE: RCIC-V-63 will be closed when either directed to manually close the valve or replace the motorfuses. Event No. .f_ Brief Event
Description:
Blown fuse on RCIC steam line isolation, RCIC-V-63. Time Position Candidates Actions / Behavior ALL See actions / behaviors for event #4 above. ALL Continues to monitor Plant conditions. j
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d J 1 ES-301 OPERATOR ACTIONS ES-301-4 l# Scenario No. _4_ Page 4 $ Event No. _fiL ) Brief Event
Description:
Small break LOCA, Reactor Recire. "A" suction line. Time Position Candidates Actions / Behavior ! CRO3 Recognize drs ' piessure going up and RPV level going down from j alarms and i wauons. Report parameters and trends to CRS. If time permits between required actions, investigate and isolate leak source if possible.
- CROI Recognize drywell pressure going up and RPV level going down from
- alarms and indications. Report parameters and trends to CRS.
f CRS Enter PPM 5.1.1, direct / perform appropriate measures. 4 ALL Continue to monitor Plant conditions. Event No. _2_ i i Brief Event
Description:
HPCS injection valve, HPCS-V-4 fails to auto open. i i Time Position Candidates Actions / Behavior j CRO3 As RPV level reaches Level 2, while verifying automatic actions, i observe HPCS-V-4 does not automatically open as required by this j situation. Report situation to CRS as attempting to open manually. j t .
- CRO3 Reports attempt to open HPCS-V-4 manually was successful. Ensures i flow is appropriate and that RPV level is trending up.
f CRS Responds to CRO3 initial report and follows-up by directing him to j attempt to manually open HPCS-V-4. i ! Verifies HPCS injecting with HPCS-V-4 opened manually and ! continues with measures directed by PPM 5.1.1 and PPM 5.2.1. 4 1 ALL Continue to monitor Plant conditions. l s i IMPORTANT SCENARIO NOTES: s i $ 1. 1 J T a _ _ _ _ _ .._ ._.
ES-301 SCENARIO EVENTS Form ES-301-3 Simulator Facility: WNP-2 Scenario No. 5 File # Drill 965HLC. pen Examiners: Candidates: Initial Conditions: Initialize IC 117. Plant at = 96%, EOC coast-down. Turnover: The plant is operating at = 96% power in a EOC coast-down. Suppression pool temperature is = 81 *F. Work in progress includes;
- 1) RPS & Isolation Reactor Vessel Low Level surveillance. Ievel 3 & Ixvel 8 CFT.
PPM 7.4.3.1.1.50, currently at step 7.1.13.
- 2) Maintenance is repairing CRD-P-1 A, it is expected to be returned to service early in this shift.
- 3) Fire Protection Jockey Pump, FP-P-11, is out of service. Maintenance is investigating.
Event No. Malf. No. Type
- Description 1 N/A N PPM 7.4.3.1.1.50 - step 7.1.13 CFT in progress.
2 N/A N Start CRD-P-1 A and secure CRD-P-1B. 3 MAL EPS6D C/I Imss of annunciators on P601, P602, and P603 due to a ground on SI-2. 4 SWI CSS 47B I/R Inadvertent initiation of HPCS due to a relay failure. 5 MAL RRS4B M Reactor recirculation suction line break. 6 PMP CSS 3 C LPCS-P-1 shaft break.
- Normal (N), Reactivity manipulation (R), Instrument malfunction (I), Component malfunction (C), Major transient (M)
Review Complete:
r ES-301 OPERATOR ACTIONS ES-301-4 ! Page 2 Scenario No. _i_ 4
- Event No.1 P Brief Event
Description:
PPM 7.4.3.1.1.50 - step 7.1.13 CFT in progress. t Time Position Candidates Actions / Behavior ; CRS Review PPM 7.4.3.1.1.50 - step 7.1.13 CFT in progress. CROI Review PPM 7.4.3.1.1.50 - step 7.1.13 CFT in progress. ALL Continue to monitor Plant conditions and perform Plant Shutdown. Event No. 2 Brief Event
Description:
Start CRD-P-1 A and secure CRD-P-1B. Time Position Candidates Actions / Behavior I CRS Direct shifting CRD pumps per PPM 2.1.1, section 5.6. CROI Shift CRD pumps per PPM 2.1.1, section 5.6. ALL Continue to monitor Plant conditions and perform Plant Shutdown. l l Event No.1 1 Brief Event
Description:
Loss of annunciators on P601, P602, and P603 due to a ground on S1-2. Time Position Candidates Actions / Behavior CRO1 Receives and acknowledges annunciator " ANNUNCIATOR 125 VDC LOSS". Report situation to CRS and monitor RPV level, RPV pressure , and all scram parameters. l CRS Acknowledge and direct CRO's to pull ARP, PPM 4.7.8.3, less of Control Room Annunciation, and PPM 4.7.8.1,125 VDC Distribution System Failure. j Review actions with CRO's, perform those that are appropriate. Classify the event per PPM 13.1.1. CRO3 Start continuous walkdown of H13-P601, H13-P602, and H13-P603. ALL Continue to m6nitor Plant conditions. , 4
ES-301 OPERATOR ACTIONS ES-301-4 Scenario No. _5_ Page 3 Event No. _4_ Brief Event
Description:
Inadvertent initiation of HPCS due to a relay failure. Time Position Candidates Actions / Behavior CRO3 Acknowledge and report HPCS has started and is injecting. Verify validity of initiation, if not valid secure HPCS. CROI Monitor all indications of Reactor power due to inadvertent injection of HPCS. Report ANY indication of increase in Reactor power due to the injection from HPCS. CRS Acknowledge and direct CRO's to pull ARP, appropriate PPM's. Review actions with CRO's, to ensure appropriate actions have been taken. ALL Continues to monitor Plant conditions. l l Event No. _5_ Brief Event
Description:
Reactor recirculation suction line break. 1 Time Position Candidates Actions / Behavior i CRO3 Recognize drywell pressure going up and RPV level going down from indications. Report parameters and trends to CRS. If time permits between required actions, investigate and isolate leak source if possible. CROI Recognize drywell pressure going up and RPV level going down from indications. Report parameters and trends to CRS. CRS Enter PPM 5.1.1 and PPM 5.2.1. Direct / perform appropriate measures. ALL Continues to monitor Plant conditions. l
l l ES-301 OPERATOR ACTIONS ES-301-4 i Scenario No. _5_ Page 4 j Event No. _ft. Brief Event
Description:
LPCS-P-1 shaft break. Time Position Candidates Actions / Behavior CRO3 As Low Pressure injection system start to inject, verify proper flows, pressures, and pump motor amps. When LPCS-P-1 reaches a discharge ; flow rate of 4,000 gpm it will experience a shaft break. Recognize and report the situation to the CRS. CRS Acknowledges report and directs securing LPCS-P-1. ALL Continue to monitor Plant conditions. IMPORTANT SCENARIO NOTES: 1. l l
- -. .. . _ - . - -- - . _ . . -- _ - _ _ - - - _ _ . _ _ - - _ _ . = - _ _ _ _ _ -
d L WASHINGTON PUBLIC POWER , SUPPLY SYSTEM INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING COURSE TITLE JOB PERFORMANCE MEASURE ] LESSON TITLE j ALIGN LPCI C TO STANDBY STATUS l { LESSON LENGTil 8 Min MAXIMUM STUDENTS 1 INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No. OJT Guide PQD Code Rev. No. Simulator Guide PQD Code Rev. No. Student Handout PQD Code Rev. No. JPM PQD Code LR000159 Rev. No. 6 Checkoff Sheet PQD Code Rev. No. Exam PQD Code Rev. No. DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED llY Larry Monroe DATE 4/20/95 REVISED BY Randy Guthrie DATE 8/20/96 TECIINICAL REViliW BY DATE l INSTRUCTIONAL RiiVIEW BY DATli APPROVED llY DATli T rammg Manager Matrix U[xlate Vision # WP U lxlate W.\
ALIGN LPCI C TO STANDBY STATUS MINOR REVISION RECORD ._ Minor Ibeription Atrected lintered I!!rective Manager Rev Number of Revision Pages lly Date Approval g I l I l l Page1of9 LR000159 Rev. 6
ALIGN LPCI C TO STANDBY STATUS JPM SETUP Simulator ICs: Any M:lfunctions/ Remote Triggers: I N/A l Overrides (Optional): N/A l Special Setup Instructions: Rotate the RHR B & C Manual Initiation Pushbutton to the ARMED position. Task Standard: Control Rmuu manipulations are performed to align RHR "C' hxip to standby readiness ihr normal power operation. 1 i Page 2 of 9 LR000159 Rev. 6
ALIGN LPCI C TO STANDBY STATUS RESULTS OF JPM: Examinee (Please Print): Evaluator (Please Print): Overall (circle one) Exam Code Evaluation SAT / UNSAT Simulator IC Used Validation / Critical Time . JPM Completion Tune N/A 8 Minutes / NA Comments: l Evaluator's Signature: Date: Page 3 of 9 LR000159 Rev. 6
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l l l
- l ALIGN LPCI C TO STANDBY STATUS 1
JPM CHECKLIST l
- Items are Critical Steps Event Control Step Element llandard Sat /Unsat RECORD START TIME:
l Note: 1 Ensures RHR manual Rotates RHR B/C initiation psubbutton S / U* l JPM steps occur at initiation pushbutton arming collar to the DISARMED H13-P601 unicss is disarmed position otherwise noted , Comments: l 2 Ensures LPCI Ensures RHR-V-41C actuator is in the S/U testable check valve neutral position (green light on) actuator is neutral Comments: 3 Ensures suppression Verifies RHR-V-2 I indicates closed S/U pool cooling test (green light on) return valve closed Verifies control switch in AUTO S/U Comments: 4 Ensures injection Verifies RHR V-42C indicates closed S/U valve closed (green light on) l Verifies amber override light off S/U i l Verifies control switch in AUTO S/U Comments: 1 i k Page 4 of 9 LR000159 Rev. 6
ALIGN LPCI C TO STANDBY STATUS I
- Items are Critical Steps l l
Event Control Step Element Standard Sat /Unsat 5 Ensures suppression Verifies RHR-V-4C indicates open (red S/U pool suction valve light on) Wrifics control switch in OPEN with S/U key removed Comments: i 6 Ensures minimum Verifics RHR-FCV-64C indicates S/U ! flow valve is closed closed (green light on) ) Verifics control switch in AUTO S/U Comments: I 7 Ensures no initiation Ensures initiating logic scal-in amber S/U signal present indicating light is off I l Comments: i
)
8 Ensures pump set for Ensures RHR-P-2C control switch is in S/U automatic start AUTO Ensures manual override scal-in amber S/U indicating light is off , Comments: 9 Verifies override At H13-P618 ensures RHR-RMS-S44B S/U i alignment of kc31ock keylock test switch is in NORMAL with l switches key removed Comments: Page 5 of 9 LR000159 Rev. 6
ALIGN LPCI C TO STANDBY STATUS
- Items are Critical Steps l
Event Control Step Element Standard Sat /Unsat 10 Wrifies override At H13-P618 casures RIIR-RMS-S70 S/U l alignment of key lock keylock test switch is in NORMAL with l switches key removed l Comments: i 11 Wrifies override At H13-P618 ensures RHR-RMS S/U alignment of keylock S101C and RHR-RMS-5103C are in switches NORMAL with keys removed Comments: 12 Wrifies system filled Ensures RilR-P-3 running S/U Ensures fil3-P601. A2-6.5, RHR C S/U PUMP DISCH PRESS HIGH/ LOW alarm is clear Comments: 13 Wrifies pump not Ensures LOCKOUT CIRCUIT S/U locked out AVAILABLE white light is on Comments: Page 6 of 9 LR000159 Rev. 6
ALIGN LPCI C TO STANDBY STATUS
- Iterns are Critical Steps Event Control Step Element Standard Sat /Unsat 14 Checks status of BISI Ensures RHR C BISI alarms are clear S/U l
Termination Cue: Comments: ! The tennination point 1 of this JPM has been l reached. RECORD Calculate JPM Completion Time: 1 TERMINATION l TIME: JPM Termination Time: JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) l l l l l l i l i i i Page 7 of 9 LR000159 Rev. 6
ALIGN LPCI C TO STANDBY STATUS l JPM INFORMATION CARD , l l 1 l Initial Conditions: A Minimum Startup Checklist is being implemented to support a reactor startup. I The RHR system has been filled and vented following system maintenance. l Cue: The CRS has directed you to align LPCI Loop C to standby status l using PPM 2.4.2 Section 5.6. 1 Tools / Equipment: None Safety items: None Task Number: RO-0199-N-RHR Validation Time: 8 Minutes Prerequisiste Training: 82-RSY-1304-L3 Time Critical: No PPM
Reference:
2.4.2 Rev 25 Location: Siinulator NUREG 1123 Ref: 203000A4.02 (4.1/4.1) Itrfortnance Method: Perforrn Prepared or Revised by: Randy Guthrie Revision Date: 8/20/96 1 1 i i Page 8 of 9 LR000159 Rev. 6
ALIGN LPCI C TO STANDBY STATtJS STUDENT JPM INFORMATION CARD Initial Conditions: A Minimum Startup Checklist is being implemented to support a reactor startup. The RHR system has been filled and vented following system maintenance. Cue: The CRS has directed you to align LPCI Loop C to standby status using PPM 2.4.2 Section 5.6. Page 9 of 9 LR000159 Rev. 6
I 4 O l ALIGN LPCI C TO STANDBY STATUS l 4 i
- 203000A4.02 RHR/LPCI: Injection Mode (Plant Specific) ;
Ability to manually operate and/or monitor in the control room system valves. (4.1/4.1) i Question: , What are the indicatiore of ECCS suction strainer plugging? ) l ? l 4 l Answer: ] Flowrate, pump discharge pressure and pump motor amps erratic or decreasing. Also frequent adjustment of test return valve at j steady state conditions. j
Reference:
PPM 4.4.7.1 - ECCS Suction Strainer Plugging Comments: Question: , Placing' keylock test switch RHR-RMS-S101C to BYPASS disables the reactor pressure injection valve open permissive. What is the j function of the permissive? I Answer: Overpressure protection for piping upstream of the injection valve. This piping is not rated for pressure above 500 psig. j
Reference:
1 RHR System Text Comments:
WASHINGTON PUBLIC POWER SUPPLY SYSTEM INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE PLACE RSCS INTO SERVICE LESSON LENGTH 9 Min MAXIMUM STUDENTS 1 INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No. OJT Guide PQD Code Rev. No. Simulator Guide PQD Code Rev. No. Student liandout PQD Code Rev. No. JPM PQD Code LR000195 Rev. No. 5 Checkoff Sheet PQD Code Rev. No. Exam PQD Code Rev. No. DIVISION TrlLE Nuclear Training del %RTMENT Operations Training PREl% RED BY Sean Dallago DKili 9/7/92 REVISED BY Randy Guthrie DAH 8/21/96 l 1 l TECilNICAL REVIEW BY DATE j INSTRUCTIONAL REVIEW BY DATE APPROVED 13Y DNI'E , i Training Manager Matrix ilpdate Vision # WP Update WA j
l PLACE RSCS INTO SERVICE 1 i l MINOR REVISION RECORD j Minor Description Atrected Entered Effective Manager l ) Rev Nurnber of Revision Pages By Date Approval l i i l l l Page1of8 LR000195 Rev. 5
PLACE RSCS INTO SERVICE JPM SETUP Simulator ICs: N/A l Milfunctions: N/A i Overrides (Optional): 1 N/A 1 Special Setup Instructions: Before JPM is intiated, deterndne which is the current rm! sequence (A or B). 3 Task Standard: Maidpluations have been completed to place the RSCS system in service per PPM. i i
- f ih l
i i i i. i 4 a iF 1 1 l Page 2 of 8 LR000195 Rev. 5
i PLACE RSCS INTO SERVICE RESULTS OF JPM: Examinee (Please Print): Evaluator (Please Print): Overall (circle one) Exam Code Evaluation SAT / UNSAT
- Simulator IC Used Validation / Critical Time JPM Completion Time {
9 Minutes / NA Conunents: i l l i Evaluator's Signature: Date: Page 3 of 8 LR000195 Rev. 5
PLACE RSCS INTO SERVICE JPM CHECKLIST
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat RECORD START TIME:
1 Ensure CLI board Ensures toggle switch on CLI board in S/U toggle switch is up the RSCS card file at lil3-P659 is in the UP position Cue: Comments: Toggle switch is in the UP position 2 Ensure substitute Ensures substitute position window on S/U display is blank RSCS control panel at Hl3-P603 is blank i Cue: Comments: Substitute position ] window is blank - 3 Determine if any rods Depresses red display control S/U* are bypassed pushbutton to illuminate the bypass light Verifies no rods bypassed (no red lights S/U lit) Cue: Comments: Display pushbutton is depressed with bypass ; light lit and no bypassed control rods are indicated 4 Select desired rod Depresses Sequence Select pushbutton S/U* withdrawal sequence (as necessary) until rod sequence specified by pull sheet is illuminated Cuc: Comments: l Sequence (determined by the current sequence) is selected Page 4 of 8 LR000195 F.ev. 5
PLACE RSCS INTO SERVICE l I I
- Items are Critical Steps l Event Control Step Elernent Standard Sat /Unsat l
- i 5 Wrify RSCS Depresses Amber Display Control S/U* correctly displays all pushbutton to illuminate ALL RODS l rods in the selected light Br up , Wrifies alllighted amber LEDS S/U l correspond to rods in the selected rod group ( uses pull sheet or PPM i 7.4.1.4.2.1 Attachment 9.1) . Cue: Comments: The amber pushbutton ].
- is depressed with ALL
- RODS lit and selected rod group displayed
)
- I j 6 Verify current rod Depress Red Display Control S/U*
1 pattern pushbutton to illuminate the Rods F.I. light Verifies all rods are consistent with S/U $ current rod pattern i
; Cue: Comments:
j The red pushbutton is
- depressed with Rods F.
- 1. lit and current rod j pattern displayed
! 7 Select first rod on the On the Reactor Manual Control Panel, S/U* i current operator's rod selects the first rod on the sequence i sequence sheet sheet which corresponds to the current
- sequence Ensures proper selection S/U Cue
- Comments:
Selected rod is selected i l ! 8 Depresses display Depresses Amber Diplay Control S/U* i control button pushbutton to illuminate Free Rods l light Cuc: Comments:
- Amber light is illuminated for the
, selected rod i
.b i t,
PLACE RSCS INTO SERVICE
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat 9 Depresses direction Depresses Direction pushbutton to S/U*
button illuminate the W/ draw light Cue: Comments: W/ draw light is lit 1 10 Ensures free rods are Ensures all rods that can be withdraw S/U indicated correctly according to the Operator's rod l sequence sheet belong to the same j group as the selected rod (uses pull i sheet or PPM Attachment 9.1 to identify group) have amber lights lit Cue: Comments: Amber lights indicating free rods and the pull sheet are in agreement 1I Ensures rod blocks Ensures RSCS Insen and Withdraw S/U are consistent blocks are consistent with the current rod pattern Cue: Comments: No rods blocks are indicated Termination Cue: ! The termination point of this JPM has been reached. RECORD Calculate JPM Completion Time: TERMINATION TIME: JPM Termination Time: JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) Page 6 of 8 LR000195 Rev. 5
PLACE RSCS INTO SERVICE JPM INFORMATION CARD Initial Conditions: The plant is shutdown, all rods are inserted. All Surveillance requirements have been met and all prerequisites are satisfied. Cue: The CRS has directed you to place the Rod Sequence Control System l In service per PPhi 2.1.5. CONTROL AIANIPULATIONS WILL NOT BE PERFORhiED. ALL ACTIONS AND STEPS H'lLL BE SinfULATED. Tools /Equiprnent: None Safety Iterns: None Task Nurnber: RO-0151-N-RSCS Validation Time: 9 Minutes Prerequisiste Training: 82-RSY-0706-Li Time Critical: No PPM
Reference:
2.1.5 Rev 9 Locatiori: Control Rmun i NUREG 1123 Ref: 201004 A402 (3.5/3.2) Perforinance Method: Simulate j l Prepared or Revised by: Randy Guthrie Revision Date: 8/21/96 l l l Page 7 of 8 LR000195 Rev. 5
I i PLACE RSCS INTO SERVICE STUDENT JPM INFORMATION CARD 1 l ,, Initial Conditions: The plant is shutdown, all rods are inserted. All Surveillance requirements have been met and all prerequisites are satisfied. Cue: The CRS has directed you to place the Rod Sequence Control System \ In service per PPhi 2.1.5. CONTROL AfANIPULATIONS WILL NOTBE PERFORhfED. ALL j ., ACTIONS AND STEPS WILL BE SihfULATED. l I l l 1 i i l i 1 i f Page 8 of 8 LR000195 Rev. 5
_ . _. .- _ . . _ . _ . _ .. .~ _. ._ _ . ._ _ [ l l PLACE RSCS INTO SERVICE l 201004A4.02 Rod Sequence Control System (Plant Specific) l Ability to manually operate and/or monitor in the control room MSCS console switches and indicators: BWR-4,5 (3.5/3.2) Question: At what reactor power level does RSCS generate rod blocks and what signal is used to determine reactor power? Answer: 20% reactor power as determined by main turbine ist stage pressure
Reference:
! RSCS Systems Text Comments: Question: How can the operation of main turbine bypass valves affect RSCS operation? Answer: With bypass valves open, turbine 1st stage pressure may be lowered, in turn causing ind.icated reactor power to be lower than actual reactor power. RSCS may generate rod blocks prior to reactor power dropping to 20%.
Reference:
RSCS Systems Text Comments: l l
1* a WASHINGTON PUBLIC POWER SUPPLY SYSTEM INSTRUCTIONAL COVER SHEET PR(XiRAM TITLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS j LESSON LENGTil 3 Min MAXIMUM STUDENTS 1 INSTRUCTIONAL MATERIALS INCLUDED k Lesson Plan PQD Code Rev. No. ! OJT Guide PQD Code Rev. No. Simulator Guide PQD Code _ Rev. No. I Student Handout PQD Code Rev. No. l JPM PQD Code LR000302 Rev. No. 3 ! Checkoff Sheet PQD Code Rev. No. l Exam PQD Code Rev. No. ! DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Donald Hughes DATE 6/29/96 REVISED BY Randy Guthrie DATE g/22/96 TECl[NICAL REVIEW 11Y DATE INSTRUCTIONAL REViliW liY DATE APPROVED BY DATl! Traisung Manager l Matrix Uplate Vnion # WP Uplate WA
l INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS l , MINOR REVISION RECORD y;,,,,, . Athsted Entered Effective yanager
**"P'f""
Rev Number of Revision Pages g) Date Approval l I i l l l l Page 1 of 8 LR000302 Rev. 3 l
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l INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS l JPM SETUP l l Simulator ICs: 1 l Any IC where the initiation of RCIC will NOT cause a reactor scrani frorn a turbine trip. 1 Milfunctions: 1 CNH RCl2 5,50,0,0,D RCIC Controller Failure Overrides (Optional): N/A ! Special Setup Instructions: ; N/A
'Ihsk Standard: i RCIC is initiated and injecting at rated flow (600 gpm) per PPM 2.4.6. l I
i I i l l l l l l l Page 2 of 8 LR000302 Rev. 3 l
INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS i RESUllIE OF JPM: Examinee (Please Print): Evaluator (Please Print): l Overall (circle one) Exam Code l l Evaluation SAT / UNSAT l l Simulator IC Used Validation / Critical Time JPM Completion Time 3 Minutes / NA l Comments: l l l l l l Evaluator's Signature: Date: i l Page 3 of 8 LR000302 Rev. 3
INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS i JPM CHECKLIST 4
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat
! RECORD START 3 TIME: l
- Note
- 1 initiate RCIC Rotates RCIC RMS-S36 to ARM and S/U*
1 All steps occur at depresses the initiation pushbutton
! lil3-P601 unless otherwise noted j Note:
- PPM 2.4.6 Section j 5.2 Step 1 is N/A
. since this is NOT a test Conunents: T 2 Verify proper system Ensures:
" E""' "' RCIC V-46 opens (red light on) S/U (Lube Oil Cooler Supply)
RCIC P-2 starts (red light on) S/U (Condenser Vacuum Pump)
- ' RCIC-V-45 opens (red light on) S/U (Steam Supply To RCIC)
RCIC-V-13 opens (red light on) S/U (RPV Injection) Comments: I Page 4 of 8 LR000302 Rev. 3
1 1 9 INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS I
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat j 3 Verify auxiliary When RCIC-V-45 opens, ensures:
l
" E"* "
RCIC-V25 closes (green light on) S/U i (Warmup Drain) l l RCIC-V-26 closes (green light on) S/U i (Warmup Drain) l i l RCIC-V-4 closes (green light on) S/U . (Discharge To EDR) l RCIC-V-5 closes (green light on) S/U : (Discharge To EDR) 1 SW-P-1B starts S/U (20 second tirne delay) Note: Comments: ' llic failure of RCIC-FIC-6(K) may be ! identified before valve alignment and ; corrective actions ! taken , 4 Verify room cooling At H13-P625, ensure SW-V-34 is open S/U ; established ; Comments: 1 5 Recognizes RCIC- Reports to CRS that RCIC-FIC-600 is S/U* FIC-600 failure not controlling properly in AUTO Cue: Comments: As CRS acknowledge controller failure and direct the operator to take actions as necessary to deliver 600 gpm to the RPV 6 Adjusts RCIC system Places RCIC-FIC-600 in MANUAL S/U*
- llow and increases system flow to 600 gpm Comments:
) l I a J Page 5 of 8 LR000302 Rev. 3
i l INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS l
- Items are Critical Steps l Event Control Step Element Standard Sat /Unsat 7 Verify minimum flow Ensures RCIC-V-19 closes when flow S/U operation is GT 150 gpm t Comments: ;
8 Maintain system flow Adjust RCIC FIC-600 as necessary to S/U ! maintain 600 gpm to the RPV i Termination Cue: Comments: ! The termination point of this JPM has been reached. RECORD Calculate JPM Completion Time: TERMINATION TIME: JPM Termination Time: 1 JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) I i i i l Page 6 of 8 LR000302 Rev. 3
1 INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS , JPM INFORMATION CARI) Initial Conditions: Plant conditions exist that require RCIC initiation to maintain RPV level. 3 The RCIC start is NOT part of a planned test. I The main turbine may trip as a result of this evolution and that has been ; determined to be acceptable by the Shift Manager. . Cue: The CRS has directed you to initiate the RCIC systernfor RPV injection by utilizing the ARM / DEPRESS rnode of operation per PPM 2.4.6. I Tools / Equipment: None Safety items: None Task Number: RO-0268-N-RCIC Validation Time: 3 Minutes R0-0658-N-RCIC Prerequisiste Training: N/A Time Critical: No PPM
Reference:
2.4.6 Rev 22 Location: Simulator NUREG 1123 Ref: 217000A2.10 (3.1/3.1) Itrformance Method: Perform 217000A2.11 (3.1/3.2) Prepared or Revised by: Randy Guthrie Revision Date: 8/22/96 Page 7 of 8 LR000302 Rev. 3
1 INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS STUDENT JPM INFORMATION CARD Initial Conditions: Plant conditions exist that require RCIC initiation to maintain RPV level. The RCIC start is NOT part of a planned test. The main turbine may trip as a result of this evolution and that has been determined to be acceptable by the Shift Manager, l l Cue: The CRS has directed you to initiate the RCIC systemfor RPV injection by utilizing the ARM / DEPRESS mode of operation per PPM . 2.4.6. l 1 I i l l l
)
l l Page 8 of 8 LR000302 Rev. 3
INITIATE RCIC FOR RPV INJECTION ARM AND DEPRESS 217000A2.10 Reactor Core Isolation Cooling System (RCIC) Ability to (a) predict the impacts of turbinc controlsystemfailurcs on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations. l (3.1/3.1) 217000A2.ll Reactor Core Isolation Cooling System (RCIC) Ability to (a) predict the impacts ofinadeguate systemflow on the REACTOR CORE ISOLATION l COOLING SYSTEM (RCIC); and (b) based on those predictions, use procedures to correct, control, or l mitigate the consequences of those abnormal conditions or operations. (3.1/3.2) ! Question: When RCIC was initiated a turbine trip occurred. What conditions must be met for this trip to occur? Answer: RCIC-V-45 and RCIC-V-13 both open. I Reference-PPM 2.4.6 & RCIC Systems Text Comments: Question: RCIC has been operating in the test mode for an extended period of time when it is noted that the containment oxygen level is increasing. Why is this happening? l Answer: Air in-leakage from the RCIC turbine gland seal system.
Reference:
PPM 2.4.6 & RCIC Systems Text l Comments.
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e
~
WASHINGTON PUBLIC POWER I N SUPPLY SYSTEM l I INSTRUCTIONAL COVER SHEET J i I i !' PROGRAM TITLE , LICENSED OPERATOR /STA REQUALIFICATION TRAINING j- : j COURSE TITLE JOB PERFORMANCE MEASURE i LESSON TITLE OVERRIDE ECCS VALVE LOGIC TO THROTTLE RPV INJECTION 1 4 l LESSON LENGTH 3 Min MAXIMUM STUDENTS I INSTRUCTIONAL MATERIALS INCLUDED ] Lesson Plan PQD Code Rev. No. ! OJT Guide PQD Code Rev. No. I Simulator Guide PQD Code Rev. No. j Student Handout PQD Code Rev. No. j JPM PQD Code LR000223 Rev. No. 4 I Checkoff Sheet PQD Code Rev. No. - 4 Exam PQD Code Rev. No. j DIVISION TITLE Nuclear Training ) 4 !' DEPARTMENT Operations Training 1 1 PREPARED BY Sean Dallago ATE 5/7/92 ] i . i ' i REVISED BY Randy Gutbrie DATE 8/22/96 TECIINICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Traming Manager Matrix Update Vision # WP Update WA
l OVERRIDE ECCS VAINE LOGIC TO TIIROTTLE RPV INJECTION MINOR REVISION RECORD Minor Description Airected Entered Etrective Manager Rev Number of Revision Pages By Date Approval . l l , l 1 i l 1 i
- l l
l l l l i l l t , I i l i l I Page1of7 LR000233 Rev. 4
OVERRIDE ECCS VALVE LOGIC TO TIIROTTLE RPV INJECTION J JPM SETUP _ i Simulator ICs: : Any 1 l Milfunctions: I N/A i Overrides (Optional): N/A
]
Special Setup Instructions: N/A Task Standard: EOP implemented to override ECCS valve logic to allow throttling RPV injection. 1 I i l l l l l i i Page 2 of 7 LR000233 Rev. 4
OVERRIDE ECCS VALVE LOGIC TO TIIROTTLE RPV INJECTION RESULTS OF JPM: l Examinee (Please Print): Evaluator (Please Print): Overall. (circle one) Exam Code j Evaluation __ SAT / UNSAT Simulator IC Used Validation / Critical Time JPM Completion Time ) 3 Minutes / NA Comments: I Evaluator's Signature: Date: Page 3 of 7 LR000233 Rev. 4
~
OVERRIDE ECCS VALVE LOGIC TO TIIROTfLE RPV INJECTION l JPM CHECKLIST I
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat RECORD START TIME:
Note: 1 Override HPCS-V-4 At H13-P625, places HPCS-RMS-S25 S/U* JPM steps may be automatic logic in the OVERRIDE position performed in any order. Comments: 2 Override LPCS-V-5 At H13-P629, places LPCS-RMS-S21 S/U* automatic logic in the OVERRIDE position Comments: 3 Override RHR-V- At H13-P629, places RHR-RMS-S105 S/U* 42A automatic logic in the OVERRIDE position Comments: 4 Oserride RHR-V-42B At H13-P618, places RHR-RMS-S196 S/U* automatic logic in the OVERRIDE position Comments: Page 4 of 7 LR000233 Rev. 4
OVERRIDE ECCS VALVE LOGIC TO TliROTTLE RPV INJECTION
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat 5 Override RHR-V- At H13-P618, places RHR-RMS-5107 S/U*
42C automatic logic in the OVERRIDE position Termination Cue: Comments: The termination point of this JPM has been reached. RECORD Calculate JPM Completion Time: TERMINATION TIME: JPM Termination Time: JPM Start Time. - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) Page 5 of 7 LR000233 Rev. 4
l OVERRIDE ECCS VALVE LOGIC TO TIIROTTLE RPV INJECTION JPM INFORMATION CARD Initial Conditions: An ATWS condition exists, PPM 5.1.2 has been entered. J Cue: The CRS has directed you to override ECCS valve logic per PPM 5.5.1 to allow :hrottlingflow. Tools /Equipinent: None i I Safety Iteins: None 1 l Task Nunnber: RO-0669-E-ECCS Validation Tinne: 3 Minutes Prerequisiste Training: 82-RMD-0901-LP Time Critical: No PPM
Reference:
5.5.1 Rev 5 IAcation: Situulator NUREG 1123 Ref: 295015GA.06 (4.1/3.9) Performance Method: Perforin Prepared or Revised by: Randy Guthrie Revision Date: 8/22/96 1 l I l i Page 6 of 7 LR000233 Rev. 4
OVERRIDE ECCS VAIXE LOGIC TO TIIROTTLE RPV INJECTION STUDENT JPM INFORMATION CARD Initial Conditions: An ATWS condition exists, PPM 5.1.2 has been entered. Cue: 1 The CRS has directed you to override ECCS valve logic per PPM 5.5.1 to allow throttlingflow. [ l l l l Page 7 of 7 LR000233 Rev. 4
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I l i OVERRIDE ECCS VALVE LOGIC TO THROTILE RPV INJECTION l 295015GA.06 Incomplete Scram Ability to locate and operate components, including local controls. (4.1/3.9) ( Question: What is the response of ECCS injection valves to RPV pressure dropping below the open permissive with its associated keylock ! switch in OVERRIDE 7 Answer; l No response, valve remains as is. i
Reference:
! RHR System Text l Comments: 1 l i Question: What condition requires these valves to be throttlabic? l f l Answer: ! PPM 5.1.4, RPV Flooding, PPM 5.1.6, RPV Flooding ATWS l t
Reference:
PPM 5.0.10 i Comments: i s 4 v , _
WASHINGTON PUBLIC POWER SUPPLY SYSTEM INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE REACTOR FEED PUMP QUICK RESTART l LESSON LENGTil 16 Min MAXIMUM STUDENTS 1 INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No. OJT Guide PQD Code Rev. No. Simulator Guide PQD Code Rev. No. Student Handout PQD Code Rev. No. JPM PQD Code LR000131 Rev. No. 2 Checkoff Sheet PQD Code Rev. No. Exam PQD Code Rev. No. DIVISION TMLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ken Elliott DATE 10/03/95 REVISED BY Randy Guthrie DATE 8/20/96 i TEClINICAL REVIEW BY DATli INSTRUCTIONAL RIiVIEW liY DATE APPROVED BY DKI'E 1rainmg Manager i Matrix Update Vision # 966 WP Update WA l
REACTOR FEED PUMP QUICK RESTART MINOR REVISION RECORD Minor Description Ati'ected Entered Ell'ective Manager Rev Number ofRevision Pages fly Date Approval I l l l l 1 l l l Page1 of8 LR000131 Rev. 2
i REACTOR FEED PUMP QUICK RESTART l JPM SETUP 1 Simulator ICs: ) i 13 Mcifunctions: j N/A Overrides (Optional): N/A i Special Setup Instructions: Verify RPV level is stable, then trip RFP "A" Perform actions per PPM 4.840.Al 1-1 i Task Standard: i RFP "A" is staned and is feeding the reactor vessel. 4 i l l l Page 2 of 8 LR00013 i Rev. 2
l REACTOR FEED PUMP QUICK RESTART
\ \
I RESULTS OF JPM: l Examinee (Please Print): Evaluator (Please Print): i ! 1 Overall (circle one) Exam Code l SAT / UNSAT Evaluation
)
' Simulator IC Used Validation / Critical Time JPM Completion Time I 16 Minutes / NA l I
1 Conunents: l 4 i I i i ! i l s i - 4 4 i i i Evaluator's Signature: Date: i Page 3 of 8 LR000131 Rev. 2
. . .. _. _ = . - - - .-___ . _ .- . _ - _ - - __ . _ _ _
i 1 REACTOR FEED PUMP QUICK RESTART JPM CIIECKLIST
- Itents are Critical Steps Event Control Step Element Standard Sat /Unsat i
RECORD START i TIME: .A 1 Ensure HIGli Ensures at least two HIGH LEVEL S/U* ] LEVEL SEAL ins SEAL ins are reset are reset i Conunents: 4 j 2 Ensure Min Flow Ensures RFW FIC-2A is in auto (A) S/U J Controller in auto Comments: 3 Check turbine "A" Checks RFW-SC-601 A is in MDVP at S/U I speed controller set 0% properly f Comments: i l 4 Ensure MS-V 105A Ensures MS-V-105A and BS-V-17A S/U j i and BS-V-17A are are open j open q Comments: i l a j i 1 J l 1 Page 4 of 8 LR000131 Rev. 2
l 1 t REACTOR FEED PUMP QUICK RESTART j
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat Note: 5 Reset the feedwater Places Turbine Emergency Trip / Reset S / U* I If the HP and LP turbine switch (P840) to the Reset position valves start to reset and trip, examinec Ensures Turbine LP and HP Stop valves S/U should consider fully open and the white TRIP starting a second CIRCUIT AVAIL light is lit, then ,
release the Trip / Reset switch turbine LO pump Comments: S/U* 6 Roll RFP "A" Depresses increase arrow as necessary turbine and slowly to increase and hold turbine speed at = increase speed to 800 800 rpm. rpm j Comments: i S/U 7 Ensure turbine if engaged, verifics the turning gear turning gear disengages. (Green light lit) automatically disengages Comments: Transfers RFW-SC-601 A to MDEM by S/U* 8 Transfer controller to MDEM depressing the MDEM pushbutton Comments: i i i l l Depresses increase arrow on RFW-SC-S/U* l 9 Increases RFP " A" speed to = 1800 rpm 601 A until turbine rpm reaches = 1800 l rpm Comments: Page 5 of 8 LR000131 Rev. 2
l REACTOR FEED PUMP QUICK RESTART
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat l
l 10 Ensure turbine Ensures turbine turning gear control S/U turning gear is off switch is in the OFF position. (green i light on) l Comments: 11 Attain desired Adjusts RFP "A" turbine speed S/U* discharge feed rate controller RFW-SC-601 A until desired feed rate is established. Termination Cue: Comments: , The termination point j of this JPM has been i reached. l RECORD Calculate JPM Completion Time: 1 TERMINATION TIME: JPM Termination Time:
)
JPM Start Time- - l l JPM Completion Time: (Transfer to RESULTS OF JPM page 3) Page 6 of 8 LR00013i Rev 2
l l REACTOR FEED PUMP QUICK RESTART 1 i ! JPM INFORMATION CARD 1 l Initial Conditions: RFP "A" was manually tripped 5 minutes ago. The RFP is fully operational. l Cue: lou have been directed by the CRS to perfonn a quick restart of the l "A " RFP and return it to service per PPM 2.2.4. ; Tools / Equipment: None l l Safety Items: None Task Number: RO-0371-N-Rav Validation Time: 16 Minutes SRX-0366-N-RFW l Prerequisiste Training: 82-BST-3000-LP Time Critical: No PPM
Reference:
2.2.4 Rev.22 Location: Simulator l NUREG 1123 Ref: 259001 A4.02 (3.9/3.7) Itrformance Method: Perform ; Prepared or Revised by: Randy Guthrie Revision Date: 8/20/96 l l l l l Page 7 of 8 LR000131 Rev. 2
a 4 . REACTOR FEED PUMP QUICK RESTART l STUDENT JPM INFORMATION CARD l Initial Conditions: RFP "A" was manually tripped 5 minutes ago. The RFP is fully operational. ; Cue: l You have been directed by the CRS to perfonn a quick restart of the
- "A " RFP and return it to service per PPM 2.2.4.
l i i t i ! l l i i I Page 8 of 8 LR000131 Rev. 2
REACTOR FEED PUMP QUICK START 259001A4.02 Reactor Feedwater System Ability to manually operate and/or monitor in the control room manually start / control a RFP/TDRFP. (3.9/3.7) Question: With RFW-SC-601 A in MDVP, how is this signal prncessed to change TDRFP speed? Answer: Controller output is processed directly through the Signal Processing Enclosure (SPE) in the Control Room to the Turbine Termination Enclosure (TTE) in the individual feedpump room where the on-line LPEHC centroller repositions the governor valve.
Reference:
LR000060, Feedwater Level Control System (Digital Feedwater Modification) 4 Cgmments: Question: The plant has increased power fro:n 40% to 45% TDRFP governor valve position indication has lowered. How is this possibic7 Answer: Extraction steam supply to the TDRFP has increased causing speed to increase. The Speed Changer responds by closing the governor valve to control TDRFP speed.
Reference:
RFW Systems Text 4 4 1 Comments:
.--. -. . - . - . . . . . . . - ..~ .. -.- . . . . . . - . . . _ . . . ~ . . ~~ ~ . - . .
l l I,
..N8 WASHINGTON PUBLIC POWER
- SUPPLY SYSTEM INSTRUCTIONAL COVER SHEET i
- PROGRAM TITLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING '
4 COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE SYNCHRONI? ' THE MAIN GENERATOR WITH THE GRID : LESSON LENGTil 9 Min MAXIMUM STUDENTS 1 , INSTRUCTIONAL MATERIALS INCLUDED 2 j Lesson Plan PQD Code Rev. No. s OJT Guide PQD Code Rev. No. Simulator Guide PQD Code Rev. No. Student Handout PQD Code Rev. No. ! JPM PQD Code LR000172 Rev. No. 7 I Checkoff Sheet PQD Code Rev. No. Exam PQD Code Rev. No. DIVISION TITLE Nuclear Training DEPARMENT Operations Training PREPARED BY Larry Monroe DAE 10/13/94 REV1 SED 13Y Randy Guthrie OATE 8/20/96 TECllNICAL REVIEW BY DATE l l INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE l Training Manager i Matrix Update Visiori # WP Update WA
l I SYNCIIRONIZE TIIE MAIN GENERATOR WITII TiiE GRID l l MINOR REVISION RECORD i Minor Description Atlected Entered Effective Manager Rev Nmnber of Revision Pages Ily Date Approval i l i i l l Page 1 of 8 LR000172 Rev. 7 l
SYNCIIRONIZE THE MAIN GENERATOR WITli THE GRID JPM SETUP Simulator ICs: 8 Milfunctions/ Remote Triggers: N/A Overrides (Optional): N/A Special Setup Instructions: Perfi>rm steps 1 - 10 of PPM 2.5.7 Section 5.4. Verify South hus (PCB 4885) is selected ihr synchronization. Task Standard: Control Room manipulations are perfbrmed to synchronize and load the main generator on the grid at the proper voltage, frequency, and phase in the AUTO mode per plant pmcedures.. l Page 2 of 8 LR000172 Rev. 7
. SYNCIIRONIZE TIIE MAIN GENERATOR WITII TIIE GRID RESULTS OF JPM:
Examinee (Please Print): Evaluator (Please Print): Overall (circle one) Exam Code Evaluation SXf / UNSAT Simulator IC Used Validation / Critical Time JPM Completion Time N/A 9 Minutes / NA Comments: l l l Evaluator's Signature: Date: Page 3 of 8 LR000172 Rev. 7
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l SYNCIIRONIZE TIIE MAIN GENERATOR WITII Tile GRID JPM CHECKLIST l
- Items are Critical Steps Event Control Step Eleinent Standard Sat /Unsat RECORD START TIME:
Note: 1 Place generator Places voltage regulator control switch S/U JPM steps occur at voltage regulation in to the ON (RESET) position lil3-P800 and H13 automatic P802 unless otherwise noted ( Cue: Comments: Dittmer dispatch has been notified of intention to synchronize the main generator. They have verified MOD positions and aligned to the South bus. 2 Adjust turbine speed Adjusts turbine speed to 1802 - 1804 rpm using DEH as follows: Depresses REF S/U* Enters speed setpoint of 1802 - 1804 S/U* rpm on numerical keyboard Depresses ENTER S/U* Depresses ACCEL RPM / MIN S/U* Enters desired acceleration rate S/U* (approximately 25 RPM / MIN) Depresses GO S/U* Comments: i 3 Aligns ASHE breaker Places ASHE breaker #1 (PCB 4885) S/U*
#1 (PCB 4885) for synch slector switch in AUTO and auto synch observes sy nchroscope and voltmeter Comments:
Page 4 of 8 LR000172 Rev. 7
SYNCIIRONIZE TIIE MAIN GENERATOR WITil Tile GRID
- Itents are Critical Steps Event Control Step Element Standard Sat /Urcsat 4 Ensure synchroscope Adjusts turbine speed as necessary to S/U is rotating slowly in ensure synchroscope rotating slowly in the fast direction the fast direction Comments:
$ Adjust generator Operates main generator exciter voltage S/U*
synch voltage adjuster as necessary until synch voltage (500 KV incoming)is equal to bus voltage (500 KV running)i 15KV Comments: 6 Select Load Rate Depresses LOAD RATE MW/ MIN S / U* Enters load rate of 200 MWe/ Min on S/U* kypad Depresces ENTER pushbutton S/U* Note: Comments: Closing PCB 4885 will cause REFERENCE and REFERENCE DEMAND windows to show calculated MW not actual MW 7 Close generator Places PCB 4885 control switch in S/U* breaker CLOSE when synchroscope pointer passes the 11 o' clock position Checks PCB 4885 closed (red light on) S/U l Comments: 1 Page 5 of 8 LR000172 Rev. 7
1 SYNCIIRONIZE TIIE MAIN GENERATOR WITII TIIE GRID l
- Items are Critical Steps i
Event Control Step Element Standard Sat /Unsat b Set generator load Depresses the REFERENCE button S/U* Enters load setpoint of 300 Mwe on S/U* keypad Depresses ENTER button S/U* Depresses GO button S/U* Depresses HOLD button when both of S / U* the following conditions exist: j Positive Mwe shown on vertical board digital display l H 13-P820.B I-4.5, , TG MOTORING alarm clears j Comments: J 9 Turn off Places SYNC SELECTOR switch for S/U synchroscope breaker 4885 in OFF 1 l Comments: l t
-- l
! 10 Place the voltage Places the Voltage Stabilizer control S/U stabilizer in senice switch in the ON position Ternination Cue: Comments: The iermi::stion point of this JPM has been reached. RECORD Calculate JPM Completion Time: TERMINATION t TIME: JPM Terminat.on Time: JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) 2 Tage 6 of 8 LR000172 Rev. 7
SYNCHRONIZE TIIE MAIN GENERATOR WITH THE GRID JPM INFORMATION CARD Initial Conditions: All prerequisites for synchronizing and initially loading the generator have been met. PPM 2.5.7 Section 5.4 has been completed through Step 10. Cue: The CRS has directed you to synchronize the generator to the grid and apply initialload. l Tools / Equipment: None Safety Items: None 4 Task Number: RO-0323-N-TG Validation Time: 9 Minutes Prerequisiste Training: 82-RSY-0502-L5 Time Critical: No PPM
Reference:
2.5.7 Rev 25 Location: Simulator NUREG 1123 Ref: 262001 A4.04 (3.6/3.7) Itrformance Method: Perform Prepared or Revised by: Randy Guthrie Revision Date: 8/20/96 4 Page 7 of 8 LR000172 Rev. 7
SYNCIIRONIZE TI1E MAIN GENERATOR WITII TIIE GRID STUDENT JPM INFORMATION CARD Initial Conditions: All prerequisites for synchronizing and initially loading the generator have been met. PPM 2.5.7 Section 5.4 has been completed through Step 10. Cue: The CRS has directed you to synchronize the generator to the grid and apply initialload. Page 8 of 8 LR000172 Rev. 7
SYNCHRONIZE THE MAIN GENERATOR WITH THE GRID 262001A4.04 AC Electrical Distribution !
- Ability to manually operate and/or monitor in the control room synchronizing and paralleling of different AC supplies.
(3.6/3.7) ) Question: The procedure cautions against unnecessary delays in loading the generator to prevent turbine rotor cooling and heating of the last stage blades. What is the mechanism that causes the heating of the turbine blades? Answer: The flow of steam provides cooling to the last stage blades, with low steam flow past these blades, they windmill in the condenser - atmosphere where the friction of the blades moving causes them to heat up. i
Reference:
PPM 2.5.7 Comments: t l I l Question: l When would manual synchronization of the generator be performed and what are the potential problems? Answer: l With the Auto Synchronization feature INOP and with Shift Manager approval. Delays in the BPA supervisory system may result in l an out of phase synchronization of the generator.
Reference:
l PPM 2.5.7 l Comments: 1 l i
. . _ _ - _ . _ -~ _ - . . _ . _ _ _ _
5 WASHINGTON PUBLIC POWER SUPPLY SYSTEM INSTRUCTIONAL COVER SHEET PR(XiRAM T' TLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE OPERATE SLC BORON SYSTEM FOR RPV INJECTION LESSON LENGTil 5 Min MAXIMUM STUDENTS I INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No. OJT Guide PQD Code Rev. No. Simulator Guide PQD Code Rev. No. Student Handout PQD Code Rev. No. JPM PQD Code LR000217 Rev. No. 7 Checkoff Sheet PQD Code Rev. No. Exam PQD Code Rev. No. I DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Carl Golightly DATE 4/4/95 REVISED 13Y Randy Guthrie DATE 8/22/96 TliClINICAL REVIEW BY DNFE INSTRUCTIONAL, REVIEW BY DATE APPROVED BY DATE Training Manager Matnx Ugxiate Vision # WP Update WA
l l
- l OPERATE SLC IlORON SYSTEM FOR RPV INJECTION l
MINOR REVISION RECORD Minor Description Allbeted Entered Etibetive Manager Rev Number of Revision Pages By Date Approval i l l l 1
)
i i l l Page1of9 LR000217 Rev. 7
OPERATE SLC BORON SYSTEM FOR RPV INJECTION JPM SETUP 1 Simulator ICs: Any , Milfunctions: MOV RWU10 5,0,D RWCU-V-4 fails to auto isolate Overrides (Optional): 3 N/A
, Special Setup Instructions:
. The above malfunction must be inserted before the examinee starts the JPM if desired. 2 SLC injection will result in a power decrease that could interfere with other JPMs being evaluated during the same session.
, The examinee may not initially use a procedure for this JPM because it is an EOP action. SLC may be initiated without a procedure and then the proce$.r- may be referenced for verification.
i Task Standard: Both SLC pumps injecting boron into the RPV and key system parameters verified. s 1 J 8 1 t 5 Page 2 of 9 LR000217 Rev. 7
OPERATE SLC BORON SYSTEM FOR RPV INJECTION RESULTS OF JPM: Examinee (Please Print): Evaluator (Please Print): Overall (circle one) Exam Code Evaluation SAT / UNSAT Simulator IC Used Validation / Critical Time JPM Completion Time 5 Minutes / NA Conunents: Evaluator's Signature: Date: Page 3 of 9 LR000217 Rev. 7
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OPERATE SLC BORON SYSTEM FOR RPV INJECTION JPM CHECKLIST j
- Itents are Critical Steps Event Control Step Eleinent Standard Sat /Unsat RECORD START TIME:
Notei . - 1 Activate SLC system Removes both blanks from and reinserts S/U If desired to prevent :' both keys into SLC System Manual t auto closure of - Control Switches RWCU-V-4, enssre: g yg, g 3,jg Q Places both Manual Controls Switches S/U* ; in OPERATE at this time Note: All steps occur ht : H13 P603 unless otherwise noted ' Comments: 2 Check squib valves Checks both squib valves actuated as actuated follows: Wrifies white circuit ready lights S/U 1 extinguished Wrifies SLC-V-4A Loss Of Continuity S/U BlSI illuminated Verifies SLC-V-4B Loss Of Continuity S/U BISI illuminated Verifies annunciator H 13-P603. A7-6.2, S/U SLC DIV 1 OUT OF SERVICE alarming Verifies annunciator H13-P603.A8-6.8, S/U SLC DIV 2 OUT OF SERVICE alarming Comments: Page 4 of 9 LR000217 Rev. 7
OPERATE SLC BORON SYSTEM FOR RPV INJECTION
- Itents are Critical Steps i Event Control Step Element Standard Sat /Unsat
, 3 Ensure SLC storage Verifies SLC-V-1 A and SLC-V-IB are S/U i tank outlet valves are open (red lights on) opCH l Comments: i i
- 4 Ensure SLC pumps Ensures both SLC pumps started (red S/U
! start lights on) Conunents: 5 Check indications of Verifics SLC system pressure increases S/U SLC system flow into to GT reactor pressure as indicated on the reactor SLC-PI-600 Verifies SLC system flow reaches =86 gpm as indicated on SLC-FI-l Comments: Not'en 64 Recognizes failure of Informs CRS of failure of RWCU-V-4. S/U*
.This stbp N/A if - RWCU to isolate to isolate'-
malfunction was not :: inserted Role Play: Comments: As CRS acknowledge f failure to isolate and : i direct attempt to manually isolate ^ 1 RWCU' Page 5 of 9 LR000217 Rev. 7
l OPERATE SLC BORON SYSTEM FOR RPV INJECTION 4 3
- Iterns are Critical Steps
- Event Control Step Element Standard Sat /Unsat
] Note: 7' Isolates RWCU > ' At H13-P602, Manually isolates RWCU .; S / U* i-2
- This step N/A if -
gggg,g g Eitbcr:/
- Places Control switch for RWCU-V-t ;-
ind7 to close (green light on, red light olf) ^ l l Or:. Places Control switch for RWCU-V-1 to close (green light on. red light ofi) Comments: 4 i 1 i [ i Note: M Ensure RWCU auto At H13-P602, verifics RWCU-V-4 S/U l This step N/A if isolates closes (green light on. red light otD malfunction was 4 inserted Comments: T J - ' 8 Monitor reactor Verifies APRMs indicate decreasmg S/U j power decreasing reactor power Note: Comments:
. Power will not bc l lowering immediately l 1 after initiation j Cue:
After 5 minutes. a reactor power has
- dropped by 2%
i ! i i I 1 4 2 4 l Page 6 of 9 LR000217 Rev. 7
OPU(ATE SLC BORON SYSTEM FOR RPV INJECTION
- Items are Critical Steps i Event Control Step Element Standard Sat /Unsat ,
9 Ensure SLC storage I l Verifies SLC storage tank level S/U i l tank level decreasing dropping as indicated on SLC-LI-601 l Note: Comments: l SLC tank level will not be lowering l immediately aller initiation ] Cue: After 5 minutes, SLC l tank level has dropped by 450 gallons Termination Cue: The termination point of this JPM has been reached. RECORD Calculate JPM Completion Time: TERMINATION TIME: JPM Termination Time: JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) Page 7 of 9 LR000217 Rev. 7
i - OPERATE SLC BORON SYSTEM FOR RPV INJECTION JPM INFORMATION CARD Initial Conditions: An ATWS condition exists, PPM 5.1.2 has been entered. Cue: The CRS has directed you to initiate SLC injection with both SLC pumps per PPM 2.4.1, Section 5.3.. Tools / Equipment: None Safety itents: None 1 Task Number: RO-0245-N-SLC Validation Time: 5 Minutes
- Prerequisiste Training
- 82-RSY-0904-L1 Tinne Critical: No i
PPM
Reference:
2.4.1 Rev 14 Location: Simulator NUREG 1123 Ref: 211000A4.04 (4.5/4.6) Itrforinance Method: Perfinm Prepared or Revised by: Randy Guthrie Revision Date: 8/22/96 1 1 a 0 Page 8 of 9 LR000217 Rev. 7
- )
OPERATE SLC BORON SYSTEM FOR RPV INJECTION i STUDENT JPM INFORMATION CARD i I Initial Conditions: An ATWS condition exists, PPM 5.1.2 has been entered. Cue: The CRS has directed you to initiate SLC injection with both SLC
\
pumps per PPM 2.4.1, Section 5.3.. I l l i l Page 9 of 9 LR000217 Rev. 7
l i 1 , 1 OPERATE SLC BORON SYSTEM FOR RPV INJECTION 211000A4.04 Standby Liquid Control System 3 Ability to monitor and/or operate in the control rooni reactor power. (4.5/4.6) i Question: What are the requirements for initiation of Standby Liquid Control (SLC)? l Answer: j Performing PPM 5.1.2 prior to Wetwell temperature reaching 110 degrees F. 1 i
Reference:
)
PPM 5.0.10 ] Comments: } i 1 1 1 j l j Question: ; j AAcr SLC initiation, when can boron injection be stopped? j a 1 4 Answer: ! { SLC-TK-1 indicating less than 100 gallons or existing control rod pattern alone can always assure reactor shutdown l 4 3
Reference:
1 PPM 5.1.2 Reactor Power Leg i j Comments: t-l a I 1 i i 4 4 t _ - - - . , , n - r -
l l
\ !
o ---s l WASHINGTON PUBLIC POWER l W) SUPPIX SYSTEM INSTRUCTIONAL COVER SHEET PROGRAM TTTLE ' LICENSED OPERATOR /STA REQUALIFICATION TRAINING 1 i COURSE TITLE JOB PERFORMANCE MEASURE' LESSON TITLE STARTUP CONTROL ROOM VENTILATION
.)
LESSON LENG111 6Mm MAXIMUM STUDENTS 1 ; INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No.. ; OJT Guide PQD Code Rev. No. Simulator Guide PQD Code Rev. No. ' Student Handout PQD Code Rev. No. JPM PQD Code LR000209 Rev. No. 5 CheckotTSheet PQD Code Rev. No. Exam PQD Code Rev. No. DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED 13Y Sean Dallago DATE 9/7/92 REVISED BY Randy Gutbrie DATE 8/21/96 TEClINICAL REVIEW 13Y DATE INSTRUCTIONAL REVIEW HY DATE APPROVED 13Y DATE Traming Manager Matnx Update Vismn # WP Update W.\
STARTUP CONTROL ROOM VENTILATION i MINOR REVISION RECORD
"'# Description All ed Entered Effective Manager
** ""I" of Revision y I) ate Approval l
\
l
' age 1 of 8 LR000209 Rev. 5
STARTUP CONTROL ROOM VENTILATION JPM SETUP Simulator ICs: Any Mrifunctions: N/A l Overrides (Optional): RWB25C 1.0,D Control switch for WM A-FN-51 A overriden to OFF RWB26C 1.0,D Control switch for WM A-FN-51B overridden to OFF . After failure to start is reported to the CRS. , RWB25C CLR ) RWB26C CLR Special Setup Instructions: Stop the rurudng control room supply fans, WM A-FN-51 A and WM A-FN-51B by placing the control switches to OFF Task Standard: ' Control room ventilation started in accordance with PPM 2.10.3. i 1 l Page 2 of 8 LR000209 Rev. 5
STARTUP CONTROL ROOM VENTILATION RESULTS OF JPM: Examinee (Please Print): Evaluator (Please Print): Overall (circle one) Exam Code Evaluation SAT / UNSAT
' Simulator IC Used Validation / Critical Time JPM Completion Time 6 Minutes / NA Comments:
4 Evaluator's Signature: Date: Page 3 of 8 LR000209 Rev. 5
STARTUP CONTROL ROOM VENTILATION JPM CHECKLIST
- Itents are Critical Steps ;
Event Control Step Element Standard Sat /Unsat RECORD START TIME: 1 Note: 1 Set d'unper control Ensures the following controls switches All steps occur at switcl.cs to auto arein AUTO: H13-P826 unless WMA-AD-54 A2 (54B2) S/U otherwise noted WMA-AD-54 A I (54B 1) S/U WMA-AD-51 A1 (51B1) S/U Note: Conunents: If failure of WMA-FN-51 A or WMA-FN-SIB is desired. cnsure overrides are inscried 2 Start control room Places control switch for WMA-FN- S/U* recirc fan $1 A (51B) to ON Comments: Note: 3 Recognizes control Informs CRS of failed control room fan S/U* This step N/A if room recire fan overrides were not failure inserted l Role Play: Comments: As CRS, take responsibility for investigation. Note: Clear overrides at this time Cue: As CRS, report probbm has been identified and corrected, direct continuation of procedure Page 4 of 8 LR000209 Rev. 5
STARTUP CONTROL ROOM VENTILATION i l 1
- Itents are Critical Uteps Event Control Step Element Standard Sat /Unsat Note: 4 Start control room Places control switch for WMA-FN- S/U* .
This step N/A if recirc fan SI A (51B) to ON ; overrides were not , ! inserted l l Conunents: , 5 Confirm outside air Confirms WMA AD-51 A1 (51B1) S/U l damper opens for the automatically opens (red light on) ) running fan : Comments: l 1 l l 6 Place standby fan Places WMA-FN-51B (51 A) control S/U ; control switch in auto switch to AUTO Comments: I l l 7 Confirms outside air Confirms WMA-AD-51BI (51 Al) S/U damper remains . remains closed (green light on) closed for the standby fan Comments: ) l l l l l l i
.' 1 1
Page 5 of 8 LR000209 Rev. 5
STARTUP CONTROL ROOM VENTILATION
- Itenis are Critical Steps j Event Control Step Element Standard Sat /Unsat l
l 8 Ensures equipment is Wrifies annunciator H13-P826.PI-5.2 S/U j operating properly (P2.-5.3) CR FLTR 51 A (51B) dP HIGH/ LOW clears following fan start Contacts EO to check cooling coil S/U WMA-CC-51 A2 (5182)in service Role Play Comments: As EO acknowledge and report cooling coil in service Termination Cue: The termination point
- of this JPM has been l reached.
RECORD Calculate JPM Completion Time: TERMINATION TIME: JPM Termination Time: JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) l l l l 1 l l I l i
- )
I l l j Page 6 of 8 LR000209 Rev. 5
l l STARTUP CONTROL ROOM VENTILATION l JPM INFORMATION CARD Initial Conditions: Control Room HVAC is shutdown following brief maintenance on a section of common ductwork. 1 Valve and power supply checklists are complete.
- Prerequisites are met for Control Room liVAC system start.
i 1 j Cue: The CRS has directed you to start Control Room ventilation per PPM , 2.10.3 Section 5.1. 1 Tools / Equipment: None l l Safety Items: None s l l Task Number: RO-N98-N-CR-HVAC Validation Time: 6 Minutes Prerequisiste Training: 82-RSY-1304-L6 Time Critical: No j PPM
Reference:
2.10.3 Rev 25 IAcation: Simulator l NUREG 1123 Ref: 290003Ga09 (3.6/3.5) Itrformance Method: Perfonn ! Prepared or Revised by: Randy Guthrie Revision Date: 8/21/96 i 1 4 e I i i l 1 1 1 d 1 Page 7 of 8 LR000209 Rev. 5
_ . . . _ .. .- . . _ _ _ - _ . . . - _ . ~ . . _ - - - . - . _ - - - _ . - _ - . - _ . - . . . . - . _ . - . -
, , t
( I STARTUP CONTROL ROOM VENTILATION l STUDENT JPM INFORMATION CARD ; Initial Conditions: Control Room HVAC is shutdown following brief maintenance on a section of i common ductwork. l Valve and power supply checklists are complete. j l L Prerequisites are met for Control Room flVAC system start. l ' l Cue: l
}
! 1he CRS has directed you to start Control Room ventilation per PPM l 2.10.3 Section 5.1. I i i i t i I i i l l l l l l l 1 Page 8 of 8 LR000209 Rev. 5
_ _ . _ . ._._ . _ _ . . . _ . _.. _ _ _ _ _ . . _ - _ . . _ _ _ - _ . _ . _ . . . . ._______.______._-..m. i j~ e O STARTUP CONTROL ROOM VENTII.ATION I 290003GA09 Control Room HVAC
- Ability to locate and operate components, including local controls. -
J (3.6/3.5) 1 1 l l Question: What permissives are required to be satisfied for WMA-TIC-II A to control temperature in the control room? -
- I Answer: ,
Emergency Chill Water Pump 1 A control switch in " AUTO", and Recirculation Fan, WMA-FN 1 A, "ON" or has " AUTO STARTED", and "RUN/ AUTO" switch on HVAC local control rack in the " AUTO" position, and Emergency Chill Water system pressure GE 100psig.
Reference:
. EWD-84E-0017 i Comments: I i i i } ! ) 4 Question: l The FSAR requires control room temperature to be maintained 72
- F to 78
- F (Normal). Why is the Technical Specification Allowable Value for control room temperature <104
- F (Emergency)? j i
Answer: 104
- F is based on Standby Senice Water supply to the air handling units (RadWaste chillers INOP).
- l 2
Reference:
1 FSAR 6.4.2.2 5 l Comments: 1 4 i l l l
a v; se WASHINGTON PUBLIC POWER SUPPLY SYSTEM INSTRUCTIONAL COVER SHEET l l l'R(XiRAM TTILE LICENSED OPERATOR /STA REQUALIFICATION TRAINING COURSE TITI.E JOB PERFORMANCE MEASURE i I LESSON TITLE RESTART OF RPS-MG-1 l LESSON LENGTil 6 Min MAXIMUM STUDENTS I INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No. OJT Guide PQD Code Rev. No. Simulator Guide PQD Code Rev. No. Student IIandout PQD Code Rev. No. JPM PQD Code LR000251 Rev. No. 4 Checkoff Sheet PQD Code Rev. No. Exam PQD Code Rev. No. DIVISION Till,E Nuclear Training DEPARTMENT Operations Training PREPARED BY Donald liughes DATE 4/14/95 REVISED BY Randy Guthrie DATE 8/22/96 l TECIINICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DNITi APPROVED BY DATE l Traimng Manager Matrix Uplate Vision # 729 WP Uplate WA
RESTART OF RPS-MG-1 MINOR REVISION RECORD Minor Description Atrected Entered Etrective Manager Rev Number ofRevision Pages By Date Approval j l l l l l i l l l l l l l Page1of8 LR000251 Rev. 4
I i 1 RESTART OF RPS-MG-1 l JPM SETUP Simulator ICs: N/A Malfunctions: N/A i Overrides (Optional): i N/A Special Setup Instructions: N/A a Task Standard: ! Simulate steps winich wuld restart RPS-MG-1 per PPM 2.7.6. j i l i l l l l Page 2 of 8 LR000251 Rev. 4
RESTART OF RPS-MG-1 RESULTS OF JPM: Examinee (Please Print): Evaluator (Please Print): Overall (circle one) Exam Code ! Evaluation SAT / UNSAT Simulator IC Used Validation / Critical Time JPM Completion Time 6 Minutes / NA l Comments: i i 4 i a 1 1 1 n 4 4 Evaluator's Signature: Date: Page 3 of 8 LR000251 Rev. 4
i l RESTART OF RPS-MG-1 JPM CIIECKLIST l
- Itents are Critical Steps Event Control Step Element Standard Sat /Unsat RECORD START TIME:
1 Note: 1 Ensure MG supply At MC7A/IB, ensures RPS-Bus Mir S/U All steps occur in Div breaker is closed Gen. Set MG-1 switch is closed i l A RPS Motor (vertical position) Generator (MG) room at RW Building 46T Cue: Comments: Switch is in vertical i position 1 2 Ensure the motor is At C72-S001-A. ensures green S/U l oft MOTOR OFF light is lit i t Cue: Comments: Green light is lit 3 Ensure MG output At RPS-MG 1 panel, ensures Generator S/U breaker is open Output breaker is in OFF Cue: Comments: Breaker is open l l 1 1 l i Page 4 of 8 LR000251 Rev. 4
RESTART OF RPS-MG-1
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat 4 Start the RPS MG set Depresses and holds MOTOR ON S / U*
pushbutton Ensures green MOTOR OFF light S/U cxtinguishes and red MOTOR ON light is lit Releases MOTOR ON pushbutton when S/U MG speed increase stops Cue: Conunents: Red light is on, green light is off, motor pitch is constant Cue: When examince checks voltage, no voltage is indicated 5 Reset the overvoltage At RPS-MG-1 panel, momentarily S/U* trip depresses then releases the MOTOR ON pushbutton Cue: Comments: Button is depressed 6 Ensure voltage Ensures voltage stable at 120 VAC on S/U stabilizes at 120 VAC AC-VOLTS meter Cue: Comments: Point out voltage stable at 120 VAC on AC-VOLTS meter t Page 5 of 8 LR000251 Rev. 4
RESTART OF RPS-MG-1
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat l 7 Close RPS MG At RPS-MG-1 panel, places S/U* )
j output breaker GENERATOR OUTPUT breaker in ON ' (pushed up) l Cue: Comments: Breaker is closed l Termination Cue: The termination point of this JPM has been reached. RECORD Calculate JPM Completion Time: TERMINATION TIME: JPM Tennination Time: JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) I l l I l l Page 6 of 8 LR000251 Rev. 4 j I
RESTART OF RPS-MG-1 l JPM INFORMATION CARD I Initial Conditions: RPS Division A has been de-energized due to a spurious fault. The fault has been identified and corrected. Cue: The CRS has directed you to restart RPS-MG-1 per PPM 2.7.6. l Tile PERFORMANCE OF Tills JPM WILL BE SIMULATED. i CONTROL MANIPULATIONS WILL NOT BE PERFORMED. l Tools / Equipment: None i Safety Itcins: None Task Number: RO4)247-N-RPS Validation Time: 6 Minutes Prerequisiste Training: N/A Time Critical: No PPM
Reference:
2.7.6 Rev ii IAcation: Plant NUREG 1123 Ref: 212000GA9 (4.2/4.2) Itrformance Method: Simulate Prepared or Revised by: Randy Guthrie Revision Date: 8/22/96 i I l l I l Page 7 of 8 LR000251 Rev. 4
RESTART OF RPS-MG-1 STUDENT JPM INFORMATION CARD Initial Conditions: RPS Division A has been de-energized due to a spurious fault. The fault has been identified and corrected. Cue: The CRS has directed you to restart RPS-hiG-1 per PPM 2.7.6. Tile PERFORhfANCE OF Tills JPAf WILL BE SinfULATED. CONTROL AfANIPULATIONS WILL NOT BE PERFORAfED. I l l l l l i Page 8 of 8 LR000251 Rev. 4
RESTART OF RPS-MG-1 212000GA9 Reactor Protection System Ability to locate and operate components, including local controls (4.2/4.2) Question: : Following an RPS actuation, the scram signal cannot be reset for 10 seconds after the mode switch is respositioned to SHUTDOWN. l What is the purpose of this time delay? i Answer: Prevents the scram valves from closing prior to full rod travel.
Reference:
RPS System Text Comments: i 1 Question: What is the purpose of the EPA breakers in the power supply to the RPS busses? Answer: I Protects the Class IE power supply from the Non-Class 1E RPS busses. I I l
Reference:
1 RPS System Text l Comments: l l
, f. WASHINGTON PUBLIC POWER W) SUPPLY SYSTEM ! INSTRUCTIONAL COVER SHEET PR(XiRAM TRLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING CotJRSE TITLE JOB PERFORMANCE MEASURE 1.ESSON TITLE REDUCE SUPPRESSION POOL LEVEL FROM THE RSD PANEL LESSON LENGTH 5 Min MAXIMUM STilDENTS I INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No. OJT Guide PQD Code Rev. No. Simulator Guide PQD Code Rev. No. Student Handout PQD Code Rev. No. JPM PQD Code LR000145 Rev. No. 7 l CheckofTSheet PQD Code Rev. No. ! Exam PQD Code Rev. No. DIVISION TITLE Nuclear Training DElWRTMENT l Operations Training PREPARED BY Roger Scott DAE 9/20/93 REVISED BY Randy Guthric DATE 8/20/96 TEClINICAI, REVIEW BY DATE INSTRlJCTIONAL REVIEW BY DATE APPROVED BY DATE
'trairung Manager Matrix Update Vision # WP Uguate W5 l
i l REDUCE SUPPRESSION POOL LEVEL FROM TIIE RSD PANEL MINOR REVISION RECORD l 1 Minor Description Au'ected Entered Etrective Manager i Rev Number ofRevision Pages By Date Approval I i l I l I 1
)
l 1 l 1 1 l 1 l Page1 of8 LR000145 Rev. 7
i l REDUCE SUPPRESSION POOL LEVEL FROM TIIE RSD PANEL i JPM SETUP l l l Simulator ICs: l N/A l Malfunctions / Remote Triggers: l N/A Overrides (Optional): N/A Special Setup Instructions: N/A Task Standard: Remote Shutdown Panel control manipulations are performed in accordance with plant procedures to reduce suppression pool level, i Page 2 of 8 LR000145 Rev. 7
REDUCE SUPPRESSION POOL LEVEL FROM Tile RSD PANEL l RESULTS OF JPM: Examinee (Please Print): l Evaluator (Please Print): Overall (circle one) Exam Code Evaluation - SAT / UNSAT Simulator IC Used Validation / Critical Time JPM Completion Time N/A 5 Minutes / NA Comments: I I Evaluator's Signature: Date: Page 3 of 8 LR000145 Rev. 7
REDUCE SUPPRESSION POOL LEVEL FROM TIIE RSD PANEL. JPM CHECKLIST
- Itenis are Critical Steps Event Control Step Element Standard Sat /Unsat RECORD START TIME:
1 Ensures RHR-B is in Ensures RHR-B is in Suppression Pool S/U Suppression Pool Cooling per Section 5.9 Cooling Cue: Comments: RHR-B is in Suppression Pool Cooling per PPM 4.12.1.1 Section 5.9 2 Notifics Radwaste Notifies Radwaste that water is to be S/U that water is being transferred to Radwaste from the transferred Suppression Pool with RHR Cue: Comments: Radwaste acknowledges transfer of water from the Suppression Pool with RHR 3 Locally opens R!lR- Directs Ops 2 to open RHR-V-40 using S/U+ V-40 to =25% the handwheel to =25% at RB 548 D HX RM Cue: Comments: Ops 2 reports that l RHR-V-40 is open 1
=25%
i l 1 Page 4 of 8 LROT)0145 Rev. 7
. - ~ . . - - - _ _ . . .- . . - - - . . - - . _ - _ . . _ . _ - - . . - .-
REDUCE SUPPRESSION POOL LEVEL FROM Tile RSD PANEL
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat 4 Opens RiiR-V-49 Places control switch for RiiR-V-49 to S/U*
, the OPEN position (Red light on. green l light ofi) Note: Comments: l ! Procedure allows l increasing reject flow , by throttling R11R-V-24B closed and opening RiiR-V-40 GT 25% r Cue: Red light is on above RHR-V-49 control switch L Cue: i Suppression Pool Level is slowly , lowering ! Cue: The CRS directs you to increase rate of reject , 5 Throttles closed Places control switch for RHR-V-24B S/U RIIR-V-24B to CLOSE and releases Cue: Conunents: Green and red lights are lit above RIIR V-24B control switch Cue: Suppression pool les el l is now lowering i l Cue: When suppression pool level is rechecked, repon level now at -2" 6 Reopens RifR-V-24B Places control switch for RiiR-V-24B S/U to OPEN and holds until fully open Cue: Comments: Red light is ht above RHR-V-24B control switch l l Page 5 of 8 LR000145 Rev. 7
REDUCE SUPPRESSION POOL LEVEL FROM Tile RSD PANEL
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat 7 Closes R11R-V-49 Places control switch for RHR-V-49 to S/U*
the CLOSE position (Green light on, red light oIT) Cue: Comments: l Green light is lit above ! RiiR-V-49 r:ontrol l switch l 8 Leaves RHR-V-40 as- Directs Ops 2 to leave RHR-V-40 as-is S/U i is , l Cue: Comments: Ops 2 acknowledges leave RHR-V-40 as-is Termination Cue: The termination point of this JPM has been ; reached. I RECORD Calculate JPM Completion Time: TERMINATION TIME: JPM Termination Time: JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3)
\
Page 6 of 8 LRO-)0145 Rev. 7
l REDUCE SUPPRESSION POOL LEVEL FROM TiiE RSD PANEL JPM INFORMATION CARD Initial Conditions: The Control Room has been abandoned due to a fire in the back panels. The Remote Shutdown Panel is manned with Div-2 equipment OPERABLE. Conditions are satisfied for the use of shutdown cooling EXCEPT that suppression pool level is high and must be lowered to -2". RHR-B is in suppression pool cooling. Cue: 1 The CRS has directed you to reduce suppression pool level to prepare ! for shutdown cooling using PPM 4.12.1.1 starting at Section 5.10. Tools / Equipment: None Safety items: None Task Number: RO4)117-A-RSP Validation Time: 5 Minutes SRO-0251-A-RSP Prerequisiste Training: 82-RSY-1304-L3 Time Critical: No PPM
Reference:
4.12.1.1 Rev 24 Location: Plant NUREG 1123 Ref: 219000A4.13 (3.9/3.8) Performance Metitod: Simulate I Prepared or Revised by: Randy Guthrie Revision Date: 8/20/96 l l l l l Page 7 of 8 LR000145 Rev. 7
REDUCE SUPPRESSION POOL LEVEL FROM THE RSD PANEL STUDENT JPM INFORMATION CARD Initial Conditions: The Control Room has been abandoned due to a fire in the back panels. l The Remote Shutdown Panel is manned with Div-2 equipment OPERABLE. . l ( Conditions are satisfied for the use of shutdown cooling EXCEPT that suppression [ pool level is high and must be lowered to -2". j R11R-B is in suppression pool cooling. l l Cue: The CRS has directed you to reduce suppression pool level to prepare , l for shutdown cooling using PPM 4.12.1.1 starting at Section 5.10. l i l i i l f l f l j Page 8 of 8 LR000145 Rev. 7
.. .. . ~. . . - - . . - - .. - ~ . - - - - - . - ~ . - . . . . . _ . ~ . - . - . . . .
l REDUCE SUPPRESSION POOL LEVEL FROM THE RSD PANEL i 21900A4.13 RHR/LPCI: Torus / Suppression Pool Cooling Mode !
?
Ability to manualy operate and/or monitor in the control room suppression pool level. (3.9/3.8) ; I Question: ' When initiating RHR shutdown cooling, suppression pool level is recommended to be low. Why is this necessary? i Answer Initial system alignment /startup will add water to the suppression pool and if not lowered initially, may cause level to rise above EOP cntry conditions. (+2") i
Reference:
PPM 4.12.1.1 [ Comments: ! Question:
- With suppression pool level off-scale low, a manual determination oflevel is required using a conversion factor for pressure. Which ;
instrument is preferred and why? l Answer: ,. l LPCS local suction pressure instrument. The range of this instrument is 0 - 30 and is more accurate for coversion purposes than other ECCS suction pressure indications.
Reference:
PPM 4.12.1.1 l Comments: l
i
', l
$ wasmucrou rusuc rowra l SUPPLY SYSTEM i INSTRUCTIONAL COVER SHEET ),
t PROGRAM TITLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING
- COURSE TITLE j JOB PERFORMANCE MEASURE l
LESSON Tm.E PERFORM MANUAL START OF HPCS DG FROM LOCAL PANEL ): 1 LESSON LENGTil 17 Min MAXIMUM STUDINTS I $ INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No.
- OJT Guide PQD Code Rev. No.
j Simulator Guide PQD Code Rev. No. j Student Handout PQD Code Rev. No. JPM PQD Code LR000199 Rev. No. 7 {
- Checkoff Sheet PQD Code Rev. No.
Exam PQD Code Rev. No. DIVISION TITLE Nuclear Training DEPARTMENT Operations Training ] PREPARED BY Randy Guthrie DATE 8/3/95 l l REVISED IlY Randy Guthric DATE 8/21/96 i 1 1 1ECIINICAL REVIEW llY DATE l INETRUCTIONAL REVIEW !!Y DATE 1 APPROVED I!Y DATE Trammg Manager i Matrix Update Vision # WP U ixlate W.\ i
PERFORM MANUAL START OF IIPCS DG FROM LOCAL PANEL MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages fly Date Approval 1 l l l Page 1 of 10 LR000199 Rev. 7
1 , 1 l PERFORM MANUAL START OF IIPCS DG FROM LOCAL PANEL j JPM SETUP Simulator ICs: N/A Malfunctions: N/A Overrides (Optional): 1 N/A l Special Setup Instructions: N/A , l i Task Standard. . The HPCS diesel generator is staned kwally per PPM 2.7.3, Section 5.6. 1 i l l l i l i l Page 2 of 10 LR000199 Rev. 7
l PERFORM MANUAL START OF IIPCS DG FROM LOCAL PANEL RESULTS OF JPM: Examinee (Please Print): Evaluator (Please Print): Overall (circle one) Exam Code Evaluation SAT / UNSAT l Simulator IC Used Validation / Critical Time. JPM Completion Time 17 Minutes / NA l Conunents: a.
\
_ _ . i i Evaluator's Signature: Date: Page 3 of 10 LR000199 Rev. 7
PERFORM MANUAL START OF llPCS DG FROM LOCAL PANEL JPM CHECKLIST
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat RECORD START TIME:
Note: I Verify alignment of Ensures Unit Mode Sci Sw is in S/U All steps occur at the Unit Mode Selector MAINT engine mounted Switch control panel E-CP-Ensures green light lit S/U DG/EP3 unless otherwise stated Cue: Comments: Unit mode selector switch is in MAINT and the green light is lit 2 Ensure annunciators At E-CP-DG/RP3 ensures annunciator S/U are clear alanns are clear except drop 1.1 and 3.1 Contacts Control Room and verifies S/U annunciator alarms cicar on lil3-P601 except drop 6.8 Cue: Comments: All alarms are clear except drop I.1,3.1 and drop 6.8 at H13-P601 3 Place DG mode Contacts Control Room to have Diesel S/U* selector switch in Generator Mode Selector Switch at local fil3-P601 place in the LOCAL position Cue: Comments: DG mode selector j switch is in LOCAL l 4 Shut down dicsci At E-CP-DG/CP3 opens f3KR 1 S/U generator space heater Cue: Comments: BKR 1 is open Page 4 of 10 LR000199 Rev. 7
PERFORM MANUAL START OF HPCS DG FROM LOCAL PANEL
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat 5 Lowers hydraulic Holds hydraulic governor control S/U*
governor setpoint 13 switch to LOWER until governor motor minimum stops (located on top of governor box on the dicsci) then releases switch Cue: Comments: Hydraulic governor setpoint is at minimum 6 Starts the HPCS DG At E-CP-DG/EPl depresses dicsci S/U* UNIT START pushbutton Cue: Comments: The IIPCS DG is running Cue: If the operator checks the red 40,250,450 and RI relay lights, they are lit 7 Confirms diesel Checks HPCS-St-DG3 to confirm DG S/U l accelerates normally speed is 425-475 rpm a Cue: Comments: $ The HPCS DG is ! running at 450 rpm Note: l Low air pressure may l alarm following DG ! start,lov water i pressure should bc
- alarmed and will clear at =700 rpm 8 Ensures air start Ensures air start motors disengaged S/U
- motors auto either visaully or audibly disengage Cue
- Comments ~
Air start motors are disengaged Note: Procedure directs diesel shutdown if start motor engagement is detected Page 5 of 10 LR000199 Rev. 7
PERFORM MANUAL START OF IIPCS DG FROM LOCAL PANEL
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat 9 Verify adequate oil in Ensures adequate oil in the starting air S/U starting air in-line in-line lubricator using the method lubricators directed in Attachment 6.5 Cue: Comments:
There is adequate oil in the starting air in-line lubricators 10 Checks governor oil Ensures governor oil level has dropped S/U level to normal operating band using the method directed in Attachment 6.3 Cuc: Comments: Governor oil level is normal Cue: 11 Ensure DSA-F-3 dP Ensures DSA-DPI-IC reads LT 10 psid S/U DSA-C-IC is running is <10 psid Cue: Comments: DSA-C-1C reads 6 psid 12 Informs Control Contacts Control Room Supervisor to S/U Room of DG start log DG start Cue: Comments: CRS acknowledges start and has classified and logged the start Cue: 13 Checks crankcase oil Checks crankcase oil level with dipstick S/U The DG has been level Logs level in Attachment 6.3 S/U idling normally for 7 minutes and the oil is now hot Cue: Comments: CrMease oil level has been logged in Attachment 6.3 1 1 Page 6 of 10 LR000199 Rev. 7
J+a----,- - - _ = di, J Jm. m J .. ,.J W_..a 34 ed.sMJ-. s . - . . - 4.. -4 ..,*__,3 -
..t__a-_a .- ,.a_ u 4-.- -- .-w-_+b - .&- - - - m - e PERFORM MANUAL START OF IIPCS DG FROM LOCAL PANEL Items are Critical Steps i
Event Control Step Element Standard Sat /Unsat
, 14 Checks starting air Checks DSA-PI-il and DSA-PI-12 S/U pressure indicating less than 206 psig
}
! Cue: Conunents:
Starting air pressure
- iridicating 190 psig on 3 both indicators i
1 Cue: 15 Raises engine speed Uses hydraulic governor control to set S/U* l The DG has been to 925 rpm engine speed to 925 rpm as indicated
; operating for 12 on HPCS-SI-DG3
]
i minutes at idle ' j
)
4 Cue: Conunents: l . IIPCS DG speed , indicates 925 rpm l i ! 3 16 Align Unit Mode At E-CP-DG/EP3 places Unit Mode Sci S/U* Selector Switch Sw in AUm i ] l Obscives red light illuminates S/U i Cue: Conunents: I Unit mode selector i suitch isin AUTO i and the red light is ; illuminated ! l 17 Ensures generator Ensures liPCS Supply Voltmeter Source S/U f voltage is cortcet Selector switch in the GEN position Ensures HPCS-VM-DG3/SM4 reads S/U 3740 - 4580 VAC l 8 l Cue: Conunents: Generator voltage indicates 4000 VAC 1 1 l l Page 7 of 10 LR000199 Rev. 7
- - -- . - . . - .- - - - -. . . _ _ -- - . . . _ _ - . . - . - = . - - - - - - ; . PERFORM MANUAL START OF IIPCS DG FROM LOCAL PANEL
- Iterns are Critical Steps i
Event Control Step Elernent Standard Sat /Unsat IM Ensures proper Ensure DEA-FN-31 and DMA-FN-31 S/U i ventilation in dicsci running a roofn l Cue: Comments: i j Diesel generator room ventilation fans are
; running l Cue:
1 Normal dicsci engine j operation has been verified using Attachement 6.3 l Termination Cue: . The termination point of this JPM has been reached. RECORD Calculate JPM Completion Time: TERMINATION . 4, TIME: JPM Termination Time: JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) i i l J l i l 5 Page 8 of 10 LR000199 Rev. 7
^ PERI'ORM MANUAL START OF HPCS DG FROM LOCAL PANEL JPM INFORMATION CARD Initial Conditions: A rnanual start of DG-3 is in progress and PPM 2.7.3 Section 5.6 has been completed through Step 17 l The plant is in a NON-EMERGENCY condition, SM-2 is energized from TR-S DG-3 Governor droop switch is in DROOP. IIPCS service water is in service. J Generator space heater breaker is open. All liPCS diesel generator annunciators are clear except for fil3-P601. Al.6-8, liPCS System Out Of Service. Cue: The CRS has directed you to continue the DG-3 local start using PPM 2.7.3, Section 5.6. CONTROL AIANIPULATIONS WILL NOT BE PERFORAfED. ALL ACTIONS AND STEPS i/LL BE SIAIULATED. Tools / Equipment: None { } Safety Items: None Task Number: RO-0706-N-DGilP Validation Time: 17 Minutes Prerequisiste Trainir.g: 82-RSY-1305-L5 Time Critical: No PPM
Reference:
2.7.3 Rev 26 Location: Plant NUREG 1123 Ref: 264000A4.04 (3.7/3.7) Performance Method: simulate Prepared or Revised by: Randy Gutitrie Revision Date: 8/21/96 Page 9 of 10 LR000199 Rev. 7
1 I PERFORM MANUAL START OF llPCS DG FROM LOCAL PANEL l STUDENT JPM INFORMATION CARD Initial Conditions: A manual start of DG-3 is in progress and PPM 2.7.3 Section 5.6 has been ! completed through Step 17 l The plant is in a NON-EMERGENCY condition, SM-2 is energized from TR-S 1 DG-3 Gove nor droop switch is in DROOP. l IIPCS scrvice water is in service. Generator space heater breaker is open. All HPCS diesel generator annunciators are clear except for H13-P601. Al.6-8, HPCS System Out Of Service. Cue: The CRS has directed you to continue the DG-3 local start using PPM 2.7.3, Section 5.6. CONTROL MANIPULATIONS WILL NOT BE PERFORMED. ALL ACTIONS AND STEPS WILL BE SIMULATED. I Page 10 of 10 LR000199 Rev. 7
j PERFORM MANUAL START OF HPCS DG FROM LOCAL PANEL i 2I4000A4.04 Emergency C.enerators (Diesel / Jet)
; Ability to manualif 9perate and/or monitor in the control room manu al start, loading, and stopping of
, emergency generator Plant Specific). (3.7/3.7) T Question: With the generator synchronized, what are the indications of an underexcited geacrator and what actions would you take for this ! condition? i 'j Answer: l KVAR meter deflected downscale, immediately unload the generator and trip the output breaker. ] ]
Reference:
i PPM 2.7.3 l Comments: i l 2 i i 4 4
- i
. I i i , Question: i Why is SM-2 transferred to TR-N prior to synchronizing DG-37 J l Answer: l To prevent parallelling DG-3 with the Main Generator ; I
Reference:
PPM 2.7.3 l l Comments: 0 l 1 l i i l 1
} , f e. WASHWGTON PUBLIC POWER SUPPLY SYSTEM INSTRUCTIONAL COVER SHEET PROGRAM TRLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING COURSE TMLE JOB PERFORMANCE MEASURE 1.ESSON TITLE GENERATOR CAPABILITY CURVE INTERPRETATION LESSON LENGTil 5 Min MAXIMUM STUDENTS I INSTRUCTIONAL MATERIALS INCLUDED ,
Lesson Plan PQD Code Rev. No. _ , OJT Guide PQD Code Rev. No. Simulator Guide PQD Code Rev. No. Student Handout PQD Code Rev. No. JPM PQD Code Rev. No. Checkoff Sheet PQD Code Rev. No. Exam PQD Code Rev. No. DIVISION TMlli Nuclear Training i del %RTMENT Operations Training ) PREl%REDI3Y Randy Guthrie DATE 8/23/96 REVISED BY DATE TECilNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW 13Y DATE APPROVED BY DATE Traimng Manager Matrix Update Vision # WP U lxlate WA
GENERNFOR CA PABILITY CURVE INTERPRETATION MINOR REVISION RECORD Minor Description Atrected Entered Etrective Manager Rev Number ofRevision Pages fly Date Approval i l i 1 l l l r i Page I of 7 Rev.O
. . - . . _ . - - . - _ - - - . - ~. .- -
GENERATOR CAPABILITY CURVE INTERPRETATION
- p t JPM SETUP Simulator ICs:
14 Malfunctions: N/A Overrides (Optional): N/A Special Setup Instructions: Trip the "A" RRC pump and alku plant conditions to stabilize. Ensure amiuniciato/ 4.800.C3 8-7, Alain Generator Overexcitation alarm is in Task Standard: Succesful determination of generator operating mnditions on the generator capability curve. Page 2 of 7 Rev.O
i
, ; ", GENERATOR CAPABILITY CURVE INTERPRETATION RESULTS OF JPM:
Examinee (Please Print): Evaluator (Please Print): Overall~ (circle one) Exam Code Evaluation . SAT / UNSAT Simulator IC Used Validation / Critical Time JPM Completion Time 5 Minutes / NA Comments: 1 1 l
\
4 l I J i Evaluator's Signature: Date: Page 3 of 7 Rev.O
GENERATOR CAPABILITY CURVE INTERPRETATION
, i JPM CIIECKLIST t
- Items are Critical Steps Event Control Step Element Standard Sat /Unsat RECORD START TIME:
1 Locates Generator Locates Generator Capability Curve: S/U Capabilty Curve (PPM 2.5.7 Attachment 6.6) Comments: 2 Determines generator Determines generator hydorgen S/U hydrogen pressure pressure: (H13-P820 H2-PI-1) Comments: l 3 Determines generator Determines generator MWE S/U l MWe (fil3-P820, Deli Panel or H13-P800 Indicators or Recorders) Comments: 4 Determines generator Determines generator MVAR S/U reactive load (Hl3-P820 DEH Panel or H13-P800 Indicators or Recorders) Comments: l Page 4 of 7 Rev.O
GENERATOR CAPABILITY CURVE INTERPRETATION
, n
- Itents are Critical Steps Event Control Step Element Standard Sat /Unsat 5 Determines plant Determines plant operating conditions S/U*
operating conditions on generator capability curve on Generator (PPM 2.5.7 Attachment 6.6) Capability curve Termination Cue: Comments: The termination point of this JPM has been reached. RECORD Calculate JPM Completion Time: TERMINATION TIME: JPM Termination Time:- JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) , i l l l Page 5 of 7 Rev.O
, ; i.
GENERATOR CAPABILITY CURVE INTERPRETATION JPM INFORMATION CARD Initial Conditions: The plant was operating at 100% rated thermal power when RRC pump "A" tripped. l Operators are responding to the pump trip.. Annunciator H13-P800.C3.8-7, MAIN GENERATOR OVEREXCITATION has just been received. Cue: The CRS has directed you to determine generator operating conditions on the Generator Capability Curve. Tools / Equipment: None Safety Items: None Task Number: Validation Time: 5 Minutes Prerequisiste Training: Time Critical: No ; PPM
Reference:
2.5.7 Rev 25 Location: Simulator 4.800.C3 8-7 Rev 5 NUREG II23 Ref: 245000Kl.01 (3.2/3.3) Pierformance Method: Perform ) 245000A4.02 (3.1/2.9) Prepared or Revised by: Randy Guthrie Revision Date: 8/23/96 I Page 6 of 7 Rev.O
, t GENERATOR CAPABII ITY CURVE INTERPRETATION STUDENT JPM INFORMATION CARD Initial Conditions: The plant was operating at 100% rated thermal power when RRC pump "A" tripped.
Operators are responding to the pump trip.. Annunciator H13-P800.C3.8-7, MAIN GENERATOR OVEREXCITATION has just been received. Cue: l The CRS has directed you to detennine generator operating conditions on the Generator Capability Cun'e. l l 1 l ; l l t l l l 1 1 I l l Page 7 of 7 Rev.O
I l J l l 8-7 MAIN GENERATOR OVEREXCITATION 8-7 WINDOW SOURCE AUTOMATIC ACTIONS j E-RLY-59/81G1 (1.26 V/HZ) GE 1.32 V/Hz causes a MAIN GENERATOR , , generator trip after a time OVEREXC1TATION delay. NOTE: This alarm is activated by a volts /hz relay and indicates either the generator voltage is high , or the frequency is low. l
- 1. Confirm V/Hz alarm by reading the digital V/Hz value on E-RLY-59/81G1 (Bd F). l !
I
- 2. If this alarm occurs during Main Turbine Generator Synchronizing with Grid or Main Turbine Generator normal shutdown, ensure indication on E-RLY-59/81G1 (H13-P842) is l LT 1.32 V/HZ and continue with PPM 2.5.7, Main Turbine Generator. !
HOIE: If system frequency is normal, then this relay is set to alarm at 105% (26,250 V, l ," 1.26 V/Hz) of generator rated voltage, and a generator trip at 110% (27,500 V,1.32 V/Hz) of generator rated voltage with a 10 minute time delay. When synchronizing the unit, it may be necessary to raise generator voltage above the alarm set point to match it with BPA 500 KV system ! voltage. This gives the Main Generator Overexcitation alarm, which is not a problem as long as the generator voltage is not kept above the trip set point (1.32V/Hz) for over 10 minutes, without . synchronizing the unit (see table below). 4 i
- 3. Determine if the alarm is due to high voltage or low frequency.
- 4. If the terminal voltage is high, lower the voltage using either the Main Generator Exciter i Voltage Adjuster or the Main Generator Exciter Base Adjuster.
- 5. If the frequency is low adjust the speed of the Main Turbine to 1800 rpm. (N/A if generator is syncronized) ;
i N/Hz Recuired TD Tolerance ; Alarm Point SA 1.26 60 See i 1 See Trip Point SI 1.32 600 See t 7 See Trip Point S2 1.35 300 Sec i 4 Sec Trip Point S3 1.38 120 See i 2.2 Sec Trip Point S4 1.41 50 Sec i 2.5 Sec Trip Point S5 1.44 6 See i 1.18 Sec Trip Point S6 1.50 2 See i 1.06 See
REFERENCES:
EWD-51E-0044 E512-1 E521-11 PROCEDURE NUMBER REVISION PAGE 4.800.C3 5 44 of 56
.' Riference SWP-PRO-o2 \ CONTROL NUMBER O iG"pTfg TEMPORARY CHANGE NOTICE 33 337 COMPLETED BY ORIGINATOR 2, ! ' 3.) POC Review Required? b rocedure P No: 2.5.7 * ' Current Rev. 25 l @ Yes O No (p
Title:
_MAtM TURRlNF nENFRATOR
@ C One-Time-Only. TCN NOT incorporated into procedure
@) Procedures pages affected by thrs TCN: 37 , J j, fD Control No. of incorporated Deviation /TCN: N/A
@ Procedure pages affected by previous Deviation /TCN: N/A
@ Summary of Changesy , %
Changed Note on p. 37 to allow use of MSR second stage reheat at less than 660 MWe on recomrnendation from system engineer. C Contmued I certify this procedure TCN is in compliance with the criteria on page 2 pf is f m nd SgW -PkO- 2 Onginated by: M. R. Johnson / L y w(, /7 7 ['[
/ [/ S ;
aol Pnnt Name j Sig'riature U Date l
.fdTERI APPROVAL OF TCN (SRO Approval required for TCNs P pproved procedure )
ggigff Manager or S, fN-% tvi or / Date l Licensed SRO / Date
}-k& 0b/
implementhtion Time & Date [* Deadline for POC revew and/or approval by Approving Authonty (14 days from implementation): %D -k OlsTRIBUTION: (Each location must be instealed by individual distnbutmg and/or iniegrating TCN)
'? ORIGINATOR / DESIGNEE INTEGRATES CONTROL ROOM (All Volumes) $J- STA DESK (CR)(Volume 13)
SHIFT MANAGER (Volumes 15.13. SWP) (Q EOP FLOWCHART (CR)(Volume 5) SCR AM BOOK (CR) CR EMERGENCY SUPPORT (Volume 5.5 senes) COMPLETED BY PROCEDURE CONTROL DisTRtBUTION: (Each ton must be mitisied by individual disintautmg and/or integrayigCg
~~ "
TSC REMOTE DG 3 SIMULATOR DG-1 OSC
~ hf[.2) _ ER L.
COfAPLETED BY PROCEDURE SPONSOR
@ Reason for Temporary Change (identify the one that most apphes)-
] Licensing issue (e g , Tech Spec, FSAR) OER ] TSSIP iISCR PMR PERA/PTL Temporary Mod. -
Minor Mod Site Wide Procedure Prgoram
! Vendor Manual Change 2 Procedure Enhanments- f, other 7 (specify)
- ALARA review required per PPM 11.2 2.77 [ Yes R None Required
_ LARA sf A Reviewer /Date
?o Have all other affected procedures been revised or changed? [ Yes 2 None Required 2' Has impact on Model Work Orders. SMS or PTURTS been ] Yes { None Required determined and resolved with a nropnate personnel?
" Procedure Sponsor -
hg _ j .,f f Pnnt Name .a . . .e [ / ~ Date l APPROVAL Approving Authority (Pla eneral Manager for POC reviewed procedures) (Date - lPOC MEe NG (if app able) I' g d ,[ h f6 Ih . 968-23342 R14 (1!96) Page1
_. . . . . .- _ - . . - - _ . . . ~ . _ . - _ - - - . . _ . - . . . . . . . . - - . . - . . . . - . - . . - - . e J 4. VERIFY PRIOR TO USE ' ' j th&SMlWGTON FO BLIC FOwra
@) SUPPLY SYSTEM l DATE .
I l WNP-2 f PLANT PROCEDURES MANUAL PROCEDURE NUMBER APPR ~ BY fi DATE
*2.5.7 ,
,Mk,p 11/10/95
[ s[ v0LUME NAME f , SYSTEM OPERA ING-PROCEDURES - SECTION \ ,' ' TURBINE GENERATOR AND AUXILIARIES TITLE ! MAIN TURBINE GENERATOR ! i t t l i i i PROCEDURE NUMBER REVIS!ON PAGE 2.5.7 25 1 of 104
TABLE OF CONTENTS EaEC 1.0 PURPOSE ................................................ 4 2.0 REFEREN C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0- PREREQUISITES ...........................................6 4.0 PRECAUTIONS AND LIMITATIONS . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . 7 5.0 PROCEDURE................... ..........................13 5.1 Main Turbine Generator Preparation For Startup (DEH Mode 1) . . . . . . . . . . 13 5.2 Main Turbine Generator Latching . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . 22 5.3 Main Turbine Generator Operator Auto Startup (DEH Mode 2) . . . . . . . . . . . 25 5.4 Main Turbine Generator Synchronizing with Grid (DEH Mode 3) . . . . . . . . . . 31 5.5 Main Turbine Generator Unit Loading (DEH Mode 4) . . . . . . . . . . . . . . . . . 36 5.6 Main Turbine Generator Moisture Separator Reheater Temperature Control . . . . 37 5.7 Main Turbine Generator Normal Shutdown .......................42 5.8 Main Turbine Generator Turning Gear Operation . . . . . . . . . . . . . . . . . . . . 47 5.9 Main Turbine Generator Trips Functional Tests . . . . . . . . . . . . . . . . . . . . . 48 5.10 Auto Controller Reset, DEH Mode 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 5.11 Auto Controller Reset, DEH Mode 4 ...........................68 5.12 500 KV Motor Operated Disconnects 89 Operation . . . . . . . . . . . . . . . . . . . 69 5.13 Governor Valve Optimization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 5.14 Recovery from Governor Valve Optimization . . . . . . . . . . . . . . . . . . . . . . 75 5.15 Power System Stabilizer Abnormal Operation . . . . . . . . . . . .. . . . . . . . . . . 76 5.16 . Manual Opening of 500 KV Breakers . . . . . . . . . . , . . . . . . . . . . . . . . . . 77 PROCEDURE NUMBER REV!510N PAGE 2.5.7 25 2 of 104
. l
.. ; r - , j t
5.17 Manual r, losing of 500 KV Breakers . . . . . . . ................... 78 l 5.18 Relatching the Main Turbine Following Turbine Shutdown . . . . . . . . . . . . . . 7.7 j i 5.19 Weekly MWIT Relaying Test . . . . . . . . . . . . ...................82 i I 6.0 A'ITATCHMENTS . . . . ......................................85 i i i 6.1 Load Changing Recommendations (HPT First Stage Temp Change) . .. . . . . . . . 86 i i 6.2 Startup Recommendations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 87 j i 6.3 Exhaust Pressure Limitations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 l i 6.4 Turbine Speed Hold Recommendation (Turbine Resonant Speed Ranges) ..... 89 ! l 6.5 Main Turbine Pedestal, Manual Test Valves, Gages, and Test Levers ....... 90 l i 6.6 Generator Capability Curve . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 l i 6.7 Monthly Main Turbine Generator Functional Test Data Sheet ............ 93 ! l 6.8 Main Turbine Test of Auxiliary Oil Pumps Data Sheet ................-94 ; i : i l 6.9 Post Refueling Outage Main Turbine Generator OPC and Actual Overspeed Trip Data Sheet . . .....................................95 l l 6.10 Manual OPC Testing Following Refueling Outage . . . . . . . . . . . . . . . . . . . 96 6.11 Off-Frequency Turbine Operation . . . . . . . . . . ..................99 6.12 Main Turbine Generator Manual Startup (DEH Mode 2) .............. 100 l l I a A 4 I l
-e.
- P9tOCEDURE NUMBER REVISION PAGE 2.5.7 25 3 of 104
? w w . - . , . 9q,. ,. _ . . -..g ----d -
t , 1.0 PURPOSE To provide operating instructions for operation of the Main Turbine Generator.
2.0 REFERENCES
2.1 LICSAR 00069 (Testing OPC, Mechanical and Electrical Overspeed Trip Mechanisms During Reactor Operation) O {2.1} 2.2 GE SIL (OER 89075) Max Combined Flow Limiter Setting of 130% 8 {2.2} ' 2.3 OER 80109C Turbine Generator Bearing Failure Caused by Turbine Lube Oil System Failure O {2.3} 2.4 OER 86009J Asymmetric Turbine Imading O {2.4} 2.5 PER 91-822, Stuck Open DEH Dump Valve O {2.5} 2.6 PER 292-266, Emergency Oil Pump Pressure Test O {2.6} 2.7 OER 91031CA, Turbine Failure Caused By Overspeed 8 {2.7} 2.8 PER 293-411, EOP Start Test 8 {2.8} 2.9 M502, Main and Exhaust Steam Flow Diagram 2.10 CVI 01-00,82, EH Fluid System & Lube Diagram (Info only) 2.11 M960, Main Turbine Oil System Flow Diagram j 2.12 M959, Electro-Hydraulic Fluid System Flow Diagram 2.13 E502, Main One Line Diagram 2.14 E503, Auxiliary One Line Diagram 2.15 E505, DC One Line Diagram i 2.16 E510, Synchronizing Diagram, Sheet 1,2 2.17 E511, Generator Station Tripping Schedule 2.18 E512, Protective Relaying and Control, Sheet 1,2 l FROCEDURE NUMBER REvts10N PAGE 2.5.7 25 4 of 104
L , 2.19 E513, Main Action One Line Diagram 500KV Relaying 2.20 E520, Turbine Generator Control, Sheets 1 to 7 2.21 CVI 02-01-000,113, O&M Manuals Vol 1 and 2, Westinghouse 2.22 Westinghouse 1250-C735 Volume II 2.23 CVI 02,01,00,145, DEH Westinghouse 2.24 PPM 1.11.8, Radiation Work Permit 2.25 PPM 2.2.4, Main Condensate and Feedwater Systems 2.26 PPM 2.2.7, Extraction Steam and Heater Vents / Drains 2.27 PPM 2.2.8, Sealing Steam System 2.28 PPM 2.5.1, Electro-Hydraulic Fluid System 2.29 PPM 2.5.2, Turbine Lube Oil Purification / Storage / Transfer System 2.30 PPM 2.5.3, Generator Seal Oil System 2.31 PPM 2.5.4, H 2/CO 2System 2.32 PPM 2.5.5, Stator Coil Cooling System 2.33 PPM 2.6.1, Circulating Water and Cooling Towers 2.34 PPM 2.6.3, Air Removal System 2.35 PPM 2.7.1A,6900 Volt and 4160 Volt AC Electrical Power Distribution System 2.36 PPM 3.2.1, Normal Shutdown to Cold Shutdown 2.37 PPM 3.2.2, Normal Shutdown to Hot Shutdown PROCEDURE NUMBER REVISION PAGE 2.5.7 25 5 of 104 1
l
<. l L ,
i I , 3.0 PREREOUISITES j 3.1 The following systems are required to be in operation to support Main Turbine: l l
- Plant Service Water System !
i
- Circulating Water System ;
i l
- Seal Oil System l l
- Steam Seal System l
- Turbine Oil System t
Stator Coil Cooling System
- Digital Electro-hydraulic System ,
I
- Isolated Bus Duct Cooling System l
- Generator Hydrogen System !
- Condenser Drains / Vents System
- Air Removal System I 1
l \
\
- Heater Vent System <
l
- Heater Drain System
- Bleed Steam System
- Auxiliary Steam System l
- Main Steam System
- Condensate Storage and Transfer System
- Control and Service Air System ;
I ' l 3.2 If the Main Turbine startup is the initial startup following a refueling outage, perform the ! OPC test per Attachment 6.10 prior to Main Turbine startup. This will confirm the operability of the OPC circuits and the turbine trip circuit. O {2.7} i i a PROCEDURE NUMBER REVISION PAGE 2.5.7 25 6 of 104
1 4.0 PRECAUTIONS AND LIMITATIONS 4.1 The turbine must be on the turning gear at least I hour prior to rolling with ; steam. ! 4.2 Do not rotate the turbine-generator rotor if the Seal Oil System is not in service. , 4.3 The DEH Controller must be energized at least two hours prior to admitting steam to turbine. , 4.4 Ensure the turbine lube oil reservoir level is adequate before starting the first oil ! pump - usually the AC Bearing Oil Pump (TO-P-BOP). [The reservoir should be j near full prior to starting the bearing oil pump (TO-P-BOP). This will provide j sufficient oil to fill the system and establish the normal reservoir operating level.] j 4.5 The generator loop seal vapor extractor must be operated continuously when i hydrogen pressure is maintained in the generator. i 4.6 Exhaust hood sprays must be available whenever the turbine is operating above 3 RPM (prior to rolling the Main Turbine with steam). ! 4.7 A turbine lube oil reservoir vapor extractor must be in service when starting and : operating the turbine and whenever hydrogen is in the generator. 4.8 The Generator must be pressurized with hydrogen gas per the Capability Curve (Attachment 6.6) before rolling the turbine with steam. l 4.9 The Main Turbine lube oil outlet temperature should be at least 100*F prior to l exceeding 500 RPM. j 4.10 Turbine speed hold recommendations (Attachment 6.4) must be referred to during the turbine roll. Holding at or close to critical speed is POSITIVELY NOT ! l permitted. 4.11 Before rolling the turbine off the turning gear the Reactor Steam Generation Rate shuold be at least 15% and Reactor Pressure should be at least 932 psig. 4.12 Rotor eccentricity should be LT 6 mils per the eccentricity recorder. The portable rotor truth dial indicator measurement at any bearing oil ring should not exceed .001 amplitude with the unit on turning gear. PROCEDURE NUMBER - REVISION PAGE 2.5.7 25 7 of 104
L , 4.13 Rotor position is based on a nominal thrust bearing clearance of .015". An alarm limit of .035" and trip limit of .040" from the center of the thrust cage clearance l (in each direction) is used. Changes in rotor position with passage of time should ' be noted; the limits given above apply relative to original setting. The System Engineer should be informed of any changes. 4.14 Turbine exhaust steam temperature should not exceed 147'F for extended operation when the exhaust hood spray is out of service. (147*F equates to 7.0" Hg Abs back pressure.) (See Attachment 6.3.) 4.15 When operating with high exhaust temperature, particular attention should be paid ! to differential expansion, vibration, and bearing metal temperature changes. ! 4.16 Use of exhaust hood sprays should be minimized by maintaining proper exhaust temperatures with as high a vacuum as possible in the main condenser. 4.17 Operation at LT approximately 65 MWe generator load should be avoided. 4.18 Do not operate with a reheat stop or interceptor valve closed at a reactor power of GT 62% during valve testing. 4.19 The L.P. exhaust hood temperature must not exceed 250 F. 4.20 The L.P. turbine steam inlet temperature should be limited to 400*F when generator load is LT 120 MWe. 4.21 Vibration limits (absolute, peak to peak - mils): 4.0 mils - satisfactory. 7.0 mils alarm. Re-balancing is indicated if vibration is continuous and of the unbalanced type. 14.0 mils - trip or other suitable action. Other suitable action may be load change / speed change, etc., according to specific conditions. PROCEDURE NUMBr_R REVISION PAGE 2.5.7 25 8 of 104
,~b O I
i l 4.22 Differential exoansion limits: t I (Trip levels are recommended manual trips. No automatic trips are installed.) 1 Generator end: 0-2000 mils scale Governor end: 0-1000 mils scale i
- a. Rotor long: a. Rotor long: )
I
- Alarm at 645 mils
- Alarm at 150 mils ;
- Trip at 615 mils
- Trip at 120 mils
- b. Rotor short: b. Rotor short: j i
- Alarm at 1289 mils
- Alarm at 550 mils '
l
- Trip at 1319 mils
- Trip at 580 mils 1 4.23 The vacuum breaker valves should not be opened until the steam flow to the condensers !
has completely stopped and the unit has either coasted down to 200 RPM or is on ; turning gear, unless rapid turbine coastdown is required. ! 4.24 If there is severe turbine vibration (18 mils or above), open main condenser vacuum . l : breakers until turbine speed decreases to less than 900 RPM. l ! 4.25 When starting with an initial rotor temperature of 300'F or greater, approximately , i 120 MWe generator load should be applied after synchronizing to avoid unnecessary l cooling of the rotor. ! 1 l 4.26 Should the unit trip from a malfunction, ensure the drains open immediately. l 4.27 Before making an actual overspeed test, operate the turbine at GT approximately
- 120 MWe generator load for at least 8 hours, f
4.28 The hydrogen should be purged from the generator before the main oil reservoir is drained. l PROCEDURE NUMDER REVISION PAGE
- 2.5.7 25 9 of 104 !
l l 1
,o ,
NOTE: Turbine back pressure is indicated on Pen 2 on MS-PR-lC H13-P820 (Bd B). NOTE: Condenser vacuum is the LOWEST operable reading (least vacuum) of l MS-PI-8A, B, C. l 4.29 Prior to exceeding any one of the following condenser back pressure limits on the Main Turbine, REDUCE TURBINE LOAD. If turbine load reduction fails to maintain condenser back pressure within limits, TRIP THE MAIN TURBINE. Administrative back pressure limit is 7" Hg Abs. Refer to Attachment 6.3.
- a. For turbine roll 5.5" Hg Abs back pressure
- b. For turbine loads LT 560 MWe 5.5" Hg Abs back pressure
- c. For turbine loads 560 - 835 MWe linear slope 5.5 - 8" Hg Abs back pressure
- d. For turbine loads GE 835 MWe 8.0" Hg Abs back pressure NOTE: High back pressure is indicated by high exhaust hood temperatures on TG-TR-SMT, H13-P820 (Bd B) points 4 through 9 or computer points T008AB, T009AB, T010AB.
NOTE: The following temperatures correlate to 5.5" Abs and 7.0" Abs. Refer to Attachment 6.3. 4.30 Prior to exceeding either of the following limits for exhaust hood tempecturc, REDUCE TURBINE LOAD:
- a. For turbine loads LT 560 MWe 135*F
- b. For turbine loads GE 560 MWe 147*F 4.31 Lut row blade and/or disc attachment fatigue damage can occur during relatively brief periods under high back pressure / low load conditions; the damage is cumulative and irreversible.
4.32 Steam supplied to the turbine glands should contain not less than 25*F superheat. 4.33 The temperature limits of steam in the low pressure turbine glands are 250*F minimum and 350* maximum. 4.34 The turbine should be taken off line if both main oil reservoir vapor extractors fail. 1 FROCEDURE NUMBER REVISION FAGE 2.5.7 25 10 of 104 I
4
,= * :
l 4.35 For periodic trip functional tests, two operators and the Shift Support Supervisor shall be assigned to the Main Turbine pedestal testing area, with communication established between testing area and Control Room. ; l 4.36 Observe the RWP requirements of PPM 1.11.8. I 4.37 Air temperature from the exciter cooler should be maintained at 45"C to 50*C. l Temperature below 45'C is not recommended due to possible condensation on the diode wheel. 4.38 A Technical Staff or Maintenance expert should be present during OPC testing when at rated speed. (After a turbine maintenance outage.) 4.39 No attempt.should be made to roll the turbine with steam when the rotor is stopped. The rotor must be either on the turning gear or being turned with steam leakage prior t to rolling the turbine with steam. 4.40 Steam Seals should NOI be put into service unless the turbine is on turning gear. The rotor gland areas could be thermally damaged. ' 4.41 When rolling the turbine off turning gear, personnel should keep clear of the turning ! gear operating lever to avoid injury when it disengages. ( 4.42 K'eep drain valves open from startup to approximately 240 MWe generator load. Ensure drain valves are open during shutdown. 4.43 Bearing oil discharge temperatures should not exceed 180*F. Minimum discharge temperature for turning gear operation is 70*F. 4.44 Journal and Thrust Bearing metal temperature should not exceed 225*F. The Turbine should be tripped if metal temperature exceeds 225"F. 4.45 Permissible pressure difference between condenser zones is 2.5 in. Hg. Trip the turbir.e if this value is exceeded. 4.46 Do not run either vapor extractor (TO-EX-1 A or TO-EX-1B) without a bearing oil pump in service because the resulting high oil level may flood the extractor suction. 4.47 All main turbine overspeed trip testing (mechanical and electrical) should be performed on governor valve control, NOT throttle valve speed control. PROCEDURE NUMBER REVISION PAGE 2.5.7 25 11 of 104
. ~ e 4.48 The Main Steam line drains (MD-V-75 through MD-V-82, and MD-V-85 through MD-V-87) must be open when governor valves' 2 or 3 are closed for more than a ;
minute, to prevent water from building up in the steam lines and causing turbine damage when the governor valves are subsequently reopened. A {2.4} 4.49 Failure of a Governor valve to respond to an open signal during testing and the standby DEH pump starting may be an indication of a stuck open hydraulic dump valve. If this occurs, a turbine trip may be required to reset the dump valve. A {2.5} 4.50 Do not exceed the Turbine Generator frequency limits per Attachment 6.11. 4.51 Operation outside the Generator Capability Curve (Attachment 6.6) is not recommended. Efforts to operate within the curve during normal system operation should be employed. During system disturbances, the generator may operate outside the curve. The generator will withstand some operation outside the capability curve (5% MVA maximum for 2 hours or less). The voltage regulator and Power System Stabilizer should be allowed to bring the generator back within the capability curve without operator intervention. If the generator, as per applicable annunciator alarm, is going to trip without operator action it is recommended to bring the generator back within the capability curve manually. ( l l l I 8 PROCEDURE NUMBEJt REVISION PAGE 2.5.7 25 12 of 104
l l
. ~ . GENERATOR CAPABILITY CURVE CURVE A B LIMITED BY FIELD WINDING TEMPERATURE CURVE B-C LIMITED BY STATOR WINDING TEMPERATURE CURVE C-D LIMITED BY STATOR CORE END IIEATING 800
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l . . . . . . . . 2.5 4.63 HYDROGEN INNER-COOLED TURBINE GENERATOR 1230.000 MVA .975 PF 25.0 KV 28400 AMPERES 3 PHASE 60 HERTZ 1800 RPM .58 SCR 75 PSIG i
. Attachment 6.6 PROCEDURE NUMBER REVISION PAGE 2.5.7 25 91 of 104
. *= e.
GENERATOR CAPABILITY CURVE TABULAR VALUES 75 PSIG 74 PSIG 73 PSIG 72 PSIG 71 PSIG 70 PSIG 68 PSIG 66 PSIG 64 PSIG MVARS Min H2 MinH2 Min H2 Min H2 Min 112 Min H2 Min H2 Min H2 Min H2 PRESS PRESS PRESS PRESS PRESS PRESS PRESS PRESS PRESS (+) OLTT/ I OVER. (1230 (1221 (1213 (1205 (1896 (1189 (1871 (1155 (1138 ) EXCTTED MVA MVA MVA MVA MVA MVA MVA MVA MVA Umit) Umit) Umit) U mit) Umit) Undi) Umit) Umit) Umit) (-) IN/ I UNDER. MWe MWe MWe MWe MWs MWe MWe MWe MWe EXCTTED Umit = Umit = Umit = Limit = Umit = Umit = Limit = u nst = Umit= GT + 140 USE GEN CURVE PREV PAGE !
+ 101 to + 140 1222 1212 1204 1896 1187 1179 1162 1146 1129
+ 61 to + 100 1225 1216 1208 1200 1l91 1183 1166 1850 1133
+31 to + 60 1228 1219 1211 1203 1194 1187 1169 !!53 1136
-30 to +30 1229 1220 1212 1204 !!95 1188 1170 1854 1137
-60 to -31 1228 1219 12Ii 1203 1194 1187 1169 1153 1136 41so.120 1224 1215 1207 1199 1189 1181 1864 1837 1131 l
-121 to -200 1213 1204 1196 1188 1179 1171 - 1153 1137 1120
!.T -200 USE GEN CURVE PREV PAGE Use Recorder E-RECT-W/ VAR /G1 on board C for MVAR and for MWe.
Due to rounding and use of broad MVAR ranges in the table, the above MWe Limit values will be consenative when I compared to limits computed using actual values of MVAR. This table is intended as a quick method of verifying compliance with the generator capability curve when plant computers are unavailable or computer data is invalid. i ( PROCEDURE NUMBER REvlSION Attachment 6.6 PAGE 2.5.7 25 92 of 104
1 , t), WASHINGTON PUBLIC POWER SUPPLY SYSTEM INSTRUCTIONAL COVER SHEET l PROGRAM TITLE LICENSED OPERATOR /STA REQUALIFICATION TRAINING COURSE TI1LE JOB PERFORMANCE MEASURE i Ll!SSON TITLE POWER / FLOW MAP INTERPRETATION , 1 l LESSON LENGTil 5 Min MAXIMUM STUDENTS I INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No. OJT Guide PQD Lode Rev. No. Simulator Guide PQD Code Rev. No. Student llandout PQD Code Rev. No. JPM PQD Code Rev. No. Checkoff Sheet PQD Code Rev. No. Exam PQD Code Rev. No. DIVISION TITLE Nuclear Training del %RTMENT Operations Training I'REl% RED BY Randy Guthric DATE 8/22/96 REVISED BY DATE TECllNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Trammg Manager Matnx Update Vnion # WP Update W.\
1 i POWER / FLOW MAP INTERPRETATION MINOR REVISION RECORD Minor Ibeription Atrected lintered liflixtive Manager Rev Ntunber of kevision l' ages fly Date Approval l l l l l l 1 Page 1 of 7 Rev. O
l 3- t POWER / FLOW MAP INTERPRETATION l JPM SETUP l Simulator ICs: 14 htlfunctions: N/A Overrides (Optional): N/A Special Setup Instructions: Trip the "A" RRC pump and allow plant conditions to stabilize. , Task Standard: Succesful detenuination of plant operating conditions on the power / flow map for single huip conditions. l l l Page 2 of 7 Rev.O
5 i POWER / FLOW MAP INTERPRETATION RESULTS OF JPM: l l J Examinee (Please Print): Evaluator (Please Print): Overail (circle one) Exam Code Evan.:'tae SAT / UNSAT I, Simulator IC Used Validation / Critical Tin JPM Completion Time l 5 Minutes / NA Comments: I ( l l Evaluator's Signature: Date: Page 3 of 7 Rev.O
s
- POWER / FLOW MAP INTERPRETATION JPM CllECKLIST
- Itents are Critical Steps Event Control Step Element Standard Sat /Unsat RECORD START TIME:
l i I Locates Single Loop Locates Single Loop Operation Power S/U Operation Flow Map Power / Flow Map Either PPM 2.2.1 Attachement 6.5 Or Operator Aid 91-31 located at 1113-P603 Comments: 2 Determines % Rated Determines % power using APRM S/U Thermal Power indications Comments: 3 Determines % Rated Determines % Rated Core Flow: S/U Core Flow Either PPM 2.2.1 Attachement 6.6 Or Operator Aid 91-31 located at 1113-P603 Comments. l l l I Page 4 of ? Rev.0
i i POWEIUFLOW MAP INTERPRETATION
- Itenis are Critical Steps Event Control Step Elernent Standard Sat /Unsat i
S/U* 4 Determines plant Correctly determines plant operating operating conditions cc'di' ions on: on Single Loop
* " '" E PI) 12.2.1 Attachement 6.5 Or Operator Aid 91 31 located ai H13-P603 Termination Cue: Comments:
The termination point of this JPM has been reached RECORD Calculate JPM Completion Time: TERMINATION TIME: JPM Termination Time: JPM Start Time: - JPM Completion Time: (Transfer to RESULTS OF JPM page 3) l Page 5 of 7 Rev.O
l 0 8 POWER / FLOW MAP INTERPRETATION JPM INFORMATION CARD l Initial Conditions: The plant was operating at 100% rated thermal power when RRC pump " A" tripped. , PPM 2.2.1 Section 4.4 has been entered and has been completed through Step 4.4.8. PPM 4.12.4.7 has been entered and all Immediate Operator Actions have been completed. Cue: The CRS has directed you to detennine plant operating conditions on the Ibwer/ Flow map. l Tools /Equipinent: None Safety Iterns: None ! Task Nurnber: Validation Tirne: 5 Minutes Prerequisiste Training: Tirne Critical: No PPM
Reference:
2.2.I Rev 25 Location: Simulator 4.12.4.7 Rev 13 NUREG 1123 Ref: 202001 A1.03 (3.6/3.6) Itrforinance Method: Perfbrm 202001GK.06 (3.0/4.1) Prepared or Revised by: Randy Guthrie Revision Date: 8/22/96 Page 6 of 7 Rev.O
i , . POWER / FLOW MAP INTERPRETATION STUDENT JPM INFORMATION CARD l Initial Conditions: The plant was operating at 100% rated thermal power when RRC pump "A" tripped. i PPM 2.2.1 Section 4.4 has been entered and has been completed through Step ! 4.4.8. 1 PPM 4.12.4.7 has been entered and all Immediate Operator Actions have been i completed. {
\
l Cue: The CRS has directed you to determine plant operating conditions on the Itnver/ Flow map. l l I l t Page 7 of 7 Rev.O
..s m.--.
,, %, . , . ,, y Referenco SWP-PRO-03 d> i ;- % , ~ ~ ' t , , CONTROL NUMBER O j @ j"f E TEMPORARY CIIANGE NOTICE g, COMPLETED BY ORIGINATOR W POC Review Required?
b rocedure P No: 2.2.1 Current Rev. 25 E Yes O No g Tme: Reactnr Recirculatinn System
@ O One-Tirne-Only.TCN NOT incorporated into procedure
@ Procedures pages affected by this TCN: 6
@ Control No. of incorporated DeviationffCN. y ,96-G3G8T tf,, -4 75
@ Procedure pages affected by previous Devfa'bCN f, 6,7,14,15,34,35,36
@ Summary of ChanDes 1
Add Precaution and Limitation 3.20 to note RRC-ASD speed at 105% (63 Hz) maximum when supplied by TR-N2. Note is added to roflect FCR #87-0244-6-22,6/8/96, identified lirnitations. { C Contmued I certify this procedure TCN is in comphance with the critena on page,2 f1 is form and SWP-PRO-02 Onginated by: GJ Freeman .V Le.v,or.s 6/9/96
@ Prmt Name -
I j- ' Signature Date NTERIM APPROVAL OF TC SRO Approval required for TCNs POC-appro procedure ) RL Koenig Y96 T)[,( g g o p ( /4 j . {70) b!Gf6h Manager or Supervisor / Date Dr.ensed SRO I Date I ~' implementation Time & Date I
@ Deadline for POC review and/or approval by Approving Authonty (14 days from implementation): 6/23/96 DISTRIBUTION: (Each locahon must be mitialed by individual distributmg and/or entegrating TCN)
O ORIGINATOR / DESIGNEE INTEGRATES l CONTROL ROOM (All Volumes) g(p STA DESK (CR)(Volume 13) N /A SHIFT MANAGER (Volumes 15,13, SWP) fd EOP FLOWCHART (CR)(Volume 5) g /A SCRAM BOOK (CR) fJ / n CR EMERGENCY SUPPORT (Volume 5.5 senes) W /, COMPLETED BY PROCEDURE CONTROL
@ g DISTRieUTION: teach location must be snitialed by andnndual distribuhne and/or integra Cg bVC ( ,(; RWCR bL DG-2 VSC C t,_ REMOTE DG-3 PREPAREDNEk SIMULATOR _ ('A DG-1 OSC O ER ,
1 COMPLETED BY PROCEDURE SPONSOR '
@ Reason for Temporary Change (identify the one that most applies)'
O Licensino issue (e o . Tecn Spec. FSAR) O OER TSSIP ISCR O PMR PERA/PTL Temporary Mod. Site Wide Procedure Prgoram O Minor Mod O Vendor Manual Change Procedure Enhanments specify) ] (Other __ O ALARA review required per PPM 11.2.2.77 O Yes @ None Required
@ ALARA Reviewer /Date
( @ Have all other affected procedures been revised or changed? Yes Y None Required l @ Has irnpact on Model Work Orders, SMS or PTL/RTS been Yes @ None Required , O P ced re Spon o ( R en s Pnnt Name [s - Signatdie 6/ Date l APPROVAL Approving Authonty (Plant General Manager for POC reviewed procedures) Date POC M NG (if applicable) E' /,- Q ' / r/f3 @ ,M 96s.23342 R14 (1196) Page1
l VERIFY PRIOR TO USE WAbMINCION PL SLAC POW 8A SUPPLY SYSTEM l DATE i WNP-2 i PLANT PROCEDURES MANUAL I PROCEDURE NUMBER APPROVED BY DATE
*2.2.1 GOS - Revision 25 06/01/96 VOLUME NAME l l
SYSTEM OPERATING PROCEDURES SECTION REACTOR COOLANT SYSTEM AND STEAM SYSTEMS l l TITLE I REACTOR RECIRCULATION SYSTEM i 1
'l l
I l I l l I r ! l PROCEDURE NUMBER REVISION PAGE 2.2.1 25 1 of 42
a l t 4 .j TABLE OF CONTENTS i Eings 1.0 PURPOSE................................................. 4 2.0 PREREQUISITES ...........................................4 j i 3.0 PRECAUTIONS AND LIMITATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 $ 4.0 PROCEDURE .............................................. 7 , Normal l l 4.1 Recirculation Pump Seal Operations .............................7
- 4.1.1 Seal Purge Startup ...................................7 4
4.1.2 Seal Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 > 4.1.3 Seal Purge Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 i q , 4.2 Recirculation Pump Startup in Modes 2, 3, 4 , 5 . . . . . . . . . . . . . . . . . . . . . . 10 ;
.( .
4.3 Recirculation Pump Startup in Mode 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.4 Single loop Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.5 RRC Pump Shutdown when Reactor is Shutdown . . . . . . . . . . . . . . . . . . . . . 19 j 4.6 ASD Glycol Cooling System Startup, Operation and Shutdown . . . . . . . . . . . . . 20 Off. normal 4.7 RRC Pump Operation With Reactor Shutdown and RCC Heat Sink Unavailable . . . 21 4.8 Restoration of a Single Drive Channel With Recire Pump Running . . . . . . . . . . . 24 4.9 Removal of a Single Drive Channel With Recire Pump Running . . . . . . . . . . . . 25 i 1
5.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 PROCEDURE NUMBER REV!slON PAGE 2.2.1 25 2 of 42
j e o i TABLE OF CONTENTS filEC i ! 6.0 ATTA CH M ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 l 1 - l 6.1 Reactor Recirculation System Valve Checklist .......................28 ' l l 6.2 Reactor Recirculation System Power Supply Checklist ..................37 ; 6.3 . Recirculation Pump Seal Operating Parameters . . . . . . . . . . . . . . . . . . . . . . . 39 . 6.4 Two-Pump Thermal Power to Core Flow Operating Map ................40 l 6.5 WNP-2 Single Loop Operation Power / Flow Map ..... ................'41 l 6.6 Single Loop Core Flow / Drive Flow Correlation . . . . . . . . . . . . . . . . . . . . . . 42 I i i r l PROCEDURE NUMBER REVISION PAGE 2.2.1 25 3 of 42
l l l 1.0 PURPOSE l To provide detailed instructions to operate the Reactor Recirculation System. 2.0 PREREOUISITES 2.1 RRC System Valve Checklist has been completed. 2.2 RRC System Power Supply Checklist has been completed. 2.3 Reactor Closed Cooling Water System in operation per PPM 2.8.3. (Except as noted in Section 5.9.) 2.4 Control Rod Drive System in operation per PPM 2.1.1. 2.5 ASD Cooling System is operating. 3.0 PRECAUTIONS AND LIMITATIONS
~
3.1 Within 15 minutes prior to startup of an idle recirculation loop, ensure that differential temperature and flow rate is within the following limits, per Technical Specification 3.4.1.4 (4.4.1.4): Steam dome to bottom head drain line coolant differential temperature LE 145*F.
'(Steam dome temperature is determined from reactor pressure and saturated steam tables. Drain line flow has to be established for accurate bottom head drain temperature.)
If both loops are idle, differential temperature between recirculation loop suction and reactor coolant is LE 50 F. i l 4 If only one loop is idle, loop-to-loop recirculation suction differential temperature is LE 50*F and operating loop Dow rate is LE 50% rated loop flow (20,860 gpm). LE 50*F between the reactor coolant within the non operating loop and coolant in ! the RPV, when thermal power is LE 25% of rated thermal power or the recirculation loop Dow in the operating loop is LE 10% of rated loop flow. 3.2 Imop flow asymmetry can result in vibration of jet pumps and riser braces. For two loop operation, maintain loop-to-loop flow mismatch LT 10% when LT 70% rated core flow, and maintain loop-to-loop Dow mismatch LT 5% when GT 70% rated core flow per Technical Specification 3.4.1.3. PROCEDURE NUMBER RIN$lON PAGE 2.2.1 25 4 of 42
3.3 Maintain power and flow within the prescribed limits of the applicable Power to Flow Operating Map (Attachments 6.4 and 6.5). 3.4 Do not start an idle RRC pump during single loop operation from initial rod pull until the mode switch is in RUN and reactor power is at least 10%. 3.5 Flexible spool pieces used to connect to the WR System are to be removed with blind flanges installed, except for chemical cleaning operations. 3.6 Secure seal purge prior to isolating a recirculation loop (CRD pump discharge pressure may cause over pressurization of an isolated reactor recirculation loop). Initiate seal purge immediately when a recirculation loop is unisolated. 3.7 If a loss of Reactor Closed Cooling Water to the pump seal heat exchangers occurs, operation can continue with adequate cooling provided by seal purge flow alone. The pump is to be secured if motor or bearing cooling water is lost, except as noted in Section 5.6. 3.8 If seal injection fluid flow to the pump stops while the pump is operating, operation can continue with adequate cooling provided by the pump seal heat exchangers. Recirculation pump operation is required to obtain cooling via the heat exchangers. 3.9 If both seal injection fluid flow and heat exchanger Reactor Closed Cooling Water flow i' are lost while the pumped fluid temperature is GT 200*F, the pump is to be shut down within onc minute. 3.10 Attempt to keep seal cooling water temperatures as stabk as possible. A sudden change in reactor closed cooling water temperature (i.e., a rapid change in RWCU non-regenerative heat load) could result in thermal shock and failure of reactor recirculation pump seals. 3.11 Do not close recirculation pump block valves RRC-V-67A(B) or RRC-V-23A(B) at reactor coolant temperatures above 310"F for more than 5 minutes unless loop isolation or shutdown cooling is required. This practice prevents valve thermal binding. Do not allow a recirculation loop to cool down with either its suction or discharge valve closed unless absolutely necessary. A {5.3} 3.12 The Recirculation Pump Motor current limit is 661 amps (8900 HP) and the stator winding temperature limits are:
- 248*F Continuous
- 266*F Intermittent PROCEDURE NUMBER REVISION PAGE 2.2.1 25 5 of 42
-_ - . - - - - - - . = _ - - . _-. . _ _ _ - . ..
i i 3.13 The recirculation pump motor guide bearing temperature limit is 205'F. Contact the i System Engineer for evaluation if this temperature is exceeded. If the. bearing l temperature is GE 215* F, the pump is to be tripped. l 3.14 During End-Of-Cycle power operations, deliberate reactor power changes by Final Feedwater Temperature Reduction are not permitted while changing recirculation flow. , 3.15 During End-of-Cycle power operation, final feedwater temperature is not permitted l below 377'F during implementation of Final Feedwater Temperature Reduction. , t 3.16 Recirculation Pump Motor starts are as follows: I The motor can be started and brought to speed whenever the motor windings are less than rated temperature (248*F) ? With other conditions permitting, it is good practice to operate for at least : 15 minutes after starting in order to cool the motor windings. i 3.17 RRC-M/A-R676A(B) and RRC-M/A-R675 signal ramp rate is dependant on the length of time the raise or lower buttons are held depressed. The ramp rates are as follows:
- 0-1.5 seconds 0.2 Hz/sec f
- GT 1.5 ~ seconds increasing 0.6 Hz/sec ;
- f. GT 1.5 seconds decreasing 3.0 Hz/sec-
- GE 15 seconds Receive " Stuck Pushbutton" alarm and pump stops !
ramping. Resets when pushbutton is released. l 3.18 During Mode 5 Refueling, with fuel bundles removed from the core, the in-core LPRMs and instrument dry tubes will not have sufficient support from the blade guides ; to protect them from damage caused by cross channel flow. To protect the incore instrumentation tubes, total core flow is restricted to LE 10,000 GPM drive flow via RHR and/or RRC. RHR (LPCI) injection is not permitted during refueling unless needed for a valid LOCA condition. O {5.6} 3(190While on the Startup Transformer (TR-S) limit RRC pump frequency to LE 30 Hz if both RRC pumps are running, or to LE 60 Hz if only one RRC pump is running. 3.207With both RRC-ASDs supplied from the TR-N2, the RRC-ASDs maximum speeds will be 105% (63 Hz). l PROCEDURE NUMBER REVISION PAGE 2.2.1 25 6 of 42 I
i l 1 l 1 4.4 Single tooo Ooeration l_ NOTE: This section takes into consideration entry into single loop by either securing l an RRC pump as part of a planned evolution, or by the automatic or manual tripping of. j an RRC pump. ! i l 4.4.1 If necessary, initiate ANNA monitoring per PPM 2.1.8. {
#' 4.4.2 Ensure both RRC-M/A-676A and RRC-M/A-676B are in MANUAL (log in Control Room leg). l f
I CAUTION: Entry into the Area of Increased i Awareness may result in core oscillations. A {5.2} { 4.4.3 Ifin or in close proximity to the Area of Increased Awareness, monitor for ; potential core oscillations per PPM 4.12.4.7. i NOTE: The Technical Specification limit for Single Loop RRC Drive flow is ! LT 41,725 gpm. ; 4.4.4 Adjust RRC drive flow in the loop to remain in service to the maximum , possible without exceeding the Jet Pump Cavitation curve on Attachment 6.5, i (, (but LT,41,725 gpm) or fuel preconditioning restraints. Iag the flow in the ! Control Room Log. {P-104550} ; t 4.4.5 As soon as possible, increase operating loop flow to GT 34,000 gpm to ! maintain adequate reverse flow through the idle loop. O 5.5} 1 3 4.4.6 If the pump in the loop to be idled is running, slowly lower RRC-P-1 A(B) speed, using RRC-M/A-R676A(B), to 15 hz demand (log in Control Room leg). 4.4.7 If the pump in the loop to be idled is running, trip the pump by depressing the STOP pushbutton. NOTE: PPM 4.12.4.7 provides guidance for determining position on the P-F , map. l l 4.4.8 . Ensure Single Loop Operation within the bounds of the Power to Flow map, i Attachment 6.5. 4.4.9 Open CB-RRA(B), GB-RPT4A(B) and CB-RPT3A(B) in the shutdown loop. ! PROCEDURE NUMBER REVISION PAGE i 2.2.1 25 16 of 42 l
CAUTION: Do not close recirculation pump block I valves RRC-V-67A(B) or RRC-V-23A(B) at reactor coolant temperatures above 310*F for more than 5 minutes unless loop isolation or shutdown cooling is required. This practice prevents valve thermal binding. 8 {5.3} I CAUTION: To prevent thermal binding, a recirculation loop should not be allowed to cool down with either its suction or discharge valve closed unless absolutely necessary. O {5.3} 4.4.10 Close the pump discharge valve, RRC-V-67A(B), in the idle loop to prevent reverse rotation of the pump (log in Control Room IAg). 4.4.11 Within 5 minutes (maximum) after closing, open RRC-V-67A(B) to maintain idle loop temperature and prevent valve binding (log in Control Room Img). 4.4.12 If in Mode 1 or 2, initiate PPM 7.4.4.1.1.1. NOTE: The following is a list of PPMs that can be used to aid in meeting the g requirements of TS 3.4.1.1: PPM 4.12.4.7 to determine location on P-F map and if additional action is required for Action items "a.1 or a.2" PPM 7.4.2.1 addresses Action items "a.3.b and c" PPM 7.4.4.1.1.2 addresses Action item "a.3.e"
# 4.4.13 Ensure the provisions of Technical Specification 3.4.1.1, Action "a", are met.
l I l ( PROCEDURE NUMBER REV1510N PAGE 2.2.1 25 17 of 42
}
CAUTION: Isolation of a recirculation loop (closure l of both suction and discharge valves) at elevated l temperature (GT 200*F) may cause damage to pump j seal clastomers, due to lack of cooling. ' l CAUTION: Seal purge into an isolated recirculation { loop over-pressurizes the loop causing the seal purge l relief to lift and discharge to the RB 501 floor. I 4.4.14 If isolation of the idle recirculation loop is required, proceed as follows: I
- a. Secure seal purge injection per step 4.1.3, Seal Purge Shutdown. j
- b. Reduce RWCU flow (if necessary due to NPSH considerations) and then close RWCU-V-100(106).
- c. Close suction and discharge isolation valves, RRC-V-23A(B) and RRC-V-67A(B), from H13-P602.
- d. When recirculation loop suction temperature decreases to LT 150'F, cooling water to the recirculation pump motor, bearings and seals may be secured.
( ; 4.4.15 When the idle Recirculation pump can be recovered, restart the idle Recirculation pump per Section 4.2 or 4.3. 1 l i ( PROCEDURE NUMBER REVISION PAGE 2.2.1 25 18 of 42
i l WNP-2 SINGLE LOOP OPERATION POWER / FLOW MAP K' s E
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b $ $3 $ 5 I i I I l l I f l I I g 8 8 R 8 8 ? R R R l CPGIEd %) Jemod i EW Jeq.1 Attachment 6.5 PROCEDURE NUMBER REVISION NUMDER PAGE NUMBER 2.2.1 24 41 of 42
S O O SINGLE LOOP CORE FLOW / DRIVE j FLOW CORRELATION ."5 50 m-
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I ? 1 s e l VERIFY PRIOR TO USE
..... cio re auc co-e.
) SUPPLY SYSTEM DATE WNP-2 PLANT PROCEDURES MANUAL PROCEDURE NUMBER APPROVED BY DATE "4.12.4.7 MMM - Revision 13 05/07/96 VOLUME NAME ABNORMAL CONDITION PROCEDURES SECTION SITE TITLE UNINTENTIONAL ENTRY INTO REGION OF POTENTIAL CORE POWER INSTABILITIES i
l l l I l l PROCEDURE NUMBER REVISION PAGE 4.12.4.7 13 1 of 7
l 1.0 ENTRY CONDITIONS l 1.1 Entry into Region A on the Power to Flow map. 1.2 Entry into Region B on the Power to Flow map. 1.3 Entry into Region C on the Power to Flow map. 1.4 Unintentional entry into Area of Increased Awareness on the Power to Flow map (s). l {P-80197} 1.5 Control system problems (RRC flow control, DEH, and Feedwater level Control (FWLC) systems). 1.6 When any of the following conditions exist: The reactor mode switch is in RUN and both RRC pumps have tripped-off (i.e., neither RRC pump is running). l One RRC pump has tripped-off and reactor power is GT 25% of rated. Both RRC pumps have ramped down to 15 hz. l A single RRC pump trip has occurred and RRC drive flow in the operating loop is l below 39,000 gpm. Periodic LPRM upscale or downscale alarms appear on the Full Core Display (H13-P603). Any unexplained significant and sustained oscillations in SRM period, LPRM or APRM levels (characteristic oscillation period is 1-3 seconds). The following are provided as examples and do not represent absolute values:
- a. Doubling of LPRM or APRM noise levels,
- b. Peak to Peak APRM limit (10% + GAF) is exceeded.
- c. Any LPRM has oscillations GT 20 watts /cm2 ,
2.0 AUTOMATIC ACTIONS None i l PROCEDURE NUMBER REVISION PpGE 4.12.4.7 13 2 of 7
l
-. ~ .
3.0 IMMEDIATE OPERATOR ACTIONS i 3.1 If at any time any of the following conditions exist, MANUALLY SCRAM the i reactor: i i The reactor mode switch is in RUN and both RRC pumps have tripped-off (i.e., neither RRC pump is running). l i
- Ifin Region A.
NOTE: It is reasonable to allow several minutes to diagnose the cause of the oscillations, if they are occurring outside of Region B, C, or the Area ofIncreased l Awareness, or the characteristic oscillation period is outside the expected period, i 1 and adequate preconditioning margin is available. Periodic LPRM upscale or downscale alarms appear on the Full Core Display
)
(H13-P603). Any unexplained sienificant and sustained oscillations in SRM period, LPRM or l APRM levels (characteristic oscillation period is inside 1-3 seconds). The
- following are provided as examples and do not represent absolute values
- a. Doubling of LPRM or APRM noise levels.
l 1
- b. Peak to Peak APRM limit (10% + GAF)is exceeded.
- c. Any LPRM has oscillations GT 20 watts /cm2, 4
i 4 l PROCEDtDtE N1TMBER REVISION PAGE 4.12.4.7 13 3 of 7
4.0 SUBSEOUENT OPERATOR ACTIONS 4.1 If at any time any of the following conditions exist, as soon as possible but within 15 minutes initiate action to exit the region by control rod insertion or core flow increase (flow changes not permitted in Single Loop Operations in Region C without ANNA). Increasing core flow by restarting a Reactor Recirculation pump is not an acceptable method of exiting any region. If entry into Region B of the Power to Flow map has occurred. If entry into Region C of the Power to Flow map has occurred. Any unplanned entry into the Area of Increased Awareness with the reactor in an unanalyzed rod pattern. CAUTION: Any core instability induced oscillations should be terminated by a manual reactor scram. 4.2 If the power transients are occurring outside of Region B, C, or the Area of Increased Awareness and are not sinusoidal, or have an oscillation period outside of the 1 to 3 second range, they may be due to a control system problem. Consider the following: NOTE: Performance of step 4.2.1 is not necessary if there is sufficient margin within the preconditioned envolope to handle the magnitude of the , power oscillations. 4.2.1 If the magnitude of the oscillations is GT 20 MWe, reduce reactor power with flow 1% for every 10 MWe that reactor power is oscillating. 4.2.2 Attempt to identify and correct the cause of power oscillations. The most probable sources of oscillations are RRC, DEH, and FWLC systems. l i 4.3 If percent core thermal power versus percent core flow is outside the Power to Flow l map, action should be initiated to return to the inside of the applicable Power to Flow l map. 4.4 Refer to Tech Specs (3.2.7,3.2.8,3.4.1). 4.5 If the plant is operating in Single Loop, refer to PPM 2.2.1, Reactor Recirculation System, and ensure that all actions for Single loop Operations have been completed, or are in progress. PROCEDURE NUMBER REVISION PAGE 4.12.4.7 13 4 of 7
5.0 DISCUSSION Uncontrolled power oscillations have occurred at WNP-2 and a number of similar events have occurred at other operating BWRs. These events have consistently occurred during operation in regions of high power and low flow. Operation with an abnormal control rod pattern, particularly during sequence exchanges, appears to have contributed to the initiation of some of these events. These neutron flux oscillations have been observed to occur core-wide (in phase) as well as locally (out of phase). Local (also called regional) oscillations typically involve two halves of the core, divided along some plane of symmetry, which oscillate out of phase with each other. It is also theoretically possible to have a larger number of local regions oscillating out of phase. In order to assure adequate protection to the MCPR safety limit, operation in Recion A, where most of the observed instabilities have occurred to date, is prohibited. Due to the relatively low margin to the MCPR safety lirnit calculated under severe postulated conditions of a regional oscillation at high power and low flow, initiation of a reactor scram is reauired as soon as posible, but in all cases within 15 minutes of entering Region A (TS 3/4.2.6). Regions B and C of the Power to Flow map (s) represent regions of lower probability, than Region A, ofinstability occurrence. If Region B cr C is unintentionally entered, action should be initiated as soon as practical, but in all cases within 15 minutes to exit the Region. Tech Specs identify either rod insertion or increasing core flow as acceptable methods of exiting these regions, except while in Region C during Single Loop Operation without ANNA available. In this case, rod insertion is the only acceptable method for exiting Region C. Increasing core flow by restarting a Reactor Recirculation pump is not an acceptable method for exiting any region. The Area ofIncreased Awareness represents a region oflow probability ofinstability occurrence. Entry into the Area of Increased Awareness of the Power to Flow map (s) is permitted when the entry is a planned evolution, provided the monitoring equipment requirements of procedure PPM 2.1.8 are satisfied. If an unplanned entry into the Area of Increased Awareness occurs with an unanalyzed rod pattern, prompt action to exit the area is required as a conservative action to prevent core oscillations. An unanalyzed rod pattern is any rod pattern that has not been analyzed as adequate p_tipr to entering the Area of Increased Awareness. The most probable event that could result in an instability during operation at power is a dual or single recirculation pump trip. Per NRC Bulletin 88-07, Supplement 1, a reactor scram is required if the plant experiences a loss of all recirculation pumps while the Mode switch is in RUN. The NRC has taken this position to simplify operator action by eliminating the need to determine rod line prior to scramming, even though instabilities are not expected below the 70% rod line. PkOCEDURE NUMllER REY!slON PAGE 4.12.4.7 13 5 of 7
3 ! F ! L i 1 The monitoring of APRM and LPRM signals, and immediate initiation of reactor scram upon detection of oscillations, ensure adequate margin to safety limits is maintained. It is important 3 i- to recognize that monitoring APRM oscillations alone is not sufficient. It is possible for ; j regional oscillations to occur where LPRMs in different core locations, which have large j magnitude oscillations, are oscillating out of phase with LPRMs at other locations. In this ! j situation the averaging process that the APRM circuitry performs on its various LPRM inputs j can effectively smooth out and hide the LPRM oscillations, resulting in APRM oscillations ! i which are much lower than for certain individual LPRMs. Periodic LPRM or APRM upscale ! or downscale alarms are indications of the presence of a core boiling instability if they alarm ] core wide every 1.5 to 2.5 seconds or regionally every 0.75 to 1.25 seconds. These alarms
- may occur core wide or only in a particular region.
J ; Reactor power oscillations outside of Region B, C, or the Area ofIncreased Awareness (AIA) ! i are unlikely, but not impossible. The most likely causes of reactor power oscillation outside j ] of the AIA are control system malfunctions. All these should first be evaluated as a potential l ! core instability oscillations. Power transients that are not sinusoidal or have a period outside ; j of the 1 to 3 second range, are probably due to control system problems. For those cases where the symptoms are not characteristic of a reactor instability the control room operators l !. need to quickly narrow down the possible causes. It is reasonable to allow several minutes to i j diagnose the cause of the oscillations if they are occurring outside of the AIA and adequate , i preconditioning margin is available. If control room Operations personnel are not able to
- i. !
promptly determine or correct the cause of the oscillations, then the reactor should be ; ' shutdown by manual scram. ! c
- ( i l
! i i i i l i i 4 I -1 l I E l 3 i i 4 i f PROCEDURE NUMBER REVISION PAGE
- 4.12.4.7 13 ti of 7 2
3 e i_. - _ _ _ ,_ _ . _ . . _ _, _ __. .__ _ _ . . _ -
_._.._-.__y e Je i
6.0 REFERENCES
6.1 NCR 292-0993, Core Oscillations {P-80197} , l 6.2 NRC Bulletin 88-07, Supplement 1 l l 6.3 BWROG OG-94078, June 6,1994 ! l l 6.4 10M, November 4,1994, D.K. Aimson to G. O. Smith ! 6.5 OER 82122T ' l 6.6 Technical Specifications 3.2.6, 3.2.7, 3.2.8, 3.4.1 l l ! 6.7 ' PPM 2.1.8, ANNA Stability Monitoring System 6.8 PPM 2.2.1, Reactor Recirculation System 6.9 i PPM 3.3.1, Reactor Scram ; i I 7.0 ATTACHMENTS None i i I i a PROCEDURE NUMBER REYlSION PAGE 4.12.4.7 13 7 of 7 l
U.S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC WRITTEN EXAMINATION g APPLICANT INFORMATION Name: Region: IV Date: October 71996 Facility / Unit: WNP-2 License Level SRO Reactor Type: GE Start Time: Finish Time: INSTRUCTIONS Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80 percent. Examination papers will be picked up 4 hours after the examination starts. All work done on this examination is my own I have neither given nor received aid. Applicant's Signature RESULTS Examination Value Points Applicant's Score Points l Applicant's Grade Points l l l l
SENIOR REACTOR OPERATOR i A N S W E R S li E E T i Multiple Choice (Circle or X your choice) NAME: Ifyou decide to change your original answer, draw a single line through the error, enter the desired i answer, and initial the change. !
- 1. abcd 21. abcd 41. abcd !
1
- 2. abcd 22. abcd 42. abcd i
- 3. abcd 23. abcd 43. abcd i
- 4. abcd 24. abcd 44. abcd l
- 5. abcd 25. abcd 45. abcd
- 6. abcd 26. abcd 46. abcd l t
- 7. abcd 27. abcd 47. abcd I i
- 8. abcd 28. abcd 48. abcd i
- 9. a b c :) 29. abcd 49. abcd
- 10. abcd 30. abcd 50. abcd 11 abcd 31. abcd 51. abcd
- 12. abcd 32. abcd 52. abcd-l l 13. abcd 33. abcd 53. abcd ;
l
- 14. abcd 34. abcd 54. abcd ;
l l
- 15. abcd 35. abcd 55. abcd i i
?
l 16. abcd 36. abcd 56. abcd j
- 17. abcd 37. abcd 57. abcd ;
- 18. abcd 38. abcd 58. abcd [
l 19. abcd 39. abcd 59. abcd
- 20. abcd 40. abcd 60. abcd Page: 2
SENIOR REACTOR OPERATOR ANSWERSHEET Multiple Choice (Circle or X your choice) NAME: If you decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the change. I
- 61. abcd 81. abcd
- 62. abcd 82. abcd ,
- 63. abcd 83- abcd
- 64. abcd 84. abcd
- 65. abcd 85. abcd
- 66. abcd 86. abcd
- 67. abcd 87. abcd
- 68. abcd 88. abcd
- 69. abcd 89. abcd
- 70. abcd 90. abcd 71 abcd 91. abcd
- 72. abcd 92. abcd
- 73. abcd 93. abcd
- 74. abcd 94. abcd 75, abcd 95. abcd i 1
- 76. abcd 96. abcd 77, abcd 97. abcd I
- 78. abcd 98. abcd
- 79. abcd 99. abcd
- 80. abcd 100. abcd Page: 3
SENIOR REACFOR OPERATOR NRC POLICIES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on any part of the examination will result in a denial of your application.
- 2. If you have any questions concerning the administration of the examination. do not hesitate asking them before starting that part of the test.
- 3. SRO applicants will be tested at the level of the responsibihty of the senior licensed shift position (i.e. Shift Manager).
- 4. You must pass every part of the examination to receive a license. Applicants for an SRO-upgrade license may require remedial training in order to continue their RO dutics if the examination reveals deficiencies in the required knowledge and abilities.
- 5. The NRC cxaminer is not allowed to reveal the results of any part of the examination until they have been reviewed and ;
approved by NRC management. Grades provided by the facilitec licensec are preliminary until approved by the NRC. l You will be informed of the official examination results about 30 days after all the examinations are complete.
- 6. After you complete the exarrjnation, sign the statement on the cover sheet indicating that the work is your own and you )
have not receive i ot given auistance in completing the examination. l I
- 7. To pass the examination, yt u must achieve a grade of 80 percent or greater. Every question is worth one point. l
- 8. The time limit for completing the examination is four hours. !
- 9. You may bring pens and calculators into the examination room. Use only black ink to ensure legible copics.
- 10. Print sour name in the blank provided on the answer sheet provided and do not leave any question blank. Use only the
, paper provided and do not write on the back side of the pages. If you decide to change your original answer, draw a single line through the error, enter the desired ansucr. and initial the change. J l 11. If the intent of a question is unclear, ask questions of the NRC cxaminer or the designated facility instructor only. < 12. Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to climinate even the appearance or possibility of cheating.
- 13. When you complete the examination, assembic a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC cxaminer or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor rcccived assistance in completing the examination. The scrap paper will be disposed of immediately aller the examination.
- 14. After you have turned in your examination. leas e the examination area as defined by the proctor or NRC cxaminer. If you are found in this area while the examination is still in progress,30ur license may be denied.
Page: 4
SENIOR REACTOR OPERATOR QUESTION: 1 (1.00) Following maintenance on Hydraulic Control Unit (HCU) 26-35, the HCU is to be returned to service. Independent verification of valve position is required. Select the one (1) statement below that correctly describes how independent verification of these valves is accomplished.
- a. The first operator opens the valves and seals them open using appropriate seals. The second operator verifies valve position from the control room by observing the " Accumulator i Trouble" light on the full core display extinguishes.
- b. The first operator opens the valves and seals them open using appropriate seals. The second operator confirms the valves are sealed open.
- c. The first operator opens the valves. The second operator observes the valves being opened.
No sealing devices are required for these valves.
- d. The first operator opens the valves. The second operator confirms the valves are open and seals them using appropriate seals.
ANSWER: b. KA: 294001K1.01 RO/SRO: 3.7 Reference- PPM l.3.29, pg 4 of 26 Comments: New question (Plant Wide Generic) LO: 6157,6159 l 1 l Page: 5 l j
. - - - . - - __ - . - - - - . . _ _ - - _ _ . . - ._ =. . =
l SENIOR REACTOR OPERATOR QUESTION: 2 (1.00) , Which one (1) of the following individuals is responsible for the initial review of plant clearance orders and work packages received from the Clearance Order Review Committee (CORC)?
- a. Control Room Operator (CRO)
- b. Shift Technical Advisor (STA)
- c. Production SRO j l l
- d. Control Room Supervisor (CRS) i ANSWER: a.
KA: 29400iK1.02 RO/SRO: 4.5
Reference:
PPM 1.3.8, pg 18 of 62 Comments: New Question (Plant Wide Generic) LO: 6238 i i l l l i Page: 6
i SENIOR REACTOR OPERATOR QUESTION: 3 (1.00) Which one (1) of the following statements would describe " Simultaneous Verification" as it applies to a danger tag clearance order? l
- a. Two qualified individuals, independently and separately checking the required status of the component or device.
- b. A Control Room Operator (CRO) verifying the required status of the component or device i using Control Room indications.
I
- c. A second qualified individual, via local panel indications showing the required status of the component or device.
- d. Two qualified operators, accompanying each other, check required status, correct identification and location prior to changing component status.
.I l
ANSWER: d. l KA: 294001K1.02 RO/SRO: 4.5
Reference:
PPM 1.3.8 Comments: Modified question (58)(Plant Wide Generic) , LO: 6231 , 1 Page: 7
t SENIOR REACTOR OPERATOR QUESTION: r 4 (1.00) h Under normal conditions, keys to locked "High Radiation" areas are issued from which one (1) of the ; following? j
- a. Control room.
i
- b. Secondary alarm station. ;
- c. Work control center. -
- d. Health physics access control.
i ANSWER: d. i KA: 294001K1.03 i RO/SRO: 3.8
Reference:
PPM i1.2.7.3 Comments: Modified question (806)(Plant Wide Generic) LO: 6390 i 1 i i Page: 8
SENIOR REACTOR OPERATOR QUESTION: 5 (1.00) A task must be performed at a location with a general area radiation level of 60 mr/hr. Previous performance of the task indicates that: One (1) worker can perform the task in I hr and 20 min. Two (2) workers can complete the task in 50 min. i Three (3) workers can complete the task in 30 min. Four (4) workers can complete the task in 25 min. Based on the above information, how many workers should be assigned to perform this task?
- a. One (1) worker :
- b. Two (2) workers
- c. Three (3) workers 7
- d. Four (4) ,vorkers ANSWER: a.
i KA: 294001K1.04 l RO/SRO: 3.6
Reference:
PPM 1.11.2 Comments: New question (Plant Wide Generic) LO: None t L J Page: 9
SENIOR REACTOR OPERATOR QUESTION: 6 (1.00) ) What is the maximum number of visitors that may accompany one (1) escort into the Main Control Room?
- a. Three (3)
- b. Five (5)
- c. Ten (10)
- d. Fifleen (15)
ANSWER: b. KA: 294001K1.05 RO/SRO: - 3.7
Reference:
GET 82-RDT-0300-HO, page 23 Comments: New question (Plant Wide Generic) LO: 6097 l 1 i i t i l l i i Page: 10 i
SENIOR REACTOR OPERATOR QUESTION: 7 (1.00) While inspecting work inside a confined space, the Designated Safety Representative (DSR) reports that combustible gas levels are above 1%. The DSR has ordered you to immediately evacuate the confined space. Which one (1) of the following actions is required prior to reentry into the confined space?
- a. Self Contained Breathing Apparatus (SCBA) and DSR authorization.
- b. Hazardous atmosphere has been eliminated and DSR has authorized reentry.
l
- c. Self Contained Breathing Apparatus (SCBA) and standby worker at the entrance to the l confined space.
- d. " Stay Time" calculated by the DSR in order to reenter the confined space and complete the job. ,
ANSWER: b. KA: 29400lKl .13 RO/SRO: 3.6
Reference:
PPM 1.9.2, pg 7 of 17 Comments: New Question (Plant Wide Generic) LC: None l l Page: 11
SENIOR REACTOR OPERATOR QUESTION: 8 (1.00) Due to an injury sustained on shift, the Fire Brigade Leader has to leave work. Which one (1) of the following individuals can take the place of the Fire Brigade Leader?
- a. Qualified security officer
- b. Shift support supervisor
- c. Qualified equipment operator
- d. Qualified health physics technician ANSWER: c.
KA: 29400lKl.16 RO/SRO: 3.8
Reference:
PPM 1.3.1, pg 40 of 86 Comments: New Question (Plant Wide Generic) LO: None Page: 12
l SENIOR REACTOR OPERATOR i QUESTION: 9 (1.00) !
' While reviewing a procedure, you notice a step with a star (*) in the left margin.
Which one (1) of the following is represented by this symbol? l
- a. Critical to plant / personnel safety step. ,
- b. FSAR Appendix F (Fire Protection Program) commitment step.
)
c Technical Specification related step. '
- d. "For Information Only" step. :
ANSWER: b. j 1 KA: 294001 A1.01 i i RO/SRO: 3.4
Reference:
SWP-PRO-03, pg. 26 of 65 Comments: New Question (Plant Wide Generic) i LO: 6054 . i l l i t l Page: 13
SENIOR REACTOR OPERATOR QUESTION: 10 (1.00) When may a task be performed without an approved procedure present? 1
- a. When the task has no safety significance.
- b. When the task procedure number on the cover sheet is proceeded by an asterisk (*).
l
- c. When the task has been previously performed during the shift by the individual and the required steps have been memorized.
i
- d. When the task consiste of simple routine actions frequently performed that don't require step sign-offs, recorded data, or specific sequence.
ANSWER: d. l KA: 294001 A1.02 RO/SRO: 4.2
Reference:
SWP-PRO-01 Comments: Modified Question (5287)(Plant Wide Generic) LO: 6058
?
?
)
l i l Page: 14
SENIOR REACTOR OPERATOR QUESTION: 11 (1.00) During the performance of a surveillance procedure the Control Room Operator (CRO) informs the Control Room Supervisor (CRS) of a potential problem associated with the next step. After discussion and with concurrence from the Shift Manager it is decided to use a Verbal Temporary Change. At what point in time does this Verbal Temporary Change have to be translated into a Temporary Change Notice (TCN)?
- a. Prior to the end of that working shin.
- b. Prior to the start of the next calendar day.
- c. Within seven (7) working days.
- d. Within fourteen (14) working days.
ANSWER: a. l KA: 294001 A1.02 i RO/SRO: 4.2 '
Reference:
SWP-PRO-02 Comments: Modified Question (5305)(Plant Wide Generic) LO: 6065 l Page: 15
- _ _ _ _ _ . _ _ . _ . _ - _ _ _ _ . _ . . _ . _ _ . _ _ . . _ _ _ __..m . _ _ . _ . _ _ . _ _ _ _ SENIOR REACTOR OPERATOR QUESTION: 12 (1.00) ; An evolution is being performed in accordance with the Reactor Water Cleanup System (RWCU) operating i procedure when an error in the procedure is noted. The procedure already has a Temporary Change Notice (TCN) cover sheet attached to the procedure. The Shift Manager (SM) has determined that completion of the work is not essential to plant operations. Select the one (1) statement below that describes the administrative requirements necessary to continue with this evolution.
- a. Develop a temporary procedure to address the problem until the existing procedure can be revised to correct the error.
- b. Implement and receive approval for a separate additional TCN to correct this error before i continuing with the evolution.
- c. Implement a new TCN which corrects the new error and incorporates and cancels the outstanding TCN.
- d. Receive verbal authorization to complete the task and record in the Operations Logging System. !
ANSWER: c. KA: 294001 A1.02 RO/SRO: 4.2
Reference:
SWP-PRO-02 Comments: Question # 758 (Plant Wide Generic) LO: 6067 l l l i Page: 16
SENIOR REACTOR OPERATOR 8 QUESTION: 13 (l.00) 1 The plant is operating in Operational Condition 1. Which one (1) of the following lists the minimum shift crew composition administrative limit as specified in plant procedures?
- a. SM, CRS, SSS, three (3) CROs, five (5) EO, ENS communicator, two i.2) 10' Techs, two (2)
- Chem Techs, two (2) I & C Techs, Duty Oflicer.
- b. SM, CRS, two (2) CROs, two (2) EOs, STA, HP Tech, Chem Tech.
I j c. SM, CRS, SSS, three (3) CROs, four (4) EOs, STA, five (5) Fire Brigade members, three (3) HP Techs, Chem Tech.
- d. SM, CRS, SSS, two (2) CROs, two (2) EOs, STA, ENS communicator, five (5) Fire Brigade j members, three (3) HP Techs, Chem Tech, Elect /l&C Tech, Mechanic, Duty OfIicer.
ANSWER: c. ! i l KA: 294001 A1.03 RO/SRO: 3.7 j!
Reference:
PPM 1.3.1, page 37 ;
- Comments
- New Question (Plant Wide Generic)
LO: 6072 (PPM 1.3.1) ] l l i 4 e 4 i
- Page
- 17
SENIOR REACTOR OPERATOR l .QUESTION: 14 (1.00) r On a case-by-case basis, line supervisors / managers can approve the use of signals for communications, if: l
- 1) the signals are not easily confused and are understood by all involved, i
- 2) the concept of three-way communication is applied to the maximum extent possible, l
- 3) a thorough pre-job briefis conducted AND J
- 4) signals are DX11 used..
l
- a. in the Control Room. ,
L
- b. in severe environments (high noise, heat, or radiation)
- c. when physical directions are the " key" elements of the task.
- d. when tSe message originator and message recipient are not readily identifiable to each other by sight ar.d voice.
ANSWER: b. KA: 294001 A1.05 RO/SRO: 3.8
Reference:
PPM 1.3.60, page 5 of 8 Comments. New Question (Plant Wide Generic) LO: None i i-Page: 18
l SENIOR REACTOR OPERATOR 1 QUESTION: 15 (1.00) l Who has the responsibility for initiating Emergency Core Cooling Systems (ECCS), if required, during a plant ) transient?
- a. Only the operator responsible for panel H13-P601. !
- b. The on-shift crew member in closest proximity to panel H13-P601.
- c. Only the operator designated by the Control Room Supervisor (CRS) to respond. .
1
- d. Any licensed operator at the control console. i 1
I ANSWER: d. i
'l 2
KA: 294001 A1.09 . RO/SRO:- 4.2
Reference:
PPM 1.3.1 ; Comments: Question (156)(Plant Wide Genetic) , LO: 6076 i 1 Page: 19
I I' SENIOR REACTOR OPERATOR QUESTION: 16 (1.00) During a plant startup with power at =62%, feedwater dissolved oxygen concentration exceeds the fuel warranty requirements. Which one (1) of the following statements describes actions that should be taken. l
- a. Initiate an immediate plant shutdown.
- b. Initiate immediate corrective actions, increase sampling frequency, and initiate a PER.
- c. Increase sampling frequency, initiate a PER, and continue with the stanup.
- d. Immediately scram the reactor.
l ' ANSWER! b. l t i KA: 294001 A1.14 i RO/SRO: 3.4 {
Reference:
PPM 1.13.1, Pg.12 of 26 -) Comments: New Question (Plant Wide Generic) l LO: 6039 , i t l i i l l 1 Page: 20
.- .. . . . - - - . ~ - - ~ - ~ - - ..
i
, SENIOR REACTOR OPERATOR I j QUESTION: 17 (1.00) l l Which one (1) of the following is the LOWEST Emergency Action Level (EAL) at which the WNP-2 i i administrative exposure hold points are automatically waived?
j
- a. Site Area Emergency (SAE).
i i
- b. Unusual Event (UE).
l
- c. General Emergency (GE). ) !
- d. Alert.
t ' i ANSWER: d. t
- t 1
- KA
- 294001 Al.16 j (
RO/SRO: 4.7 i
Reference:
PPM 13.2.1 ! Comments: Question (6703) (Plant Wide Generic) LO: 6019 ' i ; j i i I F I- ! 9 l i, 4 i .' I i i i i i i I i i Page: 21
SENIOR REACTOR OPERATOR QUESTION: 18 (1.00) The plant is operating at rated conditions with RRC-IN-ASD/l A (l A ASD UPS inverter) in service and supplying panel E-PP-ASDl/4. Pushing the red EMER POWER OFF pushbutton on the front of the invener panel will result in which one (1) of the following? j
- a. RRC-P-1 A ("A" reactor recirculation pump) will immediately trip, RRC-P-1B ("B" reactor l recirculation pump) will continue to operate.
l l b. Both reactor recirculation pumps will immediately trip.
- c. Both reactor recirculation pumps will trip aller approximately 20 minutes. I
- d. Both reactor recirculation pumps will continue to operate indermitely.
i l ANSWER: b. ! 1 KA: 202002A2.07 l RO/SRO: 3.3
Reference:
PPM 2.7.4A : Comments: Modified question (6206) (Plant Sys - Op I) l LO: 9683
)
I i l 1 Page: 22
l SENIOR REACTOR OPERATOR ' QUESTION: 19 (1.00) , The plant is operating in Mode 1. Both Reactor Recirculation pumps are operating at 45 Hz. Adjustable Speed Drive (ASD) Channel A2 is not running and its white READY light is illuminated. l Which one (1) of the following describes the response of the Recirculation Pumps if the START Pushbutton for RRC-P-1 A is pressed?
- a. The A2 ASD Channel will start. RRC-P-1 A frequency will ramp to 52.2 Hz.
l
- b. The A2 ASD Channel will start. RRC-P-1 A will continue to operate at 45 Hz.
i
- c. RRC-P-1 A frequency will ramp to 15 Hz, the A2 ASD Channel will start, and RRC-P-1 A j frequency will ramp back to 45 Hz.
- d. RRC-P-1 A frequency will ramp to 52.2 Hz and the A2 ASD Channel will stan.
ANSWER: b. i KA: 202002K1.02 RO/SRO: 4.2 i
Reference:
PPM 2.2.1 ' Comments: Modified Question (3217) (Plant Sys - Gp 1) ! LO: None l i l Page: 23
SENIOR REACTOR OPERATOR QUESTION: 20 (1.00) i Given the following indications:
- Reactor scrammed
- Reactor water level at -95" and down slow
- Drywell pressure at 1.67 psig and up fast.
Which one (1) of the following statements describes the response of the Low Pressure Coolant Injection (LPCI) mode of the Residual Heat Removal (RHR) system to the indications listed above..
- a. LPCI will initiate when reactor water level decreases to LE -129"
- b. LPCI initiated when reactor water level reached -50"
- c. LPCI will initiate when drywell pressure increases to GE 1.68 psig.
- d. LPCI initiated when drywell pressure reached 1.65 psig. ;
1 ANSWER: d. KA: 203000Kl.13 RO/SRO: 4.0
Reference:
RHR System Text Comments: New Question (Plant Sys - Gp I) LO: 5775 i Page: 24
SENIOR REACTOR OPERATOR QUESTION: 21 (1.00) i HPCS-V-1 (CST suction) is full closed and IIPCS-V-15 (suppression pool suction) is full open during the performance of a High Pressure Core Spray (HPCS) valve operability surveillance when a valid HPCS initiation signal is received.
- Which one (1) of the following statements correctly identifies the response of the HPCS system to these l conditions.
- a. HPCS-V-1 will remain closed.
HPCS-V-15 will remain open. l HPCS-P-1 stans immediately. l b. HPCS-V-15 will stan to close. ! When HPCS-V-15 is full closed, HPCS-P-1 will start and HPCS-V-1 will open. l c. HPCS-V-1 will remain closed. HPCS-V-15 will remain open. i l HPCS-P-1 must be manually started. l
- d. HPCS-V-15 will remain open.
When HPCS-V-1 starts to open, HPCS-P-1 will start after a 5 second time delay. ANSWER. a. l KA: 209002A3.01 i RO/SRO: 3.3
Reference:
HPCS Systems Text Comments: Modified question (441) (Plant Sys - Gp 1) LO: 5425 1 \ ! i i I l } l Page: 25
4 l l SENIOR REACTOR OPERATOR i . i QUESTION: 22 (l.00) Following a valid Ifigh Pressure Core Spray (liPCS) initiation on high drywell pressure., the liPCS LEVEL 8 SEALED IN light and alarm are received and IIPCS-V-4 (RPV injection valve) closes. Which one (1) of the following conditions will cause liPCS-V-4 to automatically re-open?
- a. Drywell high pressure logic reset.
) b. RPV level lowering to +12" 4 c. RPV level lowering to -51"
- d. Drywell high pressure alarm clears.
4 ANSWER: c. KA: 209002K4.02 - RO/SRO: 3.5
Reference:
liPCS System Text Comments: Modified question (443) (Plant Sys - Gp I) LO: 5429 l t f f 4 Page: 26 l J
SENIOR REACTOR OPERATOR QUESTION: 23 (1.00) i i An Anticipated Transient Without Scram (ATWS) is in progress concurrent with a loss of MC-88, ) suppression pool temperature is 118 F and up slow. Assume all required actions have been completed correctly and no other failures have occurred at this time. l Which one (1) of the following describes the Standby Liquid Control (SLC) system status? !
!l t- i
- a. SLC-P-1 A - running SLC-P-1B - running (SLCpumps) l SLC-V-1 A - open SLC-V-1B - open (SLC storage tank outlets) . ;
SLC-V-4 A - actuated SLC-V-48 actuated. (Squib valves) ! , b. SLC-P-1 A - loss of power . SLC-P-1 B - running (SLCpumps) l j SLC-V-1 A - loss of power SLC-V-1B - open (SLC storage tank outlets) j l SLC-V-4A - loss of power SLC-V-4B - actuated (Squib valves) )
- c. SLC-P-1 A - off SLC-P-lB - otT (SLCpump3)
SLC-V-1 A - closed SLC-V-1B - closed (SLC storage tank outlets) i j SLC-V-4 A - closed SLC-V-4B - closed (Squib valves) i .i 1 d. SLC-P-1 A - running SLC-?-l B - loss of power (SLCpump3) j SLC-V-1 A - open SLC-V-1 B - loss of power (SLC storage tank outlets) SLC-V-4A - actuated SLC-V-4B - loss of power (Squib valre3) ANSWER: d. 1 KA:. 211000K6.03
- RO/SRO: 3.3 i
Reference:
SLC Systems Comments: New question (Plant Sys - Op I) LO: 593th I I l l i 1 I i i 4 4 Page: 27
SENIOR REACTOR OPERATOR QUESTION: 24 (1.00) With the plant in a hydraulic Anticipated Transient Without Scram (ATWS) condition, the Control Room Operator (CRO) carries out the actions of PPM 5.5.11 and resets the scram. Annunciator P603-A8 6-4, SCRAM VALVE PILOT AIR HDR PRESS LOW, fails to clear. Which one (1) of the following could cause this condition?
- a. At least one (1) backup scram valve has failed to ENERGIZE following the scram reset.
- b. Alternate Rod Insertion (ARI) logic has not been reset.
- c. Both backup scram valves have failed to ENERGlZE following the scram reset.
- d. One (1) of the Reactor Protection System (RPS) trip signals has not been bypassed. I J
ANSWER: b. I KA: 212000A1.11 RO/SRO: 3.3
Reference:
PPM 5.0.10 Comments: Modified Question (4085) (Plant Sys - Gp I) ! LO: 8094 (5.1.2-07) j I i l l l Page: 28
_- -- .- .- - - . _ - - . - . . . - _ - .-=-. - - . - . - . - . - - _ . SENIOR REACTOR OPERATOR QUESTION: 25 (1.00) Which one (1) of the following statements desesibes all of the conditions that will cause a trip of an Reactor Protection System (RPS) Electric Power Monitoring Assemly (EPA) breaker?
- a. Underfrequency, overcurrent, and undervoltage.
- b. Overfrequency, undervoltage, and overvoltage.
- c. Undervoltage, overvoltage, and underfrequency.
- d. Overfrequency, underfrequency, and overcurrent. ;
)
ANSWER: c. KA: 212000K2.01 i RO/SRO: 3.3
Reference:
RPS System Tee.t Comments: New question (Piant Sys - Gp I) LO: 5957 , l i l i i Page: 29
1
- SENIOR REACTOR OPERATOR QUESTION: 26 (1.00)
With the plant at rated conditions, a GROUP 1 scram solenoid light for Reactor Protection System (RPS)"A" 4 is noted to be deenergized on H13-P603 and H13-P609. During your investigation a loss of RPS B occurs l followd almost immediately by a full reactor scram. i What caused the full reactor scram? I
- a. APRM INOP.
- b. Turbine trip.
- c. MSIV isolation.
i d. Scram discharge volume high level i j
- ANSWER
- d.
i KA: 212000K3.06 RO/SRO: 4.1
Reference:
RPS Text I Comments: Modified Question (6201) (Plant Sys - Gp 1) LO: 7676 l j i l Page: 30 1 1
l i SENIOR REACTOR OPERATOR QUESTION: 27 (1.00) l During a reactor startup, the reactor is subcritical with control rod withdrawal in progress. Source Range i Monitor (SRM) count rate has stabilized at 1 x 10' Counts Per Second (CPS) following the last control rod ' l withdrawal. During withdrawal of the next control rod in the sequence, the first control rod in the next Rod Worth Minimizer (RWM) group, reactor period meters deflect from infinity to =20 seconds before turning. Reactor l period is now =60 seconds increasing (approaching infinity). Which one (1) of the following actions should be taken for this condition? l a. Verify that the withdiawn control rod did not " double notch" and stop control rod withdrawal i to allow stabilization of neutron level.
- b. Monitor SRMs and retract SRMs as necessary to maintain count rate LT 1 x 10' CPS.
- c. Insert control rods until the reactor is subcritical and notify the Control Room Supervisor (CRS)/ Shift Manager (SM) and the Station Nuclear Engineer (SNE)
- d. Immediately manually scram the reactor.
1 ANSWER: a. t l KA: 215004 A4.01 RO/SRO: 3.8 l
Reference:
PPM 3.1.2, Step 4.2.2 (2nd CAUTION), page 15; Step 4.2.5 (CAUTION) , page 16; l Step 4.2.9.d (CAUTION), page 18. Comments: Modified Question (815)(Plant Sys - Gp I) ! LO: None l l l l Page: 31
- - - . _ _ . _ - _ _ _ . . ~ . _ _ _ - . . . . - .
SENIOR REACTOR GPERATOR QUESTION: 28 (1.00) Given the following:
- RPV water level 10" and steady
- Reactor pressure 200 psig and down slow
- Drywell temperature 350 F and down very slow {
- No Secondary Containment Control entry conditions exist.
- Emergency depressurization is planned.
l Which one (1) of the following is correct concerning the instrument (c) which can be used to determine RPV water level for the given conditions? l
- a. Wide range only.
- b. Narrow range only. '
- c. None.
- d. Fuel zone range only. ,
ANSWER: a. KA: 216000K5.07 RO/SRO: 3.8
Reference:
NBI Systems Text . I ' Comments: New question (Plant Sys- Gp I) LO: 5582 i l l l l Page: 32 l
SENIOR REACTOR OPERATOR QUESTION: 29 (1.00) Reactor Core Isolation Cooling (RCIC) initiated as expected on a valid low level signal raising RPV level to the Level 8 setpoint. 4 Which one (1) of the following describes the automatic restart capability of RCIC7. i RCIC will.. - l-
- a. automatically restan when RPV level drops below the Level 8 setpoint.
l b. NOT automatically restart unless a high drywell pressure signalis received. l'
- c. automatically restart when RPV level drops below the Level 2 setpoint.
4
- d. automatically restan when RPV level drops below the Level 3 setpoint. !
l ANSWER: c I l KA: 217000Al.03
- RO/SRO
- 4.0 )
1:
Reference:
RCIC Systems Text { ) Comments: New question (Plant Sys - Op 1) LO: 5071 and 8735 i i i ; t i i l l 1 i l Page: 33
SENIOR REACTOR OPERATOR QUESTION: 30 (1.00)
- A transient has resulted in RPV level dropping to -140" The level has remained stable for GT seven (7) minutes. SM-8 has deenergized and Division 1 Automatic Depressurization System (ADS) has been inhibited.
, After verifying the cause of the loss of SM-8, permission is granted to reenergize the bus. l l Assuming no operator actions, which one (1) of the following describes the respense of the ADS Logic to the I i reenergization of SM-8. i a. ADS will initiate 105 seconds after the discharge pressure of"C" Residual Heat Removal (RHR) pump reaches the ADS permissive. l [ , i ~
- b. ADS will not initiate if the operator resets the division "2" ADS timer within 105 seconds. J
- c. ADS will inithte immediately when the discharge pressure of"C" Residual Heat Removal (RliR) pump raaches the ADS permissive.
- d. ADS will not imtiate until the operator resets the division "1" ADS inhibit switch.
ANSWER: c. KA: 218000K5.01 l RO/SRO: 3.8
Reference:
ADS Systems Text Comments: New question (Plant Sys - Op I) LO: None Page: 34
SENIOR REACTOR OPERATOR l l QUESTION: 31 (1.00) What effect does manually decreasing the output on CAC-FC-67A ("A" recycle flow controller) have on the I Containment Atmosphere Control (CAC) system? l 1
- a. Reduces flow through the sembber and reduces oxygen concentration entering the recombiner. l
- b. Increases flow through the scrubber and increases oxygen concentration entering the recombiner.
l 1 i l c. Reduces flow through the scrubber and increases oxygen concentration entering the i recombiner.
- d. Increases flow through the scrubber and reduces oxygen concentration entering the i
recombiner. l ANSWER: b. l KA: 223001 A4.13 RO/SRO: 3.4
Reference:
CAC System Text , Comments: Question (4357)(Plant Sys - Op 1) l LO: 5133a l l l l l i i Page: 35 l
l SENIOR REACTOR OPERATOR I QUESTION: 32 (1.00) Which one (1) of the following is designed to prevent the differential pressure across the primary containment boundary from exceeding the design limit? - J
- a. Reactor building to wetwell vacuum breakers. ;
- b. Wetwell to dn/well vacuum breakers. j
- c. Standby Gas Treatment (SGT) system.
I
- d. Suppression pool "T" quenchers.
l t ANSWER: a. KA: 223001K4.06 RO/SRO: 3.3
Reference:
PPM 5.0.10 l Comments: New question (Plant Sys - Gp I) LO: 8352 (5.1.2-140) i l l Page: 36
j 2 SENIOR REACTOR OPERATOR QUESTION: 33 (1.00)
- A Loss of Coolant Accident (LOCA) has occurred, all Emergency Core Cooling Systems (ECCS) equipment
- - has functioned as designed. Present plant conditions are as follows
4 i - RPV level -135" and up slow
- RPV pressure 200 psig and down slow
- - Wetwell pressure 9 psig and up very slow j - RHR-V-42A (RPV injection valve) is open i k RHR-V-17A (Upper drywell spray inboard isolation valve) is opened in preparation for drywell spray. When i the Control Room Operator (CRO) takes the control switch for RHR-V-16A (upper drywell spray outboard i isolation) to OPEN, RHR-V-16A will i
i a. remain closed until RPV pressure drops below 135 psig. j b. open when RPV water level is GE -129" i ! c. open immediately. 4 i
- d. remain closed.
) ANSWER: d. KA: 22600lGK.07 j RO/SRO: 3.5
Reference:
RHR Systems Text Comments: New Question (Plant Sys - Gp 1) LO: 5781 1 Page: 37
SENIOR REACTOR OPERATOR QUESTION: 34 (1.00) Loss of DP-SI-2 power will render Safety Relief Valve f SRV) control switches INOPERABLE at the location (s) specified in which one (1) of the following?
- a. H13-P601 only l
- b. H13-P601 and H13-P631 (ADS division 2 logic panel) l
- c. H13-P628 (ADS division 1 logic panel) and E-CP-ARS ( Alternate remote shutdown panel)
- d. H13-P631 (ADS division 2 logic panel) and C61-P001 (Remote shutdown panel)
ANSWER: d. KA: 239002K4.05 l SRO: 3.7 l
Reference:
PPM 4.12.1.1 and RSD System Text Comments: Modified Question (461) (Plant Sys - Gp I) LO: 5885a l l I Page: 38
SENIOR REACTOR OPERATOR QUESTION: 35 _ (1.00)
. The plant is operating at 100% power when the Control Room Operator (CRO) reports that RPV pressure is
- trending down. Shortly after this report the reactor scrams and the Main Steam isolation Valves (MSIVs) ,
close. l Which one (1) of the following describes the cause of this transient?
- a. The selected Digital Electrohydraulic (DEH) pressure controller has slowly failed high
- b. The backup Digital Electrohydraulic (DEH) pressure controller has instantly failed low. :
) c. The selected Digital Electrohydraulic (DEH) pressure controller has instantly failed low.
- d. The backup Digital Electrohydraulic (DEH) pressure controller has slowly failed high.
ANSWER: a. KA: 241000K3.02 RO/SRO: 4.3 ; )
Reference:
DEH System Text ! 1 Comments: Modified Question (6232) (Plant Sys - Op I) , LO: 5286b 4 i 1 i l l i 4 1 Page: 39
i SENIOR REACTOR OPERATOR
, QUESTION: 36 (1.00)
I With a plant startup in progress and reactor power at 20%, #3 Turbine Bypass Valve (BPV)is declared
' INOP.
Which one (1) of the statements below describes the action (s) which must be taken under the above conditions? -
- a. Restore the inoperable BPV to OPERABLE status within I hour or reduce power to less than l ;
5% ofrated within the next 4 hours.
- b. Continue the startup but do not exceed 90% of rated power until the BPV has been restored to i OPERABLE status. t
) c. Restore the inoperable BPV to OPERABLE status prior to reaching 25% of rated power.
- ' d. Restore the BPV to OPERABLE within 12 hours, or suspend the startup and be in COLD ;
! SHUTDOWN within the next 12 hours. , ANSWER: c. ] l KA: 241000A2.03 !. RO/SRO: 4.2 i
Reference:
Tech. Spec. 3.7.9 i Comments: New question (Plant Sys - Op I) ] I LO: None i a i i I Page: 40
SENIOR REACTOR OPERATOR QUESTION: 37 (1.00) With the Reactor at 100% power, a trip of COND-P-2B (Condensate Booster Pump 2B) occurred. With no operator actions, following a time interval the following conditions exist:
- The reactor is scrammed
- RPV water level is at 60" and up very slow -
- RPV pressure being maintained using Safety Relief Valves (SRVs) ,
Which of the following describes the response of Reactor Building Ventilation (RBHVAC) and Standby Gas Treatment (SGT) systems to this transient?
- a. RBHVAC will NOTisolate RBHVAC fans will NOT trip SGT will auto start
- b. RBHVAC will isolate RBHVAC fans will trip SGT will auto start. j
- c. RBHVAC will NOT isolate ,
RBHVAC fans will NOT trip ! SGT will NOT start l
- d. RBHVAC willisolate RBHVAC fans trip SGT will NOT start.
ANSWER: b. KA: 261000K4.01 RO/SRO: 3.8
Reference:
.SGT System Text Comments: Modified Question (6240) (Plant Sys - Op 1)
LO: 5828 i Page: 41
. . .. - _ . - - - . - ~ _ - ..- -. - - _ - _ - _ . . - - -. ._
i SENIOR REACTOR OPERATOR QUESTION: 38 (l.00) Following full load operation for a routine surveillance, diesel generator #1 is being cooled down at idle speed. During this time a loss of off-site power occurs. j Which one (1) of the following statements describes the actions necessary to ensure proper operation of the f l diesel for reenergizing SM-7? l
)
( a. l Place the excitation mode selector switch in PARALLEL. Ensure that SW-P-1 A ("A" service i water pump) continues to run or manually trip the diesel. t l l b. Place the engine speed selector switch in RATED and place the control switch for SW-P-1 A to STOP to reset the auto start on the loss of off-site power.
- c. Place the excitation mode selector switch in PARALLEL and place the control switch for SW-P-1 A to STOP to reset the auto start on the loss of off-site power.
l d. Place the engine speed selector switch in RATED. Ensure that SW-P-1 A starts as soon as it's discharge valve cycles full closed to full open. ANSWER: d KA: 264000K6.08 l RO/SRO: 3.7 l
Reference:
Diesel Generator System Text and PPM 2.7.2A Comments: New Question (Plant Sys - Gp 1) LO: 5321 l l I r l l l 4 i l Page: 42
l SENIOR REACTOR OPERATOR QUESTION: 39 (1.00) 2 SM-8 is deenergized, which one (1) of the following is a permissive that MUST be satisfied in order for DG-2 1 to re-energize SM-87
- a. Relay 86DG2 (Engine lockout) must be reset. '
i '
- b. Breaker 3-8 (feed from SM-3) and 8-3 (feed to SM-8) must both be open.
- c. Engine control switch must be in the REMOTE position.
3
- d. Drywell pressure must be greater than 1.65 psig. ,
ANSWER: a. KA: 264000A3.05 i RO/SRO: 3.5 )
Reference:
Diesel Generator System Text 3 Comments: Similar Question (302) (Plant Sys - Gp I)
- LO: 5316 1
i i t d i l i l I i ; i l 4 Page: 43
. . . . . - . - . - . - .~ --.=.-~- . .. . . ~ . -.~ . . . - . .. . . - - - - . - . . . .
I SENIOR REACTOR OPERATOR l i QUESTION: 40 . (1.00) Steam tunnel cooling fans "A" and "B" are in service. A main steam line break results in steam tunnel l pressure in excess of 0.8 psi. t What actions will occur as a result of this transient?
- a. Standby gas treatment initiates.
I
- b. Reactor building ventilatico isolates.
- c. Steam tunnel cooling fan "C" :tuto starts.
i
- d. Steam tunnel blowout panels relie.ve. l l
ANSWER: d. i KA: 29000lK$.01 f l RO/SRO: 3.4 !
Reference:
Secondary Containment System Text l Comments: New question (Plant Sys - Gp I) LO: 7003 l l
-n 1
I 1 i i i Page: 44
SENIOR REACTOR OPERATOR ] QUESTION: 41 (1.00) I 4 j
- Following a reactor scram, the Control Room Operator (CRO) notes that CRD-FIC-600 (CRD system flow i j controller) output signal is going down l
Which one (1) of the following could cause this condition? f a. liigh charging header flow. 1
- b. High cooling header demand.
4
- c. Low drive header flow.
i s d. Low scram header flow. ANSWER: a. i j KA: 201001 A2.04 RO/SRO: 3.9 i
Reference:
CRDH System Text l j Comments: New Question (Plant Sys - Gp II)
- LO
- 5185b 4
l l - j i l ) i Page: 45
SENIOR REACTOR OPERATOR QUESTION: 42 (1.00) Select the power supply and logic configuration for the Alternate Rod Insertion (ARI) solenoid valves.
- a. 125 VDC (DP-SI-l A/2A)- must be ENERGIZED to vent the scram air header. l l
- b. 120 VAC (RPS A)- must be DE-ENERGIZED to vent the scram air header.
- c. 125 VDC (DP-SI-lD/2D)- must be DE-ENERGIZED to vent the scram air header. ;
- d. 120 VAC (IN-2)- must be ENERGlZED to vent the scram air header. .
ANSWER: a. ! KA: 201001K2.05 ! RO/SRO: 4.5
Reference:
DC Power Systems Text Comments: Modified question (175) (Plant Sys - Gp II) LO: 5262 i l l l i l l l Page: 46
1 I SENIOR REACTOR OPERATOR i QUESTION: 43 (1.00) ,
~
Aller completing the immediate Actions for a reactor scram, the Control Room Operator (CRO) notices a ! WITHDRAW rod block has been applied l l Which one (1) of the following is true concerning this condition? This rod block.. ! {
- a. will automatically be bypassed when the scram is reset. [
- b. can be manually bypassed by bypassing the RWM.
t
- c. will automatically be bypassed 10 seconds after placing the reactor mode switch in l SHUTDOWN. j
- d. CANNOT be bypassed. !
- ANSWER: d. . KA: 201002K4.02 f RO/SRO: 3.5
Reference:
RPS System Text , Comments: New Question (Plant Sys - Gp II) LO: 5952 I i 1 Page: 47
SENIOR REACTOR OPERATOR
- r
! QUESTION: 44 (1.00) The reactor is in Operational Condition 5, the Control Room Supervisor (CRS) has directed the Control ; Room Operator (CRO) to select control rod 30-55 to verify its position. The CRO reports that control rod 30-55 cannot be selected.
- j. Which one (1) of the following could cause this condition? ,
I !
- a. The refuel bridge is near or over the core.
f
- b. The fuel grapple is not full up.
i l c. Another control rod is withdrawn past "00", l l d. The fuel grapple is loaded. ANSWER: c. - 1 KA: 201002K3.01 RO/SRO: 3.4 l
Reference:
Fuel Handling Text Comments: New question (Plant Sys - Gp I) ! LO: 5359 & 5360 ; i i i i i l t Page: 48
l SENIOR REACTOR OPERATOR l 1 l QUESTION: 45 (1.00) With the plant at rated conditions a Group 1 isolation occurred with RPV pressure peaking at 1145 psig during the transient. j Which one (1) of the following describes the direct efTect on the reactor recirculation pump breakers? l a. No breakers are effected i l
- b. Only breakers CB-RPT-3 A and CB-RPT-3B trip.
- c. Breakers CB-RPT-3 A, CB-RPT-3B, CB-RPT-4A, CB-RPT-4B trip.
l
- d. Only breakers CB-RPT-4 A and CB-RPT-4B trip. ,
1 ANSWER: b. KA: 202001 A2.14 ) RO/SRO: 4.2
Reference:
RRC Systems Text, PPM 4.602. A6-1.2/1.6 Comments: Modified question (3893) (Plant Sys - Gp II) LO: 5023e 1 l t i Page: 49
SENIOR REACTOR OPERATOR QUESTION: 46 (1.00) The plant is operating at 100% power when the "A" reactor recirculation pump trips. After referring to the Power / Flow map, the CRS directs the CRO to increase flow on the "B" reactor recirculation pump. Of the following, which one (1) is the MAXIMUM allowable single loop recirculation flow? I
- a. 31,600 GPM l
- b. 37,500 GPM
- c. 41,500 GPM
- d. 45,000 GPM 1
ANSWER: c. l l KA: 20200lGK.06 l l RO/SRO: 4.1
Reference:
Tech. Spec. 3.4.1.1.3d Comments: Modified question (3877) (Plant Sys - Gp II) LO: None 1 I i 1 i I l 6 l I l Page: 50 l l l
SENIOR REACTOR OPERATOR QUESTION: 47 (1.00) 1 Following a loss of SM-7 and an Anticipated Transient Without Scram (ATWS) condition, boron injection is required. l What effect will Standby Liquid Control (SLC) initiation have on Reactor Water Cleanup (RWCU) system l valves? -
- a. RWCU-V-104 (RWCU system bypass) opens. ;
- b. RWCU-V-1 (RWCU inboard isolation) closes.
- c. RWCU-V-4 (RWCU outboard isolation) closes.
t
- d. RWCU-V-40 (RPV/RWCU return isolation) closes.
ANSWER: c. KA: 204000K4.04 RO/SRO: 3.6
Reference:
RWCU System Text Comments: New Question (Plant Sys - Gp II) i LO: 5035 l l l i l l Page: 51
I l SENIOR REACTOR OPERATOR QUESTION: 48 (1.00) l With the plant at 30% power, which one (1) of the following describes the effect that a loss of rod position information for a single control rod will have on the Reactor Manual Control System (RMCS)?
- a. A rod insert and withdraw block will be generated via the Rod Worth Minimizer (RWM).
- b. A rod withdraw block will be generated via the Rod Sequence Control System (RSCS) i
- c. A rod insert block will be generated via the Rod Position Indication System (RPIS).
- d. No rod blocks are generated, a loss of rod position indication only. '
I i ANSWER: d. KA: 214000K3.03 i RO/SRO: 3.2 l
Reference:
RMCS Systems Text l Comments: Modified Question (4286)(Plant Sys - Op II) l LO: 7754a I 1 l l 1-1 i Page: 52 l
SENIOR REACTOR OPERATOR QUESTION: 49 (1.00) The plant is operating at 100% power when both 500 KV generator output breakers trip. If the main turbine fails to trip. which one (1) of the following describes the shon term response of the main turbine Overspeed Protection Controller (OPC) for this condition?
- a. OPC initially actuates and then resets. Thereafter main turbine speed is controlled at 100% of rated by the Digital Electrohydaulic (DEH) control system.
( b. OPC initially actuates and does NOT reset. Main turbine speed coasts down to O rpm. b
- c. OPC repeatedly actuates and resets to control main turbine speed LT 103% of rated.
- d. OPC repeatedly actuates and resets to control pressure at Pressure Setpoint.
ANSWER: c.
. KA: 245000K4.09 9 RO/SRO: 3.2
(
Reference:
Comments: Main Turbine Systems Text Modified question (228) (Plant Sys - Gp II) LO: 5566 Page: 53
i i SENIOR REACTOR OPERATOR i : l QUESTION: 50 (1.00) i The plant is operating at 50% power with both Reactor Feed Pump Turbines (RFPTs) operating in automatic
- control when RFPT "A" governor valves become stuck in the present position. ;
l Assuming NO OPERATOR ACTIONS (other than rasing power), how will RFPT "A" speed and RPV level ! respond if reactor power is raised to 100%? !
. RFPT "A" SPEED RPV Level !
1 l a. Decrease Remain the same ! i b. Decrease Lower , I i I j c. Increase Remain the same !
- I
- d. Remain approximately the same Lower i i !
ANSWER' c. ! i i KA: 25900lK3.01 ! l RO/SRO: 3.9 Feedwater/ Condensate System Text l
Reference:
] Comments: New Question-(Plant Sys - Gp II) ! LO: None
- I i
j l i 1 4 i 4 i i J 5 l l l Page: 54 _-m
SENIOR REACTOR OPERATOR QUESTION: 51 (l.00) Which one (1) of the following will automatically transfer IN-2 to the alternate AC power supply?
- a. Inverter overvoltage Low frequency Missed conunutation pulse Low DC voltage.
- b. Low DC voltage Inverter overvoltage Low load Blown DC fuse.
t
- c. Inverter undervoltage Low frequency Missed commutation pulse Blown DC fuse.
- d. Inverter undervoltage Inverter overvoltage High load Low DC voltage.
ANSWER: d. KA: 262002K4.01 RO/SRO: 3.4
Reference:
UPS Systems Text Comments: Question (3799) (Plant Sys - Gp II) LO: 5892 Page: 55
I SENIOR REACTOR OPERATOR QUESTION: 52 (1.00) i A loss of 250 VDC Motor Control Center MC-S2-1 A has occurred. Which one (1) of the following describes the direct effect this condition will have on the Reactor Core Isolation Cooling (RCIC) System? l l a. RCIC initiation logic power is lost, but RCIC can still be manually initiated.
- b. RCIC-V-1 (RCIC turbine trip valve) indication and control will be lost rendering RCIC l INOPERABLE.
- c. RCIC flow control will not function in automatic, but can still be used in manual.
l
- d. RCIC valve indications are lost, however, all system functions still work.
ANSWER: b. KA: 263000K3.03 RO/SRO: 3.8
Reference:
DC Power System Text Comments: New question (Plant Sys - Gp II) ! LO: 7657 l l t Page: 56
d 4 SENIOR REACTOR OPERATOR j. QUESTION: 53 (1.00) A reactor stanup from cold conditions is in progress, a vacuum is being drawn in the main condenser using both AR-P-1 A and AR-P-1B (mechanical vacuum pumps). MS-RIS-610B (main steam line radiation monitor) has generated an INOP trip.
- Which one (1) of the following describes the effect of the above conditions?
- a. Both AR-P-1 A and AR-P-1B will trip.
- b. Neither AR-P-1 A or AR-P-1B will trip.
- c. Only AR-P-1B will trip.
]
- d. Only AR-P-1 A will trip.
i ANSWER: b. j KA: 272000Kl.02 RO/SRO: 3.5 , d
Reference:
PRM System Text ! Comments: New Question (Plant Sys - Op II) t
- 4 LO
- 5647f i ;
1 i i l l j i 4 l 1 i i l i l 0 Page: 57
SENIOR REACTOR OPERATOR QUESTION: 54 (1.00) ) 1 The plant is operating at 100% power when an EO, investigating an Accumulator Trouble alarm, reports that I an HCU Nitrogen Accumulator has completely de-pressurized. Which one (1) of the following describes the scram capability of the affected control rod?
- a. The rod can be scrammed because CRD Drive Header pressure is greater than Scram l Discharge Volume pressure. ]
- b. The rod can be scrammed because RPV pressure is greater than Scram Discharge Volume pressure. <
- c. The rod can NOT be scrammed because Nitrogen Accumulator pressure is less than RPV pressure.
- d. The rod can NOT be scrammed because Scram Inlet and Scram Outlet valves have lost their j pneumatic supply.
ANSWER: b. KA: 201003K4.04 RO/SRO: 3.7
Reference:
CRDM System Comments: Question (3524)(Plant Sys - Op III) LO: 5215 i l Page: 58
SENIOR REACTOR OPERATOR f QUESTION: 55 (1.00) While withdrawing control rods during a plant startup, the control room operator (CRO) reports that a 4 control rod will not move and appears to be stuck. Which one (1) of the following describes an option that could be used to attempt to move this control rod?
- a. Adjust cooling water flow to GT 80 gpm and allow the rod to be forced in.
4
- b. Use the Single Rod Insen (SR]) switches to scram the rod and then recover it.
- c. Apply continuous withdrawal signals in two minute increments.
4
- d. Apply a continuous insen signal, release, then apply a continuous withdrawal signal.
A.NSWER: d. I l
- KA
- 201003 A2.01 i RO/SRO: 3.6
Reference:
CRDH System Text l Comments: New question (Plant Sys - Gp III) ! LO: 5204 i 3 i i i n i l 2 + 3 i 1 I ? Page: 59
. - ...._ -.-.-.- __...-.. - __. - - - .....- .~.
2 L SENIOR REACTOR OPERATOR
- QUESTION: 56 (1.00) 4 Initial Conditions:
j - Reactor stanup in progress.
- RPV pressure =450 psig and going up
- Main condenser vacuum =23" 4-
- SM-1, SM-2, and SM-3 being powered from the startup transformer (TR-S)
! - COND-P-2A ("A" condensate booster pump) running
- CW-P-lC ("C" circulating water pump) running l - COND-P-1 A and COND-P-1B ("A" & "B" condensate pumps) running i
Maintenance has requested that Operations start the CW-P-1 A & CW-P-1B ("B" & "C" circulating water pumps) for post maintenance testing. Using the above information, determine which one (1) of the following statements is correct.
- a. Starting the third circulating water pump will cause an undervoltage trip of TR-S
- b. Operation of more than one (1) circulating water pump at this point in the stanup is not recommended due to tube erosion concerns.
- c. Starting two (2) additional circulating water pumps should not cause any significant problems for plant operations.
- d. If CW-P-1 A is started last, the transient on SM-1 will cause a trip of COND-P-1 A and COND-P-2A on over current.
ANSWER: a. K.A: 256000K6.02 RO/SRO: 3.1
Reference:
CW/TMU Systems Text, pg 15 & PPM 2.6.1 Sec. 5.3.13, Page 18 of 93 Comments: New question (Plant Sys - Gp III) LO; 7765 Page: 60
SENIOR REACTOR OPERATOR QUESTION: 57 (1.00) RFW-V-14 (condensate cleanup flow valve) is full closed and RFW-V-65 A & RFW-V-65B (RPV inlet isolation valves) are full open. Which one (1) of the following describes the response of these valves if the control switch for RFW-V-14 is ! taken to OPEN?
)
I
- a. RFW-V-65A & RFW-V-65B will remain open. RFW-V-14 will remain closed. !
l
- b. RFW-V-65A & RFW-V-65B will remain open. RFW-V-14 will open.
- c. RFW-V-65A & RFW-V-65B will close. RFW-V-14 will open when both RFW-V-65A &
RFW-V-65B are full closed.
- d. RFW-V-65A & RFW-V-65B will close. RFW-V-14 will immediately open. ,
ANSWER: a. KA. 256000GK.07 r.0/hRO: 3.4 i
Reference:
PPM 2.2.4, NOTE from Step 5.1.38, page 18 Comments: New question (Plant Sys - Gp til) LO: None ; 1 l 5 l I l Page: 61
SENIOR REACTOR OPERATOR QUESTION: 58 (1.00) The Plant is operating at = 100% power when the following annunciators are received on panel H 13-P800:
- BUS 2 OC LOCKOUT
- BUS 2 UNDERVOLTAGE
- BKR N1-2 TRIP Which one (1) of the following automatic actions is expected under the above conditions?
- a. Reactor scram due to loss of COND-P-1 A ("A" condensate pump).
i
- b. Auto start of the High Pressure Core Spray (HPCS) diesel generator.
- c. Reactor scram due to a main turbine trip.
- d. Auto start of diesel generator #1. j ANSWER: b.
KA: 295003K2.02 j RO/SRO: 4.2
Reference:
AC Systems Text Comments: Modified Question (340) (Emer & Abn - Gp I) j LO: 7767 l
)
l N Page: 62
_ - . - - . . _ _ - - ~ . - - . - - SENIOR REACTOR OPERATOR l l QUESTION: 59 (1.00) During a " Station Blackout" plant parameters are as follows:
- RPV water level -52" and up slow
- RPV pressure 850 psig and down slow
)
- Wetwell pressure 19 psig and up slow l
- Drywell temperature 243 F and up slow ;
- Wetwell temperature 112 F j
- Wetwell level + 3" l l
Which one (1) of the following interlocks must be defeated to allow continued Reactor Core Isolation l Cooling (RCIC) system operation under these conditions?
- a. High exhaust pressure turbine trip.
- b. RCIC exhaust diaphragm rupture isolation.
- c. Level 2 RCIC turbine trip.
- d. Drywell high temperature RCIC system isolation.
ANSWER: a. KA: 295003 A1.03 RO/SRO: 4.4
Reference:
PPM 5.6.1 Comments: Modified Question (4402) (Emer & Abn - Gp I) LO: 5722 1 Page: 63
- - . - . - - . . . - . . - . - - . . - - - . - . - , . - . . _ . ~ - . . - - _ _ . - . . .--.
I SENIOR REACTOR OPERATOR
- i. ,
! QUESTION: 60 (1.00) l A reactor scram hasjust occurred. The Rod Sequence Control System (RSCS) and Rod Worth Minimizer l (RWM) have not functioned to give the ALL RODS IN information. !
1 1 Which one (1) of the following H13-P603 indications may be used to verify rods full in?
- i j a. White Reactor Protection System (RPS) group lights deenergized. j
- b. Amber backup scram lights deenergized. l
)
^
! c. Green fullin lights energized i j d. Blue scram lights energized. , 5 ANSWER: c. ! i l KA: 295006GK.05 ! 1 RO/SRO: 4.0 i
- l
Reference:
RSCS Systems Text l j Comments: Modified question (3746)(Emer & Abn - Op I) LO: 5807 t h l i i i s 4 i 1 i 1 1 i i s 1 I Page: 64
I l SENIOR REACTOR OPERATOR ; l QUESTION: 61 (1.00) Initial plant conditions are as follows:
- Reactor power 100 % :
- RPV pressure 1020 psig :
- RPV water level 36" A reactor scram occurs and the scram inlet valve (126) of a single control rod mechanically binds and fails to open.
Which one (1) of the following describe the control rod's response to this failure?
- a. Fully inserts and its blue scram light is energized.
- b. Fails to insert and its blue scram light is energized.
- c. Fully inserts and its blue scram light is deenergized.
- d. Fails to insert and its blue scram light is deenergized.
i ANSWER: c. ; KA: 295006A1.06 i RO/SRO: 3.6
Reference:
CRDH System Text l; Comments: Modified Question (341)(Emer & Abn - Op I) LO: 5184 1 l l l I l Page: 65
- - .. ..-. . -_ - - - . . - . _ - .- ._- .. . . - . . . . _. . . _ _ . - ..= . . - . . .
j SENIOR REACTOR OPERATOR ! QUESTION: 62 (1.00) The reactor is operating at =98% power. An equipment operator reports a lube oil leak in the "B" feedwater 1 pump room. Immediately after acknowledging the report, RFW-P-1B ("B" reactor feedwater pump) trips on
- low lube oil pressure.
Which one (1) of the following describes the effect this condition has on the reactor recirculation system?
- a. Only RRC-P-1B ("B" reactor recirculation pump) will runback to 15H2.
4
- b. Both RRC-P-1 A and RRC-P-1B ("A" & "B" reactor recirculation pumps) will runback to 27Hz.
- c. Both RRC-P-1 A and RRC-P-1B ("A" & "B" reactor recirculation pumps) will runback to 15 Hz.
l_ J
- d. Only RRC-P-1B ("B" reactor recirculation pump) will runback to 52.2 Hz.
l
- ANSWER
- b.
j l KA: 295009K2.03
- RO/SRO: 3.2
{
Reference:
ASD System Text j Comments: New question (Emer & Abn - Gp I) i LO: 9683 1 i i 1 4 i d i i } l f 1 4 i ) -f Page: 66
SENIOR REACTOR OPERATOR QUESTION: 63 (1.00) . Which one (1) of the follwoing protective features is designed to actuate to ensure net positive suction head l for the Reactor Recirculation pumps?
- a. Level 1 trip l
- b. Level 2 trip
- c. Level 3 runback
- d. Level 4 runback ANSWER: b. i KA: 295009A1.03 RO/SRO: 3.1
Reference:
RRC System Text Comments: Modified Question (5556) (Emer & Abn - Op I) LO: 5023 l l l I i l i I Page: 67
SENIOR REACTOR OPERATOR l 1 QUESTION: 64 (1.00) 1 e The plant has experienced a transient, Emergency Operating Procedures (EOPs) have been entered and conditions are as follows: RPV water level -150"and down slow l RPV pressure 180 psig and down slow Wetwell temperature 110 F and up slow RHR loop "A" injecting to the RPV RHR loop "B" in suppression pool cooling All other injection sources are unavailable ) 1 Which one (1) of the following statements best describes actions that need to be taken given the above information? l
- a. Open seven (7) Automatic Depressurization System (ADS) Safety Relief Valves (SRVs) to emergency depressurize.
l
- b. RHR loop "B" should be removed from suppression pool cooling and injected into the RPV. '
i l c. RHR loop "A" should be removed from injection and placed into suppression pool cooling.
- d. No actions are required until RPV level lowers to LE -192" ANSWER: b. j
! KA: 295013K2.01 RO/SRO: 3.7
Reference:
PPM 5.2.1 Comments: New question (Emer & Abn - Op I) l LO: 8304 l l t i Page: 68
. . - . .. .~ . -- - -= ~ -- - ._ - . - . - . . . . - . - - - - . - -.. _ . - --
( SENIOR REACTOR OPERATOR l ! QUESTION: 65 (1.00) l j The control room operator (CRO) is withdrawing control rods with the reactor critical and power indicating on the IRMsjust prior to the point of adding heat. The CRO observes an unexpected rapid increase in power and a period indication of =30 seconds. ? Assuming NO OPERATOR ACTION, which one (1) of the following scram signals will terminate this transient? l 1
- a. Reactor short period.
- b. Average Power Range Monitor (APRM) neutron flux high.
- c. Source Range Monitor (SRM) upscale
- d. Intermediate Range Monitor (IRM) neutron flux high ANSWER: d.
KA: 295014 A2.01 RO/SRO: 4.2
Reference:
IRM System Text Comments: Modified Question (249)(Emer & Abn - Gp I) LO: 5459 Page: 69
- _- _. _ . _ . _ _ . _ . _ . . _ . . _ _ _ _ _ . _ _ _ _ _ _ . - _ _ _ _ . . _ _ _ . _ _ - _ _ . _1 I 1
, SENIOR REACTOR OPERATOR QUESTION: 66 (1.00) l The plant is operating at =98% power when the following indications are noted:
s
- Reactor power down slow.
j Megawatts down slow. Control air pressure down slow.
- Three (3) control rods indicate FULL-IN with scram lights energized on the full core display.
I i Which one (1) of the following statements describes the actions required to be taken given the above
- indications.
i ' 1 l a. NO actions are required until the first Main Steam Isolation Valve (MSIV) is showing dual : position indication.
- b. Close or verify closed CN-V-65 (containment instrument air crosstie shut-off valve.
i
- c. Initiate a manual reactor scram and refer to PPM 3.3 l.
l
- d. Lower core flow to reduce reactor power to LT 90% of rated core thermal power.
l } ANSWER: c. 4 1 KA: 295015Al.02
- RO/SRO: 4.2
Reference:
- CAS System Text and PPM 4.1.1.7B Comments New Question (Emer & Abn - Gp I) j LO: 7605 i i l i j l~ i j 4 a k i i 1 Page: 70.
SENIOR REACTOR OPERATOR QUESTION: 67 (1.00) Which one (1) of the following systems was specifically designed to ensure reactor power could be monitored under DBA/LOCA conditions?
- a. Source Range Monitoring (SRM) system. i
- b. Local Power Range Monitoring (LPRM) system.
- c. Wide Range Monitoring (WRM) system.
- d. Intermediate Range Monitoring (IRM) system i
ANSWER: c. ; l KA: 295015K2.08 RO/SRO: 3.7
Reference:
WRM System Text Comments: Modified Question (3707)(Emer & Abn - Gp I) LO: 5963 Page: 71
t . SENIOR REACTOR OPERATOR j QUESTION: 68 (1.00)
- The control room evacuation procedure directs isolating certain control room current meters using the
, Current Transformer (CT) shorting switches. i , ) Which one (1) of the following lists the locations of these CT shoning switches?
- a. "B" residual heat removal pump, "B" service water pump, SL-71 feeder, and SL-73 feeder breaker cubicles.
1 i b. "A" residual heat removal pump,"A" service water pump, SL-81 feeder, and SL-83 feeder , breaker cubicles. l' c. "A" residual heat removal pump, "A" service water pump, Startup Transformer (TR-S) feeder, 3 SL-71 feeder, and SL-73 feeder breaker cubicles. i d. "B" residual heat removal pump, "B" service water pump, Backup Transformer (TR-B) feeder,
- SL-81 feeder, and SL-83 feeder breaker cubicles.
e ! ANSWER: d. i KA: 295016GK.06
- RO/SRO: 4.1 l
Reference:
Remote Shutdown Panel System Text, pg.14 Comments: New Question (Emer & Abn - Op I) LO: 7737 l 1 I l i l 6 i 3. I. k-Page: 72
SENIOR REACTOR OPERATOR l QUESTION: 69 (1.00)
- A "Most Immediate" control room evacuation is required due to heavy smoke intrusion.
- Which one (1) of the following statements lists only IMMEDIATE ACTIONS that should be taken prior to
! exiting the control room? I.
- a. Manually scram the reactor, lock the reactor mode switch in SHUTDOWN and close the Main l Steam Isolation Valves (MSIVs).
- b. Manually scram the reactor, initiate Reactor Core Isolation Cooling (RCIC) and make a plant ;
i announcement. i
- c. Manually scram the reactor, close the MSIVs and transfer RPV level control to RFW-FCV-10A and RFW-FCV-10B (feedwater startup valev to the reactor).
I f d. Manually scram the reactor, lock the reactor mode switch in SHUTDOWN and start diesel generator #2 ] j ANSWER: a. J . KA: 295016K3.01 RO/SRO: 4.2 i
Reference:
PPM 4.12.1.1 Comments: Modified Question (256) (Emer & Abn - Gp I) LO: None Page: 73
SENIOR REACTOR OPERATOR QUESTION: 70 (1.00) The plant has experienced a transient, PPM 5.1.2 has been entered, plant parameters are as follows:
- RPV water level -145" and steady
- Drywell pressure 10 psig and down slow
- Wetwell temperature 110 F and up very slow
- Main Steam Isolation Valves (MSIVs) are closed
- Both Standby Liquid Control (SLC) pumps are injecting Which one (1) of the following identifies a valid annunciator that would preclude / prevent reopening the ;
MSIVs?
- a. LPCS/RHR A INIT RPV LEVEL LOW -129"
- b. DRYWELL PRESS HIGH TRIP.
- c. NSSSS ISOL MSL FLOW HIGH.
- d. RC-1 HALF TRIP.
ANSWER: c. KA: 295017K3.01 RO/SRO: 3.9
Reference:
PPM 5.1.2 and NS4 System Text Comments: Modified Question (677) (Emer & Abn - Gp I) LO: Nonc l i l Page: 74
l SENIOR REACTOR OPERATOR QUESTION: 71 (1.00) A spent fuel assembly is dropped during transport in the spent fuel pool. The bridge operator observes bubbles rising from the dropped assembly. Which one (1) of the following is an IMMEDIATE ACTION for this situation?
- a. Place all assemblies in a safe location, leave the area, and call the control room.
- b. Immediately evacuate the refuel floor of all personnel.
- c. Contact Health Physics and ask for an area survey, then inform the Control Room Supervisor )
(CRS). ]
- d. Contact the refuel floor supervisor and the system engineer, then attempt to recover the l dropped assembly.
ANSWER: b. KA: 295023K1.01 RO/SRO: 3.9
Reference:
PPM 4.12.3.1, Rev and Fuel liandling System Text i Comments: Modified Question (508)(Emer & Abn - Op I) LO: 7713 i l l i l 1 Page: 75 i
SENIOR REACTOR OPERATOR QUESTION: 72 (l.00) PPM 5.2.1 " Primary Containment Control" directs that when drywell pressure exceeds 39 psig the primary i containment is to be vented to reduce and maintain wetwell pressure below the Primary Containment Pressure Limit (PCPL). Which one (1) of the following statements describes the preferred vent path and the reason that this path is l preferred?
- a. Drywell, this is the vent path with the highest flowrate capacity.
- b. Wetwell, to take advantage of suppression pool scrubbing for minimizing the amount of radioactivity released.
- c. Drywell, in order to minimize the moisture saturation and breakdown of the Standby Gas Treatment (SGT) system charcoal adsorbers.
- d. Wetwell, in order to minimize cycling, and potential failure of the wetwell to drywell vacuum breakers.
ANSWER: b. KA: 295024K3.07 RO/SRO: 4.0 Reference. PPM 5.2.1 Comments: Question (512) (Emer & Abn - Op I) LO: 8363 (PPM 5.2.1) Page: 76
SENIOR REACTOR OPERATOR QUESTION: 73 (1.00) The plant was operating at =98% power when a leak in the discharge of a condensate booster pump caused a low suction pressure trip of the reactor feedwater pumps. RPV level dropped to -25" initially and is now going down very slow, the Control Room Supervisor (CRS) has entered PPM 5.1.1, RPV Control, and is executing alllegs concurrently. Wetwell temperature hasjust been reported at 92 F and up slow. Which one (1) of the following describes the Emergency Operating Procedure (EOP) implementation to be used under these conditions? I
- a. Continue PPM 5.1.1, RPV Control, RPV level steps, AND enter PPM 5.3.1, Secondary Containment Control.
- b. Continue PPM 5.1.1, RPV Control, AND concurrently enter PPM 5.2.1, Primary Containment !
Control.
- c. Complete PPM 5.1.1, RPV Control, RPV level steps, THEN enter PPM 5.2.1, Primary Containment Control.
i
- d. Reenter PPM 5.1.1, RPV Control, AND concurrently enter PPM 4.12.4.l A High Energy Line Break.
ANSWER: b. KA: 295024GK.I1 RO/SRO: 4.5
Reference:
PPM 5.0.10, Sect. 3.5, PPM 5.1.1, PPM 5.2.1 Comments: Modified Question (721)(Emer & Abn - Gp I) LO: 8017 Page: 77
W
! SENIOR REACTOR OPERATOR QUESTION: 74 (1.00) i A plant transient has occurred which has caused an isolated Anticipated Transient Without Scram (ATWS) condition with Average Power Range Monitors (APRMs) indicating = 15% power and RPV Pressure = 1080 psig. The Control Room Supervisor (CRS) directs the Control Room Operator (CRO) to restore and maintain l RPV pressure 800 to 1000 psig using Safety Relief Valves (SRVs).
1 l j Which one (1) of the following describes why pressure is being lowered and maintaincd below it's initial
- value?
i This pressure band., ! a. avoids Safety Relief Valve (SRV) lifting due to high pressure. i j b. ensures that the shutoff head of the Standby Liquid Control (SLC) pumps is not exceeded. i c. prevents the Heat Capacity Temperature Limit (HCTL) from being exceeded.
- d. Ensures that any potential break leak rate will be lower than the design basis for Emergency Core Cooling Systems (ECCS).
I 4 ANSWER: a. l KA: 295025K2.01 l RO/SRO: 4.1
Reference:
PPM 5.0.10, Sect. 8.23. step P-5 i Comments: Modified Question (678) (Emer & Abn - Gp I)
- LO. 8163 i
[ s 4 Page: 78 ]
SENIOR REACTOR OPERATOR QUESTION: 75 (1.00) The plant is in a condition requiring the Control Room Supervisor (CRS) to execute PPM 5.1.1, RPV Level Control, and PPM 5.2.1, Primary Containment Control, concurrently. The CRS has directed a pressure ; reduction which exceeds the normal, allowable RPV cooldown rate of 100 F/Hr. ; Which one (1) of the following describes a condition that would allow the CRS to take this action? j l
- a. Prevent RPV level from going LT Top of Active Fuel (TAF).
- b. Prevent exceeding Drywell Spray Initiation Limit (DSIL).
- c. Prevent exceeding Heat Capacity Temperature Limit (HCTL).
- d. Prevent exceeding Maximum himary Containment Water Level Limit (MPCWLL).
ANSWER: c. KA: 295025A2.03 RO/SRO: 4.1
Reference:
PPM 5.0.10 Comments: Modified Question (Emer & Abn - Op I) LO: 8048 (PPM 5.1.1) l Page: 79
SENIOR REACTOR OPERATOR QUESTION: 76 (1.00) A plant transient has caused a reactor scram. The following conditions exist:
- APRM indication 10 %
- RPV level -3 0"
- RPV pressure 1000 psig
- Drywell pressure 0.25 psig l
- Drywell temperature 125 F
- Wetwell level - 1. 5" !
- Wetwell temperature 95 F l Which one (1) of the following identifies the Emergency Operating Procedures (EOPs) that should have been entered given the above conditions?
- a. PPM 5.1.3, Emergency RPV Depressurization ,
PPM 5.3.1, Secondary Containment Control { PPM 5.1.2, RPV Control- ATWS '
- b. PPM 5.1.1, RPV Control '
PPM 5.2.,1 Primary Containment Control PPM 5.3.1, Secondary Containment Control
- c. PPM 5.1.1, RPV Control PPM 5.2.1, Primary Containment Control )
PPM 5.1.5, Emergency RPV Depressurization - ATWS
- d. PPM 5.1.1, RPV Control PPM 5.1.2, RPV Control- ATWS PPM 5.2.1, Primary Containment Control l
ANSWER: d. KA: 295026GK.12 RO/SRO: 4.5
Reference:
PPM 5.0.10 Comments: Modified Question (732)(Emer & Abn - Op I) LO: 8017 (PPM 5.0.10) Page: 80
SENIOR REACTOR OPERATOR QUESTION: 77 (1.00) A plant transient has caused a reactor scram. Plant conditions are as follows:
- Reactor power = 15%
- RPV pressure 10P3 psig and steady c - RPV level -125" and down
/ - Wetwell temperature 165
- F and up slow h - Wetwell level 32.5' and up very slow Which one (1) of the following describes the operation of the Safety Relief Valves (SRVs) with the above conditions?
- a. Heat Capacity Level Limit (HCLL) has been exceeded, emergency depressurization is required.
{
- b. Safety Relief Valve Tailpipe Level Limit (SRVTPLL) has been exceeded, emergency depressurization is required.
- c. Heat Capacity Temperature Limit (HCTL) has been exceeded, emergency depressurization is '
required.
- d. No limits 'r.ne been exceeded, cycle SRVs to maintain RPV pressure between 800 and 1000 psig.
ANSWER: c. KA: 295026K3.01 RO/SRO: 4.1 Reference. PPM 5.2.1, ADS System Ten Comments: New question (Emer & Abn - Gp I) LO: 8379 (PPM 5.2.1-77) h . 9 Page: 81
-m _ . _ , . , _ . . . - - - - - - -
. - _ . . -- - . . - . . _ . = - -. .. . - - - - - . . - - _ -
SENIOR REACTOR OPERATOR QUESTION: 78 (1.00) Which one (1) of the following is a consequence of exceeding the Heat Capacity Level Limit (HCLL)? '
- a. The Safety Relief Valve (SRV) tailpipe "T" quenchers become uncovered.
- b. RPV depressurization may challenge the pressure suppression function.
- c. Safety Rdief Valve (SRV) tailpipe chugging may occur.
- d. Emergency Core Cooling Systen. JCCS) pumps will cavitate due to the loss of available Net Positive Suction Head (NPSH).
ANSWER: b. KA: 295030K1.03 RO/SRO: 4.1
Reference:
PPM 5.0.10, Sect 7 3 : Comments: Modified Question (6099) (Emer & Abn - Op 1) LO: 8377 (PPM 5.0.10) ; i l l l l : i i Page: 82
SENIOR REACTOR OPERATOR QUESTION: 79 (1.00) Which one (1) of the following defines the minimum level for the Heat Capacity Level Limit (HCLL)?
- a. The SRV tailpipe quenchers.
- b. The RCIC Turbine Exhaust line.
1 c. The Drywell-Wetwell Downcomers.
- d. The RHR/LPCS suction Strainer.
ANSWER: c. I L ._ KA: 295030GK.07 RO/SRO: 3.9
Reference:
PPM 5.0.10, pg 87 Comments: Modified Question (685)(Emer & Abn - Gp I) LO: 7785 (PPM 5.2.1-76) i
^
Page: 83
1 SENIOR REACTOR OPERATOR QUESTION: 80 (1.00) During an Anticipated Transient withour Scram (ATWS) condition with the reactor not shutdowr, reactor ; water level is intentionally lowered to... !
- a. the point were reactor power is LT 5% and maintained at LE that level until Cold Shutdown Boron Weight (CSBW) has been injected,
- b. -65" and maintained regardless of reactor power to ensure adequate core cooling.
- c. -50", the combination of the reduced level and Reactor Core Isolation Cooling (RCIC) injection will help reduce reactor power.
- d. -65" to -192" to suppress reactor power while maintaining adequate core cooling.
ANSW'dR: d. KA: 295031 A2.01 RO/SRO: 46
Reference:
PPM 5.0.10, PPM 5.1.2 Comments: New Question (Emer & .Abn - Gp I) LO: 8149 (PPM 5.1.2-48) Page: 84
I SENIOR REACTOR OPERATOR QUESTION: 81 (1.00) ) i In PPM 5.1.4, RPV Flooding, achieving FLOODING COMPLETION TIME ensures that RPV level is GE ' t o. .
- a. the Top of Active Fuel (TAF).
- b. the Main Steam Line (MSL) openings. i
- c. 2/3 core height.
- d. the reactor head vents.. l l
ANSWER: a. i KA: 295031K1.01 RO/SRO: 4.7
Reference:
PPM 5.0.10, pg 137 , Comments: Modified Question (659) (Emer & Abn - Gp I) ' LO: 8219 (PPM 5.1.4-14) I Page: 85 l
SENIOR REACTOR OPERATOR QUESTION: 82 (1.00) When using the Reactor Core Isolation Cooling (RCIC) system for alternate boron injection, the contents of l the Standby Liquid Control (SLC) storage tank are gravity fed to the RCIC pump suction by a temporary hose connection originating at... l
- a. any drain off the SLC suction piping. !
- b. the drain off of the SLC storage tank.
i
- c. a drain on the common SLC discharge header, downstream of SLC-V-4 A & SLC-V-4B (squib i
valves). l d. the tank side of either the "A" or "B" SLC system relief valve piping. t I i l ANSWER: d. l KA: 295037K2.13 l RO/SRO. 4.1
Reference:
PPM 5.5.8 and RCIC System Text l Comments: Modified Question (3728)(Emer & Abn - Op I) l LO: 5929 I I l i 9 t Page: 86
SENIOR REACTOR OPERATOR QUESTION: 83 (1.00) Which one (1) of the following describes two (2) methods that can be used for positive confirmation that all rods are fully inserted? - i
- a. Average Power Range Monitors (APRMs) LT 5% power and Reactor Engineering calculation showing adequate shutdown margm.
! i
- b. Graphic Display System (GDS) and Plant Process Computer Replacement System (PPCRS). '
- c. Plant Process Computer Replacement System (PPCRS) ani Quick Emergency Dose Projection ,
System (QEDPS). l d. Graphic Display System (GDS) and Average Power Range Monitors (APRM)s LT 5% power. j ANSWER: b. ; I KA: 295037A2.01 RO/SRO: 4.3 l
Reference:
PPM 5.0.10, page 222 l- Comments: New Question (Emer & Abn - Gp II) l ! LO: 8182 (5.1.1-37 & 5.1.2-58) i r r ! b I $ i ! I , Page: 87 i
SENIOR REACTOR OPERATOR QUESTION: 84 (1.00) The plant is operating at 75% power and 70% core flow when an electrical malfunction in the main turbine trip circuitry causes both reactor recirculation pumps to trip off. Which one (1) of the following IMMEDIATE ACTIONS should be taken? I
- a. The recirculation pump trips will cause a RPV high pressure scram. Perform the immediate scram actions per PPM 3.3.1..
- b. Refer to the single loop operating procedure in PPM 2.2.1 to restan one of the reactor recirculation pumps.
- c. Confirm the loss of both reactor recirculation pumps and then manually scram the reactor. i
- d. Use the fast shutdown sequence control rods to exit Region "C" within 15 minutes.
1 1 ANSWER: c. KA: 295001G.10 RC'MO: 3.7
Reference:
PPM 4.12.4.7 Comments: Question (503)(Emer & Abn - Gp II) LO: 5023c Page: 88
. _ . ~. _ - .. _ _ _ _ . _ _ _ _ . . _ _
SENIOR REACTOR OPERATOR ! l QUESTION: 85 (1.00) The reactor is operating at 93% power when a loss of all circulating water pumps occurs. Assuming NO l OPERATOR ACTION, as vacuum degrades to 14" Hg, what will be the effect on RPV water level? RPV water level will..
- a. increase to +54" and then cycle t etween -50" and +54".
- b. decrease to LT 0" and then stabilize it +18"
- c. be maintained at setpoint.
t
- d. decrease to +13" and then stabilize at +36" ANSWER: b. i KA: 295002K3.01 RO/SRO: 3.8
Reference:
FWLC System Text Comments: Modified Question (6270)(Emer & Abn - Gp II) LO: 5400f j l l l l t Page: 89 l l
i SENIOR REACTOR OPERATOR QUESTION: 86 (1.00) Due to a fault, MC-7A has been deenergized and will be out of service for a minimum of eight (8) hours. Which one (1) of the following will be affected by this condition?
- a. Uninterruptable Power Supply (UPS) static inverter IN-1.
- b. Critical instrument inverter IN-2.
- c. ATWS/ARI Division 2 logic power. i 1
l
- d. DG-1 control circuit power l ANSWER: a.
KA: 295004K2.03 RO/SRO: 3.3
Reference:
DC Power System Text Comments: New question (Emer & Abn - Gp II) LO: 5263 l l l J Page: 90
SENIOR REACTOR OPERATOR QUESTION: 87 (1.00) The turbine throttle valve closure and governor valve fast closure Reactor Protection System (RPS) scrams are bypassed with reactor power LT 30%. Which one (1) of the following statements correctly describes the bases for allowing these trips to be l bypassed? I l a. A smaller void coeflicient at lower power levels adequately reduces the severity of the turbine l trip or load reject.
- b. At low power levels, adequate margin to the Minimum Critical Power Ratio (MCPR) safety limit is provided by the RPV steam dome pressure high scram.
1
- c. Load reject transients, which are most severe, cannot occur because the Digital Electrohydraulic (DEH) overspeed protection controller is bypassed.
- d. Below 30 % power, normal plant steam loads and bypass valve operation are suflicient to prevent Safety Relief Valve (SRV) operation.
ANSWER: b. KA: 295005K1.02 RO/SRO: 3.6
Reference:
Tech. Spec. Bases 2.2.1.3 Comments: Question (248)(Emer & Abn - Op II) ! LO: None l l I l l 1 4 i ( Page: 91
;_..._._ __ _-_.m._ _ ._ _ . . . . . . ._ _ _ _...-.___ -___
i i i ] SENIOR REACTOR OPERATOR QUESTION: 88 (1.00) A plant startup is in progress with reactor pressure =500 psig. RFW-FCV-10A and RFW-FCV-10B
- (feedwater startup valves to the reactor) both fail full open. RPV level is 55" and rising.
What IMMEDIATE ACTIONS should be taken to preclude flooding the main steam lines? l j l
- a. Prior to reaching an RPV level of +80", scram the reactor, and close the Main Steam Isolation I
i Valves (MSIVs). i l
- b. Stop the condensate booster pumps before RPV water level exceeds +80" l
- c. Prior to reaching an RPV level of +108", close RFW-V-118 (feedwater startup valve isolation) l l and leave it closed until RPV level is LT +54" l
! l 4 j d. Stop all condensate and condensate booster pumps before RPV water level exceeds +60" ! i ANSWER: b. j KA: 295008 Al.08 RO/SRO: 3.5
Reference:
PPM 4.2.1.2 Comments: Modified Question (504)(Emer & Abn - Op II) LO: 5400 4 Page: 92
. . - . . . - - . . . . . - . - - - . . - . - . . . - . - = - - . . . - . ._ - .- , . - - . .
SENIOR REACTOR OPERATOR 1 QUESTION: 89 (1.00)
- With reactor power at 68% and TSW-P-1B ("B" plant service water pump) Out Of Service (OOS)and danger
- tagged for maintenance, an electrical fault causes TSW-V-53 A ("A" plant service water pump discharge valve) to close and TSW-P-1 A ("A" plant service water pump) to trip. Attempts to restart the pump and open the valve have failed.
1 Which one (1) of the following IMMEDI ATE ACTIONS is required? i l l a. Monitor main generator temperatures and reduce load as necessary. 5 2
- b. Monitor control air compressors for high temperature alarms and transfer to alternate cooling as required.
r
- c. Trip RRC-P-1 A ("A" reactor recirculation pump) and refer to the single loop operating l
, procedure, i
- d. Manually scram the reactor and refer to PPM 3.3.1, l ANSWER: d.
f KA: 295018K2.02 RO/SRO: 3.6 TSW Systems Text and PPM 4.8.4.1
Reference:
. Comments: Modified Question (260) (Emer & Abn - Gp II)
)
l LO: 5854 i
)
l Page: 93
l SENIOR REACTOR OPERATOR QUESTION: 90 (1.00) Which one (1) of the following is expected to occur at a control air header pressure of 80 psig?
- a. S A-PCV-2 (control / service air crosstie valve) closes.
- b. Standby control air compressor (s) automatically start.
- c. CAS-PCV-1 (desiccant dryer bypass valve) opens.
- d. Control air header low pressure alarm is received.
ANSWER: a. KA: 295019A2.01 RO/SRO: 3.6
Reference:
CAS System Text Comments: New Question (Emer & Abn - Op II) LO: 5878 i Page: 94 1
SENIOR REACTOR OPERATOR QUESTION: 91 (l.00) CAS-C-1 A ("A" control air compressor) tripped due to high HP cylinder discharge air temperature following j a loss of Plant Service Water (TSW). Fire water has been aligned to Cooling Jack Water (CJW) heat exchanger. The control switch for CAS-C-1 A is still in RUN. Which one (1) of the following describes the conditions that will restart CAS-C-1 A? j CAS-C-1 A will restart.. 1
- a. automatically if system pressure is LE 110 psig. )
- b. when the high temperature alarm is cleared and the control switch is taken to STOP and then back to RUN.
- c. when the high temperature alarm is cleared and the RESET pushbutton at the local control cabinet is depressed.
I
- d. when the high temperature alarm is cleared. !
ANSWER: c. l I KA: 295019GK.05 1 RO/SRO: 3.3
Reference:
CAS System Text Comments: Modified Question (4078) (Emer & Abn - Gp II) LO: 5872 . Page: 95
. - . - . - ~ - . _ . . .. .- . - - _ . - . - - . _ - - - . - .-_ .- - . . . - -.
i SENIOR REACTOR OPERATOR l QUESTION: 92 (1.00) l The plant is operating at 100% reactor power when an inadvertent Group 7 Nuclear Steam Supply Shutoff i System (NS4) isolation occurs. Which one (1) of the following describes the expected plant response?
- a. Reactor Closed Cooling (RCC) system supply and return containment isolation valves close. -
- b. Reactor Water Cleanup (RWCU) system containment isolation valves close. ,
- c. Residual Heat Removal (RHR) system reactor water sample isolation valves close. '
]
- d. Primary containment recirculation fans trip. i i
1 . ANSWER: b. 1 i
- KA
- 295020K2.04 RO/SRO: 3.1
Reference:
NS4 System Text Comments: Question (259)(Emer & Abn - Op II) LO: 5598 l i Page: 96
SENIOR REACTOR OPERATOR QUESTION: 93 (1.00) A reactor shutdown to cold conditions is in progress. Plant conditions are as follows:
- Reactor mode switch positioned in SHUTDOWN 3
- RPV pressure is 45 psig
- Residual Heat Removal (RHR) "B" is being warmed up for shutdown cooling mode
- Residual Heat Removal (RHR) "A" has been removed from service for ten (10) days i Which one (1) of the following statements describes action (s) which must be taken for these conditions? ]
- d. 4
- a. Immediately place RHR Loop "B" in shutdown cooling and be in at least cold shutdown within i one (1) houc
) b. Perform a physical walkdown of the Reactor Water Cleanup (RWCU) system and then place the system in service to maintain reactor coolant temperature as low as possible. !
- c. Demonstrate operability of at least one (1) alternate method of decay heat removal.
- d. Maintain both reactor recirculation pumps in operation until RHR-P-2A ("A" residual heat ,
j removal pump)is repaired and returned to serv ce. l ANSWER: c. KA: 295021GK.08 RO/SRO: 3.9
Reference:
Tech Spec 3.4.9.1 Comments: Question (577)(Emer & Abn - Gp 11) LO: None Page: 97
SENIOR REACTOR OPERATOR QUESTION: 94 (1.00) The plant is operating at =97% when the following annunciators are received:
- H 13-P603. A7-6.7 ROD ACCUMULATOR TROUBLE
- (Thefull core display indicates this alarm isfor afidly withdrawn rod) i H 13-P603. A7-3.8 CRD CHARGE WATER PRESS LOW
- The control room operator (CRO) observes that CRD-P-1 A ("A" control rod drive pump) motor current
- indicates zero (0) amps with the red light on.
l l Which one (1) of the following describes the IMMEDIATE ACTIONS required for this situation?
- a. Place the reactor mode switch to SHUTDOWN and carry out the scram recovery per PPM 3.3.1, Reactor Scram.
- b. Place the Control Rod Drive (CRD) flow controller in MANUAL and raise controller output while monitoring CRD-P-1 A motor current.
- c. Place the standby CRD suction filter in service locally and start CRD-P-1B.
- d. Place the CRD flow controller in MANUAL, set the controller output at zero (0) and start CRD-P-1 B.
ANSWE.R: d. KA: 295022Al.01 RO/SRO: 3.4
Reference:
PPM 4.1.1.2 _ Comments: Modified Question (360)(Emer & Abn - Gp II) LO: 5192 i
)
1 Page: 98 I
SENIOR REACTOR OPERATOR ' l l ( QUESTION: 95 (1.00) Following a small steam line break inside primary containment, average drywell temperature has increased by t about 100
- F.
Assuming that actual RPV water level remains constant, indicated vessel level could be.. l
- a. higher, as heating of the reference leg decreases differential pressure.
l1 l
- b. lower, as heating of the reference leg increases differential pressure.
l i c. higher, as heating of the reference leg increases differential pressure. i
- d. lower, as heating of the reference leg decreases differential pressure.
ANSWER: a. KA: 295028Kl.01 RO/SRO: 3.7
Reference:
PPM 5.0.10 , Comments: Question (567)(Emer & Abn - Gp II) LO: 8448 (PPM 5.1.1-65) l l l l l Page: 99
. . - . . - . _ - - - . . - . . _ . . . . ~ . - - -. . . .
1 i SENIOR REACTOR OPERATOR , i ! QUESTION: 96 (1.00) ) 1 ! Given plant conditions as follows: !
- - Wetwell level 36'
- Wetwell pressure 10 psig i- - RPV pressure 1000 psig Using the attached curves, identify the possible results of Safety Relief Valve (SRV) actuation. j Actuation of an SRV..
i a. is allowed and desired given the above conditions. 4 :
- b. at this elevated wetwell level could result in damage to SRV internals.
l l
- c. will result in exceeding the suppression pool boundary design load. l
- d. could result in damage to the SRV tail pipe, quenchers, or suppons. l i ANSWER: d. i 1
KA: 295029K2.06 ! RO/SRO: 3.5
Reference:
PPM 5.0.10, Section 7.16
, Comments: New Question (Emer & Abn - Gp II)
- LO: 8381 (PPM 5.2.1-83) ,
i i i I l Page: 100
SENIOR REACTOR OPERATOR QUESTION: 97 (1.00) PPM 5.3.1, Seconday Containment Control, was entered due to confirmed high temperatures and steam in the 1 A Reactor Water Cleanup (RWCU) pump room. RWCU-V-1 & RWCU-V-4 (RWCU suction isolation valves) cannot be isolated from the control room. Maximum Safe Operating Values.for the RWCU system I have NOT been exceeded. Which one (1) of the following describes the actions to be taken for this situation?
- a. Emergency depressurize.
i b. Shutdown the reactor per PPM 3.2.1. t
- c. Continue efforts to isolate RWCU and enter PPM 5.1.1, RPV Level Control.
- d. Isolate Reactor Building Ventilation (RBHVAC) and initiate the Standby Gas Treaetment (SGT) system.
ANSWER: c. ] 4 KA: 295032K3.03 t RO/SRO: 3.9
Reference:
PPM 5.3.1 j Comments: Question (737)(Emer & Abn - Gp II) LO: 8457 (PPM 5.3.1-10) i d L Page: 101
. - . _ - - . ~ _ - - . . _ _ - . - - . - - . . . - . . . - . . - - . - - . - . -
h SENIOR REACTOR OPERATOR , QUESTION: 98 (1.00) ] Which one of the following lists actions that can be used to mitigate off site doses for an accident which j releases radioactivity inside secondary containment?
- a. Isolate primary systems leaking into the area Shutdown Reactor Building Ventilation (RBHVAC) i Isolate the Standby Gas Treatment (SGT) system
[ b. Isolate the Standby Gas Treatment (SGT) system Shut down the reactor ! Emergency depressurize the reactor
- c. . Isolate primary systems leaking into the area Shutdown Reactor Building Ventilation (RBHVAC)
! - Shut down the reactor . d. Isolate primary systems leaking into the area l Shut down the reactor l i Emergency depressurize the reactor i j 1 ANSWER: d. ' 1 KA: 295033K203 RO/SRO: 3.9 1
Reference:
PPM 5.3.1 and 5.0.10
- Comments
- New Question (Emer & Abn - Op II)
LO: 8460 (PPM 5.1.3-18) l i I , 4 t a h Page: 102
SENIOR REACTOR OPERATOR , QUESTION: 99 (1.00) A transport cask filled with Control Rod Drive (CRD)" spud end" filters has tipped over on the 501' elevation of the Reactor Building (RB). ARM-RIS-33 (RB 501' area radiation monitor)is alarming on control room panel H13-P614. Reactor building exhaust plenum radiation levels are at = 15 mr/hr and up fast. Which one (1) of the following is an " expected" response to the above conditions? l
- a. CW-P-1B & CW-P-1C ("B" & "C" circulating water pumps) trip.
l
- b. Any traversing in-core probe (TIP) inserted into the core will automatically withdraw and isolate.
- c. Drywell Equipment Drain (EDR) and Floor Drain (FDR) sumps isolate.
I
- d. Containment Nitrogen (CN) makeup isolates.
ANSWER. d. O- 295034K2.02 RO/SRO: 3.9
Reference:
PPM 4.12.4.6, page 2 Comments: New Question (Emer & Abn - Op II) LO: 5597 1 l l I Page: 103
SENIOR REACTOR OPERATOR ! QUESTION: 100 (1.00) A controller failure causes reactor building pressure to increase to GT +4.0" Wg. Which one (1) of the following describes the expected AUTOMATIC action (s) for this condition?
- a. Both the supply and exhaust fans will trip. The standby fans will NOT start.
~
- b. The supply and exhaust fans will be unaffected. The supply plenum relief damper will open to reduce pressure.
- c. Only the operating exhaust fan will trip. The standby exhaust fan will NOT start. Supply fans will be unafTected. ,
- d. Only the operating supply fan will trip'. The standby supply fan will NOT start. Exhaust fans will be unaffected.
ANSWER: a. i KA: 295035A1.02 RO/SRO: 3.8
Reference:
PPM 4.10.1.1 Comments: Modified Question (5534)(Emer & Abn - Gp II) LO: 6878 l l Page: 104
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- -- :-m- +
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I U.S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC WRITTEN EXAMINATION APPLICANT INFORMATION Name: Region: IV Date: October 71996 Facility / Unit: WNP-2 License Level RO Reactor Type: GE Start Time: Finish Time: l INSTRUCTIONS Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses afler the question. The passing grade requires a final grade of at least 80 percent. Examination papers will be picked up 4 hours after the examination starts. All work done on this examination is my owrt I have neither given nor received aid. Applicant's Signature RESULTS Examination Value Points Applicant's Score Points Applicant's Grade Points
REACFOR OPERATOR ANSWERSHEET Multiple Choice (Circle or X your choice) NAME: If you decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the change. l I
- 1. abcd 21. abcd 41. abcd ;
1
- 2. abcd 22. abcd 42. abcd
- 3. abcd 23. abcd 43. abcd
- 4. abcd 24. abcd 44. abcd
- 5. abcd 25. abcd 45. abcd
- 6. abcd 26. abcd 46. abcd l
- 7. abcd 27. abcd 47. abcd
- 8. abcd 28. abcd 48. abcd !
- 9. abcd 29. abcd 49. abcd I l
- 10. abcd 30. abcd 50. abcd 11 abcd 31. abcd 51. abcd 1
- 12. abcd 32. abcd 52. abcd
- 13. abcd 33. abcd 53. abcd
- 14. abcd 34. abcd 54. abcd
- 15. abcd 35. abcd 55. abcd
- 16. abcd 36. abcd 56, abcd
- 17. abcd 37. abcd 57. abcd
- 18. abcd 38. abcd 58. abcd 19 abcd 39. abcd 59. abcd
- 20. abcd 40. abcd 60. abcd Page: 2
REACTOR OPERATOR ANSWERSHEET Multiple Choice (Circle or X your choice) NAME: If you decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the change.
- 61. abcd 81. abcd
- 62. abcd 82. abcd
- 63. abcd 83. abcd
- 64. abcd 84. abcd
- 65. abcd 85. abcd
- 66. abcd 86. abcd
- 67. abcd 87. abcd
- 68. abcd 88. abcd
- 69. abcd 89. abcd
- 70. abcd 90. abcd 71 abcd 91. abcd I
- 72. abcd 92. abcd
- 73. abcd 93. abcd
- 74. abcd 94. abcd j
- 75. abcd 95. abcd
- 76. abcd 96. abcd
- 77. abcd 97. abcd ;
- 78. abcd 98. abcd
- 79. abcd 99. abcd
- 80. abcd 100. abcd Page: 3
. ~ .
l REACTOR OPERATOR l NRC POLICIES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on any part of the examination will result in a denial of your application.
- 2. If you have any questions concerning the administration of the examination, do not hesitate asking them before starting that part of the test.
- 3. SRO applicants will be tested at the level of the responsibility of the senior licensed shift position (i.e. Shift Manager).
- 4. You must pass every part of the examination to receive a license. Applicants for an SRO-upgrade license may require remedial training in order to continue their RO duties if the examination reveals deficiencies in the required knowledge and abilitics.
- 5. The NRC examiner is not allowed to reveal the results of any part of the examination until they have been reviewed and approved by NRC management. Grades provided by the facilitee licensec are preliminary until approved by the NRC.
You will be informed of the official examination results about 30 days after all the examinations are complete.
- 6. After you complete the examination sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.
- 7. To pass the examination. you must achieve a grade of 80 percent or greater. Every question is worth one point.
- 8. The time limit for completing the examination is four hours.
- 9. You may bring pens and calculators into the examination room. Use only black ink to ensure legible copies.
10 Print your name in the blank provided on the answer sheet provided and do not leave any question blank. Use only the paper provided and do not u rite on the back side of the pages. If you decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the change. I1. If the intent of a question is unclear, ask questions of the NRC cxaminer or the designated facility instructor only.
- 12. Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to climinate even the appearance or possibility of cheating
- 13. When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you hase neither given nor received assistance in completing the examination. The scrap paper will be disposed of immediately after the examination.
- 14. After you have turned in your examination, leave the examination area as defined by the proctor or NRC cxaminct. If you are found in this arca w hile the examination is still in progress. >our license may be denied.
i l Page: 4
REACTOR OPERATOR QUESTION: 1 (1.00) Which one (1) of the following statements would describe " Simultaneous Verification" as it applies to a danger tag clearance order?
- a. Two qualified individuals, independently and separately checking the required status of the component or device.
- b. A Control Room Operator (CRO) verifying the required status of the component or device using Control Room indications.
- c. A second qualified individual, via local panel indications showing the required status of the component or device.
- d. Two qualified operators, accompanying each other, check required status, correct identification and location prior to changing component status.
ANSWER: d. KA: 294001K1.02 RO/SRO: 3.9
Reference:
PPM 1.3.8 Comments: Modified question (58) (Plant Wide Generic) LO: 6231 l Page: 5
REACTOR OPERATOR QUESTION: 2 (1.00) l
)
A Control Room Operator (CRO)is getting ready to leave the control room to go to the Production Center l when an equipment operator calls and requests that someone bring him a "High Radiation" key. { Select the one (1) statement below that correctly describes the location from which keys to locked "Iligh i Radiation" areas are normally (non emergency conditions) issued. i
- a. The control room. j
- b. Secondary alarm station. I
- c. Work control center. l
- d. Health physics access control. i l
l i ANSWER: d. KA: 294001K1.03 RO/SRO: 3.3
Reference:
PPM 11.2.7.3 Comments: Modified question (806) (Plant Wide Generic) LO: 6390 ; t i Page: 6
REACTOR OPERATOR QUESTION: 3 (1.00) A task must be performed at a location with a general area radiation level of 60 mr/hr. Previous performance of the task indicates that: One (1) worker can perform the task in I hr and 20 min. Two (2) workers can complete the task in 50 min. Three (3) workers can complete the task in 30 n'in. Four (4) workers can complete the task in 25 min. Based on the above information, how many workers should be assigned to perfoim this task?
- a. One (1) worker
- b. Two (2) workers
- c. Three (3) workers
- d. Four (4) workers ANSWER: a.
KA: 294001K1.04 RO/SRO: 33
Reference:
PPM l.11.2 Comments: New question (Plant Wide Generic) LO. None , I l Page: 7
3 REACTOR OPERATOR (. QUESTION: 4 (1.00) What is the maximum number of visitors that may accompany one (1) escort into_ the Main Control Room?
- a. Three (3) l b. Five (5)
! ) L c. Ten (10)
- d. Fineen (15)
'l 1
ANSWER- b. 1 KA: 29400lKl.05 l RO/SRO: ' 3.2 )
Reference:
GET 82-RDT-0300-HO, page 23 ! Comments: .New question (Plant Wide Generic) LO: 6097 l
)
l l f Page: 8
i i REACTOR OPERATOR l
. QUESTION: 5 (1.00) !
l With the plant operating at rated power which one (1) of the following will require double valve isolation on a : clearance order? i
- a. COND-FCV-15B ("B" condensate booster pump minimum flow valve) l i
l b. CRD-FU-10A ("A" control rod drive pump suction filter) j - l c. FPC-HX-1B ("B" fuel pool cooling heat exchanger)
- d. SCW-P-1 A ("A" stator cooling water pump) 4 ANSWER: a.
- KA
- 29400lKl.09 i RO/SRO: 3.4 i
Reference:
PPM 1.3.8, page 24
- Comments: Modified Question (Plant Wide Generic) i l LO: 6255 e
Page: 9
REACTOR OPERATOR QUESTION: 6 (1.00) During the middle of the last shiR, maintenance completed their work in a main condenser waterbox. You l have been requested to enter the waterbox and inspect it prior to closure. Which one (1) of the following describes the requirements, if any, that must be met prior to your entry? l l
- a. No restrictions apply to this situation. .
I
- b. Perform atmospheric testing.
- c. Complete the pre-entry checklist. ;
l
- d. The Designated Safety Representative (DSR) must be present. '
ANSWER: b. KA: 29400lKl.14 RO/SRO: 3.2
Reference:
PPhi 1.9.2 Comments: New Question (Plant Wide Generic) LO: None i I J l l Page: 10
~
i i REACTOR OPERATOR l QUESTION: 7 (1.00) j Due to an injury sustained on shifl, the Fire Brigade Leader has to leave work. l l Which one (1) of the following individuals can take the place of the Fire Brigade Leader?
- a. Qualified security officer f
- b. Shift support supervisor f
- c. Qualified equipment operator
- d. Qualified health physics technician i
ANSWER: c. f KA: 29400lKl.16 RO/SRO: 3.5 i
Reference:
PPM 1.3.1, pg 40 of 86 l Comments: New Question (Plant Wide Generic) l LO: None l l l l Page: 11
. REACTOR OPERATOR l
l QUESTION: 8 (1.00) When may a task be performed without an approved procedure present? l
- a. When the task has no safety significance.
- b. When the task procedure number on the cover sheet is proceeded by an asterisk (*).
L l c. When the task has been previously performed during the shift by the individual and the ! required steps have been memorized. I
- d. When the task consiste of simple routine actions frequently performed that don't require step sign-offs, recorded data, or specific sequence.
ANSWER: d. KA: 294001 A1.02 RO/SRO: 4.2
Reference:
SWP-PRO-O I Comments: Modified Question (5287) (Plant Wide Generic) LO: 6058 l l I Page: 12
REACTOR OPERATOR l QUESTION: 9 (1.00) The plant is operating in Operational Condition 1. Which one (1) of the following lists the minimum shifl crew composition administrative limit as specified in l plant procedures? '
- a. SM, CRS, SSS, three (3) CROs, five (5) EO, ENS communicator, two (2) HP Techs, two (2) i Chem Techs, two (2) I & C Techs, Duty Omcer.
- b. SM, CRS, two (2) CROs, two (2) EOs, STA, HP Tech, Chem Tech. 1 l
- c. SM, CRS, SSS, three (3) CROs, four (4) EOs, STA, five (5) Fire Brigade members, three (3) l HP Techs, Chem Tech.
)
i
- d. SM, CRS, SSS, two (2) CROs, two (2) EOs, STA, ENS communicator, five (5) Fire Brigade !
members, three (3) HP Techs, Chem Tech, Elect /I&C Tech, Mechanic, Duty 05cer, j ANSWER: c. l KA: 294001 A1.03 ! RO/SRO: 2.7
Reference:
PPM 1.3.1, page 37 1 Comments: New Question (Plant Wide Generic) I LO: 6072 (PPM 1.3.1) ) 1 l l l l l l 1 i i Page: 13
REACTOR OPERATOR l QUESTION: 10 (1.00) l On a case-by-case basis, line supervisors / managers can approve the use of signals for communications, if:
- 1) the signals are not easily confused and are understood by all involved,
- 2) the concept of three-way communication is applied to the maximum extent possible,
- 3) a thorough pre-job briefis conducted AND
- 4) signals are ONLY used.,
- a. in the Control Room.
- b. in severe environments (high noise, heat, or radiation)
- c. when physical directions are the " key" elements of the task.
- d. when the message originator and message recipient are not readily identifiable to each other by sight and voice.
ANSWER: b. KA: 294001 A1.05 RO/SRO: 3.4
Reference:
PPM 1.3.60, page 5 of 8 Comments: New Question (Plant Wide Generic) LO; None l l l 1 i l I i l Page: 14
REACTOR OPERATOR 4 QUESTION: 11 (1.00) l Who has the responsibility for initiating Emergency Core Cooling Systems (ECCS), if required, during a plant l transient? { l
- a. Only the operator responsible for panel Ill3-P601. :
i
- b. The on-shift crew member in closest proximity to panel H13-P601. l I
- c. Only the operator designated by the Control Room Supervisor (CRS) to respond. !
I
- d. Any licensed operator at the control console. !
l 1 i ANSWER: d. KA: 294001 Al.09 RO/SRO: 3.3 :
Reference:
PPM 1.3.1 l Comments: Question (156)(Plant Wide Generic) LO: 6076 Page: 15
. _ . _ . _ . _.-_.m.____ _ _ _ . _ .._ ._ _ _ . _ _ _ . . . _ . . . . _ _ _ . . _ . . . . _
i REACTOR OPERATOR [ QUESTION: 12 (1.00) l During shift turnover and control panel walkdown, annunciator P602. A13-3.5, ASD UPS TROUBLE, is noted to be in the alarmed state. Which one (1) of the following describes the oncoming Control Room Oerators (CROs) responsibility for panel annunciators? !
- a. Verify that this annunciator correctly reflects plant conditions. ;
i
- b. Ensure that this annunciator is logged in the control room log, i l c. Place a " Problem Sticker" adjacent to this annunciator window. l l d. Remove the logic card for this annunciator. l l l ANSWER: a.
KA: 294001 Al.13 , RO/SRO: 4.5
Reference:
PPM 1.3.1, Section 4.20.5 : Comments: Modified Question # 379 (Plant Wide Generic) LO: 6087 ! l I Page: 16
1 REACTOR OPERATOR i
! l
{ QUESTION: 13 (1.00) ! i [ An Anticipated Transient Withour Scram (ATWS)is in progress. Plant conditions are as follows: - i j - RPV level -60" and down slow
- - RPV pressure 945 psig i - Reactor power 10% and down slow i
4 l i The Graphic Display System (GDS) on panel H13-P602 is selected to the Group Isolation screen. Nuclear l Steam Supply Shutoff System (NS4) Group 1 indicates red. Which one (1) of the following explains the reason GDS is providing this information? O j a.- An NS4 Group 1 isolation has occurred
- b. An NS4 logic failure has occurred.
1
- c. A GDS malfunction has occurred.
i i d. PPM 5.5.6 has been performed. ] ANSWER: d. 1 j KA: 294001 Al.15 1 RO/SRO: 3.2 4
Reference:
GDS display
- Comments: New Question (Plant Wide Generic) ,
LO: None ! l 1 1 1 l l . I l 1 Page: 17
REACTOR OPERATOR . QUESTION: 14 (1.00) Following a reactor scram, the Control Room Operator (CRO) notes that CRD-FIC-600 (CRD system flow controller) output signal is going down Which one (1) of the following could cause this condition?
- a. High charging header flow,
- b. High cooling header demand.
1
- c. Low drive header flow.
- d. Low scram header flow.
ANSWER: a. KA: 201001 A2.04 RO/SRO: 3.8
Reference:
CRDH System Text Comments: New Question (Plant Sys - Op II) LO: 5185b I i l Page: 18
REACTOR OPERATOR QUESTION: 15 (1.00) Select the power supply and logic configuration for the Alternate Rod Insenion (ARI) solenoid valves.
- a. 125 VDC (DP-SI-l A/2A)- must be ENERGIZED to vent the scram air header.
- b. 120 VAC (RPS A)- must be DE-ENERGIZED to vent the scram air header.
- c. 125 VDC (DP-SI-ID/2D)- must be DE-ENERGIZED to vent the scram air header.
- d. 120 VAC (IN-2)- must be ENERGlZED to vent the scram air header.
ANSWER: a. KA: 20100lK2.05 RO/SRO: 4.5
Reference:
DC Power Systems Text Comments: Modified question (175) (Plant Sys - Op II) LO: 5262 i l Page: 19
. - . -_ - - . _ . . . . - . . . .._.- _ . . ~ - . ~ _ . .- -.- .- -
REACTOR OPERATOR l QUESTION: 16 (1.00) l l The reactor is in Operational Condition 5, the Control Room Supervisor (CRS) has directed the Control i Room Operator (CRO) to select control rod 30-55 to verify its position. The CRO reports that control rod i 30-55 cannot be selected. Which one (1) of the following could cause this condition? 1
- a. The refuel bridge is near or over the core. l l
- b. - The fuel grapple is not full up. l I
- c. Another control rod is withdrawn past "00" l l
- d. The fuel grapple is loaded.
ANSWER: c. , I i KA: 201002K3.01 RO/SRO: 3.4
Reference:
Fuel Handling Text Comments: New question (Plant Sys - Gp 1) > LO: 5359 & 5360 i l t f l 4 Page: 20
REACTOR OPERATOR QUESTION: 17 (1.00) After completing the Immediate Actions for a reactor scram, the Control Room Operator (CRO) notices a WITHDRAW rod block has been applied Which one (1) of the following is true concerning this condition? This rod block..
- a. will automatically be bypassed when the scram is reset.
- b. can be manually bypassed by bypassing the RWM.
- c. will automatically be bypassed 10 seconds alter placing the reactor mode switch in SHUTDOWN.
- d. CANNOT be bypassed.
ANSWER: d. KA: 201002K4 02 RO/SRO: 3.5
Reference:
RPS System Text Comments: New Question (Plant Sys - Gp II) LO: 5952 l i l 1 i Page: 21
= -.
T l REACTOR OPERATOR 1 QUESTION: 18 (1.00) The plant is operating at rated conditions with RRC-IN-ASD/l A (1 A ASD UPS inverter) in service and supplying panel E-PP-ASD1/4. Pushing the red EMER POWER OFF pushbutton on the front of the inverter panel will result in which one (1) of the following?
- a. RRC-P-1 A ("A" reactor recirculation pump) will immediately trip, RRC-P-1B ("B" reactor recirculation pump) will continue to operate.
- b. Both reactor recirculation pumps will immediately trip.
- c. Both reactor recirculation pumps will trip after approximately 20 minutes.
- d. Both reactor recirculation pumps will continue to operate indefinitely.
ANSWER: b. KA: 202002A2.07 RO/SRO: 3.3
Reference:
PPM 2.7.4 A l Comments: Modified question (6206) (Plant Sys - Op I) LO: 9683 1 l Page: 22
f l REACTOR OPERATOR l QUESTION: 19 (1.00) , l The plant is operating in Mode 1. Both Reactor Recirculation pumps are operating at 45 Hz. Adjustable l Speed Drive (ASD) Channel A2 is not running and its white READY light is illuminated. ; I Which one (1) of the following describes the response of the Recirculation Pumps if the START Pushbutton ! for RRC-P-1 A is pressed? l l l J
- a. The A2 ASD Channel will start. RRC-P-1 A frequency will ramp to 52.2 Hz. ,
i
- b. The A2 ASD Channel will start. RRC-P-1 A will continue to operate at 45 Hz.
I
- c. RRC-P-1 A frequency will ramp to 15 Hz, the A2 ASD Channel will start, and RRC-P-1 A frequency will ramp back to 45 Hz.
- d. RRC-P-1 A frequency will ramp to 52.2 Hz and the A2 ASD Channel will start. j ANSWER: b. l KA: 202002Kl.02 RO/SRO: 4.2
Reference:
PPM 2.2.1 Comments: Modified Question (3217) (Plant Sys - Gp I) l LO: None , I t
]
l l l l Page: 23
- ~ , _ - . . __%, _ , + ' - - - - - - - - - ,- -r'-
_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ ._ ._ _ .m _ _. _. REACTOR OPERATOR QUESTION: 20 (1.00) Given the following indications:
- Reactor scrammed
- Reactor water level at -95" and down slow
- Drywell pressure at 1,67 psig and up fast.
Which one (1) of the following statements describes the response of the Low Pressure Coolant Injection (LPCI) mode of the Residual Heat Removal (RHR) system to the indications listed above..
- a. LPCI will initiate when reactor water level decreases to LE -129"
- b. LPCI initiated when reactor water level reached -50"
- c. LPCI will initiate when drywell pressure increases to GE 1.68 psig.
- d. LPCI initiated when drywell pressure reached 1.65 psig.
ANSWER: d. KA: 203000Kl.13 RO/SRO: 3.9 Reference. RHR System Text Comments: New Question (Plant Sys - Gp I) LO: 5775 I Page: 24
l REACTOR OPERATOR [ QUESTION: 21 (1.00) The High Pressure Core Spray (HPCS) system is operating in the Condensate Storage Tank to Condensate
- Storage Tank (CST to CST) full flow test lineup, when a loss of primary containment cooling results in drywell pressure increasing to 1.8 psig.
Which one (1) of the following statements identifies the automatic response of the HPCS system under the above conditions? f a. HPCS-V-4 (injection valve) opens immediately , HPCS-V-10 & HPCS-V-Il (full flow test valves) close immediately HPCS-V-12 (minimum flow valve) remains closed { b. HPCS-V-10 & HPCS-V-11 (full flow test valves) go full closed e HPCS-V-4 (injection valve) opens afler HPCS-V-10 atLd HPCS-V-11 fully clase HPCS-V-12 (minimum flow valve) cycles open and then closed
- c. HPCS-V-10 & HPCS-V-11 (full flow test valves) go full closed HPCS-V-4 (injection valve) opens after HPCS-V-10 or HPCS-V-11 fully close j HPCS-V-12 (minimum flow valve) cycles open and then closed !
- d. HPCS-V-4 (injection valve) opens immediately HPCS-V-10 & HPCS-V-11 (full flow test valves) remain open HPCS-V-12 (minimum flow valve) remains closed ANSWER: a.
KA: 209002A3.01 l RO/SRO: 3.3
Reference:
HPCS Systems Text Comments: Modified question (444) (Plant Sys - Op 1) , LO: 5425,5429 l l Page: 25
I REACTOR OPERATOR ! , t
- QUESTION
- 22 (1.00) i
! Following a valid High Pressure Core Spray (HPCS) initiation on high drywell pressure., the HPCS LEVEL 8 j SEALED IN light and alarm are received and HPCS-V-4 (RPV injection valve) closes. I Which one (1) of the following conditions will cause HPCS-V-4 to automatically re-open? l
- a. Drywell high pressure logic reset. i
- b. RPV level lowering to +12"
~
- c. RPV level lowering to -51" :
! d. Drywell high pressure alarm clears. i ANSWER: c. J KA: 209002K4.02 RO/SRO: 3.4
Reference:
HPCS System Text
- Comments
- Modified question (443) (Plant Sys - Gp I)
LO: 5429 i 1 i i 4 Page: 26
- REACTOR OPERATOR
( QUESTION: 23 (1.00) An Anticipated Transient Without Scram (ATWS) is in progress concurrent with a loss of MC-8B, j suppression pool temperature is 118 F and up slow. Assume all required actions have been completed correctly and no other failures have occurred at this time. Which one (1) of the following describes the Standby Liquid Control (SLC) system status?
- a. SLC-P-1 A - running SLC-P-1 B - running (SLCpumps) l SLC-V-1 A - open SLC-V-1B - open (SLC storage tank outlets) l SLC-V-4 A - actuated SLC-V-4B - actuated. (Squib valve 3)
- b. SLC-P-1 A - loss of power SLC-P-1 B - running (SLCpumps)
SLC-V-1 A - loss of power SLC-V-1B - open (SLC storage tank outlets) SLC-V-4A - loss of power SLC-V-4B - actuated (Squib ralre3)
- c. SLC-P-1 A - oft SLC-P-1 B - off (SLCpump3)
SLC-V-1 A - closed SLC-V-1B - closed (SLC storage tank outlet 3) SLC-V.4A - closed SLC-V-4B - closed (Squib valves)
- d. SLC-P-1 A - nmning SLC-P-1B - loss of power (SLCpumps)
SLC-V-1 A - open SLC-V-1B - loss of power (SLC storage tank outlets) SLC-V-4 A - actuated SLC-V-4B - loss of power (Squib valves) ANSWER: d. KA: 211000K6.03 RO/SRO: 3.2
Reference:
SLC Systems Comments: New question (Plant Sys - Gp I) ! LO: 5931h l 4 Page: 27
l REACTOR OPERATOR j i QUESTION: 24 (1.00) t ! i
- With the plant in a hydraulic Anticipated Transient Without Scram (ATWS) condition, the Control Room I i . Operator (CRO) carries out the actions of PPM 5.5.11 and resets the scram. Annunciator P603-A8 6-4, l i SCRAM VALVE PILOT AIR HDR PRESS LOW, fails to clear.
4 , ] Which one (1) of the following could cause this condition? ! 4 4 a. At least one (1) backup scram valve has failed to ENERGlZE following the scram reset. i j b. Alternate Rod Insertion (ARI) logic has not been reset. ; t l c. . Both backup scram valves have failed to ENERGlZE following the scram reset. '
- d. One (1) of the Reactor Protection System (RPS) trip signals has not been bypassed.
ANSWER; b. i KA: 212000A1.11 RO/SRO: 3.4
Reference:
PPM 5.0.10 Comments: Modified Question (4085)(Plant Sys - Op I) ! LO: 8094 (5.1.2-07) l J l I Page: 28
REACTOR OPERATOR QUESTION: 25 (1.00) The "A" Reactor Protection System (RPS) Motor Generator (MG) set has tripped. The plant is presently operating with a half scram at 100% power. The status of the white indicating lights associated with the MG SET TRANSFER SWITCH on H13-P610 is as follows:
- GENERATOR A FEED deenergized
- ALTERNATE FEED energized
- GENERATOR B FEED energized i
The MG Set Transfer switch is in NORMAL. t Based on these conditions, which one (1) of the following statements is true? , Placing the MG Set Transfer switch to...
- a. ALT B will allow resetting the Electric Power Monitoring Assembly (EPA) breakers for the "A" RPS MG
- b. ALT A will cause the GENERATOR A FEED light to energize.
- c. ALT B will cause a full scram.
- d. ALT A will allow resetting the Electric Power Monitoring Assembly (EPA) breakers for the ,
"A" RPS MG >
ANSWER: c. KA: 212000K2.01 RO/SRO: 3.2
Reference:
RPS System Text Comments: Similar question (Plant Sys - Op I) , LO: 5961 Page: 29
i REACTOR OPERATOR ! QUESTION: 26 (1.00) { The plant is operating at 100% power with AR-EX-1 A ("A" gland steam condenser exhauster) in service ] when a trip of RPS-MG-2 ("B" reactor protection system motor generator) occurs. Upon checking the full j core display, it is noted that control rod 30-07 has scrammed and is fully inserted. Assuming NO OPERATOR ACTIONS, afler a time interval, the reactor will fully scram due to which one (1) of the following?
- a. Main turbine trip (throttle valve closure / governor valve fast closure).
- b. Drywell high pressure.
- c. Main steam isolation valve (MSIV) closure.
- d. Scram discharge volume high level.
ANSWER: d. KA: 212000K3.06 RO/SRO: 4.0
Reference:
CRDH System Text, PPM 4.7.6.1 Comments: New Question (Plant Sys - Gp 1) LO: None Page: 30
1 i REACTOR OPERATOR l ! QUESTION: 27 (1.00) I l During a reactor startup, the reactor is suberitical with control rod withdrawal in progress. Source Range l Monitor (SRM) count rate has stabilized at 1 x 10' Counts Per Second (CPS) following the last control rod
- withdrawal.
During withdrawal of the next control rod in the sequence, the first control rod in the next Rod Worth l Minimizer (RWM) group, reactor period meters deflect from infinity to =20 seconds before turning. Reactor l period is now =60 seconds increasing (approaching infinity). ! Which one (1) of the following actions should be taken for this condition?
- a. Verify that the withdrawn control rod did not " double notch" and stop control rod withdrawal l to allow stabilization of neutron level.
l b. Monitor SRMs and retract SRMs as necessary to maintain count rate LT 1 x 10' CPS.
- c. Insert control rods until the reactor is subcritical and notify the Control Room Supervisor (CRS)/ Shift Manager (SM) and the Station Nuclear Engineer (SNE).
- d. Immediately manually scram the reactor.
ANS%TR: a. l KA: 215004 A4.01 RO/SRO: 3.9
Reference:
PPM 3.1.2, Step 4.2.2 (2nd CAUTION), page 15; Step 4.2.5 (CAUTION), page 16; Step 4.2.9.d (CAUTION), page 18. i Comments: Modified Question (815) (Plant Sys - Op I) l LO: None Page: 31
l REACTOR OPERATOR QUESTION: 28 (1.00) Given the following:
- RPV water level -10" and steady
- Reactor pressure 200 psig and down slow
- Drywell temperature 350 F and down very slow
- No Secondary Containment Control entry conditions exist.
- Emergency depressurization is planned.
Which one (1) of the following is correct concerning the instmment(s) which can be used to determine RPV water level for the given conditions? I
- a. Wide range only.
- b. Narrow range only.
- c. None.
- d. Fuel zone range only.
ANSWER: a. KA: 216000K5.07 RO/SRO: 3.6
Reference:
NBI Systems Text Comments: New question (Plant Sys - Gp I) LO: 5582 l i I
-i l
l l
)
Page: 32
REACTOR OPERATOR QUESTION: 29 (1.00) Reactor Core Isolation Cooling (RCIC) initiated as expected on a valid low level signal raising RPV level to l the Level 8 setpoint. I Which one (1) of the following describes the automatic restan capability of RCIC?, RCIC will..
- a. automatically restan when RPV level drops below the Level 8 setpoint.
- b. NOT automatically restan unless a high drywell pressure signal is received.
- c. automatically restart when RPV level drops below the Level 2 setpoint.
- d. automatically restart when RPV level drops below the Level 3 setpoint.
ANSWER: c. KA: 217000A1.03 RO/SRO: 4.0
Reference:
RCIC Systems Text Comments: New question (Plant Sys - Gp !) LO: 5071 and 8735 i i Page: 33
_ . _ . . _ . _ . _ _ . _ _ _ _ . - . _ _ _ _ - _ _ _ _ _ . _ _ _ . ___.__.._._._.-___m REACTOR OPERATOR QUESTION: 30 (1.00) A transient has resulted in RPV level dropping to -140" The level has remained stable for GT seven (7) minutes. SM-8 has deenergized and Division 1 Automatic Depressurization System (ADS) has been inhibited. After verifying the cause of the loss of SM-8, permission is granted to reenergize the bus. I Assuming no operator actions, which one (1) of the following describes the response of the ADS Logic to the reenergization of SM-8. f l
- a. ADS will initiate 105 seconds afler the discharge pressure of"C" Residual Heat Removal (RHR) pump reaches the ADS permissive.
l
- b. ADS will not initiate if the operator resets the division "2" ADS timer within 105 seconds.
- c. ADS will initiate immediately when the discharge pressure of"C" Residual Heat Removal (RHR) pump reaches the ADS permissive.
l ) d. ADS will not initiate until the operator resets the division "1" ADS inhibit switch. l l l ANSWER: c. l KA: 218000K5.01 l RO/SRO: 3.8
Reference:
. ADS Systems Text L Comments: New question (Plant Sys - Gp I) l . LO: None l l I I l l l l l l f Page: 34
REACTOR OPERATOR QUESTION: 31 (1.00) What effect does manually decreasing the output on CAC-FC-67A ("A" recycle flow controller) have on the Containment Atmosphere Control (CAC) system?
- a. Reduces flow through the sembber and reduces oxygen concentration entering the recombiner.
- b. Increases flow through the scrubber and increases oxygen concentration entering the recombiner.
- c. Reduces flow through the scrubber and increases oxygen concentration entering the recombiner.
- d. Increases flow through the scrubber and reduces oxygen concentration entering the recombiner.
ANSWER: b. KA: 223001 A4.13 RO/SRO: 3.4
Reference:
CAC System Text Comments: Question (4357)(Plant Sys - Gp I) LO: 5133a Page: 35
.._._ . _ . . _ . _ _ ~ . . _ . . . _ . _ . _ _ _ . __. . . . . _._ _ ._. _ __ _ . _ _ _ _ _ _ .. _.___._.
4 REACTOR OPERATOR i QUESTION: 32 (1.00) t i' Which one (1) of the following is designed to prevent the differential pressure across the primary containment boundary from exceeding the design limit?
- a. Reactor building to wetwell vacuum breakers.
- b. Wetwell to drywell vacuum breakers.
I c. Standby Gas Treatment (SGT) system. i i
- d. Suppression pool "T" quenchers. ,
t i i l ANSWER: a. l t KA: 223001K4.06 L 4 RO/SRO: 3.1 i
Reference:
PPM 5.0.10 Comments: New question (Plant Sys - Op I) ; LO: 8352 (5.1.2-140) 1 i ! 6 e s s f a
- 1 T
f . k 4 Page: 36
l , REACTOR OPERATOR QUESTION: 33 (1.00) The plant is in OPERATIONAL CONDITION 4, when fluctuations are observed in the output voltage ofIN-2. Assuming NO OPERATOR ACTIONS which one (1) of the following describes the expected result of continued IN-2 voltage degradation? I l
- a. RCIC-V-8, RCIC-V-63, & RCIC-V-76 (reactor core isolation cooling steam line isolation valves) close.
- b. Rod Worth Minimizer, (RWM), Rod Sequence Control System (RSCS) and Reactor Manual Control System (RMCS) become INOPERABLE.
l
- c. Nuclear Steam Supply Shutoff System (NS4) inboard isolations and loss of fire control panels 1, 2, & 3.
1
- d. Loss of the full core display and the ROLM telephones.
l ! ANSWER: c. l KA: 223002K6.07 l RO/SRO: 3.2
Reference:
LER 84-118 Comments: New Question (Plant Sys - Gp I) LO: None l l l I
'I Page: 37 l
REACTOR OPERATOR QUESTION: 34 (1.00) The plant was operating at 100% power when a steam leak developed in the drywell. Current plant conditions are as follows: RPV water level +12" RPV pressure 950 psig Drywell pressure 2.5 psig Drywell temperature 135'F Which one (1) of the following lists of Nuclear Steam Supply Shutoff System (NS4) groups SHOULD have received isolation / initiation signals?
- a. Group 1 Group 4 Group 5
- b. Group 2 Group 3 Group 6
- c. Group 3 Group 4 Group 5
- d. Group 1 Group 2 Group 7 ANSWER: c.
KA: 223002A2.10 SRO: 3.9
Reference:
NS4 System Text Comments: Modified Question (299) (Plant Sys - Gp I) LO: 5597 Page: 38
REACTOR OPERATOR QUESTION: 35 (1.00) Loss of DP-SI-2 power will render Safety Relief Valve (SRV) control switches INOPERABLE at the location (s) specified in which one (1) of the following?
- a. H13-P601 only
- b. H13-P601 and H13-P631 (ADS division 2 logic panel) j
- c. H13-P628 (ADS division I logic panel) and E-CP-ARS (Altemate remote shutdown panel)
- d. H13-P631 (ADS division 2 logic panel) and C61-P001 (Remote shutdown panel)
ANSWER: d. KA: 239002K4.05 SRO: 3.6
Reference:
PPM 4.12.1.1 and RSD System Text Comments: Modified Question (461) (Plant Sys - Op I) LO: 5885a l 1 l l 1 1 Page: 39
REACTOR OPERATOR QUESTION: 36 (1.00) The plant is operating at 100% power when the Control Room Operator (CRO) reports that RPV pressure is trending down. Shortly after this report the reactor scrams and the Main Steam Isolation Valves (MSIVs) close. Which one (1) of the following desenbes the cause of this transient?
- a. The selected Digital Electrohydraulic (DEH) pressure controller has slowly failed high.
- b. The backup Digital Electrohydraulic (DEH) pressure controller has instantly failed low.
- c. The selected Digital Electrohydraulic (DEH) pressure controller has instantly failed low.
- d. The backup Digital Electrohydraulic (DEH) pressure controller has slowly failed high.
ANSWER: a. KA: 241000K3.02 RO/SRO: 4.2
Reference:
DEH System Text Comments: Modified Question (6232) (Plant Sys - Op I) LO: 5286b l i l l Page: 40 l
REACTOR OPERATOR QUESTION: 37 (1.00) With a plant startup in progress and reactor power at 20%, #3 Turbine Bypass Valve (BPV) is declared INOP. Which one (1) of the statements below describes the action (s) which must be taken under the above conditions? l
- a. Restore the inoperable BPV to OPERABLE status within I hour or reduce power to less than 5% of rated within the next 4 hours.
- b. Continue the stanup but do not exceed 90% of rated power until the BPV has been restored to OPERABLE status.
- c. Restore the inoperable BPV to OPERABLE status prior to reaching 25% of rated power.
- d. Restore the BPV to OPERABLE within 12 hours, or suspend the startup and be in COLD SHUTDOWN within the next 12 hours.
ANSWER: c. KA: 241000A2.03 RO/SRO: 4.1 ,
Reference:
Tech. Spec. 3.7.9 Comments: New question (Plant Sys - Gp I) LO: None l i i Page: 41
_ _ _ . _ . . _ _ _ _ _ _ _ _ _ . _ _ _ . . . _ _ ___ ._. . _ _ _ - _ _ . _ . _ _ . ~ _ _ _ _ _ . _ _ REACTOR OPERATOR l QUESTION: 38 (1.00) ! l The plant is operating at 50% power with both Reactor Feed Pump Turbines (RFPTs) operating in automatic j control when RFPT"A" governor valves become stuck in the present position. ' ! i ! Assuming NO OPERATOR ACTIONS (other than rasing power), how will RFPT "A" speed and RPV level j respond if reactor power is raised to 100%? ; i
- RFPT "A" SPEED RPV Level l l
, a. Decrease Remain the same l 3 ) < j b. Decrease Lower l
- c. Increase Remain the same I
) d. Remain approximately the same Lower ] 3 ANSWER: c. 1 i KA: 259001K3.01 RO/SRO: 3.9 I 3
Reference:
Feedwater/ Condensate System Text l Comments: New Question-(Plant Sys - Op II) 3 i LO: None t ?
- l 1
0 l ! I l i 1 1 4 3
- l
) f f e
)
j Page: 42
)
REACTOR OPERATOR QUESTION: 39 (1.00) l SGT-V-5B-2 (exhaust to stack) fails to open upon receipt of a valid initiation signal. i l Which one (1) of the following describes the response of the Standby Gas Treatment (SGT) train "B" to this condition?
- a. SGT-FN-1B-2 (lead fan) stans and trips on low flow. SGT train "B" remains in this condition -
until the automatic start logic is reset. I i
- b. SGT-FN-1B-2 (lead fan) starts and trips on low flow. Automatic valve re-alignment and start of SGT-FN-1B-1 (lag fan) allow SGT train "B" to perform it's design function.
- c. SGT-FN-1B-2 (lead fan) and SGT-FN-1B-1 (lag fan) do NOT receive an automatic start i signal while SGT-FN-5B-2 (exhaust to stack) is in the closed position. l l
- d. SGT-FN-1B-2 (lead fan) stans and trips on low flow. SGT-FN-1B-1 (lag fan) will then stan, I but will also trip on low flow. l l
ANSWER: b. KA: 2ol000K4,01 RO/SRO: 3.7
Reference:
SGT System Text Comments: Similar Question (3587) (Plant Sys - Gp I) LO: 5828 i Page: 43
d a REACTOR OPERATOR QUESTION: 40 (1.00) i i
; Following full load operation for a routine surveillance, diesel generator #1 is being cooled down at idle 4
speed. During this time a loss of off-site power occurs. i Which one (1) of the following statements describes the actions necessary to ensure proper operation of the f diesel for reenergizing SM-77 -
- a. Place the excitation mode selector switch in PARALLEL. Ensure that SW-P-1 A ("A" service j water pump) continues to run or manually trip the diesel.
4 b. Place the engine speed selector switch in RATED and place the control switch for SW-P-1 A to j STOP to reset the auto start on the loss of off-site power. ) c. Place the excitation mode selector switch in PARALLEL and place the control switch for SW-
- P-1 A to STOP to reset the auto start on the loss of off-site power.
j d. Place the engine speed selector switch in RATED. Ensure that SW-P-1 A starts as soon as it's
- discharge valve cycles full closed to full open.
ANSWER: d. r KA: 264000K6.08 i RO/SRO: 3.6 I i
Reference:
Diesel Generator System Text and PPM 2.7.2A ; i Comments: New Question (Plant Sys - Op I) l LO: 532) l } i i i i ! l 4 4 1 J 4 1 i I 2 j Page: 44
i I j REACTOR OPERATOR l
! l ! QUESTION: 41 (1.00) 4 t l j SM-8 is deenergized, which one (1) of the following is a permissive that MUST be satisfied in order for DG-2 i
- to re-energize SM-87 l l
l .
- a. Relay 86DG2 (Engine lockout) must be reset.
- b. Breaker 3-8 (feed from SM-3) and 8-3 (feed to SM-8) must both be open. I i' c. Engine centrol switch must be in the REMOTE position.
- d. Drywell pressure must be greater than 1.65 psig.
ANSWER: a. t 5 KA: 264000A3.05
- RO/SRO
- 3.4
Reference:
Diesel Generator System Text ] Comments: Similar Question (302) (Plant Sys - Gp I) i LO: 5316 d I k 4 4 ) a i ( i I i i i I i Page: 45 i
REACTOR OPERATOR QUESTION: 42 (1.00) While withdrawing control rods during a plant startup, the control room operator (CRO) reports that a control rod will not move and appears to be stuck. Which one (1) of the following describes an option that could be used to attempt to move this control rod?
- a. Adjust cooling water flow to GT 80 gpm and allow the rod to be forced in.
- b. Use the Single Rod Insert (SRI) switches to scram the rod and then recover it.
- c. Apply continuous withdrawal signals in two minute increments.
- d. Apply a continuous insert signal, release, then apply a continuous withdrawal signal.
l ANSWER: d. KA: 201003 A2.01 RO/SRO: 3.4
Reference:
CRDH System Text Comments: New question (Plant Sys - Gp Ill) LO: 5204 1 J l l I
)
Page: 46
REACTOR OPERATOR QUESTION: 43 (1.00) - The plant is operating at 100% power when an EO, investigating an Accumulator Trouble alarm, reports that an HCU Nitrogen Accumulator has completely de-pressurized. Which one (1) of the following describes the scram capability of the affected control rod?
- a. The rod can be scrammed because CRD Drive Header pressure is greater than Scram
. Discharge Volume pressure.
, b. The rod can be scrammed because RPV pressure is greater than Scram Discharge Volume pressure.
- c. The rod can NOT be scrammed because Nitrogen Accumulator pressure is less than RPV pressure.
l d. The rod can NOT be scrammed because Scram Inlet and Scram Outlet valves have lost their pneumatic supply. I i ANSWER: b. KA: 201003K4.04
- RO/SRO: 3.6
Reference:
CRDM System Comments: Question (3524)(Plant Sys - Op III) LO: 5215 l
)
l I Page: 47
- .. ..- - -.= ._. - . . . . . . . _ . . _ . - .. . .. - . - . . . - _ . - .
REACTOR OPERATOR QUESTION: 44 (1.00) Which one (1) of the following identifies one of the boundaries of the region or trae power to flow map where operation is PROHIBITED by PPM 4.12.4.7 and Technical Specifications?
- a. 55% flow
- b. 35% flow
- c. 25% flow
- d. 45% flow ANSWER: d.
KA: 20200lGK.06 RO/SRO: 3.8
Reference:
Tech. Spec. 3.2.6 ; Comments: New Question (Plant Sys - Op II) i LO: None 1 1 i i l 4 Page: 48
REACTOR OPERATOR QUESTION: 45 (1.00) l With the plant at rated conditions a Group 1 isolation occurred with RPV pressure peaking at i145 psig during the transient. t Which one (1) of the following describes the direct effect on the reactor recirculation pump breakers? 4
- a. No breakers are effected
} b. Only breakers CB-RPT-3 A and CB-RPT-3B trip. I
- c. Breakers CB-RPT-3 A, CB-RPT-3B, CB-RPT-4 A, CB-RPT-4B trip.
1
- d. Only breakers CB-RPT-4A and CB-RPT-4B trip.
l ! ANSWER: b. I I
- KA
- 202001 A2.14 i,
RO/SRO: 3.9
Reference:
RRC Systems Text, PPM 4.602.A6-1.2/l.6 Comments: Modified question (3893) (Plant Sys - Gp II) i LO: 5023e i 4 L 4 f i i i l 1 l s Page: 49
REACTOR OPERATOR QUESTION: 46 (1.00) l Following a loss of SM-7 and an Anticipated Transient Wit' a Scram (ATWS) condition, boron injection is required. What effect will Standby Liquid Control (SLC) initiation have on Reactor Water Cleanup (RWCU) system valves?
- a. RWCU-V-104 (RWCU system bypass) opens.
- b. RWCU-V-1 (RWCU inboard isolation) closes.
- c. RWCU-V-4 (RWCU outboard isolation) closes.
- d. RWCU-V-40 (RPV/RWCU return isolation) closes.
ANSWER: c. KA: 204000K4.04 RO/SRO: 3.6 Reference. RWCU System Text Comments: New Question (Plant Sys - Gp II) LO: 5035 i i l l l l i Page: 50
i f REACTOR OPERATOR QUESTION: 47 (1.00) With the plant at 30% power, which one (1) of the following describes the effect that a loss of rod position information for a single control rod will have on the Reactor Manual Control System (RMCS)?
- a. A rod insert and withdraw block will be generated via the Rod Worth Minimizer (RWM).
b A rod withdraw block will be generated via the Rod Sequence Control System (RSCS).
- c. A rod insert block will be generated via the Rod Position Indication System (RPIS).
- d. No rod blocks are generated, a loss of rod position indication only.
ANSWER: d. KA: 214000K3.03 RO/SRO: 3.2
Reference:
RMCS Systems Text Comments: Modified Question (4286) (Plant Sys - Gp II) LO: 7754a . I i I i Page: 51
.- - . - _-. -_ . - _ = - -. .
REACTOR OPERATOR OUESTION: 48 (1.00) RHR-P-2C ("C" residual heat removal pump) is operating in the test line-up for a suiveillance when RHR-P-3 (residual heat removal B/C water leg pump) trips. Which one (1) of the following describes actions that should be taken for this condition?
- a. RHR-P-2C ("C" residual heat removal pump) should remain in operation, if possible, to ensure that the system piping remains filled and to maintain pressure to Residual Heat Removal (RHR) loop "B"
- b. RHR-P-2C ("C" residual heat removal pump) should remain in operation, if possible, until RHR-P-3 (residual heat removal B/C water leg pump) can be restored to service. RHR-P-2B
("B" residual heat removal pump) should be placed in wetwell cooling prior to the receipt of the DISCH PRESS HIGH/ LOW annunciator,
- c. RHR-P-2C ("C" residual heat removal pump) should be shut down. The control switches for both RHR-P-2C ("C" residual heat removal pump) and RHR-P-2B ("B" residual heat removal
~
pump) should be held in OFF until the control power fuses have been removed for both pump breakers.
- d. RHR-P-2C ("C" residual heat removal pump) should be shut down to facilitate repair of RHR-P-3 (residual heat removal B/C water leg pump). RHR-P-2B ("B" residual heat removal pump) should be placed in wetwell cooling prior to the receipt of the DISCH PRESS HIGH/ LOW annunciator.
ANSWER: b. KA: 219000K1.06 RO/SRO: 3.2
Reference:
RHR System Text, PPM 2.4.2, Section 4.7 and 4.8, page 8 Comments: New Question (Plant Sys - Gp II) LO: 5781f Page: 52
REACTOR OPERATOR QUESTION: 49 (1.00) RHR-V-68B (residual heat removal heat exchanger service water discharge valve) hasjust been stroked closed for a surveillance test. Which one (1) of the following conditions, if any, would result in RHR-V-68B automatically opening 7
- a. None, RHR-V-68B does not automatically reposition.
- b. Reactor water level at -60"
- c. Reactor building exhaust plenum high radiation.
- d. "B" residual heat removal pump room high temperature.
ANSWER: b. KA: 219000K4.09 RO/SRO: 3.3
Reference:
RHR System Text Comments: Modified Question (3820) (Plant Sys - Gp II) LO: 7728c Page: 53
REACTOR OPERATOR QUESTION: 50 (1.00) A Loss of Coolant Accident (LOCA) has occurred, all Emergency Core Cooling Systems (ECCS) equipment has functioned as designed. Present plant conditions are as follows:
- RPV level -135" and up slow
- RPV pressure 200 psig and down slow i
- Wetwell pressure 9 psig and up very slow J I
- RHR-V-42A (RPV injection valve) is open l
l RHR-V-17A (Upper drywell spray inboard isolation valve) is opened in preparation for drywell spray. When l the Control Room Operator (CRO) takes the control switch for RHR-V-16A (upper drywell spray outboard l isolation) to OPEN, RHR-V-16A will.. j
- a. remain closed until RPV pressure drops below 135 psig. ,
1 l
- b. open when RPV water level is GE -129"
]
- c. open immediately.
f
- d. remain closed.
ANSWER: d. ' f KA: 22600lGK.07 . RO/SRO: 3.5 i
Reference:
RHR Systems Text Comments: New Question (Plant Sys - Gp I) , LO: 5781 1 1 Page: 54
REACTOR OPERATOR QUESTION: 51 (1.00) During Loss Of Coolant Accident (LOCA) conditions with Residual Heat Removal (RHR)"A" unavailable, RIiR "B" was placed into both wetwell spray and drywell spray. What is the expected automatic system response when the high drywell pressure initiation signal subsequently clears?
- a. Drywell spray isolates and wetwell spray continues.
- b. BOTli drywell and wetwell sprays isolate.
- c. Drywell spray continues and wetwell spray isolates.
- d. BOTli drywell and wetwell sprays continue.
ANSWER: d. KA: 230000A4.02 RO/SRO: 3.8
Reference:
RIIR System Text Comments: Question (199)(Plant Sys - Gp 11) LO: 5781 i l l l Page: 55
1 l REACTOR OPERATOR QUESTION: 52 (l.00) ! The plant is operating at 100% power when both 500 KV generator output breakers trip. If the main turbine fails to trip, which one (1) of the following describes the short term response of the main turbine Overspeed Protection Controller (OPC) for this condition? i
- a. OPC initially actuates and then resets. Thereafter main turbine speed is controlled at 100% of i rated by the Digital Electrohydaulic (DEH) control system.
- b. OPC initially actuates and does NOT reset. Main turbine speed coasts down to O rpm.
j c. OPC repeatedly actuates and resets to control main turbine speed LT 103% of rated. i
- d. OPC repeatedly actuates and resets to control pressure at Pressure Setpoint.
i ANSWER: c. i KA: 245000K4.09 RO/SRO: 3.2
Reference:
Main Turbine Systems Text Comments: Modified question (228)(Plant Sys - Gp II) LO: 5566 I I 1 Page: 56
REACTOR OPERATOR QUESTION: 53 (1.00) Initial Conditions:
- Reactor startup in progress.
- RPV pressure =450 psig and going up
- Main condenser vacuum =23"
- SM-1, SM-2, and SM-3 being powered from the startup transformer (TR-S)
- COND-P-2A ("A" condensate booster pump) running
- CW-P-lC ("C" circulating water pump) running
- COND-P-1 A and COND-P-1B ("A" & "B" condensate pumps) running Maintenance has requested that Operations stan the CW-P-1 A & CW-P-1B ("B" & "C" circulating water pumps) for post maintenance testing.
Using the above information, determine which one (1) of the following statements is correct.
- a. Starting the third circulating water pump will cause an unde: voltage trip of TR-S.
- b. Operation of more than one (1) circulating water pump at this point in the startup is not recommended due to tube erosion concerns.
- c. Starting two (2) additional circulating water pumps should not cause any significant problems for plant operations.
- d. If CW-P-1 A is staned last, the transient on SM-1 will cause a trip of COND-P-1 A and COND-P-2A on over current.
ANSWER: a. KA: 256000K6.02 RO/SRO: 3.1
Reference:
CW/TMU Systems Text, pg 15 & PPM 2.6.1 Sec. 5.3.13, Page 18 of 93 Comments: New question (Plant Sys - Gp 111) LO: 7765 Page: 57
REACTOR OPERATOR QUESTION: 54 (l.00) Which one (1) of the following describes why cooling is required for a mechanical vacuum pump during it's operation.
- a. In addition to air being drawn into the suction, steam is used to seal the pump casing, which would overheat the pump.
- b. Cooling is required to enable a lower condenser vacuum to be attained.
- c. Cooling is required for the exhaust gases as they exit the mechanical vacuum pump.
- d. As air is compressed, heat is produced which would cause the pump to overheat.
ANSWER: d. KA: 256000GK.07 RO/SRO: 3.4
Reference:
Air Removal System Text Comments: Similar Question (41611) (Plant Sys - Gp II) LO: None I Page: 58 1 l
I REACTOR OPERATOR QUESTION: 55 (1.00) With bus MC-8A deenergized, what would be the consequence of pushing the REVERSE TRANSFER pushbutton on IN-27
- a. Loss of pawer to PP-7A-A.
- b. No effect, the ABT is power seeking and no transfer will occur.
- c. No effect, IN-2 would already be in a reverse power condition.
- d. Loss of power to PP-8A-A.
ANSWER: d. KA: 262002K4.01 RO/SRO: 3.4
Reference:
UPS System Text Comments: Similar Question (3491)(Plant Sys - Gp II) LO: 5892 i l Page: 59
i l REACTOR OPERATOR ! QUESTION: 56 (1.00) A loss of 250 VDC Motor Control Center MC-S2-1 A has occurred. Which one (1) of the following describes the direct effect this condition will have on the Reactor Core Isolation Cooling (RCIC) System?
- a. RCIC initiation logic power is lost, but RCIC can still be manually initiated.
- b. RCIC-V-1 (RCIC turbine trip valve) indication and control will be lost rendering RCIC INOPERABLE.
- c. RCIC flow control will not function in automatic, but can still be used in manual.
- d. RCIC valve indications are lost, however, all system functions still work.
ANSWER: b. KA: 263000K3.03 RO/SRO: 3.8
Reference:
DC Power System Text Comments: New question (Plant Sys - Gp II) LO: 7657 i i Page: 60
- - _. - . . . . - . . - ~ _ . - . . . - . - _ - . .- . - - - - .
REACTOR OPERATOR QUESTION: 57 (1.00) A reactor startup from cold conditions is in progress, a vacuum is being drawn in the main condenser using ) both AR-P-1 A and AR-P-1B (mechanical vacuum pumps). MS-RIS-610B (main steam line radiation ) j monitor) has generated an INOP trip. ) l Which one (1) of the following describes the effect of the above conditions? l a. Both AR-P-1 A and AR-P-1B will trip.
- b. Neither AR-P-1 A or AR-P-1B will trip.
- c. Only AR-P-1B will trip.
2
- d. Only AR-P-1 A will trip.
', ANSWER: b. KA: 272000Kl.02 RO/SRO: 3.5
Reference:
PRM System Text Comments: New Question (Plant Sys - Op 11) LO: 5647f Page: 61
REACTOR OPERATOR QUESTION: 58 (1.00) Due to a line break in the fire protection system header, system pressure has dropped to 108 psig and has remained there for two (2) minutes. Which one (1) of the following describes the status of the fire protection pumps after this period of time?
- a. FP-P-2A (electric pump) running FP-P-2B (electric pump) running FP-P-1 (diesel pump) running FP-P-110 (diesel pump) not running
- b. FP-P-2 A (electric pump) running FP-P-2B (electric pump) not running FP-P-1 (diesel pump) running FP-P-110 (diesel pump) not running l
not running 1
- c. FP-P-2A (electric pump)
FP-P-2B (electric pump) not running FP-P-1 (diesel pump) running FP-P-110 (diesel pump) running
- d. FP-P-2A (electric pump) not running FP-P-2B (electric pump) not running FP-P-1 (diesel pump) running l FP-P-110 (diesel pump) not running l ANSWER: a.
KA: 286000A2.06 RO/SRO: 3.1
Reference:
Fire Protection System Text Comments: Modified Question (3742)(Emer & Abn - Gp II) l LO: 5377 l 1 1 Page: 62
i REACTOR OPERATOR 1 QUESTION: 59 (1.00) l ? ' ) Steam tunnel cooling fans "A" and "B" are in service. A main steam line break results in steam tunnel pressure in excess of 0.8 psi. What actions will occur as a result of this transient? 1
- a. Standby gas treatment initiates.
J 4 I
- b. Reactor building ventilation isolates. 1
, I l
- c. Steam tunnel cooling fan "C" auto stans.
- d. Steam tunnel blowout panels relieve. l
' l ANSWER: d. KA: 29000lK5.01 RO/SRO: 3.3 4
Reference:
Secondary Containment System Text l I Comments: New question (Plant Sys - Gp I) LO: 7003 ! l l l 1 l k 4 i Page: 63
- j. REACTOR OPERATOR QUESTION: 60 (1.00) i A leak has developed in the drywell, current plant conditions are as follows:
I RPV level -40" and down slow 4
- RPV pressure 850 psig and down slow Drywell pressure 1.66 psig and up slow Reactor building exhaust plenum radiation 15 mr/hr WMA-FN-51B (recirc fan) has been observed to have automatically staned.
I
- - Which one (1) of the following describes the reason WMA-FN-51B automatically staned?
1 { a. High drywell pressure isolated normal control room ventilation, automatically starting WMA-
- j. FN-51 B.
- b. - Low RPV level directly caused an automatic start of WMA-FN-51B.
- c. Reactor building exhaust plenum high radiation automatically staned WMA-FN-54B (emergency filter unit fan) which subsequently automatically staned WMA-FN-518.
- d. High drywell pressure tripped WEA-FN-51 (toilet / kitchen exhaust fan) which caused an automatic start of WMA-FN-51B.
ANSWER: c. KA: 290003 A3.01 RO/SRO: 3.3
Reference:
CR-HVAC System Text Comments: New Question (Plant Sys - Op II) LO: 7649 l l l Page: 64
REACTOR OPERATOR QUESTION: 61 (I.00) A Traversing In-Core Probe (TIP) trace is being taken when a high drywell pressure signal is received. Which one (1) of the following describes the automatic iesponse of the TIP system?
- a. The TIP drive shifts to reverse withdrawing the detector to the "in-shield" position, then the shear valve fires.
- b. The TIP shear valve immediately fires, cutting the detector cable and sealing the guide tube.
- c. The TIP drive shifts to reverse withdrawing the detector to the "in-shield" position, then the ball valve closes.
- d. The TIP ball valve immediately closes, cutting the detector cable and sealing the guide tube.
ANSWER: c. KA: 21500lK4.01 RO/SRO: 3.4
Reference:
TIP System Text Comments: New Question (Plant Sys - Gp III) LO: 6989 l i Page: 65
.. - ----- . - . . = - _ . . - . ... - . . - . . - . _ . - . . .
l REACTOR OPERATOR QUESTION: 62 (1.00) ! ! RCC-V-129 (reactor closed cooling water supply to fuel pool cooling heat exchangers) has failed closed.
- Initial fuel pool temperature is 95 ' F and going up at 10
- F/ hour.
Which one (1) of the following describes the next automatic action, if any, that will occur in the fuel pool l cooling system? l a. When fuel pool temperature reaches 125 ' F, service water will supply cooling to the fuel pool
- cooling heat exchangers.
1
- b. As temperature rises in the spent fuel pool, evaporative cooling causes a loss ofinventory.
l The skimmer surge tank level will go down and COND-V-42 (condensate transfer makeup to j skimmer surge tank) will open.
- c. FPC-V-175 (filter demineralizer bypass) opens at 105
- F to prevent resin breakdown and j chemicalintmsion.
i
- d. - Increasing temperature in the spent fuel pool has NO effect on actions in the fuel pool cooling system.
l ANSWER: b. I ! KA: 233000A2.07 i RO/SRO: 3.0
Reference:
FPC System Text
! Comments: New Question (Plant Sys - Op III)
- LO: 5371a 1
l l
)
I I Page: 66
_ . . _ . ._ __ ___ __. - - _ _ _ . . . _ _ _ . _ _ ~ . _ __ _ _. REACTOR OPERATOR QUESTION: 63 (1.00) l ? e Drywell pressure is currently 2.0 psig and up slow. What effect, if any, does this have on the reactor building ventilation (RBHVAC) system? l
- a. - REA-V-1, REA-V-2, ROA-V-1 & ROA-V-2 (reactor building ventilation isolation valves)
! receive a close signal
- - REA-FN-1 & REA-FN-2 (reactor building ventilation supply fans) and ROA-FN-1 & ROA-FN-2 (reactor building exhaust fans) receive a trip signal
- Emergency Core Cooling System (ECCS) emergency cooling fans start 1 b. - RBHVAC continues to operate
. - The Standby Gas Treatment (SGT) system will NOT automatically start on high drywell ! pressure. 4 c. - REA-V-1, REA-V-2, ROA-V-1 & ROA-V-2 (reactor building ventilation isolation valves) remain open ;
- REA-FN-1 & REA-FN-2 (reactor building ventilation supply fans) and ROA-FN-1 & ROA- l FN-2 (reactor building exhaust fans) receive a trip signal t
- Emergency Core Cooling System (ECCS) emergency cooling fans receive a trip signal i I
- d. - RBHVAC continues to operate
- The SGT system will automatically start ;
ANSWER: a. ! KA: 288000A2.03 RO/SRO: 3.5
Reference:
RBHVAC System Text Comments: Question (3330) (Plant Sys - Op "" LO: 5679 l l l Page: 67
REACTOR OPERATOR QUESTION: 64 (1.00) Thermal limits are established to maintain fuel integrity. Which one (1) of the following statements describes the " limiting condition" and/or " failure mechanism" for one of the thermallimits?
- a. Critical Power Ratio (CPR) limits ensure that the fuel cladding will not be subjected to greater
, than 1% plastic strain. t , b. Average Planar Linear Heat Generation Rate (APLHGR) limits ensure that the fuel cladding l temperature does not exceed 2200
- F during a Design Basis Loss Of Coolant Accident (DB A LOCA).
l c. Linear Heat Generation Rate (LHGR) limits prevent fuel clad cracking due to high stress by I limiting fuel enthalpy to less than 280 calories / gram.
- d. Pre-Conditioning Interim Operating Recommendation (PCIOMR) limits prevent exceeding local power limits which could thermally fatigue the cladding.
ANSWER: b. KA: 290002K5.01 RO/SRO: 3.5
Reference:
Fuel System Text Comments: Question (236)(Plant Sys - Op III) ! LO: 5388 i l l 4 e 4 s Page: 68
i e REACTOR OPERATOR j QUESTION: 65 (1.00) l l A reactor scram hasjust occurred. The Rod Sequence Control System (RSCS) and Rod Wonh Minimizer ! (RWM) have not functioned to give the ALL RODS IN information. ; Which one (1) of the following H13-P603 indications may be used to verify rods full in?
- a. White Reactor Protection System (RPS) group lights deenergized.
i
- b. Amber backup scram lights deenergized.
- c. Green fullin lights energized i r )
- d. Blue scram lights energized.
i ANSWER: c. KA: 295006GK.05 l RO/SRO: 4.0 l
Reference:
RSCS Systems Text l Comments: Modified question (3746)(Emer & Abn - Gp I) l LO: 5807 i 1 i i i o Page: 69
l 1 1 REACTOR OPERATOR l QUESTION: 66 (1.00) , Initial plant conditions are as follows:
- Reactor power 100 % l
- RPV pressure 1020 psig
- RPV water level 36" A reactor scram occurs and the scram inlet valve (126) of a single control rod mechanically binds and fails to ;
open. . Which one (1) of the following describe the control rod's response to this failure? i
- a. Fully inserts and its blue scram light is energized.
- b. Fails to insert and its blue scram light is energized. '
l c. Fully inserts and its blue scram light is deenergized. . t
- d. Fails to insert and its blue scram light is deenergized.
ANSWER: c. l l KA: 295006Al.06 3 l RO/SRO: 3.6 ! l
Reference:
CRDil System Text ! [ Comments: Modified Question (341) (Emer & Abn - Op 1) l LO: 5184 t l I l l l 1 Page: 70
I' REACTOR OPERATOR QUESTION: 67 (1.00) The reactor is operating at =98% power. An equipment operator repons a lube oil leak in the "B" feedwater 4 pump room. Immediately after acknowledging the repon, RFW-P-1B ("B" reactor feedwater pump) trips on low lube oil pressure. l Which one (1) of the following describes the effect this condition has on the reactor recirculation system?
- a. Only RRC-P-1B ("B" reactor recirculation pump) will runback to 15Hz.
l b. Both RRC-P-1 A and RRC-P-1B ("A" & "B" reactor recirculation pumps) will runback to 27Hz
- c. Both RRC-P-1 A and RRC-P-1B ("A" & "B" reactor recirculation pumps) will runback to 15 l Hz.
l d. Only RRC-P-1B ("B" reactor recirculation pump) will runback to 52.2 Hz. ANSWER: b. KA: 295009K2.03 l RO/SRO: 3.2 i
Reference:
ASD System Text Comments: New question (Emer & Abn - Gp I) LO: 9683 l l l 1 l l: 4 1 Page: 71
I REACTOR OPERATOR QUESTION: 68 (1.00) The plant hasjust suffered a turbine trip from 100% power and main turbine bypass valves (BPVs) have stuck open. Which one (1) of the following describes the consequences of closing MS-V-146 (main steam supply to auxiliaries) to limit the uncontrolled cooldown? Loss of steam to.. l
- a. building heat nitrogen inerting bypass valves, seal steam evaporators
)
offgas preheaters
- b. steam jet air ejectors 1 reactor feed pump turbines !
seal steam evaporators ) moisture separator reheater second stage offgas preheaters. l
- c. reactor feed pump turbines l offgas preheaters -
moisture separator preheater second stage l nitrogen inerting system. I bypass valves
- d. reactor feed pump turbines moisture separator reheaters first stage seal steam evaporator i building heat bypass valves ANSWER: b.
KA: 295009A1.03 RO/SRO: 3.1
Reference:
PPM 4.2.1.14, Main Steam System Text Comments: Similar Question (5459) (Emer & Abn - Gp I) LO: 5525g Page: 72
REACTOR OPERATOR QUESTION: 69 (1.00) The control room operator (CRO) is withdrawing control rods with the reactor critical and power indicating on the IRMsjust prior to the point of adding heat. The CRO observes an unexpected rapid increase in power and a period indication of =30 seconds. Assuming NO OPERATOR ACTION, which one (1) of the following scram signals will terminate this transient?
- a. Reactor short period.
l l
- b. Average Power Range Monitor (APRM) neutron flux high.
- c. Source Range Monitor (SRM) upscale
- d. Intermediate Range Monitor (IRM) neutron flux high l
ANSWER: d. KA: 295014 A2.01 RO/SRO: 4.2
Reference:
IRM System Text Comments: Modified Question (249)(Emer & Abn - Op I) LO: 5459 l
'l l
l ; i I 4 Page: 73
REACTOR OPERATOR QUESTION: 70 (1.00) The plant is operating at =98% power when the following indications are noted: Reactor power down slow. Megawatts down slow. Control air pressure down slow. l Three (3) control rods indicate FULL-IN with scram lights energized on the full core display.
- Which one (1) of the following statements describes the actions required to be taken given the above I indications.
, a. NO actions are required until the first Main Steam isolation Valve (MSIV) is showing dual position indication. l'
- b. Close or verify closed CN-V-65 (containment instrument air crosstie shut-off valve.
I
- c. Initiate a manual reactor scram and refer to PPM 3.3.1.
- d. Lower core flow to reduce reactor power to LT 90% of rated core thermal power.
i l ANSWER: c. l l i KA: 295015Al.02 l RO/SRO: 4.2
Reference:
CAS System Text and PPM 4.1.1.7B Comments: New Question (Emer & Abn - Op I) LO: 7605 l l 1 i i i Page: 74
, -m. . _
REACTOR OPERATOR QUESTION: 71 (1.00) Which one (1) of the following systems was specifically designed to ensure reactor power could be monitored under DBA/LOCA conditions?
- a. Source Range Monitoring (SRM) system.
- b. Local Power Range Monitoring (LPRM) system. '
- c. Wide Range Monitoring (WRM) system.
- d. Intermediate Range Monitoring (IRM) system ANSWER: c.
KA: 295015K2.08 RO/SRO: 3.7
Reference:
WRM System Text Comments: Modified Question (3707) (Emer & Abn - Gp I) LO: 5963 i l Page: 75
a 1 REACTOR OPERATOR 9 QUESTION: 72 (1.00) i PPM 5.2.1 " Primary Containment Control" directs that when drywell pressure exceeds 39 psig the primary
, containment is to be vented to reduce and maintain wetwell pressure below the Primary Containment Pressure Limit (PCPL).
1 Which one (1) of the following statements describes the preferred vent path and the reason that this path is
- preferred?
- a. Drywell, this is the vent path with the highest flowrate capacity.
i b. Wetwell, to take advantage of suppression pool scrubbing for minimizing the amount of radioactivity released.
- c. Drywell, in order to minimize the moisture saturation and breakdown of the Standby Gas l Treatment (SGT) system charcoal adsorbers.
} } d. Wetwell, in order to minimize cycling, and potential failure of the wetwell to drywell vacuum j breakers. i ANSWER: b. 3 KA: 295024K3.07
- RO/SRO: 4.0 l
Reference:
PPM 5.2.1 J Comments: Question (512)(Emer & Abn - Op I) . LO: 8363 (PPM 5.2.1) i I s I 1 1 i Page: 76
l l REACTOR OPERATOR l QUESTION: 73 (1.00) The plant was operating at =98% power when a leak in the discharge of a condensate booster pump caused a j low suction pressure trip of the reactor feedwater pumps. RPV level dropped to -25" initially and is now l going down very slow, the Control Room Supervisor (CRS) has entered PPM 5.1.1, RPV Control, and is 4 executing all legs concurrently. Wetwell temperature hasjust been reported at 92 F and up slow. j i Which one (1) of the following describes the Emergency Operating Procedure (EOP) implementation to be used under these conditions?
- a. Continue PPM 5.1.1, RPV Control, RPV level steps, AND enter PPM 5.3.1, Secondary Containment Control.
- b. Continue PPM 5.1.1, RPV Control, AND concurrently enter PPM 5.2.1, Primary Containment Control. )
- c. Complete PPM 5.1.1, RPV Control, RPV level steps, THEN enter PPM 5.2.1, Primary ,
Containment Control. l l
- d. Reenter PPM 5.1.1, RPV Control, AND concurrently enter PPM 4.12.4.l A High Energy Line Break.
ANSWER: b. KA: 295024GK.I 1 RO/SRO: 4.5
Reference:
PPM 5.0.10, Sect. 3.5, PPM 5.1.1, PPM 5.2.1 Comments: Modified Question (721) (Emer & Abn - Gp I) LO: 8017 l Page: 77
REACTOR OPERATOR ! QUESTION: 74 (l.00) 4 j The plant is in a condition requiring the Control Room Supervisor (CRS) to execute PPM 5.1.1, RPV Level i Control, and PPM 5.2.1, Primary Containment Control, concurrently. The CRS has directed a pressure
- reduction which exceeds the normal, allowable RPV cooldown rate of 100 F/Hr.
l Which one (1) of the following describes a condition that would allow the CRS to take this action?
- a. Prevent RPV level from going LT Top of Active Fuel (TAF).
j b. Prevent exceeding Drywell Spray Initiation Limit (DSIL). 4 j c. Prevent exceeding Heat Capacity Temperature Limit (HCTL). j d. Prevent exceeding Maximum Primary Containment Water Level Limit (MPCWLL). i i ANSWER: c. 1 i j KA: 295025A2.03 RO/SRO: 4.1
Reference:
PPM 5.0.10 I Comments: Modified Question (Emer & Abn - Gp I) I LO: 8048 (PPM 5.1.1) l l i i i ? l k i i i 2 4 i i 1 Page: 78
I i REACTOR OPERATOR QUESTION: 75 (1.00) > In PPM 5.1.4, RPV Flooding, achieving FLOODING COMPLETION TIME ensures that RPV level is GE to.. f i
- a. the Top of Active Fuel (TAF).
{
- b. the Main Steam Line (MSL) openings. !
- c. 2/3 core height.
- d. the reactor head vents..
ANSWER: a. 3 i KA: 295031K1.01 i RO/SRO: 4.7 f
Reference:
PPM 5.0.10, pg 137 Comments: Modified Question (659) (Emer & Abn - Gp I) LO: 8219 (PPM 5.1.4-14) I l 1 I l 1 l Page: 79 1 1 l l
REACTOR OPERATOR QUESTION: 76 (1.00) When using the Reactor Core Isolation Cooling (RCIC) system for alternate boron injection, the contents of the Standby Liquid Control (SLC) storage tank are gravity fed to the RCIC pump suction by a temporary hose connection originating at..
- a. any drain off the SLC suction piping.
- b. the drain offof the SLC storage tank.
- c. a drain on the common SLC discharge header, downstream of SLC-V-4A & SLC-V-4B (squib valves).
- d. the tank side of either the "A" or "B" SLC system relief valve piping.
ANSWER: d. KA: 295037K2.13 RO/SRO: 4.1
Reference:
PPM 5.5.8 and RCIC System Text Comments: Modified Question (3728)(Emer & Abn - Gp I) LO: 5929 i l 1 Page: 80
d REACTOR OPERATOR 4
. QUESTION: 77 (1.00) 4 j Which one (1) of the following describes two (2) methods that can be used for positive confirmation that all
- rods are fully inserted?
j , i a. Average Power Range Monitors (APRMs) LT 5% power and Reactor Engineering calculation : 1 showing adequate shutdown margin. ;
- b. Graphic Display System (GDS) and Plant Process Computer Replacement System (PPCRS). ,
- c. Plant Process Computer Replacement System (PPCRS) and Quick Emergency Dose Projection .
System (QEDPS) ; ) d. Graphic Display System (GDS) and Average Power Range Monitors (APRM)s LT 5% power. i l a e ANSWER: b. 4 i KA: 295037A2.01 .
- . RO/SRO: 4.3 !
j Reference; PPM 5.0.10, page 222 i Comments: New Question (Emer & Abn - Op II) LO: 8182 (5.1.1-37 & 5.1.2-58) !- 1 4 1 1 i f 1 i I i 1 i i i 'l i i I. 4 5 j Page: 81 i ,
REACTOR OPERATOR QUESTION: 78 (1.00) The plant is operating at 75% power and 70% core flow when an electrical malfunction in the main turbine trip circuitry causes both reactor recirculation pumps to trip off. Which one (1) of the following IMMEDIATE ACTIONS should be taken?
- a. The recirculation pump trips will cause a RPV high pressure scram. Perform the immediate scram actions per PPM 3.3.1..
- b. Refer to the single loop operating procedure in PPM 2.2.1 to restart one of the reactor recirculation pumps.
- c. Confirm the loss of both reactor recirculation pumps and then manually scram the reactor.
- d. Use the fast shutdown sequence control rods to exit Region "C" within 15 minutes.
ANSWER: c. KA: 295001G.10 RO/SRO: 3.7
Reference:
PPM 4.12.4.7 Comments: Question (503)(Emer & Abn - Op II) LO: 5023c 1 Page: 82 !
l REACTOR OPERATOR QUESTION: 79 (1.00) The reactor is operating at 93% power when a loss of all circulating water pumps occurs. Assuming NO OPERATOR ACTION, as vacuum degrades to 14" Hg, what will be the effect on RPV water level? RPV water level will..
- a. increase to +54" and then cycle between -50" and +54"
- b. decrease to LT 0" and then stabilize at +18"
- c. be maintained at setpoint.
- d. decrease to +13" and then stabilize at +36" ANSWER: b.
KA: 295002K3.01 RO/SRO: 3.8
Reference:
FWLC System Text Comments: Modified Question (6270) (Emer & Abn - Gp II) LO: 5400f l i Page: 83
REACTOR OPERATOR QUESTION: 80 (1.00) During a " Station Blackout" plant parameters are as follows:
- RPV water level -52" and up slow
- RPV pressure 850 psig and down slow
- Wetwell pressure 19 psig and up slow
- Drywell temperature 243 F and up slow
- Wetwell temperature 112 F
- Wetwell level + 3" Which one (1) of the following interlocks must be defeated to allow continued Reactor Core Isolation Cooling (RCIC) system operation under these conditions?
- a. High exhaust pressure turbine trip.
- b. RCIC exhaust diaphragm rupture isolation.
- c. Level 2 RCIC turbine trip.
- d. Drfwell high temperature RCIC system isolation. j ANSWER: a.
KA: 295003 Al.03 RO/SRO: 4.4
Reference:
PPM 5.6.1 Comments: Modified Question (4402) (Emer & Abn - Gp I) LO: 5722 l I i l l I Page: 84
l 4 4 REACTOR OPERATOR QUESTION: 81 (1.00) Due to a fault, MC-7A has been deenergized and will be out of service for a minimum of eight (8) hours. Which one (1) of the following will be affected by this condition? l a. Uninterruptable Power Supply (UPS) static inverter IN-1.
- b. Critical instrument inverter IN-2.
- c. ATWS/ARI Division 2 logic power. l 1
i 4 d. DG-1 control circuit power
; l i
ANSWER: a. , 1 KA: 295004K2.03
; RO/SRO: 3.3 i
i
Reference:
DC Power System Text Comments: New question (Emer & Abn - Gp II) LO: 5263 1 i i e l l l J 1 Page: 85
REACTOR OPERATOR QUESTION: 82 (1.00) A plant startup is in progress with reactor pressure =500 psig. RFW-FCV-10A and RFW-FCV-10B (feedwater startup valves to the reactor) both fail full open. RPV level is 55" and rising. What IMMEDIATE ACTIONS should be taken to preclude flooding the main steam lines?
- a. Prior to reaching an RPV level of +80", scram the reactor, and close the Main Steam Isolation Valves (MSIVs).
- b. Stop the condensate booster pumps before RPV water level exceeds +80"
- c. Prior to reaching an RPV level of +108", close RFW-V-118 (feedwater stanup valve isolation) and leave it closed until RPV level is LT +54"
- d. Stop all condensate and condensate booster pumps before RPV water level exceeds v60" ANSWER: b.
KA: 295008A1.08 RO/SRO: 3.5 :
Reference:
PPM 4.2.1.2 l Comments: Modified Question (504) (Emer & Abn - Gp II) LO: 5400 l l I Page: 86 i
REACTOR OPERATOR QUESTION: 83 (1.00) The plant has experienced a transient, Emergency Operating Procedures (EOPs) have been entered and cond tions are as follows: RPV water level -150"and down slow RPV pressure 180 psig and down slow Wetwell temperature 110 F and up slow RHR loop "A" injecting to the RPV RHR loop "B" in suppression pool cooling All other injection sources are unavailable Which one (1) of the following statements best describes actions that need to be taken given the above information?
- a. Open seven (7) Automatic Depressurization System (ADS) Safety Relief Valves (SRVs) to emergency depressurize.
- b. RHR loop "B" should be removed from suppression pool cooling and injected into the RPV.
- c. RHR loop "A" should be removed from injection and placed into suppression pool cooling. ;
l
- d. No actions are required until RPV level lowers to LE -192" ANSWER: b.
KA: 295013K2.01 RO/SRO: 3.6
Reference:
PPM 5.2.1 Comments: New question (Emer & Abn - Gp I) LO: 8304 l i Page: 87
l j REACTOR OPERATOR j QUESTION: 84 (1.00) A "Most Immediate" control room evacuation is required due to heavy smoke intmsion. Wiiich one (1) of the following statements lists only IMMEDIATE ACTIONS that should be taken prior to exiting the control room? l
- a. Manually scram the reactor, lock the reactor mode switch in SHUTDOWN and close the Main Steam Isolation Valves (MSIVs).
I b. Manually scram the reactor, initiate Reactor Core Isolation Cooling (RCIC) and make a plant announcement.
- c. Manually scram the reactor, close the MSIVs and transfer RPV level control to RFW-FCV-10A and RFW-FCV-10B (feedwater startup valev to the reactor).
- d. Manually scram the reactor, lock the reactor mode switch in SHUTDOWN and stan diesel generator #2 ANSWER: a.
1 i KA: 295016K3.01 RO/SRO: 4.1
Reference:
PPM 4.12.1.1 Comments: Modified Question (256)(Emer & Abn - Gp I) l LO: None l \ l i i t Page: 88
REACTOR OPERATOR l QUESTION: 85 (1.00) The plant has experienced a transient, PPM 5.1.2 has been entered, plant parameters are as follows: l
- RPV water level -145" and steady
- Drywell pressure 10 psig and down slow i - Wetwell temperature 110 F and up very slow i
- Main Steam Isolation Valves (MSIVs) are closed
; - Both Standby Liquid Control (SLC) pumps are injecting j
! Which one (1) of the following identifies a valid annunciator that would preclude / prevent reopening the MSIVs? ) a. LPCS/RHR A INIT RPV LEVEL LOW -129" j b. DRYWELL PRESS HIGH TRIP.
- c. NSSSS ISOL MSL FLOW HIGH.
a j d. RC-1 HALF TRIP. l ANSWER: c. KA: 295017K3.01 ) RO/SRO: 3.6 1
Reference:
PPM 5.1.2 and NS4 System Text ! Comments: Modified Question (677)(Emer & Abn - Op I) j LO: None a f Page: 89
REACTOR OPERATOR i QUESTION: 86 (1.00) i l The plant is operating at rated conditions with Reactor Closed Cooling Water (RCC) loads being supplied by , RCC-P-1 A and RCC-P-1C ("A" & "C" reactor closed cooling water pumps) when a fuse in the 125VDC l power supply to the RCC-P-1 A breaker close logic blows. Which one (1) of the following is an action that should be taken?
- a. Start CRD-P-1B ("B" contro! rod drive pump) and trip CRD-P-1 A ("A" control rod drive :
pump).
- b. Monitor drywell temperature and pressure, enter PPM 5.2.1, Primary Containment Control )
when entry conditions are met. j
- c. Scram the reactor and trip RRC-P-1 A and RRC-P-1B ("A" and "B" reactor recirculation.
pumps).
- d. Trip RWCU-P-1 A & RWCU-P-1B ("A" & "B" reactor water cleanup pumps) and close RWCU-V-4 (reactor water cleanup outboard isolation valve).
ANSWER: d. KA: 295018K2.02 RO/SRO: 3.6
Reference:
TSW Systems, Rev. 7, PPM 4.8.3.2, EWDs IlE001, IlE017 Comments: New Question (Emer & Abn - Gp II) LO: 5706d l l Page: 90
REACTOR OPERATOR QUESTION: 87 (1.00) l l Which one (1) of the following is expected to occur at a control air header pressure of 80 psig?
- a. SA-PCV-2 (control / service air crosstie valve) closes.
- b. Standby control air compressor (s) automatically start.
- c. CAS-PCV-1 (desiccant dryer bypass valve) opens.
- d. Control air header low pressure alarm is received.
ANSWER: a. 1 i KA: 295019A2.01 j RO/SRO: 3.6
Reference:
CAS System Text , I Comments: New Question (Emer & Abn - Gp II) LO: 5878 i 1 I i-l l l l t i ) Page: 91
- i. _. . .
l ' l l REACTOR OPERATOR l QUESTION: 88 (1.00) Given the following control air compressor parameters: l
- 1. High discharge air temperature 4. High cooling water temperature
- 2. Low cooling water pressure 5. High discharge air pressure
- 3. Low oil pressure 6. High discharge flow Which one (1) of the following identifes the parameters, which if exceeded, would cause a trip of the control air compressor?
- a. 2,4 and 6
- b. 1,2 and 3 l
i ( c. 1,3 and 5
- d. 4,5 and 6 1
i ANSWER: b. KA: 295019GK.05 RO/SRO: 3.3
Reference:
CAS System Text Comments: Modified Question (4078) (Emer & Abn - Gp II) LO:. 5872 i I i I l . l l 2 4 l Page: 92
REACTOR OPERATOR QUESTION: 89 (1.00) The plant is operating at 100% reactor power when an inadvertent Group 7 Nuclear Steam Supply Shutoff System (NS4) isolation occurs. Which one (1) of the following describes the expected plant response?
- a. Reactor Closed Cooling (RCC) system supply and return containment isolation valves close.
- b. Reactor Water Cleanup (RWCU) system containment isolation valves close.
- c. Residual Heat Removal (RHR) system reactor water sample isolation valves close.
- d. Primary containment recirculation fans trip.
ANSWER: b. KA: 295020K2.04 RO/SRO: 3.1
Reference:
NS4 System Text Comments: Question (259)(Emer & Abn - Gp II) LO: 5598 Page: 93
REACTOR OPERATOR QUESTION: 90 (1.00) The plant is operating at =97% when the following annunciators are received: H13-P603. A7-6.7 ROD ACCUMULATORTROUBLE
- (7hefull core display indicates this alarm isfor afully withdrawn rod)
H 13-P603. A7-3.8 CRD CHARGE WATER PRESS LOW The control room operator (CRO) observes that CRD-P-1 A ("A" control rod drive pump) motor current indicates zero (0) amps with the red light on. Which one (1) of the following describes the IMMEDIATE ACTIONS required for this situation?
- a. Place the reactor mode switch to SHUTDOWN and carry out the scram recovery per PPM 3.3.1, Reactor Scram.
- b. Place the Control Rod Drive (CRD) flow controller in MANUAL and raise controller output while monitoring CRD-P-1 A motor current.
- c. Place the standby CRD suction filter in service locally and start CRD-P-1B.
- d. Place the CRD flow controller in MANUAL, set the controller output at zero (0) and start CRD-P-18.
ANSWER: d. KA: 295022Al.01 RO/SRO: 3.4 ,
Reference:
PPM 4.1.1.2 ; Comments: Modified Question (360) (Emer & Abn - Gp II) LO: 5192 i l Page: 94
l REACTOR OPERATOR QUESTION: 91 (1.00) A plant transient has caused a reactor scram. Plant conditions are as follows:
- Reactor power = 15%
- RPV pressure 1000 psig and steady
- RPV level -125" and down
- Wetwell temperature 165 'F and up slow
- Wetwell level 32.5' and up very slow Which one (1) of the following describes the operation of the Safety Relief Valves (SRVs) with the above conditions?
- a. Heat Capacity Level Limit (HCLL) has been exceeded, emergency depressurization is required.
- b. Safety Relief Valve Tailpipe Level Limit (SRVTPLL) has been exceeded, emergency depressurization is required.
- c. Heat Capacity Temperature Limit (HCTL) has been exceeded, emergency depressurization is l required.
- d. No limits have been exceeded, cycle SRVs to maintain RPV pressure between 800 and 1000 l psig.
)
l ANSWER: c. I l KA: 295026K3.01 i RO/SRO. 3.8 l
Reference:
PPM 5.2.1, ADS System Text l Comments: New question (Emer & Abn - Gp I) l LO: 8379 (PPM 5.2.1-77) l l d Page: 95
.. . -. . . . ~ . - . . . _- . . _ . - ~ . - _ _
)
REACTOR OPERATOR QUESTION: 92 (1.00) l Following a small steam line break inside primary containment, average drywell temperature has increased by l about 100
- F. j i
Assuming that actual RPV water level remains constant, indicated vessel level could be.. i
- a. higher, as heating of the reference leg decreases differential pressure
- b. lower, as heating of the reference leg increases differential pressure.
- c. higher, as heating of the reference leg increases differential pressure.
I
- d. lower, as heatmg of the reference leg decreases differential pressure.
l ANSWER: a. !
)
KA: 295028Kl.01 j RO/SRO: 3.7 l
Reference:
PPM 5.0.10 l Comments: Question (567)(Emer & Abn - Op II) LO: 8448 (PPM 5.1.Is.i5) I I Page: 96
REACTOR OPERATOR QUESTION: 93 (1.00) Given plant conditions as follows:
- Wetwell level 36'
- Wetwell pressure 10 psig '
- RPV pressure 1000 psig Using the attached curves, identify the possible results of Safety Relief Valve (SRV) actuation.
Actuation of an SRV..
- a. is allowed and desired given the above conditions.
- b. at this elevated wetwell level could result in damage to SRV internals.
- c. will result in exceeding the suppression pool boundary design load.
- d. could result in damage to the SRV tail pipe, quenchers, or supports.
ANSWER: d. KA: 295029K2.06 RO/SRO: 3.5
Reference:
PPM 5.0.10, Section 7.16 Comments: New Question (Emer & Abn - Gp II) LO: 8381 (PPM 5.2.1-83) I I I l l I Page: 97
REACTOR OPERATOR QUESTION: 94 (1.00) Primary containment water level cannot be maintained below the Maximum Primary Containment Water Level Limit (MPCWLL), 552'2". Injection from sources external to the primary containment must be terminated in order to prevent..
- a. failure oflow pressure Emergency Core Cooling System (ECCS) suction piping due to the static head of water.
- b. loss of the ability to determine off-site radiation release rates.
- c. loss of the ability to vent the primary containment.
- d. overloading Emergency Core Cooling System (ECCS) pump motors due to the additional flow resulting from the higher suction head.
ANSWER: c. KA: 295030GK.07 RO/SRO: 3.6
Reference:
PPM 5.0.10 Comments: Similar Question (2497) (Emer & Abn - Op II) LO: 8405 (PPM 5.2.1) i l i l l 1 i Page: 98
i l f REACTOR OPERATOR ; i QUESTION: 95 (1.00) ! l ! Which one of the following lists actions that can be used to mitigate off-site doses for an accident which i releases radioactivity inside secondary containment? :
- a. Isolate primary systems leaking into the area Shutdown Reactor Building Ventilation (RBHVAC)
Isolate the Standby Gas Treatment (SGT) system l
- b. Isolate.the Standby Gas Treatment (SGT) system !
Shut down the reactor I Emergency depressurize the reactor
- c. Isolate primary systems leaking into the area j Shutdown Reactor Building Ventilation (RBHVAC) i Shut down the reactor j
- d. Isolate primary systems leaking into the area i Shut down the reactor !
Emergency depressurize the reactor l l ! ANSWER: d. KA: 295033K203 ! RO/SRO: 3.9
Reference:
PPM 5.3.1 and 5.0.10 l Comments: New Question (Emer & Abn - Gp II) LO: 8460 (PPM 5.1.3-18) I i 1 } I . Page: 99 l
REACTOR OPERATOR QUESTION: 96 (1.00) A transport cask filled with Control Rod Drive (CRD) " spud end" filters has tipped over on the 501' elevation of the Reactor Building (RB). ARM-RIS-33 (RB 501' area radiation monitor) is alarming on control room ' panel H13-P614. Reactor building exhaust plenum radiation levels are at =15 mr/hr and up fast. i Which one (1) of the following is an " expected" response to the above conditions? I
- a. CW-P-1B & CW-P-1C ("B" & "C" circulating water pumps) trip.
- b. Any traversing in-core probe (TIP) inserted into the core will automatically withdraw and isolate.
- c. Drywell Equipment Drain (EDR) and Floor Drain (FDR) sumps isolate.
- d. Containment Nitrogen (CN) makeup isolates.
ANSWER: d. KA: 295034K2.02 RO/SRO: 3.9
Reference:
PPM 4.12.4.6, page 2 Comments: New Question (Emer & Abn - Op II) LO: 5597 Page: 100 l J
1 REACTOR OPERATOR QUESTION: 97 (1.00) l A reactor shutdown to cold conditions is in progress. Plant conditions are as follows:
- Reactor mode switch positioned in SHUTDOWN
- RPV pressure is 45 psig
- Residual Heat Removal (RHR)"B" is being warmed up for shutdown cooling mode
- Residual Heat Removal (RHR)"A" has been removed from service for ten (10) days Which one (1) of the following statements describes action (s) which must be taken for these conditions?
l
- a. Immediately place RHR Loop "B" in shutdown cooling and be in at least cold shutdown within one (1) hour.
- b. Perform a physical walkdown of the Reactor Water Cleanup (RWCU) system and then place the system in service to maintain reactor coolant temperature as low as possible.
- c. Demonstrate operability of at least one (1) alternate method of decay heat removal.
- d. Maintain both reactor recirculation pumps in operation until RHR-P-2A ("A" residual heat removal pump)is repaired and returned to service.
ANSWER: c. KA: 295021GK.08 RO/SRO: 3.2
Reference:
Tech Spec 3 4.9.1 Comments: Question (577)(Emer & Abn - Gp II) LO: None l Page: 101
l ! REACTOR OPERATOR QUESTION: 98 (1.00) A spent fuel assembly is dropped during transport in the spent fuel pool. The bridge operator observes bubbles rising from the dropped assembly. 1 Which one (1) of the following is an IMMEDIATE ACTION for this situation? I
- a. Place all assemblies in a safe location, leave the area, and call the control room.
- b. Immediately evacuate the refuel floor of all personnel.
- c. Contact Health Physics and ask for an area survey, then inform the Control Room Supervisor (CRS).
- d. Contact the refuel floor supervisor and the systera engineer, then attempt to recover the dropped assembly.
ANSWER: b. KA: 295023Kl.01 RO/SRO: 3.6
Reference:
PPM 4.12.3.1, Rev and Fuel Handling System Text Comments: Modified Question (508) (Emer & Abn - Op I) LO: 7713 l Page: 102
REACTOR OPERATOR QUESTION: 99 (1.00) PPM 5.3.1, Secondary Containment Control, was entered due to confirmed high temperatures and steam in j the l A Reactor Water Cleanup (RWCU) pump room. RWCU-V-1 & RWCU-V-4 (RWCU suction isolation valves) cannot be isolated from the control room. Maximum Safe Operating Values.for the RWCU system have NOT been exceeded. Which one (1) of the following describes the actions to be taken for this situation?
- a. Emergency depressurize.
- b. Shutdown the reactor per PPM 3.2.1.
- c. Continue efforts to isolate RWCU and enter PPM 5.1.1, RPV Level Control.
- d. Isolate Reactor Building Ventilation (RBHVAC) and initiate the Standby Gas Treaetment ,
(SGT) system. ANSWER: c. KA: 295032K3.03 RO/SRO: 3.8
Reference:
PPM 5.3.1 Comments: Question (737)(Emer & Abn - Op II) LO: 8457 (PPM 5.3.1-10) i l l l 1 Page: 103
REACTOR OPERATOR QUESTION: 100 (1.00) A failure of the reactor building ventilation system (RBHVAC) has occurred. The control room operator (CRO) has started Standby Gas Treatment (SGT) train "A" Which one (1) of the following is the SGT train "A" differential controller tape setpoint which should be set to ensure that the required negative pressure will be maintained in secondary containment?
- a. -1.7" Wg
- b. -0.6" Wg
- c. -2.5" Wg
- d. -0.25" Wg ANSWER.: a.
KA: 295035 A1.02 ; RO/SRO: 3.6
Reference:
PPM 4.10.1.1, SGT System Text Comments: Similar Question (530) (Emer & Abn - Op Ill) LO: 5828
')
i Page: 104 . l
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l J MCUT Maximum Core Uncovery Time 25 20 ; 7
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l 0 2 5 r 3.6 # 0 30 50 100 500 1000 3000 6000 Time After Reactor Shutdown (minutes) \ s
l t l l i NRC COMMENTS ON WNP-2 DRAFT WRITTEN EXAM - 10/7/96 l i I RO/SRO # COMMENT l Comments by H. Bundy l l 41/39 Distractor "d" is not believable nor compatible with other choices. 1 l Resolution: Licensee agrees. Replaced distractor. l 68/-- K/A 295009A1.03 refers to jet pump operation. Question refers to main ! I steam. l Resolution: Licensee agrees. Replaced with a question related to system. I i 1 l i l Comments by Licensee during preparation week I l 89/92 Question was inadvertently used on a practice examination. 1 l Resolution: Question was replaced on final exam. l 94 /-- Question was inadvertently used on a practice examination. Resolution: Question was replaced on final exam. l l l i t l l l i t l l L}}