ML15259A290: Difference between revisions

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(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
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New  X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:   
New  X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:   


WBN Unit 1 Reactor Trip or Safety Injection 1-E-0 Rev. 0005 Step    Action/Expected Response Response Not Obtained Page 3 of 45  
WBN Unit 1 Reactor Trip or Safety Injection 1-E-0 Rev. 0005 Step    Action/Expected Response Response Not Obtained Page 3 of 45 3.0 OPERATOR ACTIONS NOTE
 
===3.0 OPERATOR===
ACTIONS NOTE
* Steps 1 thru 4 are IMMEDIATE ACTION STEPS
* Steps 1 thru 4 are IMMEDIATE ACTION STEPS
* Status Trees / SPDS should be monitored when transitioned to another instruction.
* Status Trees / SPDS should be monitored when transitioned to another instruction.
Line 207: Line 204:
IF RCS press drops to less than 150 psig, THEN  ENSURE RHR pump flow.  
IF RCS press drops to less than 150 psig, THEN  ENSURE RHR pump flow.  


WBN System Description Document SAFETY INJECTION SYST EM WBN-SDD-N3-63-4001 Rev. 0030 Page 119 of 159  
WBN System Description Document SAFETY INJECTION SYST EM WBN-SDD-N3-63-4001 Rev. 0030 Page 119 of 159 8.2 Figure 2 - SIP Composite Curve
 
===8.2 Figure===
2 - SIP Composite Curve
: 4. Which ONE of the following describes the method used to remove decay heat 5 minutes after a LBLOCA has occurred?  
: 4. Which ONE of the following describes the method used to remove decay heat 5 minutes after a LBLOCA has occurred?  


Line 414: Line 408:
Higher  X  Lower    Question Source:
Higher  X  Lower    Question Source:
New X  Modified Bank Bank    Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank    Question History: New question for the 2015-301 NRC RO Exam Comments:
WBN Unit 1 Loss of RHR Shutdown Cooling 1-AOI-14 Rev. 0002 Page 4 of 87  
WBN Unit 1 Loss of RHR Shutdown Cooling 1-AOI-14 Rev. 0002 Page 4 of 87 3.0 OPERATOR ACTIONS 3.1 Diagnostics IF GO TO Subsection RHR Pump Trip During Midloop Operation.
 
===3.0 OPERATOR===
ACTIONS  
 
===3.1 Diagnostics===
 
IF GO TO Subsection RHR Pump Trip During Midloop Operation.
3.2 RHR Pump Cavitation During Midloop Operation.
3.2 RHR Pump Cavitation During Midloop Operation.
3.3 RCS High Pressure during RHR Shutdown Cooling. 3.4 RHR Pump 1A-A trip.
3.3 RCS High Pressure during RHR Shutdown Cooling. 3.4 RHR Pump 1A-A trip.
Line 551: Line 538:
: 3. OPEN 1-FCV-62-55, EXCESS LETDOWN ISOLATION.  
: 3. OPEN 1-FCV-62-55, EXCESS LETDOWN ISOLATION.  


WBN Unit 1 MALFUNCTION OF PRESSURIZER PRESSURE CONTROL SYSTEM 1-AOI-18 Rev. 0002 Attachment 2 (Page 2 of 2)  
WBN Unit 1 MALFUNCTION OF PRESSURIZER PRESSURE CONTROL SYSTEM 1-AOI-18 Rev. 0002 Attachment 2 (Page 2 of 2) 1.0 PLACING EXCESS LETDOWN INSERVICE (continued)
 
===1.0 PLACING===
EXCESS LETDOWN INSERVICE (continued)
Page 26 of 27 CAUTION Excess letdown design flow is 20 gpm, but can go as high as 50 gpm which could cause high letdown and CCS temperatures.
Page 26 of 27 CAUTION Excess letdown design flow is 20 gpm, but can go as high as 50 gpm which could cause high letdown and CCS temperatures.
: 4. SLOWLY OPEN 1-HIC-62-56A, EXCESS LETDN FLOW CONTROL to maintain Excess LD HX Ou tlet temperature below 195
: 4. SLOWLY OPEN 1-HIC-62-56A, EXCESS LETDN FLOW CONTROL to maintain Excess LD HX Ou tlet temperature below 195
Line 593: Line 577:
Higher    Lower  X  Question Source:
Higher    Lower  X  Question Source:
New  Modified Bank Bank  X  Question History: Bank question 029EG2.4.31  7 Comments:
New  Modified Bank Bank  X  Question History: Bank question 029EG2.4.31  7 Comments:
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 28 of 57  
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 28 of 57 2.4.4 Status Tree Rules of Usage (continued)
 
====2.4.4 Status====
Tree Rules of Usage (continued)
C. Status Trees shall be monito red in the following priority: 1. 1-FR-S, Subcriticality, 2. 1-FR-C, Core Cooling,
C. Status Trees shall be monito red in the following priority: 1. 1-FR-S, Subcriticality, 2. 1-FR-C, Core Cooling,
: 3. 1-FR-H, Heat Sink,
: 3. 1-FR-H, Heat Sink,
Line 717: Line 698:
ACTIONS: o Determine if main steamline isolation and bypass valves are closed o Close appropriate valves INSTRUMENTATION:
ACTIONS: o Determine if main steamline isolation and bypass valves are closed o Close appropriate valves INSTRUMENTATION:
Main steamline isolation and bypass valve indication CONTROL/EQUIPMENT:
Main steamline isolation and bypass valve indication CONTROL/EQUIPMENT:
Main steamline isolation and bypass valve controls E-2 Background HE2BG.doc 31 HP-Rev. 2, 4/30/2005 WBN Unit 1 Main Steam System 1-SOI-1.01 Rev. 0003 Page 16 of 91 Date________    Initials_____  
Main steamline isolation and bypass valve controls E-2 Background HE2BG.doc 31 HP-Rev. 2, 4/30/2005 WBN Unit 1 Main Steam System 1-SOI-1.01 Rev. 0003 Page 16 of 91 Date________    Initials_____
 
5.1 MSIVs and Warming Valves Closed (continued)
===5.1 MSIVs===
and Warming Valves Closed (continued)
CAUTION During initial warming with RCS T avg >300°F, opening one MSIV bypass at a time with a 2-hr wait between each opening is recommended to allow main steam header to drain and to minimize water hammer Main Steam piping Heatup Rate should NOT exceed 100
CAUTION During initial warming with RCS T avg >300°F, opening one MSIV bypass at a time with a 2-hr wait between each opening is recommended to allow main steam header to drain and to minimize water hammer Main Steam piping Heatup Rate should NOT exceed 100
°F/hr and must NOT exceed 200°F/hr. NOTE Warm-up time of up to 16 hrs may be needed to warm steam lines properly, dependent on initial temp and conditions.
°F/hr and must NOT exceed 200°F/hr. NOTE Warm-up time of up to 16 hrs may be needed to warm steam lines properly, dependent on initial temp and conditions.
[12] () WHEN desired to warm the Main Steam Header, THEN PERFORM the following:
[12] () WHEN desired to warm the Main Steam Header, THEN PERFORM the following:
NOMENCLATURE LOCATION POSITION UNID INITIAL WARMING THROTTLE VALVE S/G LOOP 4 SVR/729 (1) 1-THV-1-8  WARMING THROTTLE VALVE S/G LOOP 1 SVR/729 (1) 1-THV-1-18  WARMING THROTTLE VALVE S/G LOOP 2 NVR/729 (1) 1-THV-1-28  WARMING THROTTLE VALVE S/G LOOP 3 NVR/729 (1) 1-THV-1-48  SG 1 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-147  SG 2 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-148  SG 3 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-149  SG 4 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-150  (1) Valves may be open or throttled, as necessary, to maintain steam line warming rate to less than 100°F/Hr WBN Unit 1 Main Steam System 1-SOI-1.01 Rev. 0003 Page 17 of 91 Date________    Initials_____  
NOMENCLATURE LOCATION POSITION UNID INITIAL WARMING THROTTLE VALVE S/G LOOP 4 SVR/729 (1) 1-THV-1-8  WARMING THROTTLE VALVE S/G LOOP 1 SVR/729 (1) 1-THV-1-18  WARMING THROTTLE VALVE S/G LOOP 2 NVR/729 (1) 1-THV-1-28  WARMING THROTTLE VALVE S/G LOOP 3 NVR/729 (1) 1-THV-1-48  SG 1 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-147  SG 2 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-148  SG 3 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-149  SG 4 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-150  (1) Valves may be open or throttled, as necessary, to maintain steam line warming rate to less than 100°F/Hr WBN Unit 1 Main Steam System 1-SOI-1.01 Rev. 0003 Page 17 of 91 Date________    Initials_____
 
5.1 MSIVs and Warming Valves Closed (continued)
===5.1 MSIVs===
and Warming Valves Closed (continued)
NOTE REFER TO Appendix C for MSIVs opening guidelines.
NOTE REFER TO Appendix C for MSIVs opening guidelines.
[13] () WHEN steam lines are fully warm ed, and Criterion A, B, or C of Appendix C is met , THEN OPEN MSIVs (listed below) ONE at a time , AND  ALLOW conditions to stabilize between each opening: NOMENCLATURE LOCATION POSITION UNID INITIAL MSIV SG 1 1-M-4 OPEN 1-FCV-1-4 MSIV SG 2 1-M-4 OPEN 1-FCV-1-11 MSIV SG 3 1-M-4 OPEN 1-FCV-1-22 MSIV SG 4 1-M-4 OPEN 1-FCV-1-29
[13] () WHEN steam lines are fully warm ed, and Criterion A, B, or C of Appendix C is met , THEN OPEN MSIVs (listed below) ONE at a time , AND  ALLOW conditions to stabilize between each opening: NOMENCLATURE LOCATION POSITION UNID INITIAL MSIV SG 1 1-M-4 OPEN 1-FCV-1-4 MSIV SG 2 1-M-4 OPEN 1-FCV-1-11 MSIV SG 3 1-M-4 OPEN 1-FCV-1-22 MSIV SG 4 1-M-4 OPEN 1-FCV-1-29
[14] CLOSE MSIV Bypasses (listed below), AND ENSURE handswitches are placed in CLOSE: NOMENCLATURE LOCATION POSITION UNID INITIALS SG 1 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-147  IVSG 2 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-148  IVSG 3 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-149 IVSG 4 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-150 IV WBN Unit 1 Main Steam System 1-SOI-1.01 Rev. 0003 Page 18 of 91 Date________    Initials_____  
[14] CLOSE MSIV Bypasses (listed below), AND ENSURE handswitches are placed in CLOSE: NOMENCLATURE LOCATION POSITION UNID INITIALS SG 1 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-147  IVSG 2 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-148  IVSG 3 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-149 IVSG 4 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-150 IV WBN Unit 1 Main Steam System 1-SOI-1.01 Rev. 0003 Page 18 of 91 Date________    Initials_____
 
5.1 MSIVs and Warming Valves Closed (continued)
===5.1 MSIVs===
and Warming Valves Closed (continued)
NOTE LCO tracking of OR 14.10.2 may be discont inued when the handswitches in Step 5.1[15] are returned to OFF.
NOTE LCO tracking of OR 14.10.2 may be discont inued when the handswitches in Step 5.1[15] are returned to OFF.
[15] PERFORM the following: NOMENCLATURE LOCATION POSITION UNID INITIALS MAIN STEAM ISOL VLV LOOP 1 BYP WARMING VLV A3U/737 OFF 1-HS-1-147B  IV MAIN STEAM ISOL VLV LOOP 2 BYP WARMING VLV A4U/757 OFF 1-HS-1-148B IV MAIN STEAM ISOL VLV LOOP 3 BYP WARMING VLV A5U/757 OFF 1-HS-1-149B IV MAIN STEAM ISOL VLV LOOP 4 BYP WARMING VLV A3U/737 OFF 1-HS-1-150B IV[16] IF Steam Dumps are available to maintain RCS temperature, THEN REFER TO 1-SOI-1.02, Steam Dump System. ________
[15] PERFORM the following: NOMENCLATURE LOCATION POSITION UNID INITIALS MAIN STEAM ISOL VLV LOOP 1 BYP WARMING VLV A3U/737 OFF 1-HS-1-147B  IV MAIN STEAM ISOL VLV LOOP 2 BYP WARMING VLV A4U/757 OFF 1-HS-1-148B IV MAIN STEAM ISOL VLV LOOP 3 BYP WARMING VLV A5U/757 OFF 1-HS-1-149B IV MAIN STEAM ISOL VLV LOOP 4 BYP WARMING VLV A3U/737 OFF 1-HS-1-150B IV[16] IF Steam Dumps are available to maintain RCS temperature, THEN REFER TO 1-SOI-1.02, Steam Dump System. ________
Line 799: Line 774:
A steam generator pressure limit is set to preclude significant nitrogen injection into the RCS. To determine the steam generator pressure limit, an ideal gas expansion calculation should be performed based on nominal plant specific values for initial accumulator tank pressure (P 1), initial nitrogen gas volume (V 1), and final nitrogen gas volume (Vz). The final nitrogen gas volume should be equivalent to the total accumulator tank volume. The RCS pressure at empty tank conditions (P 2) is determined from: where y = 1.25 for ideal gas expansion.
A steam generator pressure limit is set to preclude significant nitrogen injection into the RCS. To determine the steam generator pressure limit, an ideal gas expansion calculation should be performed based on nominal plant specific values for initial accumulator tank pressure (P 1), initial nitrogen gas volume (V 1), and final nitrogen gas volume (Vz). The final nitrogen gas volume should be equivalent to the total accumulator tank volume. The RCS pressure at empty tank conditions (P 2) is determined from: where y = 1.25 for ideal gas expansion.
The steam generator pressure limit is then determined by subtracting the RCS to SG delta P from P 2* The RCS to SG delta P should be calculated as described in the RCP TRIP/RESTART section in the Generic Issues of the Executive Volume. Instrument uncertainties are not included in the determination of the steam generator pressure limit to preclude a bias toward either having more accumulator water injected into the RCS or having less nitrogen injected into the RCS. ACTIONS: Determine if SG pressures are greater than (0.07) psig INSTRUMENTATION:
The steam generator pressure limit is then determined by subtracting the RCS to SG delta P from P 2* The RCS to SG delta P should be calculated as described in the RCP TRIP/RESTART section in the Generic Issues of the Executive Volume. Instrument uncertainties are not included in the determination of the steam generator pressure limit to preclude a bias toward either having more accumulator water injected into the RCS or having less nitrogen injected into the RCS. ACTIONS: Determine if SG pressures are greater than (0.07) psig INSTRUMENTATION:
SG pressure indication for each SG ECA-0.0 Background HECAOOBG.doc 115 HP-Rev. 2, 4/30/2005 STEP DESCRIPTION TABLE FOR ECA-0.0 Step 16 STEP: Depressurize Intact SGs To (0.08) PSIG PURPOSE: To depressurize the intact steam generators BASIS: Step 16 depressurizes the intact SGs, thereby reducing RCS temperature and pressure to reduce RCP seal leakage and minimize RCS inventory loss. The advantages to performing this action, as well as restrictions that apply during the action, are detailed in Subsection  
SG pressure indication for each SG ECA-0.0 Background HECAOOBG.doc 115 HP-Rev. 2, 4/30/2005 STEP DESCRIPTION TABLE FOR ECA-0.0 Step 16 STEP: Depressurize Intact SGs To (0.08) PSIG PURPOSE: To depressurize the intact steam generators BASIS: Step 16 depressurizes the intact SGs, thereby reducing RCS temperature and pressure to reduce RCP seal leakage and minimize RCS inventory loss. The advantages to performing this action, as well as restrictions that apply during the action, are detailed in Subsection 2.3. During SG depressurization, SG level must be maintained above the top of the SG U-tubes in at least one SG. Maintaining the U-tubes covered in at least one SG will ensure that sufficient heat transfer capability exists to remove heat from the RCS via either natural circulation or reflux boiling after the RCS saturates.
 
===2.3. During===
SG depressurization, SG level must be maintained above the top of the SG U-tubes in at least one SG. Maintaining the U-tubes covered in at least one SG will ensure that sufficient heat transfer capability exists to remove heat from the RCS via either natural circulation or reflux boiling after the RCS saturates.
Step 16a requires that SG level be in the narrow range in at least one SG before SG depressurization is initiated in Step 16b. If level is not in the narrow range in at least one SG, RNO 16a instructs the operator to maintain maximum AFW flow until narrow range level is established in one SG. When narrow range level is established, SG depressurization can be started or continued via Step 16b. Step 16b instructs the operator to reduce SG pressures by depressurizing the intact SGs. Depressurization should be accomplished by opening the PORVs on the intact SGs to establish a maximum steam dump rate, consistent with plant specific constraints.
Step 16a requires that SG level be in the narrow range in at least one SG before SG depressurization is initiated in Step 16b. If level is not in the narrow range in at least one SG, RNO 16a instructs the operator to maintain maximum AFW flow until narrow range level is established in one SG. When narrow range level is established, SG depressurization can be started or continued via Step 16b. Step 16b instructs the operator to reduce SG pressures by depressurizing the intact SGs. Depressurization should be accomplished by opening the PORVs on the intact SGs to establish a maximum steam dump rate, consistent with plant specific constraints.
The step is structured assuming that the operator can open and control SG PORVs from the control room. This structure assumes that the PORVs are air-operated and have de control power and pneumatic power (i.e., either air reservoirs or nitrogen bottles) available.
The step is structured assuming that the operator can open and control SG PORVs from the control room. This structure assumes that the PORVs are air-operated and have de control power and pneumatic power (i.e., either air reservoirs or nitrogen bottles) available.
Line 1,189: Line 1,161:
AFW OPERATION
AFW OPERATION
* IF CST volume less than 5000 gal, THEN  MONITOR AFW pumps to ensure suction transfer.
* IF CST volume less than 5000 gal, THEN  MONITOR AFW pumps to ensure suction transfer.
WBN Unit 1 Loss of Secondary Heat Sink 1-FR-H.1 Rev. 0003 Step    Action/Expected Response Response Not Obtained Page 3 of 37  
WBN Unit 1 Loss of Secondary Heat Sink 1-FR-H.1 Rev. 0003 Step    Action/Expected Response Response Not Obtained Page 3 of 37 3.0 OPERATOR ACTIONS CAUTION
 
===3.0 OPERATOR===
ACTIONS CAUTION
* If total feed flow CAPABILITY of 410 gpm is available, this Instruction should NOT be performed.
* If total feed flow CAPABILITY of 410 gpm is available, this Instruction should NOT be performed.
* If an Intact S/G is avail able, feed flow should NOT be reestablished to any faulted S/G.
* If an Intact S/G is avail able, feed flow should NOT be reestablished to any faulted S/G.
Line 1,355: Line 1,324:
Higher    Lower  X Question Source:
Higher    Lower  X Question Source:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
Rod Group Alignment Limits
Rod Group Alignment Limits 3.1.5       Watts Bar-Unit 1 3.1-8          3.1  REACTIVITY CONTROL SYSTEMS  
 
====3.1.5 Watts====
Bar-Unit 1 3.1-8          3.1  REACTIVITY CONTROL SYSTEMS  


3.1.5  Rod Group Alignment Limits  
3.1.5  Rod Group Alignment Limits  
Line 1,409: Line 1,375:
5.2.4.4 "Any Rod On Bottom", "All Rods On Bottom", "Rod-to-Bank Deviation'', to-Rod Deviation", "Rod Wrong Direction", and "RPI System Trouble" alarm signals are displayed on this screen. 5.2.4.5 RPI System Trouble indicator has touch button capability and once it is touched, screen navigates to System Status Display. 5.2.4.6 Rod Demand from Passive Summer and Rod Speed indicators are provided in bar graph format. 5.2.4. 7 Under normal operating conditions, either PLC A or PLC B can be selected to provide the information on this screen. If the main PLC (PLC A) is not available, then the information on this screen will be supplied by PLC B automatically.
5.2.4.4 "Any Rod On Bottom", "All Rods On Bottom", "Rod-to-Bank Deviation'', to-Rod Deviation", "Rod Wrong Direction", and "RPI System Trouble" alarm signals are displayed on this screen. 5.2.4.5 RPI System Trouble indicator has touch button capability and once it is touched, screen navigates to System Status Display. 5.2.4.6 Rod Demand from Passive Summer and Rod Speed indicators are provided in bar graph format. 5.2.4. 7 Under normal operating conditions, either PLC A or PLC B can be selected to provide the information on this screen. If the main PLC (PLC A) is not available, then the information on this screen will be supplied by PLC B automatically.
WNA-CT-00053-WAT, Rev. 1 5-8 Westinghouse Proprietary Class 2 * *
WNA-CT-00053-WAT, Rev. 1 5-8 Westinghouse Proprietary Class 2 * *
* Watts Bar Unit 1 ARPI System Upgrade CERPI Technical Description  
* Watts Bar Unit 1 ARPI System Upgrade CERPI Technical Description 5.2.5 Shutdown Banks A, B, C, and D Rod Positions Display 5.2.5.1 This display screen shows the positions of the rods in Shutdown Banks A, B, C, and D in a bar graph format 5.2.5.2 The ability to identify a rod position value affected by a faulty analog input is provided.
 
====5.2.5 Shutdown====
Banks A, B, C, and D Rod Positions Display 5.2.5.1 This display screen shows the positions of the rods in Shutdown Banks A, B, C, and D in a bar graph format 5.2.5.2 The ability to identify a rod position value affected by a faulty analog input is provided.
5.2.5.3 An indication for the rod bottom position of each rod is provided.
5.2.5.3 An indication for the rod bottom position of each rod is provided.
5.2.5.4 "Any Rod On Bottom", "All Rods On Bottom'', "Rod-to-Bank Deviation'', to-Rod Deviation", "Rod Wrong Direction'', and "RPI System Trouble" alarm signals are displayed on this screen. 5.2.5.5 RPI System Trouble indicator has touch button capability and once it is touched, screen navigates to System Status Display. 5.2.5.6 Rod Demand from Passive Summer and Rod Speed indicators are provided in bar graph format. 5.2.5. 7 Under normal operating conditions, either PLC A or PLC B can be selected to provide the information on this screen. If the main PLC (PLC A} is not available, then the information on this screen will be supplied by PLC B automatically.
5.2.5.4 "Any Rod On Bottom", "All Rods On Bottom'', "Rod-to-Bank Deviation'', to-Rod Deviation", "Rod Wrong Direction'', and "RPI System Trouble" alarm signals are displayed on this screen. 5.2.5.5 RPI System Trouble indicator has touch button capability and once it is touched, screen navigates to System Status Display. 5.2.5.6 Rod Demand from Passive Summer and Rod Speed indicators are provided in bar graph format. 5.2.5. 7 Under normal operating conditions, either PLC A or PLC B can be selected to provide the information on this screen. If the main PLC (PLC A} is not available, then the information on this screen will be supplied by PLC B automatically.
Line 1,509: Line 1,472:
Higher X  Lower    Question Source:
Higher X  Lower    Question Source:
New X  Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:
WBN Unit 1 Nuclear Instrumentation Malfunctions 1-AOI-4 Rev. 0001  
WBN Unit 1 Nuclear Instrumentation Malfunctions 1-AOI-4 Rev. 0001 3.2 Source Range Monitor (SRM) Failure (continued)
 
===3.2 Source===
Range Monitor (SRM) Failure (continued)
Page 9 of 42 Step Action/Expected Response Response Not Obtained
Page 9 of 42 Step Action/Expected Response Response Not Obtained
: 9. IF in MODE 3, 4, or 5, THEN:
: 9. IF in MODE 3, 4, or 5, THEN:
Line 1,575: Line 1,535:
Higher  X  Lower    Question Source:
Higher  X  Lower    Question Source:
New X  Modified Bank Bank    Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank    Question History: New question for the 2015-301 NRC RO Exam Comments:
WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-II or 2-II 1-AOI-25.02 Rev. 0003  
WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-II or 2-II 1-AOI-25.02 Rev. 0003 2.1 Alarms (continued)
 
===2.1 Alarms===
(continued)
Page 4 of 24 T. SSPS-B GEN WARNING [115-A] U. CL ACCUM 1 LEVEL HI/LO [131-A]
Page 4 of 24 T. SSPS-B GEN WARNING [115-A] U. CL ACCUM 1 LEVEL HI/LO [131-A]
V. CL ACCUM 1 PRESS HI/LO [131-B]
V. CL ACCUM 1 PRESS HI/LO [131-B]
W. CL ACCUM 2 LEVEL HI/LO [132-A]
W. CL ACCUM 2 LEVEL HI/LO [132-A]
X. CL ACCUM 2 PRESS HI/LO [132-B]  
X. CL ACCUM 2 PRESS HI/LO [132-B]
 
2.2 Indications A. Power Range 1-N-42, Intermediate Range 1-N-136, and Source Range 1-N-132 failure. B. All channel II trip STATUS LIGHTS will ENERGIZE (Except Cntmt Level, Cntmt Hi-Hi Press And RWST Low Level). C. Train B CNTMT Isolation Panel 1-XX-55-6F will be DARK. D. Accumulator 3 & 4 Pressure, fails low 2.3 Automatic Actions A. Rod withdrawal block due to high flux rod stop.
===2.2 Indications===
A. Power Range 1-N-42, Intermediate Range 1-N-136, and Source Range 1-N-132 failure. B. All channel II trip STATUS LIGHTS will ENERGIZE (Except Cntmt Level, Cntmt Hi-Hi Press And RWST Low Level). C. Train B CNTMT Isolation Panel 1-XX-55-6F will be DARK. D. Accumulator 3 & 4 Pressure, fails low 2.3 Automatic Actions A. Rod withdrawal block due to high flux rod stop.
B. Charging pump suction swaps over to RWST due to de-energized separation relays. 1-LCV-62-135 and 1-LCV-62-136 open and 1-LCV-62-133 closes, which causes borated water to be injected into the RCS. C. If in service, Train A MCR, EBR, and SD BR Chillers will swap to Train B, and auto start of Train A is disabled.
B. Charging pump suction swaps over to RWST due to de-energized separation relays. 1-LCV-62-135 and 1-LCV-62-136 open and 1-LCV-62-133 closes, which causes borated water to be injected into the RCS. C. If in service, Train A MCR, EBR, and SD BR Chillers will swap to Train B, and auto start of Train A is disabled.
WBN Unit 1 Bypass, Intlk, & Pe rmissive 1-ARI-64-70 Rev. 0000 Page 8 of 47 NOTE 1 Relay K1705 is actuated when Reactor power on 2/4 PR channels is greater than or equal to10%. A. Power escalation above 10%
WBN Unit 1 Bypass, Intlk, & Pe rmissive 1-ARI-64-70 Rev. 0000 Page 8 of 47 NOTE 1 Relay K1705 is actuated when Reactor power on 2/4 PR channels is greater than or equal to10%. A. Power escalation above 10%
Line 1,615: Line 1,570:
==References:==
==References:==


WBN Unit 1 Nuclear Instrumentation Malfunctions 1-AOI-4 Rev. 0001  
WBN Unit 1 Nuclear Instrumentation Malfunctions 1-AOI-4 Rev. 0001 2.2 Indications (continued)
 
===2.2 Indications===
(continued)
Page 4 of 42 B. Intermediate Range Monitor (IRM) malfunction: 1. Erratic or loss of indication. 2. INSTRUMENT POWER ON and/or CO NTROL POWER ON lights at NIS racks DARK [1-M-13]. 3. Both startup rate channels NOT indicating same startup rate. 4. SRM does NOT energize during shutdown. 5. NON-OPERATE light LIT [1-M-13]. C. Power Range Monitor (PRM) malfunction: 1. Erratic or loss of indication.
Page 4 of 42 B. Intermediate Range Monitor (IRM) malfunction: 1. Erratic or loss of indication. 2. INSTRUMENT POWER ON and/or CO NTROL POWER ON lights at NIS racks DARK [1-M-13]. 3. Both startup rate channels NOT indicating same startup rate. 4. SRM does NOT energize during shutdown. 5. NON-OPERATE light LIT [1-M-13]. C. Power Range Monitor (PRM) malfunction: 1. Erratic or loss of indication.
: 2. Delta flux meter failed high, low, or giving erratic indications. 3. CONTROL POWER ON and/or INSTRUMENT POWER ON lights at NIS racks DARK [1-M-13]. 4. DCS Operator display for Rod Control and PR NIS indicates abnormal condition for applicable PRM. 2.3 Automatic Actions A. Source Range Monitor failure: 1. Possible source range monitor high flux Rx trip at 10 5 cps if SRM NOT blocked. 2. Possible Cntmt evacuation ala rm at 0.5 decade above background. B. Intermediate Range Monitor failure: 1. Intermediate range monitor 20% high flux rod stop (if NOT blocked above P-10). 2. Intermediate range monitor 25% high flux Rx trip (if NOT blocked above P-10). 3. Possible loss of P-6 block if intermediate range monitor greater than 1.66 X 10-4% power and intermediate range monitor fails low.
: 2. Delta flux meter failed high, low, or giving erratic indications. 3. CONTROL POWER ON and/or INSTRUMENT POWER ON lights at NIS racks DARK [1-M-13]. 4. DCS Operator display for Rod Control and PR NIS indicates abnormal condition for applicable PRM. 2.3 Automatic Actions A. Source Range Monitor failure: 1. Possible source range monitor high flux Rx trip at 10 5 cps if SRM NOT blocked. 2. Possible Cntmt evacuation ala rm at 0.5 decade above background. B. Intermediate Range Monitor failure: 1. Intermediate range monitor 20% high flux rod stop (if NOT blocked above P-10). 2. Intermediate range monitor 25% high flux Rx trip (if NOT blocked above P-10). 3. Possible loss of P-6 block if intermediate range monitor greater than 1.66 X 10-4% power and intermediate range monitor fails low.
Line 1,668: Line 1,620:
* tA A FI E PROC::R oii hl.MED OU T. THIS OO NT R.OL OA 4WI N C S U P'E:RSLOES UN ir 1 oliS*CONSTR UC TCD ORA IINC '4 716 1 0-'J0-5 REVI S IO N 1-1. 4 1 526 3 0 C J O R EVISE D PER DCA '2630-144-Z.
* tA A FI E PROC::R oii hl.MED OU T. THIS OO NT R.OL OA 4WI N C S U P'E:RSLOES UN ir 1 oliS*CONSTR UC TCD ORA IINC '4 716 1 0-'J0-5 REVI S IO N 1-1. 4 1 526 3 0 C J O R EVISE D PER DCA '2630-144-Z.
RE V CHA N GE REF P RE PARER C H£ C K ER A PPROVED DAT E SCAL£' N T S £XC£P T AS NO T E D P ROJ ECT F' A CIL I TY POW E RHO U SE U NITS 1 & 2 T ITLE ELECTRICAL CONTROL DIAGRAM RADIATION MONITORING SYSTEM W ATTS BA R N U C LEA R PL A NT TENNESSEE VA L L EY a L lil e'.-9 0*4 0+ l CO ND ENSATE V ACUU M VENTPOsrACCi'CENT MOHITORI NC UN!T 1 rM!tIDUPtNRAA
RE V CHA N GE REF P RE PARER C H£ C K ER A PPROVED DAT E SCAL£' N T S £XC£P T AS NO T E D P ROJ ECT F' A CIL I TY POW E RHO U SE U NITS 1 & 2 T ITLE ELECTRICAL CONTROL DIAGRAM RADIATION MONITORING SYSTEM W ATTS BA R N U C LEA R PL A NT TENNESSEE VA L L EY a L lil e'.-9 0*4 0+ l CO ND ENSATE V ACUU M VENTPOsrACCi'CENT MOHITORI NC UN!T 1 rM!tIDUPtNRAA
* D E SIGN !N!T!A L IS SUE ENC I NE ER INC APP ROV A L 5 6 7 8 9 CAD OlfGS: 1-<H W IU0-90-1 THFI U 4 CSSt.IF;O BY: CH EC K LR C..W. JCHNSON J.IJ. MOYU:S OE S I CNCR R EV'[ EWER t-1.L. Cl-IAPllU.N R.L. F'ORESfER RO I SS U E P ER W BEP 5. 1 7 .._RIMS 82 6 ' 90 0928 379 OCN P*OJ398*B OCN P-03 403-A Cl.P. PE R"-E'.R T. ONA L J C .C. L F O" MC.Ill WBN Unit 1 Abnormal Release of Radioactive Material 1-AOI-31 Rev. 0001 Page 3 of 27  
* D E SIGN !N!T!A L IS SUE ENC I NE ER INC APP ROV A L 5 6 7 8 9 CAD OlfGS: 1-<H W IU0-90-1 THFI U 4 CSSt.IF;O BY: CH EC K LR C..W. JCHNSON J.IJ. MOYU:S OE S I CNCR R EV'[ EWER t-1.L. Cl-IAPllU.N R.L. F'ORESfER RO I SS U E P ER W BEP 5. 1 7 .._RIMS 82 6 ' 90 0928 379 OCN P*OJ398*B OCN P-03 403-A Cl.P. PE R"-E'.R T. ONA L J C .C. L F O" MC.Ill WBN Unit 1 Abnormal Release of Radioactive Material 1-AOI-31 Rev. 0001 Page 3 of 27 1.0 PURPOSE This instruction provides a response to valid annunciations or indications of an abnormal release of radioactive material.
 
2.0 SYMPTOMS 2.1 Alarms NOTE SHLD BLDG VENT RE-400 RAD HI [268-B] in alarm should be accompanied by higher indicated radiation levels or alarm on additional monitor(s) which indicates the source of the release. A. Auxiliary Building: 1. 1-RR-90-1 AREA RAH HI [174-B] 2. ERCW DISCH HDR A 0-R M-133/140 LIQ RAD HI [180-B] 3. WDS RELEASE LINE 0-R M-122 LIQ RAD HI [181-A]
===1.0 PURPOSE===
This instruction provides a response to valid annunciations or indications of an abnormal release of radioactive material.
 
===2.0 SYMPTOMS===
 
===2.1 Alarms===
 
NOTE SHLD BLDG VENT RE-400 RAD HI [268-B] in alarm should be accompanied by higher indicated radiation levels or alarm on additional monitor(s) which indicates the source of the release. A. Auxiliary Building: 1. 1-RR-90-1 AREA RAH HI [174-B] 2. ERCW DISCH HDR A 0-R M-133/140 LIQ RAD HI [180-B] 3. WDS RELEASE LINE 0-R M-122 LIQ RAD HI [181-A]
: 4. ERCW DISCH HDR B 0-R M-134/141 LIQ RAD HI [181-B] 5. AB VENT 0-RM-101 RAD HI [183-C]
: 4. ERCW DISCH HDR B 0-R M-134/141 LIQ RAD HI [181-B] 5. AB VENT 0-RM-101 RAD HI [183-C]
: 6. WGDT REL LINE 0-RM-118 RAD HI [184-A] 7. SFP 0-RM-102/103 RAD HI [184-B]
: 6. WGDT REL LINE 0-RM-118 RAD HI [184-A] 7. SFP 0-RM-102/103 RAD HI [184-B]
Line 1,761: Line 1,705:
Higher  Lower  X Question Source:
Higher  Lower  X Question Source:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
WBN Design Criteria Document RADIATION MONITORING (UNIT 1 / UNIT 2) WBN-DCD-40-24 Rev. 0026 Page 26 of 89  
WBN Design Criteria Document RADIATION MONITORING (UNIT 1 / UNIT 2) WBN-DCD-40-24 Rev. 0026 Page 26 of 89 1.0 SCOPE This document establishes the design requirements of the Radiation Monitoring System (RMS) for the operation of Watts Bar Units 1 and 2 [Source Note 7]. The RMS provides continuous monitoring of radiation levels, provides the results of the monitoring to the operator and in some instances produces signals to initiate automatic control actions. In addition, the RMS includes equipment for off-line tritium, particulate and iodine sample collection with no real-time detection. The RMS consists of process, airborne, area, and effluent radioactivity monitors. The outdoor radiological monitoring instrumentation, laboratory radiation measurement equipment, criticality monitoring equipment, and portable survey instrumentation is outside the scope of this criteria. If a discrepancy exists between this design criteria and any other Nuclear Engineering (NE) design criteria, the appropriate Engineering Manager should be notified by a memorandum. If a discrepancy exists between this design criteria and any other document where the other document is not a NE design criteria, this design criteria shall govern.
 
2.0 DEFINITIONS Accuracy - The degree of agreement with the true value of the quantity being measured. Accessible Area - An area which is designed to be entered without special radiological controls (i.e., is not a high radiation area or airborne radioactivity area), dismantling of installed equipment or the introduction of support equipment such as staging, ladders, etc. Anticipated Operational Occurrences - Those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to the loss of power to all reactor coolant recirculation pumps, a trip of the turbine generator set, isolation of the main condenser and a loss of all off-site power. Categories 1, 2, and 3 - References to Categories 1, 2, and 3 are as stated in RG 1.97 (Reference 8.1.11). Channel - A channel includes all of the instrumentation necessary to provide the monitoring function. A channel includes associated sample lines, skid assemblies, and local readout and alarm devices. A channel also includes equipment in the control room for recording, readout, and alarm and includes the electrical cable that connects monitor components. Check Source - A test source supplied as an integral part of the monitor channel for use in determining if the channel is functional. This may be a radioactive source, or when a check of the channel exclusive of the detector scintillator is appropriate (i.e., where increased background cannot be tolerated), a pulsed light source or other similar device appropriate for the type of detector. Class 1E - The safety-related classification of the electrical equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling and containment and reactor heat removal, or otherwise are essential in preventing significant release of radioactive material to the environment. Grab Sample - A volume removed from a process or effluent for laboratory analysis.  
===1.0 SCOPE===
This document establishes the design requirements of the Radiation Monitoring System (RMS) for the operation of Watts Bar Units 1 and 2 [Source Note 7]. The RMS provides continuous monitoring of radiation levels, provides the results of the monitoring to the operator and in some instances produces signals to initiate automatic control actions. In addition, the RMS includes equipment for off-line tritium, particulate and iodine sample collection with no real-time detection. The RMS consists of process, airborne, area, and effluent radioactivity monitors. The outdoor radiological monitoring instrumentation, laboratory radiation measurement equipment, criticality monitoring equipment, and portable survey instrumentation is outside the scope of this criteria. If a discrepancy exists between this design criteria and any other Nuclear Engineering (NE) design criteria, the appropriate Engineering Manager should be notified by a memorandum. If a discrepancy exists between this design criteria and any other document where the other document is not a NE design criteria, this design criteria shall govern.  
 
===2.0 DEFINITIONS===
 
Accuracy - The degree of agreement with the true value of the quantity being measured. Accessible Area - An area which is designed to be entered without special radiological controls (i.e., is not a high radiation area or airborne radioactivity area), dismantling of installed equipment or the introduction of support equipment such as staging, ladders, etc. Anticipated Operational Occurrences - Those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to the loss of power to all reactor coolant recirculation pumps, a trip of the turbine generator set, isolation of the main condenser and a loss of all off-site power. Categories 1, 2, and 3 - References to Categories 1, 2, and 3 are as stated in RG 1.97 (Reference 8.1.11). Channel - A channel includes all of the instrumentation necessary to provide the monitoring function. A channel includes associated sample lines, skid assemblies, and local readout and alarm devices. A channel also includes equipment in the control room for recording, readout, and alarm and includes the electrical cable that connects monitor components. Check Source - A test source supplied as an integral part of the monitor channel for use in determining if the channel is functional. This may be a radioactive source, or when a check of the channel exclusive of the detector scintillator is appropriate (i.e., where increased background cannot be tolerated), a pulsed light source or other similar device appropriate for the type of detector. Class 1E - The safety-related classification of the electrical equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling and containment and reactor heat removal, or otherwise are essential in preventing significant release of radioactive material to the environment. Grab Sample - A volume removed from a process or effluent for laboratory analysis.  


Gross beta or gamma radioactivity - The combined radioactivity from the mix of nuclides without distinction as to energy levels of the emissions.
Gross beta or gamma radioactivity - The combined radioactivity from the mix of nuclides without distinction as to energy levels of the emissions.
Line 1,849: Line 1,787:
Special care must be taken to locate the compressor in areas free of dust and contaminants. At least (1) one hour air bottle is available for each self-contained breathing unit. An additional (16) one hour bottles are available onsite to provide a minimum 16-hour supply of reserve air for the self-contained breathing units provided for use by the fire brigade, damage control, and control room personnel.  
Special care must be taken to locate the compressor in areas free of dust and contaminants. At least (1) one hour air bottle is available for each self-contained breathing unit. An additional (16) one hour bottles are available onsite to provide a minimum 16-hour supply of reserve air for the self-contained breathing units provided for use by the fire brigade, damage control, and control room personnel.  


WBN System Description Document COMPRESSED AIR SYST EM WBN-SDD-N3-32-4002 Rev. 0010 Page 31 of 71  
WBN System Description Document COMPRESSED AIR SYST EM WBN-SDD-N3-32-4002 Rev. 0010 Page 31 of 71 3.1.1 General (continued)
 
====3.1.1 General====
(continued)
The C&SA compressors discharge into two r edundant headers provided with manual valves. These headers feed the two CAS receivers which in turn supply air through redundant headers to the CAS dryers. The dryers contain three trains of prefilters, dryers, and afterfilters. Each dryer train is sized to handle CAS requirements for one unit. Manual bypasses are provided around each filter assembly for abnormal operation. The air is then piped to pneumatic equipment throughout the plant.
The C&SA compressors discharge into two r edundant headers provided with manual valves. These headers feed the two CAS receivers which in turn supply air through redundant headers to the CAS dryers. The dryers contain three trains of prefilters, dryers, and afterfilters. Each dryer train is sized to handle CAS requirements for one unit. Manual bypasses are provided around each filter assembly for abnormal operation. The air is then piped to pneumatic equipment throughout the plant.
Air is supplied to the SAS receiver by a header from the CAS receivers. Service air is supplied through a valve which closes if CAS pressure drops below a minimum value of 80 psig to assure that CAS requirements take precedence over the SAS. SAS air is piped from the receiver to service outlets and miscellaneous equipment throughout the plant. Like CAS air, the SAS is delivered oil-free. SAS air is not required to be dried and filtered. A 4" permanent piping segment with a "Tee", an isolation valve and connecting flange is provided to the existing service air header for emergency or future compressed air usage.
Air is supplied to the SAS receiver by a header from the CAS receivers. Service air is supplied through a valve which closes if CAS pressure drops below a minimum value of 80 psig to assure that CAS requirements take precedence over the SAS. SAS air is piped from the receiver to service outlets and miscellaneous equipment throughout the plant. Like CAS air, the SAS is delivered oil-free. SAS air is not required to be dried and filtered. A 4" permanent piping segment with a "Tee", an isolation valve and connecting flange is provided to the existing service air header for emergency or future compressed air usage.
Line 1,912: Line 1,847:
Higher X  Lower    Question Source:
Higher X  Lower    Question Source:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
WBN Unit 1 Loss of Secondary Heat Sink 1-FR-H.1 Rev. 0003 Step    Action/Expected Response Response Not Obtained Page 3 of 37  
WBN Unit 1 Loss of Secondary Heat Sink 1-FR-H.1 Rev. 0003 Step    Action/Expected Response Response Not Obtained Page 3 of 37 3.0 OPERATOR ACTIONS CAUTION
 
===3.0 OPERATOR===
ACTIONS CAUTION
* If total feed flow CAPABILITY of 410 gpm is available, this Instruction should NOT be performed.
* If total feed flow CAPABILITY of 410 gpm is available, this Instruction should NOT be performed.
* If an Intact S/G is avail able, feed flow should NOT be reestablished to any faulted S/G.
* If an Intact S/G is avail able, feed flow should NOT be reestablished to any faulted S/G.
Line 2,087: Line 2,019:


(!) z -=-< c:::: 0 0 w z -< I-z < -0 ..I < < a: z u w-:I 00 D. z a: u. <o wZ ..I 0 o-:> .... a: u. -0 2 " 6 TIME (SECONDS) a 10 WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT NUCLEAR POWER TRANSIENT FOUR PUMPS IN OPERATION ONE PUMP COASTING DOWN FIGURE 15.2-16
(!) z -=-< c:::: 0 0 w z -< I-z < -0 ..I < < a: z u w-:I 00 D. z a: u. <o wZ ..I 0 o-:> .... a: u. -0 2 " 6 TIME (SECONDS) a 10 WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT NUCLEAR POWER TRANSIENT FOUR PUMPS IN OPERATION ONE PUMP COASTING DOWN FIGURE 15.2-16
(.!) z .... * < 0:: Q Q ..... z .... < I-z .... < ::::E Q 3 < u 2.5 0: 2 0 1.5 1 0 2 4 6 TIME (SECONDS) 8 10 WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT DNBR VERSUS TIME FOUR PUMPS IN OPERATION ONE PUMP COASTING DOWN FIGURE 15.2-17 WBN System Description Document REACTOR COOLANT SYSTEM WBN-SDD-N3-68-4001 Rev. 0034 Page 71 of 257  
(.!) z .... * < 0:: Q Q ..... z .... < I-z .... < ::::E Q 3 < u 2.5 0: 2 0 1.5 1 0 2 4 6 TIME (SECONDS) 8 10 WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT DNBR VERSUS TIME FOUR PUMPS IN OPERATION ONE PUMP COASTING DOWN FIGURE 15.2-17 WBN System Description Document REACTOR COOLANT SYSTEM WBN-SDD-N3-68-4001 Rev. 0034 Page 71 of 257 3.2.3 Reactor Coolant Pumps (continued)
 
====3.2.3 Reactor====
Coolant Pumps (continued)
: 8. Flywheel The flywheel stores energy to keep the pump operating for a short but critical period following a power failure. It provides additional inertia to extend pump coast-down. The flywheel is available for inspection by removing the cover. D. Oil Lift System An external oil lift system provides lubrication to the upper radial bearings during motor startup. This oil lift system is shown in Figure 2. High-pressure oil is also delivered simultaneously to the upper and lower sets of thrust bearing shoes. The oil lift system will provide from 0.001" to 0.003" of oil film between the runner and the thrust bearing shoes depending on the amount of pump thrust on the bearings. Thus, regardless of thrust condition (either upthrust or downthrust), an oil film between the shoes and runner is present. At the same time, high-pressure oil is being sprayed into the guide bearing chamber for lubricating the upper radial guide bearings. There are low and high oil level alarms in the upper and lower oil reservoirs. The oil lift pumps have spray shields to ensure no lube oil comes in contact with hot reactor coolant in case of component rupture. The oil lift system is required to operate approximately two minutes before RCP startup and approximately one minute after the pump reaches full speed. E. RCP Power The RCPs are powered through the 6.9KV RCP boards. Undervoltage to any two of the four pumps will cause a reactor trip above approximately 10% reactor power. Underfrequency to any two of the four pumps will trip all four RCPs and cause a reactor trip above approximately 10% power. 3.2.4 Pressurizer (PZR)
: 8. Flywheel The flywheel stores energy to keep the pump operating for a short but critical period following a power failure. It provides additional inertia to extend pump coast-down. The flywheel is available for inspection by removing the cover. D. Oil Lift System An external oil lift system provides lubrication to the upper radial bearings during motor startup. This oil lift system is shown in Figure 2. High-pressure oil is also delivered simultaneously to the upper and lower sets of thrust bearing shoes. The oil lift system will provide from 0.001" to 0.003" of oil film between the runner and the thrust bearing shoes depending on the amount of pump thrust on the bearings. Thus, regardless of thrust condition (either upthrust or downthrust), an oil film between the shoes and runner is present. At the same time, high-pressure oil is being sprayed into the guide bearing chamber for lubricating the upper radial guide bearings. There are low and high oil level alarms in the upper and lower oil reservoirs. The oil lift pumps have spray shields to ensure no lube oil comes in contact with hot reactor coolant in case of component rupture. The oil lift system is required to operate approximately two minutes before RCP startup and approximately one minute after the pump reaches full speed. E. RCP Power The RCPs are powered through the 6.9KV RCP boards. Undervoltage to any two of the four pumps will cause a reactor trip above approximately 10% reactor power. Underfrequency to any two of the four pumps will trip all four RCPs and cause a reactor trip above approximately 10% power. 3.2.4 Pressurizer (PZR)
The PZR provides a point in the RCS where liquid and vapor can be maintained in equilibrium under saturated conditions for RCS pressure control. The PZR and its sprays, heaters, PORVs, safety valves, and surge line make up the pressure control equipment.
The PZR provides a point in the RCS where liquid and vapor can be maintained in equilibrium under saturated conditions for RCS pressure control. The PZR and its sprays, heaters, PORVs, safety valves, and surge line make up the pressure control equipment.
Line 2,187: Line 2,116:
Higher    Lower  X  Question Source:
Higher    Lower  X  Question Source:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
WBN System Description Document CHEMICAL AND VOLUME CONTROL SYSTEM WBN-SDD-N3-62-4001 Rev. 0035 Page 76 of 176  
WBN System Description Document CHEMICAL AND VOLUME CONTROL SYSTEM WBN-SDD-N3-62-4001 Rev. 0035 Page 76 of 176 3.2.7 Boric Acid Blender (continued)
 
====3.2.7 Boric====
Acid Blender (continued)
The blenders were manufactured by W. They are austenitic stainless steel and consists of a conventional pipe tee. The blenders are provided with water inlet, boric acid inlet, and discharge connections. The boric acid blenders (W Spin No. CSHDBL) were purchased from W under contract No. 71C60-54114-1. 3.2.8 Orifices A. Boric Acid Transfer Pump Bypass Orifices There are three pump bypass orifices (one per BAT) located in the auxiliary building. Each orifice is designed to pass the minimum flow required for sufficient recirculation through the piping and tanks with the boric acid transfer pumps. The design operation requirements are listed in Table 11. The orifices normally operate at 100 psig and  
The blenders were manufactured by W. They are austenitic stainless steel and consists of a conventional pipe tee. The blenders are provided with water inlet, boric acid inlet, and discharge connections. The boric acid blenders (W Spin No. CSHDBL) were purchased from W under contract No. 71C60-54114-1. 3.2.8 Orifices A. Boric Acid Transfer Pump Bypass Orifices There are three pump bypass orifices (one per BAT) located in the auxiliary building. Each orifice is designed to pass the minimum flow required for sufficient recirculation through the piping and tanks with the boric acid transfer pumps. The design operation requirements are listed in Table 11. The orifices normally operate at 100 psig and  


Line 2,198: Line 2,124:
R.E. McSpadden Date: RF012 SCALING AND SETPOINT DOCUMENT LOOP DRAWING Loop Name: Boric Acid Tank A Recirc Flow  
R.E. McSpadden Date: RF012 SCALING AND SETPOINT DOCUMENT LOOP DRAWING Loop Name: Boric Acid Tank A Recirc Flow  


WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 23 of 124 Date________    Initials  
WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 23 of 124 Date________    Initials 6.3 Transfer BA from BAT C To BAT A NOTE This Section is for BAT A in service to U1 via BA Pump 1A.
 
===6.3 Transfer===
BA from BAT C To BAT A NOTE This Section is for BAT A in service to U1 via BA Pump 1A.
[1] IF BAT A in service to U1, THEN [1.1] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to STOP and RELEASE. ________
[1] IF BAT A in service to U1, THEN [1.1] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to STOP and RELEASE. ________
[1.2] CHECK 1-HS-62-140A, Green light LIT. ________
[1.2] CHECK 1-HS-62-140A, Green light LIT. ________
Line 2,215: Line 2,138:
________ [9] THROTTLE 1-FCV-62-237, BAT A RECIRC CONTROL [1-M-6], as desired to control rate of transfer.
________ [9] THROTTLE 1-FCV-62-237, BAT A RECIRC CONTROL [1-M-6], as desired to control rate of transfer.
________
________
WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 24 of 124 Date________    Initials  
WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 24 of 124 Date________    Initials 6.3 Transfer BA from BAT C To BAT A (continued)
 
===6.3 Transfer===
BA from BAT C To BAT A (continued)
CAUTION Damage may occur to BAT Heaters if leve l drops below 9.5% with heaters ENERGIZED.
CAUTION Damage may occur to BAT Heaters if leve l drops below 9.5% with heaters ENERGIZED.
[10] IF BAT C level decreases to 2270 gals (10%), THEN ENSURE the following:
[10] IF BAT C level decreases to 2270 gals (10%), THEN ENSURE the following:
Line 2,238: Line 2,158:
[19] RECORD final BAT C LEVEL from 1-LI-62-242.
[19] RECORD final BAT C LEVEL from 1-LI-62-242.
________gals. ________
________gals. ________
WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 25 of 124 Date________    Initials  
WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 25 of 124 Date________    Initials 6.3 Transfer BA from BAT C To BAT A (continued)
 
===6.3 Transfer===
BA from BAT C To BAT A (continued)
[20] REQUEST Chemistry Lab sample BAT A.
[20] REQUEST Chemistry Lab sample BAT A.
________ [21] IF 1-HS-62-140A, VCT MAKEUP CONTROL, was positioned to STOP in Step 6.3[1], THEN [21.1] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to ST ART and RELEASE.
________ [21] IF 1-HS-62-140A, VCT MAKEUP CONTROL, was positioned to STOP in Step 6.3[1], THEN [21.1] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to ST ART and RELEASE.
Line 2,292: Line 2,209:
The RHR system can be placed in operation when the pressure and temperature of the RCS are about 400 psig and 350 0F, respectively. If one of the pumps and/or one of the heat exchangers is not operable, safe operation of the plant is not affected; however, the  
The RHR system can be placed in operation when the pressure and temperature of the RCS are about 400 psig and 350 0F, respectively. If one of the pumps and/or one of the heat exchangers is not operable, safe operation of the plant is not affected; however, the  


time for cooldown is extended.  
time for cooldown is extended.
 
4.5.1 Residual Heat Removal Pumps Two identical pumps are installed in the RHR system. Each pump is sized to deliver sufficient reactor coolant flow through the residual heat exchangers to meet the plant cooldown requirements. A seal heat exchanger for each pump is supported by operation of CCS.
====4.5.1 Residual====
Heat Removal Pumps Two identical pumps are installed in the RHR system. Each pump is sized to deliver sufficient reactor coolant flow through the residual heat exchangers to meet the plant cooldown requirements. A seal heat exchanger for each pump is supported by operation of CCS.
4.5.2 RHR Safety Valves The RHR system safety valves provide RCS cold overpressure protection whenever the RHR system is in operation. The valves are located inside containment, one each on the RHR system suction and discharge path, and di scharge to the pressurizer relief tank. The valves are set at 450 psig and 600 psig, respectively.  
4.5.2 RHR Safety Valves The RHR system safety valves provide RCS cold overpressure protection whenever the RHR system is in operation. The valves are located inside containment, one each on the RHR system suction and discharge path, and di scharge to the pressurizer relief tank. The valves are set at 450 psig and 600 psig, respectively.  


===4.6 Safety===
4.6 Safety Injection System Accumulators - Key 36 During normal plant operating conditions, the safety injection system accumulators are pressurized by nitrogen gas in order to inject borated water in the RCS when RCS pressure falls below 600 psi unintentionally. During a controlled depressurization, the accumulators are isolated to prevent injection of safety injection system accumulator borated water. However, if the isolation valves remain open, the borated water will be injected and nitrogen pressure will decrease due to the increased empty volume of the accumulators. Injection of nitrogen into the RCS occurs when RCS pressure is less than 150 psi. The manual isolation of the accumulators is assumed as a post-fire activity. The isolation valve at each accumulator is closed only when the RCS is intentionally depressurized below 1000 psig. If the cables associated with these valves were damaged by fire, isolation is performed locally, governed by appropriate plant procedures (post-fire). In the event the valves are inaccessible, RCS pressure will be maintained greater than 150 psi to preclude nitrogen injection into the RCS via the accumulators.
Injection System Accumulators - Key 36 During normal plant operating conditions, the safety injection system accumulators are pressurized by nitrogen gas in order to inject borated water in the RCS when RCS pressure falls below 600 psi unintentionally. During a controlled depressurization, the accumulators are isolated to prevent injection of safety injection system accumulator borated water. However, if the isolation valves remain open, the borated water will be injected and nitrogen pressure will decrease due to the increased empty volume of the accumulators. Injection of nitrogen into the RCS occurs when RCS pressure is less than 150 psi. The manual isolation of the accumulators is assumed as a post-fire activity. The isolation valve at each accumulator is closed only when the RCS is intentionally depressurized below 1000 psig. If the cables associated with these valves were damaged by fire, isolation is performed locally, governed by appropriate plant procedures (post-fire). In the event the valves are inaccessible, RCS pressure will be maintained greater than 150 psi to preclude nitrogen injection into the RCS via the accumulators.
: 32. Given the following conditions: - Unit 1 is at 100% power.
: 32. Given the following conditions: - Unit 1 is at 100% power.
Which ONE of the following describes the physical location of 1-BKR-74-2A, LOOP 4 HOT LEG TO RHR SUCTION (1-FCV-74-2) AND the NORMAL position of this breaker?  
Which ONE of the following describes the physical location of 1-BKR-74-2A, LOOP 4 HOT LEG TO RHR SUCTION (1-FCV-74-2) AND the NORMAL position of this breaker?  
Line 2,433: Line 2,347:
_____ ...,1 68-57 2 1 11 x 2 11 The spray line is above the nominal water level. TE SPRAY LINE 19 9 9 9 L ___ sPA_Rc_ER TANK-68-PRT 68-309 -
_____ ...,1 68-57 2 1 11 x 2 11 The spray line is above the nominal water level. TE SPRAY LINE 19 9 9 9 L ___ sPA_Rc_ER TANK-68-PRT 68-309 -
E-F..v. J TANK CLASS G 68-610 68-452A CLASS G CONT ON 1-47W819-1, COORD EPRIMARY WATER C) Ul Ul 211 < ...... TV 68-601 TV 1 1-68-559 CLASS G **
E-F..v. J TANK CLASS G 68-610 68-452A CLASS G CONT ON 1-47W819-1, COORD EPRIMARY WATER C) Ul Ul 211 < ...... TV 68-601 TV 1 1-68-559 CLASS G **
* CLASS B u WBN System Description Document REACTOR COOLANT SYSTEM WBN-SDD-N3-68-4001 Rev. 0035 Page 77 of 259  
* CLASS B u WBN System Description Document REACTOR COOLANT SYSTEM WBN-SDD-N3-68-4001 Rev. 0035 Page 77 of 259 3.2.5 Pressurizer Relief Tank (PRT) (continued)
 
====3.2.5 Pressurizer====
Relief Tank (PRT) (continued)
Steam is discharged into the PRT through a sparger pipe under the water level. This condenses and cools the steam by mixing it with water that is near ambient temperature.
Steam is discharged into the PRT through a sparger pipe under the water level. This condenses and cools the steam by mixing it with water that is near ambient temperature.
The tank is equipped with an internal spray and a drain which are used to cool the tank following a discharge. The tank is protected against a discharge exceeding the design value by two rupture discs which discharge into the reactor containment. The tank is carbon steel with a corrosion-resistant coating on the wetted surfaces. A flanged nozzle is provided on the tank for the PZR discharge line connection. This nozzle and the discharge piping and sparger within the vessel are austenitic stainless steel. The tank design is based on the requirement to condense and cool a discharge of PZR steam equal to 110% of the volume above the full-power PZR water level set point. The tank is not designed to accept a continuous discharge from the PZR. The volume of water in the tank is capable of absorbing the heat from the assumed discharge. If the temperature in the tank rises above 128
The tank is equipped with an internal spray and a drain which are used to cool the tank following a discharge. The tank is protected against a discharge exceeding the design value by two rupture discs which discharge into the reactor containment. The tank is carbon steel with a corrosion-resistant coating on the wetted surfaces. A flanged nozzle is provided on the tank for the PZR discharge line connection. This nozzle and the discharge piping and sparger within the vessel are austenitic stainless steel. The tank design is based on the requirement to condense and cool a discharge of PZR steam equal to 110% of the volume above the full-power PZR water level set point. The tank is not designed to accept a continuous discharge from the PZR. The volume of water in the tank is capable of absorbing the heat from the assumed discharge. If the temperature in the tank rises above 128
Line 2,499: Line 2,410:
* 20 PSIG 1-1/4* 1.A I 19 1-70-1 . J55A ----.i--....._..,.......,.
* 20 PSIG 1-1/4* 1.A I 19 1-70-1 . J55A ----.i--....._..,.......,.
_ _. 1-70-540 TANK DIVIDER {BAFFLE) The makeup ' valve fails shut on a loss of air.
_ _. 1-70-540 TANK DIVIDER {BAFFLE) The makeup ' valve fails shut on a loss of air.
WBN Unit 1 Loss of Component Cooling Water (CCS) 1-AOI-15 Rev. 0006 Page 5 of 48  
WBN Unit 1 Loss of Component Cooling Water (CCS) 1-AOI-15 Rev. 0006 Page 5 of 48 3.0 OPERATOR ACTIONS 3.1 Diagnostics IF GO TO Subsection PAGE LOSS OF CCS FLOW:
 
===3.0 OPERATOR===
ACTIONS  
 
===3.1 Diagnostics===
 
IF GO TO Subsection PAGE LOSS OF CCS FLOW:
* Pump trip
* Pump trip
* Multiple low flow alarm
* Multiple low flow alarm
Line 2,623: Line 2,527:
Higher    Lower  X  Question Source:
Higher    Lower  X  Question Source:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
Pressurizer
Pressurizer 3.4.9          Watts Bar-Unit 1 3.4-18        3.4  REACTOR COOLANT SYSTEM (RCS)


====3.4.9 Watts====
3.4.9  Pressurizer
Bar-Unit 1 3.4-18        3.4 REACTOR COOLANT SYSTEM (RCS)
 
====3.4.9 Pressurizer====


LCO  3.4.9  The pressurizer shall be OPERABLE with:
LCO  3.4.9  The pressurizer shall be OPERABLE with:
Line 2,795: Line 2,696:
Higher  Lower X  Question Source:
Higher  Lower X  Question Source:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
RTS Instrumentation
RTS Instrumentation 3.3.1       Watts Bar-Unit 1 3.3-21 Amendment 7 Table 3.3.1-1 (page 7 of
 
====3.3.1 Watts====
Bar-Unit 1 3.3-21 Amendment 7 Table 3.3.1-1 (page 7 of
: 9) Reactor Trip System Instrumentation Note 1:  Overtemperature T The Overtemperature T Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.2% of T span.
: 9) Reactor Trip System Instrumentation Note 1:  Overtemperature T The Overtemperature T Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.2% of T span.
Where: T is measured RCS T, &deg;F. T O is the indicated T at RTP, &deg;F.
Where: T is measured RCS T, &deg;F. T O is the indicated T at RTP, &deg;F.
Line 2,804: Line 2,702:
-1.      T' is the indicated T avg at RTP,        P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure,  2235 psig    K 1  1.16  K 2  0.0183/&deg;F  K 3 = 0.000900/psig 1  33 sec  2  4 sec    4  3 sec  5  3 sec      f 1 (I) =  -2.62{22 + (q t - q b)}  when q t - q b < - 22% RTP    0  when -22% RTP  q t - q b  10% RTP    1.96{(q t - q b) - 10}  when q t - q b > 10% RTP            Where q t and q b are percent RTP in the upper and lower halves of the core, respectively, and q t + q b is the total THERMAL POWER in percent RTP.
-1.      T' is the indicated T avg at RTP,        P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure,  2235 psig    K 1  1.16  K 2  0.0183/&deg;F  K 3 = 0.000900/psig 1  33 sec  2  4 sec    4  3 sec  5  3 sec      f 1 (I) =  -2.62{22 + (q t - q b)}  when q t - q b < - 22% RTP    0  when -22% RTP  q t - q b  10% RTP    1.96{(q t - q b) - 10}  when q t - q b > 10% RTP            Where q t and q b are percent RTP in the upper and lower halves of the core, respectively, and q t + q b is the total THERMAL POWER in percent RTP.


RTS Instrumentation
RTS Instrumentation 3.3.1       Watts Bar-Unit 1 3.3-22 Amendment 7 Table 3.3.1-1 (page 8 of
 
====3.3.1 Watts====
Bar-Unit 1 3.3-22 Amendment 7 Table 3.3.1-1 (page 8 of
: 9) Reactor Trip System Instrumentation Note 2:  Overpower T  The Overpower T Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.0% of T span.      Where: T is measured RCS T, F. T O is the indicated T at RTP, F. s is the Laplace transform operator, sec
: 9) Reactor Trip System Instrumentation Note 2:  Overpower T  The Overpower T Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.0% of T span.      Where: T is measured RCS T, F. T O is the indicated T at RTP, F. s is the Laplace transform operator, sec
-1. T is the measured RCS average temperature, F. T" is the indicated T avg at RTP,  588.2 F.      K 4  1.10 K 5  0.02/F for increasing T avg K 6  0.00162/&deg;F when T > T
-1. T is the measured RCS average temperature, F. T" is the indicated T avg at RTP,  588.2 F.      K 4  1.10 K 5  0.02/F for increasing T avg K 6  0.00162/&deg;F when T > T
Line 3,098: Line 2,993:
---* 6C-1 25A A --1---,--,+ HIGH-HIGH CONTAINMENT  
---* 6C-1 25A A --1---,--,+ HIGH-HIGH CONTAINMENT  
>----t2/4 PRESSURE If 2 of the 4 containment pressure channels reach 2.8 psig (the HI-HI setpoint), then the MSIVs will automatically close. + 0 0 N F F ---l RESET TRAIN A NOTE 5 *-;;f-HS 30-640 0 0 N F F ] I ACTUATE TRAIN A/B PHASE 8 & CNTMT VENT ISOLATION LOG 509011 TRAIN 8 LOG TR B SD9013 '-"'' 6C-125C LOG SD9010 TRAIN A CONTAINMENT SPRAY DISCHARGE VALVE 1-FCV-72-2 1-47\11611-72-1 LOG TR A SD9012 CONTAINMENT SPRAY PUMP 1-47W611-72-1 MAIN STEAM LINE ISOLATION VALVES 1-47W611-1-1 r----------, : PHASE 8 : I CNTMT ISOLATION I L-----------..J SYSTEM SHEET 1-47W611-30 3, 4 1-47W611-70 3 1-47W611-32 2 1-47W611-67 3 A CONTAINMENT SPRAY PUMP AND VALVES l-47W611-72-l TRI *  
>----t2/4 PRESSURE If 2 of the 4 containment pressure channels reach 2.8 psig (the HI-HI setpoint), then the MSIVs will automatically close. + 0 0 N F F ---l RESET TRAIN A NOTE 5 *-;;f-HS 30-640 0 0 N F F ] I ACTUATE TRAIN A/B PHASE 8 & CNTMT VENT ISOLATION LOG 509011 TRAIN 8 LOG TR B SD9013 '-"'' 6C-125C LOG SD9010 TRAIN A CONTAINMENT SPRAY DISCHARGE VALVE 1-FCV-72-2 1-47\11611-72-1 LOG TR A SD9012 CONTAINMENT SPRAY PUMP 1-47W611-72-1 MAIN STEAM LINE ISOLATION VALVES 1-47W611-1-1 r----------, : PHASE 8 : I CNTMT ISOLATION I L-----------..J SYSTEM SHEET 1-47W611-30 3, 4 1-47W611-70 3 1-47W611-32 2 1-47W611-67 3 A CONTAINMENT SPRAY PUMP AND VALVES l-47W611-72-l TRI *  
'\..30* NOT ESFAS Instrumentation
'\..30* NOT ESFAS Instrumentation 3.3.2       Watts Bar-Unit 1 3.3-36 Amendment 23 Table 3.3.2-1 (page 3 of 7) Engineered Safety Feature Actuation System Instrumentation FUNCTION APPLICABLE MODE S OR OTHER SPECIFIED CONDITIONS REQUIRED  CHANNELS  CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE NOMINAL TRIP SETPOINT 4. Steam Line Isolation (continued)
 
====3.3.2 Watts====
Bar-Unit 1 3.3-36 Amendment 23 Table 3.3.2-1 (page 3 of 7) Engineered Safety Feature Actuation System Instrumentation FUNCTION APPLICABLE MODE S OR OTHER SPECIFIED CONDITIONS REQUIRED  CHANNELS  CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE NOMINAL TRIP SETPOINT 4. Steam Line Isolation (continued)
: c. Containment 1, 2(c), 3(c) 4 E SR 3.3.2.1 2.9 psig 2.8 psig  Pressure-    SR 3.3.2.4 High High    SR 3.3.2.9 SR 3.3.2.10
: c. Containment 1, 2(c), 3(c) 4 E SR 3.3.2.1 2.9 psig 2.8 psig  Pressure-    SR 3.3.2.4 High High    SR 3.3.2.9 SR 3.3.2.10
: d. Steam Line Pressure          (1) Low 1, 2(c), 3(a) (c) 3 per steam D SR 3.3.2.1 666.6(b) psig 675(b) psig      line  SR 3.3.2.4 SR 3.3.2.9 SR .3.2.10 (2) Negative 3(d) (c) 3 per steam D SR 3.3.2.1 108.5(e) ps i 100(e) psi    Rate-High  line  SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10
: d. Steam Line Pressure          (1) Low 1, 2(c), 3(a) (c) 3 per steam D SR 3.3.2.1 666.6(b) psig 675(b) psig      line  SR 3.3.2.4 SR 3.3.2.9 SR .3.2.10 (2) Negative 3(d) (c) 3 per steam D SR 3.3.2.1 108.5(e) ps i 100(e) psi    Rate-High  line  SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10
Line 3,148: Line 3,040:
Higher X  Lower    Question Source:
Higher X  Lower    Question Source:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
UFSAR - Amendment 1 PAM Instrumentation
UFSAR - Amendment 1 PAM Instrumentation 3.3.3         Watts Bar-Unit 1 3.3-41 Amendment 72 3.3 INSTRUMENTATION 3.3.3  Post Accident Monitoring (PAM) Instrumentation
 
====3.3.3 Watts====
Bar-Unit 1 3.3-41 Amendment 72
 
===3.3 INSTRUMENTATION===
 
3.3.3  Post Accident Monitoring (PAM) Instrumentation


LCO  3.3.3  The PAM instrumentation for each Function in Table 3.3.3-1 shall be OPERABLE.  
LCO  3.3.3  The PAM instrumentation for each Function in Table 3.3.3-1 shall be OPERABLE.  
Line 3,171: Line 3,056:
Immediately (continued)
Immediately (continued)


PAM Instrumentation
PAM Instrumentation 3.3.3         Watts Bar-Unit 1 3.3-44 Amendment 72 Table 3.3.3-1 (page 1 of
 
====3.3.3 Watts====
Bar-Unit 1 3.3-44 Amendment 72 Table 3.3.3-1 (page 1 of
: 2) Post Accident Monitoring Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS/TRAINS CONDITION REFERENCED FROM REQUIRED ACTION D.1        1. Intermediate Range Neutron Flux(g) 1(a), 2(b), 3 2 E 2. Source Range Neutron Flux 2(c) , 3 2 E 3. Reactor Coolant System (RCS)
: 2) Post Accident Monitoring Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS/TRAINS CONDITION REFERENCED FROM REQUIRED ACTION D.1        1. Intermediate Range Neutron Flux(g) 1(a), 2(b), 3 2 E 2. Source Range Neutron Flux 2(c) , 3 2 E 3. Reactor Coolant System (RCS)
Hot Leg Temperature (T
Hot Leg Temperature (T
Line 3,322: Line 3,204:
Watts Bar Nuclear Plant Unit 1 Abnormal Operating Instruction 1-AOI-25.01 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I Revision 0003 Quality Related Level of Use:  Continuous Use Effective Date: 04-08-2015 Responsible Organization: OPS, Operations Prepared By: Scott Warrington Approved By: William Sprinkle  
Watts Bar Nuclear Plant Unit 1 Abnormal Operating Instruction 1-AOI-25.01 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I Revision 0003 Quality Related Level of Use:  Continuous Use Effective Date: 04-08-2015 Responsible Organization: OPS, Operations Prepared By: Scott Warrington Approved By: William Sprinkle  


WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.01 Rev. 0003 Page 3 of 27  
WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.01 Rev. 0003 Page 3 of 27 1.0 PURPOSE This instruction provides actions to respond to a loss of 120V AC Vital Instrument Power Boards 1-I and 2-I. This instruction will stabilize the unit and provide a list of  
 
===1.0 PURPOSE===
This instruction provides actions to respond to a loss of 120V AC Vital Instrument Power Boards 1-I and 2-I. This instruction will stabilize the unit and provide a list of  


equipment which may NOT be available due to a loss of the board.
equipment which may NOT be available due to a loss of the board.
[C.1, C.3]  
[C.1, C.3]
 
2.0 SYMPTOMS 2.1 Alarms A. 125 DC VITAL CHGR/BATT 1 ABNORMAL [17-A] B. 125 DC VITAL BATT BD 1 ABNORMAL CKTS ISOLATED [17-B]
===2.0 SYMPTOMS===
 
===2.1 Alarms===
A. 125 DC VITAL CHGR/BATT 1 ABNORMAL [17-A] B. 125 DC VITAL BATT BD 1 ABNORMAL CKTS ISOLATED [17-B]
C. 120V AC VITAL PWR BD 1-I UV/CKT TRIP [17-D].
C. 120V AC VITAL PWR BD 1-I UV/CKT TRIP [17-D].
D. 120V AC VITAL INVERTER 1-I / 0-I ABNORMAL [17-C].
D. 120V AC VITAL INVERTER 1-I / 0-I ABNORMAL [17-C].
Line 3,349: Line 3,224:
P. DCS TROUBLE [82-F]
P. DCS TROUBLE [82-F]
Q. POWER RANGE OVERPOWER ROD WD STOP [83-A] R. POWER RANGE UPR DETECTOR FLUX DEVN [83-B] S. POWER RANGE LWR DETECTOR FLUX DEVN [83-C] Thealarmpatternseenfor120VACVitalInstrument Board1-Iisseeninthe stemofthequestion.
Q. POWER RANGE OVERPOWER ROD WD STOP [83-A] R. POWER RANGE UPR DETECTOR FLUX DEVN [83-B] S. POWER RANGE LWR DETECTOR FLUX DEVN [83-C] Thealarmpatternseenfor120VACVitalInstrument Board1-Iisseeninthe stemofthequestion.
WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.01 Rev. 0003  
WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.01 Rev. 0003 2.1 Alarms (continued)
 
===2.1 Alarms===
(continued)
Page 4 of 27 T. POWER RANGE CHANNEL DEVIATION [83-E] U. RVLIS SYS MALFUNCTION [85-F]
Page 4 of 27 T. POWER RANGE CHANNEL DEVIATION [83-E] U. RVLIS SYS MALFUNCTION [85-F]
V. CERPI TROUBLE [86-C]
V. CERPI TROUBLE [86-C]
Line 3,360: Line 3,232:
Z. CL ACCUM 2 LEVEL HI/LO [132-A]
Z. CL ACCUM 2 LEVEL HI/LO [132-A]
AA. CL ACCUM 2 PRESS HI/LO [132-B]
AA. CL ACCUM 2 PRESS HI/LO [132-B]
BB. AC VITAL PWR BD 2-I UV/CKT TRIP [135-F].  
BB. AC VITAL PWR BD 2-I UV/CKT TRIP [135-F].
 
2.2 Indications A. Power range 1-N-41, Intermediate range 1-N-135, and Source Range 1-N-131 failure. B. Rx Trip SI Status panel 1-XX-55-5 will be DARK.
===2.2 Indications===
A. Power range 1-N-41, Intermediate range 1-N-135, and Source Range 1-N-131 failure. B. Rx Trip SI Status panel 1-XX-55-5 will be DARK.
C. Train A CNTMT Isolation Panel 1-XX-55-6E will be DARK.
C. Train A CNTMT Isolation Panel 1-XX-55-6E will be DARK.
D. Loss of Channel I S/G feedwater flow and steam flow inputs to DCS.
D. Loss of Channel I S/G feedwater flow and steam flow inputs to DCS.
Line 3,501: Line 3,371:
Higher X  Lower    Question Source:
Higher X  Lower    Question Source:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 6 of 57  
WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 6 of 57 3.0 PRECAUTIONS AND LIMITATIONS A. All breakers on a board should be OFF or OPEN and all protective grounds must be removed before energizing board. B. A circuit that blows a fuse after being replaced once should have the circuit checked before replacing the fuse a second time, unless the SRO and plant conditions dictate otherwise. C. To prevent explosive mixtures of H 2 and O 2 from accumulating in battery rooms, a battery room exhaust fan should be in service at all times. D. A permanent or portable eyewash stat ion shall be available in the battery rooms. E. Acid spills should be neutralized with a solution of baking soda and water and all traces of the spill should be wiped up. F. There shall be no smoking, open flam es, or arcs in the battery rooms. G. Battery board ground alarm setpoint (90v) is set at 60% of scale. Positive or negative grounds should be maintained as low as possible.
 
===3.0 PRECAUTIONS===
AND LIMITATIONS A. All breakers on a board should be OFF or OPEN and all protective grounds must be removed before energizing board. B. A circuit that blows a fuse after being replaced once should have the circuit checked before replacing the fuse a second time, unless the SRO and plant conditions dictate otherwise. C. To prevent explosive mixtures of H 2 and O 2 from accumulating in battery rooms, a battery room exhaust fan should be in service at all times. D. A permanent or portable eyewash stat ion shall be available in the battery rooms. E. Acid spills should be neutralized with a solution of baking soda and water and all traces of the spill should be wiped up. F. There shall be no smoking, open flam es, or arcs in the battery rooms. G. Battery board ground alarm setpoint (90v) is set at 60% of scale. Positive or negative grounds should be maintained as low as possible.
H. Do NOT allow continued operation of battery chargers if output voltage exceeds 140 volts DC while supplying plant loads or the output current exceeds 250 amps. Continued operation in excess of these values can cause equipment failures; therefor e, prompt action should be taken to remove the chargers from service. I. If 480V power is removed from the charger for more than 5 minutes, the charger will automatically start up in the equalize mode of operation when power is restored. The Float pushbutton will have to be pressed and released in order to return to the float operating mode. J. When placing the charger in service, transitioning between float and equalize modes, or when making sudden large magnitude load changes, there may be alarm actuations due to momentary voltage fl uctuations on the ch arger. If these alarms occur, the operator will need to PRESS and RELEASE the PRESS TO RESET ALARMS pushbutton and ensure loca l alarms clear before proceeding to subsequent steps. K. Vital Battery V has two more cells t han the other Vital Batteries and operates at a higher voltage. Only Vital Battery Charger V can be used to perform an equalizing charge on Vital Battery V due to the other chargers having a high  
H. Do NOT allow continued operation of battery chargers if output voltage exceeds 140 volts DC while supplying plant loads or the output current exceeds 250 amps. Continued operation in excess of these values can cause equipment failures; therefor e, prompt action should be taken to remove the chargers from service. I. If 480V power is removed from the charger for more than 5 minutes, the charger will automatically start up in the equalize mode of operation when power is restored. The Float pushbutton will have to be pressed and released in order to return to the float operating mode. J. When placing the charger in service, transitioning between float and equalize modes, or when making sudden large magnitude load changes, there may be alarm actuations due to momentary voltage fl uctuations on the ch arger. If these alarms occur, the operator will need to PRESS and RELEASE the PRESS TO RESET ALARMS pushbutton and ensure loca l alarms clear before proceeding to subsequent steps. K. Vital Battery V has two more cells t han the other Vital Batteries and operates at a higher voltage. Only Vital Battery Charger V can be used to perform an equalizing charge on Vital Battery V due to the other chargers having a high  


voltage cutout function that is lower than this higher voltage. L. At no time is Vital Battery Charger V to be used to supply battery system loads.
voltage cutout function that is lower than this higher voltage. L. At no time is Vital Battery Charger V to be used to supply battery system loads.
WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 17 of 57 Date________    Initials  
WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 17 of 57 Date________    Initials 8.0 INFREQUENT OPERATIONS 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger NOTES 1) This section assumes either the normal or sp are charger is in service. If no charger is in service, one must be placed in serv ice per the applicable 0-SOI-236 series instruction. 2) At no time is Vital Battery Charger V to be used to supply battery system loads.
 
===8.0 INFREQUENT===
OPERATIONS 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger NOTES 1) This section assumes either the normal or sp are charger is in service. If no charger is in service, one must be placed in serv ice per the applicable 0-SOI-236 series instruction. 2) At no time is Vital Battery Charger V to be used to supply battery system loads.
[1] ENSURE 125V dc Vit Batt Bd V in service per Section 5.0. ________
[1] ENSURE 125V dc Vit Batt Bd V in service per Section 5.0. ________
[2] ENSURE one exhaust fan is operating. ________
[2] ENSURE one exhaust fan is operating. ________
Line 3,518: Line 3,382:
[5] PLACE 0-BKR-236-5/CB2, DC OUTPUT, [Vit Batt Chgr V] in OFF. ________
[5] PLACE 0-BKR-236-5/CB2, DC OUTPUT, [Vit Batt Chgr V] in OFF. ________
________
________
CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 18 of 57 Date________    Initials  
CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 18 of 57 Date________    Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
 
===8.1 Placing===
Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
[6] PLACE 0-BKR-236-5/CB1, AC INPUT, [Vit Batt Chgr V] in OFF. ________  
[6] PLACE 0-BKR-236-5/CB1, AC INPUT, [Vit Batt Chgr V] in OFF. ________  


Line 3,534: Line 3,395:
(N/A others) Vital Battery Board Procedure Initials 125V dc Vital Battery Board I 0-SOI-236.01 125V dc Vital Battery Board II 0-SOI-236.02 125V dc Vital Battery Board III 0-SOI-236.03 125V dc Vital Battery Board IV 0-SOI-236.04  
(N/A others) Vital Battery Board Procedure Initials 125V dc Vital Battery Board I 0-SOI-236.01 125V dc Vital Battery Board II 0-SOI-236.02 125V dc Vital Battery Board III 0-SOI-236.03 125V dc Vital Battery Board IV 0-SOI-236.04  


WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 19 of 57 Date________    Initials  
WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 19 of 57 Date________    Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
 
===8.1 Placing===
Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
NOTE Vital Batt V is 62 cells with 137 to 140 volts of float voltage
NOTE Vital Batt V is 62 cells with 137 to 140 volts of float voltage
[11] PERFORM the following:
[11] PERFORM the following:
Line 3,543: Line 3,401:
[11.2] RECORD float voltage. _____________VDC. ________
[11.2] RECORD float voltage. _____________VDC. ________
[11.3] IF recorded voltage is NOT 137 to 140 VDC, THEN NOTIFY Electrical Maintenance to adjust voltage to within 137 to 140 VDC. ________
[11.3] IF recorded voltage is NOT 137 to 140 VDC, THEN NOTIFY Electrical Maintenance to adjust voltage to within 137 to 140 VDC. ________
[12] PLACE feeder breaker to 125V DC DIST PNL 5A or 5B in ON position to set up feed to desired Vital Batt Bd (N/A other)  [125V Vit Batt Bd V, pnl 2]: Desired Battery Bd  Feed Breaker to CLOSE or place in ON PERF INITIALS VERIFIER INITIALS Feed to Battery Bds I and III 0-BKR-236-5/201 CVFeed to Battery Bds II and IV 0-BKR-236-5/202 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 20 of 57 Date________    Initials  
[12] PLACE feeder breaker to 125V DC DIST PNL 5A or 5B in ON position to set up feed to desired Vital Batt Bd (N/A other)  [125V Vit Batt Bd V, pnl 2]: Desired Battery Bd  Feed Breaker to CLOSE or place in ON PERF INITIALS VERIFIER INITIALS Feed to Battery Bds I and III 0-BKR-236-5/201 CVFeed to Battery Bds II and IV 0-BKR-236-5/202 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 20 of 57 Date________    Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
 
===8.1 Placing===
Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
CAUTION Breakers in 125V DC DIST PNL 5A or 5B (el 757, outside of Battery Bd rooms II and III) are in OFF position when placed in the UP position.
CAUTION Breakers in 125V DC DIST PNL 5A or 5B (el 757, outside of Battery Bd rooms II and III) are in OFF position when placed in the UP position.
NOTE Closing 0-BKR-236-5/301 or 0-BKR-236-5/302 will cause an alarm on annunciator  
NOTE Closing 0-BKR-236-5/301 or 0-BKR-236-5/302 will cause an alarm on annunciator  
Line 3,552: Line 3,407:
point 21-B. Closing 0-BKR-236-5/401 or 0-BKR-236-5/402 will cause an alarm on annunciator point 21-C.
point 21-B. Closing 0-BKR-236-5/401 or 0-BKR-236-5/402 will cause an alarm on annunciator point 21-C.
[13] PLACE one of the following breakers in 125V DC DIST PNL 5A or 5B [el 757, cabinets outside of Batt Bd rooms II and III] in ON to align feed to desired Vital Batt Bd.
[13] PLACE one of the following breakers in 125V DC DIST PNL 5A or 5B [el 757, cabinets outside of Batt Bd rooms II and III] in ON to align feed to desired Vital Batt Bd.
(N/A three others). Desired Battery Bd  Feed Breaker to CLOSE or place in ON PERF INITIALS VERIFIER INITIALS Feed to Battery Bds I 0-BKR-236-5/301 CV Feed to Battery Bds II 0-BKR-236-5/401 CVFeed to Battery Bds III 0-BKR-236-5/302 CVFeed to Battery Bds IV 0-BKR-236-5/402 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 21 of 57 Date________    Initials  
(N/A three others). Desired Battery Bd  Feed Breaker to CLOSE or place in ON PERF INITIALS VERIFIER INITIALS Feed to Battery Bds I 0-BKR-236-5/301 CV Feed to Battery Bds II 0-BKR-236-5/401 CVFeed to Battery Bds III 0-BKR-236-5/302 CVFeed to Battery Bds IV 0-BKR-236-5/402 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 21 of 57 Date________    Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
 
===8.1 Placing===
Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
CAUTION Board will be isolated from battery after performance of next step. Do NOT transfer shutdown boards or start large pumps until battery is on board.
CAUTION Board will be isolated from battery after performance of next step. Do NOT transfer shutdown boards or start large pumps until battery is on board.
NOTES 1) Opening breakers in the following step may cause an alarm on annunciator points 17B, 18B, 19B, or 20B. 2) In Modes 1, 2, 3, and 4, TS 3.8.4 should be applicable.
NOTES 1) Opening breakers in the following step may cause an alarm on annunciator points 17B, 18B, 19B, or 20B. 2) In Modes 1, 2, 3, and 4, TS 3.8.4 should be applicable.
Line 3,564: Line 3,416:
[14.3] IF recorded voltage is NOT 137 to 140 VDC, THEN NOTIFY Electrical Maintenance to adjust voltage to within 137 to 140 VDC. ________
[14.3] IF recorded voltage is NOT 137 to 140 VDC, THEN NOTIFY Electrical Maintenance to adjust voltage to within 137 to 140 VDC. ________
[15] OPEN one of the following to disc onnect normal vital batt from the vital batt bd [Vital Batt Bd - Pnl 1]:  (N/A three others)
[15] OPEN one of the following to disc onnect normal vital batt from the vital batt bd [Vital Batt Bd - Pnl 1]:  (N/A three others)
Assoc. Alarm BATTERY BOARD UNID PERF INITIALS VERIFIER INITIALS 17B 125v dc Vit Batt Bd I 0-BKR-236-1/109 CV18B 125v dc Vit Batt Bd II 0-BKR-236-2/109 CV19B 125v dc Vit Batt Bd III 0-BKR-236-3/109 CV20B 125v dc Vit Batt Bd IV 0-BKR-236-4/109 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 22 of 57 Date________    Initials  
Assoc. Alarm BATTERY BOARD UNID PERF INITIALS VERIFIER INITIALS 17B 125v dc Vit Batt Bd I 0-BKR-236-1/109 CV18B 125v dc Vit Batt Bd II 0-BKR-236-2/109 CV19B 125v dc Vit Batt Bd III 0-BKR-236-3/109 CV20B 125v dc Vit Batt Bd IV 0-BKR-236-4/109 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 22 of 57 Date________    Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
 
===8.1 Placing===
Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
[16] PLACE one of the following switches to appropriate position:
[16] PLACE one of the following switches to appropriate position:
(N/A others)  [Pnl 0s in each 125V Vit Batt Bd Rm]
(N/A others)  [Pnl 0s in each 125V Vit Batt Bd Rm]
Line 3,588: Line 3,437:
125V VITAL BATTERY BD V 0-DPL-236-2vital batt rm 2 0-BKR-236-2/0 CVBACKUP SUPPLY FROM 125V VITAL BATTERY BD V 0-DPL-236-3vital batt rm 3 0-BKR-236-3/0 CVBACKUP SUPPLY FROM  
125V VITAL BATTERY BD V 0-DPL-236-2vital batt rm 2 0-BKR-236-2/0 CVBACKUP SUPPLY FROM 125V VITAL BATTERY BD V 0-DPL-236-3vital batt rm 3 0-BKR-236-3/0 CVBACKUP SUPPLY FROM  


125V VITAL BATTERY BD V 0-DPL-236-4vital batt rm 4 0-BKR-236-4/0 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 23 of 57 Date________    Initials  
125V VITAL BATTERY BD V 0-DPL-236-4vital batt rm 4 0-BKR-236-4/0 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 23 of 57 Date________    Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
 
===8.1 Placing===
Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued)
NOTE In Modes 1, 2, 3 and 4, TS 3.8.4 may be exited. In Modes 5 and 6, TS 3.8.5 may be exited.
NOTE In Modes 1, 2, 3 and 4, TS 3.8.4 may be exited. In Modes 5 and 6, TS 3.8.5 may be exited.
[18] CHECK current flow toward Vital Battery V (negative or zero) to ensure normal or spare Vital Charger is carrying load and  
[18] CHECK current flow toward Vital Battery V (negative or zero) to ensure normal or spare Vital Charger is carrying load and  
Line 3,656: Line 3,502:
EMERG START (1-M-1) 0 01AEN (1-F'U-275-R73/L
EMERG START (1-M-1) 0 01AEN (1-F'U-275-R73/L
: 10) 2-HS-82-15 EMERG START (2-M-1 l A 1 D 1 CES
: 10) 2-HS-82-15 EMERG START (2-M-1 l A 1 D 1 CES
* 2 2BY 8 l-HS-8"-1 5 B part o f th e co mm o n e mer ge n cy s tart and as s u c h w ill ii ca u.-,e all fo ur DGs t o E:: s tart (\-FU-27 0 (1-f"U-27 DSL GEN TA-A REMOTE EMERG START CIRCUIT DSL GEN 1 A-A Cati 7 8 WBN Unit 0 Diesel Generator (DG) 1A-A 0-SOI-82.01 Rev. 0004 Page 12 of 104  
* 2 2BY 8 l-HS-8"-1 5 B part o f th e co mm o n e mer ge n cy s tart and as s u c h w ill ii ca u.-,e all fo ur DGs t o E:: s tart (\-FU-27 0 (1-f"U-27 DSL GEN TA-A REMOTE EMERG START CIRCUIT DSL GEN 1 A-A Cati 7 8 WBN Unit 0 Diesel Generator (DG) 1A-A 0-SOI-82.01 Rev. 0004 Page 12 of 104 3.0 PRECAUTIONS AND LIMI TATIONS (continued)
 
===3.0 PRECAUTIONS===
AND LIMI TATIONS (continued)
FF. The DG may surge from 400-450 RPM after an Idle Start, which is to be expected and does NOT have any adverse consequences. GG. The fuel oil transfer pumps will auto matically start when the gauge indicates between mark 6-9 (out of the 16 marks for tank gauge). HH. ERCW automatically aligns to a DG at 40 rpm. If alignment of ERCW CANNOT immediately be obtained, the DG should be Emergency Stopped. II. Alarm window 225-E TR A/B ERCW TO C&SS COMPR FLOW HI may alarm on some DG starts due to high diffe rential between the two ERCW headers.
FF. The DG may surge from 400-450 RPM after an Idle Start, which is to be expected and does NOT have any adverse consequences. GG. The fuel oil transfer pumps will auto matically start when the gauge indicates between mark 6-9 (out of the 16 marks for tank gauge). HH. ERCW automatically aligns to a DG at 40 rpm. If alignment of ERCW CANNOT immediately be obtained, the DG should be Emergency Stopped. II. Alarm window 225-E TR A/B ERCW TO C&SS COMPR FLOW HI may alarm on some DG starts due to high diffe rential between the two ERCW headers.
WBN Unit 1 Diesel Generator (DG) 1A-A SOI-82.01 Rev. 0090 Page 2 of 103 Revision Log Revision or Change Number Effective Date Affected Page Numbers Description of Revision/Change 81 10/04/12 2, 68 Incorporated changes associated with DCN #58383 by adding a new DG Emer gency Stop/Reset Transfer switch 1-XS-82-125, which is located on  
WBN Unit 1 Diesel Generator (DG) 1A-A SOI-82.01 Rev. 0090 Page 2 of 103 Revision Log Revision or Change Number Effective Date Affected Page Numbers Description of Revision/Change 81 10/04/12 2, 68 Incorporated changes associated with DCN #58383 by adding a new DG Emer gency Stop/Reset Transfer switch 1-XS-82-125, which is located on  
Line 3,720: Line 3,563:
Comments:   
Comments:   


WBN Unit 0 Diesel Generator (DG) 1A-A 0-SOI-82.01 Rev. 0004 Page 61 of 104 Date________    INITIALS  
WBN Unit 0 Diesel Generator (DG) 1A-A 0-SOI-82.01 Rev. 0004 Page 61 of 104 Date________    INITIALS 8.3.1 Removing DG from Service-DG Tied on to SD Bd (continued)
 
====8.3.1 Removing====
DG from Service-DG Tied on to SD Bd (continued)
CAUTION When adjusting speed and voltage, care must be taken to prevent overshooting desired values. Voltage control response is approxim ately five times faster than speed control response.
CAUTION When adjusting speed and voltage, care must be taken to prevent overshooting desired values. Voltage control response is approxim ately five times faster than speed control response.
[12] MATCH generator Incoming Fr equency (1-XI-82-2) with Running Frequency (1-XI-82-3) using 1-HS-82-13, SPEED CONTROL [0-M-26]. ________
[12] MATCH generator Incoming Fr equency (1-XI-82-2) with Running Frequency (1-XI-82-3) using 1-HS-82-13, SPEED CONTROL [0-M-26]. ________
Line 3,739: Line 3,579:
   .Note:  SOI-82.01, "Diesel Generator (DG) 1A-A" Section 8.3.2, "Removing DG from Service-DG NOT Tied on to SD Bd"REFERENCE PROVIDED (1)    (2)  correct    acceptable  correct  unacceptable incorrect    acceptable incorrect  unacceptable
   .Note:  SOI-82.01, "Diesel Generator (DG) 1A-A" Section 8.3.2, "Removing DG from Service-DG NOT Tied on to SD Bd"REFERENCE PROVIDED (1)    (2)  correct    acceptable  correct  unacceptable incorrect    acceptable incorrect  unacceptable
: a. b. c. d. Thursday, October 30, 2014 9:54:19 AM 1
: a. b. c. d. Thursday, October 30, 2014 9:54:19 AM 1
WBN Unit 0 Diesel Generator (DG) 1A-A 0-SOI-82.01 Rev. 0004 Page 61 of 104 Date________    INITIALS  
WBN Unit 0 Diesel Generator (DG) 1A-A 0-SOI-82.01 Rev. 0004 Page 61 of 104 Date________    INITIALS 8.3.1 Removing DG from Service-DG Tied on to SD Bd (continued)
 
====8.3.1 Removing====
DG from Service-DG Tied on to SD Bd (continued)
CAUTION When adjusting speed and voltage, care must be taken to prevent overshooting desired values. Voltage control response is approxim ately five times faster than speed control response.
CAUTION When adjusting speed and voltage, care must be taken to prevent overshooting desired values. Voltage control response is approxim ately five times faster than speed control response.
[12] MATCH generator Incoming Fr equency (1-XI-82-2) with Running Frequency (1-XI-82-3) using 1-HS-82-13, SPEED CONTROL [0-M-26]. ________
[12] MATCH generator Incoming Fr equency (1-XI-82-2) with Running Frequency (1-XI-82-3) using 1-HS-82-13, SPEED CONTROL [0-M-26]. ________
Line 4,006: Line 3,843:
New  Modified Bank  Bank X  Question History: Bank question 078 K4.02 number 54.
New  Modified Bank  Bank X  Question History: Bank question 078 K4.02 number 54.
Comments:
Comments:
WBN Unit 0 Loss of Control Air 0-AOI-10 Rev. 0002 Page 4 of 41  
WBN Unit 0 Loss of Control Air 0-AOI-10 Rev. 0002 Page 4 of 41 2.3 Automatic Actions A. Possible Rx trip due to MFW reg valves or heater drain valves failing closed. B. MSIVs may fail closed. C. Operating compressors load in s equence and compressors selected for auto will start on dropping press. D. Aux Air compressors st art at 83 psig dropping. E. Service air isolates from control air at 80 psig dropping.
 
===2.3 Automatic===
Actions A. Possible Rx trip due to MFW reg valves or heater drain valves failing closed. B. MSIVs may fail closed. C. Operating compressors load in s equence and compressors selected for auto will start on dropping press. D. Aux Air compressors st art at 83 psig dropping. E. Service air isolates from control air at 80 psig dropping.
F. Aux air isolates from control air at 79.5 psig dropping.
F. Aux air isolates from control air at 79.5 psig dropping.
G. Essential and non-essential air to the Rx bldg isolates at 70 psig dropping. H. ERCW to the C&SS air compressors is olate on high flow in conjunction with low press on the compressor ERCW supply headers.
G. Essential and non-essential air to the Rx bldg isolates at 70 psig dropping. H. ERCW to the C&SS air compressors is olate on high flow in conjunction with low press on the compressor ERCW supply headers.
Line 4,081: Line 3,915:
* K629Xt
* K629Xt
* K629X2 OPE N FOR OPEN fOR (R-58) SOURCE (R-58) SOURCE J RANGE J RANGE BlOCK eLOC K .. 1 9 I f'o 2  NCJ 1 ex' C LOSEO 2  l R-58) FOR S.R. H I f'LUX (R-58) AT SHUTDOWN AT SHUTDOWN ( 1-45W600-57-1J) 1 {1-45160 0-57-1 3) 1 > 1NJUX, OPEH  INJT AX , OPEN 1  FOR S.R. SHUTDN {R-58) FOR S. R. SHU TON (R-58) FLUX ALARM BLOCK F L UX ALARM BLOCK ( t-45W600-57-1 J) 2 { 1-45'#60 0-5 7-1 J) 2  1406VG 1 1-FU-99-R!58/M8 CONTAINMENT EVACUATION HORN j<eD 2 81-9) l (BD 282-B)
* K629X2 OPE N FOR OPEN fOR (R-58) SOURCE (R-58) SOURCE J RANGE J RANGE BlOCK eLOC K .. 1 9 I f'o 2  NCJ 1 ex' C LOSEO 2  l R-58) FOR S.R. H I f'LUX (R-58) AT SHUTDOWN AT SHUTDOWN ( 1-45W600-57-1J) 1 {1-45160 0-57-1 3) 1 > 1NJUX, OPEH  INJT AX , OPEN 1  FOR S.R. SHUTDN {R-58) FOR S. R. SHU TON (R-58) FLUX ALARM BLOCK F L UX ALARM BLOCK ( t-45W600-57-1 J) 2 { 1-45'#60 0-5 7-1 J) 2  1406VG 1 1-FU-99-R!58/M8 CONTAINMENT EVACUATION HORN j<eD 2 81-9) l (BD 282-B)
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 16 of 57  
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 16 of 57 2.2.5 Sequence of Step Performance (continued)
 
====2.2.5 Sequence====
of Step Performance (continued)
B. A required task need NOT be fully completed before proceeding to the next action. 1. Efficient implementation of the EOI network is important; therefore completion of the task is NOT required before continuing with the next step. 2. It is sufficient to begin a task and have assurance it is progressing satisfactorily before continuing with the next step; e.g., when opening a valve, the operator verifies t hat the red light has come on. 3. Completion of steps treated in this manner should be confirmed after sufficient time has elapsed for the acti on to complete; e.g., the operator verifies that the valve is full open based upon the red light on and the green light off. C. Substeps are identified as sequential or non-sequential by the substep designator. 1. Sequential substeps are identified by letters or numbers with a parenthesis.
B. A required task need NOT be fully completed before proceeding to the next action. 1. Efficient implementation of the EOI network is important; therefore completion of the task is NOT required before continuing with the next step. 2. It is sufficient to begin a task and have assurance it is progressing satisfactorily before continuing with the next step; e.g., when opening a valve, the operator verifies t hat the red light has come on. 3. Completion of steps treated in this manner should be confirmed after sufficient time has elapsed for the acti on to complete; e.g., the operator verifies that the valve is full open based upon the red light on and the green light off. C. Substeps are identified as sequential or non-sequential by the substep designator. 1. Sequential substeps are identified by letters or numbers with a parenthesis.
Sequential substeps must be per formed in the order listed. 2. Non-sequential substeps are identifi ed by bullets. Bulleted substeps may be performed in any order and they also have an implied "AND" logic  (i.e., all listed steps must be performed) UNLESS an "OR" conjunction is specifically used, or an im plied "OR" logic is stated in the high level action  (e.g., "PERFORM one or more of the following" or "IF an y of the following are met", etc.). 3. In general, the rules for completing a given task in steps apply to substeps. EXAMPLE 1 (Sequential Substeps)
Sequential substeps must be per formed in the order listed. 2. Non-sequential substeps are identifi ed by bullets. Bulleted substeps may be performed in any order and they also have an implied "AND" logic  (i.e., all listed steps must be performed) UNLESS an "OR" conjunction is specifically used, or an im plied "OR" logic is stated in the high level action  (e.g., "PERFORM one or more of the following" or "IF an y of the following are met", etc.). 3. In general, the rules for completing a given task in steps apply to substeps. EXAMPLE 1 (Sequential Substeps)
Line 4,090: Line 3,921:
: 3. CHECK S/G press:
: 3. CHECK S/G press:
* All S/G press controlled or rising
* All S/G press controlled or rising
* All S/G press greater than 150 psig WBN Unit 1 Refueling Operations 1-GO-7 Rev. 0000 Page 17 of 65 Date ___________    Initials  
* All S/G press greater than 150 psig WBN Unit 1 Refueling Operations 1-GO-7 Rev. 0000 Page 17 of 65 Date ___________    Initials 5.1 Reactor Disassembly (continued)
 
===5.1 Reactor===
Disassembly (continued)
NOTE If the Transfer Canal has been filled and draining is NOT required, the following two steps may be marked N/A. [9] CHECK SFP gate seal to the Transfer Canal does NOT leak excessively (leakage greater than SFP makeup flow capability) when Transfer Canal is drained for Transfer Tube Blind Flange  
NOTE If the Transfer Canal has been filled and draining is NOT required, the following two steps may be marked N/A. [9] CHECK SFP gate seal to the Transfer Canal does NOT leak excessively (leakage greater than SFP makeup flow capability) when Transfer Canal is drained for Transfer Tube Blind Flange  


Line 4,195: Line 4,023:
----------------
----------------
THIS CONFIGURATION CONTROL DRAWING DESIGNER REVIEWER RO ISSUE PER WBEP 5.17 a, RIMS 826 '90 0829 37 8 2 P.W.BUllNETT 1
THIS CONFIGURATION CONTROL DRAWING DESIGNER REVIEWER RO ISSUE PER WBEP 5.17 a, RIMS 826 '90 0829 37 8 2 P.W.BUllNETT 1
3 L .IV.BOYD CAD MAINTAINED CONTROL WBN System Description Document GASEOUS WASTE DISPOSAL SYSTEM WBN-SDD-N3-77A-4001 Rev. 0010 Page 28 of 56  
3 L .IV.BOYD CAD MAINTAINED CONTROL WBN System Description Document GASEOUS WASTE DISPOSAL SYSTEM WBN-SDD-N3-77A-4001 Rev. 0010 Page 28 of 56 3.2.1 Waste Gas Compressors (continued)
 
====3.2.1 Waste====
Gas Compressors (continued)
M. Pipe Joints All GWDS piping joints are welded except where flanged connections are necessary for maintenance (e.g., there are 16 flanged connections per compressor package). 3.2.2 Decay Tank Component Description The nine GWDS gas decay tanks are vertical, cylindrical carbon steel tanks with a volume of  
M. Pipe Joints All GWDS piping joints are welded except where flanged connections are necessary for maintenance (e.g., there are 16 flanged connections per compressor package). 3.2.2 Decay Tank Component Description The nine GWDS gas decay tanks are vertical, cylindrical carbon steel tanks with a volume of  


600 ft 3 each. The dimensions of the tanks are approximately 8 feet 8 inches outside diameter by approximately 10 feet vertical height. They are fitted with 3 nozzles, a 1-inch inlet, a 1-inch outlet, and a 3/4-inch sample line. Each tank has 2 manholes, both 16 inches in diameter:  One is located on the top of the tank and the other on the side. Tanks have a design pressure of 150 lb/in 2 g (Table 9-5), a design temperature of 180
600 ft 3 each. The dimensions of the tanks are approximately 8 feet 8 inches outside diameter by approximately 10 feet vertical height. They are fitted with 3 nozzles, a 1-inch inlet, a 1-inch outlet, and a 3/4-inch sample line. Each tank has 2 manholes, both 16 inches in diameter:  One is located on the top of the tank and the other on the side. Tanks have a design pressure of 150 lb/in 2 g (Table 9-5), a design temperature of 180
&deg;F, a normal operating pressure of 0-100 lb/in 2g, and a normal operating temperature of 40-140&deg;F. They have a service life of 40 years and are TVA Class D. 3.2.3 Waste Gas Filter Component Description A cleanup system combining a charcoal absorber and HEPA filters removes particulates and gaseous iodine compounds from the gas discharged. The system is designed to handle a maximum flow of 100 cfm at a maximum pressure and temperature of 5.3 lb/in 2 g and 180&deg;F. The maximum pressure drop across a clean element is 6 inches H 2 O at a flow rate of 100 SCFM. Collection efficiency of the particulate filter exceeds 99 percent. The iodine cartridge has a minimum collection efficiency for methyl iodine of 70 percent and 95 percent collection efficiency for elemental iodine. 3.3 Instrumentation and Controls Instrument setpoints and functions are provided in reference 7.54. 3.3.1 Vent Header Pressure System (PCV 77-89) This system indicates and controls the vent header pressure. Pressure indication is provided on the Waste Processing System (WPS) panel. When the backup waste gas compressor is in automatic mode, the controller will start the compressor on high pressure and actuate a high pressure alarm. The backup compressor is stopped automatically when the vent header pressure drops back to normal. In manual mode, both compressors may be turned on or off regardless of the vent header pressure. The vent header high pressure alarm is actuated regardless of the mode of operation of the compressor. A low pressure alarm is also provided with indication and alarms on the WPS panel. 3.3.2 Compressor Control Instrumentation The WG compressor controls are a part of the compressor packages. These consist of pressure, flow, and level controls. Refer to the simplified flow diagram (Figure 8.4). Under normal conditions one compressor package is in automatic with the second acting as backup. On high inlet pressure (2.0 lb/in 2g) the starter switch of the compressor in automatic is energized. Auxiliary contacts on this switch provide operating power for solenoid valves and pressure switches in the waste gas compressor package.
&deg;F, a normal operating pressure of 0-100 lb/in 2g, and a normal operating temperature of 40-140&deg;F. They have a service life of 40 years and are TVA Class D. 3.2.3 Waste Gas Filter Component Description A cleanup system combining a charcoal absorber and HEPA filters removes particulates and gaseous iodine compounds from the gas discharged. The system is designed to handle a maximum flow of 100 cfm at a maximum pressure and temperature of 5.3 lb/in 2 g and 180&deg;F. The maximum pressure drop across a clean element is 6 inches H 2 O at a flow rate of 100 SCFM. Collection efficiency of the particulate filter exceeds 99 percent. The iodine cartridge has a minimum collection efficiency for methyl iodine of 70 percent and 95 percent collection efficiency for elemental iodine. 3.3 Instrumentation and Controls Instrument setpoints and functions are provided in reference 7.54. 3.3.1 Vent Header Pressure System (PCV 77-89) This system indicates and controls the vent header pressure. Pressure indication is provided on the Waste Processing System (WPS) panel. When the backup waste gas compressor is in automatic mode, the controller will start the compressor on high pressure and actuate a high pressure alarm. The backup compressor is stopped automatically when the vent header pressure drops back to normal. In manual mode, both compressors may be turned on or off regardless of the vent header pressure. The vent header high pressure alarm is actuated regardless of the mode of operation of the compressor. A low pressure alarm is also provided with indication and alarms on the WPS panel. 3.3.2 Compressor Control Instrumentation The WG compressor controls are a part of the compressor packages. These consist of pressure, flow, and level controls. Refer to the simplified flow diagram (Figure 8.4). Under normal conditions one compressor package is in automatic with the second acting as backup. On high inlet pressure (2.0 lb/in 2g) the starter switch of the compressor in automatic is energized. Auxiliary contacts on this switch provide operating power for solenoid valves and pressure switches in the waste gas compressor package.
WBN System Description Document GASEOUS WASTE DISPOSAL SYSTEM WBN-SDD-N3-77A-4001 Rev. 0010 Page 45 of 56  
WBN System Description Document GASEOUS WASTE DISPOSAL SYSTEM WBN-SDD-N3-77A-4001 Rev. 0010 Page 45 of 56 9.2 TABLE 9 Gaseous Waste Systems Components Gaseous Waste Systems Components Waste Gas Decay Tanks Number 9 Volume, Each, ft 3 600 Design Pressure, lb/in 2g 150 Design Temperature, &deg;F 180 Normal Operating Pressure, lb/in 2g 0-110 Normal Operating Temperature, &deg;F 50-140 Material of Construction Carbon Steel  Vertical Cylinder Design Type Code/Class ASME III/3 ANS Safety Class D Waste Gas Compressors Number 2 Type Water Sealed  
 
===9.2 TABLE===
9 Gaseous Waste Systems Components Gaseous Waste Systems Components Waste Gas Decay Tanks Number 9 Volume, Each, ft 3 600 Design Pressure, lb/in 2g 150 Design Temperature, &deg;F 180 Normal Operating Pressure, lb/in 2g 0-110 Normal Operating Temperature, &deg;F 50-140 Material of Construction Carbon Steel  Vertical Cylinder Design Type Code/Class ASME III/3 ANS Safety Class D Waste Gas Compressors Number 2 Type Water Sealed  


Centrifugal Design Flow Rate, N 2  (at 140&deg;F, 2 lb/in 2 g)cfm 40 Design Pressure, lb/in 2g 150 Design Temperature, &deg;F 180 Normal Operating Pressure, lb/in 2 g  Suction 2.0-3.5  Discharge 0-100 Normal Operating Temperature, &deg;F 70-130 Design Code/Class ASME III/3 ANS Safety Class D Waste Gas Filter Number 1 Maximum Flow (at 5.3 lb/in 2 g, 180&deg;F) SCFM 100 Material CS Maximum Clean Pressure Drop (at 100 scfm) in H 2 O 6 Type Charcoal/HEPA Combination Design Code ASME VIII ANS Safety Class G Sequential Automatic Gas Analyzer Number 1 (shared) Oxygen Electrochemical of the Polarographic type  0-20% O 2 Hydrogen by Thermal Conductivity  0-100% H 2 Automatic stepping switch* 8 steps Recorded Readout 8 points Temperature 120&deg;F *See Table 9-4.
Centrifugal Design Flow Rate, N 2  (at 140&deg;F, 2 lb/in 2 g)cfm 40 Design Pressure, lb/in 2g 150 Design Temperature, &deg;F 180 Normal Operating Pressure, lb/in 2 g  Suction 2.0-3.5  Discharge 0-100 Normal Operating Temperature, &deg;F 70-130 Design Code/Class ASME III/3 ANS Safety Class D Waste Gas Filter Number 1 Maximum Flow (at 5.3 lb/in 2 g, 180&deg;F) SCFM 100 Material CS Maximum Clean Pressure Drop (at 100 scfm) in H 2 O 6 Type Charcoal/HEPA Combination Design Code ASME VIII ANS Safety Class G Sequential Automatic Gas Analyzer Number 1 (shared) Oxygen Electrochemical of the Polarographic type  0-20% O 2 Hydrogen by Thermal Conductivity  0-100% H 2 Automatic stepping switch* 8 steps Recorded Readout 8 points Temperature 120&deg;F *See Table 9-4.
Line 4,405: Line 4,227:
New X Modified Bank Bank    Question History: New question for the 2015-301 NRC RO Exam Comments:  
New X Modified Bank Bank    Question History: New question for the 2015-301 NRC RO Exam Comments:  


WBN System Description Document SPENT FUEL POOL COOLING AND CLEANING SYSTEM WBN-SDD-N3-78-4001 Rev. 0018 Page 38 of 71  
WBN System Description Document SPENT FUEL POOL COOLING AND CLEANING SYSTEM WBN-SDD-N3-78-4001 Rev. 0018 Page 38 of 71 5.2 System Leakage Leakage detection is provided for the SFP by leakage channels located on the back side of each welded joint of the floor and walls of SFP steel liner. Leakage into these channels will drain to the perimeter leakage channels located at the bottom of the SFP. The leakage will then flow into the SFP drain pipe to a normally open manual gate valve. Visual detection of the leakage from the SFP may be witnessed as the leakage exits the manual valve and drips into a funnel. The leakage is then routed to the tritiated drain collector tank (TDCT) of the waste disposal system. In the event of excessive leakage, the manual gate valve may be closed to prevent further leakage. Similar type design of leakage channels and visual display of leakage are also provided for the fuel transfer canal and the cask setdown pool. Nonqualified instrumentation are provided in the SFP and the TDCT with MCR and local, low and high level alarms, respectively. Leakage from the SFPCCS pumps is routed to the station drainage system. Abnormal equipment leakage is detected by visual methods. 5.3 Loss of Coolant Accident (LOCA) In the event of a LOCA, the CCS flow to the shell side of the SFP heat exchangers may be diverted to ensure an adequate flow of cooling water for safeguard loads after an accident. 5.4 Failure of the Refueling Cavity Water Seal SFP water level would decrease in the event of a leak at the Refueling Cavity Water Seal because the transfer tube would be open connecting the refueling canal, and the transfer canal with the SFP. The decreasing SFP water level would be detected by instrumentation located in the SFP that alarms in the main control room. The SFP would be isolated from the refueling cavity by closing the manual gate valve (78-ISV-600) on the auxiliary building side of the transfer tube. If the water level in the refueling cavity decreases to the elevation of the reactor vessel flange before the SFP can be isolated, there will be two feet of water above the stored fuel in the SFP. If during this time a spent fuel assembly were in transit between the reactor core and the SFP and was in the upright position, it could become uncovered. The consequences of uncovering the spent fuel assembly would be possible high radiation level, fuel cladding failures, and release of radioactivity into the containment and/or auxiliary building. For this scenario, the events would be bounded by the design basis analysis for a fuel handling accident as described in Section 15.4.5 of the WBN FSAR.
 
===5.2 System===
Leakage Leakage detection is provided for the SFP by leakage channels located on the back side of each welded joint of the floor and walls of SFP steel liner. Leakage into these channels will drain to the perimeter leakage channels located at the bottom of the SFP. The leakage will then flow into the SFP drain pipe to a normally open manual gate valve. Visual detection of the leakage from the SFP may be witnessed as the leakage exits the manual valve and drips into a funnel. The leakage is then routed to the tritiated drain collector tank (TDCT) of the waste disposal system. In the event of excessive leakage, the manual gate valve may be closed to prevent further leakage. Similar type design of leakage channels and visual display of leakage are also provided for the fuel transfer canal and the cask setdown pool. Nonqualified instrumentation are provided in the SFP and the TDCT with MCR and local, low and high level alarms, respectively. Leakage from the SFPCCS pumps is routed to the station drainage system. Abnormal equipment leakage is detected by visual methods. 5.3 Loss of Coolant Accident (LOCA) In the event of a LOCA, the CCS flow to the shell side of the SFP heat exchangers may be diverted to ensure an adequate flow of cooling water for safeguard loads after an accident. 5.4 Failure of the Refueling Cavity Water Seal SFP water level would decrease in the event of a leak at the Refueling Cavity Water Seal because the transfer tube would be open connecting the refueling canal, and the transfer canal with the SFP. The decreasing SFP water level would be detected by instrumentation located in the SFP that alarms in the main control room. The SFP would be isolated from the refueling cavity by closing the manual gate valve (78-ISV-600) on the auxiliary building side of the transfer tube. If the water level in the refueling cavity decreases to the elevation of the reactor vessel flange before the SFP can be isolated, there will be two feet of water above the stored fuel in the SFP. If during this time a spent fuel assembly were in transit between the reactor core and the SFP and was in the upright position, it could become uncovered. The consequences of uncovering the spent fuel assembly would be possible high radiation level, fuel cladding failures, and release of radioactivity into the containment and/or auxiliary building. For this scenario, the events would be bounded by the design basis analysis for a fuel handling accident as described in Section 15.4.5 of the WBN FSAR.
SFP water temperature would approach and remain at the boiling temperature until the SFPCCS service is restored. 6.0 MAINTENANCE AND TESTING COMMITMENTS Preoperational testing requirements for specific items not contained in other design output documents are provided in Appendix A. Each fuel pool gate shall be visually inspected for defects in the structure, seals or fittings prior to each use. The relief valves shall be checked to prevent over inflation of the seals. Leakage of the seal shall be checked by pressure decay test after the gate is installed.
SFP water temperature would approach and remain at the boiling temperature until the SFPCCS service is restored. 6.0 MAINTENANCE AND TESTING COMMITMENTS Preoperational testing requirements for specific items not contained in other design output documents are provided in Appendix A. Each fuel pool gate shall be visually inspected for defects in the structure, seals or fittings prior to each use. The relief valves shall be checked to prevent over inflation of the seals. Leakage of the seal shall be checked by pressure decay test after the gate is installed.
Seals may also be checked in the stored position.   
Seals may also be checked in the stored position.   
Line 4,435: Line 4,254:
I I SECTION A6-A6 (Sheets /,,? i. 3) NOTES FOR DET G6 PAINT COATING MATERIAL RED: AMER LOCK 400NT, SAFETY RED 1979, OR EQUAL WHITE: AMERLOCK 400NT, WHITE, OR EQUAL GREEN: AMERLOCK 400NT, RT-2305 GREEN, OR EQUAL THE STRIPES USED ARE GREEN TO REPRESENT THE NORMAL OPERATING RANGE FROM rnE !
I I SECTION A6-A6 (Sheets /,,? i. 3) NOTES FOR DET G6 PAINT COATING MATERIAL RED: AMER LOCK 400NT, SAFETY RED 1979, OR EQUAL WHITE: AMERLOCK 400NT, WHITE, OR EQUAL GREEN: AMERLOCK 400NT, RT-2305 GREEN, OR EQUAL THE STRIPES USED ARE GREEN TO REPRESENT THE NORMAL OPERATING RANGE FROM rnE !
m' y ALARM INDICATOR AND THE RANGE FROM 749'-1/2" TO 748'-11-1/2" AT THE LOW LEVEL ALARM I ND I CA TOR; AND UPPER AND LOWER RED STRIPES TO I ND I CATE LEVELS ABOVE THE LEVEL ALARM RANGES. (FOR ADDITIONAL HELP IN IDENTIFYING THE WATER LEVEL THE GREEN LINE INDICATING THE NORMAL LEVEL IS EXTENDED TO A 1 INCH WIDE POINTER ON THIS SECOND LEVEL INDICATOR), FOR THIS APPLICATION THE AMERLOCK 400NT PAINT IS ACCEPTABLE FOR USE ON THE STAINLESS STEEL MATERIAL AND IN THE FUEL PIT WATER. DO NOT PA I NT WI TH IN 1
m' y ALARM INDICATOR AND THE RANGE FROM 749'-1/2" TO 748'-11-1/2" AT THE LOW LEVEL ALARM I ND I CA TOR; AND UPPER AND LOWER RED STRIPES TO I ND I CATE LEVELS ABOVE THE LEVEL ALARM RANGES. (FOR ADDITIONAL HELP IN IDENTIFYING THE WATER LEVEL THE GREEN LINE INDICATING THE NORMAL LEVEL IS EXTENDED TO A 1 INCH WIDE POINTER ON THIS SECOND LEVEL INDICATOR), FOR THIS APPLICATION THE AMERLOCK 400NT PAINT IS ACCEPTABLE FOR USE ON THE STAINLESS STEEL MATERIAL AND IN THE FUEL PIT WATER. DO NOT PA I NT WI TH IN 1
* 5 I NCH DIAMETER OF HOLE CENTERLINES TO CLEAR NUTS. OCA 050006-03-0 TYP EL 749 '-1-1/2" I I I I L ______ _ W ATER LEVEL SCALE DETAILS DET GS (CUT AS SHOWN) '* PIPE ALL MATERIAL TO BE 1/4" THICK STAINLESS STEE NTS WBN Unit 0 Spent Fuel Pool Cooling And Cleaning System 0-SOI-78.01 Rev. 0007 Page 10 of 197  
* 5 I NCH DIAMETER OF HOLE CENTERLINES TO CLEAR NUTS. OCA 050006-03-0 TYP EL 749 '-1-1/2" I I I I L ______ _ W ATER LEVEL SCALE DETAILS DET GS (CUT AS SHOWN) '* PIPE ALL MATERIAL TO BE 1/4" THICK STAINLESS STEE NTS WBN Unit 0 Spent Fuel Pool Cooling And Cleaning System 0-SOI-78.01 Rev. 0007 Page 10 of 197 3.0 PRECAUTIONS AND LIMITATIONS A. SFP spills may cause contamination spread and unnecessary radiation exposure. Take immediate correct ive actions including informing Rad Protection. B. Loose items may fall into SFP, and dam age Fuel assembly, racks, etc. Loose items must be secured to personnel or another piece of stationary equipment. C. The SFP water temperature is NOT to be permitted to exceed 150
 
===3.0 PRECAUTIONS===
AND LIMITATIONS A. SFP spills may cause contamination spread and unnecessary radiation exposure. Take immediate correct ive actions including informing Rad Protection. B. Loose items may fall into SFP, and dam age Fuel assembly, racks, etc. Loose items must be secured to personnel or another piece of stationary equipment. C. The SFP water temperature is NOT to be permitted to exceed 150
&deg;F during normal operation and anticipated refueling activities, pursuant to the FSAR. D. The SFP Filter r equires replacing when P of 20 psi or greater, OR Rad Protection discretion (radiation level). E. Exceeding 100 gpm or 140
&deg;F during normal operation and anticipated refueling activities, pursuant to the FSAR. D. The SFP Filter r equires replacing when P of 20 psi or greater, OR Rad Protection discretion (radiation level). E. Exceeding 100 gpm or 140
&deg;F through SFP Demin can damage resins. During refueling outages, flows up to 180 gpm through SFP Demin using the RWPS pumps are acceptable. Ultrasonic flow-meters may be used to determine flows during refueling cleanup F. If Skimmer Loop is in service, ai r may enter skimmer head, causing pump cavitation. SFP level should be checked periodically to ensure proper relation between skimmer head and vortex created by pump suction. G. SFP Skimmer Filter is replaced at 20 psi P. H. When taking suction from Refueling Cavity with RWPS Pump(s), monitor suction press after starting pump(s) due to high loop in suction pipe which could cause loss of pump suction. I. RWPS Filter damage may occur if 200 gpm is exceeded. J. During transfer activity, RWST and SF P levels must be monitored to ensure correct valve alignments and Tech Spec requirements. K. If MCR SFP level alarm is NOT available, local monito ring of level with direct communication to MCR should be established.
&deg;F through SFP Demin can damage resins. During refueling outages, flows up to 180 gpm through SFP Demin using the RWPS pumps are acceptable. Ultrasonic flow-meters may be used to determine flows during refueling cleanup F. If Skimmer Loop is in service, ai r may enter skimmer head, causing pump cavitation. SFP level should be checked periodically to ensure proper relation between skimmer head and vortex created by pump suction. G. SFP Skimmer Filter is replaced at 20 psi P. H. When taking suction from Refueling Cavity with RWPS Pump(s), monitor suction press after starting pump(s) due to high loop in suction pipe which could cause loss of pump suction. I. RWPS Filter damage may occur if 200 gpm is exceeded. J. During transfer activity, RWST and SF P levels must be monitored to ensure correct valve alignments and Tech Spec requirements. K. If MCR SFP level alarm is NOT available, local monito ring of level with direct communication to MCR should be established.
Line 4,518: Line 4,334:
: 6. CHECK SFP area monitors 0-RM 102 and -103 NORMAL [0-M-12].
: 6. CHECK SFP area monitors 0-RM 102 and -103 NORMAL [0-M-12].
STOP and LOCKOUT Fuel Handling Area Exh Fans.
STOP and LOCKOUT Fuel Handling Area Exh Fans.
WBN Unit 1 Dropped or Damaged Fuel or Refueling Cavity Seal Failure 1-AOI-29 Rev. 0003  
WBN Unit 1 Dropped or Damaged Fuel or Refueling Cavity Seal Failure 1-AOI-29 Rev. 0003 3.4 Dropped or Damaged Irradiated Fuel Assembly in Spent Fuel Pit Area (continued)
 
===3.4 Dropped===
or Damaged Irradiated Fuel Assembly in Spent Fuel Pit Area (continued)
Page 15 of 27 Step Action/Expected Response Response Not Obtained
Page 15 of 27 Step Action/Expected Response Response Not Obtained
: 7. CHECK ABI actuated:
: 7. CHECK ABI actuated:
Line 4,590: Line 4,403:
Comments:  
Comments:  


WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 34 of 57  
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 34 of 57 2.7 Prudent Operator Actions (continued)
 
===2.7 Prudent===
Operator Actions (continued)
: 9. Adjusting a S/G PORV to 90% if a known SG tube leak/rupture exists. 10. Early isolation of auxiliary feedwater (AFW) should be performed after positive identification of a ruptured, f aulted, or ruptured and faulted S/G providing the following guidelines are met:
: 9. Adjusting a S/G PORV to 90% if a known SG tube leak/rupture exists. 10. Early isolation of auxiliary feedwater (AFW) should be performed after positive identification of a ruptured, f aulted, or ruptured and faulted S/G providing the following guidelines are met:
: a. Prior to isolation, the steps of 1-E-0 are complete through verification of heat sink. b. For a ruptured S/G, ensure t hat narrow range (NR) level on the affected S/G is >29% and heat sink minimum requirements are met for the unaffected S/Gs. c. For a faulted or ruptured and faul ted S/G, ensure heat sink minimum requirements are met fo r the unaffected S/Gs. 11. When an ATWS event is in progress and after the crew has attempted to open the reactor trip breakers from the second reactor trip hand switch and the trip is unsuccessful, the response should be as follows:
: a. Prior to isolation, the steps of 1-E-0 are complete through verification of heat sink. b. For a ruptured S/G, ensure t hat narrow range (NR) level on the affected S/G is >29% and heat sink minimum requirements are met for the unaffected S/Gs. c. For a faulted or ruptured and faul ted S/G, ensure heat sink minimum requirements are met fo r the unaffected S/Gs. 11. When an ATWS event is in progress and after the crew has attempted to open the reactor trip breakers from the second reactor trip hand switch and the trip is unsuccessful, the response should be as follows:
Line 4,602: Line 4,412:


S/G pressure is approaching 675 psig , a manual MSIV Isolation would anticipate the automatic response and is consistent with procedural guidance and plant design response.
S/G pressure is approaching 675 psig , a manual MSIV Isolation would anticipate the automatic response and is consistent with procedural guidance and plant design response.
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 14 of 57  
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 14 of 57 2.2.3 Foldout Page A. This page presents actions or transitions which are applicable at any time in the given instruction. Upon transition from an instruction, the current instruction's foldout page becomes applicable and use of the previous instruction's foldout page is discontinued. 1. In the control room, the foldout page information is presented on the back of each page of instructions for which there is a foldout page. 2. The information on the foldout page should be continuously monitored to determine when operator action is necessary. B. Transitions to other instructions allo w immediate response to new symptoms as they appear. 2.2.4 Immediate Action Steps Steps that have been designated as "Immedi ate Actions" are contained in three emergency instructions and selected AOIs. These steps are intended to be performed, if necessary, without the written instruction being available.
 
====2.2.3 Foldout====
Page A. This page presents actions or transitions which are applicable at any time in the given instruction. Upon transition from an instruction, the current instruction's foldout page becomes applicable and use of the previous instruction's foldout page is discontinued. 1. In the control room, the foldout page information is presented on the back of each page of instructions for which there is a foldout page. 2. The information on the foldout page should be continuously monitored to determine when operator action is necessary. B. Transitions to other instructions allo w immediate response to new symptoms as they appear. 2.2.4 Immediate Action Steps Steps that have been designated as "Immedi ate Actions" are contained in three emergency instructions and selected AOIs. These steps are intended to be performed, if necessary, without the written instruction being available.
A. Operators are required to be able to complete the intent of immediate operator action steps. B. During immediate operator action steps the operat ors will ensure automatic actions have occurred or initiate signals as appropriate. Diagnostic or repair actions will be delayed until the immediate actions are complete to allow for evaluation of plant response. C. The immediate actions of 1-E-0 will be addressed as follows: 1. Steps 1 through 4 will be performed in order by the OAC and completion, along with any discrepancies, will be communicated to the Procedure  
A. Operators are required to be able to complete the intent of immediate operator action steps. B. During immediate operator action steps the operat ors will ensure automatic actions have occurred or initiate signals as appropriate. Diagnostic or repair actions will be delayed until the immediate actions are complete to allow for evaluation of plant response. C. The immediate actions of 1-E-0 will be addressed as follows: 1. Steps 1 through 4 will be performed in order by the OAC and completion, along with any discrepancies, will be communicated to the Procedure  


Line 4,616: Line 4,423:
* At least one S/G NR level greater than 29%  [39% ADV].  
* At least one S/G NR level greater than 29%  [39% ADV].  
**  GO TO 1-FR-H.1, Loss of Secondary Heat Sink.
**  GO TO 1-FR-H.1, Loss of Secondary Heat Sink.
WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Step    Action/Expected Response Response Not Obtained Page 3 of 17  
WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Step    Action/Expected Response Response Not Obtained Page 3 of 17 3.0 OPERATOR ACTIONS CAUTION If a faulted S/G is NOT needed for RCS cooldown, it should remain isolated during subsequent recovery actions.
 
===3.0 OPERATOR===
ACTIONS CAUTION If a faulted S/G is NOT needed for RCS cooldown, it should remain isolated during subsequent recovery actions.
: 1. ENSURE all MSIVs and MSIV bypasses CLOSED.
: 1. ENSURE all MSIVs and MSIV bypasses CLOSED.
Manually CLOSE valves. IF valves can NOT be closed, THEN  Locally REMOVE power to valves:
Manually CLOSE valves. IF valves can NOT be closed, THEN  Locally REMOVE power to valves:
Line 4,635: Line 4,439:


WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Page 10 of 17 Attachment 1 (Page 1 of 5)
WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Page 10 of 17 Attachment 1 (Page 1 of 5)
Isolation of MSIVs and MSIV Bypass Valves  
Isolation of MSIVs and MSIV Bypass Valves 1.0 INSTRUCTIONS NOTE If additional MCR Operators are av ailable, US may direct perfo rmance of Attachment 2, MSIV and Bypass Valves Fail Open, to be performed in parallel with Attachment 1.
 
1.1 Isolation of MSIVs and MSIV Bypass Valves A. IF any MSIV will NOT close, THEN  GO TO Section 1.2 (Attachment 1).
===1.0 INSTRUCTIONS===
 
NOTE If additional MCR Operators are av ailable, US may direct perfo rmance of Attachment 2, MSIV and Bypass Valves Fail Open, to be performed in parallel with Attachment 1.  
 
===1.1 Isolation===
of MSIVs and MSIV Bypass Valves A. IF any MSIV will NOT close, THEN  GO TO Section 1.2 (Attachment 1).
B. IF any MSIV bypass valve will NOT close, THEN GO TO Section 1.3 (Attachment 1).
B. IF any MSIV bypass valve will NOT close, THEN GO TO Section 1.3 (Attachment 1).
WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Attachment 1 (Page 2 of 5)
WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Attachment 1 (Page 2 of 5)
Line 4,685: Line 4,483:
Date________    INITIALS  
Date________    INITIALS  


===5.2 Establishing===
5.2 Establishing Condenser Vacuum NOTE Since the Condenser Vacuum Exhaust (CVE) effluent must be monitored for a SG tube leak, and the CVE monitors must remain isolated until vacuum is achieved to avoid water damage, the following comp measures must be taken per guidance in System Description N3-2-4002, Section 4.10:
Condenser Vacuum NOTE Since the Condenser Vacuum Exhaust (CVE) effluent must be monitored for a SG tube leak, and the CVE monitors must remain isolated until vacuum is achieved to avoid water damage, the following comp measures must be taken per guidance in System Description N3-2-4002, Section 4.10:
* SG blowdown rad monitors should be placed in service.
* SG blowdown rad monitors should be placed in service.
* Sampling provisions of the ODCM will be used to assess radioactive effluents out of the CVE.
* Sampling provisions of the ODCM will be used to assess radioactive effluents out of the CVE.
Line 4,701: Line 4,498:
Date________    INITIALS  
Date________    INITIALS  


===5.2 Establishing===
5.2 Establishing Condenser Vacuum (continued)  
Condenser Vacuum (continued)  


[9] OPEN 1-FCV-2-255, with 1-HS-2-255, EXH BYPASS COND VAC PMPS [1-M-3].
[9] OPEN 1-FCV-2-255, with 1-HS-2-255, EXH BYPASS COND VAC PMPS [1-M-3].
Line 4,720: Line 4,516:
Date________    INITIALS  
Date________    INITIALS  


===5.2 Establishing===
5.2 Establishing Condenser Vacuum (continued)  
Condenser Vacuum (continued)  


[11.3] OPEN to vent trapped air from pump, THEN  CLOSE the following valve(s):
[11.3] OPEN to vent trapped air from pump, THEN  CLOSE the following valve(s):
Line 4,745: Line 4,540:
Date________    INITIALS  
Date________    INITIALS  


===5.2 Establishing===
5.2 Establishing Condenser Vacuum (continued)  
Condenser Vacuum (continued)  


[15] WHEN vacuum is established (C-9 permissive has been acknowledged), THEN  PLACE 1-RE-90-119, COND VAC PMP AIR EXH RAD MON, in service per 0-SOI-90.02.
[15] WHEN vacuum is established (C-9 permissive has been acknowledged), THEN  PLACE 1-RE-90-119, COND VAC PMP AIR EXH RAD MON, in service per 0-SOI-90.02.
Line 4,758: Line 4,552:
Date________    INITIALS  
Date________    INITIALS  


===5.2 Establishing===
5.2 Establishing Condenser Vacuum (continued)  
Condenser Vacuum (continued)  


NOTE 1) Ammonia is injected into the C ondensate System for pH control. 2) Ammonia injection rate is higher during unit startup; therefore maximizing Seal Water Tank turnover is necessary. Throttling open drain valves in the following step helps reduce ammonia concentration in the general ar ea by raising the turnover rate of the Seal Water Tank volume, which allows ammonia to exit via drain while still entrained in the condensate.
NOTE 1) Ammonia is injected into the C ondensate System for pH control. 2) Ammonia injection rate is higher during unit startup; therefore maximizing Seal Water Tank turnover is necessary. Throttling open drain valves in the following step helps reduce ammonia concentration in the general ar ea by raising the turnover rate of the Seal Water Tank volume, which allows ammonia to exit via drain while still entrained in the condensate.
Line 4,772: Line 4,565:
Date________    INITIALS  
Date________    INITIALS  


===5.2 Establishing===
5.2 Establishing Condenser Vacuum (continued)  
Condenser Vacuum (continued)  


CAUTION Max Hotwell and CST temp is 140
CAUTION Max Hotwell and CST temp is 140
Line 4,788: Line 4,580:
Date________    INITIALS  
Date________    INITIALS  


===5.2 Establishing===
5.2 Establishing Condenser Vacuum (continued)  
Condenser Vacuum (continued)  


[26] ISOLATE continuous drain for any pumps STOPPED: A. 1-DRV-37-572, GLAND SEAL WATER DRAIN - PUMP 1A. ________ B. 1-DRV-37-571, GLAND SEAL WATER DRAIN - PUMP 1B. ________ C. 1-DRV-37-565, GLAND SEAL WATER DRAIN - PUMP 1C. ________
[26] ISOLATE continuous drain for any pumps STOPPED: A. 1-DRV-37-572, GLAND SEAL WATER DRAIN - PUMP 1A. ________ B. 1-DRV-37-571, GLAND SEAL WATER DRAIN - PUMP 1B. ________ C. 1-DRV-37-565, GLAND SEAL WATER DRAIN - PUMP 1C. ________
Line 4,963: Line 4,754:
f I """' O-I ZS 'i Ii 1  
f I """' O-I ZS 'i Ii 1  
......
......
0-PCV-33-4 SERVICE AIR ISOLATION VLV No t e th<ll th e pre ss ur e reqmre s a mamrnl re s et ANN WOW 2C*42f' SERVICE AIR PCV-33-4 CLDS&#xa3;C (1-4!i 911 5 5*2Cl WBN Unit 0 Control Air System 0-SOI-32.01 Rev. 0012 Page 48 of 61 Date________    INITIALS  
0-PCV-33-4 SERVICE AIR ISOLATION VLV No t e th<ll th e pre ss ur e reqmre s a mamrnl re s et ANN WOW 2C*42f' SERVICE AIR PCV-33-4 CLDS&#xa3;C (1-4!i 911 5 5*2Cl WBN Unit 0 Control Air System 0-SOI-32.01 Rev. 0012 Page 48 of 61 Date________    INITIALS 8.4 Manual Alignment of Unit 2 Cons truction Air Compressor to Unit 1 Control/ Service Air Header CAUTION Performance of this section should only be performed with concurrence of US/SM with a degraded or in anticipation of a degr ading control air header pressure.
 
===8.4 Manual===
Alignment of Unit 2 Cons truction Air Compressor to Unit 1 Control/ Service Air Header CAUTION Performance of this section should only be performed with concurrence of US/SM with a degraded or in anticipation of a degr ading control air header pressure.
NOTE 2-ISV-33-543 will requ ire an extension ladder to access i
NOTE 2-ISV-33-543 will requ ire an extension ladder to access i
: t. 2-ISV-33-509 will require a scaffold to be erected.
: t. 2-ISV-33-509 will require a scaffold to be erected.
Line 4,978: Line 4,766:
________ B. WHEN Service Air pressure equals Control Air pressure, THEN  RESET 0-PS-33-4, SERVICE AIR SUPPLY ISOLATION PRESSURE [T7M/708].
________ B. WHEN Service Air pressure equals Control Air pressure, THEN  RESET 0-PS-33-4, SERVICE AIR SUPPLY ISOLATION PRESSURE [T7M/708].
________
________
WBN Unit 0 Control Air System 0-SOI-32.01 Rev. 0012 Page 49 of 61 Date________    INITIALS  
WBN Unit 0 Control Air System 0-SOI-32.01 Rev. 0012 Page 49 of 61 Date________    INITIALS 8.4 Manual Alignment of Unit 2 C onstruction Air Compressor to Unit 1 Control/ Service Air Header (continued)
 
===8.4 Manual===
Alignment of Unit 2 C onstruction Air Compressor to Unit 1 Control/ Service Air Header (continued)
NOTE 0-PCV-33-4, Service Air Supply Isolation from Control Air, will take from 5 to 10 minutes to  travel from full closed to full open position.
NOTE 0-PCV-33-4, Service Air Supply Isolation from Control Air, will take from 5 to 10 minutes to  travel from full closed to full open position.
C. ENSURE ANNUNCIATOR WINDOW 42-F [1-M-2], SERVICE AIR PCV-33-4 CLOSED, NOT LIT. ________
C. ENSURE ANNUNCIATOR WINDOW 42-F [1-M-2], SERVICE AIR PCV-33-4 CLOSED, NOT LIT. ________
Line 5,021: Line 4,806:
Higher X Lower    Question Source:
Higher X Lower    Question Source:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
NPG Standard Department Procedure Conduct of Operations OPDP-1 Rev. 0034 Page 40 of 74  
NPG Standard Department Procedure Conduct of Operations OPDP-1 Rev. 0034 Page 40 of 74 3.8.3 Briefi n gs (continued)
 
====3.8.3 Briefi====
n gs (continued)
: 3. Shift Turnover Brief Shift briefs are conducted at the beginning and middle of each shift. Shift Management directs the briefs. Personnel participation is required; with the exception of control room monitoring that is expected to continue. (See Attachment 3 for Beginning of Shift Brief and Mid-Shift Brief).
: 3. Shift Turnover Brief Shift briefs are conducted at the beginning and middle of each shift. Shift Management directs the briefs. Personnel participation is required; with the exception of control room monitoring that is expected to continue. (See Attachment 3 for Beginning of Shift Brief and Mid-Shift Brief).
* If shift activities do not facilitate 100% crew attendance, a crew member knowledgeable of the watch station status will provide turnover information on that watch station. 4. Event Based Brief Event Based Briefs are conducted as determined by the Unit Supervisor utilizing a format similar to that contained in Attachment 4. Event based briefs ensure that key information and strategies are communicated in a concise manner and that crew members are given an opportunity to share critical information. Attachment 4 is not required to be filled out. It is a general guideline to follow and does not need to be in hand during the brief. 5. Crew Updates During many conditions the communication of important information to the crew may not need a scripted brief. In these cases the "Crew Update" can be used. If a crew member has information that should be known by the majority of crew members he/she passes this information in a concise manner by:
* If shift activities do not facilitate 100% crew attendance, a crew member knowledgeable of the watch station status will provide turnover information on that watch station. 4. Event Based Brief Event Based Briefs are conducted as determined by the Unit Supervisor utilizing a format similar to that contained in Attachment 4. Event based briefs ensure that key information and strategies are communicated in a concise manner and that crew members are given an opportunity to share critical information. Attachment 4 is not required to be filled out. It is a general guideline to follow and does not need to be in hand during the brief. 5. Crew Updates During many conditions the communication of important information to the crew may not need a scripted brief. In these cases the "Crew Update" can be used. If a crew member has information that should be known by the majority of crew members he/she passes this information in a concise manner by:
Line 5,092: Line 4,874:
Higher  X  Lower    Question Source:
Higher  X  Lower    Question Source:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
AFD 3.2.3          Watts Bar-Unit 1 3.2-9          3.2    POWER DISTRIBUTION LIMITS  
AFD 3.2.3          Watts Bar-Unit 1 3.2-9          3.2    POWER DISTRIBUTION LIMITS 3.2.3   AXIAL FLUX DIFFERENCE (AFD)   
 
====3.2.3 AXIAL====
FLUX DIFFERENCE (AFD)   


LCO  3.2.3  The AFD in % flux difference units shall be maintained within the limits specified in the COLR.  
LCO  3.2.3  The AFD in % flux difference units shall be maintained within the limits specified in the COLR.  
Line 5,200: Line 4,979:
WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 47 of 48 Appendix B (Page 15 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES UT5000  1-SI-68-32 2 Y0703A 1-LPF-3-147A 1-SI-3-912 0 Y0704A 1-LPF-3-155B 1-SI-3-912 0 Y0708A 1-LPF-3-163B 1-SI-3-912 0 Y0709A 1-LPF-3-170A 1-SI-3-912 0 Y2203A 1-LPF-90-400 1-ODI-90-15 0  1-ODI-90-18 0  1-SI-0-2 0 Y8000A 0-LPF-27-98 0-PI-ENV-3.1 0 Y8001A 0-TT-27-99 0-PI-ENV-3.1 0  
WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 47 of 48 Appendix B (Page 15 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES UT5000  1-SI-68-32 2 Y0703A 1-LPF-3-147A 1-SI-3-912 0 Y0704A 1-LPF-3-155B 1-SI-3-912 0 Y0708A 1-LPF-3-163B 1-SI-3-912 0 Y0709A 1-LPF-3-170A 1-SI-3-912 0 Y2203A 1-LPF-90-400 1-ODI-90-15 0  1-ODI-90-18 0  1-SI-0-2 0 Y8000A 0-LPF-27-98 0-PI-ENV-3.1 0 Y8001A 0-TT-27-99 0-PI-ENV-3.1 0  


Containment Air Temperature
Containment Air Temperature 3.6.5      Watts Bar-Unit 1 3.6-16      3.6  CONTAINMENT SYSTEMS


====3.6.5 Watts====
3.6.5  Containment Air Temperature
Bar-Unit 1 3.6-16      3.6 CONTAINMENT SYSTEMS
 
====3.6.5 Containment====
Air Temperature


LCO  3.6.5  Containment average air temperature shall be:
LCO  3.6.5  Containment average air temperature shall be:
Line 5,248: Line 5,023:
Higher  Lower X  Question Source:
Higher  Lower X  Question Source:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
Definitions
Definitions 1.1         Watts Bar-Unit 1 1.1-7      Table 1.1-1 (page 1 of 1) MODES    MODE    TITLE    REACTIVITY CONDITION (k eff)  % RATED THERMAL POWER (a)  AVERAGE REACTOR COOLANT TEMPERATURE (F)  1  Power Operation 0.99  > 5  NA 2 Startup  0.99  5 NA 3 Hot Standby
 
===1.1 Watts===
Bar-Unit 1 1.1-7      Table 1.1-1 (page 1 of 1) MODES    MODE    TITLE    REACTIVITY CONDITION (k eff)  % RATED THERMAL POWER (a)  AVERAGE REACTOR COOLANT TEMPERATURE (F)  1  Power Operation 0.99  > 5  NA 2 Startup  0.99  5 NA 3 Hot Standby
  < 0.99 NA  350 4 Hot Shutdown (b)
  < 0.99 NA  350 4 Hot Shutdown (b)
  < 0.99 NA 350 > T avg > 200 5 Cold Shutdown (b)
  < 0.99 NA 350 > T avg > 200 5 Cold Shutdown (b)
Line 5,304: Line 5,076:


LINE VENT IC/755  AZ 30 1 , 2 CLOSED 1-VTV-68-604 IVCVCS LETDOWN RELIEF DISCH DRAIN IC/710  AZ 60 1 CLOSED 1-DRV-68-605 IV 1 VERIFY cap/blind flange installed 2 May be N/A'd if Reactor Vessel Head removed.
LINE VENT IC/755  AZ 30 1 , 2 CLOSED 1-VTV-68-604 IVCVCS LETDOWN RELIEF DISCH DRAIN IC/710  AZ 60 1 CLOSED 1-DRV-68-605 IV 1 VERIFY cap/blind flange installed 2 May be N/A'd if Reactor Vessel Head removed.
Annulus PRESSURIZER RELIEF TANK GAS ANALYZER SUPPLY OC/716  AZ 315 OPERABLE OPEN 1-FCV-68-307 IVCONTROL AIR ISOLATION VALVE TO 1-FCV-68-307 OC/716  AZ 315 OPEN  1-ISV-32-3499 IV BIT Room PRESSURIZER RELIEF TANK NITROGEN SUP PRESS CNTL A4X/713  BIT RM OPERABLE 1-PCV-68-304 IVCONTROL AIR ISOLATION VALVE TO 1-PCV-68-304 A4X/713  BIT RM OPEN  1-ISV-32-3199 IVPRESSURIZER RELIEF TANK NITROGEN SUP FLOW CNTL A4X/713  BIT RM OPEN  1-FCV-68-305 IVCONTROL AIR ISOLATION VALVE TO 1-FCV-68-305 A4X/713  BIT RM OPEN  1-ISV-32-3198 IV1-FCV-68-304 CNTL ISOL A4X/713  BIT RM OPEN 1-RTV-68-307A IVPRESSURIZER RELIEF TANK N 2 SUP HDR TEST VENT A4X/713  BIT RM CLOSED 1-TV-68-847 IV WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 7 of 50  
Annulus PRESSURIZER RELIEF TANK GAS ANALYZER SUPPLY OC/716  AZ 315 OPERABLE OPEN 1-FCV-68-307 IVCONTROL AIR ISOLATION VALVE TO 1-FCV-68-307 OC/716  AZ 315 OPEN  1-ISV-32-3499 IV BIT Room PRESSURIZER RELIEF TANK NITROGEN SUP PRESS CNTL A4X/713  BIT RM OPERABLE 1-PCV-68-304 IVCONTROL AIR ISOLATION VALVE TO 1-PCV-68-304 A4X/713  BIT RM OPEN  1-ISV-32-3199 IVPRESSURIZER RELIEF TANK NITROGEN SUP FLOW CNTL A4X/713  BIT RM OPEN  1-FCV-68-305 IVCONTROL AIR ISOLATION VALVE TO 1-FCV-68-305 A4X/713  BIT RM OPEN  1-ISV-32-3198 IV1-FCV-68-304 CNTL ISOL A4X/713  BIT RM OPEN 1-RTV-68-307A IVPRESSURIZER RELIEF TANK N 2 SUP HDR TEST VENT A4X/713  BIT RM CLOSED 1-TV-68-847 IV WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 7 of 50 2.2 Developmental References (continued)
 
===2.2 Developmental===
References (continued)
I. RCI-128, ALARA Pr ogram Implementation J. TI-134, Control of Portable Two Way Radios K. TI-229, Temporary Shielding Program L. WBN FSAR Questions 22.26, 212.116, and 212.129  3.0 PRECAUTIONS AND LIMITATIONS 3.1 General Precautions and Limitations A. All access portals (Airlocks el 757/716, equipment hatches el 757, and Annulus el 713) to the Containment buildi ng SHALL be controlled to prevent unauthorized entry while in Modes 1 through
I. RCI-128, ALARA Pr ogram Implementation J. TI-134, Control of Portable Two Way Radios K. TI-229, Temporary Shielding Program L. WBN FSAR Questions 22.26, 212.116, and 212.129  3.0 PRECAUTIONS AND LIMITATIONS 3.1 General Precautions and Limitations A. All access portals (Airlocks el 757/716, equipment hatches el 757, and Annulus el 713) to the Containment buildi ng SHALL be controlled to prevent unauthorized entry while in Modes 1 through
: 4. Radiation Protection (RP) shall maintain positive access control of Containment and Annulus in accordance with RP procedures. B. All entries into Containment or Annulus while in Modes 1 through 4 are authorized by completion of applic able sections of Appendix A, Containment/Annulus Entry Authorization. C. One Appendix A is required for eac h area entered (Upper Containment, Lower Containment, Annulus). Appendix A's ar e typically issued for a 24 hour period only. However the SM may approv e an extension beyond 24 hours.
: 4. Radiation Protection (RP) shall maintain positive access control of Containment and Annulus in accordance with RP procedures. B. All entries into Containment or Annulus while in Modes 1 through 4 are authorized by completion of applic able sections of Appendix A, Containment/Annulus Entry Authorization. C. One Appendix A is required for eac h area entered (Upper Containment, Lower Containment, Annulus). Appendix A's ar e typically issued for a 24 hour period only. However the SM may approv e an extension beyond 24 hours.
Line 5,357: Line 5,126:
RCDP-15, Dosimetry Operations Proposed references to be provided:
RCDP-15, Dosimetry Operations Proposed references to be provided:
None  Learning Objective: 3-OT-RAD0004 , Personnel Dosimetry 1. Describe the operation of Optically Stimulated Luminescent Dosimeters (IER 11-3, Having a solid understanding of plant design, engineering principles, and sciences.) Cognitive Level:      Higher    Lower  X Question Source:      New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
None  Learning Objective: 3-OT-RAD0004 , Personnel Dosimetry 1. Describe the operation of Optically Stimulated Luminescent Dosimeters (IER 11-3, Having a solid understanding of plant design, engineering principles, and sciences.) Cognitive Level:      Higher    Lower  X Question Source:      New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
NPG Standard Department Procedure Dosimetry Operations RCDP-15 Rev. 0000 Page 6 of 34  
NPG Standard Department Procedure Dosimetry Operations RCDP-15 Rev. 0000 Page 6 of 34 3.2.2 Ordering (continued)
 
====3.2.2 Ordering====
(continued)
: 1. The label example is shown below NOTE The fields shown in the example and described below may be modified to meet a specific licensee or department needs. For example, the Emergency Planning Tennessee Emergency Management Agency dosimeters will have a different label scheme than the personnel dosimeters for occupationally exposed workers. A= Customer account # specific to each site B = TVA site C= TVA D = worker first name  
: 1. The label example is shown below NOTE The fields shown in the example and described below may be modified to meet a specific licensee or department needs. For example, the Emergency Planning Tennessee Emergency Management Agency dosimeters will have a different label scheme than the personnel dosimeters for occupationally exposed workers. A= Customer account # specific to each site B = TVA site C= TVA D = worker first name  


Line 5,394: Line 5,160:
Higher    Lower  X  Question Source:
Higher    Lower  X  Question Source:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
New X  Modified Bank Bank      Question History: New question for the 2015-301 NRC RO Exam Comments:
NPG Standard Programs and Processes Radiological Controls NPG-SPP-05.1 Rev. 0005 Page 14 of 47  
NPG Standard Programs and Processes Radiological Controls NPG-SPP-05.1 Rev. 0005 Page 14 of 47 3.2.3 Exposure Control (continued)
 
====3.2.3 Exposure====
Control (continued)
TABLE 1 ADMINISTRATIVE DOSE LEVEL PROGRAM Dose Equivalent (Rem) Requirement Authorization to exceed (signatures)
TABLE 1 ADMINISTRATIVE DOSE LEVEL PROGRAM Dose Equivalent (Rem) Requirement Authorization to exceed (signatures)
Up to 0.5 TEDE (or 1.5 LDE or 5.0 SDE) at TVA  Statement of current year dose and previous years dose signed by individual Not applicable Up to 1.0 TEDE (or 3.0 LDE or 10 SDE) all sources NRC FORM-4 or equivalent to document current year and previous years dose equivalent Not applicable To exceed 1.0 TEDE (or 3.0 LDE or 10 SDE) all sources  Same as above RPM/RSO To exceed   
Up to 0.5 TEDE (or 1.5 LDE or 5.0 SDE) at TVA  Statement of current year dose and previous years dose signed by individual Not applicable Up to 1.0 TEDE (or 3.0 LDE or 10 SDE) all sources NRC FORM-4 or equivalent to document current year and previous years dose equivalent Not applicable To exceed 1.0 TEDE (or 3.0 LDE or 10 SDE) all sources  Same as above RPM/RSO To exceed   
Line 5,452: Line 5,215:
==1.0 INTRODUCTION==
==1.0 INTRODUCTION==


===1.1 Purpose===
1.1 Purpose This Instruction provides guidance fo r the use of the Emergency Operating Instructions (EOIs) and Abnormal Operat ing Instructions (AOIs) and ensures standardized implementation of the EOIs. 1.2 Applicability A. These guidelines will be followed by the Main Control Room (MCR) crew during emergency conditions as well as training exercises. B. This Instruction provides the rules of usage for the set of Emergency Operating Instructions including the Status Trees. The EOI set includes Emergency Instructions (E and ES), Emergency Contin gency Actions (ECA), and Function Restoration Instructions (FR). C. The rules explained in Section 2.2.7 (Operator Performance of Action Steps) also apply to Abnormal Operating Instru ctions (AOIs) that are written in the same two column format. Section 2.3 (Types of Action Steps in Two Column Instructions) is descriptive of EOIs and AOIs. 2.0 DETAILS The EOIs direct operator actions to the most urgent safety or operational conditions.
This Instruction provides guidance fo r the use of the Emergency Operating Instructions (EOIs) and Abnormal Operat ing Instructions (AOIs) and ensures standardized implementation of the EOIs. 1.2 Applicability A. These guidelines will be followed by the Main Control Room (MCR) crew during emergency conditions as well as training exercises. B. This Instruction provides the rules of usage for the set of Emergency Operating Instructions including the Status Trees. The EOI set includes Emergency Instructions (E and ES), Emergency Contin gency Actions (ECA), and Function Restoration Instructions (FR). C. The rules explained in Section 2.2.7 (Operator Performance of Action Steps) also apply to Abnormal Operating Instru ctions (AOIs) that are written in the same two column format. Section 2.3 (Types of Action Steps in Two Column Instructions) is descriptive of EOIs and AOIs. 2.0 DETAILS The EOIs direct operator actions to the most urgent safety or operational conditions.
This Instruction provides the rules of priority and usage which, when properly applied, will direct the operat or through the EOI network.
This Instruction provides the rules of priority and usage which, when properly applied, will direct the operat or through the EOI network.
The rules of usage define the intended use of two column format instru ctions. The rules of priority define the hierarchy of the individual instructi ons and must be followed to ensure proper implementation. 2.1 Implementation of Emergency Operating Instructions Implementation of the EOI set is limited to two specific conditions:
The rules of usage define the intended use of two column format instru ctions. The rules of priority define the hierarchy of the individual instructi ons and must be followed to ensure proper implementation. 2.1 Implementation of Emergency Operating Instructions Implementation of the EOI set is limited to two specific conditions:
A. Any time a complete loss of both tr ains of Shutdown Power occurs the MCR Team enters the EOI network by implementing 1-ECA-0.0, Loss of Shutdown Power. This includes any time during the performance of any other instruction. B. Any time a reactor trip or safety injection occurs, or is required, the MCR Team enters the EOI network by implementing 1-E-0, Reactor Trip or Safety Injection.
A. Any time a complete loss of both tr ains of Shutdown Power occurs the MCR Team enters the EOI network by implementing 1-ECA-0.0, Loss of Shutdown Power. This includes any time during the performance of any other instruction. B. Any time a reactor trip or safety injection occurs, or is required, the MCR Team enters the EOI network by implementing 1-E-0, Reactor Trip or Safety Injection.
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 14 of 57  
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 14 of 57 2.2.3 Foldout Page A. This page presents actions or transitions which are applicable at any time in the given instruction. Upon transition from an instruction, the current instruction's foldout page becomes applicable and use of the previous instruction's foldout page is discontinued. 1. In the control room, the foldout page information is presented on the back of each page of instructions for which there is a foldout page. 2. The information on the foldout page should be continuously monitored to determine when operator action is necessary. B. Transitions to other instructions allo w immediate response to new symptoms as they appear. 2.2.4 Immediate Action Steps Steps that have been designated as "Immedi ate Actions" are contained in three emergency instructions and selected AOIs. These steps are intended to be performed, if necessary, without the written instruction being available.
 
====2.2.3 Foldout====
Page A. This page presents actions or transitions which are applicable at any time in the given instruction. Upon transition from an instruction, the current instruction's foldout page becomes applicable and use of the previous instruction's foldout page is discontinued. 1. In the control room, the foldout page information is presented on the back of each page of instructions for which there is a foldout page. 2. The information on the foldout page should be continuously monitored to determine when operator action is necessary. B. Transitions to other instructions allo w immediate response to new symptoms as they appear. 2.2.4 Immediate Action Steps Steps that have been designated as "Immedi ate Actions" are contained in three emergency instructions and selected AOIs. These steps are intended to be performed, if necessary, without the written instruction being available.
A. Operators are required to be able to complete the intent of immediate operator action steps. B. During immediate operator action steps the operat ors will ensure automatic actions have occurred or initiate signals as appropriate. Diagnostic or repair actions will be delayed until the immediate actions are complete to allow for evaluation of plant response. C. The immediate actions of 1-E-0 will be addressed as follows: 1. Steps 1 through 4 will be performed in order by the OAC and completion, along with any discrepancies, will be communicated to the Procedure  
A. Operators are required to be able to complete the intent of immediate operator action steps. B. During immediate operator action steps the operat ors will ensure automatic actions have occurred or initiate signals as appropriate. Diagnostic or repair actions will be delayed until the immediate actions are complete to allow for evaluation of plant response. C. The immediate actions of 1-E-0 will be addressed as follows: 1. Steps 1 through 4 will be performed in order by the OAC and completion, along with any discrepancies, will be communicated to the Procedure  


Reader. 2. The BOP will acknowledge alarms as nec essary to reduce the noise level, and perform backup verification of steps 1 through 4. 3. Procedure Reader will read the immediate action high level step. The OAC will confirm the high level step by verbalizing the low level steps to the procedure reader. 4. When reentering 1-E-0 fr om another EOI, the first four high level steps must be reconfirmed, it is NOT necessary to reperform each low level action.
Reader. 2. The BOP will acknowledge alarms as nec essary to reduce the noise level, and perform backup verification of steps 1 through 4. 3. Procedure Reader will read the immediate action high level step. The OAC will confirm the high level step by verbalizing the low level steps to the procedure reader. 4. When reentering 1-E-0 fr om another EOI, the first four high level steps must be reconfirmed, it is NOT necessary to reperform each low level action.
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 15 of 57  
WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 15 of 57 2.2.4 Immediate Action Steps (continued)
 
====2.2.4 Immediate====
Action Steps (continued)
D. The immediate actions for 1-FR-S.1, Nuclear Power Generation/ATWS, and 1-ECA-0.0, Loss of Shutdown Power, are consistent wit h the initial actions of 1-E-0, Reactor Trip or Safety Inje ction, and will be performed in the same manner. E. The immediate actions contained in Abnormal Operating Instructions will be performed in order and completion, along with discrepancies, will be communicated to t he Procedure Reader. 1. 1-AOI-2, Malfunction of Reactor Control System 2. 1-AOI-4, Nuclear Instrumentation Malfunctions
D. The immediate actions for 1-FR-S.1, Nuclear Power Generation/ATWS, and 1-ECA-0.0, Loss of Shutdown Power, are consistent wit h the initial actions of 1-E-0, Reactor Trip or Safety Inje ction, and will be performed in the same manner. E. The immediate actions contained in Abnormal Operating Instructions will be performed in order and completion, along with discrepancies, will be communicated to t he Procedure Reader. 1. 1-AOI-2, Malfunction of Reactor Control System 2. 1-AOI-4, Nuclear Instrumentation Malfunctions
: 3. AOI-18, Malfunction of Pressu rizer Pressure Control System 2.2.5 Sequence of Step Performance The EOI steps have been written in a specified sequence. This sequence is designed to ensure timely operator action for the most limiting evolution.
: 3. AOI-18, Malfunction of Pressu rizer Pressure Control System 2.2.5 Sequence of Step Performance The EOI steps have been written in a specified sequence. This sequence is designed to ensure timely operator action for the most limiting evolution.
Line 5,531: Line 5,287:
HPFP:  1A-A  2-A-A  1B-B  2B-B  DGFP:  J. ANNOUNCE location of Incident Comm and Post over PA and radio. K. IF a confirmed fire is located in ANY of the following:
HPFP:  1A-A  2-A-A  1B-B  2B-B  DGFP:  J. ANNOUNCE location of Incident Comm and Post over PA and radio. K. IF a confirmed fire is located in ANY of the following:
: 1. Control Building  2. Auxiliary Building  3. Reactor Building  4. Annulus  THEN NOTIFY ALL AUOs to report with gear and Standby for AOI-30.2.
: 1. Control Building  2. Auxiliary Building  3. Reactor Building  4. Annulus  THEN NOTIFY ALL AUOs to report with gear and Standby for AOI-30.2.
WBN Unit 1 Plant Fires 1-AOI-30.1 Rev. 0002  
WBN Unit 1 Plant Fires 1-AOI-30.1 Rev. 0002 3.0 OPERATOR ACTIONS (continued)
 
===3.0 OPERATOR===
ACTIONS (continued)
Page 5 of 11 Step Action/Expected Response Response Not Obtained
Page 5 of 11 Step Action/Expected Response Response Not Obtained
: 5. REFER TO SOI-13.01, Fire Detection System, for ventilation systems required and USE Ap pendix A of this Instruction, as necessary, for further guidance.
: 5. REFER TO SOI-13.01, Fire Detection System, for ventilation systems required and USE Ap pendix A of this Instruction, as necessary, for further guidance.

Revision as of 10:38, 7 May 2019

301 Draft RO Written Exam
ML15259A290
Person / Time
Site: Watts Bar  Tennessee Valley Authority icon.png
Issue date: 09/16/2015
From:
Division of Reactor Safety II
To:
Tennessee Valley Authority
References
Download: ML15259A290 (540)
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{{#Wiki_filter:NRC Exam Legend Times are HH:MM:SS [ 00:00:00 - 23:59:59 ] ALARMS WINDOWS: LIT DARK LIGHT INDICATIONS: COLOR LIT DARK RED GREEN WHITE

1. In accordance with 1-E-0, Reactor Trip or Sa fety Injection, which ONE of the following describes how the OAC ENSURES the turbine is tripped and the required action(s) if the OAC cannot ENSURE the Turbine Trip?

The OAC is required to ENSURE Turbine Trip using ____(1)____. If the OAC cannot ENSURE Turbine Trip , 1-E-0 directs them to ____(2)____. A. (1) indication that all tu rbine stop valves are CLOSED (2) TRIP turbine locally B. (1) indication that all tu rbine stop valves are CLOSED (2) CLOSE MSIVs and bypasses C. (1) a GREEN light indication at 1-HS-47-24, Turbine Trip (2) TRIP turbine locally D. (1) a GREEN light indication at 1-HS-47-24, Turbine Trip (2) CLOSE MSIVs and bypasses

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: While it is co rrect that the OAC should che ck that the stop valves are closed, it is not correct that 1-E-0 directs that the turbine be tripped locally. B. Correct: 1-E-0 does require that the OAC check that all turbine stop valves are closed. Also, 1-E-0 does direct that the MSIVs and bypasses be closed given that the tu rbine failed to trip.

C. Incorrect: As mentioned in the distract or analysis for A, it is inappropriate to regard the indication on 1-HS-47-24 as the indication for turbine trip. Additionally, 1-E-0 does not mention tr ipping the turbine locally. It is plausible for an applicant to believe that it would be as this is the

response not obtained for an immediat e action in 1-ECA-0.0. Such procedure dictates: IF turbine can NOT be tripped, THEN: TRIP turbine locally: 1) TRIP from front standard. Again, this is a fairly commonly observed mistake in which either an ILT student or a LOR participant will during the immediate actions of 1-E-0, dispatch an AUO to trip the turbine upon the failure of a turbine trip.

D. Incorrect: In accordance with 1-E-0, step 2, the OAC is required to ENSURE Turbine Trip: All turbine stop valves CLOSED. 1-E-0 does not mention checking 1-HS-47-24. It is plausible for an applicant to

believe that 1-HS-47-24 is required to be checked as a green light on such is a prime indication of a turbine trip. Additionally, it is a common bad practice for both ILT students as well as LOR participants to rely on the green light on 1-HS-47-24 as an indication of a turbine trip. Often times, persons being evaluated will miss the fact that the turbine did NOT trip because they simply look for the green light indication. The distractor is also plausible because 1-E-0 does require (as a RNO for step 2) the OAC to RUNBACK tu rbine manually OR CLOSE MSIVs and bypasses.

Question Number: 1 Tier: 1 Group: 1

K/A: 007 Reactor Trip EA1 Ability to operate and monitor the following as they apply to a reactor

trip: EA1.01 T/G controls

Importance Rating: 3.7 3.4

10 CFR Part 55: (CFR 41.7 / 45.5 / 45.6)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to demonstrate that they can correctly monitor the T/G controls during a reactor trip and subsequently operate those controls during the reactor trip response. Technical

Reference:

1-E-0, Reactor Trip and Safety Injection 1-ECA-0.0, Loss of Shutdown Power .JPG picture of 1-HS-74-24

Proposed references to be provided: None Learning Objective: 3-OT-EOP0000, Emergency Instructions Reactor Trip Or Safety Injection, 1-E-0, ES-0.1, 1-ES-0.2, ES-0.3 and ES-0.4

23. CITE the immediate action steps including RNO for 1-E-0. Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

WBN Unit 1 Reactor Trip or Safety Injection 1-E-0 Rev. 0005 Step Action/Expected Response Response Not Obtained Page 3 of 45 3.0 OPERATOR ACTIONS NOTE

  • Steps 1 thru 4 are IMMEDIATE ACTION STEPS
  • Status Trees / SPDS should be monitored when transitioned to another instruction.
1. ENSURE reactor trip:
  • Reactor trip and bypass breakers OPEN.
  • RPIs at bottom of scale.
  • Neutron flux DROPPING.

Manually TRIP reactor. IF reactor will NOT trip, THEN ** GO TO 1-FR-S.1, Nuclear Power Generation / ATWS.

2. ENSURE Turbine Trip:
  • All turbine stop valves CLOSED.

Manually TRIP turbine. IF turbine will NOT trip, THEN

  • RUNBACK turbine manually OR
  • CLOSE MSIVs and bypasses LOW LOAD LIMIT % A GC ELERA T I O N RA T E RP M PE R M I N lV-1 G V-1 G IJ.J TN O P E ii O PEl4 O PEii O P E t* TV-I G V-1 G V.J N...S c IDSED c LO C ED e LO C ED e l.O O!D EH C CO NTR O L 1-XX-4 7-1 000 TURBINE MODES I vi I REFEREN C E CO NT ROL V*, T C H T f'Vl;l l S TV T O G V HIGH LOAD LIMIT % W E Kie W E JOO" uurr wrr T
  • VALVE POSITION LIMIT E R L OAD RATE % PER MI N lV-l O P Et* O PEl4 G'J-1 lV-:? CWC EO C LO C EO

TURBINE TRIP I TRIP WBN Unit 1 Loss of Shutdown Power 1-ECA-0.0 Rev. 0002 Step Action/Expected Response Response Not Obtained Page 4 of 26

2. ENSURE turbine trip:
  • All turbine stop valves CLOSED.

Manually TRIP turbine. IF turbine can NOT be tripped, THEN: a. CLOSE turbine valves, OR b. CLOSE MSIVs and bypasses, OR c. TRIP turbine locally:

1) TRIP from front standard.
2) STOP and PULL TO LOCK both EHC pumps. [Turb Bldg

T4J/745] 3. ENSURE RCPs STOPPED, AND Momentarily PLACE Handswitches in STOP to break seal-in.

2. Given the following conditions: - Unit 2 is at 100% power.

Subsequently: - PZR pressure is rapidly LOWERING - The Reactor was MANUALLY tripped. - PZR level is rapidly RISING Which ONE of the following descr ibes the accident in progress? A. Ejected control rod B. SBLOCA on a RCS cold-leg C. Feedline break inside of containment D. PZR Level Detector Condensing Pot Rupture

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: Plausible if candidate thought that the LOCA on the top of Rx vessel would cause PZR level to rise, however for all RCS LOCAs, with the exception of PZR vapor space LOCAs, PZR level would lower during the initial phase of the accident. B. Incorrect: Plausible since SBLOCA will cause RCS pressure to decrease while not initially affecting RCS temperature right after a Rx trip. However PZR level would be rapidly LOWERING not RISING. C. Incorrect: Plausible since a Feed line break would cause a rapidly LOWERING PZR pressure and after SG blowdown PZR level would rise but not initially while water in the SG is blowing down. Also RCS temperature would be rapidly LOWERING during initial phase of accident. D. Correct: PZR vapor space LOCAs are the only LOCAs that cause the indicated PZR level to rise rapidly while RCS pressure continues to rapidly decrease.

Question Number: 2 Tier: 1 Group: 1

K/A: 008 Pressurizer Vapor Space Accident AK2.02 Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Sensors and Detectors Importance Rating: 2.7*/2.7

10 CFR Part 55: (CFR 41.7) 10CFR55.43.b: Not applicable K/A Match: This question matches the K/A by asking the candidate how the level detectors would respond to a Pressurizer Vapor Space accident.

Technical

Reference:

1-E-0, Reactor Trip and Safety Injection 1-ECA-0.0, Loss of Shutdown Power .JPG picture of 1-HS-74-24

Proposed references to

be provided: None Learning Objective: 3-OT-TAA013, LOSS OF COOLANT ACCIDENTS 1. Describe the dynamic behavior of the reactor, from a thermodynamic and hydraulic point of view, following a loss of coolant accident (LOCA) for the following categories: e. Pressurizer vapor space break. Cognitive Level: Higher X Lower Question Source: New Modified Bank Bank X Question History: Bank question from the WBN Sep 2010 exam. Comments:

3. Given the following conditions: - Unit 1 has experienced a SBLOCA. - RCS pressure is 1700 psig and STABLE - Containment pressure peaked at 1.5 psig. - Safety Injection Pump 1A-A failed to AUTOMATICALLY start.

Which ONE of the following comp letes the statements listed below?

In accordance with 1-E-1, Loss of Reactor or Secondary Coolant, the control room staff will control SG NR levels between ____(1)____ and 50%. Taking ONLY the action to ____(2)____ will RAISE ECCS injection flow rate. A. (1) 29% (2) LOWER the #2 SG PORV setpoint B. (1) 39% (2) LOWER the #2 SG PORV setpoint C. (1) 29% (2) MANUALLY start the Safety Injection Pump 1A-A D. (1) 39% (2) MANUALLY start the Safety Injection Pump 1A-A

CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: In accordance with step 5 of 1-E-1, the operators will CONTROL intact SG levels between 29% and 50% [39% and 50% ADV]. As seen in the WOG background document for 1-E-1 on page 58 (step 3 in the WOG generic E-1 is step 5 in WB NP's 1-E-1), the purpose of maintaining adequate SG level is To ensure adequate feed flow or SG

inventory to ensure a secondary heat sink for small and intermediate

size LOCAs and secondary break accidents. Lowering a SG PORV setpoint will cause the PORV to open mo re to control SG pressure at a lower value. Because the S/Gs remain coupled to the RCS (as this is a SBLOCA), the depressurization of the SG will yield an RCS cooldown and thus depressurization. Because of this, injection flow from the centrifugal charging pumps will increase. RCS pressure is stated to be at 1700 psig. The great est pressure at which the safety injection pumps could have any forwar d flow is that of the Maximum Composite Curve which is found in WBN-SDD-N3-63-4001, Safety Injection System. This curve show s that the SIPs reach shutoff head at 3560 feet/H20 or 1566 psig. An it em which must be considered is the fact that 1-E-0, Reactor Trip or Safety Injection contains the step: "7. MONITOR ECCS operation: f. RCS pressure greater than 1650 psig." The response not obt ained for such step is to "f. ENSURE SI pump flow." Therefore, the question places the applicant not only above the design (as tested) shutoff head of the SIP but also above the Westinghouse EOP setpoint value used in 1-E-0 as well (within the readability of the MCR wide range pressure meter which would be used during an accident).

If the depressurization of the S/Gs continued, the running B train SIP would begin to inject below that value. B. Incorrect: While it is corre ct that LOWERING the #2 SG would increase the injection flow of the ECCS, it is not correct to maintain a SG level band of between 39% and 50%. Such band would only be required if adverse setpoints were warranted on account of a B (containment pressure of 2.8 psig) It is plausible to believe that. B occurred at 1.5 psig-the automatic setpoint of safety injection. It is also plausible to believe that containment pressure rose above 2.8 psig, caused the B and then dropped back to 1.5 psig. Howe ver, the stem st ipulates that pressure peaked at 1.5 psig. C. Incorrect: No further ECCS injection w ould result as RCS pressure is above the shutoff head of the SIP. It is plausib le to believe that starting the SIP would increase ECCS injection be cause doing so would certainly cause such result if done below the shutoff head of the pump. D. Incorrect: As previously mentioned, it is both Incorrect but plausible to believe that SG level should be maintain ed between 39% and 50% and that starting the A SIP would increase ECCS flow.

Question Number: 3 Tier: 1 Group: 1

K/A: 009 Small Break LOCA EK2 Knowledge of the interrelations between the small break LOCA and the following: EK2.03 SGs Importance Rating: 3.0 3.3*

10 CFR Part 55: (CFR 41.7 / 45.7) 10CFR55.43.b: Not applicable

K/A Match: K/A is matched because the applicant is required to demonstrate understanding of the interrelations between the SBLOCA and the SG in both halves of the question. Firstly, the applicant must understand that the SBLOCA requires that a certain level band be maintained in the S/G. Secondly, the applicant must be able to understand the effect that changing pressure in the SG has on the primary side

parameters. Technical

Reference:

1-E-1, Loss of Reactor or Secondary Coolant WBN-SDD-N3-63-4001, Safety Injection System WOG background document for 1-E-1 Proposed references to

be provided: None Learning Objective: 3-OT-EOP0100, 1-E-1, Loss Of Reactor Or Secondary Coolant

5. EXPLAIN the purpose and basis of each step of 1-E-

1, 1-ES-1.1, ES-1.2, ES-1.3 and ES-1.4.

8. Given a set of plant conditions IMPLEMENT the Action Steps, RNOs, Foldout Pages, Notes, and Cautions of 1-E-1, 1-ES-1.1, ES-1.2, ES-1.3, and ES-1.4 Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam

Comments:

WBN Unit 1 Loss of Reactor or Secondary Coolant1-E-1 Rev. 0004 Step Action/Expected Response Response Not Obtained Page 5 of 20

5. MAINTAIN Intact S/G NR levels:
a. MONITOR levels greater than 29% [39% ADV].
a. MAINTAIN total feed flow greater than 410 gpm UNTIL level is

greater than 29% [39% ADV]

in at least one S/G.

b. CONTROL intact S/G levels between 29% and 50%

[39% and 50% ADV].

b. IF level in any intact S/G continues to rise without feed flow, THEN ** GO TO 1-E-3, Steam Generator Tube Rupture.
6. CHECK secondary radiation:
  • S/G discharge monitors NORMAL.
  • Condenser vacuum exhaust rad monitors NORMAL.
  • S/G blowdown rad monitor recorders NORMAL trend prior to isolation.

IF rad monitors NOT available, THEN a. NOTIFY RADPROT to survey main steamlines and S/G

blowdown lines.

b. NOTIFY Chemistry to sample S/G activity.

IF radiation is high, THEN ** GO TO 1-E-3, Steam Generator Tube Rupture. STEP DESCRIPTION TABLE FOR E-1 Step _3_ STEP: Check Intact SG Levels PURPOSE: o To ensure adequate feed flow or SG inventory to ensure a secondary heat sink for small and intermediate size LOCAs and secondary break accidents o To provide a positive static head of water to prevent primary to secondary leakage for a large LOCA BASIS: The minimum feed flow requirement satisfies the feed flow requirement of the Heat Sink Status Tree until level in at least one SG is restored into the narrow range. Narrow range level is reestablished in all SGs to maintain symmetric cooling of the RCS. The control range ensures adequate inventory with level readings on span. The transition to E-3, STEAM GENERATOR TUBE RUPTURE, responds to an increasing level which would be observed following a SGTR. Even though a secondary heat sink is not required for large LOCAs, it is beneficial to maintain SG narrow range levels on span to ensure a positive static head of water between the secondary and primary sides of the SG since these SGs will eventually be depressurized (either later in this guideline or by natural heat losses). Water level in the SGs will prevent (or minimize) primary to secondary leakages from the intact SGs which may have leakage paths that are within the limits of the Technical Specifications. ACTIONS: o Determine if SG narrow range level is greater than (M.02)% [(M.03)% for adverse containment] o Determine if SG narrow range level in any SG increases in an uncontrolled manner o Maintain total feed flow greater than (S.02) gpm until narrow range level greater than (M.02)% [(M.03)% for adverse containment] in at least one SG o Control feed flow to maintain narrow range level between (M.02)% [(M.03)% for adverse containment] and 50% o Transfer to E-3, STEAM GENERATOR TUBE RUPTURE, step 1 INSTRUMENTATION: o SG narrow range level indication o Total feed flow indication o Feed flow control valves position indication E-1 Background HElBG. doc 52 HP-Rev. 2, 4/30/2005 WBN Unit 1 Reactor Trip or Safety Injection 1-E-0 Rev. 0005 Appendix A (Page 3 of 11) Equipment Verification Page 18 of 45 Step Action/Expected Response Response Not Obtained

7. MONITOR ECCS operation: a. Charging pumps RUNNING. a. Manually START charging pumps. b. Charging pump alignment:
  • RWST outlets 1-LCV-62-135 and 1-LCV-62-136 OPEN.
  • VCT outlets 1-LCV-62-132 and 1-LCV-62-133 CLOSED.
  • Charging 1-FCV-62-90 and 1-FCV-62-91 CLOSED.
b. ENSURE at least one valve in each set aligned. c. RHR pumps RUNNING. c. Manually START RHR pumps. d. SI pumps RUNNING. d. Manually START SI pumps. e. BIT alignment:
  • Outlets 1-FCV-63-25 and 1-FCV-63-26 OPEN.
  • Flow thru BIT.
e. ENSURE at least one valve aligned, and flow thru BIT. f. RCS pressure greater than 1650 psig.
f. ENSURE SI pump flow.

IF RCS press drops to less than 150 psig, THEN ENSURE RHR pump flow.

WBN System Description Document SAFETY INJECTION SYST EM WBN-SDD-N3-63-4001 Rev. 0030 Page 119 of 159 8.2 Figure 2 - SIP Composite Curve

4. Which ONE of the following describes the method used to remove decay heat 5 minutes after a LBLOCA has occurred?

5 minutes after the LBLOCA occurs

decay heat is removed by ________.

A. Reflux boiling B. RCS Hot Leg recirculation C. Natural circulation and SG steaming D. ECCS injection and inv entory loss from the break

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: As seen in the Westinghous e E-1 background document, reflux boiling is credited for heat removal during a LO CA of a size greater than 1" in diameter to less than 13 1/2" (1 square foot) in diameter. For LOCAs in excess of these dimensions (i.e. LBLOCAs), the secondary plant is not credited for any heat removal. B. Incorrect: While it is true that during a LBLOCA decay heat removal is afforded by the ECCS; the use of Hot leg recirculation will not be made until 3 hours after the event has occurred. C. Incorrect: As seen in the Westinghous e E-1 background doc ument, the coupling of the RCS to the S/Gs is lost most probably before the 15 second point for a Large Break LOCA. Therefore, it is incorrect that natural circulation would be possible during a large break LOCA. D. Correct: in accordance with the WOG background document for 1-E-1, during a LBLOCA, decay heat is removed by a continuous supply of water from the ECCS Question Number: 4 Tier: 1 Group: 1

K/A: Knowledge of the operational implications of the following concepts as the apply to the Large Break LOCA:

Natural circulation and cooling, including reflux boiling.

Importance Rating: 3.0 3.3*

10 CFR Part 55: (CFR 41.7 / 45.7)

10CFR55.43.b: Not applicable

K/A Match: The K/A is matched because the applicant must understand the applicability of secondary plant heat removal (natural circulation and reflux boiling) to a Large Break LOCA. Technical

Reference:

1-E-1, Loss of Reactor or Secondary Coolant WBN-SDD-N3-63-4001, Safety Injection System WOG background document for 1-E-1 Proposed references to be provided: None Learning Objective: 3-OT-TAA013, LOSS OF COOLANT ACCIDENTS 1. Describe the dynamic behavior of the reactor, from a thermodynamic and hydraulic point of view, following a loss of coolant accident (LOCA) for the following categories: f. Large-break LOCA, >1 square foot and up to double ended guillotine shear. Cognitive Level: Higher Lower X Question Source: New Modified Bank Bank X Question History: Bank Question imported from San Onofre Nuclear Generating Station Comments: Breaks -l < diameter <-13-1/2 (1FT 2) For break sizes of one to two-inch in equivalent diameter, the RCS will rapidly depressurize early in the transient, and an automatic reactor trip and safety injection signal will be generated based on low pressurizer pressure. During the early stages of the depressurization, when the system is still full of two-phase liquid, the break flow, which also will be mostly liquid, is not capable of removing all the decay heat. Therefore, the early depressurization is limited by energy removal considerations, and the RCS pressure will temporarily hang up above the steam generator safety valve set pressure, assuming no steam dump is available. The RCS pressure stays at this level in order to provide a temperature difference from primary to secondary so that core heat may be removed by the steam generator. At this energy-balance controlled pressure, however, pumped safety injection flow is less than the break flow, and there is a net loss of mass in the RCS. Voiding throughout the primary side occurs and eventually the RCS begins to drain, starting from the top of the steam generator tubes. The rate of RCS drain is determined by the net loss of liquid inventory, a function of both SI flow and break size. Prior to the occurrence of draining, heat is removed from the steam generator through continuous two-phase natural circulation, with two-phase mixture flowing over the top of the steam generator tubes. As the draining continues, the natural circulation mode of heat removal as just defined ceases, and core heat is removed through condensation of steam in the steam generator. This method of heat removal is called reflux and is discussed in Reference

2. The condensation mode of heat removal is almost as efficient as continuous two-phase natural circulation in removing heat. However, condensation heat transfer coefficients may be lower than continuous two-phase natural circulation heat transfer coefficients.

Thus, as the steam generator tubes drain, a slight increase in primary system pressure occurs to give a greater from primary to secondary in order to remove all the decay heat. The steam generator secondary side pressurizes to the safety valve set pressure E-1 Background HElBG.doc 16 HP-Rev. 2, 4/30/2005 would correspond to an event which is approaching the design basis assumptions. It should be noted that more probable delivery rates for the safety injection system (e.g., maximum safeguards or minimum safeguards with no spilling line) will yield less core uncovery or no core uncovery. For break locations other than at the cold leg, little or no core uncovery is calculated. For breaks in the crossover leg there could be an uncovery similar to the uncovery experienced for the cold leg break when steam was vented through the crossover legs. Since for crossover leg breaks safety injection is injected into all RCS cold legs and steam does not have to pass through the broken loop RCP, there is no subsequent core uncovery. For breaks in the RCS hot leg or pressurizer vapor space, steam is vented earlier then for other locations (immediately for vapor space breaks) so that essentially no core uncovery is experienced. The method used for long-term plant recovery for breaks in this category depends upon the RCS pressure at the time that the operator determines if further RCS cooldown and depressurization is required (step 13 in the E-1 guideline) . If at this time the RCS pressure is greater than the low-head SI pumps shutoff head pressure, ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION, is used for long-term plant recovery. Even if the RCS pressure is less than the shutoff head pressure of the low-head SI pumps but it cannot be verified that the low-head SI pumps are injecting flow into the RCS, the operator should transfer to ES-1.2 for long-term plant recovery. The operator would stay in E-1 if the RCS pressure is less than the shutoff head pressure of the low-head SI pumps and flow into the RCS from the low-head SI pumps has been verified. Large Break LOCA, 1 FT'< Double-Ended, Minimum Safeguards A large break LOCA is the design basis for many aspects of the NSSS design. Some of the major design considerations impacted by the large break LOCA are peak core power, containment sizing, and loop/vessel internal forces. In order to describe the large break LOCA hydraulic transient a typical SAR safety analysis for a 4-loop plant will be utilized. The phenomena described here are similar for all Westinghouse plants. E-1 Background HElBG.doc 27 HP-Rev. 2, 4/30/2005 A large break LOCA (one square foot total area up to the double-ended break) has four characteristic stages: blowdown, refill, reflood, and long-term recirculation. Blowdown starts with the assumed initiation of the LOCA and ends when the reactor coolant system pressure (initially 2250 psig) falls to essentially that of the containment atmosphere. Refill starts at the end of blowdown and ends when the addition of emergency core cooling water fills the bottom of the reactor vessel and reaches the elevation of the bottom of the fuel rods. Reflood is defined as the time from the end of refill until the reactor vessel has been filled with water to the extent that core temperature rise has been terminated and core temperatures subsequently have been reduced to their long-term steady-state levels associated with dissipating decay heat. These time divisions are established mainly for analytical convenience. As contrasted with the large break, the blowdown phase of the small break occurs over a longer period and does not result in reduction of the effective water level in the reactor vessel below the bottom of the core. Thus, for the small break LOCA there are only three characteristic stages, i.e., a gradual blowdown in which the decrease in water level is checked before the bottom of the core is uncovered (and before pressure equilibrium between reactor and containment is reached), reflood, and long-term recirculation. Figure 11 illustrates the primary system pressure transient occurring during the blowdown phase of the LOCA for a double-ended cold leg guillotine break. The primary pressure rapidly drops from an initial value of 2250 psig to a low value of 40-50 psig by the end of blowdown (-1/2 minute). Also shown on Figure 11 is the break flow transient occurring during blowdown. The break flow starts at a very high value (critical flow, -70,000 lbm/sec) and is reduced to zero by the end of blowdown. Figure 11 also includes a plot of the SI accumulators mass flow rate. Note that accumulator flow is initiated approximately 16 seconds after the break occurs. This corresponds to the time when the RCS pressure has decreased to 600 psig, which corresponds to a minimum accumulator pressure set point. E-1 Background HElBG.doc 28 HP-Rev. 2, 4/30/2005 The containment pressure transient is shown on Figure 12. As shown, the containment pressure reaches a peak value early in the transient during the blowdown phase of the transient. A safety injection signal will be initiated on a containment High-1 pressure signal in a matter of seconds after the break and containment spray may be initiated on a containment High-3 pressure signal depending on the magnitude of the break and the specific plant's containment design. The important hydraulic transient parameters during the reflood phase are downcomer water level (ZD), core water level (ZC), and the core inlet flooding rate (VIN) as shown in Figure 12. During refill the ECCS cooling water from the SI accumulators and safety injection pumps enters the top of the reactor vessel downcomer annulus and starts to fill the reactor vessel lower plenum, which is filled after 45 seconds. This is commonly called bottom of core (BOC) recovery time. After BOC occurs, the downcomer annulus starts to fill rapidly and thus provides a static head for pushing cooling water into the core. A core inlet flooding rate (inches/second) is established and the water starts to move up into the core, thus providing the mechanism for core cooling during reflood. Table 1 presents a time sequence of events for the double-ended cold leg guillotine break. After successful initial operation of the ECCS, the reactor core is once again covered with borated water. This water has enough boron concentration to maintain the core in a shutdown condition. Decay heat is removed by a continuous supply of water from the ECCS. This supply initially comes from the refueling water storage tank (RWST). When the RWST level reaches the switchover setpoint the ECCS pumps are transferred into the recirculation mode (using ES-1.3, TRANSFER TO COLD LEG RECIRCULATION) wherein water is drawn from the containment sump and is cooled in the residual heat removal heat exchangers. Thus, long-term cooling of the core is maintained by the ECCS in sump recirculation mode. The core is maintained in a shutdown state by borated water. E-1 Background HElBG.doc 30 HP-Rev. 2, 4/30/2005 TABLE 1 TIME SEQUENCE OF EVENTS FOR LARGE BREAK. Start Reactor Trip Signal Safety Injection Signal Accumulator Injection End of Blowdown Pump Injection Bottom of Core Recovery Accumulator Empty 'Double-ended cold leg guillotine break (C 0=0.6) E-1 Background HElBG. doc 32 Time (Sec) 0.00 1.10 1.10 15.8 24.9 26.10 39.5 58.8 HP-Rev. 2, 4/30/2005

5. Given the following conditions: - Unit 1 is at 100% power. FCV-62-93, CHARGING HEADER FLOW is in AUTOMATIC. - Letdown flow is 75 gpm.

Subsequently: - A failure causes 1-HIC-62-93A, CHARGING FLOW PZR LEVEL CONTROL to output a FULLY CLOSED signal. Which ONE of the following describes the MINIMUM indicated CHARGING FLOW AND the response of Annunciator CHARGING FLOW HI/LO (108-A)? The MINIMUM which 1-FI-62-93A, CHARGING FL OW will read is ____(1)____ gpm. Annunciator 108-A, will alarm at ____(2)____ value as it would IF letdown flow HAD BEEN 45 gpm. A. (1) 0 (2) the SAME B. (1) 0 (2) a DIFFERENT C. (1) 33.5 (2) the SAME D. (1) 33.5 (2) a DIFFERENT

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Correct: Considering SSD-1-LPF-62-93A, one will observ e that the instrument air to 1-FCV-62-93 passes through a pneumatic relay FM-62-93E. This relay outputs an air signal to prov ide a minimum flow of 33.5 gpm. It does this by ensuring that a mini mum air pressure is always applied to the FAIL OPEN 1-FCV-62-93. Additionally, 1-HIC-62-93A possesses software restraints which preclude it from outputting a signal less than 25.6%. This provides a minimum charging flow of 55 gpm when the controller is in automatic. It is plausible to believe that if a full closed signal were applied to 1-FCV-62-93 that it would drive full closed. This is the normal occurrence with other controllers on the main control boards. Also, in accordance with the ARI for annunciator

108-A, one may witness that three set points actuate such alarm. A high charging flow with a setpoint of 150 gpm exists and a low

charging flow (driven by the software bistable in DCS FS-62-93B). The low charging flow receives input from the position of the letdown orifices valves. If either of the 75 gpm orifices is open, then the DCS

software will utilize a 55 gpm setpoint. If both of the 75 gpm orifices are closed, then the DCS softwar e will use a 47 gpm setpoint. Therefore, it is true that the 108-A will alarm at different values for letdown flows of 75 and 45 gpm respectively. It is plausible to believe that the LOW charging alarm functions in a similar fashion to the high charging alarm; namely, that it possesses a single setpoint. B. Incorrect: While it is true that tw o setpoints exist which drive the 108-A annunciator, it is not true that charging flow would indicate 0 gpm. C. Incorrect: While it is correct that char ging flow would indicate a minimum of 33.5 gpm. It is not true that only one setpoint exists for annunciator window

108-A. Note that for this question the value of 33.5 gpm was chosen. It is

possible for the software of DCS to fail in such a manner that the restricting charging flow to greater than 55 gpm while in automatic is not operable. D. Incorrect: As mentioned two setpoints exists for the window of concern and charging flow would indicate 33.5 gpm

Question Number: 5 Tier: 1 Group: 1

K/A: 022 Loss of Reactor Coolant Makeup AA1. Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup: AA1.02 CVCS charging low flow alarm, sensor, and indicator

Importance Rating: 3.0 2.9

10 CFR Part 55: (CFR 41.7 / 45.5 / 45.6)

10CFR55.43.b: Not applicable

K/A Match: K/A is matched because the applicant is required to monitor both the CVCS charging low flow alarm and the indicator (charging flow) during a loss of reactor coolant makeup. The failure of the 1-FCV-62-93 valve as described in the stem does precipitate a loss of makeup as charging flow is insufficient to maintain letdown. Technical

Reference:

SSD-1-LPF-62-93A 1-ARI-102-108, HVAC & CVCS Proposed references to

be provided: None Learning Objective: 3-OT-STG-062A, CHEMICAL AND VOLUME CONTROL SYSTEM 3. Given plant conditions, DETERMINE if any of the following CVCS alarms would be present and actions

required by the ARI :

c. [1-XA-55-5C 108A] CHARGING FLOW HI/LO Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: W BNSSD-1-LPF-62-93A Reviewed b y: M ike Johnson App roved b y:R. S. HendersonDate:02/28/2011 Charging Header Flow Loop Name: Applicable Tech Specs: Loop Components Location 1-FC-62-93A 1-FCV-62-93 1-FI-62-93A 1-FI-62-93B

1-FM-62-93A 1-FM-62-93B

1-FM-62-93D 1-FS-62-93A/B 1-FS-62-93B/A 1-FT-62-93A

1-HIC-62-93A

1-HIC-62-93B

1-HIC-62-93C

1-HIC-62-93D 1-PX-62-93A Computer Pt F0128A 1-PI-62-93 1-FM-62-93E 1-R-18 692, A5-U 1-M-5 1-L-112 1-R-18 1-L-112 1-R-18 1-R-18 1-R-18 1-L-61 1-M-5 1-L-112 1-L-10 1-R-18 1-R-18 MCR 1-L-112 1-L-112 A ssociated Drawin gs:45W600-62

45W1635-52 & 103
Foxboro Dw g 1-47043 , CD 18;W estinghouse Dwg 108D408 SH 27; 47W600-118 SCALING AND SETPOINT DOCUMENT 09 1 25 PAGE OF REVISION LOOP COVER SHEET (1) In automatic operation, the output of 1-FC-62-93A is effectively limited by the action of 1-FM-62-93D and 1-HIC-62-93D. This provides a minimum charging flow of 55 GPM.

Pneumatic relay 1-FM-62-93E output provides a minimum flow of

33.5 GPM for Appendix R. (2) DCN M-33222-A (F-38016-A) provided pneumatic relay

(1-FM-62-93E) and setpoint, bypass valve and pressure

indicator (1-PI-62-93).(3) DCN W-36890-A removed the mechanical travel stop previously added by DCN M-17644-A. (4) DCN W-39199-A & F-39315-A disconnected output of 1-FS-62-93B/A

and added a programmed variable low alarm from ERFDS. The low alarm will have a setpoint of 55 GPM if a 75 GPM

orifice is in use OR will have a setpoint of 47 GPM if the 45 GPM orifice (PD Pump) is being used.(5) DCN 51036 replaced obsolete Foxboro E13DH transmitter with Rosemount transmitter. (6) DCN 50918 and PIC 51462 revised NE SSD Allowable Value. (7) DCN 51812 changed 1-PREG-62-93 to 48 psig -0/+2 psig. (8) DCN 51934 replaced actuator body and trim for 1-FCV-62-93 and revised 1-PREG-62-93 and 1-FM-62-93E setpoints. Remarks: W BNSSD-1-LPF-62-93A Reviewed b y: M ike Johnson App roved b y:R. S. HendersonDate:02/28/20111-R-181-M-51FK-121 TP DEV S Q RTF1056LOGF0128ASCALING AND SETPOINT DOCUMENTLOOP DRAWING LOOP NAME:CHARGING HEADER FLOW1-FI-62-93A1-M-51-FM-62-93A1-R-181-FI-62-93B1-L-1121FI-121B1-PX-62-93A1-R-18 1F Q-1211FT-1211-FT-62-93A1-R-181-FC-62-93A1-FS-62-93B/A1-R-18 1FB-121A/B1-FS-62-93A/B TP1-R-181FP/121A TP1-R-181FP/121B1-HIC-62-93A1FY-121A1FI-121A1-R-181FP/121EFINALOUTDEMAND FROM:1-LPL-68-339TO:1-XS-62-93 (SHEET 3)HISPAREXA-55-5C-108A1FY-121B1-FM-62-93D1FC-121 TP1-R-181FP/121C1-HIC-62-93D1-R-181HY-121OUT TP1-R-181FP/121D1-L-61 SPERFDSPROGRAMMED VARIABLELO ALARMXA-55-5C-108A PAGE OF REVISION225 09 W BNSSD-1-LPF-62-93A Reviewed b y: M ike Johnson App roved b y:R. S. HendersonDate:02/28/20111-M-5SCALING AND SETPOINT DOCUMENTLOOP DRAWINGLOOP NAME:CHARGING HEADER FLOW1-L-112I/P1-HIC-62-93B1-XI-62-931-XS-62-931-FM-62-93B1-HIC-62-93C1-L-101-L-11B1-L-112FROM:1-HIC-62-93A (SHEET 2)1-FCV-62-93 O1-L-1121-FM-62-93E1-L-1121-PI-62-93 PAGE OF REVISION325 09 WBN Unit 1 HVAC & CVCS 1-ARI-102-108 Rev. 0000 Page 40 of 47 A. System pipe break B. Charging pump tripped C. Malfunction of Pressurizer Level Control System [1] IF ALL the following conditions exist:

  • Any RCP Thermal Barrier Out-of-Service
  • In-Service Charging pump trips
  • RCP seal injection flow required THEN IMMEDIATELY START available charging pump to restore seal flow.

[2] CHECK 1-FI-62-93A, CHARGING FLOW [1-M-5] to determine if flow is high or low. [3] CHECK 1-LI-68-320, -335A, and -339A, PZR LEVEL [1-M-4]. [4] IF PZR Level Control System malfunctions, THEN GO TO 1-AOI-20. [5] IF charging flow is low, THEN CHECK letdown temperature AND EVALUATE increasing charging flow OR ISOLATE letdown. [6] IF charging is lost, THEN IMMEDIATELY ISOLATE letdown AND REFER to 1-AOI-20. [7] DETERMINE cause of problem AND INITIATE corrective action, as necessary. [8] REFER TO 1-SOI-62.01 for CVCS system operation. 1-47W610-62-2

1-47W809-1 1-AOI-20 1-SOI-62.01

08F734235-FD-1605 Source Setpoint HI: 1-FS-62-93A (DCS) 150 gpm LO: 1-FS-62-93B (DCS) 47 gpm with BOTH 75 gpm orifice valves CLOSED OR 55 gpm with EITHER 75 gpm orifice valve OPEN 108-A CHARGING FLOW HI/LO (Page 1 of 1) Probable Cause: Corrective

Action:

References:

6. Given the following conditions: - Tavg is 150°F following a refueling outage. - Neither 1-PCV-68-334 nor 1-PCV-68-340 will OPEN. - Unit 1 RCS pressure is 250 psig and RISINGuncontrollably. In accordance with N3-74-4001, Residual Heat Removal System description, which ONE of the following complete s the statement below?

System design will limit RCS pressure to _____(1)_____ psig. A. 370 B. 405 C. 450 D. 600 CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: The stem of the question places the Unit at 150°F (MODE 5) and thus implicitly requires that RHR be in serv ice. The stem then states that pressure is 250 psig and rising uncont rollably. This indicates that a Loss of RHR is occurring (as seen in section 3.4 of 1-AOI-14, Loss of RHR Shutdown Cooling which has an entry diagnostic of RCS High

Pressure during RHR Shutdown Cooling).

It is plausible to believe the 370 psig is correct because this value is the setpoint for a contact of a SOR pressure switch which provides an interlock for the opening of the RHR suction valves (and thus provides protection from overpressure conditions). It is not correct because this interlock does nothing but prevent the RHR suction valves from

opening. If as in this case, the valves are already open and the system is in service, then the system will not be further protected from an overpressure condition (i.e. th e main suction valves will not automatically close). B. Incorrect: The value of 405 psig is the setpoint for the pre ssurizer PORV 1-PCV-68-334 for an RCS temperat ure of 150°F. If the COMS system had performed its function successfully , then this would be the highest pressure which the RCS experienced. However, as the stem of the

question related, neither PORV operated and as such this value is not the limiting RCS pressure observed. C. Correct: As seen on print 1-47W810-1, the 1-RFV-74-505 has a lift setpoint of 450 psig. As seen in system description N3-74-4001, "RESIDUAL HEAT REMOVAL SYSTEM," this valve protects the RHR system from RCS over pressurization while the RHR system is in operation. D. Incorrect: As seen on print 1-47W811-1, the discharge reliefs for the RHR pumps have a lift setpoint of 600 psig. If 1-RFV-74-505 failed to operate, RCS pressure would be controlled by these relief valves at their lift setpoint.

Question Number: 6 Tier: 1 Group: 1

K/A: 025 Loss of Residual Heat Removal System (RHRS) AA2. Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: AA2.06 Existence of proper RHR overpressure protection Importance Rating: 3.2* 3.4* 10 CFR Part 55: (CFR: 43.5 / 45.13) 10CFR55.43.b: Not applicable K/A Match: The K/A is matched because the applicant must ascertain the correct pressure that the RCS will stabilize at (an thus verify the existence of RHR overpressure protection) given that section 3.4 of 1-AOI-14, Loss of RHR Shutdown Cooling is in effect. Technical

Reference:

1-AOI-14, Loss of RHR Shutdown Cooling 1-47W810-1

1-47W811-1 1-ARI-109-115

N3-74-4001 Proposed references to

be provided: None Learning Objective: 3-OT-STG-074, OPERATIONS THE RESIDUAL HEAT REMOVAL SYSTEM

1. DESCRIBE the design criteria, purpose and/or functions of the Residual Heat Removal System (RHR) and the major system components listed below: (IER 11-3: Having a solid understanding of plant design, engineering principles, and sciences):
g. RHR suction piping relief valve, 74-505Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Loss of RHR Shutdown Cooling 1-AOI-14 Rev. 0002 Page 4 of 87 3.0 OPERATOR ACTIONS 3.1 Diagnostics IF GO TO Subsection RHR Pump Trip During Midloop Operation. 3.2 RHR Pump Cavitation During Midloop Operation. 3.3 RCS High Pressure during RHR Shutdown Cooling. 3.4 RHR Pump 1A-A trip. 3.5 RHR Pump 1B-B trip. 3.6 RHR System Outleakage or LOCA While On Shutdown Cooling. 3.7 Loss of CCS to RHR System. 3.8 RCS Alternate Cooling Method With the Rx Vessel

Head Installed 3.9 RCS Alternate Cooling Method With the Rx Vessel

Head Off 3.10 In the event of an extended loss of AC power while in Mode 5 or 6, the use of portable FLEX equipment should be considered.

WBN Unit 1 CVCS & RHR - RPS & ESF 1-ARI-109-115 Rev. 0000 Page 34 of 49 NOTE 1-FCV-74-1, -2, -8 and -9 will NOT auto close on HI PRESS. If NOT open, these valves are prevented from being opened. A. 1-FCV-74-1, -2, -8 and -9 NOT fully closed with RCS pressure 375 psig or more B. Pressure Switch deenergization CAUTION RHR Pump suction relief valve lifts at 450 psig. [1] IF RCS water solid, THEN STOP Charging pump(s). [2] IF RHR operating and RCS pressure reaches 450 psig, THEN GO TO 1-AOI-14, LOSS OF RHR SHUTDOWN COOLING. [3] REDUCE pressure below 345 psig:

  • CONTROL Charging & Letdown
  • CONTROL PZR Heaters
  • CONTROL RHR Letdown NOTE Step [4] is N/A if RHR suction relief valve satisfies TS 3.4.12, Cold Overpressure Mitigation System.

[4] IF RHR is NOT operating, THEN ENSURE the following valves CLOSED:

  • 1-FCV-74-1, and -2, LOOP 4 HL TO RHR SUCTION
  • 1-FCV-74-8 and -9, RHR SYSTEM ISOLATION BYPASS

[5] VERIFY STATUS of 1-PS-68-63B and 1-PS-68-64B. 1-45W600-57-17 1-45W760-74-2, -4 1-AOI-14 Source Setpoint 1-PS-68-63B (PB406BX) See Probable Cause 1-PS-68-64B (PB407BX) 113-ERHR SUCT FCV-74-1, 2, 8, 9 OPEN & HI PRESS (Page 1 of 1) Probable Cause: Corrective

Action:

References:

WBN Unit 1 Loss of RHR Shutdown Cooling 1-AOI-14 Rev. 0002 Page 86 of 87 Appendix D (Page 1 of 1) Cold Overpressure Limit Curve NOTE: The values at right are for 4 RCPs in operation with RCS temp. 105°F. The most conservative values (lowest pressure, first to open) for PCV-68-340

were used to generate the

curve above. RCS press

should not approach RCS

press setpoints marked

with an asterisk due to #1 seal P concerns. Temp °F PCV 334 Setpoint, psig PCV 340 Setpoint, psig

  • Setpoint violates pump seal limit. 70 383* 373* 100 386 376* 150 405 395 200 430 420 250 510 490 275 560 540 300 590 570 350 590 570 450 2335 2335

LOOP 1 R T PUMP RCP 6" .. 1 LOOP 4 SG RCP FR 68-428A CLASS B FR STEAM GENERATOR XIS 68-70 PT PT SEE NOTE 17 a. 1 8 Of SEE 74-1 ..... NOTE 17 a. 1 8 FR 74-500 74-541 (/) (/) < ....J (J m (/) (/) < ....J (J m (/) (/) < ....J (J I I I 503 : 542 I I I I I I -CLASS A 74-2 SEE ....... NOTE 17 a. 18 SEE NOTE 17 a. 18 CLASS B ..... I CO ...-I .-L&J co ll:C "'a:::: 10 ..... u (.!) (/) (/) < ....J (J SEE NOTE 21 FCV ""'""-..,...__ COLD LEG 8" INJECTION 1-47W811-1 COORD G-8 74-543 ....--74-504 X-107 4 II TO LETDO' HT EXCH 1-47W809* COORD B-CLASS H 3/4" .. ...... , VENT 74-538 TO CHARGING PUMPS 1-47W811-1 COORD G-8 RHR SPRAY HOR 1-47W812-1 COORD F-4 Tl 74* 74-505 TO PRESSURIZER RELIEF TANK; 1-47W813-1 COORD H-2 (/) (/) (/) (/) << ....J ....J (J (J PRESSURE: 450 PSIG This is the relief valve which will provide protection. It's setpoint is 450 psig. NPG System Description Document RESIDUAL HEAT REMOVAL SYSTEM N3-74-4001 Rev. 0014 Page 41 of 84 3.2.3 RHR Valves (continued)

5. RHR HXs Bypass FCV-74-32 (W 618): This non-trained, air-operated, fail open, 8" butterfly valve is located in the HXs' common bypass line. A steel plate is installed as a protection barrier for the positioner linkage of FCV-74-32 but does not affect testing. This valve has controllers located in the MCR and ACR. By adjusting this valve in conjunction with FCV-74-16 and -28, total system flow can be held constant while changing flow through the HX(s). The line containing this valve is isolated from the RHRPs and HXs by HCV-74-36 and -37. These manual valves are opened prior to initiating RHR normal operations. 6. RHRP Mini-Flow FCV-74-12, -24 (W 610, 611): The RHRP mini-flow MOVs are safety-related, 3" globe valves located in

each pump's mini-flow loop. They are controlled by a flow indicating switch and the RHRP breaker. 7. RHR Crosstie FCV-74-33, -35 (W 8716A, 8616B): These MOVs are safety-related, 8" gate valves located in the RHR piping crosstie downstream of the HXs. These valves must be open during normal operation of the RHRS. ECCS function is addressed in Ref. 7.2.17. During Modes 5 and 6, these valves may be closed. Because of the low flow required during Modes 5 and 6, the pump-to-pump interaction could result in back pressure from one pump being strong enough to limit the weaker pump's injection flow or forward flow. Only a small break LOCA is required to be postulated during Mode 4. Per system description N3-63-4001 (Ref. 7.2.17), Section 3.1.3, the RHR pumps are not required for the injection phase of a Mode 4 LOCA. ECCS operation following a Mode 4 LOCA is bounding for a Mode 5 LOCA. Therefore, isolation of the crosstie valves does not adversely impact ECCS operation for a Mode 5 LOCA. DBA LOCA events are not required to be postulated in Mode 6 per WB-DC-40-64 (Ref. 7.2.32) and automatic SI is not required in Modes 5 and 6 per the Technical Specification bases B3.3.2. To prevent pressure locking of these valves and/or liquid entrapment a 1/4" dia. hole is located on the upstream side of the valve disc per Ref. 7.5.63. B. Relief Valves and Manual Valves 1. RHR Suction Header Relief RFV-74-505 (W 8708): This relief valve protects the RHRS from RCS over pressurization while the system is in operation (i.e., FCV-74-1 and -2 open). The valve's nozzle has a 3" inlet diameter, 4" outlet diameter and is austenitic stainless steel. NPG System Description Document RESIDUAL HEAT REMOVAL SYSTEM N3-74-4001 Rev. 0014 Page 42 of 84 3.2.3 RHR Valves (continued) The valve has a relief capacity of 900 gpm and a required flow rate of 480 gpm at 350 degrees F and 690 gpm at 200 degrees F. The setpoint for the valve is 450 psig (Ref. 7.5.45, 7.5.48, 7.5.49, 7.5.50). The valve discharges to the Pressurizer Relief Tank (PRT) and its discharge flow capacity is analyzed in Ref. 7.4.7. The valve cannot protect the RHRS from overpressure during modes other than startup or cooldown if FCV-74-1 and -2 are inadvertently open when RCS pressure is above RHRS design. Interlocks, alarms and administrative controls are used on FCV-74-1 and -2 to prevent the event from occurring (see Section 3.3

and 4.0). 2. Valves HCV-74-34, -36, -37 (W 8735, 8726A, 8726B): These valves are 8" manual gate valves. HCV-74-34: This normally closed RHR-to-RWST Return Line Isolation Valve is throttled open when pumping primary coolant to the RWST following refueling or when reducing or lowering RCS inventory or level. At all other times, this valve is locked closed (see Section 4.0). HCV-74-36, -37: The normally closed HX Bypass Isolation Valves are opened when bypass flow through FCV-74-32 is required for temperature control during startup, cooldown, and refueling. 3. RHRP Check Valves CKV-74-514, -515 (W 8730A, 8730B): These valves are 8" swing-type check va lves located downstream of RHRPs. The valves will protect against reverse flow through an RHRP when it is not operating to ensure the operating pump meets system flow requirements. 4. RHRP and HX Isolation ISV-74-520, -521, -524, -525 (W 8728A, 8728B, 8724A, 8724B): These valves are 8" gate valves loca ted between the RHRPs and the respective HXs A and B. These valves provide RHRP and RHR HX isolation for maintenance. 5. Letdown Isolation to CVCS ISV-74-530, -531 (W 8734A, 8734B): These valves are 2" Kerotest globe valves located in RHR cross-tie piping. They are opened when RHRS letdown to the CVCS (Ref. 7.2.35) is necessary (see Section 4.0). 6. Sample Root Valves SMV-74-522, -523, -532, -533 (W 8725A, 8725B, NA, NA): These valves are 3/4" Kerotest diaphragm globe valves. SMV-74-532, -533 are on mini-flow return lines A-A and B-B. SMV-74-522, -523 are in the RHR flowpath downstream of the RHRPs. These normally open valves provide for RHR sampling when required for water chemistry control prior to and during operations. Frequency and duration of sampling is controlled by the Sampling and Water Quality System (System 43).

\_SET PRESS. 600 PSI TO RHR SPRAY CONT ON 1-47W812-1. E-4 C CTOR -173A I* C 1-3 3/4* DR il& 63-629 63-693

  • 3/4* TV 6J-309A & 3 2 SET PR .... CONT ON 471813-1 I H-5 AT PEN. X-24 63-31 OA ...............

=-= -;;;;=;;t--1 N C§ .... .... '? .. 6J-3l9A a* FM PEN XJO RELIEF CONT ON 1-471811-1 0-6 3 1--...............

2 A RESTR [CTOR OR-63-92 .. SET-12* PRESS. 3" 600 PSI

  • 63-626 63-370A 14* K-15
  • The RHR pump's discharge relief has a setpoint of 600 psig. 1 TO RHR SPRAY CONT ON 1-471812-1.

F-4 3/+* a*

  • 2* .. &13----I ..... VTV-63-16 -VTV-63-17 a* .. , ...... 63-506 1* DR RESIDUAL HEAT REMOVAL PUMPS * (" ... N 0 It) I ,,., co 63-503 3/4 11 TC J :. ... 18-B RESIDUAL HEAT EXCHANGERS

..,... a* * .. 63-505 3/4* TC a* .. LETDOWN HX (RHR) 1-471809-1. B-8 a* .. a*

  • 3 * -.. RHR CONT ON 1-471810-1 8 PLACES This is the 1A-A RHR pump. The highlighted path would be for ECCS injection.

J I CONT ON 47W810-1. E-9 : I I

7. Given the following conditions: - Unit 1 is at 100% power. FCV-62-84, AUX SPRAY TO PZR has stuck OPEN. AOI-18, Malfunction of Pressurizer Pr essure Control System is in progress.

- The crew is placing excess letdown in service in accordance with 1-AOI-18 Attachment 2, Excess Letdown. Which ONE of the following describes the basis for the excess letdown flowrate limit?

The excess letdown flowrate is limited to prevent ________.

A. high letdown and CCS temperatures B. RISING dose rates in the auxiliary building C. RCP #1 seal leakoff flow RISING above the limit D. damaging the liquid radioactive waste pumps if excess letdown is diverted

CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: 1-AOI-18 contains the following caution in Attachment 2: Excess letdown design flow is 20 gpm but can go as high as 50 gpm which could cause high letdown and CCS temperatures. B. Incorrect: Attachment 2 of 1-AOI-18 contains step 1.0 D: NOT IFY Radiological Protection Shift Supervisor Excess Let down is in service. The reason for this step is that the Excess Le tdown heat exchanger is located on elevation 737' of the Aux Building. Unlike the regenerative and non-regenerative heat exchangers (which are located within a shielded

room inside of containment), the excess letdown heat exchanger exists in a location which can a ffect the dose to nuclear plant personnel. It is plausible to belie ve that an applicant would believe that given the location of the ex cess letdown heat exchanger, which the basis for the flow limit w ould be to limit personnel dose. C. Incorrect: Attachment 2 of 1-AOI-18 cont ains a caution which states: As excess letdown is placed in service, the #1 seal leakoff flow could potentially

fall below the limit. Theref ore, it is not correct t hat seal flow would rise above the limit. It is plausible to believe that it would if one believed that a siphon or eductor effect were created as the flow from excess letdown merged with the seal return flow. D. Incorrect: It is plausible to believe that the maximum flow which excess letdown flow is capable of could out r un the pumping capacity of the RCDT pumps because there are radioactive liquid waste pumps which are of insufficient capacity to keep up with excess letdown's maximum flow. Two of these pumps in particula r are one of the Tritiated Drain Collector Tank and one of the Floor Drain Collector pumps. Both of

these pumps are 20 gpm and serve tan ks much larger than the RCDT. Plausibility of this distractor is enhanced because the applicant may believe that the excess letdown flow is diverted directly to the TDCT. However, this is not the case. The TDCT does have a second pump which discharges at 100 gpm.

Question Number: 7 Tier: 1 Group: 1

K/A: 027 Pressurizer Pressure Control System (PZR PCS) Malfunction G 2.4.18 Knowledge of the specific bases for EOPs.

Importance Rating: 3.3 4.0

10 CFR Part 55: (CFR: 41.10 / 43.1 / 45.13)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to understand the specific basis for the limit on excess letdown flow when placing such in service in accordance with 1-AOI-18, Malfunction of Pressurizer

Pressure Control System. Technical

Reference:

1-AOI-18, Malfunction of Pressurizer Pressure Control System WBN-SDD-N3-77C-4001, Liquid Radwaste Processing

System Proposed references to be provided: None Learning Objective: 3-OT-STG-062A, CHEMICAL AND VOLUME CONTROL SYSTEM 6. EXPLAIN the CVCS design features and/or interlocks that provide the following: r. Excess letdown Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 MALFUNCTION OF PRESSURIZER PRESSURE CONTROL SYSTEM 1-AOI-18 Rev. 0002 Page 25 of 27 Attachment 2 (Page 1 of 2) Excess Letdown 1.0 PLACING EXCESS LETDOWN INSERVICE CAUTION As excess letdown is placed in service, the #1 seal leakoff flow could potentially fall below the limit. A. PERFORM the following: [0-M-27B]

  • OPEN 1-FCV-70-143, EXC LTDN HX SUP CIV-A.
  • OPEN 1-FCV-70-85, EXC LTDN HX OUT CIV-A. B. CHECK 1-HS-62-59A, CVCS EXCESS LETDOWN DIVERT FLOW CNTL aligned to one of the following
  • NORM
  • DIV NOTE 1-FCV-62-61 and 1-FCV-62-63 should NOT be opened if Phase A signal present.

C. PERFORM the following: [1-M-5]

1. IF 1-HS-62-59A, CVCS EXCESS LE TDOWN DIVERT FLOW CNTL in NORM, THEN
  • ENSURE 1-FCV-62-61, CVCS SEAL WATER RETURN HEADER ISOLATION, OPEN.
  • ENSURE 1-FCV-62-63, CVCS SEAL WATER RETURN HEADER ISOLATION, OPEN.
2. OPEN 1-FCV 54, EXCESS LTDN ISOL.
3. OPEN 1-FCV-62-55, EXCESS LETDOWN ISOLATION.

WBN Unit 1 MALFUNCTION OF PRESSURIZER PRESSURE CONTROL SYSTEM 1-AOI-18 Rev. 0002 Attachment 2 (Page 2 of 2) 1.0 PLACING EXCESS LETDOWN INSERVICE (continued) Page 26 of 27 CAUTION Excess letdown design flow is 20 gpm, but can go as high as 50 gpm which could cause high letdown and CCS temperatures.

4. SLOWLY OPEN 1-HIC-62-56A, EXCESS LETDN FLOW CONTROL to maintain Excess LD HX Ou tlet temperature below 195

°FAND Excess LD HX CCS Outlet Temperature below 137 °F. 5. MONITOR PZR level returning to normal. D. NOTIFY Radiological Protection Shift Supervisor Excess Letdown is in service.

8. Given the following conditions: - Unit 1 was at 100% power. - An ATWS condition has occurred. FR-S.1, Response to Nuclear Power Generation/ATWS, is in progress. - Step #2, VERIFY turbine TRIPPED is in progress.

Subsequently: - A PZR PORV OPENS. - ALL Control Rods insert to the bottom of the core. - An AUTOMATIC Safety Injection occurs. Which ONE of the following describes how the crew will implement the EOPs? The crew will ________.

A. IMMEDIATELY exit 1-FR-S.1 AND Perform 1-E-0, Reactor Trip or Safety Injection B. REMAIN in 1-FR-S.1 until completed OR directed to transition to another procedure C. EXIT 1-FR-S.1 when the RED path clears AND perform the steps of 1-E-1, Loss of Reactor or Secondary Coolant D. SHUT the block valve for the stuck OPEN PORV THEN transition to 1-E-0 at 1-FR-S.1 Step 6, MONITOR for SI Signal

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: Plausible since 1-E-0 does have the immediate actions of a Rx trip or Safety Injection, however once an FR is entered it must be completed prior to exit or an exit is directed by a procedure step. B. Correct: Per the rules of usage in TI -12.04, User's Guide for Abnormal and Emergency Operating Instructions, once 1-FR-S.1 has been entered, the actions of the FR are to be comple ted prior to exit or until directed to exit by a procedure step of FR-S.1. C. Incorrect: Plausible if candidate thought that the FR could be exited as soon as the condition which caused them to enter the FR is cleared. This is not in accordance with the TI-12.04. D. Incorrect: Plausible since at step 6 of 1-FR-S.1, the operat ors are directed to Perform the immediate action steps of 1-E-0 if SI has actuated. If candidate thought this meant for the crew to exit 1-FR-S.1 instead of performing the actions concurrently.

Question Number: 8 Tier: 1 Group: 1

K/A: 029 Anticipated Transient Without Scram (ATWS) EG 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

Importance Rating: 4.2 4.1

10 CFR Part 55: (CFR: 41.10 / 45.3) 10CFR55.43.b: Not applicable K/A Match: This question matches the K/A by testing the candidate's knowledge of the response procedures and their use for an ATWS condition. Technical

Reference:

TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions Proposed references to

be provided: None Learning Objective: 3-OT-FRS0001, Function Restoration Guidelines 1-FR-S.1 & S.2

2. Apply the rules of usage (TI-12.04) and ANALYZE plant conditions to identify any required procedure

transitions in 1-FR-S.1 and FR-S.2. Cognitive Level: Higher Lower X Question Source: New Modified Bank Bank X Question History: Bank question 029EG2.4.31 7 Comments: WBN Unit 1 & 2 User's Guide for Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 28 of 57 2.4.4 Status Tree Rules of Usage (continued) C. Status Trees shall be monito red in the following priority: 1. 1-FR-S, Subcriticality, 2. 1-FR-C, Core Cooling,

3. 1-FR-H, Heat Sink,
4. 1-FR-P, PTS,
5. 1-FR-Z, Containment,
6. 1-FR-I, Inventory. D. If a RED path is diagnosed, then the Function Restoration Instruction will be implemented IMMEDIATELY. E. If an ORANGE path is diagnosed, then the remaining Status Trees will be checked. If no RED path exits, then t he highest priority ORANGE path Function Restoration Instruction will be implemented. F. Once implemented because of any RE D or ORANGE path, that Function Restoration Instruction will be performed to co mpletion or to a point of transition UNLESS a higher priority condition develops.
1. As a Function Restoration Instruction is performed, the status of that tree may change. This change does NOT change the priority of an instruction in progress. 2. If a higher priority condition develops, the instruction in effect should be suspended and the higher priority condition addressed. G. When no RED or ORANGE path exists , a YELLOW path Function Restoration Instruction can be implemented at the operator's discretion.
9. Given the following timeline:

00:00:00 1-ES-3.3, Post-SGTR Using Steam Dump is in progress on Unit 1. 00: 0 1:00 The crew is depressurizing the rupt ured SG to 400 psig using steam dumps. 00: 1 0:00 Step 15. d. is currently being performed to STOP depressurization of the Ruptured SG. 00: 1 1 :00 1-FCV-1-103, CONDENSER A MAIN STEAM DUMP VLV, FAILS to CLOSE from the MCR. 00: 1 2:00 An AUO is dispatched to MANUALLY isolate the failed steam dump. Which ONE of the following describes the MAXIMUM number of Steam Dump Valves that CAN be OPENED AND the method the AUO will use to MANUALLY isolate the Steam Dumps? At 00:02:00 a MAXIMUM of ____(1)____ steam dump valves can be OPENED. The dispatched AUO will be able to MANUALLY shut a SINGLE valve to isolate ____(2)____. A. (1) 3 (2) 1-FCV-1-103 ONLY B. (1) 12 (2) 1-FCV-1-103 ONLY C. (1) 3 (2) ALL steam dump valves D. (1) 12 (2) ALL steam dump valves

CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: The conditions in the stem describe that the operating crew has cooledthe RCS down to less than 375°F (which would be step 10 of 1-ES- 3.3). WBNP has permissive P-12. P-12 is the LO-LO Tavg Steam Dump block. If RCS Tavg in 2 of 4 loops is less than or equal to 550°F then ALL of the steam dumps would close. This would have been the case when the crew was in 1-E-3, Steam Generator Tube Rupture. The crew would have then bypassed P-12 (still in 1-E-3) to perform their initial cooldown. The ARI for window 68-A (P-12) describes that This interlock can be bypassed for 3 of 12 steam dump valves. Therefore, when the crew was perfo rming the cooldown to less than 375°F, they would have had at most 3 steam dump valves. Additionally, as seen on print 1-47W 801-1, 1-FCV-1-103 is isolated via a downstream manual valve 1-ISV-1-627. Also seen on this print is the fact that ONLY 1-FCV-1-103 is isolated by this single manual isolation. Each steam dump is pr ovided with a single downstream isolation valve. B. Incorrect: It is not correct that 12 st eam dump valves could be open given the conditions listed in the steam. It is plausible that the applicant could believe that 12 dump valves could be opened if he failed to

comprehend the operation of P-12 or did not correctly interpret the plant conditions which would be had in 1-ES-3.3. Also, it is correct that only 1-FCV-1-103 would be isolated. C. Incorrect: While it is true that a ma ximum of 3 steam dump valves could be opened, it is not true that a single va lve could isolate all steam dumps. It would be plausible that a single valve would exist as there are other components which are supplied by main steam which can be isolated

by a single valve. Examples of th ese would be the main feed pumps' HP steam supplies and the TDAFWP. Also, the main steam piping to the steam dumps follows a path which provides a convenient location for a single isolation valve. D. Incorrect: As mentioned previously, it Is not true that 12 steam dumps could be opened or could be isolated by a single valve.

Question Number: 9 Tier: 1 Group: 1

K/A: 038 Steam Generator Tube Rupture (SGTR) EA1 Ability to operate and monitor the following as they apply to a SGTR: EA1.43 Manual isolation of steam dump valves Importance Rating: 3.6* 3.5*

10 CFR Part 55: (CFR 41.7 / 45.5 / 45.6) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to both monitor the steam dump valves in their initial state (prior to isolation) as they would exist during 1-ES-3.3. The applicant must subsequently demonstrate the knowledge required to perform a manual isolation of a failed steam dump valve. The same knowledge applies to the manual isolation of multiple steam dump valves. Technical

Reference:

1-47W801-1; 1-ARI-64-70, Bypass, Intlk, & Permissive 1-ES-3.3, Post-SGTR Cooldown Using Steam Dump

Proposed references to be provided: None Learning Objective: 3-OT-SYS001B, Steam Dump Control System 4. EXPLAIN the physical connect ions and/or caus e-effect relationships between the Steam Dump Control System and the following systems:

a. Main Steam System
6. EXPLAIN the Steam Dump Control System design features and/or interlocks that provide the following: (IER 11-3 Have a solid understanding of plant design, engineering principles, and sciences.)
c. RCS Cooldown i. Protection against excessive cooldown Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Bypass, Intlk, & Pe rmissive 1-ARI-64-70 Rev. 0000 Page 28 of 47 A. Plant cooldown NOTE The P-12 interlock prevents opening of the steam dump valves if RCS temperature is less than 550 °F. This interlock can be bypassed for 3 of 12 steam dump valves. P-12 light also indicates steam dump valves should be closed. [1] VERIFY Tavg less than or equal to 550 °F. [2] IF bypass of LO-LO TAVG interlock is desired, THEN REFER TO 1-SOI-1.02, STEAM DUMP SYSTEM. W 7246D11 Sht. 40 1-47W611-63-1 1-SOI-1.02 Source Setpoint 1-TS-68-2J (Loop 1) Tavg in 2/4 loops less than or equal to 550 °F 1-TS-68-25J (Loop 2) 1-TS-68-44J (Loop 3) 1-TS-68-67J (Loop 4)

68-A P-12 LO-LO TAVG STM DUMP BLOCK (Page 1 of 1) Probable Cause: Corrective Action:

References:

WBN Unit 1 Post-SGTR Cooldown Using Steam Dump 1-ES-3.3 Rev. 0000 Page 16 of 23 Step Action/Expected Response Response Not Obtained

14. MONITOR RCP operating parameters:
  • Number 1 seal differential pressure greater than 200 psid.
  • Number 1 seal leakoff flow greater than 0.2 gpm.

WHEN normal RCP operation parameters can NOT be maintained, THEN STOP RCP. 15. DETERMINE if RHR should be placed in service:

a. CHECK the following RHR operation limits:
  • RCS temperature less than 375°F [360°F ADV].
  • RCS pressure less than 400 psig [250 psig ADV], on RVLIS - ICCM PLASMA

DISPLAY. a. IF RCS temperature OR RCS pressure is greater than or

equal to RHR limits, THEN

    • GO TO Step 11.
b. PLACE RHR in shutdown cooling:
  • REFER TO 1-SOI-74.01, Residual Heat Removal System. c. USE RHR for RCS cooldown.
d. STOP depressurization of Ruptured S/G by CLOSING the following:
  • Ruptured S/G MSIV bypass valve
  • Ruptured S/G PORV
  • Condenser Steam Dump Valves

This is the tie in for the steam dumps to the main steam header. Notice the lack of an isolation valve. za* " I I* W I ... ... , 12*_... ... 1-689 1-558 --MAIN STEAM STOP ANO CONTROL VALVE LH ASSEMBLY 1 GV-1 GV-3 1 t

  • c
  • 0 These are the steam dumps. These are the steam dump isolation valves. 1-244A I -... *----*** --** SEE NOTE 16---* &n * ..... 0 ..... "" .., CD CD
  • f ' ---CONDENSER

.. wo ' BYPASS A DRAIN_, -208 36" -rN ..,, l'f") ""' .., CD CD CD 1 ' I ... ... ... ZONE A INTERCEPT VALVE ..., CD ' I U"J "" CD ' ... 0 = *'-----an --3 * ..... ..,, --5 MAIN STEAM DUMP DRAIN HEADER I NOTE 16 l_1. WARMING LIN BYPASS A DRAIN -COND --..&....--____,; ......_ ___ _,_ __ ....._ __ _0 ZONE B CONDENSER -INTERCEPT VALVE

10. Given the following conditions: - Unit 1 was at 100% power. - Unit 1 has a faulted SG. Which ONE of the following describes a reason for ENSURING the MSIVs and MSIV Bypass valves are CLOSED AND if OPERATOR action IS REQUIRED to perform this step? In accordance with the Westinghouse background document for 1-E-2, Faulted Steam Generator Isolation, a r eason for ensuring that the MSIVs are CLOSED is to

____(1)____. Operators ____(2)____ REQUIRED to manipulate equipment to ensure that the MSIV bypasses are CLOSED. A. (1) prevent a sustained return to power (2) ARE B. (1) prevent a sustained return to power (2) ARE NOT C. (1) isolate the break and isol ate the SGs from each other (2) ARE D. (1) isolate the break and isol ate the SGs from each other (2) ARE NOT CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: In accordance with the Westinghouse background document for 1-E-2, all MSIVs and bypass valves are checked to be closed in this step [step 1] in an attempt to isolate the break and to isolate the SGs from each other. Therefore, it is correct that the basis for the step listed is to isolate the break. Therefore, it would be plausible to believe that such basis would be included in the WOG document. However, not only is the prevention of the re-criticality not included as a basis for the step, the re-criticality of the core is accounted for by Chapter 15 of the Watts Bar FSAR. Westinghouse studies have shown that for the design basis steam line rupture, which a return to power is predicted (and will be turned by ECCS boron injection). Additionally, Westinghouse has accounted for the fact that the MSIVs will fail to close as the design basis steam line break accounts for both a Uniform cooldown (indicating that ALL SGs are feeding the break - as if the MSIVs were all OPEN) and a non-uniform (MSIVs shut) cooldown. Because of this, Westinghouse: (1) determined that for the design basis steam line break a return to power was predicted whether or not the MSIVs were closed and (2) that ECCS boron injection would provide the required negative reactivity to shut the core down., Also, it is Incorrect to believe that action is required to close the MSIV bypasses. B. Incorrect: Again it is Incorrect to believe that the basis for closing the MSIVs is to prevent a return to criticality. However it is true that action is not required to ensure that the MSIV bypasses are closed. C. Incorrect: Again, the Westinghouse background document only recognizes two bases for ensuring that the Main Steam Isolations were closed. Sound reactor principles would indicate to one that shutting such isolations would stop the plant cooldown and thus stop the positive reactivity addition to the core. However, it is not true that the control room staff is required to manipulate anything to ensure that the MSIV bypass valves are closed. As seen in pages 15-18 of 1-SOI-1.01, the MSIV bypass valves are powered and opened when steam line warming is required during plant startup. After the MSIVs are opened, the MSIV bypasses are closed and depowered. Therefore, no manipulations are required by the control room staff to ensure that the MSIV by passes are closed. It is plausible to believe that control room action would be required because even though these bypass valves are de-energized in accordance with the aforementioned SOI, as seen on print 1-45W600-2, they are designed with the same isolation circuitry as the MSIVs (i.e. that they close on a main steam isolation signal). In effect, one could believe that these valves were normally open because of the associated safeguards which were implicit to the valve design. Because the #2 MSIV failed to automatically close, it is reasonable to believe that the #2 MSIV bypass valve would have failed to close (if it were indeed open). D. Correct: As mentioned the basis for the isolation step is to isolate the break. Also, no manipulation is required to ensure that the MSIV bypasses were closed.

Question Number: 10 Tier: 1 Group: 1

K/A: 040 Steam Line Rupture AK3. Knowledge of the reasons for the following responses as they apply

to the Steam Line Rupture: AK3.01 Operation of steam line isolation valves

Importance Rating: 4.2 4.5 10 CFR Part 55: (CFR 41.5,41.10 / 45.6 / 45.13)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to .understand the reasons for ensuring that the MSIVs and MSIV bypasses are closed during step 1 of 1-E-2, Faulted Steam Generator Isolation. Technical

Reference:

Figure 15.4-12a of chapter 15 of the WBNP FSAR 1-45W600-2 1-SOI-1.01, Main Steam System Westinghouse background document for 1-E-2 Proposed references to be provided: None Learning Objective: 3-OT-STG-001A, MAIN STEAM 01. DESCRIBE the design criteria, purpose and/or functions of the Main Steam System, subsystems, and the major system components listed

below:

d. MSIVs & bypass valves Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

STEP DESCRIPTION TABLE FOR E-2 Step _1_ STEP: Check Main Steamline Isolation And Bypass Valves -CLOSED PURPOSE: To ensure that the steamline isolation and bypass valves have closed BASIS: Since the guideline is entered after symptoms of a faulted SG have been identified, the main steamline isolation signal should have been previously actuated and the MSIVs and bypass valves should have previously received a "CLOSE" signal. On the ERG reference plant, the MSIVs and bypass valves in all main steamlines receive the same isolation signal and, therefore, all SG main steamlines should be isolated. Consequently, all MSIVs and bypass valves are checked to be closed in this step in an attempt to isolate the break and to isolate the SGs from each other. If any valves have not received a CLOSE signal or if the valves failed to close, the operator is instructed to manually close the valves. Some plants may have individual loop low steamline pressure isolation capability. On these plants, the step should be modified to indicate that only the MSIVs and bypass valves on those loops for which a main steamline isolation signal is present should be checked for closure and manually closed if necessary. ACTIONS: o Determine if main steamline isolation and bypass valves are closed o Close appropriate valves INSTRUMENTATION: Main steamline isolation and bypass valve indication CONTROL/EQUIPMENT: Main steamline isolation and bypass valve controls E-2 Background HE2BG.doc 31 HP-Rev. 2, 4/30/2005 WBN Unit 1 Main Steam System 1-SOI-1.01 Rev. 0003 Page 16 of 91 Date________ Initials_____ 5.1 MSIVs and Warming Valves Closed (continued) CAUTION During initial warming with RCS T avg >300°F, opening one MSIV bypass at a time with a 2-hr wait between each opening is recommended to allow main steam header to drain and to minimize water hammer Main Steam piping Heatup Rate should NOT exceed 100 °F/hr and must NOT exceed 200°F/hr. NOTE Warm-up time of up to 16 hrs may be needed to warm steam lines properly, dependent on initial temp and conditions. [12] () WHEN desired to warm the Main Steam Header, THEN PERFORM the following: NOMENCLATURE LOCATION POSITION UNID INITIAL WARMING THROTTLE VALVE S/G LOOP 4 SVR/729 (1) 1-THV-1-8 WARMING THROTTLE VALVE S/G LOOP 1 SVR/729 (1) 1-THV-1-18 WARMING THROTTLE VALVE S/G LOOP 2 NVR/729 (1) 1-THV-1-28 WARMING THROTTLE VALVE S/G LOOP 3 NVR/729 (1) 1-THV-1-48 SG 1 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-147 SG 2 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-148 SG 3 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-149 SG 4 MSIV BYPASS WARMING LINE 1-M-4 OPEN 1-FCV-1-150 (1) Valves may be open or throttled, as necessary, to maintain steam line warming rate to less than 100°F/Hr WBN Unit 1 Main Steam System 1-SOI-1.01 Rev. 0003 Page 17 of 91 Date________ Initials_____ 5.1 MSIVs and Warming Valves Closed (continued) NOTE REFER TO Appendix C for MSIVs opening guidelines. [13] () WHEN steam lines are fully warm ed, and Criterion A, B, or C of Appendix C is met , THEN OPEN MSIVs (listed below) ONE at a time , AND ALLOW conditions to stabilize between each opening: NOMENCLATURE LOCATION POSITION UNID INITIAL MSIV SG 1 1-M-4 OPEN 1-FCV-1-4 MSIV SG 2 1-M-4 OPEN 1-FCV-1-11 MSIV SG 3 1-M-4 OPEN 1-FCV-1-22 MSIV SG 4 1-M-4 OPEN 1-FCV-1-29 [14] CLOSE MSIV Bypasses (listed below), AND ENSURE handswitches are placed in CLOSE: NOMENCLATURE LOCATION POSITION UNID INITIALS SG 1 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-147 IVSG 2 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-148 IVSG 3 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-149 IVSG 4 MSIV BYPASS WARMING LINE 1-M-4 CLOSE 1-FCV-1-150 IV WBN Unit 1 Main Steam System 1-SOI-1.01 Rev. 0003 Page 18 of 91 Date________ Initials_____ 5.1 MSIVs and Warming Valves Closed (continued) NOTE LCO tracking of OR 14.10.2 may be discont inued when the handswitches in Step 5.1[15] are returned to OFF. [15] PERFORM the following: NOMENCLATURE LOCATION POSITION UNID INITIALS MAIN STEAM ISOL VLV LOOP 1 BYP WARMING VLV A3U/737 OFF 1-HS-1-147B IV MAIN STEAM ISOL VLV LOOP 2 BYP WARMING VLV A4U/757 OFF 1-HS-1-148B IV MAIN STEAM ISOL VLV LOOP 3 BYP WARMING VLV A5U/757 OFF 1-HS-1-149B IV MAIN STEAM ISOL VLV LOOP 4 BYP WARMING VLV A3U/737 OFF 1-HS-1-150B IV[16] IF Steam Dumps are available to maintain RCS temperature, THEN REFER TO 1-SOI-1.02, Steam Dump System. ________ NOTES 1) Temporary gauges will be removed prior to Steam H eader pressure exceeding 300 psig.

2) Step 5.1[17] may be N/A d if Step 5.1[1] was N/A d. [17] NOTIFY Instrument Maintenance (M IG) to remove temporary instruments installed in Step 5.1[1] NOMENCLATURE LOCATION TEMP GAUGE UNID PERFORMED BY MAIN STEAM LOOP 1 PRESSURE A3U/713, PNL-1-L-102 REMOVED 1-PT-1-5 Inst Maint.MAIN STEAM LOOP 2 PRESSURE SBVRMR/729 PNL 1-L-106 REMOVED 1-PT-1-12 Inst Maint.MAIN STEAM LOOP 3 PRESSURE SBVRMR/729 PNL 1-L-106 REMOVED 1-PT-1-23 Inst Maint.MAIN STEAM LOOP 4 PRESSURE A3U/713 PNL-1-L-102 REMOVED 1-PT-1-30 Inst Maint.MAIN STEAM HEADER PRESSURE T5J/708 1-L-482 REMOVED 1-PT-1-33 Inst Maint.

Cl OJ >-OJ >-::> <fl TVA NO. 1SWNA1 [ FCV-1-147 STEAM LINE WARMING VALVE (TRAIN A) DESCRIPTION FCV-1-147 ST L WARNING VALVE LOOP 1 FCV-1-148 ST L WARNING VALVE LOOP 2 FCV-1-149 ST L WARNING VALVE LOOP 3 FCV-1-150 ST L WARNING VALVE LOOP 4

  • 1SW l-C43M _,,, K617 I 1 ,2 IR-48 2SW II-C48 K617 I 1 ,2 IR-51 3SW 1-C44 r K624 I 1 ,2 IR-48 4SW II-C49 t' -

K624 I 1, 2 I R-51 AMENDMENT6WATTSBARNUCLEARPLANTFINALSAFETYANALYSISREPORTTRANSIENTRESPONSETOSTEAMLINEBREAKWITHSAFETYINJECTION&WITHOUTOFFSITEPOWER(CASEB)FIGURE15.4-12a

11. Given the following conditions:

Which ONE of the following describes the event which caused the indications shown above? A. Loss of control air to the MFW Reg Valve B. A Feed line break in the steam vault room C. A failure of the Main feed pump speed control D. A DCS failure which caused the MFW Reg Valve to FAIL OPEN 

  < SEE THE NEXT PAGE FOR A MAGNIFIED VIEW OF THE RECORDERS >

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: The main feedwater regul ating valves fail CLOSED on a loss of instrument air. Theref ore, feedwater flow (t he red recorder trace) would sharply drop to 0. 00 MPPH. Steam flow (the green trace) would remain relatively constant (at 100 % power it rises to just at 4.09 MPPH). Level (the blue trace) would drop quickly. B. Correct: The indications show on t he S/G #1 trend recorder resulted from a feed line break which occurred in the main steam vault room (outside of containment). The fault caus ed feedwater to be robbed from the S/G thus causing S/G level to dr op. Feedwater flow increased (as seen) because the MFRV opened in response to a drop in S/G level and thus fed the feed break more. Steam flow rises slightly as the S/G saturation pressure rises slightly due to the reduced heat removal from RCS loop #1 (on account of t he reduced feedwater injection). C. Incorrect: A failure of the MFP speed cont rol would affect all S/Gs equally. One may mistakenly believe that the MFRVs would act quickly enough to compensate for the failure of the MFP speed control; however, if that were the case then then no large effe ct would be noted. In actuality, the MFP speed change will drive a c hange in all S/Gs' level which will in turn cause the individual MFRVs to automatically attempt to restore their respective S/G's level. D. Incorrect: A DCS level control failure which caused the MFRV to fail open would result in S/G level rising. Main Feedwater flow would rise as well.

Question Number: 11 Tier: 1 Group: 1

K/A: 054 Loss of Main Feedwater (MFW) AA2. Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW): AA2.08 Steam flow-feed trend recorder Importance Rating: 2.9 3.3

10 CFR Part 55: (CFR: 43.5 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to interpret the Steam-flow-feed trend record in order to diagnose a specific form of a loss of main feedwater (in the question given a main feedline break to S/G #1 is presented). Technical

Reference:

Proposed references to be provided: None Learning Objective: 3-OT-AOI3800, Main Steam of Feedwater Line Leak 2. RECOGNIZE Entry Conditions for 1-AOI-38, Main Steam or Feedwater Line Leak. (IER 11-3; Monitoring plant indications and conditions closely) Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

12. Which ONE of the following describes a potential implication if SG Pressure is reduced to 50 psig during depressurization of SGs in accordance with 1-ECA-0.0, Loss of All AC Power? A. A RED Path on PTS may occur B. Natural Circulation may be impeded C. Unacceptable Upper Head Voiding may occur D. The integrity of the RCP seals may be challenged

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: Plausible since the rate of cooldown may be a concern; but final pressure/temperature is being evaluated during the cooldown. Pressurized thermal shock is a concern when RCS remains pressurized, however the RCS is intentionally being de-pressurized. B. Correct: Nitrogen gas being discharged from the CLAs could accumulate in the RCS hot legs and SG U-tubes and potentially inhibit natural circulation. C. Incorrect: Plausible since upper head voidin g will occur after pressurizer level is lost. However this is an acceptable consequence to minimize RCS inventory loss through RCP seal degradation. D. Incorrect: Plausible since the cooldown and depressurization of the RCS to cool RCP seals and limit RCS inventory loss, however the RCP seal

cooling will only be a concern if th e RCS depressurization is stopped or RCS is allowed to heat up, not if the RCS is cooled down further than it should be at this point.

Question Number: 12 Tier: 1 Group: 1

K/A: 055 Station Blackout EK1.02 Knowledge of the operational implications of the following concepts as they apply to the Station Blackout: Natural Circulation Cooling

Importance Rating: 4.1 / 4.4

10 CFR Part 55: 41.8, 41.10 10CFR55.43.b: Not applicable K/A Match: This question matched the K/A by testing the candidates knowledge of the effect of LOWERING RCS pressure below the minimum value during Natural Circulation cool down. Technical

Reference:

Westinghouse background document for 1-ECA-0.0, Loss of All AC Power

Proposed references to be provided: Learning Objective: 3-OT-ECA0000, Emergency Contingency Actions, ECA-0.0, 0.1 & 0.2 3. Explain why intact SGs are depressurized to 300 psig during performance of ECA-0.0. 6. Explain why the operator is directed to maintain RCS press greater than 250 psig during RCS cooldown. Cognitive Level: Higher Lower X Question Source: New Modified Bank Bank X Question History: Bank question imported from San Onofre Nuclear Generating Station Comments: CAUTION: PURPOSE: BASIS: STEP DESCRIPTION TABLE FOR ECA-0.0 Step 16 -CAUTION 1 SG pressures should not be decreased to less than (0.07) psig to prevent injection of accumulator nitrogen into the RCS. To alert the operator that steam generator pressures must be maintained above the specified limit Steam generators should be depressurized to maximize delivery (into the RCS) of the water contained in the SI accumulators while minimizing delivery of nitrogen. Maintaining steam generator pressures above a value that prevents introduction of a significant volume of nitrogen into the RCS ensures that accumulator nitrogen will not impede natural circulation. A steam generator pressure limit is set to preclude significant nitrogen injection into the RCS. To determine the steam generator pressure limit, an ideal gas expansion calculation should be performed based on nominal plant specific values for initial accumulator tank pressure (P 1), initial nitrogen gas volume (V 1), and final nitrogen gas volume (Vz). The final nitrogen gas volume should be equivalent to the total accumulator tank volume. The RCS pressure at empty tank conditions (P 2) is determined from: where y = 1.25 for ideal gas expansion. The steam generator pressure limit is then determined by subtracting the RCS to SG delta P from P 2* The RCS to SG delta P should be calculated as described in the RCP TRIP/RESTART section in the Generic Issues of the Executive Volume. Instrument uncertainties are not included in the determination of the steam generator pressure limit to preclude a bias toward either having more accumulator water injected into the RCS or having less nitrogen injected into the RCS. ACTIONS: Determine if SG pressures are greater than (0.07) psig INSTRUMENTATION: SG pressure indication for each SG ECA-0.0 Background HECAOOBG.doc 115 HP-Rev. 2, 4/30/2005 STEP DESCRIPTION TABLE FOR ECA-0.0 Step 16 STEP: Depressurize Intact SGs To (0.08) PSIG PURPOSE: To depressurize the intact steam generators BASIS: Step 16 depressurizes the intact SGs, thereby reducing RCS temperature and pressure to reduce RCP seal leakage and minimize RCS inventory loss. The advantages to performing this action, as well as restrictions that apply during the action, are detailed in Subsection 2.3. During SG depressurization, SG level must be maintained above the top of the SG U-tubes in at least one SG. Maintaining the U-tubes covered in at least one SG will ensure that sufficient heat transfer capability exists to remove heat from the RCS via either natural circulation or reflux boiling after the RCS saturates. Step 16a requires that SG level be in the narrow range in at least one SG before SG depressurization is initiated in Step 16b. If level is not in the narrow range in at least one SG, RNO 16a instructs the operator to maintain maximum AFW flow until narrow range level is established in one SG. When narrow range level is established, SG depressurization can be started or continued via Step 16b. Step 16b instructs the operator to reduce SG pressures by depressurizing the intact SGs. Depressurization should be accomplished by opening the PORVs on the intact SGs to establish a maximum steam dump rate, consistent with plant specific constraints. The step is structured assuming that the operator can open and control SG PORVs from the control room. This structure assumes that the PORVs are air-operated and have de control power and pneumatic power (i.e., either air reservoirs or nitrogen bottles) available. Some plants may not have the capability to open the SG PORVs from the control room. These plants should evaluate their capability to accomplish this step locally via PORV handwheels. Such an evaluation should consider accessibility and communications necessary to accomplish local PORV operation. Once depressurization is initiated, maintenance of a specified rate is not critical. The depressurization rate should be sufficiently fast to expeditiously reduce SG pressures, but not so fast that SG pressures cannot be controlled. It is important that the depressurization not reduce SG pressures in an uncontrolled manner that undershoots the pressure limit, thus permitting potential introduction of nitrogen from the accumulators into the RCS. ECA-0.0 Background HECAOOBG.doc 122 HP-Rev. 2, 4/30/2005 STEP DESCRIPTION TABLE FOR ECA-0.0 Step 16 During SG depressurization, AFW flow may have to be increased to maintain the required SG narrow range level. Control of AFW flow will have to be performed from the control room or locally depending on plant specific design. Full AFW flow should be established to any SG in which level drops out of the narrow range. RCS cold leg temperatures should be monitored during SG depressurization to ensure that the depressurization does not impose a challenge to the Integrity Critical Safety Function. This check is included in Step 16c since guideline ECA-0.0 has priority over the Function Restoration Guidelines and the operator is instructed to not implement a Function Restoration Guideline even if a Critical Safety Function challenge is detected by the Critical Safety Function Status Trees. Consequently, Step 16c implicitly protects the Integrity Critical Safety Function. The SG depressurization should not result in a challenge to the Integrity Critical Safety Function since the resultant RCS cold leg temperatures should not approach the temperature limit (i.e., T2 temperature) at which a challenge will exist. Once the target SG pressure is reached, the SG PORVs and AFW flow should be controlled to maintain SG pressure at the target value until ac power is restored. The target SG pressure for Step 16 should ensure that RCS pressure is above the minimum pressure to preclude injection of accumulator nitrogen into the RCS. The target SG pressure should be based on the nominal SG pressure to preclude nitrogen addition, plus margin for controllability (e.g., 100 psi). To determine the steam generator pressure limit, an ideal gas expansion calculation should be performed based on nominal plant specific values for initial accumulator tanks pressure (P,), initial nitrogen gas volume (V,), and final nitrogen gas volume (V 2). The final nitrogen gas volume should be equivalent to the total accumulator tank volume. The RCS pressure at empty tank conditions (P 2) is determined from: where y = 1.25 for ideal gas expansion. The steam generator pressure limit is then determined by subtracting the RCS to SG delta p from P, and adding the margin to controllability. The RCS to SG delta p should be calculated as described in the RCP TRIP/RESTART section in the Generic Issues of the Executive Volume. Instrument uncertainties are not included in the determination of the steam generator pressure limit to preclude a bias toward either having more accumulator water injected into the RCS or having less nitrogen injected into the RCS. ECA-0.0 Background HECAOOBG.doc 123 HP-Rev. 2, 4/30/2005

13. Given the following portion of the wiring diagram for the 1A-A CCP:

Which ONE of the following co mpletes the statements below? The BLACKOUT (BOX/BOY) contacts which are depict ed in the print excerpt shown above are drawn in the _____(1)_____ state. The wiring diagram from which the portion shown above was taken is a _____(2)_____ series print. (1) (2) A. BLACKOUT 45W760 B. BLACKOUT 47W611 C. normal board energized 45W760 D. normal board energized 47W611 CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: The wiring diagram excerpt wh ich is presented in the stem of the question was taken from 1-45W760-62-1. Such print contains note 2 which states: "Blackout (BO) relays are shown in the reset state which is the blackout state." Therefore, during a loss of offsite power (which causes a shutdown board blackout), the charging pump (CP) blackout X (BOX) and blackout Y (BOY) contacts which are depicted will be open. It is plausible to believe that they would be shut as it is a common misconception to believe that the BOX and BOY contacts which are shown on the prints change state during a blackout event. As discussed it is correct that a 45W760 series print is the source for

the presented excerpt.

The knowledge of the print series is needed because an index is not

readily available for the lookup of the control, logic, low voltage and high voltage prints. While, for exampl e, one could refer to an index to find the single line drawing for the 480VAC Shutdown board 1A1-A, one would have to use the knowledge of the naming convention to find

the high voltage print for the 1A-A Charging pump. The naming convention for the logic print is: Unit # - 47W611 - System Number -

Sheet. Therefore, the logic sch eme for the 1A-A CCP would be depicted on a sheet of the 1-47W611-62 prints. The print excerpt could be found by finding the high vo ltage print germane to the 1A-A CCP with the naming convention: Unit # - 45W760 - System Number - sheet. Without the possession of the knowledge of the naming convention, one would leaf though t housands of prints before arriving at the correct one. B. Incorrect: While it is correct that the BOX and BOY cont acts would be open during a shutdown board blackout, it is Incorrect that the print excerpt was taken from a 47W611 series print. Such print depicts the

electrical logic diagram and does not show contact points. It is a series of electrical print and as such it is plausible to believe that it

would be the source for the excerpt. C. Incorrect: As mentioned it is Incorrect that the BOX and BO Y contacts presented would be closed after a blackout of the shutdown board. It is plausible to believe that they would be shut as it is a common misconception to

believe that the BOX and BOY contac ts which are shown on the prints change state during a blackout event. As discussed it is correct that a 45W760 series print is the sour ce for the presented excerpt. D. Incorrect: Again it is Incorrect and yet plausible that that the BOX and BOY contacts would be shut. Also, it is Incorrect and yet plausible that the print excerpt was taken from a 47W611 series print.

Question Number: 13 Tier: 1 Group: 1

K/A: 056 Loss of Offsite Power 2.2.41 Ability to obtain and interpret station electrical and mechanical

drawings. Importance Rating: 3.5 3.9

10 CFR Part 55: (CFR: 41.10 / 45.12 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant must be able to recognize the correct series of print which display the contacts for the Blackout relays and thus possess the knowledge required to obtain the station electrical drawings germane to the loss of offsite power. Also, the applicant must be able to correctly identify the status of the blackout relays' contacts (i.e. be able to interpret the drawing). Technical

Reference:

1-45W760-62-1 1-47W611-62-4 Proposed references to be provided: None Learning Objective: 3-OT-SYS-201B, BLACKOUT AND LOADSHED LOGIC RELAYS 1. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on the Blackout and Load Shed Logic Relays. 2) UVX, UVY relays

3) BOX, BOY relays Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: NOTES: These Notes are taken from the 45W760-62-1 print. The highlighted note is a fundamental knowledge item required to interpret this series of print. ETE U-211 EC. S1ATX 20 1-XS-62-108 20C NOR. ,,... < Vl - < (/) CP AX1A 0 5 SEC TOPU SA18CN XS-62-108 NOR. F 7 ! K608 -;]?A CLOSES ON (.:) 8 51 ( 1-R-+8) ( 54114. < SA18C7 7246011-20) <n CP BOX CP BOY 22 1-XS-62-108 22C NOR .. c 0 0 c 0 D 9 52STA b 7 a11X 11 30RX S 10 ----1-02-62-108A-A -< en 3 1 30X 4 2 16 16C u co ,,... < Vl 1-XS-62-108 NOR. 1 51X 7 1-XS-62-108 NOR. START 14 14C SA18CJ CP UVY CP BOY CP AX-1A INST c 0 F E 0 F NC 0 c 4C CP uvx CP BOX J CP AX-1A 1-62X-62-108-A 2 QPR 1 c 1r---------. 0 This is the source F of the excerpt. This is the 45W760 E series print. These are the "high 0 voltage" electric F schematics which as seen, do show 2 electrical contacts. 0 7 8 52 STA b CENTRIFUGAL CHARGING PUMP 1A-A (SEE NOTE 7)

14. Which ONE of the following describes the procedural guidance contained in the AOIs AND the reason for the action?

For a loss of 120V AC Vital In strument Power Board _____(1)_____ ALWAYS requires that the Unit 1 reactor be tripped.

This is based on the condition that _____(2)_____ CANNOT be controlled. A. (1) 1-I, 1-AOI-25.01 (2) SG level B. (1) 1-I, 1-AOI-25.01 (2) PZR level C. (1) 1-IV, 1-AOI-25.04 (2) SG level D. (1) 1-IV, 1-AOI-25.04 (2) PZR level NOTE: 1-AOI-25.01, Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.04, Loss of 120V AC Vital Instrument Power Boards 1-IV or 2-IV CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: DCNs 52853A and 56905 upgraded various controls to the Distributed Control System (DCS). Accordingly, 1-AOI-25.01 was revised and an immediate action to trip the reactor was remove. This may be seen in the 4/7/14 revision to 1-AOI-25.01. DCS takes power from the 120VAC preferred power boar d and converts that to a logic power which is auctioneered with a logic power created from a 120V AC Vital Instrument board. Therefore, controllers which would have previously failed in automatic and manual (as their plug-mold power source was lost) currently continue to function upon the loss of a vital AC board. This rationale is true of the Main Feed Regulating Valves. Before DCS was installed on the Main Feed System, the MFRV's controllers would fail their associated valves shut on a loss of 120V AC

Vital Instrument Boards 1-I or 1-I I. Therefore, the reactor required an immediate reactor trip to prevent an automatic reactor trip on the impending loss of S/G level. Curr ently, the loss of 120V AC Vital Instrument Board 1-I simply caus es the 120V AC Preferred Power supply to assume the load imparted by the DCS (it does so

seamlessly). Accordingly it is plausible to believe that a reactor trip is required upon the impending loss of S/G level.

B. Correct: As seen in 1-AOI-25.01, se ction 3.2, "Loss of 120V AC Vital Instrument Power Board 1-I," a reacto r trip is tripped on account of the loss of pressurizer level control. Also, it is evident that the reactor trip is not optional as it is a left hand co lumn step. The loss of pressurizer level control results from the loss of control air to the letdown isolation valves (which fail shut). Therefore, pressurizer level would continue to rise until a high pressurizer le vel automatic trip resulted. C. Incorrect: 1-AOI-25.04 contains a response not obtained step 3 which states: TRIP reactor, and ** GO TO 1-E-0, Reactor Trip or Safety Injection. Therefore, it is incorrect that a tr ip is always required as it is only conditionally required and again it is incorrect and yet plausible that a loss of S/G level control precipitates the Rx Trip. D. Incorrect: While it is true that a reactor trip is required because a loss of Pzr level control exists during the performance of 1-AOI-25.01, it is not true for an entry into 1-AOI-25.04. Howeve r, as a conditional Rx Trip is directed in 1-AOI-25.04, it is plausible to believe such.

Question Number: 14 Tier: 1 Group: 1

K/A: 057 Loss of Vital AC Electrical Instrument Bus AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: AK3.01 Actions contained in EOP for loss of vital ac electrical instrument bus

Importance Rating: 4.1 4.4 10 CFR Part 55: (CFR 41.5,41.10 / 45.6 / 45.13)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to identify the reason why the operating crew is directed by 1-AOI-25.01 trip the

reactor. Technical

Reference:

1-AOI-25.01, Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.04, Loss of 120V AC Vital Instrument Power

Boards 1-IV or 2-IV Proposed references to be provided: None Learning Objective: 3-OT-AOI2500, Loss of 125V AC Vital Instrument Power Boards 4. Given a set of plant conditions, DESCRIBE operator actions required in response to AOI-25, Loss of 120V AC

Vital Instrument Power Boards. Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.01 Rev. 0003 Step Action/Expected Response Response Not Obtained Page 6 of 27 3.2 Loss of 120V AC Vital Instrument Power Board 1-I

1. DISPATCH personnel to 120V AC Vital Instrument Power Bd 1-I to restore power using 1-SOI-235.01, 120V AC Vital Power System 1-I.
2. NOTIFY Load Dispatcher of impending Reactor Trip due to loss

of pressurizer level control. CAUTION

  • PZR level will continue to rise until Non-Essential air is restored to containment. This procedure takes

precedence over 1-AOI-20, Malfunction of Pressurizer Level Control System. NOTES

  • Actions to bypass containment air isolation valves, may be stopped, if voltage is restored to Vital Instrument Power Board and failed closed valves.
  • Note 1 of TS 3.6.3 is applic able for bypassing containment Isolation valves.
3. DISPATCH AUO to restore air to CNTMT using Appendix B.
4. () TRIP Rx. 5. **GO TO 1-E-0, Reactor Trip or Safety Injection

WHILE continuing this instruction.

6. ENSURE RCS Letdown isolated:
a. CLOSE letdown orifices(s).
b. CLOSE 1-FCV-62-69.
c. CLOSE 1-FCV-62-70.

WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.01 Rev. 0003 Page 2 of 27 Revision Log Revision or Change Number Effective Date Affected Page Numbers Description of Revision/Change 1 01/03/14 2, 4, 13, 15Clarified that false indication of LO-LO VCT level would result in closure of 1-LCV-62-132, which would result in opening of 1-LCV-62-136. Added prints to support to print list. [PER 796503-001] 02 04/07/14 2-13, 15, 17-24 Revised for implementation of DCNs 52853A & 56905 which upgraded various controls to the Distributed Control System (DCS). Changes

include, but are not limited to:

  • Rx Trip still perfo rmed but no longer as Immediate Action.
  • Removed transfer switches for PZR pressure and level controls. Single instrument failures no longer cause

transients.

  • Deleted section for if on RHR letdown.

Minor Editorial: Listed additional Tech Specs. Added System 63 to Appendix A for RWST level Instrumentation. 03 04/08/15 2, 17-19 Made corrections to Appendix A based on operator feedback [PER 937680-001].

WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-IV or 2-IV 1-AOI-25.04 Rev. 0002 Step Action/Expected Response Response Not Obtained Page 6 of 21 3.1 Loss of 1-IV 120V Vital Instrument Power Bd

1. DISPATCH personnel to 120V AC Vital Instrument Power Bd 1-IV to restore power to board using

1-SOI-235.04, 120V AC Vital Power

System 1-IV.

2. MONITOR PZR pressure NORMAL.

MAINTAIN PZR pressure using heaters and sprays.

3. MONITOR Tavg and Tref:
  • Tavg trending to Tref.
  • Mismatch less than 3

°F. MAINTAIN Tavg and Tref within 3 °F by adjusting turbine load or by rod insertion. IF Tavg and Tref mismatch can NOT be maintained less than 5 °F, THEN TRIP reactor, and

    • GO TO 1-E-0, Reactor Trip or Safety Injection.
4. PLACE steam dump controls to OFF:
  • 1-HS-1-103A, STEAM DUMP FSV "A".
  • 1-HS-1-103B, STEAM DUMP FSV "B".
5. ENSURE instrument power for AFW Pump 1A-S 1-SW-46-AC-S is in the NORMAL position (Bd 1-III) [A2T/692]
6. REFER TO Appendix A for equipment availability.

[C.2] 7. REFER TO EPIP-1, Emergency Plan Classification Flowchart.

15. Given the following conditions: - Unit 1 is at 100% power. - CCS Pump 2A-A is running.

- ALL 480V SDBDs have control power aligned to their NORMAL feeds. Subsequently: - 125V DC Vital Battery Board III is de-energized. Which ONE of the following describes t he response of CCS Pump 2A-A AND the handswitch indicating lights for CCS Pump 2A-A? CCS Pump 2A-A WILL ____(1)____. The indicating lights on 2-HS 59A, CCS PMP 2A-A are ____(2)____. A. (1) TRIP (2) LIT B. (1) TRIP (2) NOT LIT C. (1) CONTINUE RUNNING (2) LIT D. (1) CONTINUE RUNNING (2) NOT LIT CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: As seen on print 1-45W749-1A, the 2A-A CCS pump is powered from 480V SDBD 2A1-A. The circuit breaker for the 2A-A CCS pump receives control power from the NORMAL control bus for the 480V SDBD 2A1-A. This control bus is powered from the 125V DC Vital Battery Board III. As seen on print 1-45W760-70-1, both the Trip Coil (52TC) and the Closing Coil (52C) are powered from the normal control bus for the 480V SDBD 2A1-A. Therefore, when the bus (i.e. Battery Board III) is lost, the breaker is unable to either trip or close. Therefore, the 2A-A CCS pump will continue running after the loss of control power. It is plausible to believe that the 2A-A CCS pump will be secured following the loss of DC control power because of the blackout scheme for the Shutdown boards. During a loss of the normal 125V DC control power to the shutdown board the UVX (undervoltage relays which cause the individual loads to strip) are de-energized. This is the relay state which would cause the load to strip. Because the emergency control bus has not also been lost, the UVY relay will not also be lost. Because both the UVX and UVY relays must concurrently de-energize to cause a load to be stripped, the CCS pump will not receive a strip signal. The blackout logic applicable to the 2A-A CCS pump is seen on print 1-45W760-211-12. Also observed on print 1-45W760-70-1, all of the indicating lights for the 2A-A CCS pump (including those on 2-HS-70-59A) are powered by the normal control bus for the 480V SDBD 2A1-A. Therefore, when the bus (i.e. Battery Board III) is lost, the lights extinguish. It is plausible to believe that the lights do not extinguish as there are multiple controls in the MCR possessing indicating lights which are powered from a different source than the control power for the device. B. Incorrect: As described the 2A-A CCS pump will continue to run. Also, it is correct that the indicating lights will be extinguished following the loss of control power. C. Incorrect: While it is correct that the 2A-A CCS pump will continue to run, it is not correct that the indicating light will be lit for the pump controls. Again it is plausible to believe that such light would be lit as there are multiple controls in the MCR whose indicating lights are powered from a different source then the device's power supply. D. Correct: As discussed, the pump continues to run and the indicating light is extinguished.

Question Number: 15 Tier: 1 Group: 1

K/A: 058 Loss of DC Power AA2. Ability to determine and interpret the following as they apply to the Loss of DC Power: AA2.03 DC loads lost; impact on ability to operate and monitor plant systems

Importance Rating: 3.5 3.9 10 CFR Part 55: (CFR: 43.5 / 45.13)

10CFR55.43.b: Not applicable

K/A Match: K/A is matched because the applicant is required to understand that the breaker control power is lost following the loss of a Vital Battery Board (i.e. that a DC load is lost). The applicant must then understand the impact on his ability to operate the affected the component and to monitor that components status. Technical

Reference:

1-45W749-1A 1-45W760-70-1

1-45W760-211-12 Proposed references to

be provided: None Learning Objective: 3-OT-AOI2100, Loss of 125V DC Vital Battery Boards 3. IDENTIFY symptoms, including alarms and automatic actuations of AOI-21, Loss of 125V DC Vital Battery

Boards. Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: This is the normal alignment. 2A-A{2-80-211-A-A) ( 1-45W724-3) 2A2-A{2-B0-212-(1-45W749-2A) 9 10 10 OIL TEMP v v XFMR 2A-A (2-(b)(F-212-A-A) 6900-480V, 2000KVA. OA ,..,..,.""""' 7 % z ----

"r--------------------TEST ) 1 --J :z a.. Cl (.;> J: -P') '

00 a:::" an Cl ... WI LL..-a:: ..... 0 ..... :z ...... UC QCD 100A PNL 5 """""""" 2 L-4 a a-1 2 av ,..,.,.,,,.""' P NL 5 28 20 30 2L-480-120V -:::?: ...... 01< ct: .,.,, I " ..... Cl 31: .......... w I/') -CL w ... <>1--1 LL.. I I -N 1---...J N <t-t I a::: UCI cmm 1 >>-N ..n a:: C'I 1--m 48 .. ' NO ' ' ' CLOSE TRIP BKR 52E L.O. BKRS 52E & 52N, OC ANN -J :z a.. ELECTRICAL

3200A FRAME, 50.000A S' 6 38 10 41: 0 LIJ 0 w This indicates the source of control power to the 8 breaker. 0 w 180A * * < ' T 0 -I 0 P') I a:: CD I N \ ..,. -w o Wz a::: ::::::> w 1-w :::> (/) LL..-I 2P 2P 2-BKR-70-59-A------ TOPU 35 SEC TDPU .5 SEC t ANN s1 I UV! BOI ST (NOTE 2) LOW HOR PRESS .. T This is the pump discussed by the question. ,-...s. c) START AUTO STOP PNL 6 *

  • TDPU 5-50 SEC < ' ...J< ON< 0 I oo...cn 1-:J I za..o w ...... Z::?! I Owa::: CLl 0>-1 OVlN < 1 Cl "+ ' -< I N N ' a::: m I N --...J 1 w t"l.I (J) P"l I.LI,....

..... 31: can ..... >* cio-.... < I a::; -L&J< ON I.LI I.LIO a::x o=>. Z<*

16. Given the following conditions: - Unit 1 is at 100% power.

Subsequently: 

- Unit 1 experienced a LBLOCA. - The following VALID indications are noted: 
 - ECA-1.2, LOCA Outside Cont ainment has been completed. - RCS pressure is RISING  Which ONE of the following comp letes the statements listed below? The INCREASED _____(1)_____ in the applicable room caused the indicated lights to illuminate.

After the completion of ECA-1.2, RHR flow to _____(2)_____ cold legs exists. A. (1) TEMPERATURE (2) 2 B. (1) TEMPERATURE (2) 4 C. (1) PRESSURE (2) 2 D. (1) PRESSURE (2) 4 CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: As seen in SSD- 1-TS-30-5238A, RHR Hx RM B PIPE BREAK light on 1-M-6 is actuated by a temperature switch. The temperature switch detects the rise in room temperature resultant from a high energy line break. Therefore, it is correct that the light will illuminate on temperature.

During a LBLOCA, the RCS pressure will drop to a point at which the RHR pumps will provide low pressure injection flow. If one train of RHR develops a break such that some injection flow is lost to the break; RCS pressure will drop as the equilibrium achieved by the ECCS injection and the break outflow is disturbed (e.g. there will be more fluid going out of the system than there is going into the system).

Given this discussion, the performance of ECA-1.2, will be successful in detecting the fact that after the "B" train RHR ECCS injection is isolated, that the RCS pressure will rise. The procedure will then isolate the "B" RHR train and at its conclusion, two loops of cold leg injection will exist. B. Incorrect: While it is correct that a high temperature will cause the indicator lights to be lit, it is not correct to believe that injection to four cold legs will exist. It is plausible to believe that because of two reasons. First, one may simply have poor system knowledge and not understand the layout of the cross connect piping between the "A" and "B" trains of RHR. One could believe that a fault of the "B" RHR could be isolated and still allow the "A" train to inject to all four cold legs.

Also, one could fail to realize that ECA-1.2 shuts the cross connect valves 1-FCV-74-33 and 35 (or fail to realize that they are not reopened in step 13). C. Incorrect: It is incorrect and yet plausible that a rising pressure in a room would indicate a pipe break. The auxiliary building (in which all of these spaces are located) contains multiple pressure detectors which in addition to providing indication and alarm provide inputs to control loops in order that a specific pressure band be maintained.

It is correct that 2 loop injection would remain following the completion of ECA-1.2. D. Incorrect: Again, it is incorrect and yet plausible that the increased pressure in a space would cause a pipe break light to be lit. Also, it is incorrect and yet plausible that four loop injection would remain following ECA-1.2.

Question Number: 16 Tier: 1 Group: 1

K/A: Westinghouse E04 LOCA Outside Containment EK2. Knowledge of the interrelations between the (LOCA Outside Containment) and the following: EK2.2 Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations

between the proper operation of these systems to the operation of the facility. Importance Rating: 3.8 4.0

10 CFR Part 55: (CFR: 41.7 / 45.7)

10CFR55.43.b: Not applicable K/A Match: The K/A is matched because the applicant must consider the effect of the LOCA Outside of containment (in this case in the "B" RHR Hx Room) to the RHR system (the RHR Hx Room itself) and then use this interrelation to determine the final status of the ECCS's operation (the overall operation of the facility). Technical

Reference:

ECA-1.2, LOCA Outside Containment, Proposed references to be provided: None Learning Objective: 3-OT-ECA0101, 1-ECA-1.1 AND 1-ECA-1.2 Loss Of RHR Recirc, LOCA Outside Containment

8. Explain the purpose and basis of each step in ECA-

1.1 and 1.2. Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

WBN SSD-1-TS-30-5238A Reviewed b y: App roved b y:Date:01/27/2004 D. Echols R. S. HendersonSCALINGANDSETPOINTDOCUMENT PAGE OF REVISION 2 2 1Instrument No.:1-TS-30-5238A BISTABLES AND SWITCHES SOR M frr: M odel: (1)N/A V endor ID No:Contract No:PR W-7698 Instrument Ranges: Input:Type:Switch Switch No. 1 1-TS-30-5238A 134 De g F Fixed (3)CL INCR (2)N E SSD 1-T-30-5238A N E SSD 1-T-30-5238A N E SSD 1-T-30-5238A Function: Document: Document: Document: Setpoint  : Resetpoint  : A ccurac y  : Switch Action  : Document: 40-225 De g F N E SSD 1-T-30-5238AHead Suppression Data:N/A Associated Drawings:1-45W600-57-37, 45N1635-114, 45N1645-3, 45N1695-2, A ssumptions: N one 45N1689-2, 47W600-188, Manual ID WBN-VTM-S382-0010 Remarks: (1) Model: 201TA-BB125-JJTTX6 (2) Illuminates 1-XI-30-5238A (1-M-6) through relay RHR5 (1-R-74). Annunciates 1-XA-55-6A-113B through relay RHR5 (1-R-74) and relay RHS5 (1-R-80).

(3) Accuracy As Found: +7.9, -19.9 Deg F As Left: +/- 3.0 Deg F

(4) HIGH ALLOWABLE VALUE: 144.0 Deg F (5) DCN M-16271-A installed instrument & issued NE SSD.

(6) DCN S-36364-A added vent plug requirements. Function: RHR Heat Exchan g er Room A Tem p erature Location: EL 713 , COL A7-V , PNL 1-L-733A , RHR Heat Exchan g er Room A WBN SSD-1-TS-30-5238A Reviewed b y: App roved b y:Date:01/27/2004 D. Echols R. S. Henderson PAGE OF REVISION INSTRUMENT CALIBRATION RECORD W ID NO: 2 2 Remarks: 2Instrument No:1-TS-30-5238A HEAD: N/A (1) IM to functionally check 1-XI-30-5238A on 1-M-6 SAT [ ] UNSAT [ ] If UNSAT, explain.

(2) IM to functionally check 1-XA-55-6A-113B SAT [ ] UNSAT [ ] If UNSAT, explain. (3) HIGH ALLOWABLE VALUE OF 144.0 DEG F EXCEEDED. YES [ ] NO [ ] (4) IM to verify that vent plug is removed from switch housing.

YES [ ] TEST POINT ())(LO LIMIT AS FOUNDHI LIMITLO LIMIT AS LEFTHI LIMIT 1 2 3 4 5 6 7 8 9 A S FOUND A S LEFT RE Q UIRED SET.PT RESET SET.PT RESET 1-TS-30-5238A CL INCR SWITCH ID. / A CTIONDe g F ()LO LIMIT AS FOUNDHI LIMITLO LIMIT HI LIMIT A S LEFT 134.0 Fixed 114.1 N/A 141.9 N/A 131.0 N/A 137.0 N/ATEST EQUIPMENTTVA TAG NO.CAL DUE DATE BISTABLES/SWITCHES A S LEFT A S FOUND SET.PTCREW NOOOT ( )YES ( )NOPC ( )YES ( )NOPERFORMED BY/DATEREVIEWED BY/DATE (SIMF)INSTRUCTION NO.REV. NO.Function:RHR Heat Exchan g er Room A Tem p erature EL 713 , COL A7-V , PNL 1-L-733A , RHR Heat Exchan g er Room A Location: Watts Bar Nuclear Plant Unit 1 Emergency Operating Instruction ECA-1.2 LOCA Outside Containment Revision 0005 Quality Related Level of Use: Continuous Use Effective Date: 12-20-2010 Responsible Organization: OPS, Operations Prepared By: Nicholas Armour Approved By: Biran McIlnay Current Revision Description Minor/editorial revision: Conv erted to Word 2007 (PCR 4919).

WBN Unit 1 LOCA Outside Containment ECA-1.2 Rev. 0005 Page 3 of 5 Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS

1. ENSURE RHR suction from RCS CLOSED:
  • 1-FCV-74-1 and 1-FCV-74-9.

AND

  • 1-FCV-74-2 and 1-FCV-74-8.
2. ENSURE SI pumps hot leg injection 1-FCV-63-156 and 1-FCV-63-157 CLOSED. 3. ENSURE RCS letdown isolated:
  • Letdown isolation 1-FCV-62-69 and 1-FCV-62-70 CLOSED.
  • Excess letdown isolation 1-FCV-62-54 and 1-FCV-62-55 CLOSED. 4. ENSURE RHR hot leg injection 1-FCV-63-172 CLOSED.
5. CHECK RCS press DROPPING or stable.
    • GO TO Step 14.
6. CLOSE RHR crosstie valve 1-FCV-74-33 or 1-FCV-74-35.
7. CLOSE RHR Train A cold leg injection valve 1-FCV-63-93.

IF 1-FCV-63-93 failed OPEN, THEN ** GO TO Step 10. WBN Unit 1 LOCA Outside Containment ECA-1.2 Rev. 0005 Page 4 of 5 Step Action/Expected Response Response Not Obtained

8. CHECK LOCA isolated:
  • RCS press rising.

OPEN 1-FCV-63-93.

    • GO TO Step 10.
9. ISOLATE RHR Train A:
a. STOP RHR pump A-A, AND PLACE in PULL TO LOCK.
b. CLOSE RHR suction valve 1-FCV-74-3.
c. ** GO TO Step 15.
10. CLOSE RHR Train B cold leg injection valve 1-FCV-63-94.

IF 1-FCV-63-94 failed OPEN, THEN 

**  GO TO Step 13.
11. CHECK LOCA isolated:
  • RCS press rising.

OPEN 1-FCV-63-94.

    • GO TO Step 13.
12. ISOLATE RHR Train B:
a. STOP RHR pump B-B, AND PLACE in PULL TO LOCK.
b. CLOSE RHR suction valve 1-FCV-74-21.
c. ** GO TO Step 15.

WBN Unit 1 LOCA Outside Containment ECA-1.2 Rev. 0005 Page 5 of 5 Step Action/Expected Response Response Not Obtained

13. ENSURE RHR crosstie valves 1-FCV-74-33 and 1-FCV-74-35 OPEN. 14. IDENTIFY break location:
  • Radiation Protection surveys.
  • RHR pipe break lights [M-6].
  • ECCS pump flows.
  • Aux bldg flood alarms [M-15; light panel, Aux Bldg 757].
  • Radiation area monitor recorders 1-RR-90-1 and 0-RR-90-12A.

NOTIFY TSC of failure to identify break location.

15. DETERMINE appropriate Instruction:
  • IF LOCA outside cntmt isolated, THEN ** GO TO E-1, Loss of Reactor or Secondary Coolant.

NOTIFY TSC of failure to isolate break.

    • GO TO ECA-1.1, Loss of RHR Sump Recirculation. End of Section 0 FLOW RESTRtCTOR FE-63-173A 63-690 I* C ti--t 3/4
  • DR 6 rl --t....--1131&, ...............

1"

  • 63-637 \_SET PRESS. 600 PSI 6J-309A &. 3 63-310A 2 II FE '\ co* 63-151 :2;;;.
  • 2 SET PR 3*-* ... .. 6J-339A a* "'" .. . .. "'-'"" OR-63-4 TO RHR SPRAY CONT ON 1-471812-1.

E-4 CONT ON 47W81l-1. H-5 AT PEN. X-24 FM PEN XJO RELIEF CONT ON 1-471811-1 D-6 0 -,.., ....... ...... co,.., SEE NOTE 20..._.. I* l1JIA -a* Indicates the isolations provided by ECA-1.2 Area of Leak. .. &.131--I ..... VTV-63-16 a* , ...... 3" 63-506 1* DR 63-693 FLOW ...... RESTRtCTOR ............ .. \ PI 6J-57A 63-664 3/4* TC 'I 63-94 c 3/4* *I 63-696 TV co N co I ,.., co

  • N .. SET 12* PRESS. 3" 600 PSI
  • 63-626 FLOW RESTRICTOR OR-63-91 63-370A 63-665 3/4* TC 63-699 -,L __ -:=. MANIAY r.o C Iii ,., 11111 a* .. >
  • 63-730 ,.., 63-728 --.. 2* .. RESIDUAL mi HEAT EXCHANGERS a*
  • 8" .. a* .. SEE NOTE 21FcV\O T FCV -----a*
  • LETDOWN HX (RHR) 1-47W809-1.

B-8 a* .. 3 RESIDUAL HEAT REMOVAL PUMPS 1A-A

17.

Given the following conditions: - Unit 1 has experienced a LOCA. E-1, Loss of Reactor or Secondary Coolant is in progress. - CNTMT ISOL PHASE B (125-B) is LIT. - NO CCPs are RUNNING. - ALL SG WR levels are 38%. Subsequently: - The crew transitions to 1-FR-H.1, Loss of Secondary Heat Sink. Which ONE of the following describes the status of t he RCPs AND the FIRST actions the crew will perform in accordance with 1-FR-H.1? The RCPs were ____(1)____ when the crew entered 1-FR-H.1 AND The FIRST action(s) the crew will perform is to ____(2)____. A. (1) running (2) initiate RCS bleed and feed B. (1) NOT running (2) initiate RCS bleed and feed C. (1) running (2) establish MD AFW pump flow D. (1) NOT running (2) establish MD AFW pump flow

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: Given the conditions in the stem, the operating crew will reach step 2 of 1-FR-H.1, Loss of Secondary Heat Sink and ENSURE at least one charging pump RUNNING. The crew will recognize that the GREEN and WHITE lights LIT on 1-HS-62-108A, CCP A-A (ECCS) indicates that the 1A-A CCP has tripped. The crew would then recognize that because 1-EI-57-66, 6.9 SDB 1B-B VOLTS reads 0 that the 1B-B CCP is unavailable (as it has no power available). Because no CCP is available to RUN, the crew will execute the RNO for step 2 of 1-FR-H.1 which states, ** GO TO Cautions prior to Step 18 to initiate RCS bleed and feed. The RCPs would not be in operation at the time of 1-FR-H.1 entry. The RCPs would have been secured in accordance with 1-E-1, Loss of Reactor or Secondary Coolant. Most probably step 1 of 1-E-1 would have been utilized which states, CHECK if RCPs should remain in service: a. Phase B DARK [MISSP] STOP all RCPs. The crew could also have used the Foldout Page of 1-E-1 which identifies that the RCP trip criterion contains the receipt of a Phase B Isolation. Whichever step was selected for use, the crew would have secured the RCPs prior to entering 1-FR-H.1 upon recognizing that Annunciator 125-B, CNTMT ISOL PHASE B is LIT (which would be one of the attending indications that Phase B would be LIT on the MISSP). It is plausible to believe that RCPs would be in service upon the entry into 1-FR-H.1 because such pumps are normally left in service given that their trip criteria are NOT met. Additionally, 1-FR-H.1 directs the securing of the RCPs which could lead one to believe that the pumps would be secured in accordance with such procedure. B. Correct: The crew would initiate RCS bleed and feed and the RCPs would not be in operating at the time of 1-FR-H.1 entry. C. Incorrect: As mentioned RCS bleed and feed would be required as at least one CCP

was NOT running. It is plausible that the applicant believe that he would ESTABLISH MD AFW pump flow as at least three SG WR levels are greater than 36%. 1-FR-H.1 contains step 3 which explicates the bleed and feed criteria: Any THREE SG WR levels less than or equal to 26% [36% ADV]. The criteria would require the bracketed adverse value to be used as Phase B had occurred. Also, it is not correct but plausible to believe that the RCPs were in service. D. Incorrect: It is Incorrect to first attempt to restore MD AFW pump flow but correct to recognize that the RCPs were not in service at the time of entry into 1-FR-H.1. Question Number: 17 Tier: 1 Group: 1

K/A: Westinghouse E05 Loss of Secondary Heat Sink EK1. Knowledge of the operational implications of the following concepts as they apply to the (Loss of Secondary Heat Sink) EK1.3 Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of Secondary Heat Sink).

Importance Rating: 3.9 4.1

10 CFR Part 55: (CFR: 41.8 / 41.10, 45.3)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to possess the knowledge of the operational implications to both the receipt of annunciators (125-B) and conditions indicating signals (the listed board indications). The applicant must then apply this knowledge to the required actions listed in 1-FR-H.1. Technical

Reference:

1-FR-H.1, Loss of Secondary Heat Sink 1-E-1, Loss of Reactor or Secondary Coolant

Proposed references to

be provided: None Learning Objective: 3-OT-FRH0001 , FUNCTIONAL RESTORATION GUIDES 1-FR-H.1, FR-H.2, H.3, H.4 and 1-FR H.5

3. EXPLAIN the purpose for and basis of each step in 1-FR-H.1, FR-H.2, FR-H.3, FR-H.4 and 1-FR-H.5.
4. EXPLAIN the basis for tripping the RCPs during the

performance of FR-H.1. (IER 11-3, Having a solid understanding of plant design, engineering principles and sciences) Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Loss of Reactor or Secondary Coolant1-E-1 Rev. 0004 Page 20 of 20 Foldout Page (Page 1 of 1) SI REINITIATION CRITERIA Manually START ECCS pumps as necessary:

  • PZR level cannot be maintained gr eater than 15% [33% ADV], OR
  • RCS subcooling less than 65

°F [85°F ADV] RCP TRIP CRITERIA

  • Phase B Isolation, OR
  • One charging pump or one SI pump injecting AND RCS press reduced uncontrolled to less than 1500 psig. EVENT DIAGNOSTIC TRANSITIONS
  • IF any S/G press low or dropping uncontrolled AND S/G has NOT been isolated, THEN
**  GO TO 1-E-2, Faulted Steam Generator Isolation.
  • IF S/G radiation abnormal or S/

G level rising uncontrolled, THEN START SI pumps as necessary, AND ** GO TO 1-E-3, Steam Gener ator Tube Rupture. SUMP RECIRC SWITCHOVER CRITERIA

  • IF RWST level less than 34%, THEN
**  GO TO 1-ES-1.3, Transfer to RHR Containment Sump.

AFW OPERATION

  • IF CST volume less than 5000 gal, THEN MONITOR AFW pumps to ensure suction transfer.

WBN Unit 1 Loss of Secondary Heat Sink 1-FR-H.1 Rev. 0003 Step Action/Expected Response Response Not Obtained Page 3 of 37 3.0 OPERATOR ACTIONS CAUTION

  • If total feed flow CAPABILITY of 410 gpm is available, this Instruction should NOT be performed.
  • If an Intact S/G is avail able, feed flow should NOT be reestablished to any faulted S/G.
1. CHECK if secondary heat sink is required:
a. RCS pressure greater than any Intact S/G pressure.
a. RETURN TO Instruction in effect. b. RCS temperature greater than 375

°F [360°F ADV]. b. PLACE RHR System in service while continuing in this instruction.

  • REFER TO 1-SOI-74.01, Residual Heat Removal System. WHEN adequate RHR shutdown cooling established, THEN RETURN TO Instruction in effect.
2. ENSURE at least one charging pump RUNNING.

STOP all RCPs AND ** GO TO Cautions prior to Step 18 to initiate RCS bleed and feed. WBN Unit 1 Loss of Secondary Heat Sink 1-FR-H.1 Rev. 0003 Step Action/Expected Response Response Not Obtained Page 4 of 37 CAUTION RCS bleed and feed criteria must be monitored for immediate response if the criteria is exceeded.

3. DETERMINE if RCS bleed and feed required:
a. CHECK RCS bleed and feed required:
  • Any THREE S/G WR levels less than or equal to 26%

[36% ADV].

OR

  • RCS pressure greater than or equal to 2335 psig.
a. MONITOR RCS bleed and feed criteria:

WHEN criteria are met, THEN PERFORM Substep 3b. 

**  GO TO Step 4. b. STOP all RCPs, AND  **  GO TO Cautions prior to Step 18 to initiate RCS bleed

and feed.

WBN Unit 1 Loss of Secondary Heat Sink 1-FR-H.1 Rev. 0003 Step Action/Expected Response Response Not Obtained Page 5 of 37

4. ENSURE S/G blowdown ISOLATED. Manually CLOSE valves. 5. MONITOR CST volume greater than 200,000 gal.

INITIATE CST refill USING 1-SOI-59.01, Demineralized Water System. IF CST volume drops to less than 5000 gal, THEN MONITOR AFW pumps to ensure suction transfer. NOTE If the use of condensate flow is anticipated, then a higher PZR level will better accommodate the le vel shrink from S/G cooldown and depressurization.

6. CONTROL PZR level between 29% and 63%

[47% and 58% ADV].

WBN Unit 1 Loss of Secondary Heat Sink 1-FR-H.1 Rev. 0003 Step Action/Expected Response Response Not Obtained Page 6 of 37

7. ESTABLISH MD AFW pump flow:
a. CHECK MD AFW pump AVAILABLE.
a. ** GO TO Step 8. b. ENSURE both MD AFW pumps RUNNING.
b. START pumps from the shutdown boards.
c. ENSURE MD AFW LCVs OPEN. c.

OPEN MD LCVs from the

auxiliary control room, OR Locally OPEN MD LCVs and manual isolation valves USING 1-SOI-3.02, Auxiliary Feedwater

System. d. CHECK MD AFW pump flow greater than 410 gpm.

d. ENSURE AFW valve alignment USING 1-SOI-3.02, Auxiliary

Feedwater System.

e. CHECK NR level in at least one S/G greater than 29%

[39% ADV].

e. MAINTAIN total feed flow to S/Gs

greater than 410 gpm UNTIL NR

level in at least one S/G greater

than 29% [39% ADV]. WHEN NR level in at least one S/G greater than 29% [39% ADV], THEN RETURN TO instruction in effect. 

**  GO TO Step 8. f. RETURN TO Instruction in effect.
18. Given the following timeline:

09:00:00 Unit 1 trips. 09:20:00 A Small Break LOCA occurs. 09:50:00 The crew transitioned to 1-ECA-1.1, Loss of RHR Sump Recirculation. 1 0:30:00 The crew is performing the Response Not Obtained (RNO) for 1-ECA-1.1, Step 19.b, ESTABLISH minimum ECCS flow for decay heat removal. Which ONE of the following describes the MINIMUM ECCS flow rate that satisfies the RNO AND the basis for this requirement in accordance with the Westinghouse Background Document? In accordance with Figure 1 of 1-ECA-1.1, the MINIMUM ECCS flow rate is ____(1)____ gpm AND the basis for establishing this MINIMUM ECCS flow rate is to ____(2)____ . REFERENCE PROVIDED A. (1) 325 (2) delay depletion of the RWST B. (1) 325 (2) ensure adequate RVLIS level C. (1) 400 (2) delay depletion of the RWST D. (1) 400 (2) ensure adequate RVLIS level CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: From 0900 - 1030 (90 Min) Usi ng 1-ECA-1.1, figur e 1, the minimum amount of SI flow needed to match decay heat is approximately 325 gpm. The value of 325 gpm is in t he acceptable region using the graph from time of trip AND meets the requirement of Minimum Flow to delay RWST depletion. The Basis states the operator is then instructed to establish the minimum SI pump fl ow needed to match decay heat in order to further decrease SI pump flow and delay RWST depletion. B. Incorrect: Plausible since the value of 325 gpm would meet t he requirements of the step to match the SI flow needed to match decay heat. Also plausible, since maintaining RVLIS is identified in the procedure prior to step 20 and is accomplished by keeping ECCS pumps running (not

throttling) for core makeup. Inco rrect as to be in step 19.b requires that that RVLIS level was adequate at this poi nt in the procedure C. Incorrect: Plausible since this flow rate would be in the acceptable range on 1-ECA-1.1 figure 1, however this does not meet the intent of the step which is to reduce flow as low as possible to slow the rate of RWST depletion. The second part is correct. D. Incorrect: Plausible since this flow rate would be in the acceptable range on 1-ECA-1.1 figure 1, however this does not meet the intent of the step which is to reduce flow as low as possible to slow the rate of RWST

depletion. Also plausible, since main taining RVLIS is identified in the procedure prior to step 20 and is accomplished by keeping ECCS

pumps running (not throttling) for core makeup. Incorrect as to be in step 19.b requires that that RVLIS level was adequate at this point in the procedure.

Question Number: 18 Tier: 1 Group: 1

K/A: Westinghouse E11 Loss of Emergency Coolant Recirculation EK3. Knowledge of the reasons for the following responses as they apply to the (Loss of Emergency Coolant Recirculation) EK3.4 RO or SRO function within the control room team as appropriate to the assigned lposition, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

Importance Rating: 3.6 3.8

10 CFR Part 55: (CFR: 41.5 / 41.10, 45.6, 45.13)

10CFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate recall the reason for the actions that the crew will take to comply with the requirements of ECA-1.1 while attempting to reduce the depletion rate of the RWST while unable to transfer to CNMT sump

recirculation. Technical

Reference:

1-ECA-1.1, Loss of RHR Sump Recirculation Westinghouse Owners Group background document for 1-ECA-1.1

Proposed references to

be provided: Figure 1 of 1-ECA-1.1 Learning Objective: 3-OT-ECA0101, 1-ECA-1.1 AND 1-ECA-1.2 Loss Of RHR Recirc, LOCA Outside Containment

8. Explain the purpose and basis of each step in ECA-

1.1 and 1.2. Cognitive Level: Higher X Lower Question Source: New Modified Bank X Bank Question History: Modified bank question derived from Sequoyah. Used at Sequoyah for the 05/2013 exam. QUESTIONS REPORTfor SQN-WBN composite exams

1. W/E11 EK3.4 318Given the following plant conditions: - At 0900 the Unit 1 Reactor Trips. - At 0920 a small break LOCA occurs. - At 0950 the crew transitioned to ECA-1.1, "Loss of RHR Sump Recirculation", due to the failure of both RHR pumps. - Crew has established one train of ECCS flow per ECA-1.1. - SI flow cannot be terminated due to lack of subcooling. - At 1030 the crew is performing ECA-1.1 Step 20.b, "Monitor if ECCS flow should be terminated:"

Which one of the following:(1) identifies the minimum ECCS flow rate per "Curve 9" of ECA 1.1 that meets the intent of ECA-1.1, Step 20.b RNO, and(2) the reason for the minimum ECCS flow rate? REFERENCE PROVIDED(1) 325 gpm ECCS flow. (2) To delay RWST depletion.(1) 325 gpm ECCS flow. (2) To ensure adequate RVLIS level.(1) 400 gpm ECCS flow. (2) To delay RWST depletion.(1) 400 gpm ECCS flow. (2) To ensure adequate RVLIS level. A.B.C.D.Friday, May 15, 2015 1:45:02 PM 1 WBN Unit 1 Loss of RHR Sump Recirculation 1-ECA-1.1 Rev. 0003 Page 29 of 35 Figure 1 (Page 1 of 1) Minimum SI Flow for Decay Heat vs. Time After Trip

STEP DESCRIPTION TABLE FOR ECA-1.1 Step _Ji_ STEP: Check If SI Can Be Terminated PURPOSE: To determine if conditions have been established which indicate that one train of SI flow is no longer required BASIS: Following the reduction to one train of SI, RCS conditions may be within acceptable limits for SI termination to be allowed. The combination of a minimum subcooling and sufficient liquid level in the vessel to cover the core represents less restrictive SI termination criteria in this guideline because SI flow may prevent a subsequent reduction in RCS pressure and cause considerable depletion of the RWST. The subcooling criterion will ensure subcooled conditions and the RVLIS indication ensures the existence of an adequate vessel inventory such that core cooling is ensured. Refer to document SI TERMINATION/REINITIATION in the Generic Issues section of the Executive Volume. If the termination criteria are not satisfied, then SI is required to ensure core cooling and should not be terminated. If RVLIS indication is adequate but RCS subcooling is not, the operator is then instructed to establish the minimum SI pump flow needed to match decay heat in order to further decrease SI pump flow and delay RWST depletion. This is done by aligning (if necessary) and operating the appropriate SI pumps (charging/SI pumps, head SI pumps and low-head SI pumps) such that the flow required to match decay heat is established. For most Westinghouse plants, the flow through the SI lines cannot be throttled and the exact flow rate required cannot be established. Therefore, in order to establish the minimum SI flow required in this step, the operator should stop appropriate SI pump(s) to establish flow equal to or greater than the minimum SI flow required to match decay heat. The SI flow needed to match decay heat is a function of time and is obtained from Figure 1. Figure 1 is a generic curve with units for flowrate of gpm per MWt. Each utility must develop a plant specific curve for its plant from Figure l, and this curve would be included in the plant specific emergency operating procedure as Figure ECAll-1. This plant specific curve can be developed by modifying Figure 1 as follows: The Y-axis values for flowrate in gpm/MWt should be multiplied by the plant specific MWt core rating to obtain flowrate values in GPM. The X-axis values for time in minutes are used without modification. A plant specific curve is then plotted as flowrate (gpm) versus ECA-1.1 Background HECAllBG.doc 40 HP-Rev. 2, 4/30/2005 WBN Unit 1 Loss of RHR Sump Recirculation 1-ECA-1.1 Rev. 0003 Page 29 of 35 Figure 1 (Page 1 of 1) Minimum SI Flow for Decay Heat vs. Time After Trip

19. Given the following conditions: - Unit 1 is performing a power ascension from 60% to 100%. - In accordance with the Reactivity Management Plan, the OAC is withdrawing control rods and notices that CBD rod D 12 has stopped moving with its group. MON-85-5000/1, CERPI MONITOR 1 is displaying Shutdown Banks Rod Positions. MON-85-5000/2, CERPI MONITOR 2 is displaying Control Banks Rod Positions.

Subsequently: MON-85-5000/2, CERPI MONITOR 2 goes BLANK. Which ONE of the following describes the Unit 1 T/S, LC O 3.1.5, Rod Group Alignment Limits AND the Main Control Room (MCR) ONLY Indications for Control Rod Bank Positions? T/S, LCO 3.1.5 would NOT be met when D12 is ____(1)___ steps from the group step counter demand position.

1-MON-85-5000/1, CERPI MONI TOR 1 ____(2)____ be used to observe Control Banks Rod Positions. A. (1) 2 (2) can B. (1) 2 (2) can NOT C. (1) 12 (2) can D. (1) 12 (2) can NOT CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: In accordance with Unit 1 Technical Specification LCO 3.1.5, Rod Group Alignment Limits, All shutdown and control rods shall be OPERABLE, with all individual indicated rod positions within 12 steps of the group step counter demand posit ion It is plausible to believe that a difference of 2 steps is the point at which the T/S is NOT met as a similar Unit 1 Technical Requirement exists. T/R 3.1.7, Position

Indication System, Shutdown declares that The group demand position indicators shall be OPER ABLE and capable of determining within +/- 2 steps the demand positi on for each shutdown or control rod that is not fully inserted. As seen in the CERPI Technical Description, WNA-CT-00053-WAT, each of the Operator Flat Panel Displays possesses 5 operator selectable screens. The two screens which are normally in use are (1) Control Banks A, B, C, and D Rod

Positions Display and (2) Shut down Banks A, B, C, and D Rod

Positions Display. Normally, the control room staff will have the shutdown banks' information display ed on the CERPI monitor 1. The control banks' information would be displayed on CERPI monitor 2. In

the event of a failure of a monitor the control room st aff would select the required information on the remaining operable monitor. B. Incorrect: It is neither correct that t he tolerance for LCO 3. 1.5 is within 2 steps nor is it correct that th e control room staff can NOT select the required rod position information using acti ons taken from the MCR and the remaining CERPI monitor. It is plausible to be lieve that the foregoing were true as seen on print 2E10010, all of the information produced by

the two CERPI computers (PLC A and B) is passed through a

Maintenance and Test Panel (A and B). It is conceivable that one may believe that action taken at the MT P may be required. Additionally, one may simply not recollect that each of the displays can produce the positions of all control rods because of the fact that for the vast majority of the time monitor 1 displays the shutdown banks and monitor 2 displays the control banks. C. Correct: The LCO 3.1.5 requires that rods be within 12 st eps of the group demand and the MCR staff is able to select the required information on the CERPI monitors. D. Incorrect: While it is correct that LC O 3.1.5 requires that rods be within 12 steps of the group demand, it is not true but plausible that the remaining CERPI monitor can display the required information.

Question Number: 19 Tier: 1 Group: 2

K/A: 005 Inoperable/Stuck Control Rod AA1. Ability to operate and / or monitor the following as they apply to the Inoperable / Stuck Control Rod: AA1.05 RPI

Importance Rating: 3.4 3.4

10 CFR Part 55: (CFR 41.7 / 45.5 / 45.6)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to monitor and manipulate the CERPI screens (and a remaining CERPI screen) given that one rod is stuck. The applicant must also be able to detect a stuck rod by identifying the degree of misalignment required by the T/S. Technical

Reference:

TS 3.1.5 Rod Group Alignment LimitsTR 3.1.7 Position Indication System, Shutdown WNA-GO-00010-WAT CERPI OPERATION AND MAINTENANCE MANUAL Westinghouse drawing 2E10010 Proposed references to be provided: None Learning Objective: 3-OT-SYS085A, Control Rod Drive System 2. DESCRIBE the design criteria, purpose and/or functions of the Rod Position Indicating (RPI) System and the major system components listed below: a. Computer Enhanced Rod Position Indicator (CERPI) 14. DESCRIBE the following aspects of Technical Specifications and Technical Requirements for this system:

a. The conditions and required actions with completion time of one hour or less
b. The Limiting Conditions for Operation, Applicability, and Bases. Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: Rod Group Alignment Limits 3.1.5 Watts Bar-Unit 1 3.1-8 3.1 REACTIVITY CONTROL SYSTEMS

3.1.5 Rod Group Alignment Limits

LCO 3.1.5 All shutdown and control rods shall be OPERABLE, with all individual indicated rod positions within 12 steps of their group step counter demand position.

APPLICABILITY: MOD ES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod(s) untrippable. A.1.1 Verify SDM is 1.6% k/k. OR A.1.2 Initiate boration to restore SDM to within limit. AND A.2 Be in MODE

3. 1 hour 1 hour 6 hours B. One rod not within alignment limits. B.1 Restore rod to within alignment limits.

OR B.2.1.1 Verify SDM is 1.6% k/k. OR 1 hour 1 hour (continued)

Position Indication System, Shutdown TR 3.1.7 09/22/97 Watts Bar-Unit 1 3.1-13 Revision 8 Technical Requirements

TR 3.1 REACTIVITY CONTROL SYSTEMS

TR 3.1.7 Position Indication System, Shutdown

TR 3.1.7 The group demand position indicators shall be OPERABLE and capable of determining within +/- 2 steps the demand position for each shutdown or control rod that is not fully inserted.

APPLICABILITY: MODES 3, 4, and 5, when the reactor trip breakers are closed.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more group demand position indicators inoperable. A.1 Open reactor trip breakers. Immediately

TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.1.7.1 Determine that each group demand position indicator is OPERABLE by movement of the associated shutdown or control rod 10 steps in any one direction. Within 4 hours after closing the reactor trip breakers if not completed within previous 31 days. AND 31 days thereafter

Watts Bar Unit 1 ARPI System Upgrade CERPI Technical Description 5.1.14 System Tunable Parameters Display 5.1.14.1 Following tunable parameters are displayed and can be adjusted by either numeric entry or by increment I decrement buttons as appropriate:

  • Rod Bottom Alarm setpoint and hysteresis values
  • Rod Bottom Bypass setpoint and hysteresis values
  • Rod-to-Rod Deviation Limit value
  • Rod-to-Bank Deviation Limit value
  • Rod Withdrawal Limit value
  • Wrong Direction Alarm setpoint value 5.1.15 Alarm Selection Display 5.1.15.1 A display screen that contains an Enable/Disable toggle button for the following alarms is provided:
  • Rod-to-Rod Deviation Alarm
  • Rod-to-Bank Deviation Alarm
  • Rod Bottom Alarm
  • Any Rod On Bottom Alarm
  • All Rods On Bottom Alarm
  • Rod Withdrawal Limit Alarm
  • ROTC Trouble Alarm
  • Rod Wrong Direction Alarm 5.2 OPERATOR FLAT PANEL DISPLAY (Of PD) 5.2.1 General 5.2.1.1 There are 2 OFPDs in the Main Control Room. 5.2.1.2 Each OFPD consists of 5 operator selectable screens that are selectable from either the toucscreen or the trackballs, which are connected to each OFPD. 5.2.1.3 Operator functions are limited to selecting screens, monitoring individual rod positions, and evaluating the system status. No operator action is capable of affecting the CERPI system calibration or programming.

5.2.1.4 Date and time is not displayed on OFPD screens. WNA-CT-00053-WAT, Rev. 1 5-7 Westinghouse Proprietary Class 2 Watts Bar Unit 1 ARPI System Upgrade CERPI Technical Description 5.2.1.5 Each OFPD consists of a PC Node Box, a touchscreen flat panel display, and a trackball. 5.2.2 Calib.ration Adjustments 5.2.2.1 PLC tunable parameters or the PLC calibration constants are not adjustable via the Of PD. 5.2.3 Security Features 5.2.3.1 There are no access-protected screens on Of PD. 5.2.4 Control Banks A, B, C, and D Rod Positions Display 5.2.4.1 This display screen shows the positions of the rods in Control Banks A, B, C, and D in a bar graph format. 5.2.4.2 The ability to identify a rod position value affected by a faulty analog input is provided. 5.2.4.3 An indication for the rod bottom position of each rod is provided. 5.2.4.4 "Any Rod On Bottom", "All Rods On Bottom", "Rod-to-Bank Deviation, to-Rod Deviation", "Rod Wrong Direction", and "RPI System Trouble" alarm signals are displayed on this screen. 5.2.4.5 RPI System Trouble indicator has touch button capability and once it is touched, screen navigates to System Status Display. 5.2.4.6 Rod Demand from Passive Summer and Rod Speed indicators are provided in bar graph format. 5.2.4. 7 Under normal operating conditions, either PLC A or PLC B can be selected to provide the information on this screen. If the main PLC (PLC A) is not available, then the information on this screen will be supplied by PLC B automatically. WNA-CT-00053-WAT, Rev. 1 5-8 Westinghouse Proprietary Class 2 * *

  • Watts Bar Unit 1 ARPI System Upgrade CERPI Technical Description 5.2.5 Shutdown Banks A, B, C, and D Rod Positions Display 5.2.5.1 This display screen shows the positions of the rods in Shutdown Banks A, B, C, and D in a bar graph format 5.2.5.2 The ability to identify a rod position value affected by a faulty analog input is provided.

5.2.5.3 An indication for the rod bottom position of each rod is provided. 5.2.5.4 "Any Rod On Bottom", "All Rods On Bottom, "Rod-to-Bank Deviation, to-Rod Deviation", "Rod Wrong Direction, and "RPI System Trouble" alarm signals are displayed on this screen. 5.2.5.5 RPI System Trouble indicator has touch button capability and once it is touched, screen navigates to System Status Display. 5.2.5.6 Rod Demand from Passive Summer and Rod Speed indicators are provided in bar graph format. 5.2.5. 7 Under normal operating conditions, either PLC A or PLC B can be selected to provide the information on this screen. If the main PLC (PLC A} is not available, then the information on this screen will be supplied by PLC B automatically. 5.2.6 All Rods Display 5.2.6.1 This display screen shows all system rod position values simultaneously. 5.2.6.2 The ability to identify a rod position value affected by a faulty analog input is provided. 5.2.6.3 "Any Rod On Bottom", "All Rods On Bottom, "Rod-to-Bank Deviation, to-Rod Deviation", "Rod Wrong Direction", and "RPI System Trouble" alarm signals are displayed on this screen. 5.2.6.4 RPI System Trouble indicator has touch button capability and once it is touched, screen navigates to System Status Display. WNA-CT-00053-WAT, Rev. 1 5-9 Westinghouse Proprietary Class 2

20. Given the following timeline:

00:00:00 Unit 1 is performing FR-I.2 , Low Pressurizer Level. RCS Subcooling indicates 0 °F. 00:0 1 :00 A PORV OPENS. RCS Pressure LOWERS 300 psia. Which ONE of the following describes the response of PZR Level AND the reason for this response? At time 00:0 1 :05 INDICATED PZR Level will be ____(1)____ ACTUAL PZR Level at time 00:0 1 :00 due to ____(2)____. (1) (2) A. HIGHER the change in specif ic volume of the PZR water B. HIGHER Containment Temperatur e and Pressure RISING C. LOWER the change in specif ic volume of the PZR water D. LOWER Containment Temperatur e and Pressure RISING CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Correct: Assume PZR pressure is 2000 psia and the actual height of water measured in the PZR is 20 feet.

Pressure at 2000 psia

32.2 ft 20 ft lbm lbfs 2 1 ft 2 s 2 0.02565 ft 3 32.2 ft lbm 144 in 2 P = 5.415 Pressure at 1700 psia 32.2 ft 20 ft lbm lbfs 2 1 ft 2 s 2 0.02428 ft 3 32.2 ft lbm 144 in 2 P = 5.720 If initial conditions (2000 psia and 20 feet) provided an accurate indication of PZR water level, a pressure of 5.415 psia corresponds to an indicated height of 20 feet. Then at the new conditions (1700 psia and 20 feet) the higher pressure generated 5.720 psia, since the pressure is higher indicated PZR level would be higher than 20 feet. To indicate 20 feet of water, the pressure must be equal to 5.415 psia. Calculate the indicated water level if 5.415 psia was created with the PZR pressure at 2000 psia. 5.415 lbf 0.02565 ft 3 32.2 ft lbm s 2 144 in 2 in 2 lbm lbfs 2 32.2 ft 1 ft 2 L = 20.01 ft Calculate the indicated water level if 5.720 psia was created with the PZR pressure at 1700 psia. 5.720 lbf 0.02428 ft 3 32.2 ft lbm s 2 144 in 2 in 2 lbm lbfs 2 32.2 ft 1 ft2 L = 19.98 ft B. Incorrect: It is plausible that the applicant may not underst and the operation of the PORV and the PRT and may think that a 300 psia decrease in pressure may cause the PRT to rupture and therefore change the external influences on the PZR Level Detector. C. Incorrect: See explanation above. The 2 n d part is correct. D. Incorrect: See explanation above. It is plausible that the applicant may not c g gz Pc g gz P1 understand the operation of the PZR Level Detector OR how the PORV and the PRT will respond a nd may think that a 300 psia decrease in pressure may cause the PRT to rupt ure and therefore change the external influences on the PZR Level Detector.

Question Number: 20 Tier: 1 Group: 2

K/A: 028 Pressurizer (PZR) Level Control Malfunction AK3. Knowledge of the reasons for the following responses as they apply to the Pressurizer Level Control Malfunctions: AK3.03 False indication of PZR level when PORV or spray valve is open and RCS saturated

Importance Rating: 3.5 4.1

10 CFR Part 55: (CFR 41.5,41.10 / 45.6 / 45.13)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the question has established a plant condition in which PZR level has malfunctioned and a condition which the indicated PZR level is false (PZR Level Malfunction and FR-I.2, Low Pressurizer Level). The question then requires that the applicant to understand how the actual and indicated PZR level will differ and the reason for the false indication. Technical

Reference:

FR-1.2, Low Pressurizer Level GFES Sensors and Controllers Proposed references to

be provided: None Learning Objective: Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

21. Given the following conditions: - Unit 1 Core offload is in progress. - N31, Channel I SRNI FAILS. AOI-4, Nuclear Instrument ation Malfunctions is entered. - Audible count rate IS audible in the main control room. - Audible count rate IS NOT audible inside containment.

Which ONE of the following describes the requirements regarding Core Offload and the required operator action in accordance with 1-AOI-4? Core offload ____(1)____ . Audio count rate in containmen t, ____(2)____ be restored by manipul ating the switch circled in RED in the following picture. A. (1) must STOP (2) can B. (1) may CONTINUE

(2) can C. (1) must STOP

(2) can NOT D. (1) may CONTINUE

(2) can NOT CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: In accordance with 1-AOI-4, the operators mu st STOP positive reactivity changes or core alterations. A subset of core alterations is fuel offload. Therefore, it is correct that core offload must stop. It is plausible to believe that fuel offl oad may continue given the failure of only one source range inst rument because of two factors. One may believe that because there are two SRNIs that fuel movement may continue as long as one is operational. Secondly, one may believe that fuel offload is permissible becaus e the core is being placed in a more conservative condition (i.e. defueled). As seen in the excerpted page from lesson plan 3-OT-SYS092A, either SRNI N31 or SRNI N32 may feed the source range level to the scalar/timer. In the question st em, N32 (the rema ining OPERABLE SRNI) is selected. The scalar/timer out puts a count rate which is fed to two amplifiers. These amplifier channels are not named N31 and N32 but are named A1 and A2. Amplifier A1 normally only feeds the

speaker in the main control room. Amplifier A2 only feeds the speaker inside of containment. In the event of the failure listed in the question, a selector switch (located on t he rear of the drawer) may be

manipulated to select the main cont rol room amplifier to provide the containment speaker. The stem of the question, therefore, presents not only a failure of SRNI N31, but al so a failure of amplifier A2. This condition would be rectified by the second bullet of st ep 10 of 1-AOI-4

which states: IF audible audio count rate is NOT being received in CNTMT, THEN: a. TURN AMPLIFIER SELECT switch to A1 [1-M-13 switch on rear of amplifier]. The s witch circled in red on the picture in the stem of the question only controls the channel providing the signal to the scalar/timer; a common misc onception is that it selects an amplifier channel (e.g. amplifier channel N31 or N32). One of the most common misconceptions is that this switch affects the amplifier channel for the MCR and t hat the switch on the rear of the drawer affects the amplifier selecti on for the containment speaker. B. Incorrect: Again, it is incorrect and yet plausible that the core offload may continue. Also, it is incorrect and yet plausible that the switch on the front of the drawer wo uld control the signal to the speaker inside of

containment. C. Correct: As discussed, core offload must stop. Only the manipulation of the AMPLIFIER SELECT switch on the back of the drawer will restore the count rate to containment. D. Incorrect: It is Incorrect and yet plausible to belie ve that fuel offload may continue. Also, it is incorrect and yet plausible that the switch on the front of the drawer wo uld control the signal to the speaker inside of containment.

Question Number: 21 Tier: 1 Group: 2

K/A: 032 Loss of Source Range Nuclear Instrumentation 2.4 Emergency Procedures / Plan G2.4.11 Knowledge of abnormal condition procedures.

Importance Rating: 4.0 4.2

10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.13)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to present knowledge of 1-AOI-4 which contains the abnormal condition procedure for the Loss of Source Range Nuclear Instrumentation. Technical

Reference:

1-AOI-4, Nuclear Instrument ation Malfunctions Westinghouse prints for the SRNI's scalar timer and downstream amplifiers

Selected Page from 3-OT-STG-092A, EXCORE

INSTRUMENTATION Proposed references to be provided: None Learning Objective: 3-OT-SYS092A, Excore Nuclear Instrumentation 1. DESCRIBE the design criteria, purpose, and/or functions of the Excore Nuclear Instrumentation System (NIS) and subsystems, and the major system components listed below: (IER 11-3: Having a solid understanding of plant design, engineering principles, and sciences)

h. Audio Count Rate Drawer 14. DESCRIBE the following aspects of Technical Specifications and Technical Requirements for this system:
a. The conditions and required actions with completion

time of one hour or less b. The Limiting Conditions for Operation, Applicability, and Bases. Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Nuclear Instrumentation Malfunctions 1-AOI-4 Rev. 0001 3.2 Source Range Monitor (SRM) Failure (continued) Page 9 of 42 Step Action/Expected Response Response Not Obtained

9. IF in MODE 3, 4, or 5, THEN:
  • CHECK audio count rate audible in control room.
  • PLACE audio count rate CHANNEL SELECTOR switch to operable channel [1-M-13].
  • REFER TO the following Tech Specs:
  • 3.3.1,Reactor Trip System Instrumentation,
  • 3.3.3, PAM Instruments,
  • 3.3.4, Remote Shutdown System. 10. IF in MODE 6, THEN:
  • CHECK audio count rate audible in control room.
  • IF audible audio count rate is NOT being received in the control room, THEN PLACE audio count rate CHANNEL SELECTOR switch to an operable channel [1-M-13].
  • CHECK audio count rate audible in Cntmt.

IF audible audio count rate is NOT being received in Cntmt, THEN: a. TURN AMPLIFIER SELECT switch to A1 [1-M-13 switch on rear of amplifier].

b. CHECK an audible count rate in Cntmt.
  • REFER TO Tech Spec 3.9.3, Nuclear Instrumentation.

OPERATIONS EXCORE INSTRUMENTATION STUDENT TRAINING GUIDE 3-OT-STG-092A REVISION 12 PAGE 44 OF 136 Table 5 Source Range Audio Count Rate Selector Switch CONTROL SWITCH POSITION OPERATION Source Range Audio Count Rate Selector Switch SR N31 Audio count rate meter monitors source range channel N31 SR N32 Audio count rate meter monitors source range channel N32 Figure 18 Audio Count Rate

22. Given the following conditions: - Unit 1 is in MODE 2. - The following alarm is DARK: Subsequently: - 120V AC Vital Instrument Board 1-II de-energizes.

Which ONE of the following describes the Unit 1 response to the failure? IMMEDIATELY after the failure, the Unit 1 Reactor will ____(1)____. ____(2)____ of the IRNIs remain(s) energized. A. (1) TRIP (2) BOTH B. (1) TRIP (2) ONLY ONE C. (1) REMAIN CRITICAL

(2) BOTH D. (1) REMAIN CRITICAL

(2) ONLY ONE

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: The reactor would trip upon t he loss of an IRNI. As mentioned below it is plausible to believe that both IRNIs would be energized. It is plausible to believe that the Reactor would trip upon the loss of a 120V AC Vital Instrument Board for a reason other than the loss of an IRNI. Therefore, the combination of the two parts of this distractor remains plausible. B. Correct: NI-36 would fail c ausing a reactor trip. It is correct that only IRNI-135 would be energized. C. Incorrect: The reactor would trip due to IRNI 136 failure and only one intermediate range monitor would be energized. It is plausible to believe that both IRNIs be energized as if a different 120V AC vital board had been lost (e.g. III or IV) both IRNIs would have been energized. As mentioned below, it is plausible to believe that the reactor would have remained critical. D. Incorrect: 1-AOI-25.02, Loss of 120V AC Vi tal Instrument Power Boards 1-II or 2-II stipulates that Intermediate Range 1-N-136 will fail upon of loss of power board 1-II. WBNs IRNIs are manufactured by Gamma-Metrics. These NIs utilize control power and instrument power. Both control

and instrument power originate from a 120V AC vital power board. Internal to the NI are bistables which trip upon detecting a high flux condition. These bistables are no rmally energized from instrument power which maintains a contact closed. The secondary of the bistable (the normally closed contact) allows control power to be supplied to the SSPS input relays. The SSPS input relays applicable to NI trips are normally energized. Therefore, upon the loss of the vital instrument board two mechanisms exist to trip the reactor: 1. the NI drawer bistables shift to a tripped state and 2. the SSPS input relays de-energize (achieve their tripped state). As seen in 1-ARI-79-C, Rx Trip First Out, when below permissive P-10, either NC-35F (a bistable

internal to IRNI-135) or NC-36F (internal to IRNI-136) will cause an Intermediate Range Flux Trip. It is pl ausible to believe that the reactor could have remained critical because if the failure had been a different Instrument Power Board then this assumption could have been true (e.g. the loss of 120V AC vital board III or IV). Additionally, it is plausible to believe that the reactor would have remained critical because had the reactor been at a power level above P-10 (10%), the intermediate range trip would have been blocked and the reactor would have remained critical.

Question Number: 22 Tier: 1 Group: 2

K/A: 033 Loss of Intermediate Range Nuclear Instrumentation AK1. Knowledge of the operational implications of the following concepts as they apply to Loss of Intermediate Range Nuclear Instrumentation: AK1.01 Effects of voltage changes on performance

Importance Rating: 2.7 3.0

10 CFR Part 55: (CFR 41.8 / 41.10 / 45.3)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to recognize the impact which a loss of voltage to a IRNI has upon the operation of the plant when the plant is operating low in the power range (below P-10). Technical

Reference:

1-ARI-76-80, Rx Trip First Out 1-ARI-64-70, Bypass Inlk & Permissive 1-AOI-25.02, Loss of 120V AC Vital Instrument Power

Boards 1-II or 2-II 1-AOI-4, Nuclear Instrumentation Malfunctions Proposed references to

be provided: None Learning Objective: 3-OT-SYS092A, Excore Nuclear Instrumentation 13. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on the NIS: (IER

11-3, Operating the plant with a conservative bias)

b. Power suppliesCognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-II or 2-II 1-AOI-25.02 Rev. 0003 2.1 Alarms (continued) Page 4 of 24 T. SSPS-B GEN WARNING [115-A] U. CL ACCUM 1 LEVEL HI/LO [131-A] V. CL ACCUM 1 PRESS HI/LO [131-B] W. CL ACCUM 2 LEVEL HI/LO [132-A] X. CL ACCUM 2 PRESS HI/LO [132-B] 2.2 Indications A. Power Range 1-N-42, Intermediate Range 1-N-136, and Source Range 1-N-132 failure. B. All channel II trip STATUS LIGHTS will ENERGIZE (Except Cntmt Level, Cntmt Hi-Hi Press And RWST Low Level). C. Train B CNTMT Isolation Panel 1-XX-55-6F will be DARK. D. Accumulator 3 & 4 Pressure, fails low 2.3 Automatic Actions A. Rod withdrawal block due to high flux rod stop. B. Charging pump suction swaps over to RWST due to de-energized separation relays. 1-LCV-62-135 and 1-LCV-62-136 open and 1-LCV-62-133 closes, which causes borated water to be injected into the RCS. C. If in service, Train A MCR, EBR, and SD BR Chillers will swap to Train B, and auto start of Train A is disabled. WBN Unit 1 Bypass, Intlk, & Pe rmissive 1-ARI-64-70 Rev. 0000 Page 8 of 47 NOTE 1 Relay K1705 is actuated when Reactor power on 2/4 PR channels is greater than or equal to10%. A. Power escalation above 10% NOTE 2 P-10 blocks manual reset of both SR channels high voltage supply, blocks SR high flux trip, and supplies input to P-7. CAUTION The P-10 permissive will automatically reset if 3/4 PR channels decrease below 10%. The reset of P-10 will unblock the IRM rod stop and high flux trip and the PRM high flux low setpoint reactor trip. [1] VERIFY Reactor power greater than or equal to 10%. [2] IF IR high flux trip and rod stop should be blocked, THEN PLACE IR TRIP BLOCK P-10 handswitchs (1-N38A ILLUM and 1-N38B) in BLOCK and VERIFY Window 65-C illuminates. [3] IF PR high flux low setpoint should be blocked, THEN PLACE PR LO POWER TRIP BLOCK P-10 handswitchs (1-N47A and 1-N47B) in BLOCK and

VERIFY Window 64-D illuminates. 1-45W600-55-9

1-47W611-99-2

W 7246D11 Sht. 40 Source Setpoint Relay K1705 (See Note 1)

64-E P-10 NUC AT POWER PERMISSIVE (Page 1 of 1) Probable Cause: Corrective Action:

References:

WBN Unit 1 Rx Trip First Out 1-ARI-76-80 Rev. 0000 Page 27 of 40 NOTE This trip may be manually blocked above P-10 and is automatically reinstated when power drops below P-10. A. Flux level above setpoint (failure to block when above P-10) B. Low power reactivity excursion [1] GO TO 1-E-0, REACTOR TRIP OR SAFETY INJECTION. 1-47W611-99-2 W 7245D11-26 1-E-0 Source Setpoint NC-35F IR channel current or equal to 25% of NC-36F full power 79-CINTERMED RANGE FLUX HI (Page 1 of 1) Probable Cause: Corrective

Action:

References:

WBN Unit 1 Nuclear Instrumentation Malfunctions 1-AOI-4 Rev. 0001 2.2 Indications (continued) Page 4 of 42 B. Intermediate Range Monitor (IRM) malfunction: 1. Erratic or loss of indication. 2. INSTRUMENT POWER ON and/or CO NTROL POWER ON lights at NIS racks DARK [1-M-13]. 3. Both startup rate channels NOT indicating same startup rate. 4. SRM does NOT energize during shutdown. 5. NON-OPERATE light LIT [1-M-13]. C. Power Range Monitor (PRM) malfunction: 1. Erratic or loss of indication.

2. Delta flux meter failed high, low, or giving erratic indications. 3. CONTROL POWER ON and/or INSTRUMENT POWER ON lights at NIS racks DARK [1-M-13]. 4. DCS Operator display for Rod Control and PR NIS indicates abnormal condition for applicable PRM. 2.3 Automatic Actions A. Source Range Monitor failure: 1. Possible source range monitor high flux Rx trip at 10 5 cps if SRM NOT blocked. 2. Possible Cntmt evacuation ala rm at 0.5 decade above background. B. Intermediate Range Monitor failure: 1. Intermediate range monitor 20% high flux rod stop (if NOT blocked above P-10). 2. Intermediate range monitor 25% high flux Rx trip (if NOT blocked above P-10). 3. Possible loss of P-6 block if intermediate range monitor greater than 1.66 X 10-4% power and intermediate range monitor fails low.
23. Which ONE of the following describes the control room indications AND the AUTOMATIC actions (if any)

IF 0-RFV-77-743J, GAS DECAY TANK J RELIEF FAILS OPEN ? ____(1)____ will provide indication of this failure to the main control room staff AND ____(2)____isolate the failure once it reached its HIGH alarm setpoint in accordance with 1-AOI-31, Abnormal Release of Radioactive Material. NOTE: 0-RM-90-118A, WGDT RELEASE LINE 1-RI-90-400, SHIELD BLDG VENT MON A. (1) 1-RI-90-400 (2) WILL B. (1) 1-RI-90-400 (2) WILL NOT C. (1) 0-RM-90-118A (2) WILL D. (1) 0-RM-90-118A (2) WILL NOT CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: It is correct that 1-RI-90-400 would provide the operators in the main control room with indication of the listed failure. However, it is not correct that an automatic isolation of the failure would occur. As seen on the control print 1-47W610-90-5, the 400 radiation monitor performs NO automatic isolations. It is plausible to believe that it does as there are radiation monitors in the plant which do so. B. Correct: 1-RI-90-400 provides the indication to the control room staff and would not isolate the listed failure. C. Incorrect: 1-AOI-31, Abnormal Release of Radioactive Material includes the guidance for an accidental gaseous radwaste release. This AOI contains a listing of symptoms which can be seen to basically list the various radiation monitors within the plant. The operator then utilizes his knowledge of the sample points of the radiation monitors to ascertain whether the radiation problem is associated with the Auxiliary, Reactor, Turbine or Service buildings. The procedure also contains the expected automatic isolations of the radiation monitoring system. It is Incorrect that the radiation monitor for the Waste Gas release header, 0-RM-90-118 would detect a failure of a decay tank relief valve because as seen on print 1-47W830-4, the relief valve header for the tanks ties into the discharge to the plant vent DOWNSTREAM of RE-90-118. Print 1-47W866-1 shows that the discharge to the plant vent is the shield bldg exhaust vent (shield bldg vent stack); this print directs the observer to next look at print 1-47W610-90-5 which shows that 1-XIC-90-400 processes the sample effluent from the shield bldg vent stack and displays the radiation levels on 1-RI-90-400. Therefore, 1-RI-90-400 will alert the main control room staff to the failed waste decay tank relief valve and NOT 0-RM-90-118. As seen in ARI-184-A, WGDT REL LINE 0-RM-90-118 RAD HI, 0-FCV-77-119 will automatically close on a high radiation signal isolating the gas decay tank release. However, in this case, 0-RM-90-118 never produces a high radiation signal and as such 0-FCV-77-119 never receives an automatic isolation signal. It is plausible to believe that this distractor is true because 0-RM-90-118 is the radiation monitor which normally inspects the WGDT release path and it would isolate the release path (for a normal tank release) upon a high radiation condition. D. Incorrect: It is not correct that 0-RM-90-118 would detect the listed failure. Also it is correct that the release would not be isolated. It is plausible to believe that 0-RM-90-118 would not perform an automatic isolation as the majority of the radiation monitors at the plant perform no automatic isolation. Tier: 1 Group: 2

K/A: 060 Accidental Gaseous Radwaste Release 2.4 Emergency Procedures / Plan 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Importance Rating: 4.5 4.7

10 CFR Part 55: (CFR: 41.10 / 43.2 / 45.6)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to recognize which radiation monitor in the main control room would provide the abnormal indication for a system parameter which would provide for the diagnostic entry level condition for 1-AOI-31. Technical

Reference:

1-AOI-31, Abnormal Release of Radioactive Material ARI-180-187, Common Radiation Detectors Picture of 1-RI-90-400 Picture of 0-RM-90-118

Print 1-47W610-90-5

Print 1-47W866-1

Print 1-47W830-4 Proposed references to be provided: None Learning Objective: 3-OT-SYS090A, Radiation Monitoring 3. EXPLAIN the physical connections and/or cause-effect relationships between the Radiation Monitoring System and the following systems: j. Shield Bldg Vent

k. Aux Bldg Vent Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: Question Number: 23 S-06-0 1'0 FT-9 0-"?12 llLDG vt N r ST ACK <1-411610-9 0--', COOKD E"-7 3 fi .; r -------, << I NN I t---n I I I I cID m N . C> 'I' " g Ill 9 a e : ii; 1' 2 I! 8 ol ! .. e mm NN tr . u :;:g I! ... 0 ! i; MAIN STEAM LINE GENERATOR 1 DISCHAIRCE RADIATION MONITORING UNIT ' <W $ e i >>LMUM>> 12111-176( .. I lIEllJ I IHSU @M-JO I RA t--------9Q-42J I &-!I ----I PX 90-42 JB 1 : :----------- 9 0-UJ : I L**O*: RE I r:.;\, I RM so-42JA ___ ._ __ MAIN STEAM LINE GEHERATOR J DISCHARGE RADIATION MONITOR!NC utur i @&W/(!@CAIE) 2 4 1 :2.-..-17 6( r--------1, I [N.Sf ft Moli L P'IJNCTICN LOC I "°'" -------@: ':o-m c I -@h0-'1 10-422s:---1 I r---------- P IC I I I WO-<t:Z:Z I l L-4-03 I M-30 llE I 1'M s o-4 22A ___ ._ __ NOTE: Iii MAIN STEAM GENERATOR 2 D ISCMARGE RADIA TI ON MONITORING UN IT 1 WUf!rl4?)1Mt&fl 1 t A-\76E: r--------r::Mi[] I ill3!) I INSTR Mi!Lf'.UNCTION -------@:::.mc Rt I @-" a o-*2*i ___ 1 : r----------itx I I I 90-42+ t I -J O RE r:;t\' .-:M NOTE Ii MAIN STEAM LINE G E NERATOR + DISCHARGE RADIATION MO N!TOR!NC IJNtT Hjfi)tf\fifi@QM 3 4 5 5 6 r---------- 1 I I I <@-: (' : ! I I L----1------ 1 I i I I I I 0 §fil2 -N -I I i I I I I 7 8 9 I -.. -' I ' I I I I I I r-------------------1 I I I I 1g 'i : Q. < : i j © t I , 1 1 / .,-* u : ---r----1 1 ---------j I : I I 1\ I I I ,.. _____ , I : l I I I I I I\ 1 1 j \ ! ! L------------, . ! I : I I I I I I I I I I ' I < I : I :: z I *' E ti* _, I I ,.-----------' I I ! .. I I I I I L ________ J ______ J I < : : : : t---- \_ __ J : : : I I I cUrx l I 1 ! I I I I I I i I I -I I I : I I ;;; ;;; I I " .!. MT R ... 90-4'04 10_404 I P MP-*O-*o* 1 (N<iP-1) ... !I! < .. " Ill ::: u ;t ,.. "' u ' ... z ._ "' "' ;;; I I NO T ES: 1. f'OR C EN£RA L NOT ES A.N D REF'EREN.CES S E E O RJil'ING 1-oj:7W6 1 0-9 0-1. 2-. A l l C OMP'ONU ITS HA V E A UNIT P R EF I X OF " 1' F'O R U N IT I AN D " 2" FOR U NIT 1 IJN L E SS OTH E: R WJ S E N O TE'D. l. SHICLC B UILDIN G VE'Nf F LO W E LE:MCNT S A ND F LOW T RAN.S M I TTE RS AU WrTH TH E' BASE COMPO ME:l'IT AD CJRES SES Otil Y fl-9 0-"'°(l oli , r 1 4. DES IC'NATIONS E"MCLOS C D I N BllACKHS (r J) .\RE M A NUF A CT U R E R S COMPONENf IDENTIFIERS A.NO AA E I-NCLUC E D f OM ONLY. ,, IJN lT 1 A," D CCM P U T E:R PDlNT $YMl3Ql$ AR[ $i-IQ W N. i $YMiiOL S A RE AS FO L LO*S= U NIT I SYM 90 L S: U NIT 2 SYMBOLS: COM P. P T. ft910U CO W i'". l'T. H 1 01A lO-a86 COhlf'. PT. r27 0 IA C'.CO.IP. P T. F 90 1$A 16-90) A N N 12 0-191( (Ml.JK 11-4 7) AN N SC-1 02[ H.WX 3-6.j.I l A NN !C-10 1 1:'. i: MUX )-84} } 6. LOW RA NCE: l-R E-9 0-+2 1A , -.. 21A , -+2J A

  • tA A FI E PROC::R oii hl.MED OU T. THIS OO NT R.OL OA 4WI N C S U P'E:RSLOES UN ir 1 oliS*CONSTR UC TCD ORA IINC '4 716 1 0-'J0-5 REVI S IO N 1-1. 4 1 526 3 0 C J O R EVISE D PER DCA '2630-144-Z.

RE V CHA N GE REF P RE PARER C H£ C K ER A PPROVED DAT E SCAL£' N T S £XC£P T AS NO T E D P ROJ ECT F' A CIL I TY POW E RHO U SE U NITS 1 & 2 T ITLE ELECTRICAL CONTROL DIAGRAM RADIATION MONITORING SYSTEM W ATTS BA R N U C LEA R PL A NT TENNESSEE VA L L EY a L lil e'.-9 0*4 0+ l CO ND ENSATE V ACUU M VENTPOsrACCi'CENT MOHITORI NC UN!T 1 rM!tIDUPtNRAA

  • D E SIGN !N!T!A L IS SUE ENC I NE ER INC APP ROV A L 5 6 7 8 9 CAD OlfGS: 1-<H W IU0-90-1 THFI U 4 CSSt.IF;O BY: CH EC K LR C..W. JCHNSON J.IJ. MOYU:S OE S I CNCR R EV'[ EWER t-1.L. Cl-IAPllU.N R.L. F'ORESfER RO I SS U E P ER W BEP 5. 1 7 .._RIMS 82 6 ' 90 0928 379 OCN P*OJ398*B OCN P-03 403-A Cl.P. PE R"-E'.R T. ONA L J C .C. L F O" MC.Ill WBN Unit 1 Abnormal Release of Radioactive Material 1-AOI-31 Rev. 0001 Page 3 of 27 1.0 PURPOSE This instruction provides a response to valid annunciations or indications of an abnormal release of radioactive material.

2.0 SYMPTOMS 2.1 Alarms NOTE SHLD BLDG VENT RE-400 RAD HI [268-B] in alarm should be accompanied by higher indicated radiation levels or alarm on additional monitor(s) which indicates the source of the release. A. Auxiliary Building: 1. 1-RR-90-1 AREA RAH HI [174-B] 2. ERCW DISCH HDR A 0-R M-133/140 LIQ RAD HI [180-B] 3. WDS RELEASE LINE 0-R M-122 LIQ RAD HI [181-A]

4. ERCW DISCH HDR B 0-R M-134/141 LIQ RAD HI [181-B] 5. AB VENT 0-RM-101 RAD HI [183-C]
6. WGDT REL LINE 0-RM-118 RAD HI [184-A] 7. SFP 0-RM-102/103 RAD HI [184-B]
8. 0-RR-90-12 PARTICULATE RAD HI [185-B] B. Reactor Building: 1. UPR CNTMT AIR 1-RM-112 RAD HI [173-A]
2. LWR CNTMT AIR 1-R M-106 RAD HI [173-B]
3. CNTMT PURGE EXH 1-R M-130/131 RAD HI [174-A]

t-0£9MLt-2 736 -... \ ... ,. ... -SEE DETAIL C-4-I _,_. : I I I I \ SEE DETAIL D+)-\ SirMON I BREAKER I I I I I I TO DRAIN SIPHON BREAKER LOI POINT TRAP 41+A TO DRAIN :..j HSA TANK A DES PRESS. 150 PSIG o ;,I 180'F (TYP) 0-TANK-77-:SA FOR ALL TANKS G HO _,_._ '" TANK B 0-TANK-77-:SB 7460 0-ISl/-04-3-0233 2 3 _,_._ _,_._ WASTE GAS DISCHARGE 2" :-:-) H8C ;,I 741C 751C :-) _,._ TANK C 0-TANK-77-:SC 7HC H6C _,_ .. _ 3 4 5 _, .. _ ,. :s;+* 3/4' n6A1 738A 6 :s;+* WASTE GAS INLET CONN 7388 7 8 RECYCLE 2' WASTE GAS SUPPLY 756A2 J -,.-. -,-.. ___ ;._1 _______ ,_*.: _______ o_,s_u_N_Lo_AD_E_R_*_R_EL_1_EF_v_L_v_o_1s_cH _________ 1 WASTE GAS COMPRESSOR PACKAGE A CAP +O SCFlll 0 100 PSIG 737A SEE DET A4 FOR COMPR PACK DETAILS OES. PRESS l:;iO PSIG 0 180" F :-j :-:-) 7480 7510 :-) _ ,._ TANK D O-TANK-77-30 _ ,_ .. _ 7440 H60 4 $ 1*--1*-:-) 748E 741E ;,) 751E ;,) :-) _,. _ TANK E O-TANK-77-JE _ ,_ .. _ 7HE H6E GAS DECAY TANKS CAP EA TANK -600 FT 3 5 ---751F 7578 WASTE GAS COMPRESSOR PACKAGE B CAP 40 SCFM 0 100 PSIG 7378 SEE OET A+ FOR COMPR PACK DETA!LS DES. PRESS l:;iO PSIG 0 180' F ,. _ ,._ TANK F TANK G TANK H 0-TANK-77-JF O-TANK-77-:SG D-TANK-77-JH _,_ .. _ H6f H4G H6G H4H _ ,._. NOTES CONTINUED:

7. , ALL F S FOR 0 OF 4 L BE VI REMENTS ICATIO . 1' 8. SOME CIDS ON THIS ORAWING HAVE BEEN CHANGED AND MAY DIFFER FROM THE CIDS SHOWN ON OTHER DOCUMENTS FOR THE SAME COMPONENT.

THE ALTERNATE IO SCREEN {AII) IN EMS CAN BE ACCESSED AS NECESSARY TO DETERMINE IF PREVIOUS CIDS EXISTED FOR A SPECIFIC COMPONENT. '3. COt.lPRESSORS MAY BE REPLACED WITH A NON-ASME ITEM IN AN ASME SECTION III SYSTEM IN ACCORDANCE WITH NRC GENERIC LETTER as-as, PROVIDED THE COMPRESSORS ARE THE SAME FIT, FORM, ,1,ND FUNCTION. 6 7 H6H 8 9 10 11 12 ,. 416A "Z:s'--_,..,,,,.;._..;_..;,;_..,,;::;'l;::;8;::;" '-; _,_._ :...f H.SJ I I I I I I I I I I I I I I I ! I I LA __ ___ I""' <8J 7608* DETAIL A4 WASTE GAS COMPRESSOR PACKAGE A _,_._ THIS CONFIGURATION CONTROL DRAWING SUPERSEDES UNIT 1 AS-CONSTRUCTED DRAWING 47W8JD-4 REVISION S. *WASTE GAS COMPRESSOR PACKAGE B AS NOTED _,. _ TANK J $ 0-TANK-77-JJ _,_._ 744J 746J :-j _, .. _ LUS AND AUXILIARY IRED TO MAINTAIN ABSCE ATION FIEO ON DCA,5:5387-670.")

11. UNIT INTERFACE SEPARATION BOUNDARIES ARE CONTROLLED BY INSTRUCTION TI-12.08, CONTROL or UNIT INTERFACES.

ISSUED BY: CAD MAINTAINED NOTES: 21 53387 RWE JEW OW--.t>/.S REVISED PER DCA 53387-669-0. REV CHANGE REF PREPARER CHECKER SCALE, NTS EXCEPT AS NOTED PROJECT FACILITY POWERHOUSE AUX BLDG UNITS 1 &: 2 TrTLE MECHANICAL FLOW DIAGRAM WASTE DISPOSAL SYSTEM WATTS BAR NUCLEAR PLANT Q TENNESSEE VALLEY AUTHORITY DESIGN DRAFTER CHECKER DESIGNER REVIEWER J.T.COLEMAN E.E.SANOERS

INITIAL ISSUE RO ISSUE PER WBEP 5.17 a: RIMS 826 '90 0828 377 ENGINEERING APPROVAL 1 P.S.KIELB 2 P. W. BURNETTE 3 L.W.BOYO VA£ 'It l.AST H/MllR IJStD 1-JO-US 2 3 4 5 6 NO T ES CQNf t 32' VM.vE Jl. DAMll'ER FAJL CLO S E"0 1 A!JI SUPPLY ISOL ATED TO OP[llATION. V ALV[ l'O FtEMA.IJI U CLCGEO POSIHCN TO SE R'/£ AS A ut'II1' 1/UMIT Z IN T£R f AC[ POINT, F,f,,N-30-Ja LOWt R C OMPARTMt: Nl' P URGE A I R SU PP LY H. YDROGE N COLL ECTOR LOWE:FC CCMPAIUM(Nf PURCE A lft EXHllUST CF M EACH I S i rN-WC I N ACCORD A NCE WITH SMACN" T CONoST R UCTlON S T" NOAROS r'OR M!tHUM fJ SEt W A TT S B AA NUCl t AR !IEPOFilT -009 -010. N \\ ... 'll JN G AS TVA Cl.ASS !, C, C , QR H CONNECT£!;) TO TH( eul BE: l(.\K l ESi l'Ht: M OST $T "TS SPEC IFl tCl ION +.12.1 O F rtl TEft H.OU St N L BE: L EAK-T ESTED N!SO'a-1$7&. ALL lE.O:.AC£ RA!ES f°OR HOUS!NCOS SMAlt i( 7.l i.1 Cf'M , N C" +2:HFI). 3 UNIT t S H IEL O BLOC A N N U LUS COfl T AI NM EHT VESSEL 7 SEE CRAWJNC 1-+7 t'H6-1 A f'OR UNIT 2 TIU1NSf'EllltE D EOU if"MEIH z ; 0 R E FUE LING CANM. 1-ClR-.30-1112 -J 0-5 02. ....... rr---rr-.--..,.,- .. ;-1-1 0-*0 1 * *1 ,.,--.. "l O t. .-------1-i t AN-30-? ... C CU-l0-7 -'V-lli:!.O 'CI Q C C U-J 0-7 ..,I\,-_,_ -UNIT C-A lroD Olllllli'[ COOLfNG UNIT (T't'P lCAL > r AN-JD-!12/2 ,.i'I N-J O*U/\ 3 1.25 0 -r-J4-,!IO O CBYPA5$) + RE l.I C:F CAD M AI NTAINE D DRAWING **£ REV!SEO r o !:irtO W r-M QO WBN-t-Z0 1 4-0 J<l-0 04-0 1-0. REV CHA NC E REF PREPARER CHECK ER APPRO V ED C A TE SCALE' N TS ncEPT ;.s PR OJECT f'A CiLI TY R EACTOR B UILD ING UN ITS 1 a: 2 TI T L E FLOW D I AGRAM HEATING AND VENTILATION AIR FLOW WA T TS BAR N U CLEAR PLA NT T E N NESSEE VAL L EY AUTHORI T Y C O N T RO L Q WBN Unit 1 Common Radiation Detectors ARI-180-187 Rev. 0034 Page 25 of 46 A. High activity release from gas decay tanks (GDTs) B. Loss of power to control relay (WDGS) C. Loss of ratemeter power NOTE 0-RM-90-118 has associated ICS computer point R0016A. [1] ENSURE 0-FCV-77-119 CLOSED. [2] NOTIFY Chemistry to perform CM-9.09 "Effluent Radiation Monitor Alarm Guidelines" and Radiation Protection to investigate alarm: [2.1] IF 0-FCV-77-119 closure is due to high activity, THEN INITIATE new release package, OR EVALUATE postponing release to allow additional decay time. [2.2] IF 0-FCV-77-119 closure is due to malfunctioning 0-RM-90-118, THEN INITIATE corrective maintenance. [3] EVALUATE gas decay tank lineup. [4] REFER TO AOI-31, Abnormal Release Of Radioactive Material . [5] REFER TO Tech Specs/ODCM. 1-45W600-57-27 1-45W600-77-1

1-47W601-90-34

1-47W610-90-1 AOI-31 CM-9.09 ODCM Source Setpoint 0-RM-90-118A determined by Chemistry 184-A WGDT REL LINE 0-RM-118 RAD HI (Page 1 of 1) Probable Cause: Corrective Action:

References:

24. Given the following conditions: - Radiography activities are being conducted SIMULTANEOUSLY in BOTH the Waste Package Area AND the Decontamination Room Area. - The Radiographer improperly sh ields the source and causes a VALID HIGH radiation condition within the Decontamination Room Area. - The condition causes an Annunciator to alarm in the MCR.

Which ONE of the following describes the type of radiati on monitor that detected the condition AND the alarm window in the MCR? A(n) ____(1)____ radiation monitor detected the high radiation condition. The Annunciator window ____(2)____ identify the SPECIFIC room in which the high radiation condition occurred. A. (1) area (2) does B. (1) area (2) does NOT C. (1) process (2) does D. (1) process

(2) does NOT CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: As seen in the design criter ia, WBN-DCD-40-24, Radiation Monitoring (Unit 1 / Unit 2), the radiation mo nitoring system includes both process and area radioactivity monitors. Sect ion 3.1.5 of this design criteria documents the Area radiation monito rs and lists that 0-RE-90-003 monitors the Waste Packaging Ar ea and that 0-RE-90-004 monitors the Decontamination Room Area. Therefore, it is correct that an area monitor would detect the high radiation condition which is listed in the stem of the question. However, as seen in ARI-174-B, 1-RR-90-1 AREA RAD HI, the annunciator window which came into alarm does not specify which room had the high radiation condition. Step [1] of the ARI lists: CHECK 1-RR-90-1 and associated RMs [0-M-12] to determine which area radiation monitor is alarming. It is plausible to believe that the annunciator windo w would stipulate which area had the high radiation levels because th is specificity in annunciation is seen with the process radiation m onitors. These monitors have individual alarm windows. Additionally , it is plausible to believe that this is true because some of t he area radiation monitors do have individual alarm windows (e.g. 0-RM-90-135 and 0-RM-90-102/103). B. Correct: Area radiation monitors would detect the condition and the alarm window would not identify which area had the high radiation condition. C. Incorrect: It is not true that a proc ess radiation monitor would detect the condition. It is plausible to belie ve this as a high radiation condition can cause a process radiation monito r to alarm. An excellent example of this is that a refueling water purification filter change can: 1. raise local area radiation levels significantly, 2. ca use 0-RM-90-9, Waste Cond Tnk Area to alarm and 3, caus e the process monitor 0-RM 122, WDS Release Line to alarm. This later alarm occurs because the process monitor is quite close to the filter cubicle. The locations given in the question do not contain any resident process monitors and the

alarm of a process monitor is not po ssible. As described it is not correct but plausible to believe that the annunciator received would indicate the location of the high radiation. D. Incorrect: Again it is not true but plaus ible that a process monitor would detect the condition. Also, it is true that the alarm window received would not identify the location of the high radiation condition.

Question Number: 24 Tier: 1 Group: 2

K/A: 061 Area Radiation Monitoring (ARM) System Alarms AK2. Knowledge of the interrelations between the Area Radiation Monitoring (ARM) System Alarms and the following: AK2.01 Detectors at each ARM system location

Importance Rating: 2.5 2.6

10 CFR Part 55: (CFR 41.7 / 45.7)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to understand the interrelation of the Annunciation and Alarm system and two detectors with the ARM system. Technical

Reference:

WBN-DCD-40-24, Radiation Monitoring ARI-174-B, 1-RR 1 AREA RAD HI Proposed references to

be provided: None Learning Objective: 3-OT-SYS090A, Radiation Monitoring 3. IDENTIFY the following controls and indications; include main control room, auxiliary control room, and local panels as applicable:

a. Liquid Process Rad Monitors (RMs)
b. Liquid Effluent RM's c. Gaseous Process RM's d. Gaseous Effluent RMs
e. Area RMs (ARM)
f. Post Accident Area RMs (PAARM)

Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Design Criteria Document RADIATION MONITORING (UNIT 1 / UNIT 2) WBN-DCD-40-24 Rev. 0026 Page 26 of 89 1.0 SCOPE This document establishes the design requirements of the Radiation Monitoring System (RMS) for the operation of Watts Bar Units 1 and 2 [Source Note 7]. The RMS provides continuous monitoring of radiation levels, provides the results of the monitoring to the operator and in some instances produces signals to initiate automatic control actions. In addition, the RMS includes equipment for off-line tritium, particulate and iodine sample collection with no real-time detection. The RMS consists of process, airborne, area, and effluent radioactivity monitors. The outdoor radiological monitoring instrumentation, laboratory radiation measurement equipment, criticality monitoring equipment, and portable survey instrumentation is outside the scope of this criteria. If a discrepancy exists between this design criteria and any other Nuclear Engineering (NE) design criteria, the appropriate Engineering Manager should be notified by a memorandum. If a discrepancy exists between this design criteria and any other document where the other document is not a NE design criteria, this design criteria shall govern. 2.0 DEFINITIONS Accuracy - The degree of agreement with the true value of the quantity being measured. Accessible Area - An area which is designed to be entered without special radiological controls (i.e., is not a high radiation area or airborne radioactivity area), dismantling of installed equipment or the introduction of support equipment such as staging, ladders, etc. Anticipated Operational Occurrences - Those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to the loss of power to all reactor coolant recirculation pumps, a trip of the turbine generator set, isolation of the main condenser and a loss of all off-site power. Categories 1, 2, and 3 - References to Categories 1, 2, and 3 are as stated in RG 1.97 (Reference 8.1.11). Channel - A channel includes all of the instrumentation necessary to provide the monitoring function. A channel includes associated sample lines, skid assemblies, and local readout and alarm devices. A channel also includes equipment in the control room for recording, readout, and alarm and includes the electrical cable that connects monitor components. Check Source - A test source supplied as an integral part of the monitor channel for use in determining if the channel is functional. This may be a radioactive source, or when a check of the channel exclusive of the detector scintillator is appropriate (i.e., where increased background cannot be tolerated), a pulsed light source or other similar device appropriate for the type of detector. Class 1E - The safety-related classification of the electrical equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling and containment and reactor heat removal, or otherwise are essential in preventing significant release of radioactive material to the environment. Grab Sample - A volume removed from a process or effluent for laboratory analysis.

Gross beta or gamma radioactivity - The combined radioactivity from the mix of nuclides without distinction as to energy levels of the emissions. WBN Design Criteria Document RADIATION MONITORING (UNIT 1 / UNIT 2) WBN-DCD-40-24 Rev. 0026 Page 41 of 89 3.1.2 Off-Line Gas Monitors (continued) The measured noble gas activity level shall be transmitted to the Plant Computer System for display in the TSC. 3.1.3 Continuous Air Monitors (Airborne Particulate Monitors) The following continuous air (particulate) monitors shall be provided for plant areas where airborne particulate radioactivity may potentially exist and where normal occupancy is required either on a continuous basis or on an infrequent but routine basis:

  • Spent Fuel Pool Monitor
  • Shipping Bay Monitor
  • Holdup Valve Gallery Monitor
  • SI Pump Area Monitor
  • Main Control Room Monitor
  • Waste Packaging Room Monitor
  • Sample Rooms Monitors
  • Monitoring of the above listed plant areas are accomplished using portable continuous air monitors supplied and administered by Site Radcon. The requirements of this Design Criteria do not apply to these monitors. 3.1.4 Off-Line Liquid Monitors The monitors that come under this category are: Steam Generator Blowdown Liquid Monitor 1,2-RE-90-120 & -121 Waste Disposal System Liquid Monitor 0-RE-90-122 Component Cooling System Monitor 0,1,2-RE-90-123 Condensate Demineralizer Regenerant Waste Discharge Monitor 0-RE-90-225 Essential Raw Cooling Water Effluent Monitors 0-RE-90-133-, 134,-140 & -141 These monitors provide real-time monitoring of the gross gamma radioactivity in liquid samples. Selected off-line liquid monitor channels shall initiate an automatic control function to terminate releases either to the unrestricted area or to the plant environs as described in the following subsections. A. General Requirements The monitors shall have pumps or the process connection points shall provide sufficient differential pressure to ensure that an adequate flow rate from the process streams can be predictably delivered to the sample chamber.

WBN Design Criteria Document RADIATION MONITORING (UNIT 1 / UNIT 2) WBN-DCD-40-24 Rev. 0026 Page 46 of 89 3.1.4 Off-Line Liquid Monitors (continued) 3.1.5 Area Radiation Monitors The following monitors shall continuously monitor ambient radiation levels in the plant buildings to assure that work areas designed for short-term accessibility or normal occupancy have exposure rates which do not exceed the prescribed radiation zone limits, and to provide early warning of abnormal process system operations. Spent Fuel Pool Area Monitors 1,2-RE-90-001 Personnel Lock Monitors 1,2-RE-90-002 Waste Packaging Area Monitor 0-RE-90-003 Decontamination Room Monitor 0-RE-90-004 Spent Fuel Pool Pumps Area Monitor 0-RE-90-005 Component Cooling Heat Exchanger Area Monitors 1,2-RE-90-006 Sample Rooms Monitors 1,2-RE-90-007 Auxiliary Feedwater Pump Area Monitors 1,2-RE-90-008 Waste Evaporator Condensate Tank Area Monitor 0-RE-90-009 CVCS Board Area Monitors 1,2-RE-90-010 Containment Spray RHR Pump Area Monitor 0-RE-90-011 RB Upper Compartment Refueling Floor Area Monitors 1,2-RE-90-059 RB Upper Compartment Area Monitors 1,2-RE-90-060 RB Lower Compartment Instrument Room Monitors 1,2-RE-90-061 Main Control Room Monitor 0-RE-90-135 Condensate Demineralizer Area Monitor 0-RE-90-230 Condensate Demineralizer Area Monitor 0-RE-90-231 RB Upper Compartment Post Accident Monitors 1,2-RE-90-271 & -272 RB Lower Compartment Post Accident Monitors 1,2-RE-90-273 & -274 A. General Requirements The monitors shall provide real-time measurement of gross gamma ambient radiation exposure rates in plant areas. These instruments shall be installed in locations which satisfy the guidelines in RG 8.8 (Reference 8.1.6) and ANSI/ANS-HPSSC-6.8.1-1981 (Reference 8.2.3). The monitors shall have detectors with sufficient range to encompass the minimum and maximum exposure rate expected during normal and anticipated operational occurrences (considering the ambient background radiation conditions at the detectors' location). The monitors should have a minimum of five decades of range as described in ANSI/ANS-HPSSC-6.8.1-1981 (Reference 8.2.3). Certain locations may require a range in excess of five decades. Where multiple detectors are required to cover the anticipated range, the range of the detectors shall overlap. Overlap should be at least one decade. Extended ranges are discussed in the individual monitor sections below. WBN Unit 1 U-1 Radiation Det ectors 1-ARI-173-179 Rev. 0000 Page 3 of 51 UNIT 1 RADIATION DETECTORS 0-XA-55-12A 173 174 175 176 177 178 179 A UPR CNTMT AIR 1-RM-112 RAD HI CNTMT PURGE EXH 1-RM-130/131 RAD HI SG BLDN 1-RM-120/121 LIQ RAD HI CCS HX A 1-RM-123 LIQ RAD HI A B LWR CNTMT AIR 1-RM-106 RAD HI 1-RR-90-1 AREA RAD HI VAC PMP EXH 1-RM-119 RAD HI B C C D UPR CNTMT AIR 1-RM-112 INSTR MALF CNTMT PURGE EXH 1-RM-130/131 INSTR MALF SG BLDN 1-RM-120/121 INSTR MALF CCS HX A OUTLET1-RM-123 INSTR MALF D E LWR CNTMT AIR 1-RM-106 INSTR MALF 1-RR-90-1 AREA MONITORS INSTR MALF VAC PMP EXH 1-RM-119 INSTR MALF PAS MON PNL M30 / M31 INSTR MALF

E F F 173 174 175 176 177 178 179

WBN Unit 1 U-1 Radiation Det ectors 1-ARI-173-179 Rev. 0000 Page 13 of 51 A. Radiation rise in affected area B. Proximity of 0-RM-90-3 to Radwaste Mobile Demin package C. Loss of power to rate meter [1] CHECK 1-RR-90-1 and associated RMs [0-M-12] to determine which area radiation monitor is alarming. [2] CHECK Unit 2 area monitors and Aux Bldg vent 0-RM-90-101. [3] IF loss of power indicated (green light out and 174-E LIT), THEN GO TO of 174-E. [4] NOTIFY Radiation Protection to investigate alarm. [5] REFER TO 1-AOI-31, Abnormal Release Of Radioactive Material . [6] IF during refueling, THEN REFER TO AOI-29, Dropped Or Damaged Fuel Or Refueling Cavity Seal Failure. 47W610-90-2 45N1651-9, -11 1-AOI-31 AOI-29 Source Description Setpoint 1-RM-90-1 Spent Fuel Pit determined by Radiation Protection 1-RM-90-2 Personnel Access Outlet 0-RM-90-3 Waste Package Area 0-RM-90-4 Decontamination Room Area 0-RM-90-5 Spent Fuel Pit Pumps 1-RM-90-6 CCS Hx Area 1-RM-90-7 Sample Room 1-RM-90-8 AFWP Area 0-RM-90-9 Waste Cond Tnk Area 1-RM-90-10 CVCS Bd Area 0-RM-90-11 CS and RHR Pump Area 1-RM-90-59 Upper Cntmt - by equip hatch 1-RM-90-60 Upper Cntmt - by air lock 1-RM-90-61 Incore Inst Room 0-RM-90-230 Cond Demin 0-RM-90-231 Cond Demin 174-B 1-RR-90-1 AREA RAD HI (Page 1 of 1) Probable Cause: Corrective Action:

References:

25. Which ONE of the following describes the use of Self-C ontained Breathing Apparatuses (SCBAs) in accordance with the Wa tts Bar Fire Protection Report?

The operating life of the SCBAs used by control room personnel is a MINIMUM of ____(1)____.

As this time limit is approached, the wearer of the SCBA will ____(2)____. A. (1) 30 minutes (2) have the SCBA's cylinder replaced B. (1) 30 minutes (2) utilize a manifold system supplied from a storage reservoir C. (1) 2 hours (2) have the SCBA's cylinder replaced D. (1) 2 hours (2) utilize a manifold system supplied from a storage reservoir

CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Incorrect: It is correct that the operating life of the SCBA is 30 minutes. It is correct that the wearer of the SCBA will have his cylinder be replaced by a helper (another member of the control room staff) as it nears exhaustion. B. Incorrect: As mentioned it is correct that the operating life of the SCBA is 30 minutes.

Additionally, it is not correct that a manifold system be utilized. It is plausible to believe this because of several reasons. Firstly, the service air system supplies portable breathing air stations (PBAS) via quick disconnect couplings. PBAS are available to supply air to a maximum of 6 personnel (this fact is observed in the system description WBN-SDD-N3-32-4002, Compressed Air System). Secondly, chemistry personnel having breathing air systems built into their Post Accident and Primary sample labs. As seen in system description, N3-43B-4001, Sampling and Water Quality System Post accident Sampling Facility, the breathing air systems is designed to supply >6cfm/person (based upon 3 people). It would be reasonable to believe that if chemistry personnel were provided with a breathing air system that the occupants of the main control room would be as well. C. Incorrect: It is not correct that the operating life of the SCBA is 2 hours. As seen on page VIII-34 of Part VIII of the Watts Bar Fire Protection report: The operating life of the self-contained units is a minimum of one-half hour. It is plausible to believe that the operating life is 2 hours because the MSA Air Hawks which are provided to the operations personnel in the MCR do have an air bottle option which is rated at 120 minutes. This is the bottle size which is used at the TVA Nickajack fire training facility (for use with the MSA Fire and Air Hawk SCBAs).

The wearer of the SCBA would have his cylinder replaced as it became exhausted D. Incorrect: It is not correct that the operating life is 2 hours; it is not correct that a manifold system would be utilized.

Question Number: 25 Tier: 1 Group: 2

K/A: 067: Plant fire on site AA2. Ability to determine and interpret the following as they apply to the Plant Fire on Site: AA2.10 Time limit of long-term-breathing air system for control room Importance Rating: 2.9 3.6

10 CFR Part 55: (CFR: 43.5 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to determine the time limit of the long-term-breathing air system for the control room by understanding the operating life of the SCBA (the breathing air system utilized by the main control room staff during a plant fire). Technical

Reference:

Part VIII of the Watts Bar Fire Protection report N3-43B-4001, Sampling and Water Quality System Post-accident Sampling Facility

WBN-SDD-N3-32-4002, Compressed Air System

Proposed references to

be provided: None Learning Objective: 3-OT-AOI3000, AOI-30.1 &30.2, Plant Fires 8. Given the indications of a plant fire, DETERMINE the following as they apply to 1-AOI-30, Plant Fires

f. Time limit of long-term-breathing air system for control room
g. Time limit for use of respirators Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: PART VIII - CONFORMANCE TO APPENDIX A TO BTP 9.5-1 GUIDELINES VIII - 34 Rev. 10 Appendix A Guidelines Plant Conformance Alternatives Remarks Self-contained breathing apparatus, using full face positive pressure masks, approved by NIOSH (National Institute for Occupational Safety and Health - approval formerly given by the U.S.

Bureau of Mines) should be

provided for fire brigade, damage control, and control room personnel. NIOSH-approved self-contained full-face

positive pressure breathing apparatus is available for the fire brigade, damage control, and control room personnel. The operating life of the self-contained units is a minimum of one-half hour. Control room personnel may be furnished breathing air by a manifold system piped from a storage reservoir, if practical. Service or operating life should be a minimum of one-half hour for the self-contained units. Not applicable. At least two extra air bottles should be located onsite for each self-contained breathing unit. In addition, an onsite-6-hour supply of

reserve air should be provided and arranged to permit quick and complete replenishment of

exhausted supply air bottles as they are returned. If compressors are used as a source of breathing air, only units approved for breathing air should be used. Special care must be taken to locate the compressor in areas free of dust and contaminants. At least (1) one hour air bottle is available for each self-contained breathing unit. An additional (16) one hour bottles are available onsite to provide a minimum 16-hour supply of reserve air for the self-contained breathing units provided for use by the fire brigade, damage control, and control room personnel.

WBN System Description Document COMPRESSED AIR SYST EM WBN-SDD-N3-32-4002 Rev. 0010 Page 31 of 71 3.1.1 General (continued) The C&SA compressors discharge into two r edundant headers provided with manual valves. These headers feed the two CAS receivers which in turn supply air through redundant headers to the CAS dryers. The dryers contain three trains of prefilters, dryers, and afterfilters. Each dryer train is sized to handle CAS requirements for one unit. Manual bypasses are provided around each filter assembly for abnormal operation. The air is then piped to pneumatic equipment throughout the plant. Air is supplied to the SAS receiver by a header from the CAS receivers. Service air is supplied through a valve which closes if CAS pressure drops below a minimum value of 80 psig to assure that CAS requirements take precedence over the SAS. SAS air is piped from the receiver to service outlets and miscellaneous equipment throughout the plant. Like CAS air, the SAS is delivered oil-free. SAS air is not required to be dried and filtered. A 4" permanent piping segment with a "Tee", an isolation valve and connecting flange is provided to the existing service air header for emergency or future compressed air usage. The tie-in is located inside the Turbine Building. The portion of the SAS downstream of valves 2-33-509 and 2-33-543 will be supplied oil free air from temporarily installed Construction Air Compressors through the tie flange located at 2-33-675. Valves 2-33-509 and 2-33-543 will have their positions controlled as shown on the corresponding flow diagrams to prevent interaction between the Construction Air Compressors and the permanently installed station air compressors. The SAS shall be the normal source of breathing air. SAS air connects to portable breathing air stations (PBAS) via quick-disconnect couplings. PBAS are available for supplying air to a maximum of 6 personnel. These stations provide breathing air for personnel in areas with potential airborne radioactivity hazards that are not immediately hazardous to health. PBAS provide a separate source of breathing air as required by Reg

Guide 8.15 (Ref 7.5.3). Note: Associated equipment for breathing air (i.e., Post Accident Sampling Facility and Self-Contained Breathing Apparatus) have no physical interface or otherwise with the SAS, therefore it is not discussed here. The ACAS is supplied by 2 motor-driven, nonlubricated, compressors. Each compressor is sized to supply the total safety-related air requirements in the event of an accident, flood, or loss of the CAS. ACAS flow demand is 35.14 scfm for Unit 1 operation (Ref 7.4.6). The ACAS is separated into 2 independent trains, each containing a compressor, receiver, dryer, and filter. Relief valves are placed on the ACAS compressors and air receivers (Ref 7.4.2). Electric power for the ACAS is provided from normal and emergency sources. The ACAS automatically isolates from the CAS on low CAS pressure. WBN System Description Document COMPRESSED AIR SYST EM WBN-SDD-N3-32-4002 Rev. 0010 Page 69 of 71 Attachment 4 (Page 1 of 2) ABBREVIATIONS AND ACRONYMS ABGTS - Auxiliary Building Gas Treatment System ac - Alternating current ACAS - Auxiliary Control air subsystem AFW - Auxiliary feedwater ANSI - American National Standards Institute ASME - American Society of Mechanical Engineers CAS - Control air subsystem CIV - Containment Isolation Valve C&SA - Control and Station Air CNTMT - Containment Compr - Compressor Cu - Copper dc - Direct current DP - Differential Pressure EGTS - Emergency Gas Treatment System ERCW - Essential Raw Cooling Water °F - Degrees Fahrenheit FCV - Flow Control valve FMEA - Failure Modes and Effect Analysis FSAR - Final Safety Analysis Report

ft 3 - Cubic Feet gpm - Gallons per minute hp - High pressure HVAC - Heating, Ventilation, and Air-Conditioning Hz - Hertz ICFM - Inlet Cubic Feet per Minute IR - Ingersoll-Rand lp - Low pressure ph - Phase MCRHZ - Main Control Room Habitable Zone NEMA - National Electrical Manufacturer's Association Ni - Nickel NPT - Nominal pipe threads PA - Pressure alarm PAS - Post Accident Sampling PCV - Pressure Control value PI - Pressure indicator PIC - Pressure indicating controller PBAS - Portable Breathing Air Station NPG System Description Document SAMPLING AND WATER QUALITY SYSTEM POSTACCIDENT SAMPLING FACILITY N3-43B-4001 Rev. 0004

Page 31 of 34 Appendix A (Page 2 of 3) Preoperational Test Criteria TABLE 1 COMPONENT REQUIREMENTS COMPONENT REQUIREMENTS DESIGN BASIS NOTES LIQUID SAMPLE PANEL L-567 RCS AND CONTAINMENT SUMP SAMPLE FLOW OF 1900 ML/MIN DURING PURGING & 200 +/- 20 ML/MIN DURING SAMPLING N3-43B-4001 P. 3.1.1 Table 9-2 80 x 62-827371 NOTE 1 CONTAINMENT AIR CONTAINMENT SAMPLE AIRFLOW P. 3.2.11 SAMPLING PANEL L-569 0.2 ACFM HEAT TRACING MAINTAINS SAMPLES 280 F Table 9-2 80 x 62-827371 WB-DC-40-39

p. 5.3 NOTE 1, 5 SAMPLE COOLERS COOL RCS SAMPLES 102 F WITH 95 F CCS COOLING WATER N3-43B-4001 P. 3.1.1 NOTE 3 ION CHROMATOGRAPH 1&2-XAN-43-284 (U2 ONLY)

FLOW = 15 +/- 2 ML/MIN, INCREASING 80 x 62-827371 DISSOLVED OXYGEN 1&2-XAN-43-321 FLOW = 200 +/- 20 ML/MIN, INCREASING N3-43B-4001

TABLE 9-2 GAS CHROMATOGRAPH 1&2-XAN-43-338 CARRIER GAS FLOW 60 ML/MIN PURGE GAS FLOW 350 ML/MIN 827371 NOTE 4 BREATHING AIR FLOW 6 CFM/PERSON (BASED ON 3 PEOPLE) WB-DC-40-39 P. 4.3 NOTE 2 NOTES: 1. SET POINTS CONTROLLED BY 1-47B601-43 SERIES. 2. SET POINTS CONTROLLED BY 1-47B601-31 SERIES. 3. MUST BE PERFORMED DURING HOT FUNCTIONAL TEST. 4. NOT REQUIRED FOR UNIT 1. 5. MUST VERIFY THAT THE AIR SAMPLES ORIGINATE FROM CONTAINMENT.

26. Given the following timeline:

00:00:00 Unit 1 is at 100% power. 00: 0 1:00 A Steam Line Break Occurs inside CNTMT. 00: 0 1:1 5 The 1B-B MDAFWP shears its shaft. 00: 0 1:45 CNTMT Pressure is 3.0 psig and RISING 00:05:00 ALL MSIV's hand-switches have RED lights LIT. 00:07:00 1A-A MDAFWP TRIPS due to a motor fault. 00: 1 0:00 1-FR-P.1, Pressurized Thermal Shock is entered. 00: 1 0:30 SG Narrow Range levels are: SG 1 SG 2 SG 3 SG 4 1% 4% 3% 3% 00: 1 1 :00 Step 3 of 1-FR-P.1 is completed Which ONE of the following describes the status of AFW? At 00: 1 1 :00 , the operating crew has a ligned AFW to Feed ________. A. ALL SGs at 410 gpm B. ALL SGs at MINIMUM detectable flow C. ONLY the SG supplying the TDAFWP at 410 gpm D. ONLY the SG supplying the TDAFWP at MINIMUM detectable flow

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: Incorrect: Because the narrow range levels in all S/Gs are less than 39%, the status trees will indicate a red path to 1-FR-H.1, Loss of Secondary Heat Sink if total FW flow is not greater than 410 gpm. Additionally, the loss of secondary heat sink status tree is a higher priority instruction than the pressurized thermal shock tree. Therefore, it is plausible to believe that one would feed S/Gs at a value to preclude generating a red path condition to 1-FR-H.1. B. Correct: Step 3 of 1-FR-P.1 states: IF all S/Gs F aulted, THEN CONTROL feed flow at minimum detectable flow to each S/G. The WOG Basis document for this step explains that the minimum feed flow of setpoint (S.04) should be maintained to each steam generator. The EOP Setpoints Verification Document, WBN-OSG-4-188 explains that the purpose of S.04 is to allow the SG components to remain in a wet condition thereby minimizing any therma l shock effects if feed flow is increased. C. Incorrect: The stem of the question indicates that both of the MDAFWPs have failed. Therefore, the only source of AFW is the TDAFWP. A caution prior to step 3 states: If the turbine-driven AFW pump is the only available source of feed flow, steam supply to the turbine-drive AFW pump must be maintained from at l east one S/G. Therefore, it is plausible that one may believe that the SG feeding the TDAFWP's

steam would be the only SG required to have inventory. Additionally, one may couple this belief with that mentioned previously regarding the prevention of a red path to 1-FR-H.1. Therefore, because of these two items, this distractor is plausible. D. Incorrect: Again, it is plausible to believe that feeding only the SG supplying the TDAFWP's steam is correct. It is correct to feed at a minimum detectable flow rate.

Question Number: 26 Tier: 1 Group: 2

K/A: E08 Pressurized Thermal Shock EK1. Knowledge of the operational implications of the following concepts as they apply to the (Pressurized Thermal Shock) EK1.3 Annunciators and conditions indicating signals, and remedial actions associated with the (Pressurized Thermal Shock).

Importance Rating: 3.5 4.0 10 CFR Part 55: (CFR: 41.8 / 41.10, 45.3)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to rationalize the condition indicating signals (the HI-HI containment pressure which should have closed the MSIVs and the subsequent failure of the MSIVs to close) impact to the performance of 1-FR-P.1. The applicant must subsequently determine the remedial action required to combat a fault of all S/Gs during a Pressurized Thermal Shock

event.. Technical

Reference:

1-FR-H.1, Loss of Secondary Heat Sink 1-FR-P.1, Pressurized Thermal Shock WOG basis document for 1-FR-P.1 Setpoint document for the EOPs, WBN-OSG4-188

Proposed references to be provided: None Learning Objective: 3-OT-FRP0001, Function Restoration Guidelines 1-FR-P.1 & .2 3. Explain why minimum detectable flow is maintained to each S/G of all the S/Gs are faulted. Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Loss of Secondary Heat Sink 1-FR-H.1 Rev. 0003 Step Action/Expected Response Response Not Obtained Page 3 of 37 3.0 OPERATOR ACTIONS CAUTION

  • If total feed flow CAPABILITY of 410 gpm is available, this Instruction should NOT be performed.
  • If an Intact S/G is avail able, feed flow should NOT be reestablished to any faulted S/G.
1. CHECK if secondary heat sink is required:
a. RCS pressure greater than any Intact S/G pressure.
a. RETURN TO Instruction in effect. b. RCS temperature greater than 375

°F [360°F ADV]. b. PLACE RHR System in service while continuing in this instruction.

  • REFER TO 1-SOI-74.01, Residual Heat Removal System. WHEN adequate RHR shutdown cooling established, THEN RETURN TO Instruction in effect.
2. ENSURE at least one charging pump RUNNING.

STOP all RCPs AND ** GO TO Cautions prior to Step 18 to initiate RCS bleed and feed. WBN Unit 1 Pressurized Thermal Shock 1-FR-P.1 Rev. 0000 Page 4 of 25 Step Action/Expected Response Response Not Obtained CAUTION

  • If the turbine-driven AFW pump is the only available source of feed flow, steam supply to the turbine-driven AFW Pump must be maintained from at least one S/G.
  • If a faulted S/G is necessary for RCS temperature control feed flow to that generator should be maintained. NOTE
  • A Faulted S/G is any S/G t hat is depressurizing in an uncontrolled manner or is completely depressurized.
  • Minimum detectable flow is assured by observing flow indicator response to valve movement.
3. CHECK S/G pressures:
  • All S/G pressures controlled or rising.
  • All S/G pressures greater than 140 psig.

IF S/G with low or dropping press has NOT been isolated, THEN REFER TO 1-E-2, Faulted Steam Generator Isolation, while continuing

with this Instruction.

ENSURE TD AFW pump being supplied from intact S/G.

IF all S/Gs Faulted, THEN CONTROL feed flow at minimum detectable flow to each S/G. Step continued on next page WBN Unit 1 Pressurized Thermal Shock 1-FR-P.1 Rev. 0000 Page 5 of 25 Step Action/Expected Response Response Not Obtained

3. (continued)

IF Faulted S/G necessary for RCS temp control, THEN CONTROL feed flow to minimum detectable flow to that S/G.

IF Faulted S/G NOT necessary for RCS temp control, THEN ISOLATE all feedwater to Faulted S/G.

4. MONITOR CST volume greater than 200,000 gal.

INITIATE CST refill USING 1-SOI-59.01, Demineralized Water

System. IF CST volume drops to less than 5000 gal, THEN MONITOR AFW pumps to ensure suction transfer.

5. MINIMIZE RCS cooldown:
a. CHECK at least one intact S/G NR level

greater than 29% [39% ADV].

a. CONTROL total feed flow to maintain greater than 410 gpm

UNTIL NR level in at least one Intact S/G greater than

29% [39% ADV].

b. CONTROL feed flow to intact S/G(s) as necessary.
c. CONTROL S/G pressures as necessary.

STEP DESCRIPTION TABLE FOR FR-P.1 Step _2_ STEP: Check RCS Cold Leg Temperatures -STABLE OR INCREASING PURPOSE: To determine if RCS cold leg temperature is still decreasing and, if so, to attempt to stop the decrease BASIS: Cold leg temperature is the best available indication of vessel downcomer temperature. It is important to terminate, if possible, any cooldown in progress to limit the extent of possible vessel damage due to excessive thermal stresses. If the RCS cold leg temperatures are decreasing the operator is instructed to eliminate any secondary-side or RHR System instigated RCS cooldown. The items checked in this step are in a preferred order such that the most probable causes of the cooldown are checked first. Therefore, any valves that dump steam are verified to be closed. Next, any cooldown from the RHR System terminated. A cooldown caused by overfeeding the intact SGs is stopped by controlling feed flow consistent with minimum secondary heat sink requirements. The operator checks for any faulted SGs and isolates them. Finally, if a faulted SG is necessary for RCS temperature control or if all SGs are faulted, feed flow to those SGs is controlled at a minimum measurable value to minimize the effects of the RCS cooldown due to the secondary side depressurization. ACTIONS: o Determine if RCS cold leg temperatures are stable or increasing o Determine if SG PORVs are closed o Determine if condenser steam dump valves are closed o Determine if RHR System is in service o Stop any cooldown from RHR System o Determine if main steamline isolation and bypass valves are closed o Determine if a faulted SG is necessary for RCS temperature control o Determine if all SGs are faulted o Isolate all feedwater to faulted SG(s) unless necessary for RCS temperature control o Close steam supply valves from faulted SG(s) to turbine-driven AFW pump o Control feed flow to non-faulted SG(s) to stop RCS cooldown by maintaining total feed flow greater than (S.02) gpm until narrow range level greater than (M.02)% [(M.03)% for adverse containment] in at least one non-faulted SG o Control feed flow at (S.04) gpm to the faulted SG that is necessary for RCS temperature control o Control feed flow at (S.04) gpm when all SGs are faulted FR-P.1 Background HFRPlBG.doc 20 HP-Rev. 2, 4/30/2005 STEP DESCRIPTION TABLE FOR FR-P.1 Step _2_ INSTRUMENTATION: o RCS cold leg temperature indication o SG PORVs position indication o Condenser steam dump valves position indication o SG pressure indication o Main steamline isolation and bypass valves position indication o Steam supply valves to turbine-driven AFW pump position indication o Feed flow indication o SG narrow range level indication o Plant specific instrumentation to indicate if RHR System is in service CONTROL/EQUIPMENT: o Switches for: SG PORVs Condenser steam dump valves -Main steamline isolation and bypass valves Feedline isolation valves Steam supply valves to turbine-driven AFW pump Feed flow control valves (both main and AFW) o Plant Specific RHR System controls KNOWLEDGE: Thermal stresses caused by rapid RCS temperature transients are not immediately removed if cooldown is stopped or RCS heatup occurs. Therefore, additional steps to minimize RCS pressure and to perform an RCS temperature "soak" are necessary. FR-P.1 Background HFRPlBG.doc 21 HP-Rev. 2, 4/30/2005 STEP DESCRIPTION TABLE FOR FR-P.1 Step _2_ PLANT-SPECIFIC INFORMATION: o (M.02) o (M.03) o (S.02) o (S.04) SG level just in the narrow range, including allowances for normal channel accuracy and reference leg process errors. SG level just in the narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%. The minimum safeguards AFW flow requirement for heat removal, including allowances for normal channel accuracy (typically one MD AFW pump capacity at SG design pressure). This flow is equivalent to the minimum AFW flow design requirement that must be delivered to the intact steam generator as assumed in the main feedline break safety analysis. Feed flow value in plant specific units corresponding to 25 gpm. The 25 gpm value is representative of a minimum measurable feed flow to a steam generator. Plant specific values may depend upon flow instrumentation and the sensitivity of the controls on feed flow. FR-P.1 Background HFRPlBG.doc 22 HP-Rev. 2, 4/30/2005

27. Given the following timeline:

00:00:00 Unit 1 is at 100% power. 00:00:00 Loss of Offsite power results in a Reactor Trip. 0 1:00:00 ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (with RVLIS), is in progress. PZR pressure is 785 psig and bei ng LOWERED by auxiliary spray. Core Exit TCs indicate 520°F. RVLIS Upper Plenum range indicates 81%. Assuming NO additional operator action, which ONE of the following describes the expected RVLIS and PZR level trends as the depressurization continues? RVLIS Level will ____(1)____ A ND PZR Level wi ll ____(2)____. (1) (2) A. RISE RISE B. RISE LOWER C. LOWER RISE D. LOWER LOWER CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: Plausible if the candidate determines that the bubble in Rx vessel head will decrease with the use of aux spray, and PZR level would increase with the

use of aux. spray. B. Incorrect: Plausible if the candidate determines that the bubble in Rx vessel head will decrease with the use of aux spray, thus RVLIS level would increase and PZR level would decrease. C. Correct: With aux spray in service as the PZR pressure is decreased the bubble in Rx vessel head will grow. The expected plant response for this condition is that as the bubble in Rx head increases, RVLIS will decrease and PZR level will

increase D. Incorrect: Plausible since Rx vessel head bubble would grow causing RVLIS to decrease but PZR level would increase not decrease.

Question Number: 27 Tier: 1 Group: 2

K/A: Westinghouse E10 Natural Circulation with Steam Void in Vessel with/without RVLIS EA1. Ability to operate and / or monitor the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS) EA1.3 Desired operating results during abnormal and emergency situations.

Importance Rating: 3.4 3.7

10 CFR Part 55: (CFR: 41.7 / 45.5 / 45.6)

10CFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate determine the expected plant response to LOWERING RCS pressure during Natural Circulation conditions. Technical

Reference:

Westinghouse basis document for ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (with RVLIS) Proposed references to

be provided: None Learning Objective: 3-OT-EOP0000, Emergency Instructions Reactor Trip Or Safety Injection, 1-E-0, ES-0.1, 1-ES-0.2, ES-0.3

and ES-0.4

14. Using 1-ES-0.2 DETERMINE the correct response for RCS parameters indicating RCS voiding.

Cognitive Level: Higher X Lower Question Source: New Modified Bank Bank X Question History: Bank question adopted from SQN question ES-0.3-B.2 001. . Comments: STEP DESCRIPTION TABLE FOR ES-0.3 Step 2 STEP: Establish PRZR Level To Accommodate Void Growth PURPOSE: To ensure that there is adequate space in the PRZR to allow the displacement of fluid from the primary system due to the formation of a void in the vessel BASIS: In this guideline as the primary system is cooled and depressurized under natural circulation conditions, a potential for void formation in the upper head region exists. If a void does form, it will displace primary fluid from the vessel into the PRZR as it grows. Therefore, before any further cooldown/depressurization is performed, the PRZR level must be low enough to accommodate this void growth and high enough to cover the PRZR heaters and prevent letdown from isolating. A level between (D.06)% and (D.12)% satisfies these requirements. In addition, PRZR level controls are placed in manual to allow any increase in PRZR level due to void growth. ACTIONS: o Determine if PRZR level is between (D.06)% and (D.12)% o Control charging and letdown as necessary o Place PRZR level control in manual INSTRUMENTATION: o PRZR level indication o Charging flow indication o Letdown flow indication o PRZR level control setting indication o Charging flow control valve position indication o Plant specific letdown valve position indication (includes letdown isolation and orifice isolation) CONTROL/EQUIPMENT: o Charging flow control valve controls o Plant specific letdown control valves o PRZR level control ES-0.3 Background HES03BG.doc 20 HP-Rev. 2, 4/30/2005

28. Given the following timeline:

00:00:00 Unit 1 is at 100% power. 00:00: 0 1 Main Turbine Trips. 00:00:03 A Reactor Trip occurs. Which ONE of the following describes how DNBR is affected and the RCP design feature that assists in RCS heat removal? In accordance with the Final Safe ty Analysis Report (FSAR), at 00:00: 02 , the departure from nucleate boiling rati o (DNBR) is ____(1)____. The MINIMUM heat removal needed to prevent core damage after a Reactor Trip is DIRECTLY ensured by ____(2)____. A. (1) RISING (2) a flywheel attached to each RCP B. (1) LOWERING (2) a flywheel attached to each RCP C. (1) RISING (2) maintaining the RCPs connected to the generator and external network for 30 seconds D. (1) LOWERING (2) maintaining the RCPs connected to the generator and external network

for 30 seconds

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: As seen in table 15.2-1 (sheet 4 of 7) of the Watts Bar Final Safety Analysis Report (FSAR), minimum DNBR occurs 3.7 seconds after a partial loss of cooling occurs. This table considers the case in which one RCP begins coasting down at time 0. It is plausible to believe that DNBR rises at the time given in the stem because DNBR does rise

after the reactor is tripped. The case given in the stem would cause an automatic reactor trip as one or more RCPs is secured with nuclear power greater than P-8 (48%). It is correct that a flywheel attached to each RCP acts to assure the minimum heat removal needed to

prevent core damage after the reactor trip. This fact is observed in system description WBN-SDD-N3 4001, Reactor Coolant System. B. Correct: DNBR would be LOWERING at the time specified in the stem of the question. Also, it is correct that the RCP flywheel serves the purpose asked by the question. C. Incorrect: It is Incorrect and yet plausible that DNBR would be RISING at the time given in the stem. Additionally, it is Incorrect that core damage is prevented by maintaining the RCPs connected to the generator and external network for 30 seconds. Again, the RCS system description credits the flywheel of each RCP for this prevention. It is plausible to believe this because, as seen in the same system description, In the event of a turbine trip actuated by either the Reactor Trip System or the Turbine Protection Syst em, the generator and RCPs are maintained connected to the exte rnal network for 30 seconds to ensure full flow before transfer is made. This protects the turbine/generator from overs peed upon tripping the turbine. Additionally, chapter 15 of the FS AR considers that a Loss of Offsite power occurs after the reac tor trip (of the partial lo ss of flow accident). Therefore, the FSAR r egards that all RCPs are lost immediately after the reactor trip. D. Incorrect: While it is correct that the DNBR would be LOWERING at the time specified, it is not correct and yet plausible that the 30 second post trip connection of the grid, generator and RCPs would prevent core

damage. Question Number: 28 Tier: 2 Group: 1

K/A: 003 Reactor Coolant Pump System (RCPS) K5 Knowledge of the operational implications of the following concepts as they apply to the RCPS: K5.02 Effects of RCP coastdown on RCS parameters

Importance Rating: 2.8 3.2

10 CFR Part 55: (CFR: 41.5 / 45.7)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to understand the effects of the RCP coastdown on the DNBR. Also the applicant must understand what permits the removal of core decay heat during the RCP coastdown phase. Technical

Reference:

Watts Bar Final Safety Analysis Report (FSAR), chapter 15 WBN-SDD-N3-68-4001, Reactor Coolant System. Proposed references to be provided: None Learning Objective: 3-OT-TS0303, TS 3.3 "INSTRUMENTATION" 1. DESCRIBE the following aspects of Technical Specifications and Technical Requirements for Instrumentation Tech Specs: a. The Limiting Conditions for Operation, Applicability, and Bases for the LCO.

b. The conditions and required actions with completion

time of one hour or less. Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBNP-6 TABLE 15.2-1 (Sheet 4 of 7)

TIME SEQUENCE OF EVENTS FOR CONDITION II EVENTS (Cont'd) Accident Event Time (sec.) Partial Loss of Forced Reactor

Coolant Flow (four loops operating, one pump

coasting down) One pump begins coasting down

Low flow trip setpoint reached

Rods begin to drop Minimum DNBR occurs 0 1.32 2.52 3.7 Loss of External

Electrical Load

1. With pressurizer control Loss of electrical load 0 OT deltaT reactor trip point reached 8.5 Rods begin to drop 10.0 Minimum DNBR occurs 11.6 2. Without pressurizer control Loss of electrical load 0 High pressurizer pressure reactor trip point reached 4.5 Rods begin to drop 6.5 Peak Pressurizer pressure occurs 7.9

(!) z -=-< c:::: 0 0 w z -< I-z < -0 ..I < < a: z u w-:I 00 D. z a: u. <o wZ ..I 0 o-:> .... a: u. -0 2 " 6 TIME (SECONDS) a 10 WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT NUCLEAR POWER TRANSIENT FOUR PUMPS IN OPERATION ONE PUMP COASTING DOWN FIGURE 15.2-16 (.!) z .... * < 0:: Q Q ..... z .... < I-z .... < ::::E Q 3 < u 2.5 0: 2 0 1.5 1 0 2 4 6 TIME (SECONDS) 8 10 WATTS BAR NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT DNBR VERSUS TIME FOUR PUMPS IN OPERATION ONE PUMP COASTING DOWN FIGURE 15.2-17 WBN System Description Document REACTOR COOLANT SYSTEM WBN-SDD-N3-68-4001 Rev. 0034 Page 71 of 257 3.2.3 Reactor Coolant Pumps (continued)

8. Flywheel The flywheel stores energy to keep the pump operating for a short but critical period following a power failure. It provides additional inertia to extend pump coast-down. The flywheel is available for inspection by removing the cover. D. Oil Lift System An external oil lift system provides lubrication to the upper radial bearings during motor startup. This oil lift system is shown in Figure 2. High-pressure oil is also delivered simultaneously to the upper and lower sets of thrust bearing shoes. The oil lift system will provide from 0.001" to 0.003" of oil film between the runner and the thrust bearing shoes depending on the amount of pump thrust on the bearings. Thus, regardless of thrust condition (either upthrust or downthrust), an oil film between the shoes and runner is present. At the same time, high-pressure oil is being sprayed into the guide bearing chamber for lubricating the upper radial guide bearings. There are low and high oil level alarms in the upper and lower oil reservoirs. The oil lift pumps have spray shields to ensure no lube oil comes in contact with hot reactor coolant in case of component rupture. The oil lift system is required to operate approximately two minutes before RCP startup and approximately one minute after the pump reaches full speed. E. RCP Power The RCPs are powered through the 6.9KV RCP boards. Undervoltage to any two of the four pumps will cause a reactor trip above approximately 10% reactor power. Underfrequency to any two of the four pumps will trip all four RCPs and cause a reactor trip above approximately 10% power. 3.2.4 Pressurizer (PZR)

The PZR provides a point in the RCS where liquid and vapor can be maintained in equilibrium under saturated conditions for RCS pressure control. The PZR and its sprays, heaters, PORVs, safety valves, and surge line make up the pressure control equipment. This equipment is designed to accommodate changes in system volume and to limit changes in system pressure due to RCS loop temperature variation during all modes of operation. In particular, the equipment is designed to provide satisfactory operation without reactor trip when the RCS is subjected to the following transients: a. Loading or unloading at 5% of full power per minute, with automatic reactor control. b. Instantaneous load transients of +/-10% of full power (not exceeding full power), with automatic reactor control. c. Step-load reduction of 50% of full power with automatic reactor control (10%) and steam dump (40%). WBN System Description Document REACTOR COOLANT SYSTEM WBN-SDD-N3-68-4001 Rev. 0034 Page 97 of 257 3.4.2 RCS Loop Instrumentation and Control (continued) The ICCM microprocessor also provides input to a digital saturation meter (one for each train) at the main control board which provides a backup means to assess saturation temperature corresponding to RCS pressure from 0-3000 psig. The plant process computer is also utilized to continuously monitor pressure and temperature margins to saturation of the RCS. E. RCS Hot Leg Level Indication (Ref. 7.4.6) The Westinghouse Ultrasonic Level Measurement System (ULMS) is designed to provide a highly accurate indication of water level in the RCS piping during plant shutdown. Various maintenance and inspection activities performed during plant shutdown require that RCS water level be lowered into the loop piping, often referred to as "mid-loop" operation. Operation above a minimum water level in the hot leg is always necessary to prevent air ingestion by the RHP Pumps, with the potential for a resulting loss of decay heat removal capability. The ULMS was specifically developed to provide the accurate indication of water level necessary during mid-loop operations, due to the narrow operating window and the potential severity of loss of decay heat removal events. In addition to the ULMS, a differential pressure level transmitter and a liquid level gauge are provided as permanent backup instruments along with a Mansell Level Monitoring System as a temporary backup instrument to monitor the reactor vessel water level during mid-loop operations. A second D/P level transmitter provides wide range indication from the bottom of the hot leg to an elevation above the pressurizer lower level instrument tap. 3.4.3 Steam Generator Instrumentation and Control See the Main Steam and Main FeedWater System description (Ref. 7.2.1 and 7.2.46). 3.4.4 RCPs Instrumentation and Control A. RCP Power Supply The four RCPs are connected to separate buses of the 6.9KV boards which normally

receive power from the main generator. These boards receive emergency power from the 6.9KV start buses. The power supply to each RCP consists of a normally closed breaker and a normally open breaker which are electrically interlocked. If the normally closed breaker should sense a loss of power on its bus, it will open and the normally open breaker will close providing power from another bus. During plant startup and shutdown operations the four RCP buses receive power from an offsite source. Complete loss of normal ac power results in reactor trip on RCP bus undervoltage or underfrequency and turbine trip (Ref. 7.2.44). The flywheel inertia of each RCP sustains RC flow for a period of time (1-1/2 minutes) after the trip sufficient to assure the minimum heat removal needed to prevent core damage after the trip. Natural circulation of the coolant provides sufficient flow for reactor core residual heat removal until power can be restored to the pumps. WBN System Description Document REACTOR COOLANT SYSTEM WBN-SDD-N3-68-4001 Rev. 0034 Page 98 of 257 3.4.4 RCPs Instrumentation and Control (continued) In the event of a turbine trip actuated by either the Reactor Trip System or the Turbine Protection System, the generator and RCPs are maintained connected to the external network for 30 seconds to ensure full flow before transfer is made. This protects the turbine/generator from overspeed upon tripping the turbine. In case a generator trip deenergizes the pump buses, the RCP motors will be transferred to offsite power within 6 to 10 cycles. See Table 2 for RCP power supplies. B. RCP Motor Oil Lift Pump Each RCP has a normal and an alternate handswitch in the MCR for pump startup. These handswitches each have backup control handswitches and transfer switches. Both sets of controls require prior operation of the oil lift pump for approximately two minutes and are nondivisional. Each oil lift pump contains its own handswitch and transfer switch. A pressure switch is provided on the high pressure oil lift system. Upon low oil pressure, the switch will actuate a MCR alarm. In addition, the switch is part of an interlock system that will prevent starting of the RCP until the oil lift pump is started manually prior to starting the RCP motor. A local pressure gauge is also provided. The control loop for these instruments is nondivisional. C. RCP Motor Oil Level Detection Each RCP contains a control loop for oil reservoir level indication. An annunciator for high or low oil level is located in the MCR. Each control loop contains two lower oil reservoir level indicators and two upper oil reservoir level indicators. All switches of the control loop are nondivisional.

29. Given the following conditions: - Unit 1 is at 40%.

Subsequently: - The #1 RCP trips. Which ONE of the following describes the effect the RCP trip will have on INDICATED secondary plant parameters?

SG #1 Steam Flow WILL IMMEDIATELY ____(1)____.

Assuming NO OPERATOR ACTION occurs: FIVE MINUTES AFTER the RCP trip, Feed Flow WI LL be provided to the SGs via _____(2)_____ feedwater.

(1) (2) A. RISE AUX B. RISE MAIN C. LOWER AUX D. LOWER MAIN CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: When a RCP trips, the flow in the RCS loop decreases, the temperature of the S/G lowers and due to the saturation conditions existent in the S/G, the pressure in the S/G will lower. As soon as the pressure in the S/G lowers to a va lue less than that of the other S/Gs, steam flow from the affected S/G will cease.

This portion of the distractor is plausible provided that the applicant does not understand the thermodynam ic properties governing the partial loss of flow accident. It is not correct that the plant would be provided AFW as a feed source. It is plausible to believe th is as if the plant tripped, this would be the case. The conditions in the st em indicate that 3 of 4 PR NI channels are less than or equal to 48%. This may be seen in the Response instruction for annunciator 70-C. Also seen in this instruction is that the single loop loss of flow Rx trip is blocked below 48%. If the applicant did not recognize that power was below the P-8 setpoint, they would believe that t he plant would trip on the loss of the RCP and that AFW would be feeding the S/Gs (as MFW would isolate post trip). B. Incorrect: Again, it is incorrect and ye t plausible that steam flow would rise. Again, as the plant does not trip, MFW continues to supply the S/Gs. C. Incorrect: It is true that steam flow would lower. Ho wever, it is not true and yet plausible that AFW supplies the S/Gs. D. Correct: It is true that steam flow would lower. It is true that MFW continues to feed the S/Gs.

Question Number: 29 Tier: 2 Group: 1

K/A: 003 Reactor Coolant Pump System (RCPS) K5 Knowledge of the operational implications of the following concepts as they apply to the RCPS: K5.04 Effects of RCP shutdown on secondary parameters, such as steam pressure, steam flow, and feed flow

Importance Rating: 3.2 3.5 10 CFR Part 55: (CFR: 41.5 / 45.7)

10CFR55.43.b: Not applicable K/A Match: This question matches the K/A by determining if the candidate knows the effect an RCP trip will have on SG parameters (steam flow and SG level) when the plant is at power. Technical

Reference:

FSAR Ch 15, section 15.2.5.1 Westinghouse Transient Accident Analysis 1-AOI-24, RCP Malfunctions During Pump Operation, Revision 0029 Proposed references to

be provided: None Learning Objective: 3-OT-TAA015 1. Discuss loss of coolant flow theory

a. Derive the time dependent behavior of T h b. Define DNR and DNBR

3-OT-AOI2400

12. Describe basic operator actions to shutdown an RCP. Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: Comments:

WBN Unit 1 Bypass, Intlk, & Pe rmissive 1-ARI-64-70 Rev. 0000 Page 42 of 47 A. Power reduction B. Unit trip [1] VERIFY Reactor power level less than or equal to 48%. NOTES 1) P-8 illuminated blocks single loop loss of flow Rx trip. 2) P-8 will automatically reset with 2/4 PR channels greater than 48%. W 7246D11 Sht. 2, 5, 8, 11, & 40 Source Setpoint NC41N 3/4 PR Channels less than or equal to 48% NC42N NC43N NC44N 70-C P-8 LO PWR-FLOW TRIP BLOCKED (Page 1 of 1) Probable Cause: Corrective Action:

References:

30. Given the following print excerpt:

Which ONE of the following states the purpose of the flowpath highlighted in yellow in the print excerpt shown above?

The highlighted line __________. A. automatically maintains level in the Boric Acid Tank (BAT) "A" through the automatic action of 1-FCV 237, "BAT A RECIRC CONTROL." B. is used to manually makeup to the BAT A in accordance with SOI-62.05, "Boric Acid Batching, Transfer, And Storage." C. provides the required chemical mi xing of the BAT A as 1-FCV-62-237 is normally THROTTLED OPEN to achieve a flow rate specified by chemistry. D. provides miniflow protection to the Boric Acid Transfer Pumps as 1-FCV-62-237 is normally T HROTTLED OPEN to 25%. CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: As seen in SSD-1-LPF-62-237, 1-FC V-62-237 is only controlled by a manual controller. Ther e are no automatic functions associated with this valve. Therefore, it is incorrect that the level in BAT "A" will be controlled by the automatic action of this valve. It is plausible to believe that the BAT "A" has an aut omatic level control because other tanks in the plant do have such (inc luding those which are part of the CVCS and which do affect reactivity---e.g. the VCT). B. Correct: As seen in 0- SOI-62.05, "Boric Ac id Batching, Transfer, And Storage" 1-FCV-62-237 is manually controlled to raise level in the BAT "A." C. Incorrect: As the name of 1-FCV-62-237 implies ("BAT A RECIRC CONTROL"), it is used to recirculate the BAT "A." However, as seen in the system description, WBN-SDD-N3-62-4001, "Chemical and Volume Control System," the continuous recirculation of the tank is afforded by the 3/4" line which includes the normally open valve 1-BYV-62-1057 and orifice

1-OR-62-237. It is plausible to believe that 1-FCV-62-237 would be left open for chemical mixing because a fter a makeup to BAT "A," the operators will recirculat e the tank using 1-FCV-62-237 to satisfy the Chemistry department's requirements. However, after Chemistry obtains their required samples, 1-FCV-62-237 will be closed. D. Incorrect: Again, 1-FCV-62-237 is a normally cl osed valve. It is plausible to believe that it provides miniflow protection to the BATPs because if it were left throttled open, it would do so.

Question Number: 30 Tier: 2 Group: 1

K/A: 004 Chemical and Volume Control System (CVCS) K5 Knowledge of the operational implications of the following concepts as they apply to the CVCS: K5.31 Purpose of flow path around boric acid storage tank Importance Rating: 3.0 3.4

10 CFR Part 55: (CFR: 41.5/45.7) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to demonstrate the knowledge of the purpose for the flow path around the BAT. Technical

Reference:

Print 1-47W809-5 WBN-SDD-N3-62-4001, Chemical and Volume Control System 0-SOI-62.05, Boric Acid Batching, Transfer, And Storage

SSD-1-LPF-62-237 Proposed references to

be provided: None Learning Objective: 3-OT-STG-062A, CHEMICAL AND VOLUME CONTROL SYSTEM

6. EXPLAIN the CVCS design features and/or interlocks that provide the following:
i. RCS Boron concentration control and modes of operation of the Reactor Makeup Control System (blender controls) q. Boron crystallization prevention
s. Chemical injection Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN System Description Document CHEMICAL AND VOLUME CONTROL SYSTEM WBN-SDD-N3-62-4001 Rev. 0035 Page 76 of 176 3.2.7 Boric Acid Blender (continued) The blenders were manufactured by W. They are austenitic stainless steel and consists of a conventional pipe tee. The blenders are provided with water inlet, boric acid inlet, and discharge connections. The boric acid blenders (W Spin No. CSHDBL) were purchased from W under contract No. 71C60-54114-1. 3.2.8 Orifices A. Boric Acid Transfer Pump Bypass Orifices There are three pump bypass orifices (one per BAT) located in the auxiliary building. Each orifice is designed to pass the minimum flow required for sufficient recirculation through the piping and tanks with the boric acid transfer pumps. The design operation requirements are listed in Table 11. The orifices normally operate at 100 psig and

ambient temperature. The orifices were supplied by W. The boric acid transfer pump bypass orifices (W Spin No. CSORBA) were purchased from W under contract No. 71C60-54114-1. B. CCP Miniflow Orifices There are two CCP miniflow orifices (one per CCP) located in the auxiliary building. These orifices provide a minimum flow for pump protection. The design operating requirements are listed in Table 11. The orifices are shown on Drawing MF-48590 in the CCP manual (Ref. 7.3.9). These orifices (W Spin No. CSORCP) were purchased from W under contract No. 71C62-54114-1. C. Chemical Mixing Tank Orifices There is one chemical mixing tank orifice located in the auxiliary building (El. 713). An orifice is provided in the piping upstream of the mixing tank. This orifice limits the flow rate through the tank to 2 gpm to avoid slugging the pump seals with concentrated chemicals. The design operating requirements are listed in Table 11. The orifices are normally operated at ambient temperatures. The chemical mixing tank orifices (W Spin No. CSORCM) were purchased from W under contract No. 71C60-54114-1. They are shown on W drawing 271C403 (Ref. 7.3.19). D. Letdown Orifices There are four letdown orifices located in the reactor building. Two 75 gpm orifices and one 45 gpm orifice are provided to reduce letdown pressure from RCS conditions and to control the letdown flow. An approximately 3 gpm orifice is provided to limit the rate of thermal change on the Class A welds upstream of the regenerative HX. WBN SSD LPF-62-237 PAGE 2 OF 11 REVISION 01 Reviewed By: R. J. HUNT Approved By: R.E. McSpadden Date: RF012 SCALING AND SETPOINT DOCUMENT LOOP DRAWING Loop Name: Boric Acid Tank A Recirc Flow

WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 23 of 124 Date________ Initials 6.3 Transfer BA from BAT C To BAT A NOTE This Section is for BAT A in service to U1 via BA Pump 1A. [1] IF BAT A in service to U1, THEN [1.1] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to STOP and RELEASE. ________ [1.2] CHECK 1-HS-62-140A, Green light LIT. ________ [2] RECORD initial BAT A LEVEL from 1-LI-62-238 [1-M-6]________gals. ________ [3] RECORD initial BAT C LEVEL from 1-LI-62-242 [1-M-6]________gals. ________ [4] REQUEST SRO to evaluate TR 3.1. 1/3.1.2 prior to stopping BA pumps. ________ [5] PERFORM the following: NOMENCLATURE LOCATION POSITION UNID PERF INITIAL BA PMP A 1-M-6 STOP 1-HS-62-230A BA PMP B 1-M-6 STOP 1-HS-62-232A [6] OPEN 1-ISV-62-1053B, BA XFER PUMP 1B-B DISCHARGE [A12R/713]. ________ [7] START BA Pump 1B, using 1-HS-62-232A, BA PMP B [1-M-6]. ________ [8] IF desired, THEN PLACE BA pump 1B in FAST using 1-HS-62-232D. ________ [9] THROTTLE 1-FCV-62-237, BAT A RECIRC CONTROL [1-M-6], as desired to control rate of transfer. ________ WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 24 of 124 Date________ Initials 6.3 Transfer BA from BAT C To BAT A (continued) CAUTION Damage may occur to BAT Heaters if leve l drops below 9.5% with heaters ENERGIZED. [10] IF BAT C level decreases to 2270 gals (10%), THEN ENSURE the following: NOMENCLATURE LOCATION POSITION UNID PERF INITIAL BORIC ACID TANK C HEATER A A13R/713 OFF 0-HS-62-243 BORIC ACID TANK C HEATER B A12R/713 OFF 0-HS-62-246 [11] IF BAT is to be drained below heaters, THEN DRAIN and FLUSH completely to prevent precipitation. ________ [12] WHEN desired level is obtained, THEN STOP BA Pump 1B, using 1-HS-62-232A, BA PMP B. ________ [13] CLOSE 1-FCV-62-237, BAT A RECIRC CONTROL .[C.2] ________ ________ IV [14] CLOSE 1-ISV-62-1053B, BA XFER PUMP 1B-B DISCHARGE. [C.2] ________ ________ IV [15] START BA Pump 1A with 1-HS-62-230A, BA PMP A. ________ [16] ENSURE BA pump 1B in SLOW using 1-HS-62-232D. ________ [17] START BA Pump 1B with 1-HS-62-232A, BA PMP B

(N/A if BAT C removed from service). ________ [18] RECORD final BAT A LEVEL from 1-LI-62-238________gals. ________ [19] RECORD final BAT C LEVEL from 1-LI-62-242. ________gals. ________ WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 25 of 124 Date________ Initials 6.3 Transfer BA from BAT C To BAT A (continued) [20] REQUEST Chemistry Lab sample BAT A. ________ [21] IF 1-HS-62-140A, VCT MAKEUP CONTROL, was positioned to STOP in Step 6.3[1], THEN [21.1] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to ST ART and RELEASE. ________ [21.2] CHECK 1-HS-62-140A, Red light LIT. ________ [22] REQUEST SRO to verify requirement s of TR 3.1.5 or 3.1.6 met. ________ End of Section D55A 0-E ..... m-S:Z-231 c j* "'" I-LT 82-233 ...... " DRAIN

31. Which ONE of the following describes t he operation of the motor operated mini-flow valves; in regards to the RHR system flow rates, in accordance with 1-SOI-74.01, Residual Heat Removal System?

The motor operated mini-flow valve will be OPENED if RHR pump flow falls below ____(1)____ gpm AND will be CLOSED when the flow INCREASES above ____(2)____ gpm. A. (1) 250 (2) 1100 B. (1) 250 (2) 1400 C. (1) 750 (2) 1100 D. (1) 750 (2) 1400 CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: As seen in Appendix A, RHR Pump Operating Flow Limits of 1-SOI-74.01, the minimum flow operating of the RHR pumps is 750-1400 gpm. These setpoints were modified upon recommendation from Westinghouse from their original values of 250-1100 gpm. The modification occurred after two RHR pumps were seized and analysis from Westinghouse concluded that low mini-flow setpoints were subjecting the RHR pumps to a temper ature swing in excess of 200ºF. When the analysis for the Fire Safe Shutdown was performed, the setpoints for the RHR pump mini-flow was 250-1100 gpm. Therefore, as seen on page III-16 of PART III - SAFE SHUTDOWN CAPABILITIES of the Fi re Protection Report, the setpoints are listed as 250-1100 gpm. Therefore, it is pl ausible to believe that the mini-flow setpoints are 250-1100 gpm. B. Incorrect: Again it is Incorrect and yet plausible that the low setpoint for the RHR pump mini-flow valve be 250 gp

m. It is correct t hat the high setpoint for such valve is 1400 gpm.

C. Incorrect: It is correct that the low setpoint for the RHR pump mini-flow valve be 750 gpm. It is Incorrect and yet plausible that the high setpoint be 1100 gpm. D. Correct: It is correct that the RHR pump mini-flow setpoints are 750-1400 gpm.

Question Number: 31 Tier: 2 Group: 1

K/A: 005 Residual Heat Removal System (RHRS) A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: A1.02 RHR flow rate

Importance Rating: 3.3 3.4

10 CFR Part 55: (CFR: 41.5 / 45.5)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to demonstrate the knowledge of the setpoints for RHR mini-flow operation. The applicant must possess this knowledge to be able to monitor for changes in RHR flow rate to prevent exceeding design limits (specifically, RHR pump overheating). Technical

Reference:

1-SOI-74.01, Residual Heat Removal System PART III - SAFE SHUTDOWN CAPABILITIES of the

Fire Protection Report

Proposed references to

be provided: None Learning Objective: 3-OT-STG-074, THE RESIDUAL HEAT REMOVAL SYSTEM

6. EXPLAIN the RHR System design features and/or interlocks of the following: c. Function of RHR pump miniflow recirculation Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Residual Heat Removal System 1-SOI-74.01 Rev. 0002 Page 193 of 195 Appendix A (Page 1 of 1) RHR Pump Operating Flow Limits

PART III - SAFE SHUTDOWN CAPABILITIES III-16 Rev. 10 A minimum-flow return line from the downstream side of each RHR heat exchanger to the corresponding pump's suction line is provided to assure that the RHR pumps do not overheat under low flow conditions. A motor-operated valve located in each minimum flow line will be opened if RHR pump flow falls below 250 gpm and will be closed when the flow increases above 1100 gpm. The cooldown rate of the reactor coolant is controlled by regulating the flow through the tube side of the RHR heat exchangers. A bypass line, which serves both residual heat exchangers, is used to regulate the temperature of the return flow to the RCS as well as to maintain a constant flow through the RHR system.

The RHR system can be placed in operation when the pressure and temperature of the RCS are about 400 psig and 350 0F, respectively. If one of the pumps and/or one of the heat exchangers is not operable, safe operation of the plant is not affected; however, the

time for cooldown is extended. 4.5.1 Residual Heat Removal Pumps Two identical pumps are installed in the RHR system. Each pump is sized to deliver sufficient reactor coolant flow through the residual heat exchangers to meet the plant cooldown requirements. A seal heat exchanger for each pump is supported by operation of CCS. 4.5.2 RHR Safety Valves The RHR system safety valves provide RCS cold overpressure protection whenever the RHR system is in operation. The valves are located inside containment, one each on the RHR system suction and discharge path, and di scharge to the pressurizer relief tank. The valves are set at 450 psig and 600 psig, respectively.

4.6 Safety Injection System Accumulators - Key 36 During normal plant operating conditions, the safety injection system accumulators are pressurized by nitrogen gas in order to inject borated water in the RCS when RCS pressure falls below 600 psi unintentionally. During a controlled depressurization, the accumulators are isolated to prevent injection of safety injection system accumulator borated water. However, if the isolation valves remain open, the borated water will be injected and nitrogen pressure will decrease due to the increased empty volume of the accumulators. Injection of nitrogen into the RCS occurs when RCS pressure is less than 150 psi. The manual isolation of the accumulators is assumed as a post-fire activity. The isolation valve at each accumulator is closed only when the RCS is intentionally depressurized below 1000 psig. If the cables associated with these valves were damaged by fire, isolation is performed locally, governed by appropriate plant procedures (post-fire). In the event the valves are inaccessible, RCS pressure will be maintained greater than 150 psi to preclude nitrogen injection into the RCS via the accumulators.

32. Given the following conditions: - Unit 1 is at 100% power.

Which ONE of the following describes the physical location of 1-BKR-74-2A, LOOP 4 HOT LEG TO RHR SUCTION (1-FCV-74-2) AND the NORMAL position of this breaker?

1-BKR-74-2A is located on a ____(1)____ board AND is required to be ____(2)____. A. (1) Rx MOV (2) ON B. (1) Rx MOV (2) OFF & Locked C. (1) C&A Vent

(2) ON D. (1) C&A Vent (2) OFF & Locked

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: It is correct that the breaker listed is located on a Rx MOV board. It is also not correct that 1-BKR-74-2A be left ON or shut. As seen in 0-PI-OPS-17.1, 18 month Locked Breaker Verification, this breaker is positioned OFF (for Appendix R concerns). It is plausible to believe that this breaker would be left on because many MOVs have their supplies left ON. Also, circuit interlocks exist which prevent this valve from opening given that RCS parameter s are not within limits. Finally, administrative restrictions exist which prohibit manipulating this breaker's valve. B. Correct: The breaker listed is located on a Rx MOV board. Also, it is locked OFF. C. Incorrect: As seen on bot h 1-SOI-74.01 ATT 1P, Residual Heat Removal System Power Checklist 1-SOI-74.01 ATT 1P, 1-BKR-74-2A resides on the 480V Reactor MOV Board 1B1-B. It is plausible to believe that this breaker were to reside on a C&A v ent board because there are safety related MOVs which do have there supply breakers located on a C&A

vent board. Again, it is incorrect and yet plaus ible that this breaker would be left ON. D. Incorrect: Again it is Incorrect and yet plausible that the breaker listed reside on a C&A vent board. Also, it is corre ct that the breaker listed be locked OFF. Question Number: 32 Tier: 2 Group: 1

K/A: 005 Residual Heat Removal System (RHRS) K2 Knowledge of bus power supplies to the following: K2.03 RCS pressure boundary motor-operated valves

Importance Rating: 2.7 2.8

10 CFR Part 55: (CFR: 41.7)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to correctly identify on which board the supply breaker for a RHR pressure boundary valve is located (1-BKR-74-2A). The applicant must further identify the locking requirement applicable to this bus power supply breaker. Technical

Reference:

1-SOI-74.01 ATT 1P, Residual Heat Removal System Power Checklist 0-PI-OPS-17.1, 18 month Locked Breaker Verification

Proposed references to

be provided: None Learning Objective: 3-OT-STG-074, THE RESIDUAL HEAT REMOVAL SYSTEM

5. LIST the bus power supplies to the following RHR System components:
b. RHR MOVs 1-FCV-74-1, 2, 8, 9, 3, 21 Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Residual Heat Removal System Power Checklist 1-74.01-1P 1-SOI-74.01 ATT 1P Revision 0000

Effective Date: 12-03-2013

Page 1 of 1 Performed On ________ NOMENCLATURE LOCATIO N POSITION UNID PERF INITIAL VERIFIER INITIAL 480V Reactor MOV Board 1A1-A LOOP 4 HOT LEG TO RHR SUCTION (1-FCV-74-1) C/10D ON 1-BKR-74-1B CVLOOP 4 HOT LEG TO RHR SUCTION (1-FCV-74-1) C/5B 0-PI-OPS-17.1 1-BKR-74-1A 480V Reactor MOV Board 1A2-A 1-FCV-74-2 BYPASS RHR SUCTION (1-FCV-74-8) C/6D 0-PI-OPS-17.1 1-BKR-74-8 480V Reactor MOV Board 1B1-B LOOP 4 HOT LEG TO RHR SUCTION (1-FCV-74-2) C/5B 0-PI-OPS-17.1 1-BKR-74-2A LOOP 4 HOT LEG TO RHR SUCTION (1-FCV-74-2) C/10D ON 1-BKR-74-2B CV1-FCV-74-1 BYPASS RHR SUCTION (1-FCV-74-9) C/5C 0-PI-OPS-17.1 1-BKR-74-9

WBN Unit 0 18 month Locked Breaker Ve rification 0-PI-OPS-17.1 Rev. 0020 Page 9 of 17 Appendix A (Page 2 of 5) Appendix A Locked Breaker List Date________ Nomenclature LocationBoard / Compartment PositionUNID Perf Initial V eri fInitial (CV) ERCW STRAINER 1A-A ISOL (1-FCV-67-22) Rx MOV Bd 1A2-A 7B OFF & Locked 1-BKR-067-22 AB ERCW SUP HDR 1A ISOL (1-FCV-67-81) Rx MOV Bd 1A2-A 8A OFF (2) & Locked 1-BKR-067-81 AB AIR CLR ERCW SUP HDR 1A ISOL (1-FCV-67-127) Rx MOV Bd 1A2-A 10B OFF (2)& Locked 1-BKR-067-127 CCS HX C HDR 1A SUPPLY (1-FCV-67-147) Rx MOV Bd 1A2-A 11B OFF (2) & Locked 1-BKR-067-147B CCS HX C OUT ERCW HDR A FLOW CNTL (0-FCV-67-151) Rx MOV Bd 1A2-A 12A OFF& Locked 0-BKR-067-151 CCS HX C HDR 1B SUPPLY

(1-FCV-67-223) Rx MOV Bd 1A2-A 13A OFF(2) & Locked 1-BKR-067-223B 1A ESF EQUIP CCS SUP HDR (1-FCV-70-2) Rx MOV Bd 1A2-A 14A OFF(2)& Locked 1-BKR-070-2 MISC EQUIP CCS SUP HDR (1-FCV-70-4) Rx MOV Bd 1A2-A 14B OFF(2) & Locked 1-BKR-070-4 CCS HX A OUTLET (1-FCV-70-8) Rx MOV Bd 1A2-A 14D OFF(2) & Locked 1-BKR-070-8 CCS HX A/C OUTLET XTIE (1-FCV-70-10) Rx MOV Bd 1A2-A 14E OFF(2) & Locked 1-BKR-070-10 CCS HX A/C INLET XTIE (1-FCV-70-23) Rx MOV Bd 1A2-A 15B OFF(2) & Locked 1-BKR-070-23 CCS HX A INLET (1-FCV-70-25) Rx MOV Bd 1A2-A 15D OFF(2) & Locked 1-BKR-070-25 CCS HX A HDR 1B ERCW SUP (1-FCV-67-458) Rx MOV Bd 1A2-A 15E OFF(2)& Locked 1-BKR-067-458 CCP MIN FLOW VALVE (1-FCV-62-99) (SHUNT TRIP BKR) Rx MOV Bd 1B1-B 2E2 OFF & Locked 1-BKR-062-99A SIP COLD LEG INJECTION (1-FCV-63-22) Rx MOV Bd 1B1-B 2F2 OFF & Locked 1-BKR-63-22A SIS CL ACCUM 2 OUT ISOL (1-FCV-63-98) Rx MOV Bd 1B1-B 3F2 OFF & Locked 1-BKR-63-98A LOOP 4 HOT LEG TO RHR SUCTION (1-FCV-74-2) Rx MOV Bd 1B1-B 5B OFF (1) & Locked 1-BKR-074-2A 1-FCV-74-1 BYPASS RHR SUCTION (1-FCV-74-9) Rx MOV Bd 1B1-B 5C OFF (1) & Locked 1-BKR-074-9 (1) Use padlock in place of seals (2) ACB may be ON to position valve, but MUST be OFF & Locked during normal operation (Appendix R). () Positions marked with an are identified/required by NE to be in stated position. Remarks:

33. Given the following conditions: - Unit 1 has suffered a large break LOCA. - The following indications exist on 1-M-6:

Based on these conditions, it is expected that transfer to the containment sump WILL ______________.

A. occur, due to all required conditions having been met B. NOT occur due to RWST level being too high C. NOT occur due to containment sump level being too low D. NOT occur due to an SI signal still being supplied to the RHR sump valves CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: RWST level, containment sump level and the presence of the SI signal satisfy all conditions for the transfer to the containment sump signal to occur. The required logic of 2 of 4 channels is satisfied. B. Incorrect: RWST level is less than t he required 34.63% (on 2 of 4 channels) to satisfy the logic for transfer to the containment sump. C. Incorrect: Containment sump level is greater than the required 16.1% (on 2 of 4 channels) to satisfy the logic for tr ansfer to the cont ainment sump. D. Incorrect: Plausible, since an SI signal is in the logic for the automatic transfer to the containment sump and the SI reset and the transfer to the containment sump reset due to SI are separate.

Question Number: 33 Tier: 2 Group: 1

K/A: 006 Emergency Core Cooling System (ECCS) A3 Ability to monitor automatic operation of the ECCS, including: A3.08 Automatic transfer of ECCS flow paths Importance Rating: 4.2 4.3

10 CFR Part 55: (CFR: 41.7 / 45.5) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to recognize the point at which the RHR pump suction transfer will occur. Technical

Reference:

NE SSD 1-L-63-50 NE SSD 1-L-63-180

1-47W611-63-2

1-47W611-63-5 Proposed references to

be provided: None Learning Objective: 3-OT-STG-074, THE RESIDUAL HEAT REMOVAL SYSTEM

4. EXPLAIN, the physical connections and/or cause-effect relationships between the RHR System and the following systems:
b. SIS d. RWST Cognitive Level:

Higher X Lower Question Source: New Modified Bank X Bank Question History: Modified from question 006 A4.05 32 used on the 5/2009 WBN NRC EXAM Comments:

QUESTIONS REPORTfor SQN-WBN composite exams

1. 006 A4.05 032Given the following plant conditions: - Safety Injection (SI) is actuated. - RWST level is 32%. - Containment sump level is 18%.Based on these conditions, it is expected that sump swapover will ______________.NOT occur since containment sump level is too low.NOT occur because an SI signal is still present to the RHR sump valves.NOT occur since RWST level is too high.

occur, since all required conditions have been met. A.B.C.D.Monday, May 04, 2015 7:22:55 AM 1 Figure 48 - Loop accuracy and settings for 1-LS-63-50 Figure 49 - Loop accuracy and settings for 1-LS-63-180

RWST LEVEL LOOPS CNTMT FLOOR LEVEL LOOPS I I I LS\ 63-SOA LS\ 63-180 1 I I I I I I LS\ 63-51A LS\ 63-181 ' I I I I I I I I I I I I LS \ LS\ 63-52A 63-53A LS \ LS\ 63-182 63-183 -.

  • I I I I I I I I I A 6C-126C I --I -..... ----------A 6C-127E 2/4 TRAIN A LEVEL < SETPOINT 2/4 TRAIN B Note that coincidence is 2 of 4. 2/4 TRAIN A LEVEL > SETPOINT 2/4 TRAIN B I TRAIN A TRIP SIGNAL TO FCV-63-72 (8811A) 1-47W611-63-5, B-2 TRAIN B TO FCV-63-73 (88118) 1-47W611-63-5, 0-4
34. Which ONE of the following describes the operation of the Primary Relief Tank (PRT)? The PRT is cooled by dispersing primary water into the PRT ____(1)____ the quench water level.

As primary water is added to the PRT, 1-FCV-68-310, PRT Drain to the RCDT ____(2)____ AUTOMATICALLY cycle to maintain PRT level. A. (1) ABOVE (2) WILL B. (1) ABOVE (2) WILL NOT C. (1) BELOW (2) WILL D. (1) BELOW (2) WILL NOT CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: Plausible because the Primary Water is dispersed above the water level but the RCDT tank pump B wil l not require a manual start (it automatically starts when 1-FCV-68-310 gets fully open.) The B. Correct: Primary Water for cooling is dispersed through a spray header above the water level and if draining of the PRT is required to lower the level, 1-FCV-68-310, PRT Drain to RCDT is opened by the OAC. When the valves gets fully open the RCDT B pump will automatic ally start. C. Incorrect: Plausible because there is a dispersion header below the water level (sparging header) to disperse steam entering the PRT and the RCDT tank pump B automatically starti ng when 1-FCV-68-310 gets fully open is correct. D. Incorrect: Plausible because there is a dispersion header below the water level (sparging header) to disperse steam entering the PRT and the RCDT tank pump B will not require a manual start (it automatically starts when 1-FCV-68-310 gets fully open.)

Question Number: 34 Tier: 2 Group: 1

K/A: 007 Pressurizer Relief Tank/Quench Tank System (PRTS) K4 Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following: K4.01 Quench tank cooling

Importance Rating: 2.6 2.9

10 CFR Part 55: (CFR: 41.7)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to relate the knowledge of the PRT design features for accomplishing one of its functions (maintain the tank cooled during a steam discharge). . Technical

Reference:

System Description N3-68-4001, Reactor Coolant System 1-47W813-1

1-47W611-68-1

Proposed references to

be provided: None Learning Objective: 3-OT-STG-068C, PRESSURIZER AND PRESSURIZER RELIEF TANK

10. EXPLAIN, the physical connections and/or cause-effect relationships between the Pressurizer Relief Tank (PRT) and the following:
f. Primary water
11. EXPLAIN the Pressurizer Relief Tank (PRT) design features and/or interlocks that provide the following:
a. Tank cooling
c. Makeup water Cognitive Level:

Higher Lower X Question Source: New Modified Bank Bank X Question History: WBN Bank question Comments:

1 II I CONT ON 1-47W830-1. COORD C-5 r -r "" "" "" CONT ON 1-47W625-8, COORD D AUTO GAS ANALYZER.,, I" ,,, ,,, DISC-68-PRT2 3/411 3/411

  • CONT ON 1-47' 4 II
  • 3/4 11 .. .-o ISC-68-PRT 1 68-57 4 CL ASS G r-------.....;..;;......,;..;...;..

_____ ...,1 68-57 2 1 11 x 2 11 The spray line is above the nominal water level. TE SPRAY LINE 19 9 9 9 L ___ sPA_Rc_ER TANK-68-PRT 68-309 - E-F..v. J TANK CLASS G 68-610 68-452A CLASS G CONT ON 1-47W819-1, COORD EPRIMARY WATER C) Ul Ul 211 < ...... TV 68-601 TV 1 1-68-559 CLASS G **

  • CLASS B u WBN System Description Document REACTOR COOLANT SYSTEM WBN-SDD-N3-68-4001 Rev. 0035 Page 77 of 259 3.2.5 Pressurizer Relief Tank (PRT) (continued)

Steam is discharged into the PRT through a sparger pipe under the water level. This condenses and cools the steam by mixing it with water that is near ambient temperature. The tank is equipped with an internal spray and a drain which are used to cool the tank following a discharge. The tank is protected against a discharge exceeding the design value by two rupture discs which discharge into the reactor containment. The tank is carbon steel with a corrosion-resistant coating on the wetted surfaces. A flanged nozzle is provided on the tank for the PZR discharge line connection. This nozzle and the discharge piping and sparger within the vessel are austenitic stainless steel. The tank design is based on the requirement to condense and cool a discharge of PZR steam equal to 110% of the volume above the full-power PZR water level set point. The tank is not designed to accept a continuous discharge from the PZR. The volume of water in the tank is capable of absorbing the heat from the assumed discharge. If the temperature in the tank rises above 128 °F during plant operation, the tank is cooled by spraying in cool water from the primary water system and draining out the warm mixture to the waste disposal system. The spray rate is designed to cool the tank from 200 °F to 120°F in approximately one hour following the design discharge of PZR steam. The volume of nitrogen gas in the tank is selected to limit the maximum pressure following a design discharge to 50 psig. The PRT rupture discs have a relief capacity equal or greater than the combined capacity of the PZR safety valves. The tank design pressure (and the maximum rupture disc blowout point) is twice the calculated pressure resulting from the maximum safety valve discharge. This margin is to prevent deformation of the disc. The rupture disc holders (15 psig minimum external pressure) are designed for full vacuum to prevent tank collapse if the contents cool following a discharge without nitrogen being added. Failure of the PRT does not affect the integrity of the RCPB nor does it affect the capability to shut down the plant safely. PRT spray for cooling of its contents is manually controlled by opening FCV-68-303 from the MCR. Discharge from the PRT to the WDS vent header is manually controlled by the opening of PCV-68-301 from the MCR. Draining the PRT to the RCDT is manually controlled by opening FCV-68-310 from the MCR. 3.2.6 RCS Valves All RCS valves are constructed primarily of stainless steel. Where stainless steel is not utilized, such as for hard surfacing and packing, other compatible materials are in contact

with the coolant. The (MOVs) uniqueness requirements are defined in Ref. 7.1.8. The RCS valves in Table 19 are classified as active per the following rules: a. Only WBN plant specific Design Basis Events (DBEs) that can be logically coupled with an SSE were considered. These DBEs were studied to identify the active pumps and valves required to mitigate the DBE and place the plant in a safe shutdown condition. I I

  • t gr FCV\ ,; , 0 -,, t-ao 68-303 FROM PRESSURIZER SEE z -r-\. -w 1-47W 611-68-J.

8-8 > PRIMARY WATER SEE , I I I I I I I I I I I I I I I I \ I I I I I I I I I I I J --FROM REACTOR VENT SEE F-5 ..,___ _____ _ RUPTURE DISC VALVES OPEN OPEN * -HS\ CLOSE I 68-310A NORMAL # ...... XS 68-310 AUX

  • HS CLOSE I 68-310C ' * , ' _ ....... _ ...... __ ....,._....,.

____________ .....,....,. .... -----------------------------------1 ... / ,, , ,, PRESSURIZER RELIEF TANK z UJ ..... a::o :::>a... VlUJ UJ(/) a:: I a.. v I PS ,l 68-301 "-' * --DWG 47W611-81-1 E-8 -P AUTO OPEN HS\ CLOSE I 68-301A I Notice that there is no level loop as soc iated with this v a Ive (i.e. there is OPEN HS CLOSE . 68-:SOlC I "" VENT FSV 68-310 FCV ,. 68-310 ' "-(8031) x no automatic level control associated with it). COMPANION DRAWINGS: DIAGRAM l1ssUED BY: *

35. Given the following conditions: - Unit 1 is at 100% power. - The following is observed:

Which ONE of the following comp letes the statements listed below?

The status of the Unit 1 "A" train CCS surge tank indicates that __________.

NOTE: 1-AOI-15, Loss of Com ponent Cooling Water (CCS) 0-AOI-10, Loss of Control Air A. conditions are NORMAL and that NO further actions are required B. Control air has been LOST to 1-LCV-70-63 and 1-FC V-70-66; 0-AOI-10 WILL be entered. C. CCS in-leakage is probable and se ction 3.4, CCS In-Leakage of 1-AOI-15 WILL be entered D. because the valve alignment for the CCS surge tank is correct, a leak exists on the Unit 1 "A" train CCS and section 3.3, CCS Out Leakage of 1-AOI-15 WILL be entered

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: The condition of the CCS surge tank is abnormal. Firstly, the level of the CCS surge tank is seen to be 57%. This level is abnormally low; this may be seen in the diagnostics for section 3.3, "CCS Out Leakage" of 1-AOI-15 which state: 

"Surge Tank level less than 60%

and dropping uncontrolled." Additi onally, annunciator 249-A, "U1 SURGE TANK LEVEL HI/LO," would be ex pected as its LO setpoint is 57%. Other abnormalities include t he fact that both the vent and makeup valves are closed. The vent valve is normally open (it will

close on a high radiation signal). The makeup valve is designed to open when level in the tank is less than 60%. B. Correct: As seen on print 1-47W 859-1, both 1-LCV 63 and 1-FCV-70-66 are air operated valves which fail closed upon either a loss of air or signal. One may observe in ARI 249-A the probable cause which states: "Surge Tank vent closed causing positive or negative Surge Tank

press; giving erroneous level indication." The aggregate of the indications seen in the stem show that control air has been lost to both of the mentioned valves. The vent valve caused surge tank level to indicate abnormally low and even though the level was below the makeup setpoint (of < 60%) the failed closed makeup valve was unable to provide any flow. C. Incorrect: The level is less than that normally expected. However, if one believed that normal level was less t han that shown (e.g. at mid-scale or 50%) then the indications would present a potential in-leakage to the CCS system. The diagnostics for section 3.4, "CCS In-Leakage" of 1-AOI-15 contain a check for CCS surge tank level high or rising uncontrolled OR a CCS rad monitor alarm present. Because the vent valve closes on a rad monitor alarm it is plausible to believe that such is present. Accordingly, if one misunderstood the normal surge tank level, then one would believe that in-leakage (e.g. from the thermal barriers of the RCPs) was occurring. D. Incorrect: Again, two valves are observed to be out of their expected alignment. However, if the valves were in the expected alignm ent, the indicated level would indicate a CCS inventory loss and the crew would implement section 3.3, "CCS Out Leakage" of 1-AOI-15.

Question Number: 35 Tier: 2 Group: 1 K/A: 008 Component Cooling Water System (CCWS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations: A2.05 Effect of loss of instrument and control air on the position of the CCW valves that are air operated

Importance Rating: 3.3 3.5 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to predict the impact to the makeup of the CCS surge tank by understanding the

failure position of the makeup level control valve (given that it loses control air) and then be able to utilize ARI-249-A to recover from the failure. Technical

Reference:

ARI-249-A, CCS 1-47W859-1 1-AOI-15, Loss of Component Cooling Water (CCS)

Proposed references to be provided: None Learning Objective: 3-OT-STG-070A, COMPONENT COOLING WATER SYSTEM 5. EXPLAIN, the physical connections and/or cause-effect relationships between the CCS and the following systems:

e. Radiation Monitoring System
f. Sources of Make-up WaterCognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: --J* VACUUM RELIEF' VALVE 1-70-539 The vent valve fails shut on a loss of air. A (UNIT 1) \2-70-510 1 1-70-356.

  • SET
  • 20 PSIG 1-1/4* 1.A I 19 1-70-1 . J55A ----.i--....._..,.......,.

_ _. 1-70-540 TANK DIVIDER {BAFFLE) The makeup ' valve fails shut on a loss of air. WBN Unit 1 Loss of Component Cooling Water (CCS) 1-AOI-15 Rev. 0006 Page 5 of 48 3.0 OPERATOR ACTIONS 3.1 Diagnostics IF GO TO Subsection PAGE LOSS OF CCS FLOW:

  • Pump trip
  • Multiple low flow alarm
  • Low header pressure 3.2 6 Surge Tank level less t han 60% and dropping uncontrolled OR Indications of out leakage.

3.3 11 Surge Tank level greater t han 72% or rising uncontrolled, OR CCS Rad Monitor alarm. 3.4 24 Loss of ERCW to CCS HX A 3.5 33 Loss of ERCW to CCS HX C 3.6 37 Loss of CCS while RHR Shutdown Cooling is in service 1-AOI-14

WBN Unit 0 CCS 0-ARI-241-253 Rev. 0000 Page 45 of 72 A. Leak into or out of CCS B. Makeup system malfunction C. Gas bubble in system D. Valve misalignment E. Surge Tank vent closed causing positive or negative Surge Tank press; giving erroneous level indication (if level in BOTH sides of tank is less than 60%, a false level may be indicated due to interior baffle design) NOTE Auto makeup should initiate at 60% lowering level. [1] MONITOR Surge Tank level. [2] IF level is low, THEN: [2.1] ENSURE 1-LCV-70-63, U1 SURGE TANK MAKEUP LCV, OPEN. [2.2] IF Surge Tank level lowers below 52% or 1-FCV-70-183, SAMPLE HX CCS OUTLET is CLOSED, THEN NOTIFY Chemistry to suspend sampling. [2.3] IF Surge Tank loss imminent, THEN GO TO AOI-15, LOSS OF COMPONENT COOLING WATER (CCS). [C.1] [2.4] IF level is NOT maintained due to loss of makeup, THEN CONSIDER installing ERCW spool piece for emergency source. [2.5] IF level is NOT maintained due to loss of air to makeup valve, THEN CONSIDER installing bottled gas on 1-LCV-70-63. Continued on Next Page Source Setpoint 1-LS-70-99A/B 85% 1-LS-70-99B/A 57% 249-AU1 SURGE TANK LEVEL HI/LO (Page 1 of 2) Probable Cause: Corrective Action:

36. Which ONE of the following describes t he required actions for a Loss of Component Cooling Water (CCS) with Unit 1 operating at 100% power; in accordance with 1-AOI-

15, Loss of Component Cooling Water?

The operating crew will initiate a Unit 1 reactor trip if the ____(1)____ CCS train is lost.

RCPs can be operated for a MAXIMUM of ____(2)____ minutes after loss of CCS flow.

(1) (2) A. 1A 10 B. 1A 12 C. 1B 10 D. 1B 12 CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: In accordance with 1-AOI-15, the reactor is tri pped upon the loss of the 1A CCS train (with t he Unit at power). This is seen in several instances within the contents of the AOI. For example, step 3. States: CHECK 1A Train flows NORM AL - e. TRIP Reactor. f. STOP RCPs. Additionally, it is correct that the RCPs can be operated for up to 10 minutes after loss of CCS flow. This fact is seen in the CAUTION in section 3.2 of 1-AOI-15 (prior to step 3.). B. Incorrect: While it is correct that the reactor is tripped upon the loss of the 1A CCS train, it is not correct that the RCPs can be operated for up to 12 minutes after loss of CCS flow. It is plausible to believe this because the same caution in 1-AOI-15 whic h describes the impact to the RCPs contains the following: CCP may survive for only 10 to 12 minutes after loss of CCS to lube oil cooler. C. Incorrect: It is Incorrect that the loss of 1B CCS train would cause the crew to initiate a reactor trip. It is plausib le to believe this because if 1B CCS train provided cooling to the RCPs then its loss would require tripping

the reactor (with the Unit at power). Additionally, it is correct that the RCPs can be operated for up to 10 mi nutes after loss of CCS flow. D. Incorrect: Again it is Incorrect but plaus ible that the loss of 1B CCS train would cause the crew to initiate a reactor trip. It is also, as mentioned, Incorrect and yet plausible that the RCPs could be operated for up to 12 minutes after loss of CCS flow.

Question Number: 36 Tier: 2 Group: 1

K/A: 008 Component Cooling Water System (CCWS) 2.4 Emergency Procedures / Plan 2.4.11 Knowledge of abnormal condition procedures. Importance Rating: 4.0 4.2

10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to utilize the guidance provided in 1-AOI-15 to ascertain which CCS train's loss will precipitate a reactor trip. Furthermore, the applicant must recall the guidance provided in the same AOI to the operational impact to the RCPs given that a loss of their CCS occurs. Technical

Reference:

1-AOI-15, Loss of Component Cooling Water (CCS) Proposed references to be provided: None Learning Objective: 3-OT-STG-070A, COMPONENT COOLING WATER SYSTEM

5. EXPLAIN, the physical connections and/or cause-effect relationships between the CCS and the following systems:
a. RCP Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Loss of Component Cooling Water (CCS) 1-AOI-15 Rev. 0006 Step Action/Expected Response Response Not Obtained Page 6 of 48 3.2 Loss of CCS Flow

1. CHECK at least one U-1 Train A header supply pump RUNNING AND pumping forward:
  • 1A-A
  • 1B-B START available U-1 Train A CCS pump. 2. IF loss of CCS is due to a loss of AC power train, THEN REFER to AOI-43 series and continue with this procedure.

CAUTIONS

  • CCP may survive for only 10 to 12 minutes after loss of CCS to lube oil cooler.
  • RCPs can be operated for up to 10 minutes after loss of CCS flow.
3. CHECK 1A Train flows NORMAL
  • CCS Low Flow alarms on 0-M-27 CLEAR. IF 1A Train CCS flow lost, THEN PERFORM the following:
a. ENSURE CCP 1B-B Running THEN STOP and LOCKOUT CCP 1A-A.
b. INITIATE alignment of ERCW to CCP 1A-A lube oil heat exchanger

USING Attachment 1 (may use placard posted locally in CCP room

1A-A). Step continued on next page

WBN Unit 1 Loss of Component Cooling Water (CCS) 1-AOI-15 Rev. 0006 Step Action/Expected Response Response Not Obtained 3.2 Loss of CCS Flow (continued) Page 7 of 48 Step 3 continued.

c. ISOLATE letdown and charging.
1) CLOSE Orifices valves
2) CLOSE 1-FCV-62-69 & 70
3) CLOSE 1-FCV-62-90 & 91
4) ENSURE excess letdown 1-FCV-62-54 & 55 isolated.
d. STOP and LOCKOUT the following pumps:
  • TBBPs 1-A & 1-B,
  • CCS pumps 1A-A & 1B-B (If aligned to A Trn)
  • CS pump 1A-A,
  • SI pump 1A-A,
  • RHR pump 1A-A,
  • CCP 1A-A (IF ERCW not aligned and greater than 10 min has elapsed)
e. TRIP Reactor.
f. STOP RCPs. g. **GO TO 1-E-0, Reactor Trip or Safety Injection,

WHILE CONTINUING this Instruction.

37. Which ONE of the following describes the power supply to the 1C and 1D PZR heaters?

A. 480V Unit Boards B. 6.9kV Unit Boards C. 480V Shutdown Boards D. 6.9kV Shutdown Boards

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: It is not correct that the 1C and 1D heater groups are powered from the 480V Unit Boards. Two items lend plausibility to t he belief that the Unit Boards would power the 1C and 1D heater groups. First, as seen in the labeling for these heater groups, there is NO train designation (i.e. a -A or -B). This would lead one to believe that the component is NOT safety related and as such NOT powered from shutdown power. Additionally, as seen in T/S LCO 3.4. 9, Pressurizer: The pressurizer shall be OPERABLE with:

b. Two groups of pressurizer heaters OPERABLE with the capacity of eac h group > 150kW. Because the Unit's T/S do not require that four groups of heaters be operable, it is plausible to believe that only tw o would receive shutdown power. Also, it is plausible to believe t hat because the pre ssurizer heaters are actually 480VAC heater s that they would be powered from a 480VAC board. B. Incorrect: It is not correct that the 1C and 1D heater groups are powered from the 6.9kV Unit Boards. It is plausible to believe this for a reason tantamount to the "A" distractor with the exception that an applicant may understand the voltage of the s upply to the heaters is 6.9kV.

C. Incorrect: While it is correct that shutdown power supplies the 1C and 1D heaters, it is not correct that the supply originat es from a 480V shutdown board. It is plausib le to believe that because the

pressurizer heaters are actually 480V AC heaters that they would be powered from a 480VAC board. D. Correct: It is correct that the 1C backup group and 1D control group of pressurizer heaters are powered from the 6.9kV Shutdown Boards. Question Number: 37 Tier: 2 Group: 1

K/A: 010 Pressurizer Pressure Control System (PZR PCS) K2 Knowledge of bus power supplies to the following: K2.01 PZR heaters Importance Rating: 3.0 3.4

10 CFR Part 55: (CFR: 41.7) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to possess the knowledge that ALL of the groups of pressurizer heaters are powered from the 6.9kV shutdown boards. Technical

Reference:

1-45W724-1 1-45W724-2 Unit 1 Technical Specifications Proposed references to

be provided: None Learning Objective: 3-OT-STG-068C, PRESSURIZER AND PRESSURIZER RELIEF TANK

5. LIST the power supplies to the following Pressurizer

Level Control System components:

a. PZR Heaters Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: Pressurizer 3.4.9 Watts Bar-Unit 1 3.4-18 3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.9 Pressurizer

LCO 3.4.9 The pressurizer shall be OPERABLE with:

a. Pressurizer water level 92%; and
b. Two groups of pressurizer heaters OPERABLE with the capacity of each group 150 kW.

APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water level not within limit. A.1 Be in MODE 3 with reactor trip breakers open. AND A.2 Be in MODE

4. 6 hours 12 hours B. One required group of pressurizer heaters inoperable.

B.1 Restore required group of pressurizer heaters to OPERABLE status. 72 hours C. Required Action and associated Completion Time of Condition B not met. C.1 Be in MODE

3. AND C.2 Be in MODE
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38. Which ONE of the following describes the EAGLE 21 system response to a Channel I input failure and the effect on the system's ability to calculate a trip setpoint?

If the channel I input for ____(1)____ FAILS , EAGLE-21 will be unable to calculate a VALID setpoint for the channel I ____(2)____ trip. (1) (2) A. the PRNI OPT B. Loop 1 Flow OPT C. PZR Pressure OTT D. Turbine Impulse Pressure OTT CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: The OPT trip uses the T of each loop as a measure of reactor power with a setpoint that is automatically varied with reactor coolant average temperature and the rate of change of reactor coolant average temperature. These facts may be found in multiple locations; provided as supporting material is the T/S basis discussion for T/S LCO 3.3.1. It is plausible to believe that the loss of a channel I PRNI is detrimental to the ability of the OPT calculator because as seen in Note 2: Overpower T of table 3.3.1-1 (of the Unit's T/S), the setpoint calculation contains the term -f 2 (I). Delta Flux (I) is produced from the appropriate Power Range Nuclear Instrument (PRNI). As applied at WBN, delta flux (and thus, PRNI) is not included in the OPT calculation because as seen in the Note 2: f 2 (I)=0 for all f 2I. Incidentally, it is a common misconception that OPT utilizes the PRNIs as the name overpower implies that it uses power as an input. B. Incorrect: As seen in the basis for the Unit's T/S (again these facts may be found in multiple locations), the OTT setpoint is automatically calculated from reactor coolant average temperature, pressurizer pressure and axial power distribution f(I). The inputs to OTT may also be seen in print 1-47W610-68-8. Also given by this reference (and multiple others) is that The Over

temperature T trip Function is provided to ensure that the design limit DNBR is met. Therefore, it is plausible to believe that a RCS parameter which does impact the DNBR (namely RCS flow) would input into the OTT setpoint. However, the OTT setpoint assumes normal full loop flow. This is one of the reasons why the Ts present in the RCS must be verified at the beginning of a operation cycle. C. Correct: As described pressurizer pr essure is an input which affects the setpoint for the OTT trip. D. Incorrect: As mentioned OPT uses only temperatures as inputs (even though its name is over power). It is plausible that it would utilize an indication of power as an input (i.e. Turbine Impulse Pressure). In the case of channel I, 1-R-2 (Unit 1's rack 2) of the EAGLE-21 protective system, calculates the setpoints for both OPT and OTT. This fact is seen on print 1-47W610-99-1. 1-R-2 is included in the collection of channel I racks collectively known as protection set I. One of the channel I (or protection set I) racks, 1-R-4, receives turbine impulse pressure and uses such for a protective system function. This fact is seen on print 1-47W610-99-2. Therefore, it is additionally plausible that a channel I rack would use turbine impulse pressure for the production of the OPT setpoint.

Question Number: 38 Tier: 2 Group: 1

K/A: 012 Reactor Protection System A1 Ability to predict and/or monitor Changes in parameters (to prevent exceeding design limits) associated with operating the RPS controls including: A1.01 Trip setpoint adjustment

Importance Rating: 2.9* 3.4*

10 CFR Part 55: (CFR: 41.5 / 45.5)

10CFR55.43.b: Not applicable

K/A Match: K/A is matched because the applicant is required to have knowledge of the parameters associated with the RPS and then be able to monitor the correct parameter predicting what its failure will have upon a RPS setpoint. The failures described which if unresolved would place the plant outside of design limits. Technical

Reference:

Unit 1 Technical Specifications Unit 1 Technical Specifications Basis 1-47W610-68-8 1-47W610-99-1

1-47W610-99-2

Proposed references to be provided: None Learning Objective: 3-OT-TS0303, TS 3.3 "INSTRUMENTATION" 1. DESCRIBE the following aspects of Technical Specifications and Technical Requirements for Instrumentation Tech Specs:

a. The Limiting Conditions for Operation, Applicability, and Bases for the LCO. b. The conditions and required actions with completion

time of one hour or less. Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: RTS Instrumentation 3.3.1 Watts Bar-Unit 1 3.3-21 Amendment 7 Table 3.3.1-1 (page 7 of

9) Reactor Trip System Instrumentation Note 1: Overtemperature T The Overtemperature T Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.2% of T span.

Where: T is measured RCS T, °F. T O is the indicated T at RTP, °F. s is the Laplace transform operator, sec -1. T' is the indicated T avg at RTP, P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, 2235 psig K 1 1.16 K 2 0.0183/°F K 3 = 0.000900/psig 1 33 sec 2 4 sec 4 3 sec 5 3 sec f 1 (I) = -2.62{22 + (q t - q b)} when q t - q b < - 22% RTP 0 when -22% RTP q t - q b 10% RTP 1.96{(q t - q b) - 10} when q t - q b > 10% RTP Where q t and q b are percent RTP in the upper and lower halves of the core, respectively, and q t + q b is the total THERMAL POWER in percent RTP.

RTS Instrumentation 3.3.1 Watts Bar-Unit 1 3.3-22 Amendment 7 Table 3.3.1-1 (page 8 of

9) Reactor Trip System Instrumentation Note 2: Overpower T The Overpower T Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.0% of T span. Where: T is measured RCS T, F. T O is the indicated T at RTP, F. s is the Laplace transform operator, sec

-1. T is the measured RCS average temperature, F. T" is the indicated T avg at RTP, 588.2 F. K 4 1.10 K 5 0.02/F for increasing T avg K 6 0.00162/°F when T > T " 0/F for decreasing T avg 0/°F when T T" 3 5 sec 4 3 sec 5 3 sec f 2 (I) = 0 for all I.

  • '°'I --.. I 11\
  • N T-1' I I tj -,* -*-. -. -. . --(RTD CURRENT) 'J A/D A/O 1 I DFP-99-0?13 0 211 .. CALC POWER RANGE CHECK L , N-411A QL TRIP DELAY .. CALC FLUX IMBALANCE f 1 .. *411 UXIMB f 1 --r -+ SUM N-411 CALCULATE DELTA I CAL .. A/D I UI" t"-99-0213 CALC POWER RANGE CHECK N-4118 QU DELTA I N-411 DELTA I CAL CALC FLUX IMBALANCE f 2 l{'boo--N-411 I D/A DFP-99-0? 1? CALC PRESS. RANGE CHECK ' UFP-99-02 CALC T l .RANGE Cl P-4550 PZR PRESS T-4118 NR TCOLDl "' -T-412

-T5-68-2K "' TS-68-2J .. DEENG DECR DEENG DECR 1+T1S 1+T2S T-411 TAVGLIL j 1.===f .. CAL C OVER-TEMP SP ' ' N-411 FLUXIMB f 1 .. 1 1 T* OT T-412 ,------------------ TAVG

  • T3S 1 +TJS FLUXIMB f 2 The three inputs to OT delta Tare Temperature, Pressure and Delta I. T-411 TAVGR/l "' CALCULATE 1------------

...... ---------t-OVERPOWER SF.J---.. ,C-D/A DAC-J?15 99-0215 O/A I I __ T-411 DAC-99-0? 15 I OPSP LO TAVG TB-412G DID DDC-99-0214 I LO-LO TAVG TB-4120 D/A T-411 OTSP DAC-99-n? 15 I DID TS-68-DEENG I 0 T uoc-99-021 I RTS Instrumentation B 3.3.1 Bases (continued) Watts Bar-Unit 1 B 3.3-16 Revision 17 APPLICABLE

5. Source Range Neutron Flux (continued) SAFETY ANALYSES, LCO, and In MODE 3, 4, or 5 with the reactor shut down, the Source Range APPLICABILITY Neutron Flux trip Function must also be OPERABLE. If the CRD System is capable of rod withdrawal, the Source Range Neutron Flux trip must be OPERABLE to provide core protection against a rod withdrawal accident.

If the CRD System is not capable of rod withdrawal, the source range detectors are not required to trip the reactor. However, their monitoring Function must be OPERABLE to monitor core neutron levels and provide visual indication and audible alarm of reactivity changes that may occur as a result of events like a boron dilution. The requirements for the NIS source range detectors in MODE 6 are addressed in LCO 3.9.3, "Nuclear Instrumentation."

6. Overtemperature T

The Overtemperature T trip Function is provided to ensure that the design limit DNBR is met. This trip Function also limits the range over which the Overpower T trip Function must provide protection. The inputs to the Overtemperature T trip include pressurizer pressure, coolant temperature, axial power distribution, and reactor power as indicated by loop T assuming full reactor coolant flow. Protection from violating the DNBR limit is assured for those transients that are slow with respect to delays from the core to the measurement system. The Function monitors both variation in power and flow since a decrease in flow has the same effect on T as a power increase. The Overtemperature T trip Function uses each loop's T as a measure of reactor power and is compared with a setpoint that is automatically varied with the following parameters:

reactor coolant average temperaturethe Trip Setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature;

pressurizer pressure the Trip Setpoint is varied to correct for changes in system pressure; and

RTS Instrumentation B 3.3.1 Bases (continued) Watts Bar-Unit 1 B 3.3-17 Revision 13 Amendment 7 APPLICABLE

6. Overtemperature T (continued) SAFETY ANALYSES, LCO, and axial power distribution - the f(I) Overtemperature T APPLICABILITY Trip Setpoint is varied to account for imbalances in the axial power distribution as detected by the NIS upper and lower power range detectors. If axial peaks are greater than the design limit, as indicated by the difference between the upper and lower NIS power range detectors, the Trip Setpoint is reduced in accordance with Note 1 of Table 3.3.1-
1.

Dynamic compensation is included for delays associated with fluid transport from the core to the loop temperature detectors (RTDs), and thermowell and RTD response time delays. T 0, as used in the Overtemperature and Overpower T trips, represents RTP Tavg value as measured by the plant for each loop. T 0 T setpoint to the actual operating conditions existing at the time of measurement, thus forcing the setpoint to reflect the equivalent full power conditions as assumed in the accident analyses. Differences in RCS loop T and Tavg can be due to several factors, e.g., measured RCS loop flow greater than minimum measured flow, and slightly asymmetric power distributions between quadrants. While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific T and Tavg values. Loop specific values of T 0 must be determined at the beginning of each fuel cycle at full power, steady-state conditions (i.e., power distribution not affected by xenon transient conditions) and will be checked quarterly and updated, if required. Tolerances for T 0 determination of the Overtemperature T setpoint.

The Overtemperature T trip Function is calculated for each loop as described in Note 1 of Table 3.3.1-1. Trip occurs if Overtemperature T is indicated in two

RTS Instrumentation B 3.3.1 Bases (continued) Watts Bar-Unit 1 B 3.3-18 Revision 13 Amendment 7 APPLICABLE

6. Overtemperature T (continued) SAFETY ANALYSES, LCO, and loops. The pressure and temperature signals are used for other control APPLICABILITY functions. The actuation logic must be able to withstand an input failure to the control system, which may then require the protection function actuation, and a single failure in the other channels providing the protection function actuation. Note that this Function also provides a signal to generate turbine runback prior to reaching the Trip Setpoint. A turbine runback will reduce turbine power and reactor power. A reduction in power will normally alleviate the Overtemperature T condition and may prevent a reactor trip.

The LCO requires all four channels of the Overtemperature T trip Function to be OPERABLE. Note that the Overtemperature T Function receives input from channels shared with other RTS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.

In MODE 1 or 2, the Overtemperature T trip must be OPERABLE to prevent DNB. In MODE 3, 4, 5, or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient heat production to be concerned about DNB.

7. Overpower T

The Overpower T trip Function ensures that protection is provided to ensure the integrity of the fuel (i.e., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions. This trip Function also limits the required range of the Overtemperature T trip Function and provides a backup to the Power Range Neutron FluxHigh Setpoint trip.

RTS Instrumentation B 3.3.1 Bases (continued) Watts Bar-Unit 1 B 3.3-19 Revision 13 Amendment 7 APPLICABLE

7. Overpower T (continued) SAFETY ANALYSES, LCO, and The Overpower T trip Function ensures that the allowable heat APPLICABILITY generation rate (kW/ft) of the fuel is not exceeded. It uses the T of each loop as a measure of reactor power with a setpoint that is automatically varied with the following parameters:

reactor coolant average temperature the Trip Setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature; and

rate of change of reactor coolant average temperature including dynamic compensation for delays associated with fluid transport from the core to the loop temperature detectors (RTDs),and thermowell and RTD response time delays.

T 0, as used in the Overtemperature and Overpower T trips, represents 100% RTP Tavg value as measured by the plant for each loop. T 0 and T setpoint to the actual operating conditions existing at the time of measurement, thus forcing the setpoint to reflect the equivalent full power conditions as assumed in the accident analyses. Differences in RCS loop T and Tavg can be due to several factors, e.g., measured RCS loop flow greater than minimum measured flow, and slightly asymmetric power distributions between quadrants. While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific T and Tavg values. Loop specific values of T 0 must be determined at the beginning of each fuel cycle at full power, steady-state conditions (i.e., power distribution not affected by xenon transient conditions) and will be checked quarterly and updated, if required. Tolerances for T 0 determination of the Overtemperature T setpoint.

RTS Instrumentation B 3.3.1 Bases (continued) Watts Bar-Unit 1 B 3.3-20 APPLICABLE

7. Overpower T (continued) SAFETY ANALYSES, LCO, and The Overpower T trip Function is calculated for each loop as per APPLICABILITY Note 2 of Table 3.3.1-1. Trip occurs if Overpower T is indicated in two loops. The temperature signals are used for other control functions.

Therefore, the actuation logic must be able to withstand an input failure to the control system, which may then require the protection function actuation and a single failure in the remaining channels providing the protection function actuation. Note that this Function also provides a signal to generate a turbine runback prior to reaching the Trip Setpoint. A turbine runback will reduce turbine power and reactor power. A reduction in power will normally alleviate the Overpower T condition and may prevent a reactor trip.

The LCO requires four channels of the Overpower T trip Function to be OPERABLE. Note that the Overpower T trip Function receives input from channels shared with other RTS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.

In MODE 1 or 2, the Overpower T trip Function must be OPERABLE. These are the only times that enough heat is generated in the fuel to be concerned about the heat generation rates and overheating of the fuel.

39. Which ONE of the following descri bes the power supply to a Reactor Trip Breaker UV Trip Coil?

A Reactor Trip Breaker UV Trip Coil power is supplied by a ________ Board. A. 48V-DC Plant Battery B. 125V-Vital DC Battery C. 120V-AC Preferred Power D. 120V-AC Vital Instrument Power

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: Plausible since 48 VDC power is used for the UV relays (This 48V power supply provides power for the RPS UV relays which cause a Reactor Trip (52 UV relay in breaker trip logic)) however this supply provides power for plant phone system. B. Incorrect: Plausible since 125 VDC power is the control power to the Reactor Trip breakers control power. C. Incorrect: Plausible since 120 VAC is us ed in RPS logic, howev er it is 120 VAC vital Instrument power that used for RPS. 120 VAC-Preferred power provides power primarily to secondary 120V control functions D. Correct: 120V AC is converted to 48V DC in the logic cabinets. This 48V power supply provides power for the RPS UV relays which cause a Reactor Trip (52 UV relay in breaker trip logic)

Question Number: 39 Tier: 2 Group: 1

K/A: 012 Reactor Protection System K2 Knowledge of bus power supplies to the following: K2.01 RPS channels, components, and interconnections

Importance Rating: 3.3 3.7

10 CFR Part 55: (CFR: 41.7)

10CFR55.43.b: Not applicable

K/A Match: This question matches the K/A by having testing the knowledge of power supply to portions of the RPS and which of those power supplies are used to cause RTA &/or RTB to trip.

Technical

Reference:

1-45W600-99-1 Proposed references to

be provided: None Learning Objective: 3-OT-SYS085A, Control Rod Drive System 13. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on the Rod Control or Rod Position Indicating Systems: e. Loss of control power reactor trip breakers Cognitive Level: Higher Lower X Question Source: New Modified Bank Bank X Question History: Bank question used on the SQN 2/2010 exam. Comments:

40. Given the following conditions: - Unit 1 is at 100%. - The following is observed:

Which ONE of the following describes the response of SSPS?

SSPS will ________. A. initiate a Reactor Trip ONLY B. initiate a Safety Injection ONLY C. NOT initiate ANY actuation signals D. initiate BOTH a Reactor Trip AND a Safety Injection CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: Reactor trip occurs due to 2/4 channels failing low (322 had failed low as given in the stem of the question and now a subsequent channel fails low). No safety injection will occur as it uses channels I,II and III. The SI logic will see 1 of 3 channels meeting the initiation criteria. B. Incorrect: Reactor trip occurs due to 2/4 channels (322 had failed low as given in the stem of the question). Plausible since, if the original failure had been a different channel (II or III) it would have caused an SI. If the swaps the logic between the safety in jection and the reactor trip, that individual may believe that a safety injection could occur but a rx trip would not. C. Incorrect: As mentioned, the reactor trip logic would be met (2/4 channels). If the applicant believes that both the trip and SI logics were 2 of 3 (using channels I, II and III) then they would a rrive at this distractor as the correct. D. Incorrect: This distractor is incorrect and yet plausible given the assumption that both the trip and safety injection logics were 2 of 4.

Question Number: 40 Tier: 2 Group: 1

K/A: 012 Reactor Protection System K6 Knowledge of the effect of a loss or malfunction of the following will have on the RPS: K6.01 Bistables and bistable test equipment

Importance Rating: 2.8 3.3

10 CFR Part 55: (CFR: 41.7 / 45/7)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to understand the impact which is had on the protective system of two different Eagle bistables. Technical

Reference:

1-47W611-99-2 1-47W611-63-1 Proposed references to

be provided: None Learning Objective: 3-OT-TS0303, TS 3.3 "INSTRUMENTATION" 1. DESCRIBE the following aspects of Technical Specifications and Technical Requirements for Instrumentation Tech Specs:

a. The Limiting Conditions for Operation, Applicability, and Bases for the LCO. b. The conditions and required actions with completion time of one hour or less. Cognitive Level:

Higher X Lower Question Source: New Modified Bank X Bank Question History: Modified from Bank question 013 K5.02 39 which was used on the SQN 09/2010 NRC exam. Comments:

QUESTIONS REPORTfor SQN-WBN composite exams

1. 013 K5.02 039Given the following plant conditions: - Unit 1 is at 100% power. - Pressurizer pressure Channel I, 1-PT-68-340, has been removed from service for surveillance testing with its associated bistables tripped. - Pressurizer pressure Channel IV, 1-PT-68-322, fails high.

Which ONE of the following describes the immediate result of these conditions?Both Reactor trip and Safety Injection occur.PZR PORV, 1-PCV-68-340 opens; PZR PORV, 1-PCV-68-334 opens.A Reactor trip but NO Safety Injection occurs.PZR PORV, 1-PCV-68-340 opens; PZR PORV, 1-PCV-68-334 remains closed.Both Reactor trip and Safety Injection occur.PZR PORV, 1-PCV-68-340 remains closed; PZR PORV, 1-PCV-68-334 opens.A Reactor trip but NO Safety Injection occurs.PZR PORV, 1-PCV-68-340 remains closed; PZR PORV, 1-PCV-68-334 remains

closed.A.B.C.D.Friday, May 15, 2015 1:50:02 PM 1 LOG P0492D I --I' V"T/.JU! STEAM LINE ISOLATION SIGNAL SEE 1-47W611-1-1 PRESSURIZER PRESSURE GREATER THAN NORMAL SETPOINT PS-68-3408 LOG II PRESSURIZER PRESSURE GREATER THAN NORMAL SETPOINT PS-68-3348 P0493D LOG P0494D The low pressure SI uses 2 of 3 coincidence. --e-----1 .... I I I PRESSURIZER PRESSURE GREATER THAN NORMAL SETPOINT PS-68-3238 2/3 HS BLOCK 63-136A OTHER LOGIC TRAIN "" 4A-698 A -----TO SIS ACCUMULATOR TANKS 1-4 SHEET 6 7_ 4A-69A A 4 P-11 LOG P0495D TRAIN A LOG P0496D TRAIN B PRESSURIZER LOG P0491D LOG P04890 LOG P0490D 6C-124DA l ... ' A 2/3 4D-77G

  • F PRESSURE -I I I I II LOG Y04800 LO PRESSURIZER PRESSURE PS-68-3230 LO PRESSURIZER PRESSURE PS-68-3400 LO PRESSURIZER PRESSURE PS-68-3340 I 6C-124B PRESSURIZER HIGH PRESSURE LOG P0480D II LOG P0481D 2/4 I I I PS 68-323A LOG P0482D nz: PS 68-322A LOG P04970 6C-124C A The High and Low Pzr Pressure trips use 2 of 4 coincidence.

40-77C LOG P0483D PRESSURIZER HIGH PRESSURE REACTOR TRIP 1-47W611-99-1 I PS 68-340E PRESSURIZER LOW PRESSURE {FIXED SET POINT LEAD/LAG COMPENSATION) LOG :0: PS 68-334E LOG I I I PS 68-323E LOG P04840 P0485D P0486D I 40-770 ...._L_OG_.....,._ ____ ...... A P0488D PRESSURIZER LOW PRESSURE REACTOR TRIP 1-47W611-99-1 DZ: PS 68-322E LOG P0487D QUESTIONS REPORTfor SQN-WBN composite exams

1. 013 K5.02 039Given the following plant conditions: - Unit 1 is at 100% power. - Pressurizer pressure Channel I, 1-PT-68-340, has been removed from service for surveillance testing with its associated bistables tripped. - Pressurizer pressure Channel IV, 1-PT-68-322, fails high.

Which ONE of the following describes the immediate result of these conditions?Both Reactor trip and Safety Injection occur.PZR PORV, 1-PCV-68-340 opens; PZR PORV, 1-PCV-68-334 opens.A Reactor trip but NO Safety Injection occurs.PZR PORV, 1-PCV-68-340 opens; PZR PORV, 1-PCV-68-334 remains closed.Both Reactor trip and Safety Injection occur.PZR PORV, 1-PCV-68-340 remains closed; PZR PORV, 1-PCV-68-334 opens.A Reactor trip but NO Safety Injection occurs.PZR PORV, 1-PCV-68-340 remains closed; PZR PORV, 1-PCV-68-334 remains

closed.A.B.C.D.Monday, May 04, 2015 12:23:35 PM 1

41. Given the following conditions: - Unit 1 is at 100% power. - ALL Containment Cooling is FULLY Operational in accordance with the SOIs.

Which ONE of the following descr ibes an event which causes the following containment pressure transient?

A. Starting an additional lo wer containment cooler B. A RCS leak inside of containment C. Securing one of the running lower containment coolers D. Instrument Air leak inside containment which does NOT result in the isolation of instrument air to containment CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: If an additional lower containm ent cooler is started, the available cooling capacity available to lower containment is increased. The cooler will immediately act to lower the temperature of lower containment and because of the effe ct which can be approximated by the ideal gas law, will lower pressure slightly. The originally running LCCs will sense the lowered temperat ure and will reduce their cooling. Therefore, the temperature will return to that originally seen and as such the pressure will return to that originally seen. The LCCs may be simplified as a re circulation fan which draws air across a cooling coil. If one believe that the LCCs drew air from somewhere other than the containm ent atmosphere, then they could reasons that the small pressure spike was due to this initial increased admission (until the other LCCs reduced their air intake flow).

Also, one may understand that the LCCs discharge air at a velocity

which will literally blow a person over. Because of this, it is plausible to believe that starting this very power blower will initially cause a

pressure spike. B. Incorrect: A RCS leak inside of containment will cause containment pressure to rise. If the leak is small enough, containment pressure will rise and establish a higher equilib rium value. The equili brium value is one in which the energy input into the co ntainment atmosphere is in balance with the heat removal capability of the containment coolers. This distractor is plausible if the app licant believes that the containment coolers (through automatic action of their temperature control valves) will cause containment pressure to turn and return to near 0 psig. C. Correct: When the Lower Containment Cool er A-A is secured, overall cooling to containment is reduced. The te mperature and pr essure of the containment will rise. The rise in temperature will cause the other

running LCCs' Temperature control valves to open further. The remaining running LCCs will rest ore containment temperature and pressure to their values bef ore the securing of the LCC. D. Incorrect: An instrument air break inside of containment will cause containment pressure to rise. It would be plausible to believe that this distractor is correct if air had isolated (on a low ai r header pressure). If the break were large enough to cause the applicable reactor compartment air header to isolate, then the observed containment pressure trend would

be similar to that shown in the ti me plot. Pressure would rise and subsequently be restored through the actions of both the containment pressure reliefs and the annulus vacuum subsystem.

Question Number: 41 Tier: 2 Group: 1

K/A: 022 Containment Cooling System (CCS) A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: A1.02 Containment pressure

Importance Rating: 3.6 3.8

10 CFR Part 55: (CFR: 41.5 / 45.5)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to interpret a time plot of containment pressure (monitor this parameter) and determine the initiating cause of the time plot. Thus, the applicant demonstrates the capability to understand the impact on containment pressure when securing (or operating) a Lower Containment Cooler. Technical

Reference:

Proposed references to

be provided: None Learning Objective: 3-OT-STG-030C, REACTOR BUILDING VENTILATION SYSTEM

7. EXPLAIN the Reactor Building Ventilation System design features and/or interlocks that provide the following: h. Temperature Indicating Controller settings for control of TCVs Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

42. Regarding the following hand-switches:

Located on 1-M-5 Located on 1-M-6 Which ONE of the following describes the logic for a MANUAL actuation of a phase B Isolation?

A. The operation of ANY ONE of the 4 hand-switches will actuate BOTH trains B. The SINGLE operation of either 1-HS-30-64A or 1-HS-30-68A will actuate ONLY the A train C. The SEQUENTIAL operation of one pair of hand-switches (either those on 1-M-5 or those on 1-M-6) will actuate BOTH trains D. The SIMULTANEOUS operation of one pair of hand-switches (either those on 1-M-5 or those on 1-M-6) will actuate BOTH trains CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: It is not correct that the oper ation of one of the f our hand-switches will produce a phase B signal (which in turn yields the containment spray actuation). The plausibility for this distractor is seen if the applicant believes that the manual actuation of phase B is similar to the manual actuation of phase A. To manually actuate phase A, the operator would take either of the two avai lable hand-switches to ACTUATE. This fact may be observed on print 1-47W611-88-1. Also seen on this print is the operation of the manual phase B signal. Phase B requires that two hand-switches be simultaneously taken to Actuate. The fact that the switches be simultaneously positioned is indicated by the arrow appearing above and to the left of the switch symbol. If one switch were taken and then released and then the other taken and then released, the AND circuit downstr eam of the switc hes would at no

time receive a signal of other t han one 0 and one 1. Additionally, the pair of switches operated must be in the same pair. There are two

pairs of switches. One pair is on 1-M-5 and one is on 1-M-6. One could not, as seen in print 1-47W 611-88-1 simultaneously operate one

of the switches on 1-M-5 and one of the switches on 1-M-6 to produce a phase B. B. Incorrect: Again it is Incorrect that the operation of only one switch would actuate a single train of phase B. The plausibility of this distractor is that this is exactly how two of the other isolation signals function. Specifically, the

Control Room Isolation signal has two pairs of hand-switches. One

pair is on panel 1-M-6 and the other pair is on panel 2-M-6. Either pair of hand-switches contains an A train and a B train switch. For example the Control Room Isolation switches on 1-M-6 includes 1-HS-31-177A, MCR ISOL TR-A and the switches on 2-M-6 includes 2-HS-31-177A, MCR ISOL TR-A. If either of the two hand-switches (assuming U2 were in operation) were taken to ACTUATE, then a CRI would occur. The other isolation which functions in the manner described above is the Auxiliary Building Isolation. It too has a pair of switches on 1-M-6 and a pair on 2-M-6. By design, the operation of an A train switch on either Unit would cause the appropriate train of ABI to occur. Another method by which th is distractor is plausible is the fact that isolation signal rese ts operate exactly in the manner described above. C. Incorrect: As previously discussed, t he sequential operation of a pair of phase B switches will not cause an actuation. The plausibility of this distractor

originates not only from the fact the design of phase B could certainly have stipulated that a sequential operation of two hand-switches be utilized for the signals actuation. This design would certainly have befit the human factors design basis that only single handed operations be allowed. Also, the distractor is plausible as the sequential operation of the Safety Injection reset pushbuttons is required to produce the desired a nnunciator (and system) response. Take, for example, step 12 of 1-E-1, Loss of Reactor or Secondary Coolant which states: RESET SI AND CHECK the following: SI ACTUATED permissive DARK AUTO SI BLOCKED permissive LIT. To achieve this step (which appears throughout the Emergency procedure set), the operator will push t he A train SI reset button, check that the Annunciator 70-B, AUTO SI BLOCKED is LIT, push the B train SI reset button, and finally check that Annunciator 70-A, SI ACTUATED is DARK. This sequential operation of safeguards switches is practiced time and time again. D. Correct: As discussed the simultaneous operation of either pair of hand-switches will cause the phase B actuation. Question Number: 42 Tier: 2 Group: 1

K/A: 026 Containment Spray System (CSS) A4 Ability to manually operate and/or monitor in the control room: A4.01 CSS controls

Importance Rating: 4.5 4.3

10 CFR Part 55: (CFR: 41.7 / 45.5 to 45.8)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to demonstrate the capability to operate the phase B controls in the main control room. The phase B controls are those which initiate a containment spray signal. Technical

Reference:

1-47W611-88-1 Proposed references to

be provided: None Learning Objective: 3-OT-TS0303, TS 3.3 "INSTRUMENTATION" 1. DESCRIBE the following aspects of Technical Specifications and Technical Requirements for Instrumentation Tech Specs:

a. The Limiting Conditions for Operation, Applicability, and Bases for the LCO.
b. The conditions and required actions with completion time of one hour or less. Cognitive Level:

Higher X Lower Question Source: New Modified Bank X Bank Question History: Modified bank question 026 A4.01 43 which was used on Sequoyah's 09/2010 NRC exam. Comments: QUESTIONS REPORTfor SQN-WBN composite exams

1. 026 A4.01 043Which ONE of the following is correct regarding the logic for MANUAL actuation of aphase B Isolation?Operation of any one of 4 handswitches will actuate both trains.Operation of either 1-HS-30-64A (M-6) or 1-HS-30-68A (M-5) will actuate train A ONLY.Operation of the paired handswitches (M-5 or M-6) operated sequentially willactuate both trains.Operation of the paired handswitches (M-5 or M-6) operated simultaneously willactuate both trains.

A.B.C.D.Friday, May 15, 2015 1:51:31 PM 1 A CTUATE SIS 1-47W611-6J-1 SYSTEM SHEET 1-+7'1f611-JO 1, 6 1-+7'1f6t1-61 2 1-+7W6t1-62 t 1-+7W611-6J J, 5, 6, 8 1-+7W611-65

1. 2, J 1-+7W6t1-68 t 1-+7W611 70 J '* 5 t-+7W611-81 1-+7W611 Jt +A-Ei90 4A-680 CONTAINMENT SPRAY DISCHARGE VALVE 1-FCV-72-2 1-471¥611-72-1 CONTAINMENT SPRAY' PUMP 1-47161 t-72-1 MAIN STEAM LINE ISOLATION VALVES 1-47W611-1-1 4A-670 + SYSTEM t-+7W611-JO t-47W611-70 1-47'111611-67 4A-660 + L 5 t-47W611-88 SHEET " . NORMAL 1-HS REFUEL 90-410 NOTES: RADIATION DETECTION IN THE REFUELING AREA (1-47WEilt-J0-5)
1. FOR TRAIN 8, 1-RE-90-1J1, 1-HS-90-415, ANO O-HS-90-1J6A2 ARE USED. O-HS-90-136111

& 136A2 ARE MULTI SELECT SW'S USED TO TEST ONE CHANNEL AT A TIME, J. ALL ISOLATION SICNALS AND CIR.CUI TRY' THAT ARE PART OF ENGINEERED SAFEGUARDS ARE REDUNDANT, 4. NOTE DELETED. 5 *

6. VALVES 1-FCV-4J-22 AND t-FCV-4J-2J MAY BE NORMALLY' OPEN WITH 1-FSV-4J-2J ENERGIZED TO ALLOW CONTINUOUS RCS FLOW DUE TO THERMAL CYCLING CONCERNS (REF. EOC-5113+).

QUESTIONS REPORTfor SQN-WBN composite exams

1. 026 A4.01 043Which ONE of the following is correct regarding the logic for MANUAL actuation of aphase B Isolation?Operation of any one of 4 handswitches will actuate both trains.Operation of either 1-HS-30-64A (M-6) or 1-HS-30-68A (M-5) will actuate train A ONLY.Operation of the paired handswitches (M-5 or M-6) operated sequentially willactuate both trains.Operation of the paired handswitches (M-5 or M-6) operated simultaneously willactuate both trains.

A.B.C.D.Monday, May 04, 2015 1:47:47 PM 1

43. Given the following timeline:

00:00:00 Unit 1 is at 100% power. 00:0 1 :00 A LOCA occurs. 00:02:00 The following is observed: 00:03:00 The following is observed: 

< QUESTION CONTINUED ON THE NEXT PAGE >

00:04:00 The following is observed: 00:05:00 The following is observed: 

 < QUESTION CONTINUED ON THE NEXT PAGE >

From the times and indications listed above, which ONE of the following co mpletes the statement listed below?

The MSIVs would FIRST be expected to AUTOMATICALLY CLOSE at __________. A. 00:02:00 B. 00:03:00 C. 00:04:00 D. 00:05:00 CORRECT ANSWER: C A. Incorrect: This distractor presents two of the four c hannels of containment pressure in excess of the High cont ainment pressure setpoint of 1.5 psig. While the coincidence is correct, the setpoint of 1.5 is incorrect. As seen in T/S table 3.3.2-1 (thi s value may be found in many other references), the containment pressu re-High High setpoint applicable to the automatic steam line isolation is 2.8 psig. The Containment High setpoint of 1.5 would be applicable to the automatic safety injection. B. Incorrect: This distractor presents th ree of the four channels of containment pressure in excess of the High cont ainment pressure setpoint of 1.5 psig. Again, both the coincidenc e required and the setpoint are incorrect. C. Correct: This distractor presents tw o of the four channel s of containment pressure in excess of the High-High containment pressure setpoint of 2.8 psig. As seen on print 1-47W611-88-1, the coincidence for the automatic steam line is olation is 2 of 4. D. Incorrect: This distractor presents th ree of the four channels of containment pressure in excess of the High-High containment pressure setpoint of 2.8 psig. While the MSIVs would be expected to be closed at this

time, they would have first closed at 0400.

Question Number: 43

Tier: 2 Group: 1

K/A: 039 Main and Reheat Steam System A3 Ability to monitor automatic operation of the MRSS, including: A3.02 Isolation of the MRSS Importance Rating: 3.1 3.5

10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b: Not applicable K/A Match: The K/A is matched because the applicant must be able to monitor for the automatic isolation of the Main Steam System during a casualty which causes containment pressure to rise. Technical

Reference:

1-47W611-88-1 T/S Table 3.3.2-1, Engineered Safety Feature Actuation

System Instrumentation

Proposed references to

be provided: None Learning Objective: 3-OT-TS0303, TS 3.3 "INSTRUMENTATION" 1. DESCRIBE the following aspects of Technical Specifications and Technical Requirements for Instrumentation Tech Specs:

a. The Limiting Conditions for Operation, Applicability, and Bases for the LCO. b. The conditions and required actions with completion time of one hour or less Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: 4A-69D 4A-68D 4A-67D I TEST BYPASS --ACTUATE 'f-PdS' 30-42A I I --TEST BYPASS --4A-660 I --TEST BYPASS I I if,-Hs"' ..... TRAIN A/B \.30-638 ,, 1'-PdS ' f-PdS f-PdS !_ ACTUATE TRAIN A/8 30-64A AC" ' 0 0 N F F ACTUATE if,-HS TRAIN A/B I ,30-63A PHASE A ct. CNTMT VENT ISOLATION ---* 6C-1 25A A --1---,--,+ HIGH-HIGH CONTAINMENT >----t2/4 PRESSURE If 2 of the 4 containment pressure channels reach 2.8 psig (the HI-HI setpoint), then the MSIVs will automatically close. + 0 0 N F F ---l RESET TRAIN A NOTE 5 *-;;f-HS 30-640 0 0 N F F ] I ACTUATE TRAIN A/B PHASE 8 & CNTMT VENT ISOLATION LOG 509011 TRAIN 8 LOG TR B SD9013 '-" 6C-125C LOG SD9010 TRAIN A CONTAINMENT SPRAY DISCHARGE VALVE 1-FCV-72-2 1-47\11611-72-1 LOG TR A SD9012 CONTAINMENT SPRAY PUMP 1-47W611-72-1 MAIN STEAM LINE ISOLATION VALVES 1-47W611-1-1 r----------, : PHASE 8 : I CNTMT ISOLATION I L-----------..J SYSTEM SHEET 1-47W611-30 3, 4 1-47W611-70 3 1-47W611-32 2 1-47W611-67 3 A CONTAINMENT SPRAY PUMP AND VALVES l-47W611-72-l TRI * '\..30* NOT ESFAS Instrumentation 3.3.2 Watts Bar-Unit 1 3.3-36 Amendment 23 Table 3.3.2-1 (page 3 of 7) Engineered Safety Feature Actuation System Instrumentation FUNCTION APPLICABLE MODE S OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE NOMINAL TRIP SETPOINT 4. Steam Line Isolation (continued)

c. Containment 1, 2(c), 3(c) 4 E SR 3.3.2.1 2.9 psig 2.8 psig Pressure- SR 3.3.2.4 High High SR 3.3.2.9 SR 3.3.2.10
d. Steam Line Pressure (1) Low 1, 2(c), 3(a) (c) 3 per steam D SR 3.3.2.1 666.6(b) psig 675(b) psig line SR 3.3.2.4 SR 3.3.2.9 SR .3.2.10 (2) Negative 3(d) (c) 3 per steam D SR 3.3.2.1 108.5(e) ps i 100(e) psi Rate-High line SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10
5. Turbine Trip and Feedwater Isolation
a. Automatic 1, 2(f), 3(f) 2 trains H SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays b. SG Water 1, 2(f), 3(f) 3 per SG I SR 3.3.2.1 83.1% 82.4% Level-High SR 3.3.2.4 High(P-14) SR 3.3.2.9 SR 3.3.2.10 (h) c. Safety Refer to Function 1 (Safety Injection) for all initiation Injection functions and requirements.
d. North MSV Vault 1, 2(f), (g) 3/vault O SR 3.3.2.6 5.31 inches 4 inches Room Water Room SR 3.3.2.9 Level - High e. South MSV Vault 1, 2(f), (g) 3/vault O SR 3.3.2.6 4.56 inches 4 inch es Room Water Room SR 3.3.2.9 Level - High (continued)

(a) Above the P-11 (Pressurizer Pressure) interlock. (b) Time constants used in the lead/lag controller are t 1 50 seconds and t 2 5 seconds. (c) Except when all MSIVs are closed a nd de-activated. (d) Function automatically blocked above P-11 (Pressurizer Interlock) setpoint and is enabled below P-11 when safety injection on Steam Line Pressure Low is manually blocked. (e) Time constants utilized in the rate/lag controller are t 3 and t 4 50 seconds. (f) Except when all MFIVs, MFRVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve. (g) MODE 2 if Turbine Driven Main Feed Pumps are operating. (h) For the time period between February 23, 2000, and prior to turbine restart (following the next time the turbine is removed from service), the response time test requirement of SR 3.3.2.10 is not applicable for 1-FSV-47-027.

44. Given the following conditions: - Unit 1 is at 100% power.

Subsequently: - EHC POWER FAILURE (72-C) is FLASHING. - The light above 1-HS-47-24, TURBINE TRIP is GREEN. - Unit 1 is in MODE 3. - The position of SG 2 MFW isol vlv is UNKNOWN. Which ONE of the following describes T/S LCO 3.3.3, Post Accident Monitoring (PAM) Instrumentation AND when the valves go CLO SED in reference to the picture above?

In accordance with T/S LCO 3.3.3, actions ____(1)____ required to be taken.

The valves received a closure signal when RCS Tavg dropped to less than ____(2)____ °F.

(1) (2) A. are 550 B. are 564 C. are NOT 550 D. are NOT 564 CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: As seen in Table 6.2.4-1 of the stations Final Safety Analysis Report, containment penetration is in part isolated by motor operated valve 3-47. Therefore, this valve is a containm ent isolation valve (contained within system 3, Main Feedwater). In accordance with T/S LCO 3.3.3, Post Accident Monitoring (PAM) Instrumentation, whenever One or more Functions with one channel [is] inoperable, then action A.1 must be completed. This action states to Restore required channel to Operable status and has a completion time of 30 days. The function which has the inoperable component in this case is function 11-Containment Isolation Valve Position- which is found in Table 3.3.3-1 of the Unit 1 technical specifications. Therefore is is correct that the loss of the valve position indication in the question is a failure to meet T/S LCO 3.3.3. The stem of the question presents a normal and uncomplicated reactor trip. The reactor trip resulted from a Turbine Trip which was caused by an EHC power failure (this sequence of events may be substantiated using 1-ARI-71-75, Turbine Trip First Out.). A normal and uncomplicated reactor trip will always result in a Feedwater Isolation because the automatic steam dump system will always stabilize RCS Tavg at a value less than 564°F. A reactor trip coupled with Tavg less than 564°F will produce a Feedwater isolation. The value of 564°F may be found (in addition to many other places) in Nuclear Engineering Scaling and Setpoint Document SSD-1-T-68-2. Therefore, it is Incorrect to select 550°F for this setpoint. It is plausible to do so as 550°F is the Lo Lo Tavg (P-12) setpoint. Lo Lo Tavg, as seen in 1-ARI-64-70, Bypass Intlk, & Permissive, is the steam dump block and even though its name implies a more severe action (the Lo Lo versus Lo), it simply causes the steam dumps to close versus actuating a Feed Water Isolation. B. Correct: Again, it is correct that T/S LCO actions are required and that the setpoint for Feedwater isolation is 564°F. C. Incorrect: It is not correct that the conditions listed do not incur T/S LCO required actions. It is plausible to believe so for several reasons. Firstly, if the Unit had stabilized in Mode 4 (when the failure of indication occurred), then Table 3.3.3-1 would demonstrate that no actions were required as the Modes of applicability are seen to be 1,2 and 3. Secondly, it is plausible to believe that the valve with the failed position indication is not a containment isolation valve and thus not included in the T/S LCO. A good example of this would be the Main Feed Regulating valve which is in series with 1-ISV-3-47. The MFRV also closes on a FWI but is not a containment isolation valve. Also, as discussed it is Incorrect and yet plausible that the FWI would occur at 550°F. D. Incorrect: While it is correct that the FWI occurs at 564°F, it is not correct and yet plausible that T/S LCO actions are not required.

Question Number: 44 Tier: 2 Group: 1

K/A: 059 Main Feedwater (MFW) System G2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. Importance Rating: 3.9 4.6

10 CFR Part 55: (CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to recognize that a failure of the position indication for a main feed water valve represents entry-level conditions for the unit's Technical Specifications.. Technical

Reference:

1-ARI-71-75, Turbine Trip First Out 1-ARI-64-70, Bypass, Intlk, & Permissive SSD-1-T-68-2 FSAR Table 6.2.4-1, Containment Penetrations and

Barriers T/S LCO 3.3.3, Post Accident Monitoring (PAM) Instrumentation

Proposed references to

be provided: None Learning Objective: 3-OT-TS0303, TS 3.3 "INSTRUMENTATION" 1. DESCRIBE the following aspects of Technical Specifications and Technical Requirements for Instrumentation Tech Specs:

a. The Limiting Conditions for Operation, Applicability, and Bases for the LCO. b. The conditions and required actions with completion

time of one hour or less. Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: UFSAR - Amendment 1 PAM Instrumentation 3.3.3 Watts Bar-Unit 1 3.3-41 Amendment 72 3.3 INSTRUMENTATION 3.3.3 Post Accident Monitoring (PAM) Instrumentation

LCO 3.3.3 The PAM instrumentation for each Function in Table 3.3.3-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.3-

1.

ACTIONS -----------------------------------------------------------------------NOTE--------------------------------------------------------- Separate Condition entry is allowed for each Function. ---------------------------------------------------------------------------------------------------------------------------------------- CONDITION REQUIRED ACTION COMPLETION TIME A. ----------NOTE-------------- Not applicable to Functions 3, 4, 1 4, and 16. -------------------

One or more Functions with one required channel inoperable . A.1 Restore required channel to OPERABLE status. 30 days B. Required Action and associated Completion Time of Condition A not met. B.1 Initiate action in accordance with Specification 5.9.8. Immediately (continued)

PAM Instrumentation 3.3.3 Watts Bar-Unit 1 3.3-44 Amendment 72 Table 3.3.3-1 (page 1 of

2) Post Accident Monitoring Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS/TRAINS CONDITION REFERENCED FROM REQUIRED ACTION D.1 1. Intermediate Range Neutron Flux(g) 1(a), 2(b), 3 2 E 2. Source Range Neutron Flux 2(c) , 3 2 E 3. Reactor Coolant System (RCS)

Hot Leg Temperature (T -Hot) 1,2,3 1 per loop E 4. RCS Cold Leg Temperature (T -Cold) 1,2,3 1 per loop E 5. RCS Pressure (Wide Range) 1,2,3 3 E 6. Reactor Vessel Water Level (f) (g) 1,2,3 2 F 7. Containment Sump Water Level (Wide Range) 1,2,3 2 E 8. Containment Lower Comp. Atm. Temperature 1,2,3 2 E 9. Containment Pressure (Wide Range) (g) 1,2,3 2 E 10. Containment Pressure (Narrow R ange) 1,2,3 4 E 11. Containment Isolation Valve Position (g) 1,2,3 2 per penetration flow path (d)(i) E 12. Containment Radiation (High Range) 1,2,3 2 upper containment 2 lower containment F 1 3. RCS Pressurizer Level 1,2,3 3 E 14. Steam Generator (SG) Water Level (Wide Range)(g) 1,2,3 1/SG E 15. Steam Generator Water Level (Narrow Range) 1,2,3 3/SG E 16. AFW Valve Status (j) 1,2,3 1 per valve E 17. Core Exit Temperature -Quadrant 1(f) 1,2,3 2 (e) E (continued)

WBN Unit 1 Turbine Trip First Out 1-ARI-71-75 Rev. 0000 Page 13 of 40 A. Loss of +/-15V dc power supply (Set A or Set B) B. Loss of 48V dc power supply (Set A or Set B) [1] IF greater than or equal to 50% power (P-9), THEN VERIFY Turbine Trip/Rx Trip, AND GO TO 1-E-0, REACTOR TRIP OR SAFETY INJECTION. [2] IF less than 50% power (P-9), THEN VERIFY Turbine Trip, AND GO TO 1-AOI-17, TURBINE TRIP. 1-45W600-47-2 1-45W600-55-12 WBN-VTM-W120-2872 1-AOI-17 1-E-0 Source DC Bus Trip 72-CEHC POWER FAILURE (Page 1 of 1) Probable Cause: Corrective Action:

References:

LOOP ID: PROCESS RANGE: SAFETY RELATED: LOOP FUNCTION: LOOP ACCURACY Comparator Functions Lo TAvg (FW Isolation) Lo Lo TAvg (P-12) Overtemp Delta T Trip SSD-1-T-68-2 REV. 6 Page 27 of 63 SCALING and SETPOINT DOCUMENT ACCURACY & CALIBRATION SECTION DELTA TIT AVG LOOP 1 PS I 530.0 to 650.0°F -Narrow Range THot 510.0 to 630.0°F -Narrow Range TCold 0 to 150% power -Delta T 530.0 to 630.0°F -T Avg 1700 to 2500 psig -Pressurizer Pressure 0 to 60 PU -NIS Upper/Lower Flux Yes. Up to and including the isolators (EAO boards or SSPS). The peripheral devices are categorized as follows: All MCR indicators -quality-related; Plant Computer log points and MCR annunciators, non-safety related; XS-68-2B and TR-68-2A -quality related; Compliance: See TI-49 (Reference 56). Perform measurement and processing for Delta and Average Temperature. Provides Comparator Outputs to Low TAvg (Feedwater Isolation), Low-Low TAvg (Permissive P-12), Overtemperature and Overpower Delta-T Reactor Trip, Turbine Runback and Block Rod Withdrawal. Also provides Analog Outputs to the LOG (Plant Computer), Temperature Recorder and to the Main Control Board Indicators and to the ICCM. (8) Safety Analysis Limit Setpoint Tech Spec Allowable Value (1) 564.0°F (l) (13) (1) (14) (l) 550.0°F (l) (13) (1) (14) FSAR Figure 15.1-1 (2) 0 PU (2) (13) (2) (14) OverTemp Turb Rbk & Blk (l) OT Delta T trip setpoint -1 PU (1) (14) Rod Withdrawal (1) (13) Overpower Delta T Trip FSAR Figure 15.1-1 (3) 0 PU (3) (13) (3) (14) Overpower Turb Rbk & Blk (l) OP Delta T trip setpoint -1 PU (1) (14) Rod Withdrawal (l) (13) Loop Indication Functions Technical Specification/ Main Control Board TRM/ODCM Sections Indication Uncertainty Applicable LCO Value RCS Average Temperature S 593.2°F (LCO 3.4.1) NIA (21) (21) RCS Average Temperature 551°F (LCO 3.4.2) +/- 3.6°F (22) R6 WBN Unit 1 Bypass, Intlk, & Pe rmissive 1-ARI-64-70 Rev. 0000 Page 28 of 47 A. Plant cooldown NOTE The P-12 interlock prevents opening of the steam dump valves if RCS temperature is less than 550 °F. This interlock can be bypassed for 3 of 12 steam dump valves. P-12 light also indicates steam dump valves should be closed. [1] VERIFY Tavg less than or equal to 550 °F. [2] IF bypass of LO-LO TAVG interlock is desired, THEN REFER TO 1-SOI-1.02, STEAM DUMP SYSTEM. W 7246D11 Sht. 40 1-47W611-63-1 1-SOI-1.02 Source Setpoint 1-TS-68-2J (Loop 1) Tavg in 2/4 loops less than or equal to 550 °F 1-TS-68-25J (Loop 2) 1-TS-68-44J (Loop 3) 1-TS-68-67J (Loop 4)

68-A P-12 LO-LO TAVG STM DUMP BLOCK (Page 1 of 1) Probable Cause: Corrective Action:

References:

45. Given the following conditions: - Unit 1 is at 3% power. - The MDAFWPs are in service supplying the SGs.

POS-3-164, positioner to 1-LCV-3-164, air output FAILS to 0 psig. POS-3-164A, positioner to 1-LCV-3-164A, air output FAILS to 0 psig.

Which ONE of the following descri bes the response of the NORMAL AND BYPASS Level Control Valves?

The NORMAL level control valve 1-LCV-3-164 wil l ____(1)____. The BYPASS level control valve 1-LC V-3-164A wi ll ____(2)____. NOTE: 1-LCV-3-164, MD AFW PUMP 1A-A SG 1 LEVEL CONTROL 1-LCV-3-164A, SG 1 AUX FEEDWATER 1-FCV-3-164 BYPASS (1) (2) A. OPEN OPEN B. OPEN CLOSE C. CLOSE OPEN D. CLOSE CLOSE

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: As seen in SSD-1-LPL-3-164 (the normal level control valve), 1-LCV-3-164 fails open on a loss of control air and 1-LCV-3-164A (the bypass level control valve) fails closed on a loss of control air. The purpose of the valve positioners in this case is to receive a pneumatic signal from 1-LM-3-164A and translate that into the appropriate pneumatic output signal to control the valve actuator. Therefore, when the positioners fail and output 0 psig, the valves in effect suffer a loss of control air and travel to their failed position. It is plausible to believe that 1-LCV-3-164A fails open as its sister valve does (as well as many other air operated valves). B. Correct: As mentioned the failure positions of both valves is correct. C. Incorrect: It is neither correct; that 1-LCV-3-164 will close, nor that 1-LCV-3-164A will open. It is plausible to believe this for the aforementioned rationale. An item of note is that it is completely plausible that an AFW level control valve would fail closed as this is the case with the TDAFWP's LCVs. On a loss of control air, the TDAFWP's LCVs fail closed. D. Incorrect: While it is correct that 1-LCV-3-164A will close, it is not correct and yet plausible that 1-LCV-3-164 will close. Question Number: 45 Tier: 2 Group: 1

K/A: 061 Auxiliary / Emergency Feedwater (AFW) System K6 Knowledge of the effect of a loss or malfunction of the following will have on the AFW

components: K6.01 Controllers and positioners (CFR: 41.7 / 45.7)

Importance Rating: 2.5 2.8

10 CFR Part 55:

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to understand the effect on the AFW level control valves 1-LCV-3-164 and 1-LCV 164A that failures of the associated positioners have Technical

Reference:

SSD-1-LM-3-164A Proposed references to

be provided: None Learning Objective: 3-OT-SYS003B, Auxiliary Feedwater System 12. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on

the Auxiliary Feedwater System:

a. Instrumentation failures
b. Control Air Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: W BNSSD-1-LPL-3-164 Reviewed b y:R. S. Henderson App roved b y:K. D. GarrisonDate:10/13/2009E/I 1-L-11A 1-LM- 3-164D 1-L-11 A1-L-11A 1-LM- 3-164C 1-L-11A 1-LS- 3-164D/B 1-LS- 3-164B/D 1-LC- 3-164B 1-L-11A 1-LM- 3-164 1-L-183D 1-LT- 3-164 1-L-11A 1-PX- 3-164 STEAM GENERATOR 1 MOTOR DRIVEN AFW LEVEL CONTROL SCALING AND SETPOINT DOCUMENT LOOP DRAWING LOOP NAME: 1-XA 3C-60B SPARE I/I 1-M-3 1-LI- 3-164 1-L-379 1-LM- 3-164B 1-L-10 1-LI- 3-164C 1-LIC- 3-164B 1-L-11A 1-XS- 3-164A 1-LIC- 3-164A 1-LC- 3-164A I/I 1-L-11A 1-LCV- 3-164 1-LCV- 3-164A O X S 1-LSV- 3-164 S 1-LSV- 3-164A I/P 1-LM- 3-164A 1-L-11A NEST 1 SLOT 10, 111-PX- 3-164A TO 15Vdc & -NEST BUSSES I/E 1-M-4 TP 1-L-11A TP-8 PAGE OF REVISION225 10 W BNSSD-1-LPL-3-164 Reviewed b y:R. S. Henderson App roved b y:K. D. GarrisonDate:10/13/2009 SCALING AND SETPOINT DOCUMENT PAGE OF REVISION 10 10Instrument No.:1-LCV-3-164 PROCESS CONTROL VALVES AND CONTROL DEVICES M asoneilan M frr: M odel: 37-20721 N/A V endor ID No: Contract No: 83577 Function: CV A ir to Close V alve Action: M asoneilan Positioner Mfrr: Model No:7432-713 Cam Type: U/ACalibrated Ranges:Input: 9-15 PSIG (4)Document: Document: Document: Document: 47B601-3 Output: 0-100% OPEN M FRR Stroke: N/A A ccuracy:+/- 6%N/A M IG A ssociated Interlock: 1-LSV-3-164

ENG to close and DEENG to modulate 1-LCV-3-164.

Associated Drawings: W BN-VTM-M120-0010 25Remarks:(1) Pressure Regulator: 1-PREG-3-164 Setting: 65.0 PSIG As-Found 63 to 65.5 PSIG / As-Left 64.5 to 65.5 PSIG (2) DCN F-29644-A revised action of valve positioner from direct to

reverse. This was done by rolling the output wires of 1-LM-3-164C.

The input wires at the valve positioner were not changed, leaving

black to (+) terminal and white to (-) terminal.

(3) DCN W-32189-B changed positioner (1-POS-3-164) from current in/pressure out to pressure in/pressure out. (4) Input is to positioner (1-POS-3-164) which is reverse acting.

Valve fails open with loss of supply air.

(5) DCN W-37869-A changed regulator setting.

(6) DCN 51815-A changed 1-PREG-3-164 setpoint. Function: Steam Generator 1 Motor Driven AFW Level Control Location: EL 737 , COL A3-5 W BNSSD-1-LPL-3-164 Reviewed b y:R. S. Henderson App roved b y:K. D. GarrisonDate:10/13/2009 SCALING AND SETPOINT DOCUMENT PAGE OF REVISION 10 11Instrument No.:1-LCV-3-164A PROCESS CONTROL VALVES AND CONTROL DEVICES Leslie M frr: M odel: M ark P N one V endor ID No: Contract No: 87379 Function: CV A ir to O p en V alve Action: Baile y Positioner Mfrr: M odel No: AV1121000Cam Type: LinearCalibrated Ranges:Input: 3-9 PSIG (5)Document: Document: Document: Document: 47B601-3 Output: 0-100% OPEN M FRR Stroke: N/A A ccuracy:+/- 6%N/A M IG A ssociated Interlock: 1-LSV-3-164A

ENG to close and DEENG to modulate 1-LCV-3-164A Associated Drawings

W BN-VTM-B045-0470 , Manual ID WBN-L170-0010 25Remarks:(1) Pressure Regulator: 1-PREG-3-164A1 Setting: 59 +0/-4 PSIG (2) DCN S-21327-A revised 1-PREG-3-164A1 setting. (3) DCN W-28198-A revised 1-PREG-3-164A1 setting.

(4) DCN W-32189-B changed positioner (1-POS-3-164A) range and changed action of 1-LS-3-164A. (5) Input is to positioner (1-POS-3-164A).

Valve fails closed with loss of supply air.

(6) EDC 51696-A revised 1-PREG-3-164A1 setpoint.

(7) DCN 52127-A replaced positioner for 1-LCV-3-164A. Function: Steam Generator 1 Motor Driven AFW Level Control Location: EL 737 , COL A3-S

46. Given the following conditions: - An electrical board has failed - The following indications are observed:

Which ONE of the following describes the electrical board that failed? 120 VAC Vital Instrument Board _______ has FAILED. A. 1-I B. 1-II C. 1-III D. 1-IV CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: The indications depicted occur upon the loss of 120 VAC Vital Instrument Board 1-I. B. Incorrect: The annunciator pattern which would be received is similar. Channel II bistables (i.e. a different row of bistables) would be lit versus Channel I bistables. Also, both of the CERPI monitors would remain powered. C. Incorrect: Again, the annunciator pattern would be different. Channel III bistables would be lit versus Channel I bistables. Both of the CERPI monitors would remain powered. D. Incorrect: Again, the annunciator pattern would be different. Channel IV bistables would be lit versus Channel I bistables. The right hand

CERPI monitors would be deenergized and thus dark.

Question Number: 46

Tier: 2 Group: 1

K/A: 062 A.C. Electrical Distribution 2.4.46 Ability to verify that the alarms are consistent with the plant conditions. Importance Rating: 4.1 4.4

10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.3 / 45.12) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to verify that the alarms and indications are consistent with the loss of 120 VAC Vital Instrument Board 1-I. Technical

Reference:

1-AOI-25.01, Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I Proposed references to be provided: None Learning Objective: Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

Watts Bar Nuclear Plant Unit 1 Abnormal Operating Instruction 1-AOI-25.01 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I Revision 0003 Quality Related Level of Use: Continuous Use Effective Date: 04-08-2015 Responsible Organization: OPS, Operations Prepared By: Scott Warrington Approved By: William Sprinkle

WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.01 Rev. 0003 Page 3 of 27 1.0 PURPOSE This instruction provides actions to respond to a loss of 120V AC Vital Instrument Power Boards 1-I and 2-I. This instruction will stabilize the unit and provide a list of

equipment which may NOT be available due to a loss of the board. [C.1, C.3] 2.0 SYMPTOMS 2.1 Alarms A. 125 DC VITAL CHGR/BATT 1 ABNORMAL [17-A] B. 125 DC VITAL BATT BD 1 ABNORMAL CKTS ISOLATED [17-B] C. 120V AC VITAL PWR BD 1-I UV/CKT TRIP [17-D]. D. 120V AC VITAL INVERTER 1-I / 0-I ABNORMAL [17-C]. E. NORTH/SOUTH VALVE VAULT RM LEVEL HI/MFW ISOLATION [57-C] F. SG FEEDWATER FLOW HI [58-B] G. SG 1 LEVEL HI [60-B] H. SG 1 STM-FW FLOW MISMATCH [60-C] I. SG 2 LEVEL HI [61-B] J. SG 2 STM-FW FLOW MISMATCH [61-C] K. SG 3 STM-FW FLOW MISMATCH [62-C] L. SG 4 LEVEL HI [63-B] M. SG 4 STM-FW FLOW MISMATCH [63-C] N. C-5 LO TURB IMPULSE PRESS ROD BLOCK [66-A] O. RCS COPS ARM PERMISSIVE [67-B] P. DCS TROUBLE [82-F] Q. POWER RANGE OVERPOWER ROD WD STOP [83-A] R. POWER RANGE UPR DETECTOR FLUX DEVN [83-B] S. POWER RANGE LWR DETECTOR FLUX DEVN [83-C] Thealarmpatternseenfor120VACVitalInstrument Board1-Iisseeninthe stemofthequestion. WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.01 Rev. 0003 2.1 Alarms (continued) Page 4 of 27 T. POWER RANGE CHANNEL DEVIATION [83-E] U. RVLIS SYS MALFUNCTION [85-F] V. CERPI TROUBLE [86-C] W. RWST LEVEL LO-LO [126-D] X. CL ACCUM 1 LEVEL HI/LO [131-A] Y. CL ACCUM 1 PRESS HI/LO [131-B] Z. CL ACCUM 2 LEVEL HI/LO [132-A] AA. CL ACCUM 2 PRESS HI/LO [132-B] BB. AC VITAL PWR BD 2-I UV/CKT TRIP [135-F]. 2.2 Indications A. Power range 1-N-41, Intermediate range 1-N-135, and Source Range 1-N-131 failure. B. Rx Trip SI Status panel 1-XX-55-5 will be DARK. C. Train A CNTMT Isolation Panel 1-XX-55-6E will be DARK. D. Loss of Channel I S/G feedwater flow and steam flow inputs to DCS. E. Loss of letdown due to Non-Essential air isolation to CNTMT. F. 1-LT-68-339, PZR Level indication fails low. G. 1-PT-68-340, PZR pressure indication fails low. 2.3 Automatic Actions A. Rod withdrawal block due to high flux and C-5 interlock. B. Charging suction swap to RWST due to de-energized separation relay causing 1-LCV-62-132, VCT OUTLET VALVE, to close and 1-LCV-62-135 and 1-LCV-62-136, RWST CVCS SUPPLY HDR ISOLATION, to open injecting borated water to the RCS. C. If in service, Train B MCR, EBR, and SDBR Chillers will swap to Train A, and auto start of Train B is disabled. Andbecausetheinputrelayisdeenergized(for SSPS),atrip bistablewillbe generatedforthe lossofFandG (andmanyother triporESFAS

functions). WBN Unit 1 Loss of 120V AC Vital Instrument Power Boards 1-I or 2-I 1-AOI-25.01 Rev. 0003 Appendix A (Page 5 of 6) 120V AC Vital Instrument Power Board 1-I Loads Page 19 of 27 SYSTEM 74 - RHR:

  • 1-FCV-74-1 will require placing 1-XS-74-1 to AUX position [Rx MOV Bd 1A1-A, Compt 5B] to open valve when RCS pressure is less than 380 psig.
  • 1-FCV-74-16 & -32 will require placing 1-XS-74-16 & -32 to AUX position [1-L-11-A, Aux Control Room] to control RHR cool down temperature. SYSTEM 82
  • Emergency start for 1A-A and 2A-A D/Gs on 1-M-1 does not work. D/Gs will Emergency start from 0-M-26. SYSTEM 85 - CERPI
  • 1-MON-85-5000/1, CERPI Monitor A. SYSTEM 88 - Containment Isolation:
  • Train A CRI and CVI will occur after restoration of the board.
  • If the refueling logic switch, 1-HS-90-410-A [1-R-73], is in the REFUEL position, then an A Train ABI (partial ABI associated with high rad in refuel area) will also occur when the CVI is initiated. SYSTEM 90 - Radiation Monitors:
  • 1-RM-90-106, Lower Containment Air Monitor control power NOT operable.
  • 0-RM-90-125, MCR Train A Air Intake Monitor control power NOT operable.
  • 1-RM-90-130, Containment Train A Purge Exhaust Monitor control power NOT operable. SYSTEM 92 - Excore Nuclear Instrumentation:
  • All Channel I NIS Control and Instrument Power. SYSTEM 98-Distributed Control System:
  • Primary power supply to DCS cabinets 1-L-981, 983, and 1-R-193. (ref 1-SOI-98.01 for specific loads)

SYSTEM 8 5 - C ERPI

  • 1-MON-85-5000/1, CERPI Monitor A.
47. Given the following timeline:

00:00:00 Unit 1 is at 100% power. 00:00:0 1 The 1A-A SDBD experiences a BLACKOUT. At 00:00: 1 5, which ONE of the following combinations of alarms will be seen in the MCR? A. B. C. D. CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: It is not correct that a BLACKOUT on one 6.9kV SDBD will cause none of the EDGs to start. An undervoltage condition on one 6.9kV Shutdown Board will cause all of the EDGs to start. In the case presented in the question, the 27LV relays which are monitoring the 1A-A 6.9kV SDBD (loss of voltage relays) will actuate at <87% of nominal shutdown board voltage. T herefore, as the board voltage decays to 0, the 27LV relays will ac tuate to pick up two timer relays LV-1 and LV-2. Only 1 of these two timers is required to cause (after

0.75 seconds) the LV relay to energi ze. When the LV relay energizes, the Normal (from CSST C), the al ternate (from CSST D) and the maintenance feeder breaker (from a Unit board-this breaker is normally racked down) will trip open. At this point, the shutdown board is truly blacked out (0 volts). One half second after voltage drops below 70% of nominal (remember that the board is blacked out at this time), the 27D1A relays will act to energize the 27D1AX and 27D1AY

relays. These two relays will open contacts in the ES1AY relay circuit

which de-energizes ES1AY and ES1AY

1. Note that ES stands for emergency start. With the diesels aligned in standby, ES1AY will act to start the diesel generator for the specif ic shutdown board and ES1AY1 will operate to st art the other three di esels. Print 1-45W760-211-1 shows an excellent summary of the high level relay actions affecting the shutdown board. It demonstrates the 27LV, 27S and 27D relays' operation. Print 1-45W760-82-6 demonstrates in detail the emergency start of both the affe cted shutdown board's EDG and the common emergency start of the three other EDGs. In accordance with multiple documents in the combined license basis, the EDGs are designed to start and reenergize the SDBDs within 10 seconds.

It is plausible to believe this because 14 seconds has elapsed

between the blackout of the SDBD and the time at which the indications would be seen. If one believed that greater than 14 seconds was required for an EDG to start and repower the SDBD, then one would arrive at this answer. B. Incorrect: Again, it is incorrect that only the 1A-A EDG would be running. It is plausible to believe this because one may misunderstand the

operation of the common emergency start. C. Incorrect: It is incorrect that only the "A" train EDGs would be running. It is plausible to believe this simply because of the concept of affiliating the trains (e.g. believing that the "A " train EDGs would start on the blackout of an "A" train SDBD). D. Correct: Correct, as discussed, a blackout of one 6.9kV SDBD will initiate the Common Emergency start of all of the EDGs.

Question Number: 47

Tier: 2 Group: 1

K/A: 062 A.C. Electrical Distribution K1 Knowledge of the physical connections and/or cause effect relationships between the ac distribution system and the following systems: K1.02 ED/G

Importance Rating: 4.1 4.4

10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.3 / 45.12)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to understand the cause effect relationship between the blackout of the 1A-A 6.9kV SDBD and its associated ED/G. Technical

Reference:

1-45W760-211-1 1-45W760-82-6 Proposed references to be provided: None Learning Objective: 3-OT-SYS-201B, BLACKOUT AND LOADSHED LOGIC RELAYS 9. Given specific plant conditions, ANALYZE the effect that a loss of shutdown boards or malfunction of the Blackout and Load Shed Logic Relays will have on the following:

a. Diesel Generators Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

1E 1-XS-6K (1-FU-211-A/C2) 11 D1AESP ( 1-FU-211-A/01) K609 0 1-R-+8 12<S1l {1-R-51.TRB) AUX CONT 01 AESl ... D D1AES2 D D1AES3A 5G 1-XS-82-121 DlAESJ ES1AY3 D1AES+ +T3£Ts\L) Ill @ 11K Ill PJ JJ-N 1-XS-82-121 AUX CONT RM 18 A Rl2l 1-HS-82-16C 6 EMERG ST ART Rll:-l'. (0-L-4) 8 "' c J3-N @ ., (1-FU-211-A/C2) (1-FU-211-A/Bl) t o th e l A-A EDG D1AEN ( 1-FU-275-R73/L

10) 0 DlACSP ( 1-FU-275-R75/K3)

,.,.._ l-FU-82-DA1/2 DCP2 I 68 ¢ 43T(l) 6 1NOR .. @0 J4-N I I DCPl I 1-FU-82-DAl/1 I I 1-IPS-82-0CPAIAl-A 'Y I OP( I ENG 1 I {PS-39) I ES1AY1 0 I L, §I :'i1 ii'I _, 0 0 g :;'I 0 01 u 1.o 1 r-16 11 -T I I I I 1 -FU-82-DA 1 /1 A DCN2 1 -FU-82-DAl /2 DC LUBE OIL CIRCULATION th e nffecte<l bfo c k o ut c unnt t o th e o ther v w <i co mm o n 0 DlACSN ( 1-FU-275-R75/K4) DSL GEN lA-A REMOTE EMERG START CIRCUIT DSL GEN lA-A COMMON START RELAY CIRCUIT Tiu:-; t o th e o th e r tlu ee Al TEllN.tiT! FOR FRCIM CSST D (43N71*) AtrlN. !YI" ICAL rCA CICT A l PE.:Rt.41T A UTO CLOSlNG Of IEMEFI COt.IT BIJ'S KEY DIAGRAM 69DOV SHUTDOWN BO 1A-A M.AINTENANCE F'DR FRCM 6900\I UMl T Bl) 1e <1-+nn1-1 l (El.ECTRlCAL INTERLOCK sit.tILA FI FOR 80AFllJS U-t,2A-A,.Ze-e> io:mr1r.1 Dt1 itM.ut FOR l l<lt itr10R. fOFI &!Cit T'l'l'PICAL FOR CK T NQFl:MAL FD<< FltOM CSST C (lj.SN710 N O T ES: 1. WHICk ITS ASS CCIAUD LOAD 2* AND U NIJERVOL TAGE RELAY S EE

  • 16. J. ll'USE N U Ml(RS SHOWN IN TABLES SHOIJLCI ee COM IHN£0 WlTH THE 90ARO 'm IM L lSTrn BELO W TO FORMULA T E CO MPLETE UNIQU E r u sE JDENTlFlCATI CN ri.9K'Y $!1TCIN BC.S: 1-F'U-21 1* E'XIWP'Ll'.1

'lllll CHANCE l'O 2 FCR UNI T 2. 4. AUf L R F IUlN NORMAL TO A.L T AT{ COM ORM ER OR l .. 7. AlTE RN.-, TC ANO !'OW C l'I S OU"IC!:S. &. D E L EtED 10. T H[ TERMINAL NUMBERS F08 MCC-IN.T ERNA.L COMPONENTS A RE: THE lIRE: R AT tiCR THAN T H C ACT U AL. 11. MA I NT(NAtrl.CE FitOER BRE',..l<E:RS 171.a, 11111, 17 26 41 1U6 W IL L AOMJ NIST A.ATJ:\l[l"f l C:CKEC " N.ORMAlLY OP£N" DUA: IM G NORMAL O PERATI ON iliNO 11t 1U. BE U'SEC ONLY D Ufl:I NC COLD Sil'IU T COWN rG'I MAINTE N AMCt P U RPOSE

  • lRAMS:FE R nus SOURCE TO ANO T HE:R SOURCE W I L L iE MANUAL. 12. FOR GENERAL NOTES. I NCL U D lNCi TCO!NICA L I NFOKMAltc:

JN AliOUT CIRCUIT O E5rGN, SE!! OR.AllNG 1-4"&W7tiD-O-l. 1J. LI SE Of 1'H[ 69COV lMW F LEX DIESl'L CENE'.R A T ORS: 8YP'l\S:S TH l S PERM ISSIVE. SEE NOTES 16 f. 17, DRA.WINC 1-1 S00-2.

48. Given the following conditions: - Spare charger 6-S has been placed in service to Vital Battery Board I, in accordance with 0-SOI-236.01, Section 8.1.

Subsequently: - The operating crew is directed to place Vi tal Battery V in service to Vital Battery Board II. Which ONE of the following describes the Battery V voltage and which charger is in service? PRIOR to Vital Battery V being placed in service, Vital Battery V voltage will be ____(1)____ than that of Vital Battery II. In accordance 0-SOI-236.05, AFTER Vital Battery V is placed in service on Battery Bd II, 125V VITAL BATTERY CHARGER ____(2)____ will be in service to Vital Battery Board II. NOTE: 0-SOI-236.01, 125V DC Vital Battery Board I 0-SOI-236.01, Section 8.1, 125V Battery Bd I Transfer to Spare 125V DC Vital Battery Charger 6-S

0-SOI-236.05, 125V DC Vital Battery Board V (1) (2) A. higher than II B. higher than V C. the same as II D. the same as V

CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: As seen in 0-SOI-236.05, 125V DC Vital Battery Board V, Vital Battery V has two more cells than the other Vital Batteries and as such operates at a higher voltage. Therefore, it is correct th at prior to Vital Battery V being placed in service, Vital Battery V voltage will be higher than that of Vital Battery II. As seen in 0-SOI-236.05, 125V DC Vital Battery Board V, an assumption is made (prior to placing Vital Battery

V in service to a Vital Battery Board) that either the normal or spare

charger is in service. If no charger is in service, the user of the procedure is directed to use the 0-SOI-236 series to place either the normal or spare charger in service. The reason for this is that (as noted in the procedure): "At no time is Vital Battery Charger V to be used to supply battery system loads." Vital Battery Charger V is not a safety related charger and is not allowed to be connected to an operable DC source (i.e. Battery Board). B. Incorrect: While it is correct that the normal operating voltage of VB V is higher than that of its counterparts, it is not correct that Vital Battery Charger V is allowed to be aligned to Vital Battery Board II. As seen in the simplified drawing in 0-SI-0-3, Weekly Log, it is fully physically possible to align the Vital Battery Charger V to another Vital Battery Board. However, also seen in the same procedure is the note which says that such alignment is not allowed. C. Incorrect: While it is corre ct that Vital Battery Charger II must be used, it is not correct that VB V has the same volt age as the other 4 vital batteries. D. Incorrect: It is incorrect that both the VB V has the same nominal voltage as the other four batteries and that Vital Battery Charger V may be aligned to VBB II.

Question Number: 48 Tier: 2 Group: 1

K/A: 063 D.C. Electrical Distribution K1 Knowledge of the physical connections and/or cause effect relationships between the DC electrical system and the following systems: K1.03 Battery charger and battery Importance Rating: 2.9 3.5

10 CFR Part 55: (CFR: 41.2 to 41.9 / 45.7 to 45.8) 10CFR55.43.b: Not applicable K/A Match: The K/A is matched because the applicant must understand cause effect relationships of the Vital Battery V and Vital Battery Board II in that one must understand that the operating voltage of VB V is higher than that of the normally in service VB II and that the charger normally affiliated with VB V is not allowed to be used with VBB II. Technical

Reference:

0-SOI-236.01, 125V DC Vital Battery Board I 0-SOI-236.05, 125V DC Vital Battery Board V 1-45W703-1 Proposed references to be provided: None Learning Objective: 3-OT-SYS057P, Plant DC Systems

4. EXPLAIN the physical connections and/or cause-effect relationships between the DC Systems and the following systems:
c. Battery charger and battery
9. Given plant conditions, IDENTIFY the applicable DC Systems Precautions and Limitations related to the following: (IER 11-3, Operating the plant with a conservative bias)
a. SOI-236.01 125V DC Vital Battery Board I
b. SOI-239.01 250V Battery Board I c. SOI-100.01 Communications System Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 6 of 57 3.0 PRECAUTIONS AND LIMITATIONS A. All breakers on a board should be OFF or OPEN and all protective grounds must be removed before energizing board. B. A circuit that blows a fuse after being replaced once should have the circuit checked before replacing the fuse a second time, unless the SRO and plant conditions dictate otherwise. C. To prevent explosive mixtures of H 2 and O 2 from accumulating in battery rooms, a battery room exhaust fan should be in service at all times. D. A permanent or portable eyewash stat ion shall be available in the battery rooms. E. Acid spills should be neutralized with a solution of baking soda and water and all traces of the spill should be wiped up. F. There shall be no smoking, open flam es, or arcs in the battery rooms. G. Battery board ground alarm setpoint (90v) is set at 60% of scale. Positive or negative grounds should be maintained as low as possible. H. Do NOT allow continued operation of battery chargers if output voltage exceeds 140 volts DC while supplying plant loads or the output current exceeds 250 amps. Continued operation in excess of these values can cause equipment failures; therefor e, prompt action should be taken to remove the chargers from service. I. If 480V power is removed from the charger for more than 5 minutes, the charger will automatically start up in the equalize mode of operation when power is restored. The Float pushbutton will have to be pressed and released in order to return to the float operating mode. J. When placing the charger in service, transitioning between float and equalize modes, or when making sudden large magnitude load changes, there may be alarm actuations due to momentary voltage fl uctuations on the ch arger. If these alarms occur, the operator will need to PRESS and RELEASE the PRESS TO RESET ALARMS pushbutton and ensure loca l alarms clear before proceeding to subsequent steps. K. Vital Battery V has two more cells t han the other Vital Batteries and operates at a higher voltage. Only Vital Battery Charger V can be used to perform an equalizing charge on Vital Battery V due to the other chargers having a high

voltage cutout function that is lower than this higher voltage. L. At no time is Vital Battery Charger V to be used to supply battery system loads. WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 17 of 57 Date________ Initials 8.0 INFREQUENT OPERATIONS 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger NOTES 1) This section assumes either the normal or sp are charger is in service. If no charger is in service, one must be placed in serv ice per the applicable 0-SOI-236 series instruction. 2) At no time is Vital Battery Charger V to be used to supply battery system loads. [1] ENSURE 125V dc Vit Batt Bd V in service per Section 5.0. ________ [2] ENSURE one exhaust fan is operating. ________ [3] IDENTIFY below which 125V DC Vital Batt Bd to be placed in service with 125V Vital Batt V. (N/A others) Board to be Aligned to Vital Battery V INITIALS125V dc Vital Battery Board I 125V dc Vital Battery Board II 125V dc Vital Battery Board III 125V dc Vital Battery Board IV NOTE If selected board aligned to spare charger then next step may be N/Ad. [4] ENSURE 125V Vital Batt Bd to be placed in service with 125V Vital Batt V is in service. (N/A others) VITAL BATTERY BOARD PROCEDURE INITIALS125V dc Vital Battery Board I 0-SOI-236.01 125V dc Vital Battery Board II 0-SOI-236.02 125V dc Vital Battery Board III 0-SOI-236.03 125V dc Vital Battery Board IV 0-SOI-236.04 [5] PLACE 0-BKR-236-5/CB2, DC OUTPUT, [Vit Batt Chgr V] in OFF. ________ ________ CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 18 of 57 Date________ Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued) [6] PLACE 0-BKR-236-5/CB1, AC INPUT, [Vit Batt Chgr V] in OFF. ________

________ CV [7] PLACE 0-BKR-236-5/102, SUPPLY FROM 125V VIT BATT CHGR V [5th Vital Batt Bd], OFF. ________

________ CV [8] NOTIFY UO prior to switching on Vital Battery Boards. ________ [9] IF normal charger is to remain in service with Vital Battery V, THEN GO TO Step 8.1[12]. ________ NOTE If spare charger in service then nex t steps 8.1[10] and 8.1[11] may be N/A d. [10] PLACE Spare Charger in service using appropriate procedure. (N/A others) Vital Battery Board Procedure Initials 125V dc Vital Battery Board I 0-SOI-236.01 125V dc Vital Battery Board II 0-SOI-236.02 125V dc Vital Battery Board III 0-SOI-236.03 125V dc Vital Battery Board IV 0-SOI-236.04

WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 19 of 57 Date________ Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued) NOTE Vital Batt V is 62 cells with 137 to 140 volts of float voltage [11] PERFORM the following: [11.1] ENSURE 6-S, 7-S, 8-S, or 9-S charger (as appropriate) set to FLOAT. ________ [11.2] RECORD float voltage. _____________VDC. ________ [11.3] IF recorded voltage is NOT 137 to 140 VDC, THEN NOTIFY Electrical Maintenance to adjust voltage to within 137 to 140 VDC. ________ [12] PLACE feeder breaker to 125V DC DIST PNL 5A or 5B in ON position to set up feed to desired Vital Batt Bd (N/A other) [125V Vit Batt Bd V, pnl 2]: Desired Battery Bd Feed Breaker to CLOSE or place in ON PERF INITIALS VERIFIER INITIALS Feed to Battery Bds I and III 0-BKR-236-5/201 CVFeed to Battery Bds II and IV 0-BKR-236-5/202 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 20 of 57 Date________ Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued) CAUTION Breakers in 125V DC DIST PNL 5A or 5B (el 757, outside of Battery Bd rooms II and III) are in OFF position when placed in the UP position. NOTE Closing 0-BKR-236-5/301 or 0-BKR-236-5/302 will cause an alarm on annunciator

point 21-B. Closing 0-BKR-236-5/401 or 0-BKR-236-5/402 will cause an alarm on annunciator point 21-C. [13] PLACE one of the following breakers in 125V DC DIST PNL 5A or 5B [el 757, cabinets outside of Batt Bd rooms II and III] in ON to align feed to desired Vital Batt Bd. (N/A three others). Desired Battery Bd Feed Breaker to CLOSE or place in ON PERF INITIALS VERIFIER INITIALS Feed to Battery Bds I 0-BKR-236-5/301 CV Feed to Battery Bds II 0-BKR-236-5/401 CVFeed to Battery Bds III 0-BKR-236-5/302 CVFeed to Battery Bds IV 0-BKR-236-5/402 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 21 of 57 Date________ Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued) CAUTION Board will be isolated from battery after performance of next step. Do NOT transfer shutdown boards or start large pumps until battery is on board. NOTES 1) Opening breakers in the following step may cause an alarm on annunciator points 17B, 18B, 19B, or 20B. 2) In Modes 1, 2, 3, and 4, TS 3.8.4 should be applicable. In Modes 5 and 6, TS 3.8.5 may be applicable. [14] PERFORM the following: [14.1] ENSURE in-service charger (as appropriate) set to FLOAT. ________ [14.2] RECORD float voltage. _____________VDC. ________ [14.3] IF recorded voltage is NOT 137 to 140 VDC, THEN NOTIFY Electrical Maintenance to adjust voltage to within 137 to 140 VDC. ________ [15] OPEN one of the following to disc onnect normal vital batt from the vital batt bd [Vital Batt Bd - Pnl 1]: (N/A three others) Assoc. Alarm BATTERY BOARD UNID PERF INITIALS VERIFIER INITIALS 17B 125v dc Vit Batt Bd I 0-BKR-236-1/109 CV18B 125v dc Vit Batt Bd II 0-BKR-236-2/109 CV19B 125v dc Vit Batt Bd III 0-BKR-236-3/109 CV20B 125v dc Vit Batt Bd IV 0-BKR-236-4/109 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 22 of 57 Date________ Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued) [16] PLACE one of the following switches to appropriate position: (N/A others) [Pnl 0s in each 125V Vit Batt Bd Rm]

  • 0-XS-236-1/0, MCR AMMETER SELECTOR SWITCH, to BTRY V position.

________ ________ CV

  • 0-XS-236-2/0, MCR AMMETER SELECTOR SWITCH, to BTRY V position.

________ ________ CV

  • 0-XS-236-3/0, MCR AMMETER SELECTOR SWITCH, to BTRY V position.

________ ________ CV

  • 0-XS-236-4/0, MCR AMMETER SELECTOR SWITCH, to BTRY V position.

________ ________ CV [17] ENSURE selected breaker placed in ON position. (N/A others) [Pnl 0 in associated vital batt bd rms] NOMENCLATURE LOCATION UNID PERF INITIALS VERIFIER INITIALS BACKUP SUPPLY FROM

125V VITAL BATTERY BD V 0-DPL-236-1vital batt rm 1 0-BKR-236-1/0 CVBACKUP SUPPLY FROM

125V VITAL BATTERY BD V 0-DPL-236-2vital batt rm 2 0-BKR-236-2/0 CVBACKUP SUPPLY FROM 125V VITAL BATTERY BD V 0-DPL-236-3vital batt rm 3 0-BKR-236-3/0 CVBACKUP SUPPLY FROM

125V VITAL BATTERY BD V 0-DPL-236-4vital batt rm 4 0-BKR-236-4/0 CV WBN Unit 0 125V DC Vital Battery Board V 0-SOI-236.05 Rev. 0002 Page 23 of 57 Date________ Initials 8.1 Placing Vital Battery V in service on Battery Bd I, II, III or IV Using Normal or Spare Charger (continued) NOTE In Modes 1, 2, 3 and 4, TS 3.8.4 may be exited. In Modes 5 and 6, TS 3.8.5 may be exited. [18] CHECK current flow toward Vital Battery V (negative or zero) to ensure normal or spare Vital Charger is carrying load and

Vital Battery V tied to board (selected Battery Board ammeter). ________ NOTES 1) The Battery Board is now connected to Vita l Batt V. Transfers of Shutdown Boards or starting of large pumps may be resumed as necessary. 2) The normal Vital Battery is now disconnec ted without a charger. If a spare battery charger is in service to the 5th Vital Ba ttery, then the normal charger should be placed on the normal vital battery unless a condition exists which would prohibit the battery from remaining on a charger. Sections 8. 5 through 8.8 should be used for placing the normal battery charger in service to its disconnected vital battery. End of Section WBN Unit 0 Weekly Log 0-SI-0-3 Rev. 0057 Page 21 of 36 Appendix A (Page 10 of 24) INITIALS OF DATA COLLECTOR: DATE REMARKS: 125V VITAL BATT BD V NO (101) 125V VITAL BATT V NO (202) NC (201) NO (301) NO (302) NO (401) NO (402) NO (4/0) NO (2/0) NO (3/0) NO (1/0) 125V DC DIST PNL 5A 0-DPL-236-1 125V VITAL BATT BD IV 125V VITAL BATT BD II 125V VITAL BATT BD III 125V VITAL BATT BD I 125V DC DIST PNL 5B 0-DPL-236-3 0-DPL-236-2 0-DPL-236-4 SR 3.8.9.1 SR 3.8.10.1 NOTE 1: Only one battery board may be feed from the 5th battery board at any one time. X A ICLOSEDAVAILABLEINOPERABLE (102) NC 125V Vit Chgr V 480V Aux Bldg Com MCC C 0-BKR-236-5 CHG V NC CB1 NC CB2 NC NOTE 2: 0-BKR-236-5/102 must be OPEN when 5th battery is inservice for battery bd I, II, III, or IV See note 2

49. Given the following conditions:
- Unit 1 is at 100%.

HANDSWITCH ON 0-M-26

Which ONE of the following describes the Emergency Diesel Generators response to depressing the handswitch shown above and the Essential Raw Cooing Water (ERCW) alignment?

Depressing the switch wil l start ____(1)____ EDG(s).

ERCW ____(2)____.

A. (1) ALL (2) AUTOMATICALLY aligns to the starting DG B. (1) ALL (2) is NORMALLY flowing to the standby DGs C. (1) ONLY ONE (2) AUTOMATICALLY aligns to the starting DG D. (1) ONLY ONE (2) is NORMALLY flowing to the standby DGs CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: While it is correct that E RCW would be automatically aligned to a DG, it is not correct but plausible that the handswitch depicted would start all of the DGs. B. Incorrect: It is not correct that all of the DGs will start when 1-HS-82-16A is depressed. It is plausible to believe that becaus e there are MCR hand-switches (1-HS-82-15 and 2-HS-82-15) which will cause all four

of the DGs to emergency start. This is seen on print 1-45W760-82-6. It is also plausible to believe this because on this same print, 1-HS 16A appears in-line with the other causes of a common emergency start of the DGs; howev er, it appears downstream of the branch to ES1AY1 and as such will not cause the common start. Also, it is Incorrect and yet plausible that ERCW would be normally flowing to the standby DGs. C. Correct: It is correct that ERCW automatically aligns to t he DGs and that only one DG will start when 1-HS 16A is depressed. D. Incorrect: As seen on print 1-45W760-82-6, it is correct that depressing 1-HS 16A will only emergency start one dies el generator (specifically, the 1A-A). Because 1-HS-82-16A opens the current flow path to relay ES1AY after the branch to re lay ES1AY1, the common emergency start of the diesel generators is not actuated. It is not correct that ERCW is normally flowing to the standby DGs. As seen in 0-SOI-82.01, Diesel Generator (DG) 1A-A precaution and limitation HH: ERCW automatically aligns to a DG at 40 rpm. It is plausible to believe that ERCW would be normally flowing to the standby DGs as

this was the case before Design Change 56341 was implemented in November of 2013. This design change considered the overall two unit flow balancing requirements for ERCW and required that the

ERCW supplies to the DGs be isolat ed when the DGs were in standby.

Currently, ERCW to the DGs is autom atically aligned as the DG comes to speed and is manually secured as part of the standby alignment.

Question Number: 49 Tier: 2 Group: 1

K/A: 064 Emergency Diesel Generators (ED/G) A4 Ability to manually operate and/or monitor in the control room: A4.06 Manual start, loading, and stopping of the ED/G Importance Rating: 3.9 3.9

10 CFR Part 55: (CFR: 41.7 / 45.5 to 45.8) 10CFR55.43.b: Not applicable

K/A Match: K/A is matched because the applicant is required to understand which hand-switches in the MCR will manually start all of the DGs and which will manually start only one DG. The applicant must also be able to monitor the proper operation of the DG cooling water (ERCW) during a manual diesel start. Technical

Reference:

SOI-82.01, Diesel Generator (DG) 1A-A revision 0090 (HISTORIC) 0-SOI-82.01, Diesel Generator (DG) 1A-A

1-45W760-82-6 Proposed references to be provided: None Learning Objective: 3-OT-SYS082A, DIESEL GENERATOR OVERVIEW 1. EXPLAIN the DG design features and/or operational interlocks that provide the following: (IER 11-3: Having a Solid Understanding of Plant Design and System Interrelationships)

a. Normal & Emergency starts (automatic and

manual)

b. DG control during emergency start Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

§ 8 I !'.! G Q 2 1-XS-6K -.B.:111__ (1-FU-211-A/C2} K609o 1-R-48 , 2 CS1} (1-R-51,TRB) C 1-FU-278-L 11 A/K5) (1-FU-211-A/81) D1AR4 1-M-1 1'A 1'A ¢ ! R12 R12! A HS-82-17Ai EMER STOP i1-HS-82-17C (0-M-26) R11 R1! (O-l-4 ) @ 11K 11L PJ JJ-N 1-XS-82-121 AUX CONT RM 2 18 Note that l-HS-82-16A appears downstream of the branch wire to ESlAYl. Therefore , depressing l-HS-82-16A will 01tly start one diesel (in this case the lA-A). A RlZl1-HS-82-16C 6 EMERG START (0-L-4) 8 g > 3 SHUTDOWN RELAY RESET 4 D1ARN8 { 1-FU-278-L 11A/K4l 812 11-XS-82-125 AUX CONT RM 811 = All = DIAL 1 ( 1-FU-27 8-L 11 A/KG l DlAGNB EMERGENCY STOF' 5 R G 01AL2 6 0 D1AEP ( 1-FU-275-R7 J/L9) A 81 HD I D ,m \__ EMERG START (1-M-1) 0 01AEN (1-F'U-275-R73/L

10) 2-HS-82-15 EMERG START (2-M-1 l A 1 D 1 CES
  • 2 2BY 8 l-HS-8"-1 5 B part o f th e co mm o n e mer ge n cy s tart and as s u c h w ill ii ca u.-,e all fo ur DGs t o E:: s tart (\-FU-27 0 (1-f"U-27 DSL GEN TA-A REMOTE EMERG START CIRCUIT DSL GEN 1 A-A Cati 7 8 WBN Unit 0 Diesel Generator (DG) 1A-A 0-SOI-82.01 Rev. 0004 Page 12 of 104 3.0 PRECAUTIONS AND LIMI TATIONS (continued)

FF. The DG may surge from 400-450 RPM after an Idle Start, which is to be expected and does NOT have any adverse consequences. GG. The fuel oil transfer pumps will auto matically start when the gauge indicates between mark 6-9 (out of the 16 marks for tank gauge). HH. ERCW automatically aligns to a DG at 40 rpm. If alignment of ERCW CANNOT immediately be obtained, the DG should be Emergency Stopped. II. Alarm window 225-E TR A/B ERCW TO C&SS COMPR FLOW HI may alarm on some DG starts due to high diffe rential between the two ERCW headers. WBN Unit 1 Diesel Generator (DG) 1A-A SOI-82.01 Rev. 0090 Page 2 of 103 Revision Log Revision or Change Number Effective Date Affected Page Numbers Description of Revision/Change 81 10/04/12 2, 68 Incorporated changes associated with DCN #58383 by adding a new DG Emer gency Stop/Reset Transfer switch 1-XS-82-125, which is located on

1-PNL-276-L011A in ACR. 82 02/05/13 2, 13, 42, 43, 95 Added steps to use strobotac for failure of single Air

Intake Damper group and commitment statement including WBPER458545.[PER598595-001] 83 03/25/13 2, 12 Added P&L GG to clarify when the DG transfer pump will automatically st art. [PER 620257-001] 84 06/07/13 2, 88-90 PER 699018: Add sections to shutdown and restore DG room exhaust fans as part of a tagout. 85 08/22/13 2, 91, 92 Added Section 8. 15 for manual exhaust fan start in low temperature conditions. Requested by Work

Control for PMT perform ance following clearance. 86 11/18/13 2, 12, 29, 42 Added P&L and step to alert operator that

1-FCV-67-66 will automatically open at 40 rpm and

emergency stop required if this does not occur. [DCN56341] 87 11/27/13 2, 29 Added step to cl ose 1-FCV-67-66 for standby alignment. 88 02/27/14 2, 13, 27-35 Minor/editorial revision to ad notes regarding the

effects on ERCW header pressure from the new DG ERCW alignment. [PER 817463-001] [PER 826190-

001] Moved IV steps to end of 5.2 Verification of Standby Alignment to help prev ent human error traps. 89 04/06/14 2, 93 Revised note to clarify normal position of Exhaust Fan 1 and 2 iaw 1-SI-0-2 se ries. [PER 827633-001] 90 ALL Cancelled to 0-SOI-82.01, Rev 0. ////

50. Given the following timeline:

00:00:00 Unit 1 is at 100% power. 00:0 1 :00 An error by the Protective Relay te st group causes the 1A-A 6.9kV SDBD to experience a Blackout. 00: 1 0:00 In accordance with 0-SOI-82.01, section 8.3.1, an operator is preparing to parallel the 1A-A EDG to offsite power. 1-XI-82-1, TRAIN 1A-A SYNCHROSCOPE is rotating CLOCKWISE as shown below. The synchroscope has NOT completed a full revolution.

Synchroscope at 00: 1 0:00 Synchroscope at 00: 1 0:07 Which ONE of the following describes the direction and speed of the synchroscope rotation? The direction of synchroscope rotation is ____(1)____ AND the speed of synchroscope rotation is ____(2)____. NOTE: 0-SOI-82.01, Diesel Generator (DG) 1A-A,Section 8.3.1, Removing DG from Service-DG Tied on to SD Bd REFERENCE PROVIDED A. (1) CORRECT (2) CORRECT B. (1) CORRECT (2) TOO FAST C. (1) INCORRECT (2) CORRECT D. (1) INCORRECT (2) TOO FAST CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: As seen in step [15] of section 8.3.1 of 0-SOI-82.01, the operator is to ADJUST 1-HS-82-13, SPEED CONTROL, [0-M-26] to obtain desired clockwise rotation (15 or more seconds) on 1-XI-82-1, TRAIN 1A-A SYNCHROSCOPE. Therefore, the direction of synchroscope rotation is correct. The two pictures of the synchroscope depict two states of the instrument over a 7 second span. Because the needle has travelled about 120°, the total time for one rotation of the meter is 360°/120° x 7 seconds = 21 seconds. Even if one assumes that the needle has travelled 150°, the time for one rotation becomes 16.8 seconds. Because the rotation is longer than 15 seconds, it is acceptable. The actual angle of displacement of the

needle is just at 135°. B. Incorrect: Again the direction of synchroscope rotation is correct. However, the rotation of the meter is not too fast. It is plausible to believe that it is if one calculates the speed of rotation based upon a counterclockwise rotation of the synchroscope. If one did so one would determine that the speed for rotation is 10.5 seconds per rotation (based upon an interpretation that the synchroscope needle rotated 240° in 7 seconds) or 12.0 seconds (based upon an interpretation that the synchroscope needle rotated 210° in 7

seconds). C. Incorrect: Again, it is incorrect that the direction of rotation is incorrect. However, as demonstrated the speed of rotation is correct. D. Incorrect: It is both incorrect that the direction of rotation is incorrect and that the speed

of rotation is too fast.

Question Number: 50 Tier: 2 Group: 1

K/A: 064 Emergency Diesel Generator (ED/G) System A4 Ability to manually operate and/or monitor in the control room: A4.09 Establishing power from the ring bus (to relieve ED/G)

Importance Rating: 3.2 3.3

10 CFR Part 55: (CFR: 41.7 / 45.5 to 45.8)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to monitor the proper conditions needed to bring the normal offsite power supply to the 6.9kV shutdown board back into service (i.e. relieving the ED/G) following a blackout condition. . Technical

Reference:

0-SOI-82.01, Diesel Generator (DG) 1A-A Proposed references to

be provided: Redacted page 61 of 0-SOI-82.01 Learning Objective: 3-OT-SYS082A, DIESEL GENERATOR OVERVIEW 8. Given plant conditions, IDENTIFY the applicable Diesel Generator Precautions and Limitations contained in SOI-82 series. Cognitive Level: Higher X Lower Question Source: New Modified Bank X Bank Question History: Modified bank Question (original Q was SYS082I.04). The original Q started with offsite power on the SDBD. This Q starts with the DG as the only power source on the SDBD. Comments:

WBN Unit 0 Diesel Generator (DG) 1A-A 0-SOI-82.01 Rev. 0004 Page 61 of 104 Date________ INITIALS 8.3.1 Removing DG from Service-DG Tied on to SD Bd (continued) CAUTION When adjusting speed and voltage, care must be taken to prevent overshooting desired values. Voltage control response is approxim ately five times faster than speed control response. [12] MATCH generator Incoming Fr equency (1-XI-82-2) with Running Frequency (1-XI-82-3) using 1-HS-82-13, SPEED CONTROL [0-M-26]. ________ [13] MATCH running voltage (1-EI-82-5) to incoming voltage (1-EI-82-4) with 1-HS 12, VOLTAGE REGULATOR [0-M-26]. ________ [14] PLACE 1-HS-82-18, DG MODE SELECTOR, in PARALLEL [0-M-26]. ________ [15] ENSURE DG Frequency and Voltage are MATCHED with 6.9 kV SD Bd, AND ADJUST 1-HS-82-13, SPEED CONTRO L, [0-M-26] to obtain desired clockwise rotation (15 or more seconds) on

1-XI-82-1,TRAIN 1A-A SYNCHROSCOPE. ________ ________ CV [16] IF a white breaker disagreement light is lit for any of the following handswitches, THEN CLEAR the disagreement by taking the handswitch to TRIP. (N/A handswitches with no disagreement) NOMENCLATURE LOCATION POSITION UNID PERF INITIAL 1716-NORMAL FROM CSST C 0-M-26 TRIP 1-HS-57-41B 1932 ALTERNATE FROM CSST D 0-M-26 TRIP 1-HS-57-97B 1718-MAINTENANCE FROM 6.9 UNIT BD 1B 0-M-26 TRIP 1-HS-57-44B QUESTIONS REPORTfor ILT EXAM BANK NOVEMBER 2013 CLOSED

1. Given the following timeline:0000 - Unit 1 is at 100% power.0001 - In accordance with section 8.3.2 of SOI-82.01, an operator is preparing toparallel the 1A-A DG to the 1A-A 6.9kV SDBD. He notes that the 1A-A DG'ssynchroscope is rotating clockwise as shown below. He notes that thesynchroscope has NOT completed a full revolution. Synchroscope at 0001:00 Synchroscope at 0001:05Which ONE of the following completes the statements below:

The direction of synchroscope rotation is (1) 

 .The speed of synchroscope rotation is (2)     
 .Note:  SOI-82.01, "Diesel Generator (DG) 1A-A" Section 8.3.2, "Removing DG from Service-DG NOT Tied on to SD Bd"REFERENCE PROVIDED (1)    (2)  correct    acceptable  correct  unacceptable incorrect    acceptable incorrect  unacceptable
a. b. c. d. Thursday, October 30, 2014 9:54:19 AM 1

WBN Unit 0 Diesel Generator (DG) 1A-A 0-SOI-82.01 Rev. 0004 Page 61 of 104 Date________ INITIALS 8.3.1 Removing DG from Service-DG Tied on to SD Bd (continued) CAUTION When adjusting speed and voltage, care must be taken to prevent overshooting desired values. Voltage control response is approxim ately five times faster than speed control response. [12] MATCH generator Incoming Fr equency (1-XI-82-2) with Running Frequency (1-XI-82-3) using 1-HS-82-13, SPEED CONTROL [0-M-26]. ________ [13] MATCH running voltage (1-EI-82-5) to incoming voltage (1-EI-82-4) with 1-HS 12, VOLTAGE REGULATOR [0-M-26]. ________ [14] PLACE 1-HS-82-18, DG MODE SELECTOR, in PARALLEL [0-M-26]. ________ [15] ENSURE DG Frequency and Voltage are MATCHED with 6.9 kV SD Bd, AND ADJUST 1-HS-82-13, SPEED CONTRO L, [0-M-26] to obtain desired clockwise rotation (15 or more seconds) on

1-XI-82-1,TRAIN 1A-A SYNCHROSCOPE. ________ ________ CV [16] IF a white breaker disagreement light is lit for any of the following handswitches, THEN CLEAR the disagreement by taking the handswitch to TRIP. (N/A handswitches with no disagreement) NOMENCLATURE LOCATION POSITION UNID PERF INITIAL 1716-NORMAL FROM CSST C 0-M-26 TRIP 1-HS-57-41B 1932 ALTERNATE FROM CSST D 0-M-26 TRIP 1-HS-57-97B 1718-MAINTENANCE FROM 6.9 UNIT BD 1B 0-M-26 TRIP 1-HS-57-44B

51. Given the following conditions: - CASK DECON COLL TANK(CDCT) LEVEL HI (0-L-2D-9) is LIT. The following windows alarm and are acknowledged:

Which ONE of the following comp letes the statements listed below?

____(1)____ would cause the annunciators shown above. In accordance with 0-SOI-77.01, Liquid Waste Disposal, the high level in the CDCT ____(2)____ A. (1) High activity in the liquid rad waste piping (2) is unable to be REDUCED unt il 0-RM-90-122 is restored B. (1) A loss of power to the radiation monitor (2) is unable to be REDUCED unt il 0-RM-90-122 is restored C. (1) High activity in the liquid rad waste piping (2) can be REDUCED after fuses are remov ed and a temporary jumper is installed D. (1) A loss of power to the radiation monitor (2) can be REDUCED after fuses are remov ed and a temporary jumper is installed CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: As seen in ARI-180-187, Common Radiation Detectors, both the loss of power to the rate meter and the loss of signal from the detector will cause Annunciator 181-C, WDS RELEASE LINE 0-RM-122 INSTR MALF to alarm. Also contained in this alarm response instruction is that the causes for Annunciator 181-A, WDS RELEASE LINE 0-RM-122 LIQ RAD HI are as follows: A. High activity in release line B. Loss of power to control relay (WDRS) C. Loss of power to rate meter. D. Background radiation rise at monitor To further develop the causes of Annunciator 181-A, one may refer to the vendor manual for the readout modules manufactured by Sorrento Electronics, WBN-VTD-G292-1150. The radiation monitor instrument loop contains (in order of signal progression) a detector, preamplifier, and readout module. The preamplifier delivers a sequence of negative pulses to the readout module. These negative pulses are filtered by the readout module and then converted into a dc current which is proportional to the counting rate of input pulses. Finally, the dc current is fed to a meter amplifier circuit which provides a current to voltage transfer function. This voltage is fed into the summing junction of an operational amplifier which controls the trip circuit of the readout module. In the foregoing, one may observe that if the signal from the detector is lost, the sequence of negative pulses will not exist. Because the pulse rate is 0, the dc current which is generated will be 0. As a result, the voltage fed to the summing junction of the operational amplifier will be 0 and will never reach the high radiation trip setpoint. Therefore, a loss of signal from the detector cannot cause a High Radiation alarm. It is plausible to believe that a loss of signal would place the readout module's Bistable Trip Circuits into their tripped state because of the existence of the Operate (Instrument Malfunction) circuit. If the detector signal should fail, the Operate relay of the readout module, K101 will de-energize causing its contacts to change state. The applicant may therefore believe that the trip bistables are affected in the same mode as the operate relay by a loss of signal. It is also Incorrect that the high level in the CDCT would be unable to be corrected until the radiation monitor is restored. Section 8.16, Release of the Cask Decontamination Collector Tank to Cooling Tower Blowdown of 0-SOI-77.01 contains step [3] which takes action IF 0-RE-90-122 is INOPERABLE. These actions include: REMOVE fuses N-17 and N-18 in Panel 1-R-76. PLACE temporary jumper (clip-type) at TB-248 across terminals 8 and 10 in Panel 1-R-72. It is plausible to believe that the high level cannot be corrected until the radiation monitor is restored because there are procedures such as SOI-77.02, Waste Gas Disposal System which do not contain steps for the conditions listed in the question. During a gas tank release, 0-RM-90-118 monitors the effluent. If 0-RM-90-118 were to lose power, the release would be terminated (tantamount to the response of the liquid radioactive waste system). The steps in SOI-77.02 in part state: NOTIFY Instrument Maintenance to adjust setpoint of 0-RE-90-118 to avoid inadvertent isolation of release. This step would not allow the gas decay tank release to continue. B. Incorrect: It is correct that a loss of power to the radiation monitor would produce the annunciators depicted. It is not correct and yet plausible that a procedural compliant liquid release would be impossible until the radiation monitor was restored. C. Incorrect: As mentioned it is Incorrect and yet plausible that a loss of signal would produce a High Radiation alarm. It is correct that steps within 0-SOI-77.01 could be used to compensate for the loss of the radiation monitor.

D. Correct: As discussed, a that a loss of power to the radiation monitor would produce the annunciators depicted. It is correct that steps within 0-SOI-77.01 could be used to compensate for the loss of the radiation monitor.

Question Number: 51 Tier: 2 Group: 1

K/A: 073 Process Radiation Monitoring (PRM) System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations: A2.01 Erratic or failed power supply Importance Rating: 2.5 2.9

10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to first predict that a loss of power to the 0-RM-90-122 monitor has on that monitor's output to the RONAN system. Subsequently, the applicant must be able to use 0-SOI-77.01 to mitigate the consequence imposed by the loss of power and correct the high level existent in a liquid radioactive waste tank. Technical

Reference:

SOI-77.02, Waste Gas Disposal System ARI-0-L-2D, Liquid Waste Panel 0-SOI-77.01, Liquid Waste Disposal ARI-180-187, Common Radiation Detectors WBN-VTD-G292-1150, Vendor Manual for the General

Atomics Radiation Analyzer Readout Module Proposed references to

be provided: None Learning Objective: 3-OT-SYS090A, Radiation Monitoring 3. Given plant conditions, DETERMINE if any of the

following Radiation Monitoring System alarms would be present and the actions required by the ARI: (IER 11-3, Monitoring plant indications and controls closely)

b. [0-XA-55-12B, ALL] All annunciators on COMMON RADIATION DETECTORS panel from ARI-180-187Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Common Radiation Detectors ARI-180-187 Rev. 0034 Page 9 of 46 A. High activity in release line B. Loss of power to control relay (WDRS) C. Loss of power to ratemeter D. Background radiation rise at monitor NOTE 0-RM-90-122 has associated ICS computer point R1022A. [1] ENSURE 0-RCV-77-43, WASTE CONDENSATE DISCHARGE RADIATION ISOLATION VALVE closed. [2] CHECK 0-RE-90-9 (ARM) [3] NOTIFY Chemistry to perform CM-9.09 "Effluent Radiation Monitor Alarm Response Guidelines". [4] NOTIFY Radiation Protection to investigate alarm. [5] IF release was in progress, THEN EVALUATE discharge lineup. [6] REFER TO AOI-31, Abnormal Release Of Radioactive Material. [7] REFER TO EPIP-1. 1-45W600-57-27 1-45W600-77-1

1-47W610-90-2 AOI-31 EPIP-1 CM-9.09 Source Setpoint 0-RM-90-122A determined by Chemistry 181-AWDS RELEASE LINE 0-RM-122 LIQ RAD HI (Page 1 of 1) Probable Cause: Corrective Action:

References:

WBN Unit 1 Common Radiation Detectors ARI-180-187 Rev. 0034 Page 11 of 46 A. Loss of signal from detector B. Failure of readout module's internal power supply C. Loss of power to ratemeter D. Loss of power to control relay E. Function switch in ALARM ADJ [1] ENSURE 0-RCV-77-43, WASTE CONDENSATE DISCHARGE RADIATION ISOLATION VALVE, CLOSED. [2] CHECK 0-RM-90-122A and 0-RR-90-122 [0-M-12] OR ICS computer points R1022A. [3] NOTIFY Chemistry to perform CM-9.09 "Effluent Radiation Monitor Alarm Guidelines". [4] NOTIFY Instrument Maintenance to investigate alarm. [5] REFER TO ODCM Section 1/2.1.1 AOI-31 1-45W600-57-18

1-45W600-77-1 CM-9.09 ODCM Source Setpoint 0-RM-90-122A Instrument Failure 181-CWDS RELEASE LINE 0-RM-122 INSTR MALF (Page 1 of 1) Probable Cause: Corrective Action:

References:

WBN Unit 0 Liquid Waste Dispo sal 0-SOI-77.01 Rev. 0006 Page 78 of 166 Date ________ 8.16 Release of the Cask Decontaminat ion Collector Tank to Cooling Tower Blowdown NOTES 1) The Unit SRO should confirm there are no planned interruptions in Hydro Plant operation which would result in isolation of normal CTBD diffuser flow, in order to minimize transportation of contaminants to the yard holding pond. 2) MI-0.018, NSSS FILTER CHANGEOUT, may be referenced if 0-RM-90-122 alarms on hi rad levels with filters installed on CDCT system. [1] ENSURE Section 8.15 complete. ________ [2] IF 0-RE-90-122 is OOS, THEN (N/A part of this Step NOT performed): PERFORM 0-RE-90-122 INOPERABLE steps, OR ________ REQUEST Instrument Maintenan ce (IM) to ENSURE 0-RE-90-122 properly a ligned for service. [C.8] ________ [3] IF 0-RE-90-122 is INOPERABLE, THEN PERFORM the following: [3.1] REMOVE fuses N-17 and N-18 in Panel 1-R-76. ________ ________ CV [3.2] PLACE temporary jumper (clip-type) at TB-248 across terminals 8 and 10 in Panel 1-R-72. ________ ________ CV [3.3] IDENTIFY jumper with an Information Tag noting procedure and step number. ________ WBN Unit 1 Waste Gas Disposal System SOI-77.02 Rev. 0040 Page 60 of 72 Date________ INITIALS 8.9 Gas Decay Tank Release Followed by Nitrogen Header Purge (continued) [13] OBTAIN Decay Tank Release Permit (0-ODI-90-5) from Chemistry, AND COMPLETE in conjunction with this Instruction. ________ NOTE IV for Step [14] may be performed by Chemistry. [14] ENSURE 0-ODI-90-5 Release Permit approved by SM/SRO. ________ ________ IV [15] OBTAIN SRO approval and verification that release is authorized, and instructions are correct for release of the applicable Gas Decay Tank. ________

SRO [16] IF 0-RE-90-118 is INOPERABLE, THEN ENSURE Chemistry Countroom is performing compensatory measures,

AND NOTIFY Instrument Maintenance to adjust setpoint of 0-RE-90-118 to avoid inadverte nt isolation of release, AND PERFORM Independent Verification fo r Steps [19], [20], and [21]. [C.4] ________ CAUTION ABGTS should NOT be operated for testing purposes or for releases of waste decay tank at the same time both containment purge exhaust fans are in operation. Operation of one purge exhaust fan with the ABGTS is acceptable. [17] START ABGTS Fan A-A per SOI-30.06. ________

52. Given the following conditions:
- Unit 1 is at 100%.

- The following indication is observed on 0-M-27B. Which ONE of the following describes the cause of the above indications?

A Loss of the ________ ERCW header.

A. 1A B. 1B C. 2A D. 2B CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: The closed cooling water system is called Component Cooling Water (CCS) at WBN. The service water system at WBN is called Essential Cooling Water System (ERCW). CCS has three heat exchangers which are all cooled by ERCW. The A trained CCS Hxs are A and B (even though the B nomenclature would imply affiliation with the B train). The B trained CCS HXs is therefore C. The A and B CCS Hx are normally cooled by ERCW header 2A and alternately cooled by ERCW 1A. The C CCS Hx is normally cooled by ERCW header 2B and alternately cooled by ERCW header 1B. The isolations between the same trained ERCW headers are locked shut in accordance with the fire protection report. Use of the alternate headers will therefore incur Operating Requirements (ORs). Use of the alternate headers will also incur required actions in accordance with the technical specifications as two ERCW headers will be cross connected and thus rendering both inoperable. These alternate headers are used during outages and the abnormal operating procedures. All of the aforementioned may be found on print 1-47W845 sheets 2 and 5 (for two notes). It is plausible to believe that the 1A ERCW header supplies the B CCS Hx because even though it normally is not aligned, it physically can supply the Hx B. Incorrect: The 1B ERCW header does not supply the B CCS Hx. It is plausible to believe such because as discussed the very name of the B CCS Hx leads one to believe that it is affiliated with the B train and thus cooled by a B

trained ERCW header. C. Correct: The 2A ERCW header does supply the B CCS Hx. D. Incorrect: The 2B ERCW header does not supply the B CCS Hx. It is plausible to believe such because as discussed the very name of the B CCS Hx leads one to believe that it is affiliated with the B train and thus cooled by a B trained ERCW header. Furthermore, it is plausible to believe this if one remembers that Unit 2's ERCW supplies the CCS Hxs.

Question Number: 52 Tier: 2 Group: 1

K/A: 076 Service Water System (SWS) K3 Knowledge of the effect that a loss or malfunction of the SWS will have on the following: K3.01 Closed cooling water

Importance Rating: 3.4* 3.6

10 CFR Part 55: (CFR: 41.7 / 45.6)

10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to associate a trend observed on a CCS Hx's temperature recorder with the appropriate malfunction of the ERCW system. Technical

Reference:

1-47W845-2 1-47W845-5

Proposed references to

be provided: None Learning Objective: 3-OT-STG-070A, COMPONENT COOLING WATER SYSTEM

5. EXPLAIN, the physical connections and/or cause-effect relationships between the CCS and the following systems:
c. ERCW Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

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t N-SEE .._ _____

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  • 1-67-.:5 388 i O .9.6' ;, N L OSE[). T KE'I' AR E OPEN.EC f OR 1H OC ICE IMJE C'\'IDN O R OTHER M E AC HV IT lES. MUST e T O T ME H T HE CP PR IOR TO I PlNG SY S fE'.M MOOES 5
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  • RIMS 826 '89 10 1$ J 7.5 S.W. H U NCATE RE V IEWER 2 D.l. BE<LS 3 r .A. KOOM TZ CAD MAINTAINED DRAWING I ( CONFIGURATION CONTROL A E
53. Given the following timeline:

00:00:00 Unit 2 is at 100% power. 00:0 1:00 The station's Non-essential control air is lost and Non-essential air header pressure is 0 psig.

Which ONE of the following describes the response of the Unit 2 MSIVs to this condition?

At 00:0 1:0 1, the Unit 2 MSIVs will ________. A. have CLOSED B. remain OPEN as the MSIVs are provided with essential control air C. remain OPEN until the MSIVs air accumulator's inventory is depleted D. remain OPEN as a N 2 backup is provided to the MSIVs' instrument air supply

CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: In accordance with WBN system description NPG-SDD-WBN2-1-4002, An air accumulator is provided at each MSIV air supply line to prevent MSIV closure on loss of control air. [ page 55 of NPG-SDD-WBN2-1-4002] This distractor is plausible from two perspectives. Firstly, if the applicant simply fails to recall that the MSIVs possess an air accumulator then the applicant would deduce that a instrument air pressure of 0 psig would result in the MSIVs travelling to their failed position of closed. Secondly, it is a common misconception to believe that the MSIVs have actuators which are similar to

other common Air Operated Valves. Typical AOVs have an air admission port and a bleed orifice. A continuous supply of air is required to maintain the AOV in the desired position (otherwise the bleed orifice will depressurize the valve operator). The MSIV has a pneumatic cylinder for an operator. Air is admitted via a solenoid valved line and air may be vented off via a solenoid valved line. If the air supply is lost and the vent line is not opened, the valve will not change state as there is no exit path for the contained instrument air. B. Incorrect: As seen on print 2-47W848-9, the MSIVs (shown on the print as FCV-1-11 and FCV-1 the other two MSIVs are depicted on 2-47W848-6) are afforded Non-essential control air. It is plausible to believe that these valves be afforded essential control air as are many other air operated valves in the plant. Amplifying the plausibility of this is that the MSIVs fulfill a T/S LCO; that being, T/S LCO 3.7.2, MSIVs. The applicant may assume that due to the fact that the valves are safety related that their air supply must be safety

related. C. Correct: Given the information presented in the analysis for distractor A, one may understand that the MSIVs will remain open until their air accumulators inventory deplete due to operating cylinder leakage. (which as seen in 0-MI-1.002, Main Steam Isolation Valve Maintenance is miniscule). D. Incorrect: As seen on prints 2-47W848-9 and 2-47W848-6 there is no backup N2 supply to the instrument air for the MSIVs. It is plausible to believe that one does exist because several valves impacting the steam generators do receive a backup N2 supply. Examples of such are SG PORVs and AFW LCVs and PCVs. Additional strengthening to the plausibility for the existence of a backup N2 supply is that U2 was selected for the Unit affected. More air operated valves are receiving backup N2 supplies on U2 than on U1. Most of these backups are being provided due to either Appendix R analysis or Fukishima response.

Question Number: 53 Tier: 2 Group: 1

K/A: 078 K1.05 078 Instrument Air System (IAS) K1 Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: K1.05 MSIV air

Importance Rating: 3.4* 3.5*

10 CFR Part 55: CFR: 41.2 to 41.9 / 45.7 to 45.8

10CFR55.43.b: Not applicable K/A Match: K/A is matched because it requires the applicant to both understand the physical connection of the instrument air system to the MSIVs and to comprehend the cause-effect relationship that a loss of the instrument air system would have on the MSIVs (and the MSIVs' air accumulators). Technical

Reference:

2-47W848-6 2-47W848-9 NPG-SDD-WBN2-1-4002 0-MI-1.002 T/S LCO 3.7.2 Proposed references to

be provided: None Learning Objective: 3-OT-STG-001A, MAIN STEAM 04. EXPLAIN the physical connections and/or cause-effect relationships between the Main Steam System and the following systems:

k. Instrument Air Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

NPG System MAIN STEAM SYSTEM NPG-SDD-WBN2-1-4002 Description Rev.0004 Document Paae 55of118 3.3.2 MSIV Controls (continued) The redundant exhaust valves are energize-to-close and they are arranged in parallel paths so that one train of the exhaust valves needs to be deenergized to exhaust air from the operator. The exhaust valves in a train are arranged in series to preclude inadvertent closing of the MSIV. The test valves are energize-to-close and they are normally opened (deenergized) for exhaust air from the operator to go through a low resistance path for MSIV quick closing. By closing a test exhaust valve and an exhaust valve in the same train, exhaust from MSIV operator will be routed through a high resistance flow path for slow MSIV closing. All solenoid valves, except the test solenoid valves, are energized to open the MSIV. Redundant sets of solenoid valves are independently operated. Power supplies for Unit 1 train A and train B solenoid valves come from 125V DC Vital Battery Boards I and II respectively (Table 2.2). Similarly, power supplies for Unit 2 train A and train B solenoid valves come from the 125V DC Vital Battery Boards Ill and IV, respectively (Table 2.2). The supply solenoid valve, one exhaust valve and the test solenoid valve in the same train are powered from a fused circuit of the applicable battery board. The second exhaust solenoid valve is powered from another fused circuit of the same board. Separate fused power supply circuits for the two exhaust valves are provided to prevent MSIV closure in the event of a power supply circuit fault. Control air (System 32) is used to supply all MSIV operations. An air accumulator is provided at each MSIV air supply line to prevent MSIV closure on loss of control air. The control air is a quality air which has been filtered and dried to assure that the solenoid valves will not be degraded by poor air quality and failed to respond on deenergization. The MSIV is provided with position switches to indicate its fully closed, 90 and 100% open positions; and for MSIV test closing control circuit. The MSIV positions are indicated in the MCR. Redundant solenoid valves, limit switches, wiring and terminals for field interface cables are physically separated in protective enclosures and conduits to assure that effects of harsh environment resulting from high energy line break will not prevent the solenoid valves from performing their safety function. The solenoid valve safety function is to deenergize and remain in this state to keep MSIV closed. These solenoid valves are subjected to Environmental Qualification Program and their qualification is documented in Ref. 7.2.53. The main steam isolation signal is generated by the Solid State Protection System (SSPS) which is a part of System 99 -Reactor Protection System (Ref. 7.2.44). The main steam isolation signal is generated from either of the following conditions: a) High-high containment pressure, b) Low steamline pressure in any steamline (lead-lag compensated, P-11 permissive), c) High steamline negative pressure rate in any steamline (rate-lag compensated, P-11 permissive). Refer to Figure 3.3-1 for the main steam isolation signal development logic. NIT 2 This is the non-essential air header. Class G is non-safety related piping. (V806) 2-FC0-30-295 ....__ (V444) 3 211 3212 53178-045-1 HOLD-----........ 53178-046-0 ............ (V236) 2-FCV-3-185 .......... 3258 (239) 2-FCV-3-188 .___, (V724) 2-FCV-3-245 (V404) 2-FCV-1-150 .......... 3259 3261 2-32-592 2-32-592 AIR 1 /2" (V402) FCV-1-4 1 .......... 2-32-5 These are the MSIVs {V403) 2-FCV-1-29 1 1 2 ............ 2-32-594 -(V807) _____. 2-FC0-30-294 These two check valves prevent the accumulators (air tanks) from depressurizing during a loss of the non-essential air header. l

MSIVs 3.7.2 Watts Bar-Unit 1 3.7-5 3.7 PLANT SYSTEMS

3.7.2 Main Steam Isolation Valves (MSIVs)

LCO 3.7.2 Four MSIVs shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3 except when all MSIVs are closed and de-activated.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One MSIV inoperable in MODE 1. A.1 Restore MSIV to OPERABLE status. 8 hours B. Required Action and associated Completion Time of Condition A not met. B.1 Be in MODE

2. 6 hours C. -------------NOTE---------------- Separate Condition entry is allowed for each MSIV.
 -------------------

One or more MSIVs inoperable in MODE 2 or 3. C.1 Close MSIV. AND C.2 Verify MSIV is closed and de -activated. 8 hours Once per 7 days (continued)

WBN Unit 1 & 2 Main Steam Isolation Valve Maintenance 0-MI-1.002 Rev. 0001 Page 54 of 79 Date ________ 6.6 Air Cylinder Refurbishment and Bench Pressure Testing (continued) NOTES 1) Air cylinder should now be leak test ed for leakage past piston and all static seals (2 and 17 included), in addition to rod seal (11). 2) 125 psi supply pressure required. Air or nitrogen may be used. For air, Control Air System air may be used, or Service Air may be used if a dryer and 50 micron filter are used. To conserve nitrogen, cylinder may be filled with air, then fi nal pressure obtained with nitrogen, if desired. [36] ENSURE cylinder cap port (above piston) is open to atmosphere, THEN: A. CONNECT test rig to cylinder head port (below piston). B. PRESSURIZE to 125 psig. C. USING Snoop leak detector, TEST tube end seal and piston rod seal. D. VERIFY no leakage. Initials [37] ENSURE rod end still pressurized to 125 psig, THEN CLOSE test rig shutoff valve to isolate air in cylinder, AND PERFORM following: A. MONITOR test gauge for 1 minute, AND OBSERVE pressure drop and RECORD. B. IF pressure drop is 0 psi, piston seal leakage is acceptable, THEN GO TO Step 6.6[3].

54. Given the following conditions: - Unit 1 is at 100% power. - A leak develops on the Control Air system. - Control air pressure is 72 psig and LOWERING Which ONE of the following identifies the system response that occurred to maintain Train A Essent ial Air Header pressure?

A. 1-FCV-32-80, AUX AIR TO RX BLDG TR A, SHUTS B. Auxiliary air compressor A-A STARTED and LOADED C. 1-FCV-32-110, NON-ESS AUX AIR TO RX BLDG, SHUTS D. 0-PCV-33-4, Service Air Receiver isolation valve, OPENED

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: Plausible since this valve does go closed to isolate essential air from control air, however this valve doesn't go closed until system pressure decreases to < 70 psig. B. Correct: At 83 psig decreasing the A-A Aux air compressor starts in an attempt to maintain A essential air header pressure above 70 psig to prevent auto closure of the containment supply valves. C. Incorrect: Plausible if the candidate thought that by closing the nonessential air to CNMT that this would maintain essential air header pressure; however, this valve does not go closed until pressure is < 70 psig. D. Incorrect: Plausible, if candidate thought that Service air could be used to supply essential air, however PCV-33-4 goes closed on decreasing pressure (<80 psig) not open.

Question Number: 54 Tier: 2 Group: 1

K/A: 078 Instrument Air System (IAS) K4 Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following: K4.02 Cross-over to other air systems Importance Rating: 3.2 3.5 10 CFR Part 55: (CFR: 41.7) 10CFR55.43.b: Not applicable K/A Match: Question matches K/A by having the candidate determine when the standby Aux air compressors automatically start and supply essential air. Technical

Reference:

0-AOI-10 Proposed references to be provided: None Learning Objective: 3-OT-SYS032A, Control Air System 6. EXPLAIN the Control and Service Air System design features and/or interlocks that provide the following: (IER 11-3, Having a solid understanding of plant design, engineering principles and sciences)

b. Cross-over to other air systems
d. Automatic isolation of sections of the air system
f. Automatic control of station air pressure Cognitive Level:

Higher Lower X Question Source: New Modified Bank Bank X Question History: Bank question 078 K4.02 number 54. Comments: WBN Unit 0 Loss of Control Air 0-AOI-10 Rev. 0002 Page 4 of 41 2.3 Automatic Actions A. Possible Rx trip due to MFW reg valves or heater drain valves failing closed. B. MSIVs may fail closed. C. Operating compressors load in s equence and compressors selected for auto will start on dropping press. D. Aux Air compressors st art at 83 psig dropping. E. Service air isolates from control air at 80 psig dropping. F. Aux air isolates from control air at 79.5 psig dropping. G. Essential and non-essential air to the Rx bldg isolates at 70 psig dropping. H. ERCW to the C&SS air compressors is olate on high flow in conjunction with low press on the compressor ERCW supply headers.

55. Given the following timeline:

00:00:00 The polar crane is secured after placing the reactor head on the head stand. 00:0 1 :00 SOURCE RANGE HI FLUX AT SHUTDOWN(81-B) alarms and is acknowledged. 00: 0 1 : 1 5 Rising counts are observed on both 1-NI-131 and 1-NI-132. Which ONE of the following describes if a horn can be heard and the required actions in accordance with 1-ARI-81-87, NIS & Rod Controls? A horn ____(1)____ be heard in upper containment. The crew will FIRST ____(2)____. NOTE: 1-AOI-34, Immediate Boration A. (1) WILL (2) ENSURE all personnel are evacuated from CNTMT B. (1) WILL NOT (2) ENSURE all personnel are evacuated from CNTMT C. (1) WILL (2) INITIATE an immediate borati on in accordance with 1-AOI-34 D. (1) WILL NOT (2) INITIATE an immediate borati on in accordance with 1-AOI-34

CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: As seen in the Westinghous e Technical Manual for the Nuclear Instrumentation system, WBN-VTD-W120-2826, The evacuation horn assembly is also mounted near the reactor. This horn works in conjunction with the Source Range hi gh flux at shutdown bistable relay driver assembly to provide an evacuation alarm for on-site personnel. Referring to the annunciator response for window 81-B.

SOURCE RANGE HI FLUX AT SHUTDOWN, one will find that such alarm is generated by bistables NC 31CX and NC32CX. On print 1-45W600-57-13, one may find that these two bistable relays are normally energized during the time t hat their associated source range nuclear instrument detects that the high flux at shutdown is zero (or does not exist). When such a condition occurs, the bistable relay will de-energize causing a contact observed on print 1-45W600-57-17 to shut. This contact causes 120V AC power to be supplied to the containment evacuation horn. This alarm horn is located in upper

containment. Considering the st eps contained in the annunciator response for window 81-B, one will s ee that the operating crew will ENSURE all personnel are evacuated from CNTMT. This is step [2.1] in such instruction. The next st ep directs the crew to 1-AOI-34, Immediate Boration. In accordance with TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions, numbered procedure steps are to be performed in the order enum erated within the procedure. B. Incorrect: It is not correct that a horn would not sound in upper containment. It is plausible to believe that this would be the case because one may believe that an alarm horn does not exist because of the facts that the plant paging system allow for the announcement of emergencies and that during core alterations a f uel handling supervisor is stationed (such individual is a SRO and is in constant comm unication with the main control room). He would be able to direct the containment evacuation upon the first indication of RISING counts. Also, as seen in 1-GO-7, Refueling O perations, step [11] dire cts that, Demonstrate that Plant P/A system AUDIBLE in CNTMT OR A means for containment evacuation has been established. This coupled with the fact that 1-GO-7 does not test the alarm horn inside of upper containment would again lead one to believe that the P/A were the primary mode by which a containment evacuation were conducted. Therefore, it would be reasonable to believe that a separate alarm horn would not be required. Again, it is correct that the first action which the control room staff took is to ensure that personnel were evacuated from containment. C. Incorrect: While it is correct that the alarm horn will sound within upper containment, it is not correct that the operating crew will initiate a boration prior to ensuring that all personnel are evacuated from CNTMT. It is plausible to believe t hat this is the case because of the fact that in accordance with TI-12.04, the term ensure is defined as To observe that an expected characterist ic or condition exists and, if necessary, to take actions to make the condition occur. Therefore, in accordance with the ARI, the operat ing crew would only begin a boration after containment was evac uated. In accordance with TI-12.04, initiate is To commence or begi

n. Generally used to cause the start or beginning of an effort which can NOT be completed in a short period of time. Therefore it woul d be reasonable to believe that the operating crew would initiate a containment evacuation and then initiate a boration prior to ensuring that containment were indeed evacuated.

D. Incorrect: As discussed it is Incorrect and yet plausible that an alarm horn would not sound and that the operating crew w ould first initiate a boration of the RCS. Question Number: 55 Tier: 2 Group: 1

K/A: 103 Containment System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations A2 04 Containment evacuation (including recognition of the alarm) Importance Rating: 3.5* 3.6* 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to associate the SOURCE RANGE HI FLUX AT SHUTDOWN annunciator with an alarm horn within upper containment. This association demonstrates the knowledge that a high flux at shutdown will impact the containment system by causing an alarm physically within the containment. The applicant must then use knowledge of 1-ARI-81-87 to identify the action which they would first take. Technical

Reference:

WBN-VTD-W120-2826 TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions 1-GO-7, Refueling Operations 1-ARI-81-87, NIS & Rod Controls

1-45W600-92-1 1-45W600-57-13 1-45W600-57-17 Proposed references to

be provided: None Learning Objective: 1-AOI-29, Dropped or Damaged Fuel or Refueling Cavity Seal Failure 3. IDENTIFY Alarms, symptoms, automatic actuations, and other indications of AOI-29, Dropped or Damaged Fuel or Refueling Cavity Seal Failure.

4. Given a set of plant conditions, DESCRIBE operator actions required in response to the following per AOI-29, Dropped or Damaged Fuel or Refueling Cavity Seal

Failure: Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN-VTD-W120-2826 Pa g e 97 of 479 Table 1-8. Front Panel Control and Indicator Functions (Continued) Name Reference Designation Function DETECTOR CURRENT COMPARATOR SECTION (Continued) UPPER SECTION switch LOWER SECTION switch 1.6.3 REMOTE EQUIPMENT. S605 S606 A five position, make-before-break switch which when placed in the N41, N42, N43, or N44 position enables the removal of a faulty upper section power range channel. The circuit automatically adjusts the gain of the averaging amplifier alarm so that the set point stays the same. When the switch is in the NORMAL position, all four upper power range channels are averaged and compared. A five-position, make-before-break switch which when placed in the N41, N42, N43 or N44 position enables the removal of a faulty lower section power range channel. The circuit automatically adjusts the gain of the averaging amplifier alarm so that the set point stays the same. When the switch is in the NORMAL position all four lower power range channels are averaged and compared. A preamplifier, a loud speaker, a volume control, and an evacuation horn assembly are supplied as part of the system for installation at a remote location. The description of the preamplifier appears in paragraph

1. 2. 3. 11 under the Source Range channel discussion.

The volume control and speaker are mounted near the reactor and operate in conjunction with the remote speaker output of the Audio Count Rate Channel drawer assembly. The volume control is mounted in a housing and has an impedance of 8 ohms and a power rating of 50 watts. The volume control provides a means of adjusting the sound level of the remote speaker over a range of approximately 0 to 100 decibels. The evacuation horn assembly is also mounted near the reactor. This horn works in conjunction with the Source Range high flux at shutdown bistable relay driver assembly to provide an evacuation alarm for on-site personnel. The horn is screw mounted and has two horns separated by 180 degrees. The horn requires 118 volts at 50/60 Hertz. 1-62 WBN Unit 1 NIS & Rod Controls 1-ARI-81-87 Rev. 0004 Page 5 of 52 A. Positive reactivity insertion from: 1. RCS cooldown 2. RCS dilution

3. Rod withdrawal
4. Xenon decay

[1] ENSURE CNTMT evacuation alarm initiates. [2] IF count rate continues to rise, THEN [2.1] ENSURE all personnel are evacuated from CNTMT. [2.2] GO TO 1-AOI-34, IMMEDIATE BORATION. [3] GO TO 1-AOI-4, NUCLEAR INSTRUMENTATION MALFUNCTION. 1-45W600-92-1

1-45W600-92-2

1-45W600-57-13 1-AOI-4 1-AOI-34 Source Setpoint Bistables: 0.5 decade increase above background count rate at shutdown (variable) 1N31AX 1N32AX NC31CX NC32CX 81-BSOURCE RANGE HI FLUX AT SHUTDOWN (Page 1 of 1) Probable Cause: Corrective Action:

References:

1128VL c LAY!:> (M-13) -o OV> TBl 11-2 {M-13) I 6 s (R-58) NCJICX w 8 < j: 1 17 l. r r T8211-1 (M-13) II --;! s (R-58) NCJ2CX -(M-1.J.) SOURCE RANGE HI FLUX AT SHUTDOWN SEPARATION RELAYS I 7 I 8 -" 8 < f1j: 1 7 l. r r I 9 9 10 I I I NOTES: , 406Vl 1 1. FOR C 1-F'U-99-R58/M7

2. * -l JR
  • 4. FOR l ON Sf z CSPR-92-1 CONTAINMENT E VACUATION i HORN "' "'
  • K629Xt
  • K629X2 OPE N FOR OPEN fOR (R-58) SOURCE (R-58) SOURCE J RANGE J RANGE BlOCK eLOC K .. 1 9 I f'o 2 NCJ 1 ex' C LOSEO 2 l R-58) FOR S.R. H I f'LUX (R-58) AT SHUTDOWN AT SHUTDOWN ( 1-45W600-57-1J) 1 {1-45160 0-57-1 3) 1 > 1NJUX, OPEH INJT AX , OPEN 1 FOR S.R. SHUTDN {R-58) FOR S. R. SHU TON (R-58) FLUX ALARM BLOCK F L UX ALARM BLOCK ( t-45W600-57-1 J) 2 { 1-45'#60 0-5 7-1 J) 2 1406VG 1 1-FU-99-R!58/M8 CONTAINMENT EVACUATION HORN j<eD 2 81-9) l (BD 282-B)

WBN Unit 1 & 2 User's Guide for Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 16 of 57 2.2.5 Sequence of Step Performance (continued) B. A required task need NOT be fully completed before proceeding to the next action. 1. Efficient implementation of the EOI network is important; therefore completion of the task is NOT required before continuing with the next step. 2. It is sufficient to begin a task and have assurance it is progressing satisfactorily before continuing with the next step; e.g., when opening a valve, the operator verifies t hat the red light has come on. 3. Completion of steps treated in this manner should be confirmed after sufficient time has elapsed for the acti on to complete; e.g., the operator verifies that the valve is full open based upon the red light on and the green light off. C. Substeps are identified as sequential or non-sequential by the substep designator. 1. Sequential substeps are identified by letters or numbers with a parenthesis. Sequential substeps must be per formed in the order listed. 2. Non-sequential substeps are identifi ed by bullets. Bulleted substeps may be performed in any order and they also have an implied "AND" logic (i.e., all listed steps must be performed) UNLESS an "OR" conjunction is specifically used, or an im plied "OR" logic is stated in the high level action (e.g., "PERFORM one or more of the following" or "IF an y of the following are met", etc.). 3. In general, the rules for completing a given task in steps apply to substeps. EXAMPLE 1 (Sequential Substeps)

12. ENSURE ERCW operation: a. At least four ERCW pumps RUNNING, one on each shutdown board preferred. b. D/G ERCW supply valves OPEN EXAMPLE 2 (Non-Sequential Substeps)
3. CHECK S/G press:
  • All S/G press controlled or rising
  • All S/G press greater than 150 psig WBN Unit 1 Refueling Operations 1-GO-7 Rev. 0000 Page 17 of 65 Date ___________ Initials 5.1 Reactor Disassembly (continued)

NOTE If the Transfer Canal has been filled and draining is NOT required, the following two steps may be marked N/A. [9] CHECK SFP gate seal to the Transfer Canal does NOT leak excessively (leakage greater than SFP makeup flow capability) when Transfer Canal is drained for Transfer Tube Blind Flange

removal. ________ [10] IF water is in the Transfer Canal and draining is required, THEN DEWATER Fuel Transfer Canal per SOI-78.01. ________ [11] INITIATE Setup of Communications between the following stations: ________

  • Main Control Room (MCR)
  • SFP-Side Fuel Transfer Control Console
  • Rx-Side Fuel Transfer C ontrol Console, if manned.
  • Refueling Machine
  • Demonstrate that Plant P/A system AUDIBLE in Cntmt. OR or A means for containment evacuation has been established.

[12] ENSURE BOTH cavity drain plugs INSTALLED per MI-68.001. ________ WBN Unit 1 & 2 User's Guide for Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 41 of 57 Appendix A (Page 3 of 8) EOI Glossary DETERMINE - To evaluate the status of plant parameters or a system in order to establish whether or NOT an action should be performed. Action steps containing this verb usually end with a substep di recting performance of the desired action. DISPATCH - To send to a particular destination. DO NOT - Used to emphasize that somethi ng should NOT be done or implemented at that specific time. DUMP - The operator action of opening steam dump valves in order to cooldown the RCS. EMERGENCY (OR IMMEDIATE) BORATE - To initiate actions to inject a high concentration of boric acid solution into the RCS through the normally recognized emergency boration flow path. EMERGENCY START - Operator action to depress the "Emergency Start" push button for a diesel generator or ot herwise initiate an emergency start. EMERGENCY STOP - Operator action to depress the "Emergency Stop" push button for a diesel generator.

ENERGIZE - To take actions necessary to supply electrical power to equipment or circuitry. ENSURE - To observe that an expected characte ristic or condition exists and, if necessary, to take actions to make the c ondition occur. Typically, the expectation comes from some previous automatic or operator action.

EQUALIZE - The process of making two or more variables or parameters the same. ESTABLISH - To bring about. To take necessary actions to cause a specified set of conditions to exist. EVALUATE - To determine the significance or wo rth of something usually by careful appraisal or study. Generally, to formulate a decision regarding a course of action by careful study of applicable conditi ons. For example, "Evaluate the need for continued containment spray system operation". FAULTED - Used to describe a secondary system component, e.g., steam generator, with a feedwater line or a steam line break. This term is initial letter capped when used in this manner. GO TO - Transition point during pr ocedure performance. The transition may be to another step or another instruction. WBN Unit 1 & 2 User's Guide for Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 42 of 57 Appendix A (Page 4 of 8) EOI Glossary IDENTIFY - To determine which of several si milar components is applicable to a specified condition as outlined by the instruction. IMMEDIATE (OR EMERGENCY) BORATE - To initiate actions to inject a high concentration of boric acid solution into the RCS through the normally recognized emergency boration flow path.

IMMINENT - About to occu r at any moment IMPLEMENT - To carry out. To initiate and maintain a prescribed course of action(s). INITIATE - To commence or begin. Generally used to cause the start or beginning of an effort which can NOT be completed in a short period of time. INSPECT - To examine closely in a critical manner, generally used to require a search for a potential problem or error. INTACT - Describes a steam generator which is neither ruptured nor faulted, the S/G pressure boundary has no breaks. This term is initial letter capped when used in this manner.

INVESTIGATE - To observe or study by close examination in a systematic manner. Generally used to cause a search for problems or information. ISOLATE - To separate one component from another. Generally used to require the securing of flow to and from a component.

LEAVE - Maintain. LOCAL OR LOCALLY - An action performed by an operator outside the Main Control Room. A member of the Control Room Team normally requests personnel other than a team member to perform these actions. Feedback from personnel performing local actions should be r equired by the Control Room Team. LOWER - To cause a parameter to become smaller in magnitude. MAINTAIN - To continuously control a given plant parameter wit hin a specified range or other specified limit as required by the instruction.

MANUAL OR MANUALLY - An action performed by the operator in the control room. These actions are normally in lieu of actions which are normally performed automatically.

56. Given the following timeline:

00:00:00 1-PDI-30-42, CNTMT PRESS is 0 psig and STABLE 00:0 1 :00 1-PI-77-2, RCDT PRE SSURE is RISING 1-LI-77-1, RCDT LEVE L is 5% and RISING 00:20:00 1-LI-77-1 is 45% and STABLE 0-PIS-77-88, WASTE GAS VENT HEADER PRESSURE is 2.2 psig and STABLE Which ONE of the following describes a cause of the above conditions AND the expected Radioactive Collecting Drain Tank (RCDT) pressure? The observed trend was caused by flushing the ____(1)____. Given ONLY the events listed in the timeline, at 00:25:00 , 1-PI-77-2 will indicate ____(2)____ psig. A. (1) RCP #3 seal leakoff lines (2) 0 B. (1) RCP #3 seal leakoff lines

(2) 2.2 C. (1) instrument lines at panel 1-L-171 (2) 0 D. (1) instrument lines at panel 1-L-171 (2) 2.2 CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: It is not correct that flushing the instrument lines at panel 1-L-171 would cause the events listed in the timeline. As seen on print 1-47W851-1, the drain headers from panels 1-L-170 and 1-L-171 (and any other instrument panel inside of containment for that matter) are admitted to the Reactor Building Floor and Equipment Drain Sump's Pocket Sump. It is plausible to believe that these would enter the RCDT because both the Pocket Sump and the RCDT are liquid radioactive waste tanks within co ntainment and both are designed to receive Tritiated water.

The final pressure of the RCDT would not be 0 psig. It would be 0 psig if it were vented to the containment atmosphere (such as is the RBFES). The RCDT is vented to the waste gas header and as such would remain at its pressure (nominally 2.2 psig). B. Incorrect: As mentioned, it is not correct and yet plausible that the instrument line flush would cause the RCDT parameters to increase. It is correct that the pressure of the RCDT would stabiliz e at 2.2 psig. The RCDT is piped to the waste gas vent header. The vent header leads to the suction of the wast e gas compressors which act to

maintain their suction and thus the waste gas header within a pressure band of 2.0 to 3.5 psig. Therefore, sending water inventory to the RCDT will initially cause RCDT pressure to rise and then cause the waste gas vent header to equalize RCDT pressure to its that of its own. C. Incorrect: Prints 1-47W830-1 and 1-47W809-1 demonstrate t hat the RCP #3 seal leakoff lines are piped to t he 4" outlet piping of the RCDT. Therefore, it is correct that any flushing of #3 seal leakoff lines would admit water to the RCDT. Again, it is incorrect and yet plausib le that the stable final pressure of the RCDT would be 0 psig. D. Correct: As discussed, it is correct that flushing seal leakoff lines would affect the RCDT and that the final pressure of the RCDT would be that of the waste gas header.

Question Number: 56 Tier: 2 Group: 2 K/A: 002 Reactor Coolant System K5 Knowledge of the operational implications of the following concepts as they apply to the RCS: K5.05 Reason for drain tank pressure rise during water inventory operations Importance Rating: 2.9 3.3 10 CFR Part 55: (CFR: 41.5 / 45.7) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is to understand the implication had on the reactor coolant drain tanks Technical

Reference:

WBN-SDD-N3-77A-4001, Gaseous Waste Disposal System 1-47W830-1 1-47W809-1 1-47W851-1 Proposed references to be provided: None Learning Objective: 3-OT-STG-077A, LIQUID RADWASTE PROCESSING SYSTEM 4. EXPLAIN the physical connections and/or cause-effect relationships between the Liquid Radwaste System and the following systems: e. Waste gas vent header

f. PRT
g. RCDT
h. Reactor Coolant System (RCS)Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam L-L!;;BMLt-L r FL EL 716.0 EL 70J'-6 J/8" EL 70J'-2-J/8" Fl EL 702.78 t. EL 674'-2-J/S" CONT ON 1-47W845-J COORD LATER CRANE W A LL ;, 1-SUMP-77-129 2 PLAN EL 702.78 SHIELD SLOG ANNULUS DRAINS REACTOR 81..0G t. EL 79!5.0 (15P) ;, 1-77-JOI 3/+"' DIAPHRAM VLV FOR VENTING (BY FIELD) 3 ;, ....... EL 67:5.0 EL SH.O CONTAINMENT PIT SUMP VENT a:. PURCE 2 3 4 r CONT ON 1-47WS4!5-3 COORD LATER GR EL 728.0 TO YARD DRAINAGE v.,. ;'.'l,,..; 5 PLAN EL 729.0 EQUIPMENT .to FLOOR DRAINS 56.63 L----SEAL TA8LE DR I EL 720'-0" -CONT ON 1-'4-7W600-LATER COORD LATER '-------. I ,,>*-1 *----------- I I I I I I I I I e : ;, I :S" FO ;,j -'.:. -'.:.

,I -'.
.

4 HIGH LEVEL ALAR'-& -EL 699'-6" PUt.iPS STOP -EL 697'-l" RB FLOOR EQUIP DRAIN POCKET SUMP 1-SllllP-77-+10 5 6 6 7 CANAL LINER (TYP) LEAK DETECTOR DR (TYP) § § ii _ _f_LAN DR FRQ.A REACTOR PUMPS (TYP 4 Pl.C'S.) FL EL 752.96 NOTE A -fSi if EL ct: ORIFICE ;, .i-rr--,---BOTTct.4 SUMP EL 696.27 RB FLOOR a. EQUIPMENT DRAIN SUMP 1-SUMP-77-125 7 47W476 47W 4 82 47W830 47W8S2 47W801 8 9 10 rm "m .. f ::::.,,=, '""" 11 REACTOR FUEL CAVITY EL 713.88 RCP #+ DR A IN FROM REACTOR PUMPS ( TYP 4 PLACES) 12 ifEAcfroi:f CAVITY LOWER COMPARTMENT PENETRATION LEAK DETECTOR DRAINS CONT ON 1-47W845-J COORD LATER CRANE WALL ,._ NOTE A 295") AZ 280"1 318°-30') EL 795.0 CRANE WALL ,._ -c CONT ON 1-47W866-LATER COORD LATER TO I 4:. C PANELS CARBON TO SS CONT ON 47'#600-SERIES EL 723.0 :(ORIFICE PLATE E

, (42') ;, ;, ILDING AS EM. THE SOCK NT RADIO-'CTIVE HE RE-'CTOR PRIOR TO 12. ITEM AND ASSOCIATED COf.IPONENTS ARE REQUIRED TO REMAIN I GURA TION AS TO PROTECT THE ABSCE PRESSURE BOUNDARY.

CONFIGURATIONS MAY INCLUDE COMPONENTS ON THE ANNULUS AND AUXILIARY ABSCE 13. VALVE F-'ILED CLOSED; AIR SUPPLY ISOLATED TO PREVENT OPERATION. VALVE TO REMAIN IN CLOSED POSITION FOR ABSCE BOUNDARY RETENTION. 14-. UNIT INTERFACE SEPARATION BOUNDARIES ARE CONTROLLED BY INSTRUCTION NOTE A: TI-12.08, CONTROL OF UNIT INTERFACES. DRAINS ARE PERMANENTLY PLUGGED. 0 RCP CAVITY EL 723'-10-1/2" CARBON TO SS \ EL 703'-6 3/8' FL EL 702.78 NOTES: EMBEDDED 1 *

2. PREFIX VALVE NUMBERS AND EQUIPMENT AND STRUMENT REFERENCE OESICNATIONS WITH THE UNIT AND THE SYSTEM NUMBER (77) Cl-77-12
  • D-77-125; ETC) 0 -ON USE, BOTH UN ITS ' 7.

OF THE 47821 SERIES DRAWINGS 'PIPING SYSTEM CLASSIFIC N3-77A-+001

GASEOUS WASTE DISPOSAL SYSTEM NJ-778-4001

SOLID RADWASTE DISPOSAL SYSTEM NJ-77C-4001

LIQUID RADWASTE DISPOSAL SYSTEM WB-DC-40-61

EQUIPMENT ANO FLOOR DRAINS S. DENOTE UNIT 1 2 INTERFACE POINTS. THE STRUCTURAL THE FIRST ANCHORED EQUIPMENT OR PIPE ANCHOR ON THE UNIT 2 INTERFACE POINTS WILL 9. CIID AND!JJIJ DENOTE SAFETY RELATED AND NON-SAFETY RELATED UNIT 1/2 INTERFACE POINTS. SAFETY RELATED INTERFACE POINTS WILL BE LOCKEO CLOSEO, OR CLOSED WI1l1 1HE f.IETHOD OF ACTUATION REMO\IED OR DISABLED.

NON-SAFETY INTERFACE POINTS WILL BE LOO<ED CLOSED OR CLOSED WITH HANOWHEEl..S REMOVED. 10. SOME CIDS ON THIS DRAWING HAVE BEEN CHANGED AND MAY DIFFER FROM THE CIDS SHOWN ON OTHER DOCUMENTS FOR THE SAf.IE COMPONENT. THE ALTERNATE ID SCREEN CAIIl IN EMS CAN BE ACCESSED AS NECESSARY TO DETERMINE IF PREVIOUS CIOS EXISTED FOR A SPECIFIC COMPONENT. 30 53387 RWE JEW REVISED PER DCA 53387-665-0. REV CHANGE REF PREPARER CHECKER APPRO V ED DATE SCALE, NTS EXCEPT AS NOTED PROJECT FACILITY POWERHOUSE UNITS 1 &:. 2 TITLE MECHANICAL FLOW DIAGRAM FLOOR & EQUIPMENT DRAINS WATTS BAR NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY DESIGN INITIAL ISSUE DRAFTER CHECKER Q ENGINEERING APPROVAL C.W.JOHNSON D.L.COOPER 1 P.S.KIRKHAM

THIS CONFIGURATION CONTROL DRAWING DESIGNER REVIEWER RO ISSUE PER WBEP 5.17 a, RIMS 826 '90 0829 37 8 2 P.W.BUllNETT 1 3 L .IV.BOYD CAD MAINTAINED CONTROL WBN System Description Document GASEOUS WASTE DISPOSAL SYSTEM WBN-SDD-N3-77A-4001 Rev. 0010 Page 28 of 56 3.2.1 Waste Gas Compressors (continued) M. Pipe Joints All GWDS piping joints are welded except where flanged connections are necessary for maintenance (e.g., there are 16 flanged connections per compressor package). 3.2.2 Decay Tank Component Description The nine GWDS gas decay tanks are vertical, cylindrical carbon steel tanks with a volume of

600 ft 3 each. The dimensions of the tanks are approximately 8 feet 8 inches outside diameter by approximately 10 feet vertical height. They are fitted with 3 nozzles, a 1-inch inlet, a 1-inch outlet, and a 3/4-inch sample line. Each tank has 2 manholes, both 16 inches in diameter: One is located on the top of the tank and the other on the side. Tanks have a design pressure of 150 lb/in 2 g (Table 9-5), a design temperature of 180 °F, a normal operating pressure of 0-100 lb/in 2g, and a normal operating temperature of 40-140°F. They have a service life of 40 years and are TVA Class D. 3.2.3 Waste Gas Filter Component Description A cleanup system combining a charcoal absorber and HEPA filters removes particulates and gaseous iodine compounds from the gas discharged. The system is designed to handle a maximum flow of 100 cfm at a maximum pressure and temperature of 5.3 lb/in 2 g and 180°F. The maximum pressure drop across a clean element is 6 inches H 2 O at a flow rate of 100 SCFM. Collection efficiency of the particulate filter exceeds 99 percent. The iodine cartridge has a minimum collection efficiency for methyl iodine of 70 percent and 95 percent collection efficiency for elemental iodine. 3.3 Instrumentation and Controls Instrument setpoints and functions are provided in reference 7.54. 3.3.1 Vent Header Pressure System (PCV 77-89) This system indicates and controls the vent header pressure. Pressure indication is provided on the Waste Processing System (WPS) panel. When the backup waste gas compressor is in automatic mode, the controller will start the compressor on high pressure and actuate a high pressure alarm. The backup compressor is stopped automatically when the vent header pressure drops back to normal. In manual mode, both compressors may be turned on or off regardless of the vent header pressure. The vent header high pressure alarm is actuated regardless of the mode of operation of the compressor. A low pressure alarm is also provided with indication and alarms on the WPS panel. 3.3.2 Compressor Control Instrumentation The WG compressor controls are a part of the compressor packages. These consist of pressure, flow, and level controls. Refer to the simplified flow diagram (Figure 8.4). Under normal conditions one compressor package is in automatic with the second acting as backup. On high inlet pressure (2.0 lb/in 2g) the starter switch of the compressor in automatic is energized. Auxiliary contacts on this switch provide operating power for solenoid valves and pressure switches in the waste gas compressor package. WBN System Description Document GASEOUS WASTE DISPOSAL SYSTEM WBN-SDD-N3-77A-4001 Rev. 0010 Page 45 of 56 9.2 TABLE 9 Gaseous Waste Systems Components Gaseous Waste Systems Components Waste Gas Decay Tanks Number 9 Volume, Each, ft 3 600 Design Pressure, lb/in 2g 150 Design Temperature, °F 180 Normal Operating Pressure, lb/in 2g 0-110 Normal Operating Temperature, °F 50-140 Material of Construction Carbon Steel Vertical Cylinder Design Type Code/Class ASME III/3 ANS Safety Class D Waste Gas Compressors Number 2 Type Water Sealed

Centrifugal Design Flow Rate, N 2 (at 140°F, 2 lb/in 2 g)cfm 40 Design Pressure, lb/in 2g 150 Design Temperature, °F 180 Normal Operating Pressure, lb/in 2 g Suction 2.0-3.5 Discharge 0-100 Normal Operating Temperature, °F 70-130 Design Code/Class ASME III/3 ANS Safety Class D Waste Gas Filter Number 1 Maximum Flow (at 5.3 lb/in 2 g, 180°F) SCFM 100 Material CS Maximum Clean Pressure Drop (at 100 scfm) in H 2 O 6 Type Charcoal/HEPA Combination Design Code ASME VIII ANS Safety Class G Sequential Automatic Gas Analyzer Number 1 (shared) Oxygen Electrochemical of the Polarographic type 0-20% O 2 Hydrogen by Thermal Conductivity 0-100% H 2 Automatic stepping switch* 8 steps Recorded Readout 8 points Temperature 120°F *See Table 9-4. 1-TANK-77-1 HIGH LEVEL ALARM EL 707'-3-3/8" PUMP 8 START EL 706'-11-3/8" PUMP A START EL 706'-7-3/8" LOW LEVEL ALARM EL 705'-3-3/8" REACTOR COOLANT DRAIN TANK CAP: 350 GAL DES PRESS. 25 PSIG {INT) 60 PSIG (EXTERNAL) o 267*F , 1-77-512 43/4" --PUMP Aa.B STOP EL 705'-7-J/8" -I a... :J-o... IT ml uoo a:: co ;=Cl (/),........a:: u ... o >10 U--0 ..

0 2" Note: All four of the RCPs' #3 seal leak off go to COORD D-5 of 47W830-1. I* ol 1 ... 62-JOJA VENT [SEAL TI ........ H 1 II -.. CLASS G u < ..... u UN[T 1 ONLY CLASS STAND ""'-"'""*PIPE REACTOR COOLANT PUMP (LOOP #1) 62-302A 62-624-_... ....... _ ... 10 14 A1-1"t

57. Which ONE of the following describes the bus power supply to the 1A-A Centrifugal Charging Pump? A. 1A-A 480V SDBD B. 1A 480V UNIT BD C. 1A-A 6.9kV SDBD D. 1A 6.9kV UNIT BD

CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: The power supply to the 1A-A CCP is the 6.9kV SDBD 1A-A and not the 480V SDBD 1A-A. It is plausible to believe that it would derive power from a 480V board because there are plants whose large motors are 480V powered. B. Incorrect: The CCPs were originally installed as the "High Head Safety Injection Pumps." WBNP's original design in volved the continuous use of a positive displacement charging pump which would serve the non-safety related purpose of chemical and volume control. This charging pump did not possess a safety relat ed qualification. As use of the positive displacement pump was phased out, the CCPs assumed its non-safety related role of chemical and volume control. The relation of this non-safety related role to a power supply generates the plausibility of selecting the Unit Boards as an answer. C. Correct: As seen on print 1-45W724-1, the power supply to the 1A-A CCP is the 1A-A 6.9kV SDBD. D. Incorrect: Again, while it is correct t hat a 6.9kV board powers the 1A-A CCP, it is not correct that a unit board is such supply.

Question Number: 57 Tier: 2 Group: 2

K/A: 011 Pressurizer Level Control System K2 Knowledge of bus power supplies to the following: K2.01 Charging pumps Importance Rating: 3.1 3.2 10 CFR Part 55: (CFR: 41.7) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to identify the bus power supply for the 1A-A Charging Pump. Technical

Reference:

1-45W724-1 Proposed references to be provided: None Learning Objective: 3-OT-STG-062A, CHEMICAL AND VOLUME CONTROL SYSTEM 5. LIST the bus power supplies to the following CVCS components:

a. Centrifugal Charging pumps (CCPs) Cognitive Level:

Higher Lower X Question Source: New Modified Bank Bank X Question History: WBN Bank question used for ILT periodic exams. Comments: " " .. FSAR FIG 8.3-16 ISSUEO BY: 5 9 SYM&OLS: 11 ***** t:OUIPMEN'T LOCATED ON G.91(V $C SIO L.QQJC R(L A Y l'o OIEI. 1A-A * ----- LOCATED ON A STATtlS JMDJt:ArIXC" EIO I O R C IN THE I AUXIL I AR'I' CONJRO I. FIOOM. ti -*-** EOUIF'MEN"!' NOT SAFETY RELAJEO ---Sl'STEM NUMSCR Of £CUIPMCN T l"'\I? ---Sl((<Y NUMoE* (Jf SCHEM.t!C lN SERIES 1-1'"760 I I I ,_,o-* I ooo-l-_ _J nus COMTROL D"AllM: SUPERSEDES UNIT , DRAWINC OWJZt,.-1 ttEVISION FF. 28 60853 RWE R EVISED PER OCA 60!53-77-1. REV REF PREPARER SCALE: NTS PROJECT f ACil HY AUXILIARY BUILDING UNITS 1 & 2 TIT LE WIRING DIAGRAMS GJO CHECKER APPROVEO OHE EXCEPT AS NOTEO 6900V SHUTDOWN SOARD tA-A SINGLE LINE 1 WATTS BAR NUCLEAR PLANT TEMNESSEE VALLEY AUTHORITY DES!CN D R ArTER CHECKER O.H. IA.NC D.L. COClllEFI INITlAL ISSUE RO I SSUE PER WBEP 5.17 -R[hlS 826 '90 0829 377 a ENGINEER INC APPROVAL c."'. SUDCUrH 5 .... HQNHR c.c. 58 Which ONE of the following describes how CERPI calculates Rod Insertion Limits? The CERPI system calculates the Rod In sertion Limit using ________ as an input. A. Calorimetric Power B. Nuclear Instrument Power C. Turbine Impulse Pressure D. Delta Temperature Percent Power

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: It is Incorrect that the plant calorimetric provides an input into the rod insertion limit. It is plausible to believe this for the exact reason seen in the analysis for distractor B. B. Incorrect: As discussed, it is Incorrect that NIS provides an input into the rod insertion limit. It is plausible to believe this because as seen in WNA-CT-00053-WAT: The Rod Insertion Li mit (RIL) algorithm calculates the minimum allowable control bank pos itions as a function of plant power. Therefore, because this dist ractor provides a measure of plant power the distractor is plausible. C. Incorrect: It is Incorrect that the turbi ne impulse pressure provides an input into the rod insertion limit. It is plausible to believe this for the exact reason seen in the analysis for distractor B. D. Correct: As seen in the CERPI Tec hnical Description, WNA-CT-00053-WAT: The Rod Insertion Limit (RIL) algorithm calculates the minimum allowable control bank positions as a function of plant power to ensure sufficient shutdown reactivity margin. The calculated RIL is compared to the control bank demand positions by two bistables; LO and LO-LO. The MTP [maintenance test panel ] receives Delta Temperature Percent Power (%T) from the ICS in ev ery second and it updates PLCs via AF100 bus. %T shall be processed by the following algorithm: RIL LO-LO Setpoint = K 1 x %T + K 2. Question Number: 58 Tier: 2 Group: 2

K/A: 014 Rod Position Indication System K1 Knowledge of the physical connections and/or cause effect relationships between the RPIS and the following systems:

K2.02 NIS Importance Rating: 3.0 3.3 10 CFR Part 55: (CFR: 41.3 to 41.9 / 45.7 to 45.8) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the K/A asks for the power input into the RPI system. The question requires that the applicant recall the power input which is used to produce the Rod Insertion Limit. Technical

Reference:

CERPI Technical Description, WNA-CT-00053-WAT Proposed references to be provided: None Learning Objective: 3-OT-SYS085A, Control Rod Drive System 2. DESCRIBE the design criteria, purpose and/or functions of the Rod Position Indicating (RPI) System and the major system components listed below:

e. Rod Insertion Limit (RIL, CERPI Software)
5. EXPLAIN the physical connections and/or cause-effect relationships between the Rod Control system and the following systems: a. Rod Position Indicators (RPIs)
b. Excore Nuclear Instrumentation (NIS) Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: Watts Bar Unit 1 ARPI System Upgrade CERPI Technical Description 4.2.6.3.5 If any one of the 8 bistable outputs (one for each bank) is active, the output of the Rod to Bank Deviation Alarm relay will be energized to alarm. 4.2.6.3.6 If the quality of any of the Bank Demand Position transmitted by the ICS has Bad quality, then the algorithm uses the value of the last Good quality value of that Bank Demand. 4.2.6.3.7 Rod to Bank Deviation Alarm can be disabled by using the access restricted (by utilizing Function Enable Keylock Switch} Alarm Selection display on the MTP. 4.2.7 Rod Insertion Limit Alarms 4.2.7.1 The Rod Insertion Limit (AIL) algorithm calculates the minimum allowable control bank positions as a function of plant power to ensure sufficient shutdown reactivity margin. The calculated RIL is compared to the control bank demand position by two bistables; LO and LO-LO. 4.2.7.2 The MTP receives Delta Temperature Percent Power (%Lff) from the ICS in every second and it updates PLCs via AF1 oo bus. 4.2.7.3 shall be processed by the following algorithm: RIL LO-LO Setpoint = K1 x %i:1 T + K2 where K1 and K2 are as follows per Reference 9: Control Bank A B c D Coefficient K 1 (steps/%L'1T) 0 2.30 2.29 2.29 K2 211 214 86 -42.45 Table 4.2.7.2-1 -RIL LO-LO Setpoint Algorithm Constants 4.2.7.4 The AIL LO-LO Setpoint is limited to a maximum value of 211 steps per Reference 1 O. 4.2.7.5 RIL LO Setpoint is 10 steps greater than the RIL LO-LO Setpoint and it is limited to a maximum value of 221 steps and minimum value of 0 step. WNA-CT-00053-WAT, Rev. 1 4-7 Westinghouse Proprietary Class 2 Watts Bar Unit 1 ARPI System Upgrade CERPI Technical Description 4.2.7.6 Control bank demand position is an input to RIL LO and RIL LO-LO bistables which trips when the demand position goes below RIL LO Setpoint and RIL LO-LO Setpoint values, respectively. 4.2.7.7 The hysteresis value for both RIL LO and RIL LO-LO bistables is 5 steps. 4.2.7.8 K 1 and K 2 coefficients, RIL maximum and minimum limit values, bistable setpoint and hysteresis values are adjustable via the PLC programming tool. 4.2.7.9 If the quality of any of the Control Bank Demand Position transmitted by the ICS has Bad quality, then the algorithm uses the value of the last Good quality value of that Control Bank Demand. 4.2.7.10 If the quality the Delta Temperature Percent Power transmitted by the ICS has Bad quality, then the algorithm uses the value of the last Good quality Delta Temperature Percent Power. 4.2.7.11 AIL LO Alarm and RIL LOLO Alarm outputs have re-flash capability. Thus, while an alarm condition exists, if a new alarm condition occurs then RIL LO Alarm and RIL LOLO Alarm outputs are energized for 2 Seconds and de-. energized (to alarm) again to indicate the new alarm condition. 4.2.8 Rod Speed Indicator 4.2.8.1 Rod Speed Indicator algorithm calculates the rod speed during a rod motion based on the analog rod speed indication signal. Rod speed shall be between 8 Steps/minute to 72 Steps/minute during a rod motion and 0 Steps/minute otherwise. 4.2.8.2 Analog Rod Speed signal is* processed by the following algorithm: Rod Speed during a rod movement= - + 1.21) I 0.115 4.2.9 Rod Demand from Passive Summer Indicator 4.2.9.1 Rod demand from passive summer analog signal is received, scaled, and displayed in order to help the Operator to know the magnitude of rod movement to be expected when the rod control system is switched from Manual Rod Control to Automatic Rod Control. WNA-CT-00053-WAT, Rev. 1 4-8 Westinghouse Proprietary Class 2 * * *

59. Which ONE of the following descri bes the In-core Monitoring System?

The T/C Map Display of 1-XI-68-100, RVLIS-ICCM PLASMA DISPLAY will show ________ for a BAD thermocouple input? A. 0000 B. 9999 C. BBBB D. XXXX B CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: As seen in the equipment description for the Inadequate Core Cooling Monitor-86, WBN-VTD-W120-3020, 0000 is displayed for the MAX value for a quadrant with no good thermo couples. It is plausible to believe that this would be displaye d as a thermocouple does indeed fail low. One could believe that th is failure would be displayed on the ICCM. B. Incorrect: As seen in the equipment description for the Inadequate Core Cooling Monitor-86, WBN-VTD-W120-3020, 9999 is displayed for the MIN value for a quadrant with no good thermo couples. It is plausible to believe that this would be displayed if one believe that a thermocouple failed high and that such would be displayed on the ICCM. C. Incorrect: It is not correct that " BBBB" will be displayed in the place of a thermocouple reading. Rather, a reverse video "B" will. It is plausible to believe this because the Distri buted Control System (installed on both the primary and secondary plants) displays "BAD" for a bad input. D. Correct: As seen in the equipment de scription for the Inadequate Core Cooling Monitor-86, WBN-VTD-W120-3020, "if a calculated analog input point quality code of BAD-is assigned to any of these values, then a reverse video B-will be displayed next to the new value of X's, for a BAD quality code." As seen in t he vendor manual for the ICCM, the T/C Map Display page displays temperatur es to four digits. Therefore, the correct mode of showing a BAD thermocouple input would be to display XXXXB. Question Number: 59 Tier: 2 Group: 2

K/A: 017 In-Core Temperature Monitor System K6 Knowledge of the effect of a loss or malfunction of the following ITM system components: K6.01 Sensors and detectors Importance Rating: 2.7 3.0 10 CFR Part 55: (CFR: 41.7 / 45.7) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to understand the effect which a failed Thermocouple has on the ICCM display.

Technical

Reference:

Equipment description for the Inadequate Core Cooling Monitor-86, WBN-VTD-W120-3020 Proposed references to be provided: None Learning Objective: 3-OT-STG-068F, RVLIS AND ICCM

10. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on RVLIS:
b. RTD failure Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

INADEQUATE: CORE: COOLING MONITOR-86 Rl:MOTE: INl-OklVIA i IUN \\.1 no _ 3020 IBN-VTD--VESSEL LEVEL DIR6NDSTIC PAGE TinE 12:16:3ct nESSR6E 100 LT-68-36ct -5.331 30 00 UPPER RRNiE 13.025 31 00 LT-68-368 0.000 60 80 LDYER RRNiE 10.000 61 00 LT-68-361 18.121 30 00 DYNmIC HERD 85.22ci 31 00 TE-68-313 1ct.126 30 00 TE-68-316 80.3ct3 30 00 RVLIS LEVEL 13.025 31 00 TE-68-311 80.135 30 00 TE-68-318 80.250 30 00 TE-68-31ct 1ct.101 30 00 PEN 1: RVLIS LEYEL 13.025 1PY-106R 2501.113 30 00 PEN 2: RVLIS LEYEL 13.025 PEN 3: RVLIS LEYEL 13.025 THDT 600.185 31 00 1TY-i13T 600.i85 30 00 RCP-1 00 00 1TY-i23T i23.22i 30 00 RCP-2 00 00 1TY-i11C ct3.012 30 00 RCP-3 00 00 RCP-i 00 00 nRLFUNCT I DN 00 XIS-68-381 00 00 DIAGNOSTIC INFDRMRTIDN XIS-68-388 00 00

  • XIS-68-38ct 00 00 WBNREV b WBN PAG-E _/

Figure 5-10 Typical RCS Level Diagnostics Display Page 5.10.4.1 TIC Map Display Pages The core map display pages (figures 5-11 and 5-12) give a train-oriented view of the core exit thermocouple layouts (CETC). These graphic examples are presented only to familiarize the operator with the mechanics of the ICCM-86 System, and do not necessarily represent exact replicas of the display pages that actually appear on the screen. The rounded whole number value of each CETC is shown in its assigned core location. If a CETC value becomes greater than or equal to some critical setpoint, then it will be displayed in reverse video. If a calculated analog input point quality code of BAD or SUSPECT (without a manually entered value) is assigned to any of these values, then a reverse video "B" 29748 5-27 Rev. O INADEQUATE CORE COOLING MONITOR-86 REMOTE DISPLAY INFORMATION DN -VTD -lf1"0. or "S," respectively, will be displayed next to the new value of X's, for a "BAD" quality code, or D's, for a "SUSPECT" quality code. If a quality code of SUSPECT (with a manually entered value) is assigned to any of these values, then a reverse video "S" will be displayed next to the current numeric value. A calculated analog input point quality code of POOR assigned to any of these values will cause a reverse video "P" to be displayed next to the current numeric value. 5.10.4.2 lncore TIC Quadrant Summary Display Pages WBN REV_...!:O:t_ ___ .' 180° +-N WBNPAGE 117 ICTC QUAD

SUMMARY

(OF) 210° SUBCOOL C TC) QUAD I QUAD II +88 OF MAX 21qo MAX 2160 AVG 2180 AVG 2150 LOOP 3 MIN 2110 : 111 MIN 21"10 LOOP '\ *--*-**************************************:........................................... 0 0 LOOP 2 QUAD IV QUAD I I I LOOP 1 MAX 2100 MAX 2130 AVG 2oqo AVG 2120 MIN 2080 MIN 2110 ....._ _____ _._ _____ __,

  • HI C UAP AVG qoo Figure 5-13 Typical ICTC Quadrant Summary Display Page, Unit 1 The incore TIC quadrant summary pages are shown in figures 5-13 and 5-14. These graphic examples are presented only to familiarize the operator with the mechanics of the ICCM-86 System, and do not necessarily represent exact replicas of the display pages that actually appear on the screen. In a train-oriented view of the reactor vessel, the TIC quadrant summary page displays the rounded whole number values for each quadrant's minimum and maximum thermocouples.

The average TIC temperature will also be displayed for each of the four quadrants. An asterisk will 29748 5-29 Rev. 0 d INADEQUATE CORE COOLING MONITOR-86 REMOTE DISPLAY INFORMATION WBN _ VTD -W1?0 -3020 ICTC QUAD

SUMMARY

(OF) 210° SUBCOOL C TC J QUAD I QUAD I I *BB OF MAX 21qo MAX 2160 AVG 2180 AVG 2150 LOOP MIN 2110 MIN LOOP

  • 3 0 -******************************************:*----***-**--**-*-----**-*-**-***---*----*-

0 180° LOOP 1 QUAD JY l QUAD I I I LOOP 2 MAX 2100 ! MAX 2130 AVG 2oqo AVG 2120 MIN 2080 MIN 2110 N-+ ....._ ___________ __,

  • HI 0 UA 0 AVG qoo Figure 5-14 Typical ICTC Quadrant Summary Display Page, Unit 2 appear in the quadrant with the high auctioneered average temperature and the note,* HI QUAD AVG, will appear in the lower right-hand corner of the page. The corresponding loop numbers will appear in each quadrant's displayed section. In addition to the TIC values being displayed, the rounded whole number value for SUB COOL (TIC or TH) will also appear in the upper right-hand corner of the page. If the subcooling value becomes less than an entered setpoint, its respective value will be displayed in reverse video. If a calculated analog input point quality code of BAD is assigned to any average quadrant value, then a reverse video "B" will be displayed next to the new value of X's for a "BAD" quality code. If a quality code of POOR is assigned to any average quadrant value, then a reverse video "P" will be displayed next to the current numeric value. If there are no "GOOD" thermocouples in any one quadrant, then the MAX and MIN values for that quadrant will be 0000 and 9999 respectively.

29748 WBNREV o WBN 5-30 Rev. O INADEQUATE CORE COOLING MONITOR-86 EQUIPMENT DESCRIPTION WBN -\'TD -W120 -:J020 2300°F) from the field wiring. Each signal is then converted to a -5 to 5-volts input voltage range. An open thermocouple will cause the output signal to go offscale low. The thermocouple input board is described in section 15 of this volume (volume I). 1.7 REMOTE DISPLAY MODULE The train A and train B modular remote displays are units containing an alphanumeric graphic plasma display, pushbuttons which enable page selection, printed circuit boards, and other electronics that process data and format the information for display. Detailed descriptions of the display modules and graphic displays can be found in section 5 of this volume (volume I). 1.8 MAINTENANCE TERMINAL The Maintenance Terminal is a portable personal computer which comes packaged in a conveniently sized suitcase along with an AC adaptor, interconnecting cable for connection to the ICCM cabinet, and an XT ALK communications package on a 3-1 /2 inch diskette. Verifying or modifying scaling values and other constants is accomplished via the Maintenance Terminal and the transaction results are stored on disk. The Maintenance Terminal is disconnected during normal System operation. WBN WBN PAGE :J ?-... *'* 29748 1-16 Rev. 0

60. Given the following conditions: - Unit 1 is at 100%. - The B train CNTMT Purge is in service to Lower CNTMT.

Subsequently: - The following indications are noted: Which ONE of the following describes the status of the CNTMT Purge Fans and the Purge Damper position indications?

The running supply and exhaust CNTMT Purge Fans will ____(1)____.

Purge Damper position indica tion ____(2)____ available on panel 1-M-6. A. (1) STOP (2) IS B. (1) STOP (2) IS NOT C. (1) continue running (2) IS D. (1) continue running (2) IS NOT CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: Print 1-47W611 1 demonstrates the logic for the operation of the containment purge supply and exhaust f ans. One may observe on this print that the supply an d exhaust operate as a pai r (i.e. the same start or stop signal will operate both the supply and exhaust fan of a given train). This print shows that either a Containment Vent Isolation Train A or a CVI Train B will cause both of the trains of containment purge fans to receive a stop signal. Theref ore, it is correct that the A train CVI will cause the B train contai nment purge supply and exhaust fans to stop. One may note that the dam pers immediately adjacent to the supply and exhaust fans receive the sa me signal that the fan does. This logic does not hold true for t he remainder of the containment purge dampers. It is correct that containment purge damper position indication is available on 1-M-6. The containment isolation status panels on 1-M-6

(1-XX-55-6E and 6F) show those dam pers and solenoid valves which are isolated on a CVI. B. Incorrect: It is true t hat both trains of contai nment purge are stopped with the actuation of either trai n of CVI. It is not true that position indication only available on 1-M-9. 1-M-9 is the MCR panel which possesses the controls for containment purge and thus the indications. However, again, 1-M-6 does possess the status panels which do reflect the damper positions for containment purge. C. Incorrect: It is not correct that the B tr ain purge fans will continue to run. It is plausible to believe this because if t he logic attributed to the majority of the containment purge damper s were attributed to the fans then this would be the case. The cont ainment purge dampers which are located either in the annulus or inside of containment are actuated only by the train of CVI to which t hey are affiliated. This is seen on print 1-47W611-30-1. It is correct that containment purge damper position indication is available on 1-M-6. D. Incorrect: It is not correct but plausible t hat the B train purge fans will continue to run. It is not true but plausible th at position indication only available on 1-M-9. Question Number: 60 Tier: 2 Group: 2 K/A: 029 Containment Purge System (CPS) A3 Ability to monitor automatic operation of the Containment Purge System including: A3.01 CPS isolation Importance Rating: 3.8 4.0 10 CFR Part 55: (CFR: 41.7 / 45.5) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to demonstrate the ability to anticipate the response of the containment purge system to a single train of the containment vent isolation signal. Also, the applicant is required to demonstrate the ability to monitor the CPS isolation by understanding on which control board to observe such isolation. Technical

Reference:

1-47W611-30-1 .jpg picture of 1-M-6 Proposed references to be provided: None Learning Objective: 3-OT-TS0303, TS 3.3 "INSTRUMENTATION" 1. DESCRIBE the following aspects of Technical Specifications and Technical Requirements for Instrumentation Tech Specs: a. The Limiting Conditions for Operation, Applicability, and Bases for the LCO.

b. The conditions and required actions with completion

time of one hour or less. Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: !YAAT !ICMA l P"ft<M lttCCM: RM SUPP\. Y r AA E-.> sr.ARf rlll<'M HM 1 9 _____________ __ -l,--'----l l"" Ctl F AM tA

61. Given the following conditions:

GATE-79-5, Spent Fuel Pit and Fuel Transfer Canal Gate is INSTALLED.

- SFP level is 749' 1 1/2".

- The Unit 1 Reactor cavity is EMPTY.

- The flange on the Unit 1 Containment side of the fuel transfer tube is REMOVED.

- Due to a failure of 0-LS-78-3, SFP LEVEL HI/LO (128-A), is DISABLED. Which ONE of the following describes the response of the SFP level AND the means by which SFP level will be monitored as a re sponse to the failure of annunciator window 128-A, in accordance with 0-SOI-78.01, Spent Fuel Pool Cooling and Cleaning System?

SFP level will ____(1)____ IF 1-ISV-78-600, Fuel Transfer Tube Isol is OPENED.

SFP level will be m onitored ____(2)____. (1) (2) A. LOWER at the Spent Fuel Pit itself B. LOWER using a meter on 1-M-6 C. REMAIN the same at the Spent Fuel Pit itself D. REMAIN the same using a meter on 1-M-6

CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: As seen in the system description WBN-SDD-N3-78-4001, Spent Fuel Pool Cooling and Cleaning System , the manual gate valve1-ISV-78-600 is located on the auxiliary buildi ng side of the transfer tube. One may also see this on mechanical print 47W455-1 as well as the intertie between the transfer canal and the equipm ent pit of the reactor cavity; also of note on this print is the fact that the containment side of the fuel transfer tube has a flange which is installed when the plant is not refueling. Prints 47W454-2 and 47W 454-3 show a top down view of the transfer canal and spent fuel pi

t. Print 47W454-3 also shows the level placard located on the south we st wall of the spent fuel pit and the gate which can be installed to separate the transfe r canal from the spent fuel pit. Print 47W454-6 shows the details of the level placard.

From these resources one may s ee that the opening of 1-ISV-78-600 will allow water to flow from either the auxiliary building or the containment given that either po ssesses a water inventory greater than the other. In the question pos ed, the auxiliary building (SFP) possesses a greater inventory than the containment side but because the SFP is isolated from the flow pat h by 0-GATE-79-5, Spent Fuel Pit and Fuel Transfer Canal Gate the level of the SFP will remain

constant. It is plausible to believe that level in the SFP would lower if one did not understand the intertie between the various refueling

volumes. For example, if one believed that 1-ISV-78-600 were attached directly to the SFP or to the fuel cask loading area then level would lower. Also, if the gate between the SFP and the transfer canal were not installed, then the le vel in the SFP would lower. 0-SOI-78.01 contains precaution and limitation K. which states: If MCR SFP alarm is NOT available, local monitoring of level with direct communication to MCR should be established. MCR panel 1-M-6, while it contains the annunciator 128-A does not contain any meter or indicator which presents the actual le vel of the SFP. It is plausible to believe that it does because there ar e level indicators which are used during outage and which would indica te the SFP as it is cross connected with the cavity. Therefore, when the annunciator is lost, the appropriate action is to direct an oper ator to monitor the SFP level at the SFP itself. B. Incorrect: As discussed it is Incorrect and yet plausible that the level in the SFP would not lower. Also, it is Inco rrect and yet plausible that an operator would monitor the SFP level at 1-M-6.. C. Correct: It is correct that level in the SFP would remain the same. It is also correct that level would be monitored from the SFP itself. D. Incorrect: While it is correct that level in the SFP would remain the same, it is not correct and yet plausible that the level of the SFP would be monitored at 1-M-6.

Question Number: 61 Tier: 2 Group: 2

K/A: 033 Spent Fuel Pool Cooling System (SFPCS) A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including: A1.01 Spent fuel pool water level Importance Rating: 2.7 3.3 10 CFR Part 55: (CFR: 41.5 / 45.5) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to first predict the change in SFP level which would result from the opening of 1-ISV 600. Additionally the applicant must be able to select the correct location to monitor the level of the SFP given a loss of annunciation related to the spent fuel cooling system. Technical

Reference:

WBN-SDD-N3-78-4001, Spent Fuel Pool Cooling and Cleaning System 47W455-1; 47W454-2; 47W454-3; 47W454-6 0-SOI-78.01, Spent Fuel Pool Cooling And Cleaning System Proposed references to

be provided: None Learning Objective: 3-OT-SYS079A, FUEL HANDLING 1. STATE the location of the following Fuel Handling System components:

a. New Fuel Pit
b. Spent Fuel Pit
c. Cask Loading area d. Transfer Canal e. Upenders (including equipment, controls and

power supply)

f. Refueling Machine (Including equipment, controls and power supplies)
g. RCCA Change Fixture h. Fuel Transfer Tube gate valveCognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

WBN System Description Document SPENT FUEL POOL COOLING AND CLEANING SYSTEM WBN-SDD-N3-78-4001 Rev. 0018 Page 38 of 71 5.2 System Leakage Leakage detection is provided for the SFP by leakage channels located on the back side of each welded joint of the floor and walls of SFP steel liner. Leakage into these channels will drain to the perimeter leakage channels located at the bottom of the SFP. The leakage will then flow into the SFP drain pipe to a normally open manual gate valve. Visual detection of the leakage from the SFP may be witnessed as the leakage exits the manual valve and drips into a funnel. The leakage is then routed to the tritiated drain collector tank (TDCT) of the waste disposal system. In the event of excessive leakage, the manual gate valve may be closed to prevent further leakage. Similar type design of leakage channels and visual display of leakage are also provided for the fuel transfer canal and the cask setdown pool. Nonqualified instrumentation are provided in the SFP and the TDCT with MCR and local, low and high level alarms, respectively. Leakage from the SFPCCS pumps is routed to the station drainage system. Abnormal equipment leakage is detected by visual methods. 5.3 Loss of Coolant Accident (LOCA) In the event of a LOCA, the CCS flow to the shell side of the SFP heat exchangers may be diverted to ensure an adequate flow of cooling water for safeguard loads after an accident. 5.4 Failure of the Refueling Cavity Water Seal SFP water level would decrease in the event of a leak at the Refueling Cavity Water Seal because the transfer tube would be open connecting the refueling canal, and the transfer canal with the SFP. The decreasing SFP water level would be detected by instrumentation located in the SFP that alarms in the main control room. The SFP would be isolated from the refueling cavity by closing the manual gate valve (78-ISV-600) on the auxiliary building side of the transfer tube. If the water level in the refueling cavity decreases to the elevation of the reactor vessel flange before the SFP can be isolated, there will be two feet of water above the stored fuel in the SFP. If during this time a spent fuel assembly were in transit between the reactor core and the SFP and was in the upright position, it could become uncovered. The consequences of uncovering the spent fuel assembly would be possible high radiation level, fuel cladding failures, and release of radioactivity into the containment and/or auxiliary building. For this scenario, the events would be bounded by the design basis analysis for a fuel handling accident as described in Section 15.4.5 of the WBN FSAR. SFP water temperature would approach and remain at the boiling temperature until the SFPCCS service is restored. 6.0 MAINTENANCE AND TESTING COMMITMENTS Preoperational testing requirements for specific items not contained in other design output documents are provided in Appendix A. Each fuel pool gate shall be visually inspected for defects in the structure, seals or fittings prior to each use. The relief valves shall be checked to prevent over inflation of the seals. Leakage of the seal shall be checked by pressure decay test after the gate is installed. Seals may also be checked in the stored position. ,. 1.,,*, '* I; *,J 1 ,, .. , .... . . ..: **. : .... Ii £ NtJ. t'k:.°kr-rR Mr a 'ftf "M *az.>. 'r'*iUH.£i ,HICK . .. ..* **\* ...

  • t * .... t. . . $$ 8AR1*t10 fd' "IJ.I. #I 8DrTOlf °' RI.Sr C.f>llfCLUTl(JN . CIJ/80N SrltL PIP! /Jf*' UJit' '32/4*QD..

$$ 8ARl"00, -AIUl1,IJ1FID.D .IHNPr . ... : ... ..... *: .. .*: Secr10N C*C NTS FOR SOILT I NC MATERIA.l. SCE OCA-22321-0! OA OftI*Ctr'tM. BIL\. OF MHE'RIAL OFIAWlKC HB!ol455-1 Cl< SA,_. iJO ANO SA 194 QR i. r 11f -Fuel P. 911* EL 72S'-1 t* For $JWnf fNf 5 S. /mer (See Srruct. Sr clwgs * 'llJN12.!4 fhru I M.3862.3-11-1 .3 1J-2t ocA ocA MJ (.r93)E"t. 750'-<> v.::::; -Z.. F11ef setdown area boffom El. . I 3'-10" TVA C101ss' c -c_ (Sli ,-: 4 1,,'-1t. ..... * .... *. . t ,' . :* ..... . .. . ... .. . . . 27'-O" .* Locati o n of 1-ISV-78-600 t.f*-rr L A6 .. **. *; '* . t1'-<r OCA It ] JI A ROD !M 1 3 -0"'] O"l 050006-02-0 "' I * *

  • 1 *. ,14 ., I l I
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5'-'" FROM AS 4'-0" fROM AS 7'-0 "FROM A5 705'-O" . ,' I \ \ IL____ ""='" ' ',, ___ //BOTTOM OF REDUCtR'-', ----,EL 742'-0" ; '--. 1:1 < WEST WALL 60oo/ fil] \--VENDOR SUPPLIED I ' I SEE DWG # 1710 1 I CONTRACT

  • 163402 \ 1--t THK WALL ',, 14-3/8"/oD DET FS '----/ NTS El mdximum operdf1011 , .. ,,1 .,,,1;,,,.,, ,,,....,; t:d$k. , . "'11n. 50{8

£1. SS. /111e..-, 48Nlci'f BOTTOM OF SCALE" EL 745:9f.f W4TCR UVCL SCAL[ (FAB BV FICLOJ SS BAR n 'd,'x4:71/.I-MARKEO AT E"L 75o:r T!-IRIJ CL 74s:10:* PLATC MTL MAV 8£ SIJBSTIT()T[O FOR SAR AS fi'F()O. (SA 240 TYPE 304 OR SA 216, TYPE !04.) ---------1 SPENT FUEL PIT Liner fli. *1 I I Bottom of slot 725 tf"-:::;___ I I SECTION A6-A6 (Sheets /,,? i. 3) NOTES FOR DET G6 PAINT COATING MATERIAL RED: AMER LOCK 400NT, SAFETY RED 1979, OR EQUAL WHITE: AMERLOCK 400NT, WHITE, OR EQUAL GREEN: AMERLOCK 400NT, RT-2305 GREEN, OR EQUAL THE STRIPES USED ARE GREEN TO REPRESENT THE NORMAL OPERATING RANGE FROM rnE ! m' y ALARM INDICATOR AND THE RANGE FROM 749'-1/2" TO 748'-11-1/2" AT THE LOW LEVEL ALARM I ND I CA TOR; AND UPPER AND LOWER RED STRIPES TO I ND I CATE LEVELS ABOVE THE LEVEL ALARM RANGES. (FOR ADDITIONAL HELP IN IDENTIFYING THE WATER LEVEL THE GREEN LINE INDICATING THE NORMAL LEVEL IS EXTENDED TO A 1 INCH WIDE POINTER ON THIS SECOND LEVEL INDICATOR), FOR THIS APPLICATION THE AMERLOCK 400NT PAINT IS ACCEPTABLE FOR USE ON THE STAINLESS STEEL MATERIAL AND IN THE FUEL PIT WATER. DO NOT PA I NT WI TH IN 1

  • 5 I NCH DIAMETER OF HOLE CENTERLINES TO CLEAR NUTS. OCA 050006-03-0 TYP EL 749 '-1-1/2" I I I I L ______ _ W ATER LEVEL SCALE DETAILS DET GS (CUT AS SHOWN) '* PIPE ALL MATERIAL TO BE 1/4" THICK STAINLESS STEE NTS WBN Unit 0 Spent Fuel Pool Cooling And Cleaning System 0-SOI-78.01 Rev. 0007 Page 10 of 197 3.0 PRECAUTIONS AND LIMITATIONS A. SFP spills may cause contamination spread and unnecessary radiation exposure. Take immediate correct ive actions including informing Rad Protection. B. Loose items may fall into SFP, and dam age Fuel assembly, racks, etc. Loose items must be secured to personnel or another piece of stationary equipment. C. The SFP water temperature is NOT to be permitted to exceed 150

°F during normal operation and anticipated refueling activities, pursuant to the FSAR. D. The SFP Filter r equires replacing when P of 20 psi or greater, OR Rad Protection discretion (radiation level). E. Exceeding 100 gpm or 140 °F through SFP Demin can damage resins. During refueling outages, flows up to 180 gpm through SFP Demin using the RWPS pumps are acceptable. Ultrasonic flow-meters may be used to determine flows during refueling cleanup F. If Skimmer Loop is in service, ai r may enter skimmer head, causing pump cavitation. SFP level should be checked periodically to ensure proper relation between skimmer head and vortex created by pump suction. G. SFP Skimmer Filter is replaced at 20 psi P. H. When taking suction from Refueling Cavity with RWPS Pump(s), monitor suction press after starting pump(s) due to high loop in suction pipe which could cause loss of pump suction. I. RWPS Filter damage may occur if 200 gpm is exceeded. J. During transfer activity, RWST and SF P levels must be monitored to ensure correct valve alignments and Tech Spec requirements. K. If MCR SFP level alarm is NOT available, local monito ring of level with direct communication to MCR should be established. [C.1] L. Pump Starting Guidelines are in GOI-7. [C.1] M. Spent Fuel Pit minimum temperature assumed for criticality analysis is 39°F(4°C), based on Sys 79 System Description-Nuclear Fuel Report PFE-R07, Revision 0, dated October 1996, "Criticality Analysis Summary Report, Watts Bar Nuclear Plant (WBN)" (L38 961015 802). N. Instrument maintenance department should be notified to ensure required instruments will be placed in service as necessary to support system operation.

62. Given the following timeline:

00:00:00 Unit 1 is at 100% power. Spent fuel movement is in progress in the SFP. The FH AREA EXH FAN A is RUNNING. 00:15:00 A fuel assembly is dropped while trying to extract it from its storage rack. Visual inspection shows that it is slightly bent. 00:20:00 0-RM-90-102 and 0-RM-90-103 display a STABLE reading as shown. 00:2 1:00 The control room staf f enters 1-AOI-29, Dropped or Damaged Fuel or Refueling Cavity Seal Failure section 3.4, Dropped or Damaged Irradiated Fuel Assembly in Spent Fuel Pit Area. Which ONE of the following descri bes the status of FH AREA EXH FAN A AND the FIRST action the crew will implement in accordance with 1-AOI-29? At 00:20:0 1, the FH AREA EXH FA N A is ____(1)____. The FIRST action which will be taken is to ____(2)____ . A. (1) running (2) CHECK ABI actuated B. (1) secured (2) CHECK ABI actuated C. (1) running (2) EVACUATE the affected area D. (1) secured (2) EVACUATE the affected area CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: It is correct that the "A" FHE fan will continue running. The rate meters shown in the stem of the questi on do not show a high radiation condition. Also, step 1 of 1-AOI-29 is "EVACUATE the affected area." Step 7 of the AOI initiates ABI. Therefore, it is incorrect to believe that the first action would be to evacuate. It is plausible to believe that the initiation of the signal which by design is cr edited for protecting the health and safety of the public (the ABI whic h limits the release to the public) would be the immediate action. B. Incorrect: The conditions shown in the st em of the question do not indicate that a high radiation signal has been reached. If such had been the case, the Rate Meters shown would have t he red alarm light LIT. As seen in prints: 1-45W600-3, 1-45W 760-30-14 and 1-45W600-57-30, a high radiation signal from eit her of the spent fuel pi t radiation monitors will cause both of the FHE fans to be automatically secured. It is plausible to believe that the FHE fan would be secured as damage to the fuel assembly was noted.

Again, it is incorrect and yet plausible that the initiation of the ABI would be the first action taken. C. Correct: It is correct that the "A" FHE fan would remain runni ng. It is also correct that the first action taken in 1-AOI-29 is to evacuate the affected area. D. Incorrect: Again, it is incorrect and yet plausible that the "A" FHE fan would be secured. It is correct that the fi rst action taken in 1-AOI-29 is to evacuate the affected area.

Question Number: 62 Tier: 2 Group: 2

K/A: 034 Fuel Handling Equipment System (FHES) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling System ; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations: A2.01 Dropped fuel element Importance Rating: 3.6 4.4 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to predict the impact that a dropped fuel element has upon the FH AREA EXH FAN A (a component of the fuel handling equipment system). Subsequently, the applicant must be able to prioritize the actions contained in the abnormal response instruction for the dropped fuel assembly in the spent fuel pit area; thus, demonstrating their ability to use the procedure to mitigate the consequences of the accident. Technical

Reference:

1-45W600-30-4 1-AOI-29, Dropped or Damaged Fuel or Refueling Cavity Seal Failure Proposed references to be provided: None Learning Objective: 1-AOI-29, Dropped or Damaged Fuel or Refueling Cavity Seal Failure 3. IDENTIFY Alarms, symptoms, automatic actuations, and other indications of AOI-29, Dropped or Damaged Fuel or Refueling Cavity Seal Failure.

4. Given a set of plant conditions, DESCRIBE operator actions required in response to the following per AOI-29, Dropped or Damaged Fuel or Refueling Cavity Seal

Failure:Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

A27CT4 A27CTP A27CCP (1-M-9) 0-[ A6 0-HS-30-136A A-P AUTO *

  • B6 *
  • A1 G w ' 0-HS-30-136A
  • L
  • START B1 ,,.., (.) B , "" A '"" (.) " N "' <( (.) " [ N 5 . <( 62A . / D 1 / ' 30X '
  • 11
  • G
  • 1 6 30X 1 0 TEST TRIP 0-HS-30-136B L STOP -, I [ 0-HS-30-136A STOP , "-F A 11
  • B 11 (1-M-9) 0-HS-30-136A A-P AUTO An identical contact appears on the "B" trained FHE fan. It ._-1 serves to allow (through the train separation E relays) the "A" train OR RADN DET IN REFUEL AREA [ O-FSV-30-136 ENERGIZE TO OPEN A L FHF D 0 -1< UVX 4 VKAR-4 D This came from (1-R-73) (

step 4b. When this 30-4) A contact shuts, the * ... F HF trip coil is actuated 0 -' UVY causing the FHE fan 'D "A" to secure. N (.) (.) " N (.) -/ logic to secure the "B" train FHE fan. . . H A27CC1 <( c 0 D c 0 D FHF BOX FHF BOY L? (.) (.) " N <( x 0 ,,.., (.) " N <( " N <( This line actuates the trip coil of the _____ l ______ , ______________________ " ------------- breaker. 13 14 (FOR BREAKER INTERNALS 8 SEE 1-45W760-30-8) 12 AB1B-2) 1 -R-7 2 1 (45'/1600-57-30) 2 A27CT1 \ / IF SM 400a,600V A27CT3 D AB1 OR RADN DET IN REFUEL AREA A2 7CTRL 1 -.L -------------60P 9 29 I 5 1 0 30 -------------------------------------------------------------------------- --FUEL HANDLING EXHAUST FAN A (O-MTR-30-136) (SEE NOTE 3) (480V SHUTDOWN BOARD 2A2-A ONLY) (.) " N <( 6 [ 62B 2 A27CTRL 2-FU-212-A27/52 A27BTRL This excerpt is from 1-45W760-30-14. -R ' N " I a::: I --0 l.&J .... < ..... 8 en en < < z ..... < a::: t-...J < z_ o_ """"' I (/) " ..... an >, ..... 0 00 I <O z. 0 Ill Z+ o' <->LLJ 0W N (Jl 132 VL --1320VG 1 1 a ABA1 L2l 2 a ( 1-R-73) KAR-3 CLOSE FOR ABI OR HIGH RAO IN REFUELING AREA ABA2 This contact shuts (as seen in 4a) causing train separation relay AB1 A-2 to energize (which in turn causes the "B" FHE fan to be secured). AUX BLDG GEN AUX BLDG GEN EXHAUST FAN SUPPLY FAN (1-45W760-l0-12) (1-45W760-30-11) __ __,,A I \ I \ 1 B 28 18 28 A > IL >......__ ,-- ' CO N CO N CO N CO N 0 t-0 .... 0 .... 0 < < < < m m m m I' I' I' ...... N N N N 5m m m m m m m m a (1-R-72) J 7 9 11 13 15 17 D L24 2 4 10 12 14 16 18 6 ____ ,....._,....._ u t-u ...... 0 ...... u ...... < < < < m m m m l'l'l'l'NNNN N N N N ..... ..--..... m m m m m m m m ( 1-R-72) m z < ..... W* -"' zo ...... ,,., ...J I co Zco <,..... Ill ...J+ W1 ...__ ...... .... 0 ...... N m 1 2 3 7 8 9 10 AUXILIARY BUILDING GENERAL EXH AND SUPPLY FANS 18 & 28 & FUEL HANDLING EXH FAN 8 SEPARATION RELAYS (COMMON TO PLANT) WBN Unit 1 Dropped or Damaged Fuel or Refueling Cavity Seal Failure 1-AOI-29 Rev. 0003 Page 14 of 27 Step Action/Expected Response Response Not Obtained 3.4 Dropped or Damaged Irradiated Fuel Assembly in Spent Fuel Pit Area CAUTION

  • Maintenance must be notified to IMMEDIATELY ensure at least one door is closed on both upper and lower personnel air locks, and close any other open penetrations required per

TI-68.002.

  • Entry to the Refuel Floor and other affected areas shall be coordinated with Radi ation Protection.
1. EVACUATE the affected area:
  • Fuel Handling SRO NOTIFY personnel in the general area of radiation concern.
  • ANNOUNCE for all personnel to evacuate the affected area.
2. NOTIFY Shift Manager and Radiation Protection of radiation release.
3. EVALUATE conditions and necessary

protective measure prior to reentry into affected area.

4. ENSURE closure initiated for personnel air locks and any other open penetrations in accordance with

TI-68.002.

5. CHECK 0-RM-90-101A, B, C Aux Bldg Vent monitor NORMAL [0-M-12].

NOTIFY Radiation Protection to begin surveys and monitoring of Refuel Floor.

6. CHECK SFP area monitors 0-RM 102 and -103 NORMAL [0-M-12].

STOP and LOCKOUT Fuel Handling Area Exh Fans. WBN Unit 1 Dropped or Damaged Fuel or Refueling Cavity Seal Failure 1-AOI-29 Rev. 0003 3.4 Dropped or Damaged Irradiated Fuel Assembly in Spent Fuel Pit Area (continued) Page 15 of 27 Step Action/Expected Response Response Not Obtained

7. CHECK ABI actuated:
  • ABI Train A window LIT [MISSP]
  • ABI Train B window LIT [MISSP]

INITIATE ABI by placing 1-HS-30-101A and 1-HS-30-101B to ACTUATE [1-M-6]. 8. CHECK the following fans OFF [1-M-9]:

  • Aux Bldg Supply Fans
  • Aux Bldg Exhaust Fans
  • Fuel Handling Area Exh Fans Manually STOP fans as necessary.
9. CHECK Train A and Train B ABGTS running [0-M-25].

START ABGTS Train A and Train B.

10. CHECK the following ABGTS dampers OPEN:
  • 1-FCO-30-146B, Train A Suct
  • 1-FCO-30-146A, Train A Disch
  • 2-FCO-30-157B, Train B Suct
  • 2-FCO-30-157A, Train B Disch Manually OPEN dampers.
11. IF Containment or Annulus is OPEN to the Aux Bldg, THEN CHECK Containment Vent Isolation has occurred:
  • CVI Train A window LIT [MISSP]
  • CVI Train B window LIT [MISSP]
  • 1-XX-55-6E CVI Train A green
  • 1-XX-55-6F CVI Train B green IF CVI has NOT occurred, THEN: a. STOP Purge Air Supply & Exh Fans. b. CLOSE Purge Air Cntmt Isol Dampers. c. CLOSE Cntmt Press Relief Valves [1-M-9]. d. MONITOR Cntmt Pressure.
63. Given the following conditions: - Unit 1 SG #1 is faulted. - Unit 1 has tripped and Safety Injection has actuated.

- When the MSIVs hand-switches were placed to CLOSED , the #1 MSIV failed to CLOSE. - The control room is MINIMALLY staffed. Which ONE of the following describes when AFW flow may be isolated AND the indications on 1-M-4? In accordance with 0-TI-12.04, User's Guide Fo r Abnormal And Emergency Operating Instructions, AFW flow to SG #1 may be isolated afte r the ____(1)____ of 1-E-0 are complete. In accordance with 1-E-2, the FIRST MCR indication that compensator y measures for the failure of the #1 MSIV to close HAVE been taken is _____(2)_____. NOTE: 1-E-0, Reactor Trip or Safety Injection 1-E-2, Faulted Steam Generator Isolation A. (1) immediate actions (2) B. (1) immediate actions (2) C. (1) steps through the verification of heat sink (2) D. (1) steps through the verification of heat sink (2) CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: 0-TI-12.04 states: "Early isolation of auxiliary feedwater (AFW) should be performed after positive identificat ion of a ruptured, faulted, or ruptured and faulted S/G providing t he following guidelines are met: a. Prior to isolation, the steps of 1-E-0 are complete through verification of heat sink." 0-TI-12. 04 also states: "Diagnostic or repair actions will be delayed until the immediate actions are complete to allow for

evaluation of plant response." Theref ore, it is incorrect that AFW may be isolated to the faulted S/G immedi ately following the first four steps (the immediate steps) of 1-E-0. It is plausible to believe this as the second excerpt of 0-TI-12.04 shows a requirement which would

independently support the isolation of AFW immediately following the

completion of the immediate actions.

As seen in 1-E-2, the first action ta ken to compensate for a failed open MSIV (by a minimally staffed crew) is to manipulate AUX transfer switches. The resulting indication for such component operation is that of the "ACR PNL 1-L-11A" and "ACR PNL 1-L-11B" lights lit on 1-XX-55-6F, FUSES & AUX SWITCHES indicating panel. B. Incorrect: Again, it is incorrect and yet plausible that the isolation of AFW to a faulted S/G would occur following t he completion of the immediate action steps of 1-E-0.

It is incorrect that a green position would be seen on 1-FCV-1-135 before the ACR PNL lights were observed. The manipulation of 1-HS-1-135A is directed by Attachment 2 of 1-E-2. If the control room is minimally staffed, Attachment 2 is always performed at the conclusion of Attachment 1 (if no resolution is had to the failure of the MSIV to close). It is plausible to believe that the green position is seen first because if an extra operator were available, then the Attachment 2

could be performed in parallel with . A board operator could certainly manipulate a HS on 1-M-2 prior to an AUO reaching the ACR and manipulating tw o transfer switches. C. Correct: As discussed, it is correct that AFW may be isolated to a faulted S/G after the verification of heat si nk. It is also correct that the ACR PNL lights would be observed prior to the green indication on 1-FCV-1-135.

D. Incorrect: While it is correct that AFW may be isolated to a faulted S/G after the verification of heat sink, it is not correct and yet plausible that the green indication on 1-FCV-1-135 woul d be noted prior to the ACR PNL

lights. Question Number: 63 Tier: 2 Group: 2

K/A: 035 Steam Generator System A4 Ability to manually operate and/or monitor in the control room: A4.06 SG isolation on steam leak or tube rupture/leak Importance Rating: 4.5 4.6 10 CFR Part 55: (CFR: 41.7 / 45.5 to 45.8) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires ability to appropriately operate components to isolate AFW to a faulted S/G and then to monitor the correct indications for the isolation of the steam flowpath. Technical

Reference:

1-E-2, Faulted Steam Generator Isolation 1-E-0, Reactor Trip or Safety Injection

0-TI-12.04, User's Guide For Abnormal And Emergency Operating Instructions, Revision Proposed references to

be provided: None Learning Objective: 3-OT-EOP0200, 1-E-2, Faulted Steam Generator Isolation

4. LIST from memory all the components/valves that must be isolated in order to ISOLATE a faulted S/G. (No valve numbers required). Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New Question for the 2015-301 NRC RO exam. Comments:

WBN Unit 1 & 2 User's Guide for Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 34 of 57 2.7 Prudent Operator Actions (continued)

9. Adjusting a S/G PORV to 90% if a known SG tube leak/rupture exists. 10. Early isolation of auxiliary feedwater (AFW) should be performed after positive identification of a ruptured, f aulted, or ruptured and faulted S/G providing the following guidelines are met:
a. Prior to isolation, the steps of 1-E-0 are complete through verification of heat sink. b. For a ruptured S/G, ensure t hat narrow range (NR) level on the affected S/G is >29% and heat sink minimum requirements are met for the unaffected S/Gs. c. For a faulted or ruptured and faul ted S/G, ensure heat sink minimum requirements are met fo r the unaffected S/Gs. 11. When an ATWS event is in progress and after the crew has attempted to open the reactor trip breakers from the second reactor trip hand switch and the trip is unsuccessful, the response should be as follows:
a. The RO starts inserting control rods.
b. The BOP/CRO immediately trips the turbine, then starts AFW.
12. If NO SI condition exists, heat sink may be verified and AFW throttled as necessary to stabilize RCS temperature prior to transition out of 1-E-0. B. In deciding if taking prudent action is appropriate, the following elements should be considered as a whole:
1. Plant safety status should be main tained or enhanced. Prudent mitigation or preemptive action should NOT degrade plant status or put the plant in a less safe state or challenge it more than the initiating event. It should NOT cause a RED or ORANGE path critical safety function condition. For example, closing the MSIVs to isol ate the steam dumps at 100% power would challenge the plant more than the original problem by lifting Pressurizer and S/G safeties and causing an overheat and load rejection event. However, in Mode 3, closi ng the MSIVs to stop an uncontrolled cooldown and positive reactivity addition would have minimal impact on plant status, while enhancing reactor safety. 2. Prudent operator actions should be c onsistent with procedural guidance for similar situations. For example, if a steamline break occurs in Mode 3 and

S/G pressure is approaching 675 psig , a manual MSIV Isolation would anticipate the automatic response and is consistent with procedural guidance and plant design response. WBN Unit 1 & 2 User's Guide for Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 14 of 57 2.2.3 Foldout Page A. This page presents actions or transitions which are applicable at any time in the given instruction. Upon transition from an instruction, the current instruction's foldout page becomes applicable and use of the previous instruction's foldout page is discontinued. 1. In the control room, the foldout page information is presented on the back of each page of instructions for which there is a foldout page. 2. The information on the foldout page should be continuously monitored to determine when operator action is necessary. B. Transitions to other instructions allo w immediate response to new symptoms as they appear. 2.2.4 Immediate Action Steps Steps that have been designated as "Immedi ate Actions" are contained in three emergency instructions and selected AOIs. These steps are intended to be performed, if necessary, without the written instruction being available. A. Operators are required to be able to complete the intent of immediate operator action steps. B. During immediate operator action steps the operat ors will ensure automatic actions have occurred or initiate signals as appropriate. Diagnostic or repair actions will be delayed until the immediate actions are complete to allow for evaluation of plant response. C. The immediate actions of 1-E-0 will be addressed as follows: 1. Steps 1 through 4 will be performed in order by the OAC and completion, along with any discrepancies, will be communicated to the Procedure

Reader. 2. The BOP will acknowledge alarms as nec essary to reduce the noise level, and perform backup verification of steps 1 through 4. 3. Procedure Reader will read the immediate action high level step. The OAC will confirm the high level step by verbalizing the low level steps to the procedure reader. 4. When reentering 1-E-0 fr om another EOI, the first four high level steps must be reconfirmed, it is NOT necessary to reperform each low level action. WBN Unit 1 Reactor Trip or Safety Injection 1-E-0 Rev. 0005 Step Action/Expected Response Response Not Obtained Page 6 of 45

5. PERFORM Appendixes A and B , 1-E-0, pages 16-26
6. ANNOUNCE reactor trip and safety injection over PA system.
7. ENSURE secondary heat sink available with either:
  • Total AFW flow greater than 410 gpm, OR
  • At least one S/G NR level greater than 29% [39% ADV].
    • GO TO 1-FR-H.1, Loss of Secondary Heat Sink.

WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Step Action/Expected Response Response Not Obtained Page 3 of 17 3.0 OPERATOR ACTIONS CAUTION If a faulted S/G is NOT needed for RCS cooldown, it should remain isolated during subsequent recovery actions.

1. ENSURE all MSIVs and MSIV bypasses CLOSED.

Manually CLOSE valves. IF valves can NOT be closed, THEN Locally REMOVE power to valves:

  • DISPATCH NAUO to perform Attachment 1 (1-E-2).

NOTE If it is known that a steam leak exists in the Turbine building, the following step should NOT be performed until the affected steam header is depressurized.

2. PLACE steam dump controls OFF:
  • 1-HS-1-103A, STEAM DUMP FSV "A".
  • 1-HS-1-103B, STEAM DUMP FSV "B". 3. DETERMINE if Cntmt spray should be stopped: a. MONITOR Cntmt pressure less than 2.0 psig.
a. WHEN Cntmt pressure less than

2.0psig, THEN PERFORM substeps 3b thru e.

b. CHECK at least one Cntmt spray pump running.
b. **GO TO Step 4 c. RESET Cntmt spray signal.

Step continued on next page.

WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Page 10 of 17 Attachment 1 (Page 1 of 5) Isolation of MSIVs and MSIV Bypass Valves 1.0 INSTRUCTIONS NOTE If additional MCR Operators are av ailable, US may direct perfo rmance of Attachment 2, MSIV and Bypass Valves Fail Open, to be performed in parallel with Attachment 1. 1.1 Isolation of MSIVs and MSIV Bypass Valves A. IF any MSIV will NOT close, THEN GO TO Section 1.2 (Attachment 1). B. IF any MSIV bypass valve will NOT close, THEN GO TO Section 1.3 (Attachment 1). WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Attachment 1 (Page 2 of 5) Isolation of MSIVs and MSIV Bypass Valves Page 11 of 17 1.2 MSIV Isolation A. PLACE affected MSIV transfer control switch in AUX position: [Auxiliary Control Room, P anels 1-L-11A and 1-L-11B] S/G (Circle affected S/G) EQUIPMENT AUX TRANSFER SWITCH AUX POSITION 1 MSIV Loop 1, Train A, 1-FCV-1-4 1-XS-1-4A MSIV Loop 1, Train B, 1-FCV-1-4 1-XS-1-4B 2 MSIV Loop 2, Train A, 1-FCV-1-11 1-XS-1-11A MSIV Loop 2, Train B, 1-FCV-1-11 1-XS-1-11B 3 MSIV Loop 3, Train A, 1-FCV-1-22 1-XS-1-22A MSIV Loop 3, Train B, 1-FCV-1-22 1-XS-1-22B 4 MSIV Loop 4, Train A, 1-FCV-1-29 1-XS-1-29A MSIV Loop 4, Train B, 1-FCV-1-29 1-XS-1-29B B. CONSULT UO to verify affected MSIV closed. WBN Unit 1 Faulted Steam Generator Isolation 1-E-2 Rev. 0002 Page 15 of 17 Attachment 2 (Page 1 of 3) MSIV and Bypass Valves Fail Open Date ________ 1.0 MSIV AND BYPASS VALVES FAIL OPEN A. IF affected MSIV or MSIV Bypass still open after completion of or performance directed by US, THEN: 1. CLOSE the following HP steam and bypass isolation valves. MSR A2 HP STM ISOL 1-M-2 1-HS-1-135A MSR B2 HP STM ISOL 1-M-2 1-HS-1-137A MSR C2 HP STM ISOL 1-M-2 1-HS-1-139A MSR A1 HP STM ISOL 1-M-2 1-HS-1-141A MSR B1 HP STM ISOL 1-M-2 1-HS-1-143A MSR C1 HP STM ISOL 1-M-2 1-HS-1-145A MSR A2 HP STM BYPASS ISOL 1-M-2 1-HS-1-235A MSR B2 HP STM BYPASS ISOL 1-M-2 1-HS-1-237A MSR C2 HP STM BYPASS ISOL 1-M-2 1-HS-1-239A MSR A1 HP STM BYPASS ISOL 1-M-2 1-HS-1-241A MSR B1 HP STM BYPASS ISOL 1-M-2 1-HS-1-243A MSR C1 HP STM BYPASS ISOL 1-M-2 1-HS-1-245A 2. CLOSE the following HP steam warming valves. MSR A1 WARMING LINE 1-M-2 1-HS-1-142 MSR A2 WARMING LINE 1-M-2 1-HS-1-136 MSR B1 WARMING LINE 1-M-2 1-HS-1-144 MSR B2 WARMING LINE 1-M-2 1-HS-1-138 MSR C1 WARMING LINE 1-M-2 1-HS-1-146 MSR C2 WARMING LINE 1-M-2 1-HS-1-140 3. ENSURE BOTH steam Seal Supply valves in CLOSED position: a. 1-HS-47-180A, HP SEAL STEAM SUPPLY ISOL [1-M-2] b. 1-HS-47-181A, HP SEAL STEAM SUPPLY BYPASS [1-M-2]

64. Which ONE of the following describes if condenser vacuum can be established AND the effect a loss of all condenser vacuum pumps WILL have on saturation temperature?

In accordance with 1-SOI-2&3.01, Condensate and Feedwater System, condenser vacuum ____(1)____ be established followi ng a refueling outage if no condenser vacuum pumps are available.

With Unit 1 at 100% power, a loss of all condenser vacuum pumps WILL cause the saturation temperature of the condenser to ____(2)____. (1) (2) A. CAN RISE B. CAN LOWER C. CAN NOT RISE D. CAN NOT LOWER CORRECT ANSWER: C DISTRACTOR ANALYSIS: A. Incorrect: It is not correct that initial conden ser vacuum could be draw n without the use of the condenser vacuum pumps. It is correct t hat saturation temperature would rise. Admitting a initial amount of steam via the Steam Dumps would initially draw a vacuum due to the condensing action of the steam, but the air and non condensable gasses would build up and vacuum would degrade subsequently. B. Incorrect: Again, it is not co rrect and yet plausible that initia l condenser vacuum could be drawn without the vacuum pumps. This may be obser ved in section 5.2 of 1-SOI-2&3.01, Condensate and Feedwater System. It is impracticable to build a large condenser which is air tight. As su ch, some method must exist to remove air and other non-condensable gases which migrate into the co ndenser. Plants either utilize air ejectors or vacuum pumps for this purpose. Addition ally, some method must exist to initially cause a vacuum to exist in the condenser. A gain, this necessitates the use of either an air ejector or a vacuum pump. It is plaus ible to believe that the vacuum pumps are not required to establish condenser vacuum following a refueling outage because one may either believe that air ejectors are pr esent and thus used for this purpose or one may not understand that the condensing ac tion of steam (which maintains the vacuum at power) is the method by which initial vacuum is formed.

It is not correct that saturation temperature lowers. If the inevitably present non-condensable gases are not removed from the condenser, then the gas concentration present will increase. As this occurs, the heat transfer surface ar ea of the condenser will decrease. This occurs because the gases tend to blanket the tubes of the condenser. A reduction of the surface area will cause the heat transfer to become

reduced and thus the vacuum of the condenser to lower. Because of the relationship of temperature and pressure within a saturated system, the saturation temperature at which the steam will condense increases. It is plausible to believe that the saturation temperature lowers because one may believe that saturation temperature varies as does the saturation pressure (i.e. that as one lowers the other does as well). C. Correct: It is correct that in itial condenser vacuum could not be achieved without the CVPs. It is correct that saturati on temperature would rise. D. Incorrect: It is correct that in itial condenser vacuum could not be achieved without the CVPs. It is not correct and yet plausible that saturation temperature would lower.

Question Number: 64 Tier: 2 Group: 2

K/A: 055 Condenser Air Removal System (CARS) K3 Knowledge of the effect that a loss or malfunction of the CARS will have on the following: K3.01 Main condenser Importance Rating: 2.5 2.7 10 CFR Part 55: (CFR: 41.7 / 45.6) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to understand the effect that a loss of the condenser vacuum pumps (the CARS at WBN) would have on both the initial production of main condenser vacuum and on the saturation temperature of the main condenser when the Unit is at power. Technical

Reference:

1-SOI-2&3.01, Condensate and Feedwater System Generic article regarding the non-condensable gases in a condenser. Proposed references to be provided: None Learning Objective: 3-OT-AOI1100, Loss of Condenser Vacuum 3. IDENTIFY Alarms, symptoms, automatic actuations, and other indications of AOI-11, Loss of Condenser

Vacuum Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Condensate and Feedwater System 1-SOI-2&3.01 Rev. 0011 Page 33 of 261

Date________ INITIALS

5.2 Establishing Condenser Vacuum NOTE Since the Condenser Vacuum Exhaust (CVE) effluent must be monitored for a SG tube leak, and the CVE monitors must remain isolated until vacuum is achieved to avoid water damage, the following comp measures must be taken per guidance in System Description N3-2-4002, Section 4.10:

  • SG blowdown rad monitors should be placed in service.
  • Sampling provisions of the ODCM will be used to assess radioactive effluents out of the CVE.

[1] ENSURE Section 5.1, Short Cycle Operation, COMPLETE. ________ [2] ENSURE Atmospheric Condensate Drain Tank is available to receive water per 1-SOI-5&6.01, THEN PLACE Injection Water in service to the Condensate Booster Pumps (if ready to start) and Main Feedwater Pumps per applicable steps of 1-SOI-54.01, if NOT already in service. ________ [3] ENSURE 1-RE-90-119, COND VAC PMP AIR EXH RAD MON, is NOT in service. ________ [4] PLACE Turbine on Turning Gear per 1-SOI-47.01. ________ [5] ESTABLISH Steam Seals per 1-SOI-47.03. ________ [6] ESTABLISH Gland Seal Water to Boot Seal and Vacuum Breaker per 1-SOI-37.01. ________ [7] ENSURE water level in each Condens er Vacuum Pump (CVP) reservoir is approx. 50% [T3H/685]: A. 1-LG-2-168, COND VAC PMP A SEAL WATER LEVEL. ________ B. 1-LG-2-173, COND VAC PMP B SEAL WATER LEVEL. ________ C. 1-LG-2-178, COND VAC PMP C SEAL WATER LEVEL. ________ [8] PLACE 1-HS-30-883, CONDENSER VACUUM PUMP AREA COOLER [T3F/685], in AUTO. ________ WBN Unit 1 Condensate and Feedwater System 1-SOI-2&3.01 Rev. 0011 Page 34 of 261

Date________ INITIALS

5.2 Establishing Condenser Vacuum (continued)

[9] OPEN 1-FCV-2-255, with 1-HS-2-255, EXH BYPASS COND VAC PMPS [1-M-3]. ________ [10] OPEN the following: A. 1-ISV-24-589, COND VACUUM PUMP CLR 1C RCW INLET ISOL [T3F/685]. ________ B. 1-ISV-24-593, CVP SEAL WTR COOLER 1A RCW INLET ISOL [T3G/685]. ________ C. 1-ISV-24-591, CVP SEAL WTR COOLER 1B RCW INLET ISOL [T3G/685]. ________ NOTE Venting seal water pump is necessary if t he pump and/or seal water tank has been drained. [11] IF necessary, THEN VENT CVPs' seal water pump(s) by performing the following: (N/A substeps if venting NOT performed) [11.1] REMOVE cap from the following valve(s):

  • 1-VTV-2-1045, COND VACUUM WATER PMP A VENT. ________
  • 1-VTV-2-1046, COND VACUUM WATER PMP B VENT. ________
  • 1-VTV-2-1047, COND VACUUM WATER PMP C VENT. ________

[11.2] ENSURE water level in seal water tank is above level of vent valve. ________ WBN Unit 1 Condensate and Feedwater System 1-SOI-2&3.01 Rev. 0011 Page 35 of 261

Date________ INITIALS

5.2 Establishing Condenser Vacuum (continued)

[11.3] OPEN to vent trapped air from pump, THEN CLOSE the following valve(s):

  • 1-VTV-2-1045.

________

  • 1-VTV-2-1046.

________

  • 1-VTV-2-1047.

________ [11.4] REINSTALL cap on the following valve(s):

  • 1-VTV-2-1045.

________

  • 1-VTV-2-1046.

________

  • 1-VTV-2-1047.

________ [12] PERFORM the following: NOMENCLATURE LOCATION POSITION UNID PERF INITIALS COND VACUUM PMP A CONTROLS T2F/680.5 RESET 1-HS-2-171B COND VACUUM PMP B CONTROLS T2F/680.5 RESET 1-HS-2-176B COND VACUUM PMP C CONTROLS T2F/680.5 RESET 1-HS-2-181B [13] START CVP(s) as necessary [1-M-3]: A. 1-HS-2-171A, COND VACUUM PMP A. ________ B. 1-HS-2-176A, COND VACUUM PMP B. ________ C. 1-HS-2-181A, COND VACUUM PMP C. ________ [14] ENSURE 1-FCV-6-330, VACUUM BREAKER COND A, CLOSED. ________ WBN Unit 1 Condensate and Feedwater System 1-SOI-2&3.01 Rev. 0011 Page 36 of 261

Date________ INITIALS

5.2 Establishing Condenser Vacuum (continued)

[15] WHEN vacuum is established (C-9 permissive has been acknowledged), THEN PLACE 1-RE-90-119, COND VAC PMP AIR EXH RAD MON, in service per 0-SOI-90.02. ________ [16] ENSURE Chemistry notified when condenser vacuum exhaust system and/or radiation monitors returned to service. ________ [17] IF additional CVPs are needed to maintain backpressure 0 to 5 in. Hg on 1-P/TR-2-2 [1-M-3], THEN START CVPs as necessary: A. 1-HS-2-171A, COND VACUUM PMP A. ________ B. 1-HS-2-176A, COND VACUUM PMP B. ________ C. 1-HS-2-181A, COND VACUUM PMP C. ________ WBN Unit 1 Condensate and Feedwater System 1-SOI-2&3.01 Rev. 0011 Page 37 of 261

Date________ INITIALS

5.2 Establishing Condenser Vacuum (continued)

NOTE 1) Ammonia is injected into the C ondensate System for pH control. 2) Ammonia injection rate is higher during unit startup; therefore maximizing Seal Water Tank turnover is necessary. Throttling open drain valves in the following step helps reduce ammonia concentration in the general ar ea by raising the turnover rate of the Seal Water Tank volume, which allows ammonia to exit via drain while still entrained in the condensate. [18] ESTABLISH continuous drain for the in-service pump(s) to reduce ammonia concentration by throttling the drain valve to

a position which maximizes drain flow within the capabilities of the drain system and the auto makeup: A. 1-DRV-37-572, GLAND SEAL WATER DRAIN - PUMP 1A. ________ B. 1-DRV-37-571, GLAND SEAL WATER DRAIN - PUMP 1B. ________ C. 1-DRV-37-565, GLAND SEAL WATER DRAIN - PUMP 1C. ________ CAUTION CVP Discharge Reliefs open if 1-FCV-2-255 is closed before condenser pressure reaches 5 in. Hg. [19] WHEN 1-P/TR-2-2, COND PRESS & TEMPERATURE, is 0 to 5 in. Hg, THEN CLOSE 1-FCV-2-255 with 1-HS-2-255, EXH BYPASS COND VAC PMPS [1-M-3]. ________ WBN Unit 1 Condensate and Feedwater System 1-SOI-2&3.01 Rev. 0011 Page 38 of 261

Date________ INITIALS

5.2 Establishing Condenser Vacuum (continued)

CAUTION Max Hotwell and CST temp is 140 °F. [20] IF Sparging Steam is required, THEN THROTTLE 1-PFV-12-513, CONDENSER SPARGING SUPPLY [T8F/708], to maintain less than 140 ° on 1-P/TR-2-2, COND PRESS & TEMPERATURE. ________ [21] IF Startup filter s are required, THEN PLACE Startup filters in service per 1-SOI-2.02. ________ [22] IF Condensate Polishers are required, THEN PLACE Condensate Polishers in service per 1-SOI-14.01. ________ [23] VERIFY Condensate is within chemical limits. ________ [24] IF MFPT Condenser Drain Pumps are required, THEN PLACE in service per 1-SOI-5 & 6.01. ________ [25] WHEN main condenser vacuum is established and stable, THEN STOP one of the following Cond enser Vacuum Pumps and place in P-AUTO: (Pumps NOT selected may be N/A'd) A. 1-HS-2-171A, COND VACUUM PMP A. ________ B. 1-HS-2-176A, COND VACUUM PMP B. ________ C. 1-HS-2-181A, COND VACUUM PMP C. ________ WBN Unit 1 Condensate and Feedwater System 1-SOI-2&3.01 Rev. 0011 Page 39 of 261

Date________ INITIALS

5.2 Establishing Condenser Vacuum (continued)

[26] ISOLATE continuous drain for any pumps STOPPED: A. 1-DRV-37-572, GLAND SEAL WATER DRAIN - PUMP 1A. ________ B. 1-DRV-37-571, GLAND SEAL WATER DRAIN - PUMP 1B. ________ C. 1-DRV-37-565, GLAND SEAL WATER DRAIN - PUMP 1C. ________ End of Section

Heat Exchanger Knowledge l Heat Exchanger Companies Suppliers The steam condenser, shown in Figure 9 below, is a major component of the steam cycle in power generation facilities. It is a closed space into which the steam exits the turbine and is forced to give

up its latent heat of vaporization. It is a necessary component of the steam cycle for two reasons.

One, it converts the used steam back into water for return to the steam generator or boiler as

feedwater. This lowers the operational cost of the plant by allowing the clean and treated

condensate to be reused, and it is far easier to pump a liquid than steam. Two, it increases the

cycle's efficiency by allowing the cycle to operate with the largest possible delta- T and delta-P

between the source (boiler) and the heat sink (condenser). Because condensation is taking place, the term latent heat of condensation is used instead of latent heat of vaporization. The steam's latent heat of condensation is passed to the water flowing through

the tubes of the condenser. After the steam condenses, the saturated liquid continues to transfer heat to the cooling water as it falls to the bottom of the condenser, or . This is called subcooling, and a certain amount is

desirable. A few degrees subcooling prevents condensate pump cavitation. The difference between

the saturation temperature for the existing condenser vacuum and the temperature of the

condensate is termed . This is expressed as a number of degrees condensate

depression or degrees subcooled. Excessive condensate depression decreases the operating

efficiency of the plant because the subcooled condensate must be reheated in the boiler, which in

turn requires more heat from the reactor, fossil fuel, or other heat source. There are different condenser designs, but the most common, at least in the large power generation

facilities, is the straight-through, single-pass condenser illustrated Figure 9 above. This condenser

design provides cooling water flow through straight tubes from the inlet water box on one end, to

the outlet water box on the other end. The cooling water flows once through the condenser and is termed a single pass. The separation between the water box areas and the steam condensing area is accomplished by a tube sheet to which the cooling water tubes are attached. The cooling water

tubes are supported within the condenser by the tube support sheets. Condensers normally have a

series of baffles that redirect the steam to minimize direct impingement on the cooling water tubes. The bottom area of the condenser is the hotwell. This is where the condensate collects and the condensate pump takes its suction. If noncondensable gasses are allowed to build up in the

condenser, vacuum will decrease and the saturation temperature at which the steam will condense increases. Non-condensable gasses also blanket the tubes of the condenser, thus reducing the heat transfer surface area of the condenser. This surface area can also be reduced if the condensate level is allowed to rise over the lower tubes of the condenser. A reduction in the heat transfer surface has the same effect as a reduction in cooling water flow. If the condenser is operating near its design

capacity, a reduction in the effective surface area results in difficulty maintaining condenser vacuum. The temperature and flow rate of the cooling water through the condenser controls the temperature of the condensate. This in turn controls the saturation pressure (vacuum) of the condenser. To prevent the condensate level from rising to the lower tubes of the condenser, a hotwell level control system may be employed. Varying the flow of the condensate pumps is one method used to

accomplish hotwell level control. A level sensing network controls the condensate pump speed or pump discharge flow control valve position. Another method employs an overflow system that

Front Page Engineering Store Engineering Forum Engineering News Engineering Videos Engineering Calculators Site Directory Engineering Jobs GD&T Training ASME Y14.5-2009 DFM DFA Training Advertising Center Copyright Notice Share More lShare Share Share Share Page 1 of 2 Large Steam System Condensers - Heat Exchanger - Engineers Edge 11/ 8/ 2014 http://www.engineersedge.com/heat_exchanger/large_steam_condenser.htm spills water from the hotwell when a high level is reached. Condenser vacuum should be maintained as close to 29 inches Hg as practical. This allows maximum expansion of the steam, and therefore, the maximum work. If the condenser were

perfectly air-tight (no air or noncondensable gasses present in the exhaust steam), it would be

necessary only to condense the steam and remove the condensate to create and maintain a vacuum.

The sudden reduction in steam volume, as it condenses, would maintain the vacuum. Pumping the

water from the condenser as fast as it is formed would maintain the vacuum. It is, however, impossible to prevent the entrance of air and other noncondensable gasses into the condenser. In

addition, some method must exist to initially cause a vacuum to exist in the condenser. This

necessitates the use of an air ejector or vacuum pump to establish and help maintain condenser

vacuum. Air ejectors are essentially jet pumps or eductors, as illustrated in Figure 10 below. In operation, the jet pump has two types of fluids. They are the high pressure fluid that flows through the nozzle, and the fluid being pumped which flows around the nozzle into the throat of the diffuser. The high velocity fluid enters the diffuser where its molecules strike other molecules. These molecules are in turn carried along with the high velocity fluid out of the diffuser creating a low pressure area

around the mouth of the nozzle. This process is called entrainment. The lowpressure area will draw

more fluid from around the nozzle into the throat of the diffuser. As the fluid moves down the diffuser, the increasing area converts the velocity back to pressure. Use of steam at a pressure

between 200 psi and 300 psi as the high pressure fluid enables a singlestage air ejector to draw a

vacuum of about 26 inches Hg. Normally, air ejectors consist of two suction stages. The first stage suction is located on top of the

condenser, while the second stage suction comes from the diffuser of the first stage. The exhaust

steam from the second stage must be condensed. This is normally accomplished by an air ejector

condenser that is cooled by condensate. The air ejector condenser also preheats the condensate

returning to the boiler. Two-stage air ejectors are capable of drawing vacuums to 29 inches Hg. A vacuum pump may be any type of motor-driven air compressor. Its suction is attached to the condenser, and it discharges to the atmosphere. A common type uses rotating vanes in an elliptical

housing. Single-stage, rotary-vane units are used for vacuums to 28 inches Hg. Two stage units can

draw vacuums to 29.7 inches Hg. The vacuum pump has an advantage over the air ejector in that it

requires no source of steam for its operation. They are normally used as the initial source of

vacuum for condenser start-up. © Copyright 2000 - 2014, by Engineers Edge, LLC All rights reserved Disclaimer l Feed Back l Advertising l Contact Page 2 of 2 Large Steam System Condensers - Heat Exchanger - Engineers Edge 11/ 8/ 2014 http://www.engineersedge.com/heat_exchanger/large_steam_condenser.htm

65. Given the following conditions: - Unit 1 is in MODE 3. - A Loss of Control Air has occurred. - Control Air header pressure is 60 psig and LOWERING - Power is available to the A Station Air compressor. - Unit 2 Construction Air is available. Which ONE of the following descr ibes the response of the AUO?

In accordance with 0-AOI-10, section 3.3, an AUO will be dispatched to ____(1)____. After the AUO completes this task and control air header pressu re is restored, 0-PCV-33-4, Service Air Isol Valve ____(2)____ AUTOMATICALLY RE-OPEN. NOTE: 0-AOI-10, Loss of Control Air Section 3.3, Loss of Control Air in Mode 1,2,3 or 4 A. (1) START the 'A' Station Air compressor (2) WILL B. (1) START the 'A' Station Air compressor (2) WILL NOT C. (1) ALIGN Unit 2 Construction Air Compressor to Unit 1 Control Air header

(2) WILL D. (1) ALIGN Unit 2 Construction Air Compressor to Unit 1 Control Air header

(2) WILL NOT CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: Section 3.3 of 0-AOI-10, Loss of Control Air addresses a loss of control air in mode 1, 2, 3 or 4. Step 3. of this section directs: START C&SS Compressors locally as required USING Attachment 1, Local Restart of the C&SS Compressors. Investigation of Attachment 1 will reveal that it provides direction to start the A and/or B station air compressor and alerts the operator that If 480V Aux Bldg Common Bd is energized, Compressors C will also be available and will start if there is a demand. It does not provide any guidance for starting the D station air compressor. Print 1-45W600-32-1 shows the circuitry for 0-PCV-33-4. Contained on this print is PS-33-4 which opens on a decreasing pressure (nominally set to 80 psig). When this pressure switch opens, the solenoid valve which admits control air to 0-PCV-33-4 de-energizes and thus causes 0-PCV-33-4 to shut. This action isolates control air from service air. PS-33-4 requires a manual reset; this is seen on print 1-45W600-32-1. It is plausible to believe that it does not as numerous valves in the plant will close and open automatically. B. Correct: It is correct that the A station air compressor would be started. It is also correct that 0-PCV-33-4 would require a manual reset to cause it to reopen. This manual reset would be conducted by an AUO (the reset is local to the valve's control panel). C. Incorrect: It is not correct that Unit 2 construction air would be aligned to restore the Unit 1 control air header pressure. It is plausible to believe this because 0-SOI-32.01, Control Air System contains section 8.4, Manual Alignment of Unit 2 Construction Air Compressor to Unit 1 Control / Service Air Header. This section directs the AUO to open 1 valve to restore control air header pressure. Before the replacement of the reciprocating air compressors (the old A,B and C) which occurred this cycle, any time that the D air compressor was tagged out for maintenance, the on-shift crews would direct AUOs to walk down (i.e. locate the 2-ISV-33-543 valve) such that they could restore control air pressure quickly. Therefore, because physically, the quickest was to restore air header pressure is to cross connect Unit 2 construction air to Unit 1 it is plausible that this would be directed by the AOI-10. However, 0-AOI-10 does not make allowance for using this method. Again, it is plausible but Incorrect to believe that 0-PCV-33-4 would automatically reopen. D. Incorrect: It is Incorrect and yet plausible that Unit 2 construction air would be aligned to restore the Unit 1 control air header pressure. It is correct 0-PCV-33-4 would not automatically reopen.

Question Number: 65 Tier: 2 Group: 2

K/A: 079 Station Air System G 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. Importance Rating: 3.8 4.0 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to possess the knowledge that an AUO would start the A station air compressor and that as a result of his action the service air isolation valve would not automatically reopen.. Technical

Reference:

0-AOI-10, Loss of Control Air 0-SOI-32.01, Control Air System 1-45W600-32-1 Proposed references to be provided: None Learning Objective: 3-OT-AOI1000, Loss of Control Air 5. DESCRIBE the reasons and applicable conditions for the notes, cautions, and major actions of AOI-10, Loss of Control

Air. (IER 11-3, Operating the plant with a conservative

bias) Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 0 Loss of Control Air 0-AOI-10 Rev. 0002 Step Action/Expected Response Response Not Obtained Page 11 of 41 3.3 Loss of Control Air in Mode 1, 2, 3 or 4

1. DISPATCH personnel to check C&SS Compressors.
2. CHECK 480V power available to at least 2 C&SS compressors. REFER TO 1-AOI-35, Loss of Offsite Power. 3. START C&SS Compressors locally as required USING Attachment 1, Local

Restart of the C&SS Compressors.

4. Locally CHECK air dryers and filters operating properly with 2 stations in

service:

  • Regeneration time is 3 minutes for A & B Dryers only.
  • Filter diff press less than 5 psid.
  • One tower per station is drying.

OPEN the following to bypass any filter which is malfunctioning: a. 0-BYV-32-233, Stat ion A Prefilter Bypass. b. 0-BYV-32-228, Stat ion A Afterfilter Bypass. c. 0-BYV-32-222, Stat ion B Prefilter Bypass. d. 0-BYV-32-357, Stat ion B Afterfilter Bypass. e. 0-BYV-32-199, Stat ion C Prefilter Bypass. f. 0-BYV-32-207, Station C Afterfilter Bypass. 5. ENSURE Emergency Purge Shutoff valve not closed as indicated on Dryer Panel for C Dyer only. ENSURE remaining dryers are in service and remove dryer from service.

6. DISPATCH personnel to identify and isolate air leak.
7. REFER TO Appendix A for list of potentially inoperable valves and equipment.

WBN Unit 0 Loss of Control Air 0-AOI-10 Rev. 0002 Page 38 of 41 Attachment 1 (Page 1 of 2) Local Restart of C&SS Air Compressors Step Action/Expected Response Response Not Obtained NOTE 1 Continuous contact between UO and NAUO should be maintained to aid quick recovery of air compressors. NOTE 2 Attachment 1 is posted on panel 0-L-240 near the station air compressors. The posting must be re vised if this Attachment is revised. Changes in revision level or page number will NOT require re-posting, as long as the content does NOT change 1. Locally CHECK 0-PCV-33-4, SERVICE AIR SUPPLY ISOLATION, CLOSED [T7M/708]. CLOSE 0-PCV-33-4 usin g 0-HS-33-4. NOTE A local audible alarm will sound when 0-HS-32-25E or 0-HS-32-26B is pressed in the following steps.

2. PRESS 0-HS-32-25E to restore power to Compressor A.
3. PRESS 0-HS-32-26B to restore power to Compressor B.
4. ENSURE 0-HS-32-25D is in the CLOSE position.

PLACE 0-HS-32-25D in CLOSE.

WBN Unit 0 Loss of Control Air 0-AOI-10 Rev. 0002 Attachment 1 (Page 2 of 2) Local Restart of C&SS Air Compressors Page 39 of 41 Step Action/Expected Response Response Not Obtained

5. ENSURE 0-HS-32-26A is in the CLOSE position.

PLACE 0-HS-32-26A in CLOSE.

6. ENSURE A and/or B Station Air Compressor is running and supplying air pressure.

IF neither Air Compressor is running, ENSURE Header pressure is <setpoint on 0-CNTL-32-25 and/or -26.

7. MONITOR Compressor operation:
  • Oil press >34 psig.
  • Compressors auto-loading.

NOTE If 480V Aux Bldg Common Bd is energized, Compressors C will also be available and will st art if there is a demand.

8. CHECK 480V Aux Bldg Common Bd ENERGIZED.
    • GO TO Step 10.
9. ENSURE 0-HS-32-27A is in the CLOSE position.

PLACE 0-HS-32-27A in CLOSE.

10. RETURN TO Instruction in effect.

S MX { 2 MS*ll.-4 r!I'"'° { f I """' O-I ZS 'i Ii 1 ...... 0-PCV-33-4 SERVICE AIR ISOLATION VLV No t e th<ll th e pre ss ur e reqmre s a mamrnl re s et ANN WOW 2C*42f' SERVICE AIR PCV-33-4 CLDS£C (1-4!i 911 5 5*2Cl WBN Unit 0 Control Air System 0-SOI-32.01 Rev. 0012 Page 48 of 61 Date________ INITIALS 8.4 Manual Alignment of Unit 2 Cons truction Air Compressor to Unit 1 Control/ Service Air Header CAUTION Performance of this section should only be performed with concurrence of US/SM with a degraded or in anticipation of a degr ading control air header pressure. NOTE 2-ISV-33-543 will requ ire an extension ladder to access i

t. 2-ISV-33-509 will require a scaffold to be erected.

[1] UNLOCK and OPEN one of the following:

  • 2-ISV-33-543 SERVICE AIR TURBINE BLDG SUPPLY HDR ISOL.[T9D/708] ________

OR

  • 2-ISV-33-509 SERVICE AIR SUPPLY HDR ISOL.

[T9K/708] ________ [2] IF 0-PCV-33-4 SERVICE AIR SUPPLY ISOLATION is NOT OPEN, THEN THROTTLE 0-ISV-33-502 SERVICE AIR 0-PCV-33-4 BYPASS OPEN. [T7M/708] ________ A. ENSURE 0-HS-33-4, SERVICE AIR SUPPLY ISOLATION [T7M/708], in AUTO. ________ B. WHEN Service Air pressure equals Control Air pressure, THEN RESET 0-PS-33-4, SERVICE AIR SUPPLY ISOLATION PRESSURE [T7M/708]. ________ WBN Unit 0 Control Air System 0-SOI-32.01 Rev. 0012 Page 49 of 61 Date________ INITIALS 8.4 Manual Alignment of Unit 2 C onstruction Air Compressor to Unit 1 Control/ Service Air Header (continued) NOTE 0-PCV-33-4, Service Air Supply Isolation from Control Air, will take from 5 to 10 minutes to travel from full closed to full open position. C. ENSURE ANNUNCIATOR WINDOW 42-F [1-M-2], SERVICE AIR PCV-33-4 CLOSED, NOT LIT. ________ D. CLOSE 0-ISV-33-502, SERVI CE AIR 0-PCV-33-4 BYPASS [T7M/708]. ________ [3] CONTACT Unit 2 Operations and not ify them that Unit 1 is receiving air from Unit 2 cons truction air compressors. ________ End of Section

66. Given the following conditions:

00:00:00 Unit 1 is at 100%. Observed at 00:00:00 Observed at 00:00:05 In accordance with OPDP-1, Conduct of O perations, which ONE of the following completes the MINIMUM standard for the OAC's reporting the information depicted above to the operating crew? Announce: "Crew update."

____(1)____ to ensure ALL crew members are listening. Announce: "Pressurizer pressure is 2200 ____(2)____."

Announce: "End of update" A. (1) Wait (2) psig B. (1) Wait (2) psig and LOWERING C. (1) DO NOT wait (2) psig D. (1) DO NOT wait (2) psig and LOWERING CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: OPDP-1 defines the conduct of operations at WBNP. It contains section 5.3 which addresses various forms of briefings used by the control room staff. Among these is the crew update. As seen in OPDP-1, If a crew mem ber has information that should be know by the majority of crew members he/she pa sses this information in a concise manner by: Obtaining the crew's a ttention by stating crew update / crew announcement Ensuring crew me mbers are listening through the use of raised hands or crew members stating ready. Concisely announcing the information and Stating end of update / end of announcement. Therefore, one may obse rve that the sender of the information (in this question the sender is the OAC) is required to wait to ensure that all of the members of the crew are ready before passing the information.

OPDP-1 also contains section 7.2 which discusses the following: Communication of indicator reading s hould be provided in the format of PARAMETER - VALUE - UNITS - TREND (with rate when

appropriate). Therefor e, in accordance with OPDP-1, the minimum standard is not met by simply stating that pressure is 2200 psig. It is plausible to believe that this would be the MINIMUM standard because of the fact that one is also required to Concisely announce the information to be passed. Also, it is very well shown in the written simulator comments from both Lic ensed operator requalification training and Initial License training that the provision of trend is not uniformly made. A classic argument for the concept that trend would

not be required in this case would be that stating that pressure is 2100 psig is sufficient because this is a standard trigger value used to initiate a manual reactor trip and it does not matter what the trend is if while at 100% power, pressurizer pressure is noted to be 2100 psig. It would be patently obvious to any licensed operator that pressures were LOWERING. It is plausible to believe that one would give the crew update and immediately announce to the Unit Supervisor: US, Tripping the Unit 1 reactor. However, in accordance with OPDP-1, this is not the case.

B. Correct: As discussed it is correct t hat one would wait to obtain the recognition of all of the other members of the crew. It is corre ct that the trend must be provided in accordance with OPDP-1. C. Incorrect: As seen it is not correct that the information of the crew update can be passed before the remainder of the cr ew acknowledges such. This is continually a contentious issue both in LOR and ILT and in this case one may observe that waiting for all of the other crew members to announce ready or raise their hand may not allow the OAC time to insert a manual reactor trip (and safety injection) before an automatic action occurs. However, it is the requirement of OPDP-1 that this be done. It is very plausible that one not wait and simply pass the information because as previously discussed, the OAC would make the crew update and summarily notify t he Unit Supervisor that he was tripping the Unit 1 reactor. D. Incorrect: Again it is plausible but Incorrect that the OAC not wait before passing the information of the crew update. It is correct though that he provide value and trend.

Question Number: 66 Tier: 3 Group:

K/A: 2.1 Conduct of Operations 2.1.18 Ability to make accurate, clear, and concise logs, records, status boards, and reports. Importance Rating: 3.6 3.8 10 CFR Part 55: (CFR: 41.10 / 45.12 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to make the most concise and accurate report (crew update) procedurally allowed by OPDP-1, Conduct of Operations. Technical

Reference:

OPDP-1, Conduct of Operations Proposed references to be provided: None Learning Objective: 3-OT-SPP1000, OPDP-1, CONDUCT OF OPERATIONS and NPG-SPP-10.0 14. State the types of briefings and associated standards as described in OPDP-1.(IER 11 Precisely Controlling Plant Evolutions) Cognitive Level: Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: NPG Standard Department Procedure Conduct of Operations OPDP-1 Rev. 0034 Page 40 of 74 3.8.3 Briefi n gs (continued)

3. Shift Turnover Brief Shift briefs are conducted at the beginning and middle of each shift. Shift Management directs the briefs. Personnel participation is required; with the exception of control room monitoring that is expected to continue. (See Attachment 3 for Beginning of Shift Brief and Mid-Shift Brief).
  • If shift activities do not facilitate 100% crew attendance, a crew member knowledgeable of the watch station status will provide turnover information on that watch station. 4. Event Based Brief Event Based Briefs are conducted as determined by the Unit Supervisor utilizing a format similar to that contained in Attachment 4. Event based briefs ensure that key information and strategies are communicated in a concise manner and that crew members are given an opportunity to share critical information. Attachment 4 is not required to be filled out. It is a general guideline to follow and does not need to be in hand during the brief. 5. Crew Updates During many conditions the communication of important information to the crew may not need a scripted brief. In these cases the "Crew Update" can be used. If a crew member has information that should be known by the majority of crew members he/she passes this information in a concise manner by:
  • Obtaining the crew's attention by stating "crew update" / "crew announcement"
  • Ensuring crew members are listening through the use of raised hands or crew members stating "ready."
  • Concisely announcing the information and
  • Stating "end of update" / "end of announcement". 3.9 Understanding Plant Design And Interaction 3.9.1 Training and Qualification A. Operations training and qualification programs help ensure the knowledge and skills needed by non-licensed operators Assistant Unit Operators (AUOs), licensed reactor operators (UOs), licensed Senior Reactor O perators (SROs), Shift Technical Advisors (STAs), and Shift Managers (SMs) to effectively perform routine operations activities in the plant and properly respond to and combat unusual incidents and to anticipate and

prevent such events. B. Operations managers and supervisors are directly involved in the training and qualification of Operations personnel. This involvement includes close coordination

with the Training Department to: NPG Standard Department Procedure Conduct of Operations OPDP-1 Rev. 0034 Page 45 of 74 3.10.2 Error Prevention Tools (continued) Verif." Column. 7. Slash through the circled IV from the step just performed. D. Communication 1. Three-Way Communication is used for plant equipment manipulations, procedure performance and plant /equipment parameter information that will result in decision making, direction being given or actions being taken. See NPG-SPP-

22.202, Human Performance Tools.

  • Ensure all communications are clear, concise, and free of ambiguity.
  • For non-face-to-face verbal communication or to gain the individual's attention the sender identifies themselves by stating their name or title.
  • Use equipment noun names when practical and the phonetic alphabet when communicating component, train, channel, or procedure step designators.
  • Avoid words during verbal communication that could be mistaken for each other, like "increase' and "decrease."
  • Communication of indicator readings should be provided in the format of PARAMETER - VALUE - UNITS - TREND (with rate when appropriate).
  • Use the appropriate unit designator, system designator, or noun name and appropriate phonetic alphabet component or train designator when

communicating equipment nomenclature. FOR EXAMPLE Letdown Isolation 1-FCV-62-69 should be verbalized as "Letdown Isolation one FCV sixty two sixty nine" or "Letdown Isolation one FCV sixty two tack sixty nine." a. Sender Responsibilities (Direction)

  • Address intended receiver by name or title
  • Obtain the attention of the intended receiver and speak clearly
  • Send the intended message in just enough words to minimize the chance of receiver misunderstanding
  • Require confirmation from the intended receiver of the information by ensuring a repeat back b. Receiver Responsibilities (Repeat Back/Answer)
  • Repeat direction/request back to sender (paraphrasing is allowed).
  • If responding to a request for information, respond with "yes" or "no"
  • Maintain a questioning attitude, if you question or do not understand a person's direction then reconfirm with sender that instructions

were correct as given. NPG Standard Department Procedure Conduct of Operations OPDP-1 Rev. 0034 Page 46 of 74 3.10.2 Error Prevention Tools (continued)

  • Take no action until message is clearly understood and confirmed. c. Sender Responsibilities (Confirmation)
  • If the receiver's repeat-back is satisfactory, reply in the affirmative.
  • If unsatisfactory, reply in the negative and repeat the original communication.

EXAMPLE: THREE -WAY COMMUNICATIONS Sender (Request for Information) SRO: "Name or title, verify Reactor Coolant pressure is 1000 psig and stable" Receiver (Answer) UO: " No. Pressure is 900 psig and slowly going down." Sender (Acknowledge) SRO: "Understand" EXAMPLE: THREE -WAY COMMUNICATIONS Sender (Direction) SRO: "Name or title, place RCIC SYSTEM FLOW TRANSFER SWITCH 3-XS-71-36 Bravo in the EMERGENCY position" Receiver (Repeat Back) UO: " Place 3-XS-71-36 Bravo in the EMERGENCY position" Sender (Confirmation) SRO: "That's correct." 2. Phone and Radio Communications a. Phone Communications

  • Phone communications utilize 3 Way communications as described in this procedure.
  • Names or titles are used to ensure sender and receiver are understood.
  • Formality and professional language are the standard for phone communications to ensure miscommunications do not occur. b. Radio Communications
  • Radio communications utilize 3-Way communications as described in this procedure.
  • If reception is poor or radio usage is not allowed other methods of
67. Given the following conditions:

00:00:00 Unit 1 is at 100% power. 00:0 1:00 Unit 1 has a Turbine runback to 80%. At the times specified, ALL Excore channels indicate the following: Rated Thermal Power (%) AFD (%) 00:0 1 :05 95 (-) 3.0% 00:0 1 : 1 0 90 (-) 6.0% 00:0 1 : 1 5 85 (-) 15.5% 00:0 1 :20 80 (-) 20.0% Which ONE of the following identifies the FIRST time that T/S LCO 3.2.3, AFD will be NOT MET. REFERENCE PROVIDED A. 00:0 1 :05 B. 00:0 1 : 1 0 C. 00:0 1 : 1 5 D. 00:0 1 :20 CORRECT ANSWER: D DISTRACTOR ANALYSIS:

Question Number: 67 Tier: 3 Group:

K/A: G 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. Importance Rating: 3.9 4.2 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.12) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to use the AFD graph to determine whether or not the T/S is violated. Technical

Reference:

Unit 1 Nuclear Operating Book (NOB) Proposed references to be provided: Redacted page A-1 of the Nuclear Operating Book (NOB) Learning Objective: 3-OT-TS0302 TS 3.2 "Power Distribution Tech Specs" 2. Given a copy of the Technical Specifications and the Technical Requirements Manual, DETERMINE the following for Power Distribution Technical Specification: (IER 11-3, Controlling Plant Evolutions Precisely, Monitor Plant Indications Closely)

a. Applicable Limiting Condition for Operation for a given plant condition. Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: AFD 3.2.3 Watts Bar-Unit 1 3.2-9 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits specified in the COLR. 

-----------------------------------------------NOTE-----------------------------------------------------

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILITY: MODE 1 with THERMAL POWER > 50% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL POWER to < 50% RTP. 30 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore channel. 7 days AND Once within 1 hour and every 1 hour thereafter with the AFD monitor alarm inoperable

-dP -QI :t 0 ..... "' e QI .c: E-4 'O QI +> "' 110 100 90 80 70 60 50 r 'fl 40 30 20 10 0 -30 I I Target -.. 4.11'1 . :j I J *

  • ll r , , Unacceptable I I " " Unacceptable Operation .i , Operation
  • r ' I I l ' ' ., , , , * ' 1 ., 1 .. r r , ' ' I , , , ' I * ' .i .i , ., I ' Acceptable

, , .i .i : I : 1 Acceptable ., ., , , Operation I Operation ' ' I J, 1 ' ' " * , , L * , i: * ., 1 r 1 Desired Target Band t--20 3 0 +3 10 20 Axial Flux Difference (%) WBN 1 -dP -QI :t 0 ..... "' e QI .c: E--4 'O QI .µ "' 120 110 1 00 f-+--90 80 70 60 ' 50 , ' " 40 30 20 10 0 -30 J NUCLEAR OPERATING BOOK (NOB) Target Band vs. Power Level WBN Unit 1 Cycle 13 Target -:i.* 4.m l 1 2. D ... ' * .. r , NOB SHEET A-1 Revision 161 Page 1 of 1 '* , Unacceptable I ' I ' ' Unacceptable Operation

  • l Operation r r * ' ' ' ' , ' , , ii< l l ., I ' r , , ... ' ' . f ' , ' ' * ' \ ., ' ' l l ... ' Acceptable

'I Acceptable ., ., , , Operation I ' Operation I i .. l *

  • I L \ ' , , *-* l J , , ' l , * * ., Des i red Target Band -11
  • I ** *1 --20 3 0 +3 10 20 30 Axial Flux D ifference

(%) K2024X CURRENT AFD AND TARGET BAND CONSTANTS -28%, K2025X = -12%, K2026X = +7%, K2027X K2028 = -1.0%, K2029 = -3.0 , K2030 = 3.0 +21%,

68. Which ONE of the following describes the MCR Log Taking in accordance with 1-SI-0-2B-01, 0700-0900, Shift and Daily Surveillance Log Mode One?

The Reactor Operator will obtain data available in the MCR from _____(1)_____.

This data will be compared to criteria which are _____(2)_____ that contained in the Unit 1 Technical Specifications. A. (1) ONLY the MCR board indications (2) ALWAYS identical to B. (1) ONLY the MCR board indications (2) SOMETIMES different than C. (1) the MCR board indications and the ICS (2) ALWAYS identical to D. (1) the MCR board indications and the ICS (2) SOMETIMES different than

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: It is not correct that dat a supporting the performance of 1-SI-0-2B-01 be solely obtained from the MCR boar d indications. In fact, this surveillance will specify that some points are preferentially obtained from the Integrated Com puter System (ICS). Take, for example, CNTMT Lower Compartment Average Air Temperature. This is reference 63 in 1-SI-0-2B-01 and list s the calculated point U9020 as the source for data. U9020 is a dat a point which is produced by the ICS. It would be plausible to be lieve that one would only use the pedigreed indicators on the MCR panel s for the completion of this surveillance as such is being used to satisfy the Unit's technical specifications. As the ICS is not safety related, it would be plausible to believe that it would not be used to survey the license requirements of the Unit. However, as seen in TI-49, Compliance Inst ruments, the ICS does provide sundry inputs into the various surveillance instructions of the plant. Because the ICS is used fo r such purposes, it is subject to the calibration requirements of NPG-SPP-06.7, Instrumentation Setpoint, Scaling And Calibration Program and as long as such requirements are met, may be used to satisfy the surveillance procedures. It is also not correct that the Techni cal Specification Operability Limits are always equal to the Technical Specification Surveillance Limits. While this may be the case at many other nuclear sites, it is not the case at WBNP. Two design output drawings 47W605-242 and 47W605-243 list those parameters who possess Surveillance Limits which differ from their Operability Limits. Take the Lower Containment Average Air Temper ature for consideration again.

T/S LCO 3.6.5, Containment Ai r Temperature states that: Containment average air temperature shall be: b. > 100°F and < 120°F for the containment lower compartment. Reference 63 of 1-SI-0-2B-01 lists that the T S Limi t of the same parameter is > 102.5°F and < 117.5°F. Therefore, an average temperature of 119.5 would render the lower containment inoperable even though its temperature was inside of the band contained within t he T/S LCO. The two drawings of

the 47W605 series captured the co mmitment which WBNP made to incorporate instrument errors into the surveillance instructions. WBNP did not solely use the instrument errors already built into the standardized improved Technical Specifications. It is plausible to believe that the surveillance satisfying the Technical Specifications would use the requirement s of the Technical Specifications as this would be common sense as well as the practice at many other nuclear

plants. B. Incorrect: Again, it is not correct and yet plausible that data supporting the performance of 1-SI-0-2B-01 be solely obtained from the MCR board indications. Also, it is correct that the acceptance criter ia of the SI are not always the same as that c ontained in the Unit's technical specifications. C. Incorrect: It is correct that the IC S will be used in conc ert with the MCR board indications to obtain data for 1-SI-0-2B-01. Also, it is not correct and yet plausible that the acc eptance criteria of the SI is the same as that contained in the Unit's technical specifications. D. Correct: It is correct that the ICS will be used in concer t with the MCR board indications to obtain data for 1-SI-0-2B-

01. Also, it is correct that the acceptance criteria of the SI ar e not always the same as that

contained in the Unit's technical specifications.

Question Number: 68 Tier: 3 Group:

K/A: G 2.2.12 Knowledge of surveillance procedures. Importance Rating: 3.7 4.1 10 CFR Part 55: (CFR: 41.10 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to have the fundamental (generic) knowledge of which sources of data are allowed to be used for surveillance procedures. 1-SI-0-2B-01 was selected for specificity and clarity but the concept is applicable to the surveillance procedures at WBN. The applicant is then required to understand the source of the acceptance criteria.. Technical

Reference:

1-SI-0-2B-01 TI-49, Compliance Instruments T/S LCO 3.6.5, Containment Air Temperature 47W605-242 Proposed references to

be provided: None Learning Objective: 3-OT-TS0000 Technical Specification Overview 12. EXPLAIN briefly the information that can be found in the Technical Specification and Technical Requirements "Surveillance Requirements" section. Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 0700 - 1900 Shift And Daily Surveillance Log Mode One 1-SI-0-2B-01 Rev. 0066 Page 40 of 47 Data Sheet 1 (Page 20 of 24) Mode 1 Surveillance Log: 0700 - 1900 Date ________ Ref Location Description Instrument # T S Limit Data Acc NOT Met 60 ICS SGBD flow to CTBD F0619A OR 1-FIT-15-42 Operable and Channel Check gpm If inoperable, SRO is to be notified and Section 6.1L should be referenced. A flow reading from either the computer point or the local indicator will satisfy the channel check requirement. 1-PNL-276-L657 T8J/712 gpm If either the computer point or the local indicator is inoperable, a SR is to be initiated and SR number recorded in remarks. 61 T8J/685 OR ICS Station Sump Discharge Radiation Mon 0-RI-90-212A OR point R1010A Operable and Channel Check cpm If monitor is inoperable, notify SRO, and Chemistry Countroom per Section 6.1K. 62 S4Sn/741 0-L-399 OR ICS SBVS Duct Monitor Stack Volume 0-FI-90-320/1B OR point F2702A Operable and Channel Check scfm (4) If inoperable, SRO is to be notified and Section 6.0. L should be referenced. Perform Data Sheet 62 every 4 hours until declared Operable. 63 ICS CNTMT Upper Compartment Average Air Temperature U9019 87.1°F to 107.9°F (3) °F (1) If ICS is INOPERABLE, Data Sheet 63 must be performed. CNTMT Lower Compartment Average Air Temperature U9020 102.5°F to 117.5°F (2) °F (1) (1) Minimum number of operable channels is verified by computer point routines. Points will indicate unreliable if less than the required number of inputs are available. A print value review of points T1000A through T1033A can be performed to determine which point s are invalid. (2) Maximum Lower Compartment average air temperature can be raised to 119 °F, and Minimum Lower Compartment average air temperature can be lowered to 101 °F, if ALL of the temperature inputs and U9020 are reliable (NONE of the computer points have a quality code of "BAD"). The points can be displayed on computer Primary Mimic CONTAINMENT TEMPERATURES. (3) Maximum Upper Compartment average air temperature can be raised to 108.5 °F, and Minimum Upper Compartment average air temperature can be lowered to 86.5 °F, if ALL of the temperature inputs and U9019 are reliable (NONE of the computer points have a quality code of "BAD"). The points can be displayed on computer Primary Mimic CONTAINMENT TEMPERATURES. (4) Inoperable if display is 3,000 SCFM Initials indicate acc met OR action in "Acc not Met" column initiated. Performer's Initials SRO'sInitials Remarks:

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 33 of 48 Appendix B (Page 1 of 15) Compliance ICS Computer Point List Date ________ This list consists of all ICS co mputer points used for compliance. The following column headings are used: COMPUTER POINT - The ICS computer point used as a Compliance device by the listed Instruction. LOOP/INSTRUMENT - The loop number associated with the ICS computer point. INSTRUCTION - The number(s) of the Instruction where the IC S computer point is used. NOTES 1) These Offsite Dose Instructions (ODIs) record data taken from their associated System Operating Instruction (SOI). The applicable computer points are listed in the SOI. 2) Devices which can NOT be calibrated. Some examples are: thermocouples which input directly to the plant computer; computer calculated points where the individual inputs to the calculation are calibrated under their specific loops; or computer points normalized by the computer such as Incore flux detectors.

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 34 of 48 Appendix B (Page 2 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES F0125A 1-LPF-62-40 1-SI-68-33 0 F0127A 1-LPF-62-27 1-SI-68-33 0 F0129A 1-LPF-62-14 1-SI-68-33 0 F0131A 1-LPF-62-1 1-SI-68-33 0 F0400A 1-LPF-68-6A 1-SI-0-2 0 F0401A 1-LPF-68-6B 1-SI-0-2 0 F0402A 1-LPF-68-6D 1-SI-0-2 0 F0403A 1-LPF-3-35A 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-41 0 F0404A 1-LPF-3-35B 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-41 0 F0407A 1-LPF-3-235 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0408A 1-LPF-1-152 1-SI-0-20 0 1-SI-0-24 0 1-SI-1-905 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0420A 1-LPF-68-29A 1-SI-0-2 0 F0421A 1-LPF-68-29B 1-SI-0-2 0 F0422A 1-LPF-68-29D 1-SI-0-2 0 F0423A 1-LPF-3-48A 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0424A 1-LPF-3-48B 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0427A 1-LPF-3-238 1-SI-0-24 0 1-SI-92-1 0 WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 35 of 48 Appendix B (Page 3 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES F0427A (cont) 1-LPF-3-238 TI-6 0 TI-41 0 F0428A 1-LPF-1-156 1-SI-0-20 0 1-SI-0-24 0 1-SI-1-905 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0440A 1-LPF-68-48A 1-SI-0-2 0 F0441A 1-LPF-68-48B 1-SI-0-2 0 F0442A 1-LPF-68-48D 1-SI-0-2 0 F0443A 1-LPF-3-90A 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0444A 1-LPF-3-90B 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0447A 1-LPF-3-241 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0448A 1-LPF-1-160 1-SI-0-20 0 1-SI-0-24 0 1-SI-1-905 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0460A 1-LPF-68-71A 1-SI-0-2 0 F0461A 1-LPF-68-71B 1-SI-0-2 0 F0462A 1-LPF-68-71D 1-SI-0-2 0 F0463A 1-LPF-3-103A 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 36 of 48 Appendix B (Page 4 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES F0463A (cont) 1-LPF-3-103A TI-6 0 TI-41 0 F0464A 1-LPF-3-103B 1-SI-0-20 0 1-SI-0-24 0 1-SI-68-31 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0467A 1-LPF-3-244 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0468A 1-LPF-1-164 1-SI-0-20 0 1-SI-0-24 0 1-SI-1-905 0 1-SI-92-1 0 TI-6 0 TI-41 0 F0619A 1-FIT-015-42 1-ODI-90-2 0 1-SI-0-2 0 F2702A 0-LPF-90-320 0-ODI-90-8 0 F2704A 0-LPF-90-300 0-ODI-90-22 0 F9015A 2-LPF-90-400 1-ODI-90-26 0 L0112A 1-LPL-62-130 1-SI-68-32 0 L0480A 1-LPL-68-339 1-SI-68-32 0 L0481A 1-LPL-68-335 1-SI-68-32 0 L0482A 1-LPL-68-320 1-SI-68-32 0 L0485A 1-LPL-68-300 1-SI-68-32 0 L1040A 1-LPL-63-50 1-SI-63-911 0 1-SI-63-912 0 L1041A 1-LPL-63-51 1-SI-63-911 0 1-SI-63-912 0 LEFMSTAT 1-SI-92-1 2 N0001A 1-SI-0-15 2 1-SI-0-20 2 1-SI-0-22 2 1-SI-92-2 2 1-SI-92-3 2 1-SI-92-4 2 TI-41 2 N0002A 1-SI-0-15 2 1-SI-0-20 2 1-SI-0-22 2 1-SI-92-2 2 1-SI-92-3 2 1-SI-92-4 2 TI-41 2 WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 37 of 48 Appendix B (Page 5 of 15) Compliance ICS Computer Point List N0003A 1-SI-0-15 2 1-SI-0-20 2 1-SI-0-22 2 1-SI-92-2 2 1-SI-92-3 2 1-SI-92-4 2 TI-41 2 N0004A 1-SI-0-15 2 1-SI-0-20 2 1-SI-0-22 2 1-SI-92-2 2 1-SI-92-3 2 1-SI-92-4 2 TI-41 2 N0005A 1-SI-0-15 2 1-SI-0-20 2 1-SI-0-22 2 1-SI-92-2 2 1-SI-92-3 2 1-SI-92-4 2 TI-41 2 N0006A 1-SI-0-15 2 1-SI-0-20 2 1-SI-0-22 2 1-SI-92-2 2 1-SI-92-3 2 1-SI-92-4 2 TI-41 2 N0041A 1-INM-92-1/301 1-SI-0-11 0 1-SI-0-21 0 1-SI-0-22 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0042A 1-INM-92-1/302 1-SI-0-11 0 1-SI-0-21 0 1-SI-0-22 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0043A 1-INM-92-2/301 1-SI-0-11 0 1-SI-0-21 0 1-SI-0-22 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 38 of 48 Appendix B (Page 6 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES N0043A (cont) 1-INM-92-2/301 TI-41 0 N0044A 1-INM-92-2/302 1-SI-0-11 0 1-SI-0-22 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0045A 1-INM-92-3/301 1-SI-0-11 0 1-SI-0-21 0 1-SI-0-22 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0046A 1-INM-92-3/302 1-SI-0-11 0 1-SI-0-21 0 1-SI-0-22 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0047A 1-INM-92-4/301 1-SI-0-11 0 1-SI-0-21 0 1-SI-0-22 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0048A 1-INM-92-4/302 1-SI-0-11 0 1-SI-0-21 0 1-SI-0-22 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0049A 1-INM-92-1/303 1-SI-0-10 0 1-SI-0-11 0 1-SI-0-20 0 1-SI-0-24 0 1-SI-0-28 0 1-SI-92-1 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0050A 1-INM-92-2/303 1-SI-0-10 0 1-SI-0-11 0

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 39 of 48 Appendix B (Page 7 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES N0050A (cont) 1-INM-92-2/303 1-SI-0-20 0 1-SI-0-24 0 1-SI-0-28 0 1-SI-92-1 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0051A 1-INM-92-3/303 1-SI-0-10 0 1-SI-0-11 0 1-SI-0-20 0 1-SI-0-24 0 1-SI-0-28 0 1-SI-92-1 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 N0052A 1-INM-92-4/303 1-SI-0-10 0 1-SI-0-11 0 1-SI-0-20 0 1-SI-0-24 0 1-SI-0-28 0 1-SI-92-1 0 1-SI-92-2 0 1-SI-92-3 0 1-SI-92-4 0 TI-41 0 P0142A 1-LPP-62-92 1-SI-68-33 0 P0400A 1-LPP-1-2A 1-SI-0-20 0 1-SI-0-24 0 1-SI-68-31 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0401A 1-LPP-1-2B 1-SI-0-20 0 1-SI-0-24 0 1-SI-68-31 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0402A 1-LPP-1-5 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 40 of 48 Appendix B (Page 8 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES P0403A 1-LPP-3-37 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0420A 1-LPP-1-9A 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0421A 1-LPP-1-9B 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0422A 1-LPP-1-12 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0423A 1-LPP-3-50 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0440A 1-LPP-1-20A 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0441A 1-LPP-1-20B 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0442A 1-LPP-1-23 1-SI-0-20 0 1-SI-0-24 0

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 41 of 48 Appendix B (Page 9 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES P0442A (cont) 1-LPP-1-23 1-SI-92-1 0 TI-6 0 TI-41 0 P0443A 1-LPP-3-92 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0460A 1-LPP-1-27A 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0461A 1-LPP-1-27B 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0462A 1-LPP-1-30 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0463A 1-LPP-3-105 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 P0480A 1-LPP-68-340 1-SI-0-23 0 1-SI-68-32 0 P0481A 1-LPP-68-334 1-SI-68-33 0 P0482A 1-LPP-68-323 1-SI-68-33 0 P0483A 1-LPP-68-322 1-SI-68-33 0 P0499A 1-LPP-68-66A 1-SI-68-44 0 P0701A 1-LPP-68-63 1-SI-68-44 0 P4005A 1-LPP-30-126 1-SI-0-2 0 P4006A 1-LPP-30-127 1-SI-0-2 0 WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 42 of 48 Appendix B (Page 10 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES P0702A 1-LPP-68-64 1-SI-68-44 0 Q0340A 1-ET-57-16A 1-SI-0-23 2 P1000A 1-LPD-30-42 1-SI-0-2 0 P1001A 1-LPD-30-43 1-SI-0-2 0 P1002A 1-LPD-30-44 1-SI-0-2 0 P1003A 1-LPD-30-45 1-SI-0-2 0 R0001A 1-LPR-90-119 1-ODI-90-25 0 R0011A 0-LPR-90-132B 0-ODI-90-8 0 R0016A 0-LPR-90-118 0-ODI-90-5 1 R0020A 0-LPR-90-101B 0-ODI-90-22 0 R1010A 0-LPR-90-212 0-ODI-90-3 0 R1011A 0-LPR-90-225 0-ODI-90-23 1 R1020A 1-LPR-90-120 1-ODI-90-2 1 R1021A 1-LPR-90-121 1-ODI-90-2 1 R1022A 1-LPR-90-122 0-ODI-90-1 1 R1027A 1-LPR-90-130 1-ODI-90-15 0 1-ODI-90-18 0 R1028A 1-LPR-90-131 1-ODI-90-15 0 1-ODI-90-18 0 R9101A 1-LPR-90-400 1-ODI-90-15 0 1-ODI-90-18 0 1-ODI-90-26 0 R9102A 2-LPR-90-400 1-ODI-90-26 0 T0400A 1-LPT-68-2 1-SI-0-10 0 1-SI-0-11 0 1-SI-0-23 0 1-SI-0-27 0 1-SI-68-32 0 T0403A 1-LPT-68-2 1-SI-0-23 0 T0406A 1-LPT-68-18 1-SI-0-10 0 T0418A 1-LPT-3-36 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 TI-6.001 0 T0419A 1-LPT-68-1 1-SI-68-44 0 T0420A 1-LPT-68-25 1-SI-0-10 0 1-SI-0-11 0 1-SI-0-23 0 1-SI-0-27 0 1-SI-68-32 0

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 43 of 48 Appendix B (Page 11 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES T0423A 1-LPT-68-25 1-SI-0-23 0 T0426A 1-LPT-68-41 1-SI-0-10 0 T0438A 1-LPT-3-49 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 TI-6.001 0 T0439A 1-LPT-68-24 1-SI-68-44 0 T0440A 1-LPT-68-44 1-SI-0-10 0 1-SI-0-11 0 1-SI-0-23 0 1-SI-0-27 0 1-SI-68-32 0 T0443A 1-LPT-68-44 1-SI-0-23 0 T0446A 1-LPT-68-60 1-SI-0-10 0 T0458A 1-LPT-3-91 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 TI-6.001 0 T0459A 1-LPT-68-43 1-SI-68-44 0 T0460A 1-LPT-68-67 1-SI-0-10 0 1-SI-0-11 0 1-SI-0-23 0 1-SI-0-27 0 1-SI-68-32 0 T0463A 1-LPT-68-67 1-SI-0-23 0 T0466A 1-LPT-68-83 1-SI-0-10 0 T0478A 1-LPT-3-104 1-SI-0-20 0 1-SI-0-24 0 1-SI-92-1 0 TI-6 0 TI-41 0 TI-6.001 0 T0479A 1-LPT-68-65 1-SI-68-44 0 T0480A 1-LPT-68-319 1-SI-68-44 0 T0481A 1-LPT-68-324 1-SI-68-44 0 T0497A 1-LPT-68-2 1-SI-92-1 0 T0630A 1-LPT-74-14 1-SI-68-44 0 T0631A 1-LPT-74-25 1-SI-68-44 0

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 44 of 48 Appendix B (Page 12 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES T1000A 1-TE-30-210A 1-SI-0-2 2 T1001A 1-TE-30-210B 1-SI-0-2 2 T1002A 1-TE-30-210C 1-SI-0-2 2 T1003A 1-TE-30-210D 1-SI-0-2 2 T1004A 1-TE-30-210E 1-SI-0-2 2 T1005A 1-TE-30-210F 1-SI-0-2 2 T1014A 1-TE-30-210O 1-SI-0-2 2 T1015A 1-TE-30-210P 1-SI-0-2 2 T1016A 1-TE-30-210Q 1-SI-0-2 2 T1017A 1-TE-30-210R 1-SI-0-2 2 T1018A 1-TE-30-210S 1-SI-0-2 2 T1019A 1-TE-30-210T 1-SI-0-2 2 T1020A 1-TE-30-210U 1-SI-0-2 2 T1021A 1-TE-30-210V 1-SI-0-2 2 T1022A 1-TE-30-210W 1-SI-0-2 2 T1023A 1-TE-30-210X 1-SI-0-2 2 T1024A 1-TE-30-210Y 1-SI-0-2 2 T1025A 1-TE-30-210Z 1-SI-0-2 2 T1026A 1-TE-30-210AA 1-SI-0-2 2 T1027A 1-TE-30-210AB 1-SI-0-2 2 T1028A 1-TE-30-210AC 1-SI-0-2 2 T1029A 1-TE-30-210AD 1-SI-0-2 2 T1030A 1-TE-30-210AE 1-SI-0-2 2 T1031A 1-TE-30-210AF 1-SI-0-2 2 T1032A 1-TE-30-210AG 1-SI-0-2 2 T1033A 1-TE-30-210AH 1-SI-0-2 2 T0418A 1-LPT-3-36 TI-6.001 0 T0438A 1-LPT-3-49 TI-6.001 0 T0458A 1-LPT-3-91 TI-6.001 0 T0478A 1-LPT-3-104 TI-6.001 0 T2612A 1-LPT-67-455 1-SI-0-2 0 T2613A 1-LPT-67-456 1-SI-0-2 0 T2614A 2-LPT-67-455 1-SI-0-2 0 T2615A 2-LPT-67-456 1-SI-0-2 0 T2816A 0-TE-65-56 0-SI-65-6-A 2 T2817A 0-TE-65-57 0-SI-65-6-A 2 T2819A 0-TE-65-62 0-SI-65-6-B 2 T2836A 0-TE-65-61 0-SI-65-6-B 2

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 45 of 48 Appendix B (Page 13 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES U0049 Stp Ct CBAG1,2 1-SI-0-15 2 U0050 Stp Ct CBBG1,2 1-SI-0-15 2 U0051 Stp Ct CBCG1,2 1-SI-0-15 2 U0052 Stp Ct CBDG1,2 1-SI-0-11 2 1-SI-0-12 2 1-SI-0-15 2 1-SI-0-20 2 1-SI-0-23 2 1-SI-0-24 2 TI-41 2 U0053 Stp Ct SBAG1,2 1-SI-0-15 2 U0054 Stp Ct SBBG1,2 1-SI-0-15 2 U0055 Stp Ct SBCG1 1-SI-0-15 2 U0056 Stp Ct SBDG1 1-SI-0-15 2 U0484 1-SI-0-12 2 1-SI-0-24 2 U0900 1-LPF-2-256 1-ODI-90-25 2 U0902 0-LPF-90-320 0-ODI-90-8 2 U0903 1-LPF-90-400 1-ODI-90-26 2 U0904 0-LPF-90-300 0-ODI-90-22 2 U0905 1-SI-0-10 2 1-SI-0-11 2 U0906 1-SI-0-10 2 1-SI-0-11 2 U0907 1-SI-0-10 2 1-SI-0-11 2 U0908 1-SI-0-10 2 1-SI-0-11 2 1-SI-0-23 2 U0909 1-SI-0-10 2 1-SI-0-11 2 1-SI-0-23 2 U0910 1-SI-0-10 2 1-SI-0-11 2 U1118 1-SI-0-10 2 1-SI-0-11 2 1-SI-0-12 2 1-SI-0-20 2 1-SI-0-24 2 1-SI-92-1 2 1-SI-92-2 2 1-SI-92-3 2 TI-41 2 U1125 1-SI-0-2 2 U1126 1-SI-0-2 2

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 46 of 48 Appendix B (Page 14 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES U1127 1-SI-0-10 2 1-SI-0-11 2 1-SI-0-12 2 1-SI-0-20 2 1-SI-0-24 2 1-SI-92-1 2 1-SI-92-2 2 1-SI-92-3 2 TI-41 2 U1141 N43 DELTA-FLUX 1-SI-0-21 2 1-SI-92-3 2 U1142 N42 DELTA-FLUX 1-SI-0-21 2 1-SI-92-3 2 U1143 N44 DELTA-FLUX 1-SI-0-21 2 1-SI-92-3 2 U1144 N41 DELTA-FLUX 1-SI-0-21 2 1-SI-92-3 2 U1145 1-SI-0-21 2 U1146 1-SI-0-21 2 U1151 1-INM-92-3/302 1-SI-0-21 2 U1152 1-INM-92-2/302 1-SI-0-21 2 U1153 1-INM-92-4/302 1-SI-0-21 2 U1154 1-INM-92-1/302 1-SI-0-21 2 U1159 1-INM-92-3/301 1-SI-0-21 2 U1160 1-INM-92-2/301 1-SI-0-21 2 U1161 1-INM-92-4/301 1-SI-0-21 2 U1162 1-INM-92-1/301 1-SI-0-21 2 U1169 1-SI-0-10 2 U1254 1-SI-92-1 2 U2118 1-SI-0-10 2 1-SI-0-11 2 1-SI-0-12 2 1-SI-0-20 2 1-SI-0-24 2 1-SI-92-1 2 1-SI-92-2 2 1-SI-92-3 2 TI-41 2 U2125 1-SI-0-2 2 U2126 1-SI-0-2 2 U9019 1-SI-0-2 2 U9020 1-SI-0-2 2 UF1015 2-LPF-90-400 1-ODI-90-26 2 UL1005 1-SI-68-32 2 UN2000 1-SI-68-32 2

WBN Unit 1 Compliance Instruments TI-49 Rev. 0042 Page 47 of 48 Appendix B (Page 15 of 15) Compliance ICS Computer Point List ICS COMPUTER POINT LOOP /INSTRUMENT INSTRUCTION NOTES UT5000 1-SI-68-32 2 Y0703A 1-LPF-3-147A 1-SI-3-912 0 Y0704A 1-LPF-3-155B 1-SI-3-912 0 Y0708A 1-LPF-3-163B 1-SI-3-912 0 Y0709A 1-LPF-3-170A 1-SI-3-912 0 Y2203A 1-LPF-90-400 1-ODI-90-15 0 1-ODI-90-18 0 1-SI-0-2 0 Y8000A 0-LPF-27-98 0-PI-ENV-3.1 0 Y8001A 0-TT-27-99 0-PI-ENV-3.1 0

Containment Air Temperature 3.6.5 Watts Bar-Unit 1 3.6-16 3.6 CONTAINMENT SYSTEMS

3.6.5 Containment Air Temperature

LCO 3.6.5 Containment average air temperature shall be:

a. 85F and 110F for the containment upper compartment, and
b. 100F and 120F for the containment lower compartment.
  -------------------------------------------------NOTE-------------------------------------------------

The minimum containment average air temperatures in MODES 2, 3, and 4 may be reduced to 60 F. -----------------------------------------------------------------------------------------------------------

APPLICABILITY: MODES 1, 2, 3, and

4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air temperature not within limits. A.1 Restore containment average air temperat ure to within limits. 8 hours B. Required Action and associated Completion Time not met. B.1 Be in MODE

3. AND B.2 Be in MODE
5. 6 hours 36 hours
69. Which ONE of the following describes the Modes of Oper ation in accordance with the Unit 1 Technical Specifications?

Unit 1 will enter MODE 5 from MODE 4 when ____(1)____ is ____(2)____. (1) (2) A. Tavg 200 °F B. Tavg 350 °F C. Tcold 200 °F D. Tcold 350 °F

CORRECT ANSWER: A DISTRACTOR ANALYSIS: A. Correct: As seen in Table 1.1-1 of the Unit 1 Technical Specifications, Average Reactor Coolant Temperature (Tavg) is used as a factor into the Unit's mode of operation. Also seen in this table is that Mode 5 is achieved when Tavg < 200°F. B. Incorrect: It is correct that Tavg is used as a factor into the Unit's mode of operation. It is not correct that Mode 5 is achieved when Tavg <350°F. When Tavg is <350, the Unit enters Mode 4. C. Incorrect: It is not correct that Tcold be us ed as a factor into the Unit's mode of operation. It is plausible to believe that it is because Tcold is used to satisfy the monitoring of several other Technical Specifications. For example, Tcold is used to determi ne the heatup and cooldown rates of the RCS and thus, the monitoring of T/S LCO 3.4.3, RCS Pressure and Temperature (P/T) Limits. Tcold is also an input into the cold over-pressure mitigation system and thus used to satisfy T/S LCO 3.4.12. It is correct that Mode 5 is achieved when Tavg < 200°F. D. Incorrect: Again it is Incorrect and yet plausible that Tcold be used as a factor into the Unit's mode of operation. It is not correct that Mode 5 is achieved when Tavg < 350°F. When Tavg is <350, the Unit enters Mode 4. Question Number: 69 Tier: 3 Group:

K/A: G 2.2.35 Ability to determine Technical Specification Mode of Operation. Importance Rating: 3.6 4.5 10 CFR Part 55: (CFR: 41.7 / 41.10 / 43.2 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to correctly identify the appropriate point at which Mode 5 is entered. The applicant must also identify the correct temperature (Tavg vs. Tcold) to use when determining a mode change. In doing so, the applicant is exercising the knowledge required to determine the Mode of Operation.. Technical

Reference:

Unit 1 Technical Specifications Proposed references to be provided: None Learning Objective: 3-OT-TS0000 Technical Specification Overview 9. State the number and respective title of the different operational modes according to Technical Specifications. Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: Definitions 1.1 Watts Bar-Unit 1 1.1-7 Table 1.1-1 (page 1 of 1) MODES MODE TITLE REACTIVITY CONDITION (k eff)  % RATED THERMAL POWER (a) AVERAGE REACTOR COOLANT TEMPERATURE (F) 1 Power Operation 0.99 > 5 NA 2 Startup 0.99 5 NA 3 Hot Standby 

< 0.99 NA  350 4 Hot Shutdown (b)
< 0.99 NA 350 > T avg > 200 5 Cold Shutdown (b)
< 0.99 NA  200 6  Refueling (c)

NA NA NA (a) Excluding decay heat.

(b) All reactor vessel head closure bolts fully tensioned.

(c) One or more reactor vessel head closure bolts less than fully tensioned.

70. Given the following conditions:
- Unit 1 is at 7% power

In accordance with TI-12.07A, Containment Access Modes 1-4, which ONE of the following tasks does NOT pose a radiological hazard r equiring authorization for Special Access into containment?

A. Physically checking a lower ice condenser door shut B. Locally inspecting the RCP #2 oil reservoir sight glasses C. Isolating a leak on the letdown line by closing 1-ISV-68-580 (IC/705 AZ 221) D. Venting the RHR suction header using 1-TV-74-504 and 1-TV-74-543 (716-AZ 301 #4 Accum Rm)

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: The lower ice condenser doors are inside of the polar crane wall. This distractor is actually the product of a common ILT misconception which is realized when an operator receives an Ice Condenser Door Open annunciator. With no other indications except this annunciator, the student will inevitably dispatch a simulated operator to inspect the door (as the existence of such presents a short term technical specification LCO required action). Additionally, it is possible to inspect some ice condenser lower doors by simply stepping a foot or two out of the Upper polar crane wall door in Accumulator room #4. B. Incorrect: The inspection of the RCP oil sight glasses requires an operator to enter inside of the polar crane wall. Concerning RCP #2, one would cross the seal table area and then open the seal table area to Loop #2 polar crane wall door. The operator could then wall a few feet past that door before looking up and to his left to inspect the sight glasses. C. Incorrect: 1-ISV-68-580 is a manual valve located just inside the polar crane wall door on the Loop 2/3 side of lower containment. This isolation is the first letdown isolation off of the RCS loop #3. Because this valve is located inside of the polar crane wall the operation of such when in Modes 1 or 2 will require special authorization in accordance with TI-12.07A. As seen in TI-12.07A, Special Access authorization is also required for entries inside the Polar Crane Wall or Regen Hx Room during Modes 1 or 2. It is plausible to believe that this task would require general access authorization because it is a task which would be of short duration and would require an operator to step only about two feet through the Loop 2/3 door. Also, one may believe that 1-ISV-68-580 were located in an area outside of the polar crane wall. Letdown isolations are provided which are outside of the polar crane wall. D. Correct: The venting of the RHR common suction header is performed almost at the center of the accumulator room #4. This room is between the steel containment vessel and the polar crane wall. 1-TV-74-504 and 543 are 1/2

kerotest valves and do not provide an exceptionally fast venting of the large common suction line piping. Additionally, the location of these valves subjects the operator to a fairly high radiation field. Therefore, it is outside of the polar crane wall and even though that the venting would be relatively slow and present the operator to a fairly substantial dose, not subject to any other authorization but general access.

Question Number: 70 Tier: 3 Group:

K/A: 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Importance Rating: 3.4 3.8 10 CFR Part 55: (CFR: 41.12 / 43.4 / 45.10) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to understand the radiological risks of several abnormal tasks and then to be able to recognize which permissions are required for those tasks. Technical

Reference:

TI-12.07A, Containment Access Modes 1-4 Proposed references to be provided: None Learning Objective: 3-OT-RAD0003 , RADIATION PROTECTION STANDARDS AND GUIDELINES 8. Identify the responsibilities of the following concerning the ALARA program: a. Radiation Protection Manager

b. TVA NPG Organization
c. Employee Cognitive Level:

Higher X Lower Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments:

  • .. 1' I I I These are the lower ice doors which are located inside of the polar crane wall. This one is located just inside of the door in the polar crane wall which connects the upper level of the #4 accumulator room to the ice deck.

This is the oil sightglass on the #2 RC P. It is located off of a balcony inside of the polar crane wall. ' ( This valve is located just inside of the Polar Crane wall door for the Loop 2/3 side of containment. It is INSIDE of the polar crane wall. 1-ISV-68-580 I I

  • WBN Unit 1 Reactor Coolant System Pressurizer Relief Tank Operations Valve Checklist 1-68.01-1V 1-SOI-68.01 ATT 1V Revision 0000 Effective Date: 04-07-2014

Page 4 of 6 Performed On ________ NOMENCLATURE LOCATION POSITION UNID INITIALS INITIALS Inside Polar Crane Wall (Cont.) PRESSURIZER VENT IC/788 AZ 105 1CLOSED 1-VTV-68-577 IVRCS LOOP 3 HOT LEG SAMPLE IC/720 AZ 194 OPEN 1-SMV-68-578 IV RCS LOOP 3 EXCESS LETDOWN ISOL IC/716 AZ 235 OPEN 1-ISV-68-579 IVRCS LOOP 3 LETDOWN ISOL IC/705 AZ 221 OPEN 1-ISV-68-580 IVRCS LOOP 3 DRAIN IC/704 AZ 221 CLOSED 1-DRV-68-581 IVRCS LOOP 3 DRAIN IC/704 AZ 221 CLOSED 1-DRV-68-582 IVCVCS LETDOWN RELIEF DISCH DRAIN IC/712 AZ 60 CLOSED 1-DRV-68-584 IV RCS LOOP 2 PZR SPRAY LINE DRAIN IC/732 AZ 110 1 CLOSED 1-DRV-68-593 IV PRESSURIZER SPRAY LINE

VENT IC/783 AZ 105 1CLOSED 1-VTV-68-594 IVREACTOR VESSEL HEAD VENT

LINE VENT IC/755 AZ 30 1 , 2 CLOSED 1-VTV-68-604 IVCVCS LETDOWN RELIEF DISCH DRAIN IC/710 AZ 60 1 CLOSED 1-DRV-68-605 IV 1 VERIFY cap/blind flange installed 2 May be N/A'd if Reactor Vessel Head removed. Annulus PRESSURIZER RELIEF TANK GAS ANALYZER SUPPLY OC/716 AZ 315 OPERABLE OPEN 1-FCV-68-307 IVCONTROL AIR ISOLATION VALVE TO 1-FCV-68-307 OC/716 AZ 315 OPEN 1-ISV-32-3499 IV BIT Room PRESSURIZER RELIEF TANK NITROGEN SUP PRESS CNTL A4X/713 BIT RM OPERABLE 1-PCV-68-304 IVCONTROL AIR ISOLATION VALVE TO 1-PCV-68-304 A4X/713 BIT RM OPEN 1-ISV-32-3199 IVPRESSURIZER RELIEF TANK NITROGEN SUP FLOW CNTL A4X/713 BIT RM OPEN 1-FCV-68-305 IVCONTROL AIR ISOLATION VALVE TO 1-FCV-68-305 A4X/713 BIT RM OPEN 1-ISV-32-3198 IV1-FCV-68-304 CNTL ISOL A4X/713 BIT RM OPEN 1-RTV-68-307A IVPRESSURIZER RELIEF TANK N 2 SUP HDR TEST VENT A4X/713 BIT RM CLOSED 1-TV-68-847 IV WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 7 of 50 2.2 Developmental References (continued) I. RCI-128, ALARA Pr ogram Implementation J. TI-134, Control of Portable Two Way Radios K. TI-229, Temporary Shielding Program L. WBN FSAR Questions 22.26, 212.116, and 212.129 3.0 PRECAUTIONS AND LIMITATIONS 3.1 General Precautions and Limitations A. All access portals (Airlocks el 757/716, equipment hatches el 757, and Annulus el 713) to the Containment buildi ng SHALL be controlled to prevent unauthorized entry while in Modes 1 through

4. Radiation Protection (RP) shall maintain positive access control of Containment and Annulus in accordance with RP procedures. B. All entries into Containment or Annulus while in Modes 1 through 4 are authorized by completion of applic able sections of Appendix A, Containment/Annulus Entry Authorization. C. One Appendix A is required for eac h area entered (Upper Containment, Lower Containment, Annulus). Appendix A's ar e typically issued for a 24 hour period only. However the SM may approv e an extension beyond 24 hours.

Authorization requirements for entries are described as follows:

1. General Access is authorized by the comple tion of Appendix A, Section 1.0, which requires the incore detectors to be TAGGED and in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). This ensures that a detector is in a location which would not expose entry personnel. In addition, this section also permits entry inside the Polar Crane Wall when below Mode 2.
2. Special Access is authorized by completion of Appendix A, Sect ion 2.0, for any entry into Containment or Annulus when the incore detectors are NOT TAGGED or NOT in their approved storage location. Special Access authorization is also required for ent ries inside the Polar Crane Wall or

Regen Hx Room during Modes 1 or 2. 3. All entries requiring Special Access Authorization shall require approvals in accordance with RCI-128, ALARA Program Implementation to ensure that appropriate controls are established to prevent personnel overexposure (i.e., Special RWP and APR). These valves are #4 Accumulato located in the room is outsid r room. This crane wall. e of the polar 1*TV*14*543 TESi vtrrr WI Di 11 \ I \\,t-I \ \'.\ \\\'.\ \I I\ '\. \ ' \ WBN Unit 1 Residual Heat Removal System Valve Checklist 1-74.01-1V 1-SOI-74.01 ATT 1V Revision 0000

Effective Date: 12-03-2013

Page 1 of 2 Performed On ________ NOMENCLATURE LOCATION POSITION UNID PERF INITIAL VERIFIER INITIAL Containment RHR SUCTION HDR VLV LOW POINT DRAIN 702-AZ315 Inside Polar Crane Wall CLOSED 1-DRV-74-542 IVRHR SUCTION HDR VLV LOW POINT DRAIN 702-AZ315 Inside Polar Crane Wall CLOSED 1-DRV-74-503 IVRHR SUCTION HDR DRAIN 702-AZ345 Inside Polar Crane Wall CLOSED 1-DRV-74-500 IVRHR SUCTION HDR DRAIN 702-AZ345 Inside Polar Crane Wall CLOSED 1-DRV-74-541 IVLOOP 4 HOT LEG TO RHR SUCTION 706-AZ345 Inside Polar Crane Wall 0-PI-OPS-17.0 1-FCV-74-1 1-FCV-74-1 BYPASS RHR SUCTION 706-AZ345 Inside Polar Crane Wall 0-PI-OPS-17.0 1-FCV-74-9 RHR SUCTION HEADER

TEST VENT 716-AZ301

  1. 4 Accum Rm 0-PI-OPS-17.0 1-TV-74-504 RHR SUCTION HEADER

TEST VENT 716-AZ301

  1. 4 Accum Rm 0-PI-OPS-17.0 1-TV-74-543 LOOP 4 HOT LEG TO RHR SUCTION 716-AZ301
  2. 4 Accum Rm 0-PI-OPS-17.0 1-FCV-74-2 1-FCV-74-2 BYPASS RHR SUCTION 716-AZ304
  3. 4 Accum Rm 0-PI-OPS-17.0 1-FCV-74-8
71. FRONT BACK Which ONE of the following descri bes Dosimeter use at Watt Bar?

In order to check out an electronic dosimeter, the operator mu st scan the barcode on the ____(1)____ of the OSL.

The OSL is returned to Radiation Protection ev ery ____(2)____ months so that it may be read. A. (1) BACK (2) THREE B. (1) BACK (2) SIX C. (1) FRONT (2) THREE D. (1) FRONT (2) SIX CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: WBNP uses a Landauer InLi ght Optically Stimulated Luminescence Dosimeter. This dosimeter has tw o bar codes. The bar code on the front only contains either a four or five digit number. This is the unique number assigned to the employee which Rad Con uses for tracking purposes. Therefore, when a certain employee obtains an electronic dosimeter, he scans this bar c ode such that the computerized dosimetry issuing station recognize s his specific account. The bar code on the back of the OSL is affixed to the actual dosimetry element (i.e. the actual element which records the radiation received). In this way, each employee will have their unique number assigned to one or more protective cases which house the dosimetry elements. The dosimetry elements are tied electroni cally to the specific case by relating the dosimetry serial number to the employees unique number.

As seen the serial number s are not equivalent in va lue. It is plausible to believe that the bar code on t he back would be used because both bar codes are of the same font (7 of 9 bar code) and general appearance. To this extent, a st andard bar code reader is able to process both equally well.

Previous to the use of the OSL, WBNP issued TLDs which were read on a quarterly basis. With the advent of the OSL, WBNP has shifted to a six month read frequency. B. Incorrect: Again it is incorrect that the bar code on the back would be used for dosimetry issue. It is correct that the OSL is returned to Rad Con each six months. C. Incorrect: While it is co rrect that the bar code on the front of the OSL is used for dosimetry issue, it is not correct that the OSL be returned to Rad Con every 3 months. D. Correct: It is correct t hat both the bar code on the fr ont of the OSL be used and that the OSL be returned every six months for processing.

Question Number: 71 Tier: 3 Group:

K/A: 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and l alarms, portable survey instruments, personnel monitoring equipment, etc. Importance Rating: 2.9 3.1 10 CFR Part 55: (CFR: 41.12 / 43.4 / 45.9) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to possess the practical knowledge of a piece of personnel monitoring equipment (Optically Stimulated Luminescence Dosimeter) Technical

Reference:

RCDP-15, Dosimetry Operations Proposed references to be provided: None Learning Objective: 3-OT-RAD0004 , Personnel Dosimetry 1. Describe the operation of Optically Stimulated Luminescent Dosimeters (IER 11-3, Having a solid understanding of plant design, engineering principles, and sciences.) Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: NPG Standard Department Procedure Dosimetry Operations RCDP-15 Rev. 0000 Page 6 of 34 3.2.2 Ordering (continued)

1. The label example is shown below NOTE The fields shown in the example and described below may be modified to meet a specific licensee or department needs. For example, the Emergency Planning Tennessee Emergency Management Agency dosimeters will have a different label scheme than the personnel dosimeters for occupationally exposed workers. A= Customer account # specific to each site B = TVA site C= TVA D = worker first name

E = Worker last name

F = Plant ID G = Bar Code for plant ID H= Wear period I = Slot number

2. Background color is White for worker, area and environmental badges and yellow for controls. 3. Frequency Color (H bar) the color around the worker icon:

Blue for Semi-annual (January - June)

72. Which ONE of the following co mpletes the statement below?

In accordance with 10 CFR 20. 1201, Occupational dose limit s for adults, individual worker's total effective dose equivalent (TEDE) shall be maintained less than __________ REM.

A. 0.5 B. 1.0 C. 3.0 D. 5.0 CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: NPG-SPP-05.1, Radiological Controls details the Nuclear Power Groups administrative dose level progr am. This program lists five levels of administrative dose limits. The first of these is a dose equivalent of up to 0.5 TEDE. This limit does not require a signed authorization. This fact demonstrates the plausibility for th is distractor. This distractor is not correct how ever because the next level of the administrative dose level program also does not require a signed authorization and is up to 1.0 TEDE. B. Incorrect: Correct: the administrative dose level program allows an individual to receive up to 1.0 TEDE before obt aining signed authorization. C. Incorrect: It is not correct that an individual is allowed to receive up to 3.0 TEDE before obtaining signed authorization. It is plausible to believe this because this is the Lens Dose Equivalent (LDE) listed in the second

step of the administrative dose level program. The complete second step of the program is Up to 1.0 TEDE (or 3.0 LDE or 10 SDE) all sources. D. Incorrect: It is not correct that an individual is allowed to receive up to 5.0 TEDE. It is plausible to believe this because, as seen in 10 CFR 20.1201(a)(1)(i), The total effectiv e dose equivalent being equal to 5 REM. Therefore, by law, the occupati onal dose limit for an adult is 5.0 TEDE. However, by administrative pr ocess, the dose limit is less than this. Question Number: 72 Tier: 3 Group:

K/A: G 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. Importance Rating: 3.2 3.7 10 CFR Part 55: (CFR: 41.12 / 43.4 / 45.10) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to possess the knowledge of the radiation exposure limits for normal conditions. Technical

Reference:

10 CFR 20.1201, Occupational dose limits for adults NPG-SPP-05.1, Radiological Controls Proposed references to be provided: None Learning Objective: 3-OT-RAD0003 , RADIATION PROTECTION STANDARDS AND GUIDELINES

4. List TVA radiation dose goals. Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: NPG Standard Programs and Processes Radiological Controls NPG-SPP-05.1 Rev. 0005 Page 14 of 47 3.2.3 Exposure Control (continued) TABLE 1 ADMINISTRATIVE DOSE LEVEL PROGRAM Dose Equivalent (Rem) Requirement Authorization to exceed (signatures) Up to 0.5 TEDE (or 1.5 LDE or 5.0 SDE) at TVA Statement of current year dose and previous years dose signed by individual Not applicable Up to 1.0 TEDE (or 3.0 LDE or 10 SDE) all sources NRC FORM-4 or equivalent to document current year and previous years dose equivalent Not applicable To exceed 1.0 TEDE (or 3.0 LDE or 10 SDE) all sources Same as above RPM/RSO To exceed

5.0 TEDE 3 all sources Form-4 information must be verified and a Planned Special Exposure initiated RPM/RSO, Plant Manager 1 , and Site VP 2 or SED as appropriate. To exceed 1N 4 all sources Form-4 must be verified RPM/RSO, Plant Manager 1 , and Site VP 2 or SED as appropriate. 1 At non-nuclear plant sites, this will be the RSO's immediate supervisor. 2 At non-nuclear plant sites, this will be the applicable TVA VP. 3 Authorizations for a planned special exposure will only be considered in an exceptional situation when alternatives that might avoid the dose estimated to result from the planned special exposure are unavailable or impractical. 4 Total effective dose equivalent should not exceed 1N rem, where N equals the individual's age in years at last birthday, without the authorization signatures delineated. 4. ADLs are based on dosimeter(s) used in determining the reported dose. Results which exceed an administrative level, based on other dosimeter data, do not violate the ADL. NOTE Exceeding an ADL as a result of an incorrect signed documentation of current year or previous years dose does not constitute a violation of the intent of the ADL. 5. When visitor or contract personnel have more restrictive dose limits than TVA - the more restrictive limits will be used. It is the responsibility of the contractor to provide written notification to RP (or RSO) of any company administrative limit. 6. To ensure that ADLs are not exceeded, an administrative control system shall be maintained. 7. Individuals under the age of 18 shall not enter radiologically controlled areas. Home > NRC Library 

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> § 20.1201 Occupational dose limits for adults.

Subpart C--Occupational Dose Limits Source: 56 FR 23396, May 21, 1991, unless otherwise noted. § 20.1201 Occupational dose limits for adults. (a) The licensee shall control the occupational dose to individual adults, except for planned special exposures under § 20.1206, to the following dose limits. (1) An annual limit, which is the more limiting of-- (i) The total effective dose equivalent being equal to 5 rems (0.05 Sv); or (ii) The sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the eye being equal to 50 rems (0.5 Sv). (2) The annual limits to the lens of the eye, to the skin of the whole body, and to the skin of the extremities, which are: (i) A lens dose equivalent of 15 rems (0.15 Sv), and (ii) A shallow-dose equivalent of 50 rem (0.5 Sv) to the skin of the whole body or to the skin of any extremity. (b) Doses received in excess of the annual limits, including doses received during accidents, emergencies, and planned special exposures, must be subtracted from the limits for planned special exposures that the individual may receive during the current year (see § 20.1206(e)(1)) and during the individual's lifetime (see § 20.1206(e)(2)). (c) When the external exposure is determined by measurement with an external personal

monitoring device, the deep-dose equivalent must be used in place of the effective dose equivalent, unless the effective dose equivalent is determined by a dosimetry method approved by the NRC. The assigned deep-dos e equivalent must be for the part of the body receiving the highest exposure. The assigned shallow-dose equivalent must be the dose averaged over the contiguous 10 square centimeters of skin rece iving the highest exposure. The deep-dose equivalent, lens-dose equivalent, and shallow-dose equivalent may be assessed from surveys or other radiation measurements for the purpose of demonstrating compliance with the occupational dose limits, if the individual monitoring device was not in the region of highest potential exposure, or the results of indi vidual monitoring are unavailable. (d) Derived air concentration (DAC) and annual limit on intake (ALI) values are presented in table 1 of appendix B to part 20 and may be used to determine the individual's dose (see § 20.2106) and to demonstrate compliance with the occupational dose limits. (e) In addition to the annual dose limits, the licensee shall limit the soluble uranium intake by an individual to 10 milligrams in a week in consideration of chemical toxicity (see footnote 3 of appendix B to part 20). (f) The licensee shall reduce the dose that an individual may be allowed to receive in the current year by the amount of occupational dose received while employed by any other person (see § 20.2104(e)). [56 FR 23396, May 21, 1991, as amended at 60 FR 20185, Apr. 25, 1995; 63 FR 39482, July 23, 1998; 67 FR 16304, Apr. 5, 2002; 72 FR 68059, Dec. 4, 2007]

73. Which ONE of the following comp letes the statements listed below?

There _____(1)_____ which may be DIRECTLY entered. The Immediate Actions listed in certain EOPs ______(2)_____ without the written instruction being available. A. (1) is ONLY ONE EOP (2) ONLY require that the operator per form the left hand column step B. (1) is ONLY ONE EOP (2) require that the operator per form both the left hand column step AND the right hand column-"response not obtained"-step C. (1) are TWO EOPs (2) ONLY require that the operator per form the left hand column step D. (1) are TWO EOPs (2) require that the operator per form both the left hand column step AND the right hand column-"response not obtained"-step

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: Again it is incorrect and yet plausible that there is only one EOP which may be directly entered. Also, it is incorrect that an immediate action only requires the operator to execute the left hand step from memory. It is plausible to believe such as some of the immediate actions have response not obtained steps which are quire lengthy and difficult to remember (such as the RNO for step 3 of 1-E-0). B. Incorrect: It is incorrect that ther e is only one EOP which may be directly entered. As seen in 0-TI-12.04, implementation of the EOI set is limited to the two cases of: implementing 1-ECA-0.0 and implementing 1-E-0. It is plausible to belie ve that the entry into 1-ECA-0.0 would be made much in the manner that the entry into 1-FR-

S.1, ATWS is (i.e. while the im plementation of 1-FR-S.1 appears direct, the implicit EOI network entry is made via the response not obtained of the immediate action step 1 of 1-E-0). Along such lines, one could reason that the entry into 1-ECA-0.0 while appearing direct was actually made via the implicit transition from the immediate action step 3 of 1-E-0. As seen in 0-TI-12.04, the immediate action steps are "intended to be performed, if necessary, without the written instruction being available." There is no relaxation providing the need to execute only the left hand step and not the right hand "response not obtained" step. C. Incorrect: It is correct that there are two EOPs which are directly entered. Also, it is incorrect and yet plausible t hat only the left hand column of an immediate action step be performed (if needed) without the written instruction.. D. Correct: It is correct that there are two EOPs which are directly entered. Also, it is correct that both columns of an immediate action are required to be performed (if needed) without t he written instruction.

Question Number: 73 Tier: 3 Group:

K/A: G 2.4.1 Knowledge of EOP entry conditions and immediate action steps.

Importance Rating: 4.6 4.8 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to recognize the entry conditions for 1-ECA-0.0 and 1-FR-S.1 while simultaneously recalling the immediate actions of 1-ECA-0.0. . Technical

Reference:

1-ECA-0.0, Loss of Shutdown Power 1-E-0, Reactor Trip or Safety Injection TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions WOG background document for 1-ECA-0.0 Proposed references to be provided: None Learning Objective: 3-OT-TI1204, USER's GUIDE FOR AOIs and EOIs 1. Describe the conditions for entry into the EOP network 11. Identify the three (3) Emergency Instructions which contain "Immediate Actions." Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 & 2 User's Guide for Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 5 of 57

1.0 INTRODUCTION

1.1 Purpose This Instruction provides guidance fo r the use of the Emergency Operating Instructions (EOIs) and Abnormal Operat ing Instructions (AOIs) and ensures standardized implementation of the EOIs. 1.2 Applicability A. These guidelines will be followed by the Main Control Room (MCR) crew during emergency conditions as well as training exercises. B. This Instruction provides the rules of usage for the set of Emergency Operating Instructions including the Status Trees. The EOI set includes Emergency Instructions (E and ES), Emergency Contin gency Actions (ECA), and Function Restoration Instructions (FR). C. The rules explained in Section 2.2.7 (Operator Performance of Action Steps) also apply to Abnormal Operating Instru ctions (AOIs) that are written in the same two column format. Section 2.3 (Types of Action Steps in Two Column Instructions) is descriptive of EOIs and AOIs. 2.0 DETAILS The EOIs direct operator actions to the most urgent safety or operational conditions. This Instruction provides the rules of priority and usage which, when properly applied, will direct the operat or through the EOI network. The rules of usage define the intended use of two column format instru ctions. The rules of priority define the hierarchy of the individual instructi ons and must be followed to ensure proper implementation. 2.1 Implementation of Emergency Operating Instructions Implementation of the EOI set is limited to two specific conditions: A. Any time a complete loss of both tr ains of Shutdown Power occurs the MCR Team enters the EOI network by implementing 1-ECA-0.0, Loss of Shutdown Power. This includes any time during the performance of any other instruction. B. Any time a reactor trip or safety injection occurs, or is required, the MCR Team enters the EOI network by implementing 1-E-0, Reactor Trip or Safety Injection. WBN Unit 1 & 2 User's Guide for Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 14 of 57 2.2.3 Foldout Page A. This page presents actions or transitions which are applicable at any time in the given instruction. Upon transition from an instruction, the current instruction's foldout page becomes applicable and use of the previous instruction's foldout page is discontinued. 1. In the control room, the foldout page information is presented on the back of each page of instructions for which there is a foldout page. 2. The information on the foldout page should be continuously monitored to determine when operator action is necessary. B. Transitions to other instructions allo w immediate response to new symptoms as they appear. 2.2.4 Immediate Action Steps Steps that have been designated as "Immedi ate Actions" are contained in three emergency instructions and selected AOIs. These steps are intended to be performed, if necessary, without the written instruction being available. A. Operators are required to be able to complete the intent of immediate operator action steps. B. During immediate operator action steps the operat ors will ensure automatic actions have occurred or initiate signals as appropriate. Diagnostic or repair actions will be delayed until the immediate actions are complete to allow for evaluation of plant response. C. The immediate actions of 1-E-0 will be addressed as follows: 1. Steps 1 through 4 will be performed in order by the OAC and completion, along with any discrepancies, will be communicated to the Procedure

Reader. 2. The BOP will acknowledge alarms as nec essary to reduce the noise level, and perform backup verification of steps 1 through 4. 3. Procedure Reader will read the immediate action high level step. The OAC will confirm the high level step by verbalizing the low level steps to the procedure reader. 4. When reentering 1-E-0 fr om another EOI, the first four high level steps must be reconfirmed, it is NOT necessary to reperform each low level action. WBN Unit 1 & 2 User's Guide for Abnormal and Emergency Operating Instructions 0-TI-12.04 Rev. 0000 Page 15 of 57 2.2.4 Immediate Action Steps (continued) D. The immediate actions for 1-FR-S.1, Nuclear Power Generation/ATWS, and 1-ECA-0.0, Loss of Shutdown Power, are consistent wit h the initial actions of 1-E-0, Reactor Trip or Safety Inje ction, and will be performed in the same manner. E. The immediate actions contained in Abnormal Operating Instructions will be performed in order and completion, along with discrepancies, will be communicated to t he Procedure Reader. 1. 1-AOI-2, Malfunction of Reactor Control System 2. 1-AOI-4, Nuclear Instrumentation Malfunctions

3. AOI-18, Malfunction of Pressu rizer Pressure Control System 2.2.5 Sequence of Step Performance The EOI steps have been written in a specified sequence. This sequence is designed to ensure timely operator action for the most limiting evolution.

A. The operator is expected to perform a ll high level action steps in the specified sequence. 1. High level steps and substeps for wh ich action may be delayed until plant conditions allow performance are identified by the step wording. 2. The Procedure Reader is responsible for keeping track of steps for which performance must be delayed. 3. When one operator is involved in co mpleting a step, the Procedure Reader may direct the other operator to perform a subsequent step provided the subsequent step does NOT depend upon completion of the previous step. WBN Unit 1 Reactor Trip or Safety Injection 1-E-0 Rev. 0005 Step Action/Expected Response Response Not Obtained Page 4 of 45

3. CHECK 6.9 kV shutdown boards: a. At least one board energized from:
  • CSST (offsite), OR
  • D/G (blackout) with ERCW flow to running DG(s) [0-M-27A]:
  • 1-HS-67-66A OR 1-HS-67-68A
  • 2-HS-67-66A OR 2-HS-67-68A
  • 1-HS-67-67A OR 1-HS-67-65A
  • 2-HS-67-67A OR 2-HS-67-65A RESTORE power to at least one train of shutdown boards:
1) EMERGENCY START All D/G's [1-M-1]. 2) IF D/G did NOT start from 1-M-1, THEN EMERGENCY START All D/G's

[0-M-26] 3) IF both trains of shutdown boards remain de-energized, THEN PLACE 6.9kV SD Bd transfer switch in MAN [1-M-1], AND CLOSE supply breaker from energized source.

4) ENSURE ERCW to running DG(s)

[0-M-27A]:

  • 1-HS-67-66A OR 1-HS-67-68A
  • 2-HS-67-66A OR 2-HS-67-68A
  • 1-HS-67-67A OR 1-HS-67-65A
  • 2-HS-67-67A OR 2-HS-67-65A
5) IF ERCW flow CANNOT be aligned, THEN EMERGENCY STOP affected DG(s)

IF power can NOT be restored to at least one train of shutdown boards, THEN ** GO TO 1-ECA-0.0, Loss of Shutdown Power.

74. Given the following conditions: - Unit 1 is at 100% power. - A contract worker uses a plant phone and the x3911 line to report that a temporary pump located outsi de of the Unit 2 TDAFWP room has caught fire. - Heavy smoke is present at the scene of the fire. Which ONE of the following descr ibes the appropriate response in accordance with 1-AOI-30.1, Plant Fires? The RO actuating the fire alarm will ____(1)____.

The control room staff will ____(2)____. NOTE: 0-SOI-13.01, Fire Detection System A. (1) INITIATE Fire Alarm and then MANUALLY RESET Fire Alarm (2) initiate an ABI B. (1) INITIATE Fire Alarm and then MANUALLY RESET Fire Alarm (2) secure ventilation in accordance with 0-SOI-13.01 C. (1) ONLY INITIATE Fire Alarm, the alarm will AUTOMATICALLY RESET (2) initiate an ABI D. (1) ONLY INITIATE Fire Alarm, the alarm will AUTOMATICALLY RESET (2) secure ventilation in accordance with 0-SOI-13.01

CORRECT ANSWER: B DISTRACTOR ANALYSIS: A. Incorrect: It is correct that in accordance with 1-AOI-30.1, t hat the operator must INITIATE Fire Alarm and then RESET Fire Alarm. This is seen in steps B. and C. of Appendix B of the AOI. The reason for this is that the fire alarm is actuated and secured by one red pushbutton which sits on the reactor operator's desk in the MCR. If one does not push the button a second time, the fire alarm will not automatically secure itself. It is Incorrect that the main control room will initiate an ABI (in order to secure the auxiliary building ventilation). Step 5 of 1-AOI-30.1 directs: REFER TO 0-SOI-13.01 Fire Detection System, for ventilation systems required and USE Appendix A of this Instruction, as

necessary, for further guidance. Appendix A addresses fire dampers which fail to automatically close. 0-SOI-13.01, Fire Detection System contains Appendix B, Operator Actions. Appendix B of the SOI directs which ventilation fans are to be secur ed for a fire in a given location. For general areas in the auxiliary build ing, one may see that either the Unit 1, Unit 2 or both Units G eneral Supply and Exhaust Fans are secured. It is plausible to believe that would actuate an ABI because doing so would secure the Auxiliary building ventilation in a predictable manner. The WARNING which appears above the Operator Actions for the Auxiliary Building in A ppendix B was included because of a corrective action which resulted fr om the use of the SOI. The operating crew using the Appendix realized that the alterations to the Aux Bldg ventilation being directed were not (at the time) existent in a procedure and would cause a gross imbalance in Aux Bldg pressure-to the extent that damage to dam pers and doors could occur. The crew asked the fire protection engi neer (who had input into the SOI) about how the securing of ventilation could be done and the crew asked about the feasibil ity of using an ABI. The decision was made to include the warning discussed. B. Correct: Again, it is correct that one must actuat e and subsequently reset the fire alarm. Also, it is correct that the crew w ould secure ventilation in accordance with 0-SOI-13.01. C. Incorrect: It is not correct that a manual re set of the fire alarm is not required. It is plausible to believe this because the accountability alarm does have an automatic reset. The accountability alarm is located in the vicinity of the Fire Detection panel, 0-M-29 and will cease several minutes after being actuated. It is also Incorrect and yet plausible that an ABI would be used to secure Aux Bldg ventilation. D. Incorrect: It is not correct and yet plausib le that a manual rese t of the fire alarm is not required. It is correct that the crew woul d secure ventilation in accordance with 0-SOI-13.01.

Question Number: 74 Tier: 3 Group:

K/A: 2.4 Emergency Procedures / Plan 2.4.27 Knowledge of fire in the plant procedures.

Importance Rating: 3.4 3.9 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.13) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to possess the knowledge of how the fire in the plant procedure 1-AOI-30.1 addresses the actuation of the fire alarm as well as how 1-AOI-30.1 secures ventilation for a space affected by smoke. Technical

Reference:

0-SOI-13.01, Fire Detection System 1-AOI-30.1, Plant Fires Proposed references to be provided: None Learning Objective: 3-OT-AOI3000, AOI-30.1 & 30.2, "PLANT FIRES" 4. Given a set of plant conditions, DESCRIBE operator actions required in response to the following per 1-AOI-30, Plant Fires. Cognitive Level: Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: WBN Unit 1 Plant Fires 1-AOI-30.1 Rev. 0002 Page 10 of 11 Appendix B (Page 1 of 1) Operator Actions Guideline A. OBTAIN the following: NAME of person reporting fire (if verbal): ________________________ LOCATION of fire: ________________ __________________________ TYPE of fire: ________________ ________________________ ______ SEVERITY of fire: __________________ ________________ ________ DATE & TIME: _____________________________ ________________ B. INITIATE Fire Alarm. C. RESET Fire Alarm. D. NOTIFY Fire Operations Unit by radio. E. ANNOUNCE fire location over PA. F. REPEAT fire announcement. G. INITIATE fire alarm. H. RESET fire alarm when necessary. TIME of reset: ________________ I. ENSURE fire pumps running as necessary: HPFP: 1A-A 2-A-A 1B-B 2B-B DGFP: J. ANNOUNCE location of Incident Comm and Post over PA and radio. K. IF a confirmed fire is located in ANY of the following:

1. Control Building 2. Auxiliary Building 3. Reactor Building 4. Annulus THEN NOTIFY ALL AUOs to report with gear and Standby for AOI-30.2.

WBN Unit 1 Plant Fires 1-AOI-30.1 Rev. 0002 3.0 OPERATOR ACTIONS (continued) Page 5 of 11 Step Action/Expected Response Response Not Obtained

5. REFER TO SOI-13.01, Fire Detection System, for ventilation systems required and USE Ap pendix A of this Instruction, as necessary, for further guidance.
75. Given the following conditions:
- Unit 1 is at 100% power. - Supplemental Condenser Circulating Water System (SCCW) has been placed in service in accordance with SOI-27.03, section 5.2.

Subsequently: - COOLING TOWER BASIN LEVE L HI (164-D) is LIT. Which ONE of the following descri bes the appropriate alarm response? The operator responding to this annunciator will validate the ala rm by comparing the

alarm setpoint to the actual reading obtained ____(1)____. In accordance with ARI-159-165, Sumps & CCW, the operator responding to this annunciator will ADJUST 0-FCV-112, SUPPLEMENTAL CCW SYSTEM SUPPLY FLOW CONTROL by ____(2)____. NOTE: SOI-27.03, Supplemental Conden ser Circulating Water System; Section 5.2, Placing SCCW In Service A. (1) on the ICS (2) operating the valve from 1-M-15 B. (1) on the ICS (2) dispatching an AUO to operate the valve from a local control C. (1) locally at the cooling tower basin (2) operating the valve from 1-M-15 D. (1) locally at the cooling tower basin (2) dispatching an AUO to operate the valve from a local control

CORRECT ANSWER: D DISTRACTOR ANALYSIS: A. Incorrect: It is not correct that the operat or would refer to IC S to obtain cooling tower basin level. Watts Bar is a closed loop condenser circulating water system. The CCW pumps take suction from the cooling tower basin, provide cooling water to the condenser and then the return water to the condenser is discharged back to the cooling tower. To compensate for evaporative losses as well as cooling tower blowdown, three forms of CCW makeup are prov ided. These are the raw cooling water bypass strainers, essential raw cooling water discharge and the supplemental condenser circulating water system (SCCW). The SCCW system was a modification to the original plant design and takes water originally designed to provide the Watts Bar Coal Plant (taken from over twenty feet below the surface of Watts Bar Lake) and discharges such into the Unit 2 cooling tower basin. This may be seen on print 1-47W831-1. The Unit 2 c ooling tower basin is connected to the Unit 1 cooling tower and the cold injection water is fed into the suction of the Unit 1 CCW pumps. T he hot water in the Unit 1 basin is discharged to the original discharge point of the Watts Bar Coal plant. Coupled with the normal operatio n of CCW, this feed and bleed provides a very effective cooling. If the supply valve and the discharge valve for SCCW are not balanced pr operly, then the Unit 1 basin can either drain or overfill. Because balancing a 90" butterfly valve and a 78" butterfly valve is not an easy ta sk, this evolution can taken well over a shift (and require the next shi ft to respond to cooling tower level alarms). Both the SOI for placing S CCW in service as well as the ARI for addressing the cooling tower high le vel alarm direct that the SCCW inlet be throttled (0-FCV-27-112). As this was a design modification, controls for these valves are not on the water service panel of the MCR (1-M-15). Additionally, cooling tower basin level is not a

parameter monitored on the ICS. This may be seen on 1-47W610 2. It is plausible to believe that the ICS would monitor cooling tower basin level because disastrous results come from either a high or low level. A high level will cause an overfill which is an environmental reportable event. A low level will cause the CCW pumps to lose suction and trip; causing a turbine and reactor trip. Parameters such as holding pond level are monitored on ICS and as such it would be reasonable to monitor the basin leve

l. One of the original makeup methods for the cooling tower basin level is RCW makeup via the

bypass strainer. This makeup passes through the control valve 1-FCV-24-191. The handswitch for this valve is on 1-M-15 (seen on print 1-47W611-24-1) and as such it is plausible to believe that controls for SCCW would be provided on 1-M-15 especially given the consequences that a loss of basin level control could have. B. Incorrect: Again, it is not correct and yet plausible t hat ICS would be used to obtain cooling tower basin level. Al so, it is correct that 0-FCV-27-112 would be operated locally. C. Incorrect: It is correct that basin le vel need be obtained locally. Also, it is Incorrect and yet plausible that SCCW would be adjusted locally. D. Correct: It is correct that basin level need be obtained locally. Also, it is correct that SCCW would be adjusted locally.

Question Number: 75 Tier: 3 Group:

K/A: 2.4 Emergency Procedures / Plan 2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

Importance Rating: 4.2 4.0 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.3) 10CFR55.43.b: Not applicable K/A Match: K/A is matched because the applicant is required to demonstrate the ability to validate the alarm setpoint in an ARI by correctly identifying the source of the data. Next the applicant must demonstrate the ability to operate the controls identified in the ARI. Technical

Reference:

1-47W831-1; 1-47W611-24-1; 1-47W610-27-2 ARI-159-165, Sumps and CCW SOI-27.03, Supplemental Condenser Circulating Water System Proposed references to be provided: None Learning Objective: 3-OT-SYS027A, CONDENSER CIRCULATING WATER 4. EXPLAIN the physical connections and/or cause-effect relationships between the Condenser Circulating Water System and the following systems:

a. Supplemental CCW System (SCCW) Cognitive Level:

Higher Lower X Question Source: New X Modified Bank Bank Question History: New question for the 2015-301 NRC RO Exam Comments: 4r Ell.O'IOO'I N f'Ra.I ---COOL fNC lO'll'ERS WBN Unit 1 Sumps & CCW ARI-159-165 Rev. 0038 Page 37 of 49 A. CT blowdown terminated via closure of all the following valves:

  • 0-FCV-27-97, HOLDING POND ISOL
  • 0-FCV-27-100, 54" DIFFUSER A SUP
  • 0-FCV-27-101, 48" DIFFUSER B SUP B. Misalignment of SCCW supply or discharge valves C. SCCW evolution in progress D. Excessive debris on Trash rack 1-RACK-27-1 and/or 1-RACK-27-3 NOTE If both HI level and LO level (window 164-E) are lit, serious blockages on trash racks could exist.

[1] REFER TO SOI-24.01 for operation of 1-FCV-24-191, RCW BYPASS FLOW CONTROL VALVE. [2] CHECK position of 1-FCV-27-97, -100, and -101. [3] ENSURE SCCW supply and discharge valves BOTH open OR BOTH closed:

  • 0-FCV-27-112, SUPPLEMENTAL CCW SYSTEM SUPPLY FLOW CONTROL.
  • 0-FCV-27-111, SUPPLEMENTAL CCW SYSTEM DISCHARGE FLOW CONTROL.

[4] CHECK Cooling Tower Trash Racks and CONTACT maintenance to clean trash racks if required. [5] ADJUST 0-FCV-27-112, SUPPLEMENTAL CCW SYSTEM SUPPLY FLOW CONTROL, as necessary to control basin level:

  • REFER TO SOI-27.03, Supplemental Condenser Circulating Water System.

[6] ESTABLISH CT blowdown per SOI-27.01, Condenser Circulating Water System.

1-47W610-27-2 SOI-27.01 SOI-27.03

Source Setpoint 1-LS-27-93B el 730' 6" (6 inches below top of basin and 6 inches above normal level)

164-D COOLING TOWER BASIN LEVEL HI (Page 1 of 1) Probable Cause: Corrective Action:

References:

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