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{{#Wiki_filter:Click Here to expand Bookmarks Each Question has an expandable subset which in explained on the following page of this document.
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All needed supporting references are included and marked up with the question.
 
Question PDF Bookmark Anatomy A.0 - Question A.1 - Distractor Analysis A.2 - Misc Comments/Feedback/History A.3 - NUREG ES4015 A.4 - Original Question (if question was MODIFIED from a BANK question)
A.5 - Supplied Reference B.0 - Regulator Documents B.0 - Tech Specs B.1 - Tech Spec Bases B.2 - TRM B.3 - COLR B.4 - PTLR B.5 - ODCM B.6 - PLS B.7 - PTDB B.8  FSAR C.0 - Procedures C.1 - EOP C.2 - AOP C.3 - SOP C.4 - UOP C.5 - ARP C.6 - Admin C.7  Surveillance C.9 - Other D.0 - Drawings D.1 - P&IDs D.2 - Oneline D.3 - Elementary D.4 - Logic D.5  Other E.0 - Misc Other E.1 - Photographs E.2 - Maps E.3 - Rad Surveys E.4 - Lesson Plan E.5  OE
: 1. 001K2.03 001/LOIT/RO/M/F 2.7/3.1/001K2.03/LO-TA-27008///
Given the following:
          - Unit 1 is at 100% reactor power.
Which one of the following completes the following statement?
The Rod Control Logic Cabinet is energized from __(1)__ or __(2)__.
__(1)__                                    __(2)__
A.        Vital 125 VDC Panel                          Vital 120 VAC Panel (1AD11)                                    (1AY1A)
B.          Vital 125 VDC Panel                Regulated 120 VAC Instrument Panel (1AD11)                                      (1NYS)
C.            1A MG Set via a                          Vital 120 VAC Panel 120 VAC transformer                                (1AY1A)
D.            1A MG Set via a                Regulated 120 VAC Instrument Panel 120 VAC transformer                                  (1NYS)
K/A 001              Control Rod Drive System K2.03            Knowledge of bus power supplies to the following:
                        - One-line diagram of power supplies to logic circuits.
K/A MATCH ANALYSIS The question tests the candidate's knowledge of the two power supplies to the Rod Control Logic and Power cabinets.
EXPLANATION OF REQUIRED KNOWLEDGE The candidate must recall where the power source originates for the power supplies in the Rod Control Logic and Power cabinets. The source is the same for both cabinets, however the power supply output voltages are different from the cabinets. The primary source comes from the output of the MG sets. A single phase of the 260 VAC MG Set output is transformed and regulated to 120 VAC, which splits and feeds both cabinets.
The second source comes from the 120 VAC Regulated Instrument Bus NYS, breaker 27, which also splits and feeds both cabinets. Reference one-line drawing 1X3D-AA-G05A and lesson plan simplified one-line V-LO-PP-27101, slide 32.
ANSWER / DISTRACTOR ANALYSIS Wednesday, February 26, 2014 8:26:15 AM                                                    1
 
A. Incorrect. Plausible. The first part is incorrect. The Logic and Power cabinets are powered from the MG Sets and 1NYS-27 only. However, AD11 is a common 125 VDC feed to many safety-related components and bistables. AD11 is the control power supply to the 'A' RTBs. Additionally, the Rod Control MG Set Controllers are powered from non-1E 125 VDC. A candidate without specific knowledge of the rod control system could find it reasonable for the cabinets to be powered from the same source as the MG Set controllers or the reactor trip breakers.
The second part is incorrect. The Logic and Power cabinets are powered from the MG Sets and 1NYS-27 only. However, AY1A is the primary power feed to the SSPS logic bays. A candidate without specific knowledge of the rod control system could find it reasonable for the cabinets to also be powered from the 120 VAC Vital panel which is fed from AD1, especially if the candidate knows that rod control utilizes both DC and AC power.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. The Logic and Power cabinets are powered from the MG Sets and 1NYS-27.
C. Incorrect. Plausible. The first part is correct. The Logic and Power cabinets are powered from the MG Sets and 1NYS-27.
The second part is incorrect. The Logic and Power cabinets are powered from the MG Sets and 1NYS-27 only. However, AY1A is the primary power feed to the SSPS logic bays. A candidate without specific knowledge of the rod control system could find it reasonable for the cabinets to also be powered from this common control related power supply.
D. Correct.                  Both parts are correct. The control and logic cabinets are powered from the MG set 260 VAC output via a 260/120 VAC transformer.
Wednesday, February 26, 2014 8:26:15 AM                                                            2
 
Level:                          RO Tier # / Group #                T2 / G2 K/A#                            001K2.03 Importance Rating:              2.7 / 3.1 Technical
 
==Reference:==
V-LO-PP-27101 Rev 2.0 1X3D-AA-G05A Rev 49.0 References provided:            None Learning Objective:              LO-PP-27101-02 State the power supplies for the Rod Control System.
LO-TA-27008      Draw and label a one-line diagram of the Control Rod Drive Power Supply Question origin:                BANK - Catawba 2007 Question #56 Cognitive Level:                M/F 10 CFR Part 55 Content:          41.7 Comments:                        This K/A was a replacement for 75K2.03 from the original sample plan.
You have completed the test!
Wednesday, February 26, 2014 8:26:15 AM                                                          3
 
(ND31-09)                    (ND31-04)
CONTROL POWER - SSPS                    CONTROL POWER - SSPS CHARGING SPRING - AD11-09              CHARGING SPRING - BD11-09 NB09-10                                                                  RTA      RTB MG #1 260vac                        BYA      BYB                        DC HOLD NB08-10                                                                                                      CABINET 1CB (ND32-01)          MG #2 (ND32-06)                                                                          S1 150vac/120vac 125vdc 70vdc SURGE PROTECTOR fuses FU1a&b fuses FU3a&b 100vdc    +16.5 vdc -16.5 vdc battery PS1              +28vdc PS1
                                                                                            -24vdc Auto/manual                        PS3 fuse F7A reset LOGIC CABINET fuse F5A CARD FRAME 100vdc    +16.5 vdc -16.5 vdc fuse F6A                                              OFF LATCH battery PS2                                                        HOLD fuse F8A        -24vdc PS4
                                                                        +28vdc PS2 SURGE PROTECTOR                              Auto/manual                                                      POWER reset                        fuses FU4a&b NYS 27 fuses FU2a&b CABINET LIFT COIL MOVABLE                      STATIONARY GRIPPER                      GRIPPER V-LO-PP-27101 Rev-2.0                                                                                                              32
: 1. 003A3.04 001/LOIT AND LOCT/RO/C/A 3.6/3.6/003A3.04/LO-PP-28103-03///
Initial condition:
          - Unit 1 is at 8% reactor power.
Current condition:
          - 1NAA de-energizes due to a fault on the bus.
Which one of the following completes the following statement?
When conditions stabilize, __(1)__ flow on 1FI-414, RCS Flow Loop 1 will be observed, and the reactor __(2)__ automatically trip due to the event.
__(1)__                                  __(2)__
A.                  ~113%                                      will B.                  ~113%                                    will NOT C.                    ~6%                                      will D.                    ~6%                                    will NOT K/A 003              Reactor Coolant Pump A3.04            Ability to monitor automatic operation of the RCPS, including:
RCS flow.
K/A MATCH ANALYSIS The question requires the candidate to monitor the status of the RCPs following undervoltage on a 13.8 kV bus and determine which RCS loops will still have flow.
EXPLANATION OF REQUIRED KNOWLEDGE Above P-7 (10% NI), RCPs will trip on an underfrequency signal. The actuation of the undervoltage signal blocks the underfrequency relays to prevent all RCPs from tripping when voltage is lost on a single bus. Therefore, only two RCPs will be affected.
RCPs 1 and 3 are supplied by 13.8KV bus NAA, RCPs 2 and 4 are supplied by NAB.
Above P-7 and below P-8 (48% NI), loss of flow in two RCS loops will result in an automatic reactor trip. At 48% and above, a loss of flow in a single loop would result in Wednesday, February 26, 2014 8:28:02 AM                                                        1
 
an automatic reactor trip.
RCPs normally indicate approximately 106% flow while in operation at NOPT. Once RCPs trip and reverse flow stabilizes, running RCPs indicate approximately 113% due to decreased backpressure and increased flow in the loops. The tripped RCPs indicate approximately 6% flow due to backflow in the loop.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. 1NAA feeds RCPs 1 & 3 and 6% flow would be observed. However, RCP power supplies are easily confused. If the candidate gets the two buses confused, RCPs 2&4 would be assumed to remain in operation and this choice would be correct.
The second part is incorrect. NI's are indicating less than the P-7 setpoint so the "at power trips" are not enabled and a loss of two RCPs will not result in an automatic reactor trip.
However, if reactor power was greater than or equal to 10%, the reactor would trip.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. NI's are indicating less than the P-7 setpoint so the "at power trips" are not enabled and a loss of two RCPs will not result in an automatic reactor trip.
C. Incorrect. Plausible. The first part is correct. Bus 1NAA powers RCPs 1&3. The tripped RCPs indicate approximately 6% flow due to backflow in the loops.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 8:28:02 AM                                                              2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            003A3.04 Importance Rating:              3.6 / 3.6 Technical
 
==Reference:==
13503A-1 Rev 7.2 page 33 1X3D-AA-C01A Rev 23.0 1X6AA02-00229 Rev 1.0 References provided:            None Learning Objective:              LO-PP-16401 Describe the following for the RCP supply breakers:
: a. Breaker arrangement
: b. Power supply for each pump
: c. Protection features LO-PP-28103 List all reactor trip set points, coincidences, permissives, and blocks.
Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 / 45.5 Comments:                        None You have completed the test!
Wednesday, February 26, 2014 8:28:02 AM                                                        3
 
Approved By                                                                                                  Procedure    Version J.B. Stanley                        Vogtle Electric Generating Plant                                        13503A-1 7.2 Effective Date                TRAIN A REACTOR CONTROL SOLID-STATE PROTECTION                                Page Number 6/21/13                                                      SYSTEM                                                  33 of 38 ATTACHMENT C                                        Sheet 1 of 6 PERMISSIVES, CONTROL INTERLOCKS, REACTOR TRIPS AND ESF ACTUATIONS PERMISSIVES Permissive                Setpoint/Coincidence              Function P-4                      Train related Rx trip &          Trips Main Turbine Bypass breaker open                        Train A - mechanical Train B - electrical FWI if Lo Tavg (2/4  564 F) present Seals in FWI if caused by SI or P-14 (Hi Hi Level)
Must be present to block auto SI after SI reset.
P-4 Train A arms Steam Dumps P-4 Train B swaps Steam Dumps to plant trip controller P-6                      1/2 IR Detectors  2.0 E -5      Allows manual block of SR High  trip
                                % Rx Power P-7                      P-10 (2/4 PR NIs  10%            Unblocks "At Power" Trips Rx power) or                              Przr Low Pressure P-13 (PT-505 or 506                      Przr High Level 10% turbine power)                        RCS Two Loop Low Flow RCP UF RCP UV P-8                      2/4 PR NIS  48% Rx power        Enables Single Loop Low Flow Rx Trip P-9                      2/4 PR NIS  40% Rx power        Enables Turbine trip Rx trip P-10                      2/4 PR NIS  10% Rx power        Auto block of SR High  trip Enables P-7 Allows manual block of IR rod stop and Hi  trip Allows manual block of PR Hi  trip Lo Setpoint P-11                      2/3 Przr Pressure channels        Auto enables Lo Przr Press SI & Lo Steamline Press 2000 psig                      SI/SLI & sends signal to open Accum Isolation Valves when P-11 resets. P-11 allows operator to block PRZR & Steamline low pressure SI & SLI. Also activates "Not Full Open" annunciators for Accumulator MOVs (ALB16; A5, B5, C5 &
D5) and HV-8806; (ALB16 E03).
P-12                      2/4 NR Tavg  550 F              Interlocks Steam Dumps closed ( Cooldown Dump Valves PV-507A, B & C) may be reopened by use of Bypass Interlock switches)
P-13                      1/2 Turbine Impulse channels    Enables P-7 10%
P-14                      2/4 NR SG Level channels  82%    Actuates FWI Actuates MFP and Main Turbine trip Printed September 6, 2013 at 15:12
: 1. 004K6.09 001/LOIT/RO/C/A - 2.8/3.1/004K6.09/LO-PP-09100-03///
Initial conditions:
          - Unit 1 is at 100% reactor power.
          - VCT level is 50%.
          - VCT makeup control is in AUTO.
Current condition:
          - VCT level transmitter, 1LT-112, fails HIGH.
Assuming no operator action, which one of the following completes the following statements?
VCT Divert Valve, 1LV-112A, is aligned to the __(1)__.
If actual VCT level lowers to 29%, automatic makeup __(2)__ occur.
__(1)__                                  __(2)__
A.                    PRT                                      will B.                    PRT                                    will NOT C.                    RHUT                                      will D.                    RHUT                                    will NOT K/A 004              Chemical Volume Control System (CVCS):
K6.09            Knowledge of the effect of a loss or malfunction on the following CVCS components:
                        - Purpose of VCT divert valve.
K/A MATCH ANALYSIS The question addresses the effect of a failure of VCT Level Transmitter LT-112A (which results in the VCT divert valve aligning to the HUT), and the effect the transmitter failure will have on automatic VCT makeup.
EXPLANATION OF REQUIRED KNOWLEDGE Letdown Divert valve LV-112A is controlled by both LT-185 and LT-112. Per ARP 17007-1 ALB07-E05, LV-112A will divert to the HUT if in AUTO and VCT level is >97%
Wednesday, February 26, 2014 9:08:58 AM                                                          1
 
on LT-112, or modulate to the HUT if LT-185 is greater than the setpoint on PIC-185 (normally 87%).
Per ARP 17007-1, annunciator ALB07-E05 automatic makeup starts when LT-112 indicates <30% and stops when LT-112 is >50%. With LT-112 failed failed high, automakeup will not start regardless of actual VCT level as sensed by LT-185.
Furthermore, the automatic swap-over to the RWST suctions will not occur since only one transmitter will indicate less than 5.7%. CCPs will eventually cavitate without manual operator action.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per 17007-1, LT-112 >97% level will result in LV-112A tripping to the HUT position. However, all relief and divert valves in the letdown system are directed to either the PRT or the HUT based on location. A candidate with insufficient knowledge of the VCT level control circuit may assume that flow would be diverted to the PRT since this is the more common location within letdown.
The second part is incorrect. LT-0112 transmitter has failed high. Per ARP 17007-1, LT-112 will never indicate <30% and therefore auto makeup will not start. However, a candidate with insufficient knowledge of the VCT auto makeup control circuit may believe VCT makeup is controlled by LT-185, or both LT-185 and LT-112, and assume VCT makeup will not be affected.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. LT-0112 transmitter has failed high.
Per ARP 17007-1, LT-112 will never indicate <30% and therefore auto makeup will not start.
C. Incorrect. Plausible. The first part is correct. Per 17007-1, LT-112 >97% level will result in LV-112A tripping open to the HUT position.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first choice of part C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 9:08:58 AM                                                              2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            004K6.09 Importance Rating:              2.8 / 3.1 Technical
 
==Reference:==
17007-1/2 Rev 29.1 1X4DB115 Rev 34.0 LO-PP-09100 CVCS Letdown (pages 113 and 129 in particular)
References provided:            None Learning Objective:              LO-PP-09100-03 State the purpose and describe the control signals, setpoints, and any interlocks for the following:
: b. VCT divert valve, LV-112A LO-PP-09300-05 Describe the automatic control functions for the Reactor Makeup Control System components during:
: a. automatic makeup to the VCT Question origin:                BANK Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 / 45.7 Comments:
You have completed the test!
Wednesday, February 26, 2014 9:08:58 AM                                                        3
 
Approved By                                                                              Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                      17007-1      29.1 Effective  Date            ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 07 ON PANEL          Page Number 07/25/2012                                        1A2 ON MCB                                  41 of 51 WINDOW E05 ORIGIN                          SETPOINT VCT 1-LT-0112                        92% HI                    HI/LO LEVEL 1-LT-0185                        20% LO 1.0              PROBABLE CAUSE
: 1.        High level:
: a. Volume Control Tank (VCT)/Hold-up Tank (HUT) Divert Valve 1-LV-0112A malfunction OR aligned to the VCT position,
: b. VCT makeup greater than charging flow.
: 2.        Low level:
: a. Makeup Control NOT in automatic,
: b. VCT/HUT Divert Valve 1-LV-0112A malfunction OR aligned to the Recycle Holdup Tank position,
: c. System leak.
Printed September 9, 2013 at 14:15
 
Approved By                                                                                Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                        17007-1      29.1 Effective  Date            ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 07 ON PANEL            Page Number 07/25/2012                                          1A2 ON MCB                                    42 of 51 WINDOW E05 (Continued) 2.0              AUTOMATIC ACTIONS NOTE VCT automatic makeup should have started at 30 percent or stopped at 50 percent.
: 1.        Letdown flow diverts to the HUT WHEN 1-HS-0112A is in AUTO with VCT high level of 97 percent.
: 2.        Charging Pump suction auto swaps to the Refueling Water Storage Tank (RWST) upon a Lo-Lo VCT level of 5.7 percent.
: 3.        A summary of instrument setpoints associated with the VCT levels include:
LI-0112                          VCT LEVEL    LI-0185 Trip open 112A                97%            Modulate 112A full divert (if LIC-0185 pot @8.70)
Hi level alarm                92%
112A Trip Open signal          87%            112A starts to divert (if LIC-0185 pot Resets                                        @8.70)
Auto Makeup stops              50%
Auto Makeup starts            30%
Low level alarm                20%            Low level alarm RWST auto swapover            5.7%(2 of 2)  RWST auto swapover Printed September 9, 2013 at 14:15
 
Approved By                                                                                Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                        17007-1      29.1 Effective  Date            ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 07 ON PANEL            Page Number 07/25/2012                                            1A2 ON MCB                                  43 of 51 3.0              INITIAL OPERATOR ACTIONS
: 1.        Check VCT level using 1-LI-0185 on the QMCB AND compare to 1-LI-0112 on the IPC OR on Trend Recorder XR-40053.
: 2.        IF equipment failure is indicated by EITHER LT-0185 OR LT-0112 failed high, perform the following:
: a. Place 1HS-112A to the VCT position.
NOTE Pump cavitation may be indicated by fluctuating discharge pressure and/or erratic flow.
: b. Monitor charging pump(s) for signs of cavitation. IF cavitation is observed:
(1)    Isolate letdown, (2)    Stop any running charging pumps, (3)    Initiate 18007-C Section B.
: c. Initiate Manual VCT Makeup per 13009-C.
: d. Contact maintenance to initiate repairs.
: 3.        IF level is low AND makeup is lost, initiate 18007-C, "Chemical And Volume Control System Malfunction."
: 4.        IF level is low due to system leakage, initiate 18004-C, Reactor Coolant System Leakage.
Printed September 9, 2013 at 14:15
 
Approved By                                                                                Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                        17007-1      29.1 Effective  Date            ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 07 ON PANEL            Page Number 07/25/2012                                          1A2 ON MCB                                  44 of 51 WINDOW E05 (Continued) 4.0              SUBSEQUENT OPERATOR ACTIONS
: 1.        IF VCT level is high:
: a.      Stop Makeup,
: b.      Divert letdown flow to the Recycle Holdup Tank (HUT position) using 1-HS-0112A on the QMCB,
: c.      Operate makeup per 13009-1, "CVCS Reactor Makeup Control System."
: 2.        IF equipment failure is indicated, initiate maintenance as required.
: 3.        IF an operating charging pump fails due to suspected gas binding (fluctuating discharge pressure AND flow), THEN the standby pump SHALL NOT be started UNTIL the cause of the gas binding is understood AND all effected piping and components have been vented.
5.0              COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB115, 1X4DB116-1, 1X3D-BD-C02E, 1X6AU01-184, PLS, 1X5DT0012 Printed September 9, 2013 at 14:15
: 1. 005A2.02 001/LOIT AND LOCT/RO/C/A 3.5/3.7/005A2.02/LO-TA-60007///
Initial conditions:
          -  Unit 1 is in Mode 5 with solid plant conditions.
          -  RHR Train 'B' is in service aligned to low pressure letdown.
          -  RHR Train 'A' is in standby.
          -  CCP 'A' is in service.
Current condition:
          - Instrument air header depressurizes due to an air line break.
Which one of the following completes the following statements?
With no operator action, RCS pressure will __(1)__.
Per 18028-C, "Loss of Instrument Air," to mitigate the pressure transient the crew will stop the __(2)__ pump.
__(1)__                                  __(2)__
A.                  increase                                    RHR B.                  increase                                charging C.                decrease                                    RHR D.                decrease                                  charging K/A 005              Residual Heat Removal A2.02            Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Pressure transient protection during cold shutdown.
K/A MATCH ANALYSIS (a) The question requires the candidate to predict the pressure transient caused by a malfunction of the RHR system from an instrument air loss during cold shutdown conditions with the primary "water solid".
(b) The question then requires the candidate to utilize Abnormal Operating Procedure (AOP) guidance to control/mitigate the resulting pressure transient.
Wednesday, February 26, 2014 9:14:14 AM                                                        1
 
EXPLANATION OF REQUIRED KNOWLEDGE The candidate is required to recall the failure modes of RHR letdown valves, RHR outlet and bypass valves, and Charging flow control valves.
The candidate is required to integrate failure modes against plant response, including contradictory responses from RHR and Charging.
The candidate is required to recall AOP response for loss of instrument air in lower modes with the plant in a solid condition.
Narrative of Response:
A loss of instrument air will cause RHR low pressure letdown flow control valve HV-128 to fail closed, resulting in a loss of letdown flow. Simultaneously, charging flow control valve FV-121 will fail open resulting in maximum charging flow (>150 gpm). In a solid plant condition, pressure will increase rapidly until the RHR suction reliefs lift at 450 psig. Per AOP 18028-C guidance on steps B3 and B4, all charging pumps will be stopped.
Additionally, the RHR outlet valve will fail open and the bypass valve will fail closed, resulting in maximum RHR flow with maximum cooling. In a solid plant condition, the pressure increase from the charging flow with no letdown has a far greater magnitude than the pressure reduction from the increased cooling (1 degree F is approximately 100 psig). In accordance with AOP18028-C guidance on steps B7 thru B9, the one running RHR pump will NOT be stopped, and instead an operator will be dispatched to the pump discharge valve to limit the cooldown.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. A loss of instrument air will result in RHR low pressure letdown flow control valve HV-128 failing closed, resulting in a loss of letdown flow. Simultaneously, charging flow control valve FV-121 will fail open resulting in maximum charging flow (>150 gpm). In a solid plant condition, pressure will increase rapidly until the RHR suction reliefs lift at 450 psig to control pressure The second part is incorrect. Step B9 will not stop the RHR pump and instead will throttle the discharge flow. The distractor is plausible since the candidate may not consider the loss of letdown and increase in charging flow, and only consider the RHR flow and cooling increase. As such, 18028 step B9.b gives direction to stop an RHR pump.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. Per AOP 18028-C guidance on steps B3 and B4, if the RCS is solid, all charging pumps will be stopped.
C. Incorrect. Plausible. The first part is incorrect. The pressure increase caused by a Wednesday, February 26, 2014 9:14:15 AM                                                                2
 
loss of letdown and max rate charging is significantly greater than the pressure reduction caused by an increase in RHR cooling. However, if the candidate does not consider the loss of letdown and max charging and only considers the increase in RHR flow and cooling, then RCS pressure would be expected to decrease due to density changes from the cooldown.
The second part is incorrect. AOP 18028-C does not give direction to stop the last RHR pump. 18028-C step B9.b. gives direction to stop AN RHR pump. To a novice operator, this direction may seem to be consistent with a pressure decrease caused by excessive cooling.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 9:14:15 AM                                                            3
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            005A2.02 Importance Rating:                3.5 / 3.7 Technical
 
==Reference:==
AOP 18028-C Rev 26.2 P&ID 1X4DB114 Rev 41.0 P&ID 1X4DB116-1 Rev 50.0 P&ID 1X4DB122 Rev 51.0 References provided:            None Learning Objective:              LO-LP-60321-08 Describe the effects on RCS pressure due to a loss of instrument air while solid on RHR.
LO-LP-60321-05 Describe why the RHR pump discharge should not be fully closed while throttling RHR flow to maintain RCS temperature during a loss of instrument air when in modes 4, 5, or 6.
LO-LP-60321-11 Given the entire AOP, describe:
: a. Purpose of selected steps
: b. How and why the step is being performed
: c. Expected response of the plant/parameter(s) for the step LO-TA-60007        Respond to a Loss of Instrument Air per 18028-C Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.5 / 43.5 / 45.3 / 45.13 Comments:
You have completed the test!
Wednesday, February 26, 2014 9:14:15 AM                                                              4
 
Approved By                                                                          Procedure Number Rev JB Stanley                            Vogtle Electric Generating Plant              18028-C          26.2 Date Approved                                                                        Page Number LOSS OF INSTRUMENT AIR 09/23/09                                                                                  22 of 31 ATTACHMENT B                        Sheet 5 of 7 LOSS OF INSTRUMENT AIR IN MODES 4, 5, OR 6 ACTION/EXPECTED RESPONSE                            RESPONSE NOT OBTAINED B1
__B1.      Check Instrument Air supply header          __B1. Go to Step B12.
pressure on PI-9361 - LESS THAN 100 PSIG.
B2
__B2.      IF a temporary air compressor is              B2.
connected to the Service Air Header.
Perform Attachment C while continuing with Attachment B.
CAUTION Loss of instrument air will cause CHARGING LINE CONTROL FV-0121 and SEAL FLOW CONTROL HV-0182 to fail open.
B3
__B3.      Check RCS inventory - SOLID.                  B3. Perform the following:
B3.a
__a. IF needed to maintain RCS level, THEN establish safety grade charging by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM.
B3.b
__b. Go to Step B6.
B4
__B4.      Trip all charging pumps.                      B4.
B5
__*B5.      Monitor No. 1 seal leakoff                    B5.
temperature and flow until charging pump is restarted.
 
S Printed September 9, 2013 at 16:16                  7
 
Approved By                                                                          Procedure Number Rev JB Stanley                            Vogtle Electric Generating Plant              18028-C          26.2 Date Approved                                                                          Page Number LOSS OF INSTRUMENT AIR 09/23/09                                                                                    23 of 31 ATTACHMENT B                          Sheet 6 of 7 LOSS OF INSTRUMENT AIR IN MODES 4, 5, OR 6 ACTION/EXPECTED RESPONSE                            RESPONSE NOT OBTAINED
 
CAUTION Loss of instrument air pressure will cause the RHR HX outlet valves to fail full open and the HX bypass valves to fail fully closed.
B6
__B6.      Check plant Mode - MODE 4 OR                __B6. Suspend all fuel movement.
MODE 5.
B7
__B7.      Check RCS temperatures -                        B7. Perform the following:
LOWERING.
__a. Control temperature using          B7.a ARVs.
B7.b
__b. IF ARVs NOT available, THEN stop all but one RCP.
B8
__*B8.      Check RCS cooldown rate -                      *B8. Perform the following:
GREATER THAN 100&deg;F/HR.
__a. Monitor cooldown rate.              B8.a B8.b
__b. IF cooldown rate can NOT be maintained less than 100&deg;F/hr, THEN perform Step B9.
B8.c
__c. Go to Step B10.
 
S Printed September 9, 2013 at 16:16                  7
 
Approved By                                                                              Procedure Number Rev JB Stanley                            Vogtle Electric Generating Plant                  18028-C          26.2 Date Approved                                                                            Page Number LOSS OF INSTRUMENT AIR 09/23/09                                                                                      24 of 31 ATTACHMENT B                          Sheet 7 of 7 LOSS OF INSTRUMENT AIR IN MODES 4, 5, OR 6 ACTION/EXPECTED RESPONSE                                  RESPONSE NOT OBTAINED
 
NOTE A key will be necessary to unlock 1205-U6-019 (Train A) or 1205-U6-020 (Train B).
B9
      *B9. Control RCS cooldown:                                B9.
B9.a
__a. Check RHR pump status - TWO                            __a. Go to Step B9.c.
PUMPS RUNNING.
B9.b
__b. Stop one RHR pump.                                      b.
B9.c
__c. Dispatch an operator to establish                        c.
communications at the in service RHR heat exchanger inlet valve and local flow indicator:
UNIT 1              UNIT 2 1-1204-U6-019        2-1204-U6-019 TRAIN A        1-FIS-0610          2-FIS-0610 (AB-C122)            (AB-C38) 1-1204-U6-020        2-1204-U6-020 TRAIN B        1-FIS-0611          2-FIS-0611 (AB-C92)            (AB-C27)
 
Step 9 continued on next page Printed September 9, 2013 at 16:16                    7
 
Approved By                                                                          Procedure Number Rev JB Stanley                            Vogtle Electric Generating Plant                18028-C          26.2 Date Approved                                                                          Page Number LOSS OF INSTRUMENT AIR 09/23/09                                                                                    25 of 31 ATTACHMENT B                          Sheet 8 of 7 LOSS OF INSTRUMENT AIR IN MODES 4, 5, OR 6 ACTION/EXPECTED RESPONSE                            RESPONSE NOT OBTAINED B9.d
: d. Unlock and throttle                                d.
1205-U6-019 (Train A) or 1205-U6-020 (Train B) to maintain:
RHR flow rate - GREATER THAN 750 GPM.
RCS cooldown rate - LESS THAN 100&deg;F/HR.
CCW temperature at RHR HXs - LESS THAN 195&deg;F.
[ALB61-A01 NSCW CCW ACCW TRAIN A TEMP ALARM extinguished.]
[ALB61-A02 NSCW CCW ACCW TRAIN B TEMP ALARM extinguished].
B10
__*B10. Check Instrument Air header                      __*B10. Dispatch an operator to close pressure - REMAINS GREATER                          Turbine Building Instrument Air THAN 70 PSIG.                                        isolation valve:
UNIT 1:      1-2420-U4-512 (TB-1-TE12)
UNIT 2:      2-2420-U4-512 (TB-1-TE10)
B11
__B11. Check main turbine turning gear -                  __B11. Engage turning gear if necessary ENGAGED.                                            by initiating 13800, MAIN TURBINE OPERATION.
 
S Printed September 9, 2013 at 16:16                  7
 
LV459 &460 fail CLOSED to isolate letdown flow
 
FV-0121 fails OPEN for max flow
 
HV606 fails OPEN causing max cooling HV618 fails CLOSED causing max flow
: 1. 005AK3.05 001/LOIT/RO/M/F 3.4/4.2/005AK3.05/LO-TA-60037///
Initial conditions:
          - Unit 1 is at 95% reactor power with a power ascension in progress.
          - CBD is at 200 steps.
Current conditions:
          -  ALB10-D06 ROD DEV is received.
          -  CBD rod M12 DRPI indicates 109 steps.
          -  18003-C, "Rod Control System Malfunction," is entered.
          -  Reactor power has been lowered per 18003-C guidance.
Which one of the following completes the following statement?
Per 18003-C, control rod M12 is considered __(1)__,
and the reason for the power reduction is to minimize __(2)__ heat rates and power distribution variances.
__(1)__                              __(2)__
A.                  dropped                                core B.                  dropped                                local C.                misaligned                              core D.                misaligned                              local K/A 005              Inoperable/Stuck Control Rod AK3.05          Knowledge of the reasons for the following responses as they apply to the Inoperable / Stuck Control Rod:
                        - Power limits on rod misalignment.
K/A MATCH ANALYSIS The question tests the candidate's knowledge of the reason power is reduced with a misaligned/dropped rod. The question also tests the candidate's knowledge of when a rod is considered dropped versus misaligned.
EXPLANATION OF REQUIRED KNOWLEDGE Wednesday, February 26, 2014 9:15:41 AM                                                  1
 
Per AOP 18003-C Section A NOTE prior to step 1 and the section entry conditions, a rod misaligned greater than 110 steps is considered dropped.
Per AOP 18003-C Section A NOTE prior to step 11, power reduction as soon as practical after the rod drop occurs minimizes local fuel power distribution variances and the chances of fuel damage.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. A rod that is misaligned by more than 110 steps is considered dropped per 18003-C. Rod M12 is only misaligned 91 steps. However, rod M12 is only at a position of 109 steps. It is reasonable to assume a candidate who does not fully understand the NOTE in 18003-C may conclude that since the rod is <110 steps inserted into the core, the rod is considered dropped. Therefore, this distractor is plausible.
The second part is incorrect. When a rod is significantly misaligned, the power peaking that could result in fuel damage is a local area issue as opposed to an entire core issue.
However, depending on location, a dropped or misaligned rod can significantly alter the core flux shape and drive AFD and QPTR outside limits. The flux issues seen over the core tend to be slower moving changes and are bounded by core design and Tech Spec limits. As such, these changes do not challenge fuel integrity. The localized flux changes are more significant and do challenge fuel integrity. A candidate with insufficient knowledge of core behavior on a dropped or misaligned rod could find it reasonable that core flux changes could damage fuel because Tech Spec limits could be approached.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. When a rod is misaligned, the peaking of power that could result in fuel damage is a local area issue as opposed to the entire core issue. This is described in the Note prior to step 11.
C. Incorrect. Plausible. The first part is correct. Rod M12 is only misaligned by 91 steps from its bank. Per the NOTE in 18003-C, the rod is considered "misaligned'.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 9:15:41 AM                                                              2
 
Level:                          RO Tier # / Group #                T1 / G2 K/A#                            005AK3.05 Importance Rating:              3.4 / 4.2 Technical
 
==Reference:==
AOP 18003-C Rev 26.4 References provided:            None Learning Objective:              LO-LP-60303-01 Describe how the following Rod Control System malfunctions could result in xenon oscillations in the core:
: a. dropped rod
: b. uncontrolled continuous rod motion
: c. misaligned rod LO-LP-60303-14 Describe how the retrieval of a misaligned rod can affect the power distribution limits of the core. Include a effects on why Reactor Engineering must be consulted if the rod has been misaligned for longer than one hour.
LO-LP-60303-16 Describe how the radial flux profile may be affected by a misaligned rod.
LO-TA-60037      Respond to a Misaligned Control Rod per 18003-C Question origin:                NEW Cognitive Level:                M/F 10 CFR Part 55 Content:          41.5, 41.10, 45.6, 45.13 Comments:
You have completed the test!
Wednesday, February 26, 2014 9:15:41 AM                                                              3
 
Approved By                                                                          Procedure    Version J. B. Stanley                        Vogtle Electric Generating Plant              18003-C        26.4 Effective Date                                                                        Page Number ROD CONTROL SYSTEM MALFUNCTION 6/14/13                                                                                    6 of 30 A. DROPPED RODS IN MODE 1 ACTION/EXPECTED RESPONSE                            RESPONSE NOT OBTAINED A10 A10. Check Annunciator ALB10-F02,                  A10. On the affected PR NI Drawer, reset Power Range Hi Neutron Flx                        the Positive Rate Trip as follows:
Rate Alert is clear.
TSLB-4 NI Hi rate bistables not                                                    A10.a illuminated.                                a. Turn the RATE MODE Switch momentarily to RESET.
A10.b
: b. Verify POSITIVE RATE TRIP Drawer Light is NOT lit.
A10.c
: c. Check PR HI RATE bistable not illuminated on TSLB-4 8.1 - PR HI Q RATE NC 41U 8.2 - PR HI Q RATE NC 42U 8.3 - PR HI Q RATE NC 43U 8.4 - PR HI Q RATE NC 44U NOTE Power reduction as soon as practical after the rod drop occurs minimizes local fuel power distribution variances and the chances of fuel damage. Although TS 3.1.4 requires power be reduced to less that 75% within 2 hours, a target of achieving reactor power level less than 75% in one hour meets the as soon as practical objective.
A11 A11. Reduce Thermal Power to less than              A11.
75% within 1 hour from time of Rod drop using 12004-C POWER OPERATION.
 
S Printed November 13, 2013 at 09:18
 
Approved By                                                                        Procedure    Version J. B. Stanley                        Vogtle Electric Generating Plant              18003-C        26.4 Effective Date                                                                      Page Number 6/14/13 ROD CONTROL SYSTEM MALFUNCTION                            1 of 30 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure provides instructions for malfunctions of the Rod Control System resulting in uncontrolled rod motion, dropped or misaligned rods.
SYMPTOMS SECTION A, DROPPED RODS IN MODE 1 ALB10-E5 ROD AT BOTTOM ALB10-F2 POWER RANGE HI NEUTRON FLX RATE ALERT ALB10-C2 POWER RANGE CHANNEL DEVIATION Rod bottom LED on digital rod position indication.
Rod misaligned greater than 110 steps from demand position Tavg dropping.
SECTION B, UNCONTROLLED CONTINUOUS ROD MOTION IN ALL MODES Rod motion with invalid demand from the Automatic Rod Control System.
Failure of rods to stop moving when the Rod Motion Switch is released.
SECTION C, MISALIGNED RODS IN MODE 1 ALB10-C2 POWER RANGE CHANNEL DEVIATION ALB10-D2 POWER RANGE UP DET HI FLX DEV ALB10-E2 POWER RANGE LWR DET HI FLX DEV Failure of ALB10-C4 ROD BANK LO LIMIT or ALB10-D4 ROD BANK LO-LO LIMIT to reset during rod withdrawal.
Rod misaligned greater than 12 steps and less than or equal to 110 steps from demand position.
Quadrant power tilt ratio calculation exceeds 1.02.
Printed November 13, 2013 at 09:14
 
Approved By                                                                        Procedure    Version J. B. Stanley                        Vogtle Electric Generating Plant            18003-C        26.4 Effective Date                                                                      Page Number ROD CONTROL SYSTEM MALFUNCTION 6/14/13                                                                                  4 of 30 A. DROPPED RODS IN MODE 1 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED
 
NOTE A Rod misaligned greater than 110 steps should be considered dropped and this section performed.
A1 A1. Stop any turbine loading changes.          A1.
A2 A2. Check the following:                        A2. Perform the following:
A2.a DRPI - AVAILABLE.                          a. Trip the Reactor and Go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.
0 Only one Rod dropped by observing DRPI.
A3 A3. Check rod misaligned greater than          A3. Go to Section C MISALIGNED 110 steps.                                      RODS IN MODE 1.
A4 A4. Initiate TS 3.1.4.                          A4.
A5 A5. Initiate The Continuous Actions Page.      A5.
A6
        *A6. Maintain Tavg at program by                A6.
performing the following as appropriate:
Adjust Turbine load.
Dilute or borate.
Use manual Rod control.
 
S Printed November 13, 2013 at 09:14
 
Approved By                                                                          Procedure    Version J. B. Stanley                        Vogtle Electric Generating Plant              18003-C      26.4 Effective Date                                                                        Page Number ROD CONTROL SYSTEM MALFUNCTION 6/14/13                                                                                    16 of 30 C. MISALIGNED RODS IN MODE 1 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED C1 C1. Stop any turbine loading changes.            C1.
C2 C2. Check only one Rod - MISALIGNED.            C2. Trip the Reactor and Go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.
C3 C3. Check rod misaligned less than or            C3. Go to Section A, DROPPED RODS equal to 110 steps.                              IN MODE 1.
C4 C4. Check misaligned Rod -                      C4. Go to 13502, CONTROL ROD MISALIGNED BY GREATER THAN                        DRIVE AND POSITION INDICATION 12 STEPS.                                        SYSTEM to address restoration of the misaligned rod.
C5 C5. Initiate TS 3.1.4.                          C5.
NOTE Power reduction as soon as practical after the rod drop occurs minimizes local fuel power distribution variances and the chances of fuel damage. Although TS 3.1.4 requires power be reduced to less that 75% within 2 hours, a target of achieving reactor power level less than 75% in one hour meets the as soon as practical objective.
C6 C6. Reduce Thermal Power to less than            C6.
75% within 1 hour from time of discovery of Rod misalignment using 12004-C, POWER OPERATION (MODE 1).
 
S Printed November 13, 2013 at 09:14
: 1. 005K5.09 001/LOIT/RO/M/F 3.2/3.4/005K5.09/LO-LP-39213-01///
Initial conditions:
            - Unit 1 is in Mode 6 for refueling.
            - RHR pump 'B' is in service.
            - RHR pump 'A' is tagged out.
Current conditions:
            - RHR pump 'B' is stopped to place a fuel assembly in the vicinity of the RCS hot leg nozzle.
            - Chemistry requests permission to make a chemical addition to the RCS.
Which one of the following completes the following statement?
Per Tech Spec 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," the Chemistry technician __(1)__ allowed to make the chemical addition to the RCS, and RHR pump 'B' can remain stopped for a MAXIMUM of __(2)__ hour(s) in an 8 hour period.
__(1)__                                __(2)__
A.                      is                                    1 B.                      is                                    4 C.                  is NOT                                    1 D.                  is NOT                                    4 K/A 005              Residual Heat Removal K5.09            Knowledge of the operational implications of the following concepts as they apply the RHRS:
                        - Dilution and boration considerations.
K/A MATCH ANALYSIS The RHR system is in service for shutdown cooling in Mode 6, and requires the candidate to evaluate proposed plant activities with regards to the Tech Spec Thursday, March 06, 2014 11:14:33 AM                                                          1
 
requirements. One of the items evaluated is the operational implication of a chemical addition (dilution activity) in the current plant configuration.
EXPLANATION OF REQUIRED KNOWLEDGE The candidate is required to recall the Limiting Condition of Operation for Residual Heat Removal (RHR) and Coolant Circulation - High Water Level and apply it to current plant configuration.
Only one train of RHR is required to be OPERABLE and in operation per Tech Spec 3.9.5. Additionally, per the note in the LCO, the train of RHR can be removed from operation for up to 1 hour per 8 hour period provided the RCS boron concentration is not reduced. The bases for this note specifically discusses stopping RHR to allow placement of fuel assemblies near the loops.
Chemical additions are made via a chemical addition pot, which is aligned to the suction of the charging pumps. Demin water from the RMWST is aligned to the pot and is used to mix and flush the chemicals. Therefore, any addition of chemicals to the RCS will result in a dilution.
The Tech Spec knowledge requirement is above the line and therefore is applicable to RO knowledge level. Additionally, ROs are expected to consider impacts to plant equipment and Tech Specs prior to altering system alignments. The specific Bases knowledge mentioned above is not required to answer the question, it is included in the explanation to justify the selection of the current conditions framed by the question for realism.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. A chemical addition utilizes RMWST water; a dilution (regardless of how insignificant) is being performed. Therefore, it would not be allowed. However, the RCS volume in this configuration is substantial. The chemical addition would produce a negligible change in boron concentration. A candidate not familiar with the specifics of the Tech Spec note could find it reasonable to make such a negligible change. Additionally, Tech Spec 3.4.8 allows chemical additions in Mode 5 provided Shutdown Margin is verified and HFASA is OPERABLE..
The second part is correct. Per 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level NOTE, the RHR Pump may be stopped for <1 hour in an 8 hour period B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level the RHR Pump may be stopped for <1 hour in an 8 hour period.
However, 4 hours is a completion time associated with TS 3.9.5 condition A.
Thursday, March 06, 2014 11:14:33 AM                                                                2
 
C. Correct.                  The first part is correct. A chemical addition utilizes RMWST water; a dilution (regardless of how insignificant) is being performed. Therefore, it would not be allowed. Per LCO 3.9.5 NOTE, no activities are allowed that would reduce the RCS boron concentration.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            005K5.09 Importance Rating:              3.2 / 3.4 Technical
 
==Reference:==
Tech Spec 3.9.5 Tech Spec Bases 3.9.5 P&ID 1X4DB116-1 Rev 50.0 References provided:            None Learning Objective:            LO-LP-39213-01 For any given item in section 3.9 of Tech Specs, be able to:
: a. State the LCO.
: b. State any one hour or less required actions.
LO-LP-39213-02 Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode:
: a. Whether any Tech Spec LCOs of section 3.9 are exceeded.
: b. The required actions for all section 3.9 LCOs.
LO-LP-39213-04 Describe the bases for any given Tech Spec in section 3.9.
Question origin:                NEW Cognitive Level:                M/F 10 CFR Part 55 Content:        41.5 / 45.7 Comments:
You have completed the test!
Thursday, March 06, 2014 11:14:33 AM                                                                  3
 
RHR and Coolant Circulation  High Water Level 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation  High Water Level LCO 3.9.5            One RHR loop shall be OPERABLE and in operation.
                    -------------------------------------------NOTE-------------------------------------------
The required RHR loop may be removed from operation for  1 hour per 8 hour period, provided no operations are permitted that would cause a reduction of the Reactor Coolant System boron concentration.
                    -----------------------------------------------------------------------------------------------
APPLICABILITY:      MODE 6 with the water level  23 ft above the top of reactor vessel flange.
ACTIONS CONDITION                              REQUIRED ACTION                          COMPLETION TIME A. RHR loop requirements              A.1          Suspend operations                  Immediately not met.                                        involving a reduction in reactor coolant boron concentration.
AND A.2          Suspend loading                      Immediately irradiated fuel assemblies in the core.
AND A.3          Initiate action to satisfy          Immediately RHR loop requirements.
AND (continued)
Vogtle Units 1 and 2                          3.9.5-1                            Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
 
RHR and Coolant Circulation  High Water Level 3.9.5 ACTIONS CONDITION                        REQUIRED ACTION              COMPLETION TIME A.  (continued)                    A.4        Close all containment    4 hours penetrations providing direct access from containment atmosphere to outside atmosphere.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.9.5.1        Verify one RHR loop is in operation and circulating  In accordance with reactor coolant at a flow rate of  3000 gpm.        the Surveillance Frequency Control Program Vogtle Units 1 and 2                      3.9.5-2                  Amendment No. 158 (Unit 1)
Amendment No. 140 (Unit 2)
 
RHR and Coolant Circulation  High Water Level B 3.9.5 BASES APPLICABLE          RHR and Coolant Circulation - High Water Level satisfies Criterion 4 SAFETY ANALYSES      of 10 CFR 50.36 (c)(2)(ii).
(continued)
LCO                  Only one RHR loop is required for decay heat removal in MODE 6, with the water level t 23 ft above the top of the reactor vessel flange.
Only one RHR loop is required to be OPERABLE, because the volume of water above the reactor vessel flange provides backup decay heat removal capability. At least one RHR loop must be OPERABLE and in operation to provide:
: a. Removal of decay heat;
: b. Mixing of borated coolant to minimize the possibility of criticality; and
: c. Indication of reactor coolant temperature.
An OPERABLE RHR loop includes an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.
The LCO is modified by a Note that allows the required operating RHR loop to be removed from service for up to 1 hour per 8 hour period provided no operations are permitted that would cause a reduction of the RCS boron concentration. Boron concentration reduction is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to RHR isolation valve testing.
During this 1 hour period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.
APPLICABILITY        One RHR loop must be OPERABLE and in operation in MODE 6, with the water level t 23 ft above the top of the reactor vessel flange, to provide decay heat removal and mixing of the borated coolant. The 23 ft water level was selected (continued)
Vogtle Units 1 and 2                    B 3.9.5-2                                Rev.1-10/01
: 1. 006A1.06 001/LOIT AND LOCT/RO/C/A 3.6/3.9/006A1.06/LO-TA-37008///
Initial conditions:
          -  Unit 1 experienced a small break LOCA.
          -  19012-C, "Post-LOCA Cooldown and Depressurization," is in progress.
          -  Both RHR pumps are stopped.
          -  RCS temperature and pressure are stable.
Current condition:
          - CCP 'A' is stopped per 19012-C.
Which one of the following completes the following statements?
When CCP 'A' is stopped, RCS subcooling margin will initially __(1)__.
Subsequently, if the OATC observes subcooling margin is 22&deg;F and lowering after stopping SIP 'B', then the OATC is required to __(2)__ per 19012-C guidance.
__(1)__                                  __(2)__
A.                  increase                              re-start SIP 'B' B.                  increase                              re-actuate SI C.                  decrease                              re-start SIP 'B' D.                  decrease                              re-actuate SI K/A 006              Emergency Core Cooling A1.06            Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including:
                        - Subcooling Margin K/A MATCH ANALYSIS -
The question requires the candidate to predict the subcooling trend in response to stopping the first CCP during 19012-C, "Post-LOCA Cooldown and Depressurization".
The candidate must determine the required actions when subcooling margin lowers below procedural limits.
EXPLANATION OF REQUIRED KNOWLEDGE 19012-C, "Post-LOCA Cooldown and Depressurization", establishes a 100 F/hour Wednesday, February 26, 2014 9:28:14 AM                                                    1
 
cooldown and then secures ECCS pumps based on subcooling requirements. The question places the candidate in 19012-C after the completion of Step 25, when the CCP is stopped. The candidate must predict the effect of stopping the CCP (and subsequent pressure reduction) on subcooling.
There is a note just prior to Step 25 that reminds the operating crew to allow RCS pressure to stabilize or rise after an ECCS pump is secured before stopping another.
The candidate is then asked to determine the correct action to mitigate the plant response. With subcooling <24F and lowering, ECCS flow must be increased. Step 36 gives the guidance to operate ECCS pumps as necessary and utilize Attachment C if needed to re-establish CCP Cold Leg Injection. The procedure then loops you back through to allow the cooldown to restore subcooling to a point that the ECCS pump can then be secured.
This response is contrasted to the guidance in other EOPs (ex. 19001-C) which direct the candidate to actuate SI if subcooling is <24F or Przr Lvl cannot be maintained >9%.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. RCS pressure will initially lower as the CCP is stopped, and subcooling will lower. However, a 100F/Hr cooldown rate is in progress priot to stopping the CCP, increasing subcooling as RCS temperature and pressure lower.
The second part is correct. Per EOP 19012-C step 36 guidance, if subcooling continued to decrease to <24 F after plant condtions stabilized when the CCP was stopped, the RNO directs opperating ECCS pumps and re-establishing CCP Cold Leg Injection as necessary. Since SI 'B' was just stopped, it would therefore be re-started.
B. Incorrect. Plausible The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per step 36 and foldout page guidance, ECCS pumps are operated and Attachment C utilized to re-align to cold leg injection as necessary. The distractor is plausible since other EOPs direct the operator to actuate SI based on subcooling of <24F. Additionally, a candidate unfamiliar with the guidance of 19012-C could find it reasonable to re-actuate SI since the only actions that have occurred to this point were to stop one CCP and align through normal charging.
Re-actuating SI would realign both of these without negatively affecting the remaining ECCS equipment.
C. Correct.                  The first part is correct. When the CCP is stopped, RCS pressure is expected to lower as injection flow is reduced. The reduction in RCS pressure results in a corresponding reduction in RCS subcooling. This is reflected in the NOTE prior to step 25 of EOP 19012-C.
The second part is correct. See the second part of choice A above.
Wednesday, February 26, 2014 9:28:14 AM                                                              2
 
D. Incorrect. Plausible The first part is correct. See the first part of choice C above.
The second part is in correct. See the second part of choice B above.
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            006A1.06 Importance Rating:              3.6 / 3.9 Technical
 
==Reference:==
19012-C Rev 33.3 References provided:            None Learning Objective:              LO-LP-37112-01 Using EOP 19012 as a guide, briefly describe how each step is accomplished.
LO-LP-37112-04 Given a NOTE or CAUTION statement from the EOP, state the bases for that NOTE or CAUTION statement.
LO-TA-37008      Perform Post-LOCA Cooldown and Depressurization of the RCS per 19012-C Question origin:                MODIFIED - LOIT Bank question LO-LP-37112-01-6 Cognitive Level:                C/A 10 CFR Part 55 Content:          41.5 / 45.5 Comments:
You have completed the test!
Wednesday, February 26, 2014 9:28:14 AM                                                          3
 
Name: ________________________________                                    Export Temp Test Form: 0 Version: 0
: 1. LO-LP-37112-01 006/LOLP37112/LO-TA-13008/000EK3.06/////
A small break LOCA has occurred on Unit 1. The operating crew has transitioned to 19012-C, "ES-1.2 POST-LOCA COOLDOWN AND DEPRESSURIZATION", and are currently at the step which determines if a CCP may be stopped. The minimum subcooling requirements are met and the SS directs the OATC to stop CCP 1A.
Which ONE of the following parameters CORRECTLY describes the RCS response when Train 'A' CCP is stopped?
A. RCS pressure increases.
B. RCS pressure decreases.
C. RCS break flow increases.
D. Pressurizer level increases.
Thursday, September 12, 2013 3:42:28 PM                                                    1
 
Approved By                                                                              Procedure Versi on J. B. Stanley                                Vogtle Electric Generating Plant            19012-C 33.3 Effective Date                                                                            Page Number ES - 1.2 POST-LOCA COOLDOWN AND 05/01/2013                                          DEPRESSURIZATION                          11 of 43 ACTION/EXPECTED RESPONSE                                    RESPONSE NOT OBTAINED 12
        *12. Initiate RCS cooldown to cold                    12.
shutdown:
12.a
: a. Monitor shutdown margin by                          a.
initiating 14005, SHUTDOWN MARGIN AND KEFF CALCULATIONS.
Cooldown continuously                            12.b
: b. Maintain cooldown rate in RCS                      b.
increases subcooling, cold legs - LESS THAN                which allows ECCS 100&deg;F/HR.                            pumps to be stopped sequentially.                                    12.c
: c. Use RHR system if in service.                      c.
12.d
: d. Dump steam to Condenser from                        d. Dump steam from intact intact SG(s) using Steam Dumps:                      SG(s) using SG ARV(s).
12.d.1
: 1)    Place PIC-507 in Manual.                      1) 12.d.2
: 2)    Match demand on SG                            2)
Header Pressure Controller PIC-507 and SD demand meter UI-500.
12.d.3
: 3)    Transfer Steam Dumps to                        3)
STM PRESS mode.
12.d.4
: 4)    Open available Steam                          4)
Dumps by slowly raising demand on PIC-507.
13
: 13. Check RCS subcooling - GREATER                    13. Go to Step 36.
THAN 24F [38F ADVERSE].
 
S Printed September 17, 2013 at 12:41
 
Approved By                                                                                Procedure Versi on J. B. Stanley                                  Vogtle Electric Generating Plant            19012-C 33.3 Effective Date                                                                              Page Number ES - 1.2 POST-LOCA COOLDOWN AND 05/01/2013                                          DEPRESSURIZATION                          17 of 43 ACTION/EXPECTED RESPONSE                                    RESPONSE NOT OBTAINED 24.c
: c. PRZR Level - GREATER THAN                            c. Return to Step 15.
19% [50% ADVERSE].
24.d
: d. Start an RCP using ATTACHMENT A. (RCP 4 or RCP 1 preferred).
24.e
: e. Close PRZR Spray Valve(s) for                        d.
stopped RCP(s):
RCP 1: PIC-0455C RCP 4: PIC-0455B Reminder on pressure response from stopping            NOTES ECCS pumps.
After stopping any ECCS Pump, RCS pressure should be allowed to stabilize or rise before stopping another ECCS pump.
The CCPs and SI Pumps should be stopped on alternate ECCS trains when possible.
25
: 25. Check if one CCP should be stopped:                  25.
25.a
: a. Two CCPs - RUNNING.                                  a. Go to Step 26.
25.b
: b. Determine required RCS                                b.
subcooling from table:
Subcooling Criteria (F)
SI Pump Status              With Any RCP Running      With No RCP Running NORMAL ADVERSE NORMAL ADVERSE None Running                    82          99          95          108 One Running                      44          61          53          67 Two Running                      40          57          49          63
 
Step 25 continued on next page Printed September 17, 2013 at 12:41
 
Approved By                                                                              Procedure Versi on J. B. Stanley                              Vogtle Electric Generating Plant              19012-C 33.3 Effective Date                                                                            Page Number ES - 1.2 POST-LOCA COOLDOWN AND 05/01/2013                                      DEPRESSURIZATION                            18 of 43 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 25.c
: c. RCS Subcooling - GREATER                          c. IF RCS WR Hot Leg THAN REQUIRED                                    temperature greater than SUBCOOLING.                                      350&deg;F [340&deg;F ADVERSE],
THEN go to Step 36.
IF RCS WR Hot Leg temperature less than 350&deg;F
[340&deg;F ADVERSE],
THEN perform the following:
25.c.1
: 1)  Verify at least one RHR Pump running, 25.c.2
: 2)  IF RHR Pump running, THEN go to Step 25.d.
IF RHR Pumps can NOT be operated, THEN go to Step 36.
25.d
: d. PRZR Level - GREATER THAN                        d. Return to Step 15.
19% [50% ADVERSE]. Just after this step is where the question                                            25.e
: e. Stop one CCP.              takes place.          e.
26
: 26. Check if one SI Pump should be                26.
stopped:
26.a
: a. Any SI Pump - RUNNING.                            a. Go to Step 27.
 
Step 26 continued on next page Printed September 17, 2013 at 12:41
 
Approved By                                                                                  Procedure Versi on J. B. Stanley                                  Vogtle Electric Generating Plant            19012-C 33.3 Effective Date                                                                                Page Number ES - 1.2 POST-LOCA COOLDOWN AND 05/01/2013                                            DEPRESSURIZATION                            25 of 43 ACTION/EXPECTED RESPONSE                                    RESPONSE NOT OBTAINED 34.c
: c. Depressurize RCS until EITHER                          c.
of the following conditions is satisfied:
PRZR level - GREATER THAN 75% [52% ADVERSE].
                                      -OR-RCS subcooling - LESS THAN 34&deg;F [48&deg;F ADVERSE].
35
: 35. Verify adequate shutdown margin for                35.
xenon free cold shutdown by initiating 14005, SHUTDOWN MARGIN AND KEFF CALCULATIONS.
36
: 36. Check ECCS flow not required:                      36.
No                                              36.a
: a. RCS Subcooling - GREATER                              a. Perform the following:
THAN 24F [38F ADVERSE].
Direction to re-established                            Operate ECCS Pumps ECCS flow based on                                      as necessary.
subcooling <24F.
Initiate ATTACHMENT C as necessary to re-establish CCP Cold Leg Injection.
Go to Step 37.
 
Step 36 continued on next page Printed September 17, 2013 at 12:41
 
Approved By                                                                        Procedure Versi on J. B. Stanley                            Vogtle Electric Generating Plant          19012-C 33.3 Effective Date                                                                      Page Number ES - 1.2 POST-LOCA COOLDOWN AND 05/01/2013                                    DEPRESSURIZATION                        26 of 43 ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED 36.b
: b. PRZR level - GREATER THAN                  b. Perform the following:
9% [37% ADVERSE].
Operate ECCS Pumps as necessary.
Initiate ATTACHMENT C as necessary to re-establish CCP Cold Leg Injection.
Return to Step 15.
37
        *37. Check if SI Accumulators should          37.
be isolated:
37.a
: a. RCS Subcooling - GREATER                  a. WHEN at least two RCS WR THAN 24F [38F ADVERSE]                    Hot Leg temperatures less than 380&deg;F, THEN go to Step 38.
IF at least two RCS WR Hot Leg temperatures NOT less than 380&deg;F, THEN go to Step 39.
37.b
: b. PRZR level - GREATER THAN                  b. Return to Step 15.
9% [37% ADVERSE].
 
S Printed September 17, 2013 at 12:41
 
Approved By                                                                              Procedure Versi on J. B. Stanley                              Vogtle Electric Generating Plant              19012-C 33.3 Effective Date                                                                            Page Number ES - 1.2 POST-LOCA COOLDOWN AND 05/01/2013                                        DEPRESSURIZATION                          43 of 43 FOLDOUT PAGE      Guidance is also repeated on the foldout page.
: 1.        SI REINITIATION CRITERIA Operate ECCS pumps as necessary if EITHER condition listed below occurs. Initiate ATTACHMENT C if it is necessary to re-establish CCP cold leg injection.
RCS subcooling - LESS THAN 24F [38F ADVERSE].
PRZR level - CANNOT BE MAINTAINED GREATER THAN 9% [37% ADVERSE].
: 2.        SECONDARY INTEGRITY CRITERIA Go to 19020-C, E-2 FAULTED STEAM GENERATOR ISOLATION, if any SG pressure is lowering in an uncontrolled manner or has been completely depressurized, and has not been isolated.
: 3.        E-3 TRANSITION CRITERIA Go to 19030-C, E-3 STEAM GENERATOR TUBE RUPTURE, if any SG level rises in an uncontrolled manner or any SG has abnormal radiation.
: 4.        COLD LEG RECIRCULATION SWITCHOVER CRITERION Go to 19013-C, ES-1.3 TRANSFER TO COLD LEG RECIRCULATION, if RWST level lowers to less than 29%.
: 5.        AFW SUPPLY SWITCHOVER CRITERION Switch to alternate CST by initiating 13610, AUXILIARY FEEDWATER SYSTEM, when CST level lowers to less than 15%.
 
S Printed September 17, 2013 at 12:41
 
Approved By                                                                        Procedure    Version M.G. Brill                              Vogtle Electric Generating Plant          19001-C      34.1 Effective Date                                                                      Page Number ES - 0.1 REACTOR TRIP RESPONSE 08/28/2013                                                                                26 of 26 FOLDOUT
: 1. SI ACTUATION CRITERIA Actuate SI and go to Procedure 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION, if EITHER condition listed below occurs:
RCS subcooling - LESS THAN 24F.
PRZR level - CANNOT BE MAINTAINED GREATER THAN 9%.
: 2. AFW SUPPLY SWITCHOVER CRITERION Switch to alternate CST by initiating 13610, AUXILIARY FEEDWATER SYSTEM when CST level lowers to less than 15%.
: 3. Monitor SPENT FUEL POOL COOLING conditions:
Verify annunciators 17005-A6, SPENT FUEL PIT HI TEMP and 17005-E2, SPENT FUEL PIT LOW LEVEL are both clear.
IF SPENT FUEL POOL LEVEL OR COOLING alarms are NOT clear, THEN initiate 18030-C, LOSS OF SPENT FUEL POOL LEVEL OR COOLING.
IF applicable, Using PTDB TAB 26, determine time to restore SFP LEVEL OR COOLING is < time to reach 200&deg;F in Spent Fuel Pool.
IF NOT initiate 18030-C, LOSS OF SPENT FUEL POOL LEVEL OR COOLING.
 
S Printed September 17, 20139/17/2013 at 12:5612:56 PM 1
: 1. 006G2.4.47 001/LOIT AND LOCT/RO/C/A 4.2/4.2/006G2.4.47/LO-TA-37020///
At time 0100:
          - Unit 1 reactor tripped.
At time 0530:
          - Safety Injection occurred.
          - All RCPs were stopped.
At time 0730:
          - 19111-C, "Loss of Emergency Coolant Recirculation," is in progress.
          - The crew is at Step 18, "Check if ECCS can be terminated," with the following conditions:
                    - Containment pressure is 3.9 psig and stable.
                    - Total ECCS flow is 625 gpm.
                    - RCS WR pressure is 1045 psig and lowering.
                    - CETC temperature is 475oF and lowering.
                    - RVLIS Full Range is 65% and stable.
Which one of the following completes the following statement?
Based on the given conditions, the crew is required to _________ per 19111-C.
REFERENCE PROVIDED A. terminate ECCS flow and establish normal charging B. reduce ECCS flow to approximately 225 gpm C. reduce ECCS flow to approximately 320 gpm D. raise charging / ECCS flow as necessary K/A 006              Emergency Core Cooling G2.4.47          Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
K/A MATCH ANALYSIS The question requires the candidate to utilize plant parameters and trends during a Loss of Emergency Recirculation to diagnose ECCS effectiveness and determine the minimum amount of ECCS flow which is required based on guidance of 19111-C, Step Wednesday, February 26, 2014 9:29:48 AM                                                    1
 
18, Table 1, and/or Figure 1 to conserve RWST inventory.
Subcooling is required to be calculated utilizing RCS WR Press, Core Exit Thermocouples, and steam tables. Minimum flow is determined by approximation on a graph or by linear interpolation from a table.
EXPLANATION OF REQUIRED KNOWLEDGE Loss of Emergency Recirculation 19111-C checks for the ability to "terminate" SI in step
: 18. The EOP utilizes a relaxed termination criteria based on RVLIS Full Range and Subcooling. RVLIS requirements are modified by the number of RCPs running and subcooling is modified by the presence of adverse containment. These parameters determine the effectiveness of ECCS to cool the core. If sufficient cooling and inventory exist, SI is terminated and CCPs are aligned to the normal flow path. If sufficient inventory exist, but not sufficient subcooling, the EOP directs the operator to minimum ECCS flow in order to conserve RWST inventory. The minimum ECCS flow required is determined using either Table 1 or Figure 1 and is based on the decay heat, using time since Reactor Trip as the determining parameter. The flow determined is a minimum value and therefore the operator is expected to start and stop pump and operator valves within the control room as necessary to be at or above the minimum flow.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. RVLIS level is greater than required and subcooling is greater than the required non-adverse value. If the candidate does not recognize that adverse containment conditions exist, this would be the correct choice. Therefore, this distractor is plausible.
B. Correct.                  The candidate should determine that RVLIS level is greater than required, but subcooling is below the adverse value. This would require the candidate to determine a minimum flow from either the graph or table. The time since the event started is approximately 390 minutes which would result in a required flow of approximately 215 gpm, however this exact value is not given. The candidate must choose a flow rate greater than the required flow.
C. Incorrect. Plausible. The candidate should determine that RVLIS level is greater than required, but subcooling is below the adverse value. This would require the candidate to determine a minimum flow from either the graph or table. If the candidate incorrectly uses the time since SI reset instead of the time from reactor trip, 120 minutes would be used and a flow of approximately 315 gpm determined, however this exact value is not given. The candidate must choose a flow rate greater than the required flow.
D. Incorrect. Plausible. A common error candidates make is to navigate step 18 incorrectly. When evaluating RVLIS level, candidates go to the RNO when level is greater than specified instead of continuing down in the step. If this was occurred, the candidate would Wednesday, February 26, 2014 9:29:48 AM                                                              2
 
continue to step 24 which would direct the operator to raise charging flow as necessary to maintain RVLIS level. Therefore, this distractor is plausible.
Level:                            RO Tier # / Group #                  T2 / G1 K/A#                              006G2.4.47 Importance Rating:                4.2 / 4.2 Technical
 
==Reference:==
19111-C Rev 33.2 WOG Background ECA-1.1 Rev 2, 4/30/2005 References provided:              19111-C Rev 33.2 step 18, Figure 1 and Table 1; pages 13-16, 45, and 46 Learning Objective:              LO-LP-37114-12 State the intent of EOP 19111, Loss of Emergency Coolant Recirculation.
LO-PP-37115-02 Describe the actions taken to conserve RWST inventory for a loss of emergency coolant recirculation.
LO-PP-37115-04 List the ECCS termination criteria and their bases for a loss of emergency coolant recirculation.
LO-PP-37115-05 Discuss the ECCS injection flow control methods for a loss of emergency coolant recirculation.
LO-PP-37115-07 Discuss parameters used to confirm adequate ECCS injection flow is being maintained during a loss of emergency coolant recirculation.
LO-PP-37115-08 Desribe why the minimum required ECCS flow decreases with time following a reactor trip.
LO-TA-37020          Respond to a Loss of Emergency Coolant Recirculation Capability per 19111-C Question origin:                  BANK - LOIT question HL-SR-00000-02-6 Cognitive Level:                  C/A 10 CFR Part 55 Content:          41.10 / 43.5 / 45.12 Comments:
You have completed the test!
Wednesday, February 26, 2014 9:29:48 AM                                                            3
: 1. HL-SR-00000-02 006/HLSR00000/LO-TA-28004/061A2.04/////
A reactor trip with SI occurred at 0100. At 0600, the control room operators were directed to enter 19111-C, "Loss of Emergency Coolant Recirculation". The operators are now determining if SI flow can be terminated.
The following conditions currently exist at 0700:
        - Total SI flow rate = 600 gpm
        - CNMT Press = 3.8 psig
        - RCS WR Pressure = 1035 psig
        - CETC temperature = 500 degrees F
        - All RCPs are OFF
        - RVLIS Full Range = 65%
Based on these indications, the control room operators should:
A. Terminate SI and establish normal charging flow.
B. Reduce total SI flow rate to approximately 175 gpm.
C. Reduce total SI flow rate to approximately 200 gpm.
D. Reduce total SI flow rate to approximately 540 gpm.
Tuesday, January 14, 2014 10:24:33 AM                                                    1
 
Approved By                                                                              Procedure      Version C. S. Waldrup                            Vogtle Electric Generating Plant                19111-C        33.2 Effective Date                                                                              Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                        RECIRCULATION                                13 of 49 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 17.b
: b. RCS Subcooling based on Core                          b. Close Spray Valve for idle Exit TCs - GREATER THAN 24F                          RCP:
[38F ADVERSE].
RCP 1: PIC-0455C RCP 4: PIC-0455B Go to Step 18.
17.c
: c. Start RCP 4 or RCP 1 or other RCP(s) as necessary to provide Normal Spray using ATTACHMENT D.
17.d
: d. Close Spray Valve for idle RCP:
RCP 1: PIC-0455C RCP 4: PIC-0455B 18
        *18. Check if ECCS can be terminated:                    18.
18.a
: a. Applicable RVLIS indication:                          a. Go to Step 24.
RCP(s) running      Required Indication 0      Full Range greater than 63%
1      Dynamic Range greater than 25%
2      Dynamic Range greater than 34%
3      Dynamic Range greater than 50%
4      Dynamic Range greater than 72%
 
Step 18 continued on next page Printed January 14, 2014 at 13:43                    27
 
Approved By                                                                          Procedure    Version C. S. Waldrup                          Vogtle Electric Generating Plant              19111-C        33.2 Effective Date                                                                        Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                    RECIRCULATION                                14 of 49 ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED 18.b
: b. RCS Subcooling based on Core                    b. Establish minimum ECCS Exit TCs - GREATER THAN 74&deg;F                    flow to remove decay heat
[88&deg;F ADVERSE].                                  by performing the following:
18.b.
: 1)  Determine minimum ECCS flow required using the following:
TABLE 1 or FIGURE 1 18.b.2
: 2)  Throttle ECCS flow to minimum value.
18.b.3
: 3)  Go to Step 24.
CAUTION Repositioning Phase A Isolation Valves may cause radiation problems throughout the plant.
19
: 19.      Reset Containment Isolation Phase            19.
A.
20
: 20.      Establish Instrument Air to                  20.
Containment:
20.a
: a. Instrument Air pressure -                      a. Start additional Air GREATER THAN 100 PSIG.                          Compressors as necessary.
20.b
: b. Open INSTR AIR CNMT ISO                        b.
VLV HV-9378.
20.c
: c. Verify PRZR Spray Valves                        c.
operating as required.
 
S Printed January 14, 2014 at 13:43                27
 
Approved By                                                                                Procedure    Version C. S. Waldrup                            Vogtle Electric Generating Plant                19111-C        33.2 Effective Date                                                                              Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                        RECIRCULATION                                15 of 49 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 21
: 21.      Stop ECCS Pumps and place in                        21.
standby:
RHR Pumps SI Pumps All but one CCP 22
: 22.      Establish charging flow:                            22.
22.a
: a. Check Instrument Air -                                a. Establish Safety Grade AVAILABLE.                                            Charging by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM.
Go to Step 23.
22.b
: b. Verify CCP alternate miniflow valves in ENABLE PTL position:
HV-8508A - CCP-A RV TO RWST ISOLATION HV-8508B - CCP-B RV TO RWST ISOLATION Verify white Pressure Control Mode light - LIT 22.c
: c. Close BIT DISCH ISOLATION valves:
HV-8801A HV-8801B 22.d
: d. Set SEAL FLOW CONTROL HC-182 to maximum seal flow (HV-0182 closed).
 
Step 22 continued on next page Printed January 14, 2014 at 13:43                    27
 
Approved By                                                                          Procedure    Version C. S. Waldrup                            Vogtle Electric Generating Plant          19111-C        33.2 Effective Date                                                                        Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                        RECIRCULATION                            16 of 49 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 22.e
: e. Open CHARGING TO RCS ISOLATION valves:
HV-8105 HV-8106 22.f
: f. Establish desired charging flow using HV-0182 and FV-0121.
23
        *23. Maintain Seal Injection flow to all            23.
RCPs - 8 TO 13 GPM.
24
        *24. Check adequate charging/ECCS                    24.
flow:
24.a
: a. Applicable RVLIS indication:                    a. Raise charging/ECCS flow to maintain RVLIS indication as necessary.
RCP(s) running      Required Indication 0      Full Range greater than 63%
1      Dynamic Range greater than 25%
2      Dynamic Range greater than 34%
3      Dynamic Range greater than 50%
4      Dynamic Range greater than 72%
24.b
: b. Core Exit TCs - STABLE OR                        b. Raise charging/ECCS flow LOWERING.                                        to maintain TCs stable or lowering.
 
S Printed January 14, 2014 at 13:43                    27
 
Approved By                                                              Procedure      Version C. S. Waldrup                          Vogtle Electric Generating Plant 19111-C        33.2 Effective Date                                                            Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                    RECIRCULATION                  45 of 49 TABLE 1            Sheet 1 of 1 MINIMUM ECCS FLOW VERSUS TIME Time Since      ECCS Reactor Trip    Flow Rate (Minutes)        (GPM) 10              615 15              555 20              515 30              471 40              435 50              399 60              381 70              363 80              344 90              337 100              326 150              294 200              268 300              232 400              214 500              205 600              196 800              181 1000            170 2000            141 3000            127 4000            116 5000            109 10000            91 Printed January 14, 2014 at 13:43                  1
 
Approved By                                                                        Procedure    Version C. S. Waldrup                                Vogtle Electric Generating Plant    19111-C          33.2 Effective Date                                                                    Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                        RECIRCULATION                          46 of 49 FIGURE 1                Sheet 1 of 1 MINIMUM ECCS FLOW VERSUS TIME 1000 900 800 700 ECCS FLOW IN GPM 600 500 400 300 200 100 0
1          10          100        1000    10000 TIME SINCE REACTOR TRIP (MINUTES)
Printed January 14, 2014 at 13:43                      1
 
Approved By                                                                                Procedure      Version C. S. Waldrup                              Vogtle Electric Generating Plant                19111-C        33.2 Effective Date                                                                                Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                          RECIRCULATION                                13 of 49 ACTION/EXPECTED RESPONSE                                    RESPONSE NOT OBTAINED 17.b
: b. RCS Subcooling based on Core                              b. Close Spray Valve for idle Exit TCs - GREATER THAN 24F                              RCP:
[38F ADVERSE].
RCP 1: PIC-0455C RCP 4: PIC-0455B Go to Step 18.
17.c
: c. Start RCP 4 or RCP 1 or other RCP(s) as necessary to provide Normal Spray using ATTACHMENT D.
17.d
: d. Close Spray Valve for idle RCP:
RCP 1: PIC-0455C RCP 4: PIC-0455B 18
        *18. Check if ECCS can be terminated:                      18.
18.a
: a. Applicable RVLIS indication:                              a. Go to Step 24.
Distractor D RCP(s) running        Required Indication 0          Full Range 65%
greater than 63%
1          Dynamic Range greater than 25%
2          Dynamic Range greater than 34%
3          Dynamic Range greater than 50%
4          Dynamic Range greater than 72%
YES
 
Step 18 continued on next page Printed September 18, 2013 at 08:32                      27
 
Approved By                                                                          Procedure    Version C. S. Waldrup                          Vogtle Electric Generating Plant              19111-C        33.2 Effective Date                                                                        Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                    RECIRCULATION                                14 of 49 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 18.b
: b. RCS Subcooling based on Core NO                b. Establish minimum ECCS Exit TCs - GREATER THAN 74&deg;F                    flow to remove decay heat
[88&deg;F ADVERSE].                                  by performing the following:
18.b.
: 1)  Determine minimum 1045 psig =>                  ECCS flow required 1060psia => 551.8              using the following:
Containment                    F Tsat => 76 F Pressure =>                    subcooling                    TABLE 1 or FIGURE 1 Adverse values 18.b.2 used.                                                    2)  Throttle ECCS flow to minimum value.
18.b.3
: 3)  Go to Step 24.
CAUTION Repositioning Phase A Isolation Valves may cause radiation problems throughout the plant.
19
: 19. Reset Containment Isolation Phase              19.
A.
20
: 20. Establish Instrument Air to                    20.
Containment:
20.a
: a. Instrument Air pressure -                      a. Start additional Air GREATER THAN 100 PSIG.                          Compressors as necessary.
20.b
: b. Open INSTR AIR CNMT ISO                        b.
VLV HV-9378.
20.c
: c. Verify PRZR Spray Valves                        c.
operating as required.
 
S Printed September 18, 2013 at 08:32              27
 
Approved By                                                                                Procedure    Version C. S. Waldrup                            Vogtle Electric Generating Plant                19111-C        33.2 Effective Date                                                                              Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                        RECIRCULATION                                15 of 49 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 21
: 21.      Stop ECCS Pumps and place in                        21.
standby:
RHR Pumps SI Pumps All but one CCP 22
: 22.      Establish charging flow:                            22.
22.a
: a. Check Instrument Air -                                a. Establish Safety Grade AVAILABLE.                                            Charging by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM.
Go to Step 23.
22.b
: b. Verify CCP alternate miniflow valves in ENABLE PTL position:
HV-8508A - CCP-A RV TO RWST ISOLATION HV-8508B - CCP-B RV TO RWST ISOLATION Verify white Pressure Control Mode light - LIT 22.c
: c. Close BIT DISCH ISOLATION valves:
HV-8801A HV-8801B 22.d
: d. Set SEAL FLOW CONTROL HC-182 to maximum seal flow (HV-0182 closed).
 
Step 22 continued on next page Printed January 14, 2014 at 13:02                    27
 
Approved By                                                                          Procedure    Version C. S. Waldrup                            Vogtle Electric Generating Plant          19111-C        33.2 Effective Date                                                                        Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                        RECIRCULATION                            16 of 49 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 22.e
: e. Open CHARGING TO RCS ISOLATION valves:
HV-8105 HV-8106 22.f
: f. Establish desired charging flow using HV-0182 and FV-0121.
23
        *23. Maintain Seal Injection flow to all            23.
RCPs - 8 TO 13 GPM.
24
        *24. Check adequate charging/ECCS                    24.
flow:
24.a
: a. Applicable RVLIS indication:                    a. Raise charging/ECCS flow to maintain RVLIS indication Distractor D as necessary.
RCP(s) running      Required Indication 0      Full Range 65%
greater than 63%
1      Dynamic Range greater than 25%
2      Dynamic Range greater than 34%
3      Dynamic Range greater than 50%
4      Dynamic Range greater than 72%
24.b
: b. Core Exit TCs - STABLE OR                        b. Raise charging/ECCS flow LOWERING.                                        to maintain TCs stable or lowering.
 
S Printed January 14, 2014 at 13:02                    27
 
Approved By                                                                          Procedure      Version C. S. Waldrup                          Vogtle Electric Generating Plant              19111-C        33.2 Effective Date                                                                        Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                    RECIRCULATION                                45 of 49 TABLE 1                          Sheet 1 of 1 MINIMUM ECCS FLOW VERSUS TIME Time Since      ECCS Reactor Trip    Flow Rate (Minutes)        (GPM) 10              615 15              555 20              515 30              471 40              435 50              399 60              381 70              363 80              344 Interpolate for 90              337 120min =>
100              326      approximately 150              294      315gpm.
200              268 300              232 Interpolate for 400              214    390min =>
500              205    approximately 600              196    215gpm.
800              181 1000            170 2000            141 3000            127 4000            116 5000            109 10000            91 Printed September 18, 2013 at 08:32                1
 
Approved By                                                                            Procedure    Version C. S. Waldrup                                Vogtle Electric Generating Plant          19111-C          33.2 Effective Date                                                                          Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                        RECIRCULATION                              46 of 49 FIGURE 1                    Sheet 1 of 1 MINIMUM ECCS FLOW VERSUS TIME 1000 900 800 700 ECCS FLOW IN GPM 600 500                                        120 minutes = approx 315 degrees F.
400 390 minutes = approx 300                                              215 degrees F.
200 100 0
1          10          100        1000          10000 TIME SINCE REACTOR TRIP (MINUTES)
Printed September 18, 2013 at 08:32                    1
 
STEP DESCRIPTION TABLE FOR ECA1.1          Step 15 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, highhead SI pumps and lowhead 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 1, and this curve would be included in the plant specific emergency operating procedure as Figure ECA111. This plant specific curve can be developed by modifying Figure 1 as follows: The Yaxis values for flowrate in gpm/MWt should be multiplied by the plant specific MWt core rating to obtain flowrate values in GPM. The Xaxis values for time in minutes are used without modification. A plant ECA1.1 Background              43              HPRev. 2, 4/30/2005 HECA11BG.doc
 
specific curve is then plotted as flowrate (gpm) versus ECA1.1 Background              44            HPRev. 2, 4/30/2005 HECA11BG.doc
 
STEP DESCRIPTION TABLE FOR ECA1.1          Step 15 time (minutes). Note that Figure ECA111 in guideline ECA1.1 has been developed for a plant with a core rating of 3411 MWt and is included as an example.
ACTIONS:
o  Determine if RVLIS indication is greater than the full range or dynamic head range value, as applicable.
o Determine if RCS subcooling (based on core exit TCs) is greater than (R.12)&deg;F [(R.13)&deg;F for adverse containment]
o Determine minimum SI flow required from Figure ECA111 o Establish minimum SI flow INSTRUMENTATION:
o  RCS pressure indication o  Core exit TCs temperature indication o  RVLIS indication CONTROL/EQUIPMENT:
o  Highhead SI pump switches o  Charging/SI pump switches o  Lowhead SI pump switches KNOWLEDGE:
o Understanding of RVLIS function, configuration, and interpretation o Due to the less restrictive SI termination and reinitiation criteria provided in this guideline the operator should be especially alert for any decrease in RCS subcooling or vessel level that warrants SI reinitiation o This step is a continuous action step while in this guideline ECA1.1 Background              45              HPRev. 2, 4/30/2005 HECA11BG.doc
 
STEP DESCRIPTION TABLE FOR ECA1.1          Step 15 PLANTSPECIFIC INFORMATION:
o  (K.02) RVLIS full range value which is the top of the core, including allowances for instrument uncertainties.
o (L.05) RVLIS dynamic range value corresponding to an average system void fraction of 25 percent with 4 RCPs running, including allowances for instrument uncertainties.
o (L.06) RVLIS dynamic range value corresponding to an average system void fraction of 25 percent with 3 RCPs running, including allowances for instrument uncertainties.
o (L.07) RVLIS dynamic range value corresponding to an average system void fraction of 25 percent with 2 RCPs running, including allowances for instrument uncertainties.
o (L.08) RVLIS dynamic range value corresponding to an average system void fraction of 25 percent with 1 RCP running, including allowances for instrument uncertainties.
o (R.12) The sum of temperature and pressure measurement system errors, including allowances for normal channel accuracies, translated into temperature using saturation tables, plus 50&deg;F.
o (R.13) The sum of temperature and pressure measurement system errors, including allowances for normal channel accuracies and post accident transmitter errors, translated into temperature using saturation tables, plus 50&deg;F.
o If RVLIS is not available, RCS subcooling based on core exit TCs is sufficient for terminating SI since a 50&deg;F margin has been added to instrument uncertainties. This 50&deg;F margin allows sufficient time for operator action to reinitiate SI before core uncovery.
o As long as the RVLIS dynamic range uncertainty for the Westinghouse RVLIS design is less than +/6%, the uncertainty does not need to be included in the calculation of the plantspecific EOP setpoints.
ECA1.1 Background              46              HPRev. 2, 4/30/2005 HECA11BG.doc
 
ECA1.1 Background 47 HPRev. 2, 4/30/2005 HECA11BG.doc
: 1. 007A2.02 001/LOIT/RO/M/F 2.6/3.2/007A2.02/LO-PP-16301-11//HL17 NRC/
Initial conditions:
            - Unit 1 is in Mode 5 with solid plant conditions.
            - Pressurizer bubble is being established.
Current conditions:
            - A transient results in an RCS pressure spike.
            - ALB12-E02 PRZR REL TANK HI PRESS is received.
            - ALB12-E03 PRZR REL TANK HI TEMP is received.
Which one of the following completes the following statement?
The RHR pump __(1)__ relief valve lifting caused the PRT high pressure condition, and per 13004-1, "Pressurizer Relief Tank Operation," recirculation through the RCDT heat exchanger to lower PRT temperature is expected to take approximately __(2)__ .
__(1)__                                    __(2)__
A.                  discharge                                    1 hour B.                discharge                                    8 hours C.                  suction                                    1 hour D.                  suction                                    8 hours K/A 007              Pressurizer Relief / Quench Tank A2.02            Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Abnormal pressure in the PRT K/A MATCH ANALYSIS The question requires the candidate to determine the cause of the high pressure and high temperature annunciators associated with the PRT. This is backward logic from the first part of the specified K/A, which requires the candidate to predict how the malfunction will impact the PRT. This was necessary in order to write a technically valid question. The question as written requires the candidate to have knowledge of the Thursday, March 06, 2014 11:22:05 AM                                                          1
 
relationship between the malfunction and the PRT pressure. The only other possible topic that could be asked would be associated with the PRT rupture pressure. However, the rupture pressure does not lend itself to an A2 since there is no plausible distractor to pit against rupture as an impact.
The candidate is then required to choose the procedurally directed process which will correct / control the temperature and pressure increase based on the conditions depicted by the annunciators.
This question is a re-use from the HL-17 NRC exam.
EXPLANATION OF REQUIRED KNOWLEDGE Based on the PRT annunciators, both temperature and level are elevated. Coupled with the given RCS pressure spike, the most likely source of mass addition to the PRT would be from the RHR suction relief valves, which lift at 450 psig. These valves provide cold over-pressure protection in Mode 5 and pass a significant volume of water, which will raise PRT level, pressure, and temperature. The RHR discharge relief valves are sized for thermal protection and lift at 600 psig. These valves relieve to the RHUT.
With RHR pumps in service, the RHR discharge relief valves will lift first since the RHR pump creates approximately 200 psid (450psig + 200psid=650psig > 600psig).
However, RCS pressure will continue to rise and the suction relief valves will open due to the small flow rate of the discharge relief valvess.
Per 17012-1 E02 and E03, PRT pressure and temperature should be restored per 13004-1. 13004-1 gives direction on how to control PRT pressure, level, and temperature. Per Limitation 2.2.1, if the PRT High Temperature Alarm is annunciated, the contents of the PRT shall be cooled. Cooling the PRT can be accomplished by two methods - spray or recirculation. Per the procedure titles and the note at the beginning of each section, cooling by spray is designed to take 1 hr and cooling by recirculation is designed to take 8 hrs. Therefore, the use of spray is fastest.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. The RHR discharge valves relieve to the RHUT and therefore could not have caused the rise in PRT temperature and pressure. However, the candidate may believe the RHR discharge relief valves are located inside containment and therefore relieve to the PRT. With the added discharge pressure, the disscharge relief valves will lift before the suction relief valves and result in the described PRT conditions.
The second part is incorrect. SOP-13004 describes the 1 hour cooldown method as the spray and drain and the 8 hours method as cooldown using the RCDT heat exchanger.
However, a candidate without specific knowledge of the PRT cooling methods times could find it reasonable that recirculation is a faster method since it would cool the bulk liquid temperature faster than a spray method, which tends to create thermal stratification.
Thursday, March 06, 2014 11:22:05 AM                                                                  2
 
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. SOP-13004 describes the 1 hour cooldown method as the spray and drain and the 8 hour cooldown method as using the RCDT heat exchanger.
C. Incorrect. Plausible. The first part is correct. The RHR suction relief valves are located inside Containment and, if they lift during a RCS pressure transient, will result in the described PRT conditions.
The second part is incorrect. See the second part of choice A above D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            007A2.02 Importance Rating:              2.6 / 3.2 Technical
 
==Reference:==
17012-1 Rev 21 13004-1 Rev 20 1X4DB121 Rev 42.0 1X4DB122 Rev 51.0 References provided:            None Learning Objective:              LO-PP-16301-01 List the sources that input into the PRT.
LO-PP-16301-09 Describe the methods for cooling the PRT.
Question origin:                MODIFIED - HL17 # 007A2.06 Cognitive Level:                M/F 10 CFR Part 55 Content:          41.5 / 43.5 / 45.3 / 45.13 Comments:
You have completed the test!
Thursday, March 06, 2014 11:22:05 AM                                                                  3
: 1. 007A2.06 001/2/1/PRT VACUUM PROCS/H-2.6/2.8/NEW/HL-17 NRC/RO/SRO/TNT/GCW Initial conditions:                          Original Question
          - Unit 1 is solid plant.
          - PRZR bubble is being drawn per 12001-C, "Unit Heatup to Hot Shutdown".
Current conditions:
          - A transient results in an RCS pressure spike.
          - ALB12-E03 PRZR REL TANK HI TEMP illuminates Which one of the following correctly completes the statement below?
The RHR pump ___(1)___ relief valve lifting caused the PRT High Temperature and per 13004-1, "Pressurizer Relief Tank Operation", the FASTEST way to cooldown the PRT is using ___(2)___.
A. (1) discharge (2) spray from RMWST and drain to the RCDT B. (1) discharge (2) recirculation through the RCDT heat exchanger C. (1) suction (2) spray from RMWST and drain to the RCDT D. (1) suction (2) recirculation through the RCDT heat exchanger Thursday, March 06, 2014 11:19:27 AM                                                    1
 
Approved By                                                                              Procedure      Version W.R. Dunn                            Vogtle Electric Generating Plant                    17012-1      21 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON                Page Number 05/06/2013                                      PANEL 1C1 ON MCB                                    3 of 52 (1)                  (2)          (3)          (4)            (5)            (6)
RCP LOOP 1                          RC LOOP      RC LOOP        TAVG/TREF      OVERTEMP A      LOW FLOW                            T/AUCT T    TAVG/AUCT TAVG DEVIATION      T ALERT ALERT HI-LO DEV    HI-LO DEV RCP LOOP 2                                        AUCT TAVG      TAVG            OVERPOWER B      LOW FLOW                                          HIGH          LO-LO ALERT    T ALERT ALERT RCP LOOP 3                                        A COLD OP      B COLD OP      TERR C      LOW FLOW                                          LOW AUCT RCS  LOW AUCT RCS    (TAVG - TREF)
ALERT                                            TEMP          TEMP LO RCP LOOP 4                          PRZR PRESS    A RCS PRESS    B RCS PRESS    RV VENT D      LOW FLOW                            LO PORV      APPROACHES    APPROACHES      HI TEMP ALERT                              BLOCK        COLD OP LIMIT  COLD OP LIMIT PRZR RELIEF          PRZR REL TANK PRZR REL TANK PV-0455A                      A COLD OP ACTU E      DISCH                HI PRESS      HI TEMP OPEN SIGNAL                    VLV HV-8000A HI TEMP NOT FULL OPEN PRZR SAFETY          PRZR REL TANK RV FLG        PV-0456A                      B COLD OP ACTU F      RELIEF DISCH          HI/LO LEVEL  LKOF HI TEMP OPEN SIGNAL                    VLV HV-8000B HI TEMP NOT FULL OPEN Printed September 18, 2013 at 12:55
 
Approved By                                                                                    Procedure    Version W.R. Dunn                          Vogtle Electric Generating Plant                          17012-1    21 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON                    Page Number 05/06/2013                                      PANEL 1C1 ON MCB                                      33 of 52 WINDOW E02 ORIGIN                            SETPOINT PRZR REL TANK 1-PT-0469                        8 psig                        HI PRESS 1.0              PROBABLE CAUSE
: 1.        One or more of the following valves has lifted or is leaking to the Pressurizer Relief Tank:
: a.      Pressurizer Safety Valves,
: b.      Pressurizer (PRZR) Power Operated Relief Valves (PORV)s,
: c.      Chemical and Volume Control System (CVCS) Letdown Relief Valve 1-PSV-8117,
: d.      CVCS Seal Return Relief Valve 1-PSV-8121,
: e.      Residual Heat Removal (RHR) Relief Valves 1-PSV-8708A and B during shutdown conditions.
: 2.        Nitrogen Regulator malfunction.
: 3.        Safety grade letdown in use and aligned to the Pressurizer Relief Tank.
2.0              AUTOMATIC ACTIONS NONE 3.0              INITIAL OPERATOR ACTIONS NONE Printed September 18, 2013 at 12:55
 
Approved By                                                                                  Procedure    Version W.R. Dunn                          Vogtle Electric Generating Plant                          17012-1    21 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON                    Page Number 05/06/2013                                        PANEL 1C1 ON MCB                                    34 of 52 WINDOW E02 (Continued) 4.0              SUBSEQUENT OPERATOR ACTIONS CAUTION If PRT pressure increases the PRT rupture disk will fail at 86 to 100 psig, opening the PRT to containment.
: 1.        Determine actual Pressurizer Relief Tank pressure using 1-PI-0469 on the QMCB.
: 2.        Monitor Pressurizer Relief Tank temperature, level, and pressure.
: 3.        Check tailpipe temperatures for the Pressurizer Safety Valves, Power Operated Relief Valves, and CVCS Letdown Relief Valve.
: 4.        IF a PRZR PORV OR Safety Valve has actuated, check valve closure when pressure is lowered in the Reactor Coolant System.
: 5.        IF a nitrogen supply malfunction has occurred, isolate the supply by shutting valves 1-HV-8033 and 1-HV-8047.
: 6.        IF a Pressurizer Safety Valve is open OR fails to close following an actuation, Go To 18004-C, "Reactor Coolant System Leakage."
: 7.        IF a PRZR PORV 455A/456A is open OR fails to close following an actuation:
: a.      Place the Control Switch for the affected valve to the closed position,
: b.      IF the affected valve will NOT close, close the associated Block Valve,
: c.      Refer to Technical Specification LCO 3.4.11.
: 8.        IF the pressure rise is due to the CVCS Letdown Relief Valve being open isolate letdown, and initiate 18007-C, "Chemical And Volume Control System Malfunction."
Printed September 18, 2013 at 12:55
 
Approved By                                                                                  Procedure    Version W.R. Dunn                          Vogtle Electric Generating Plant                          17012-1    21 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON                    Page Number 05/06/2013                                      PANEL 1C1 ON MCB                                    35 of 52 WINDOW E02 (Continued)
: 9.        IF the pressure rise is due to a failed RHR Relief Valve, isolate the affected Train of RHR and initiate 18019-C, "Loss Of Residual Heat Removal."
: 10.      IF the pressure rise is due to a failed Seal Return Relief Valve, attempt to isolate the leak.
: 11.      IF pressure rise is due to a hard bubble, i.e., no temperature or level change, notify Chemistry and place Pressurizer Steam Space Sample in service and control RCS pressure using 12004-C.
: 12.      Restore pressure in the Pressurizer Relief Tank to normal per 13004-1, "Pressurizer Relief Tank Operation."
: 13.      IF equipment failure is indicated, initiate maintenance as required.
5.0              COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB112, PLS Printed September 18, 2013 at 12:55
 
Approved By                                                                                    Procedure    Version W.R. Dunn                          Vogtle Electric Generating Plant                          17012-1    21 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON                    Page Number 05/06/2013                                      PANEL 1C1 ON MCB                                      36 of 52 WINDOW E03 ORIGIN                            SETPOINT PRZR REL TANK 1-TE-0468                        115&deg;F                          HI TEMP 1.0              PROBABLE CAUSE
: 1.        One or more of the following valves has lifted or is leaking to the Pressurizer Relief Tank:
: a.      Pressurizer Safety Valve,
: b.      Pressurizer (PRZR) Power Operated Relief Valves (PORV)s,
: c.      Chemical and Volume Control System (CVCS) Letdown Relief Valve 1-PSV-8117,
: d.      CVCS Seal Return Relief Valve 1-PSV-8121,
: e.      Residual Heat Removal (RHR) Relief Valves 1-PSV-8708A and B during shutdown conditions.
: 2.        Safety grade letdown in use and aligned to the Pressurizer Relief Tank.
2.0              AUTOMATIC ACTIONS NONE 3.0              INITIAL OPERATOR ACTIONS NONE 4.0              SUBSEQUENT OPERATOR ACTIONS
: 1.        Determine the actual temperature of the Pressurizer Relief Tank using 1-TI-0468 on the QMCB.
: 2.        Check tailpipe temperatures for the Pressurizer Safety Valves, PRZR PORVs, and CVCS Letdown Relief Valve.
Printed September 18, 2013 at 12:55
 
Approved By                                                                                  Procedure    Version W.R. Dunn                          Vogtle Electric Generating Plant                          17012-1    21 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON                    Page Number 05/06/2013                                        PANEL 1C1 ON MCB                                    37 of 52 WINDOW E03 (Continued)
: 3.        IF a Pressurizer Safety OR PRZR PORV Valve has actuated, check valve closes when system pressure is reduced.
: 4.        IF a Pressurizer Safety Valve is open OR fails to close following an actuation, initiate 18004-C, "Reactor Coolant System Leakage."
: 5.        IF a PRZR PORV 455A/456A is open OR fails to close following an actuation:
: a.      Place the Control Switch for the affected valve to the closed position,
: b.      IF the affected valve will NOT close, close the associated Block Valve,
: c.      Refer to Technical Specification LCO 3.4.11.
: 6.        IF the temperature rise is due to the CVCS Letdown Relief Valve being open, isolate letdown and initiate 18007-C, "Chemical And Volume Control System Malfunction."
: 7.        IF the temperature rise is due to a failed RHR Relief Valve, isolate the affected Train of RHR and initiate 18019-C, "Loss Of Residual Heat Removal."
: 8.        IF the temperature rise is due to a failed Seal Return Relief Valve, attempt to isolate the leak.
: 9.        Restore the Pressurizer Relief Tank temperature to normal per 13004-1, "Pressurizer Relief Tank Operation."
: 10.      IF equipment failure is indicated, initiate maintenance as required.
5.0              COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB112, PLS Printed September 18, 2013 at 12:55
 
Approved By                                                                              Procedure    Version M.C. Henry                          Vogtle Electric Generating Plant                      13004-1      20 Effective Date                                                                            Page Number 04/18/2013                        PRESSURIZER RELIEF TANK OPERATION                                3 of 43 1.0                PURPOSE This procedure provides the necessary instructions for operation of the Pressure Relief Tank (PRT) and supporting equipment. Procedure instructions include the following:
4.1.2        Placing The PRT In Service 4.2.1        Pressure Control Of The PRT 4.2.2        Level Control Of The PRT 4.4.1        Purging The PRT To The WGS 4.4.2        Purging The PRT To Containment HVAC 4.4.3        PRT Cooldown Using Spray And Drain (One Hour Cooldown) 4.4.4        PRT Cooldown Using RCDT Heat Exchanger (Eight Hour Cooldown) 4.4.5        Draining The PRT 4.4.6        Venting PRT For PRZ Code Safeties OR Manway removal Printed January 14, 2014 at 15:17
 
Approved By                                                                            Procedure    Version M.C. Henry                          Vogtle Electric Generating Plant                  13004-1      20 Effective Date                                                                        Page Number 04/18/2013                        PRESSURIZER RELIEF TANK OPERATION                            4 of 43 INITIALS 2.0                PRECAUTIONS AND LIMITATIONS 2.1                PRECAUTIONS 2.1.1              Equipment that has been exposed to air must be purged before it is connected to the Vent Header to prevent explosive mixtures of hydrogen and oxygen.                                                ________
2.1.2              A nitrogen gas blanket should be maintained in the PRT to exclude air and prevent the formation of an explosive hydrogen and oxygen mixture.                                                            ________
2.1.3              Before opening the PRT to atmosphere, verify that the gas space hydrogen content is less than 4% and the oxygen content is less than 5%.                                                            ________
2.2                LIMITATIONS 2.2.1              If the PRT High Temperature Alarm (115&deg;F) is annunciated, the contents of the PRT shall be cooled.                                ________
2.2.2              The PRT rupture disk will fail at 86 to 100 PSIG.                  ________
3.0                PREREQUISITES AND INITIAL CONDITIONS 3.1                The Gaseous Waste Processing System is available to provide processing of gases from the PRT.                                  ________
3.2                The Auxiliary Gas System-Nitrogen or nitrogen from the Waste Gas Decay Shutdown Tank is available to provide a nitrogen blanket for the PRT.                                                            ________
3.3                Reactor Make-Up Water is available to provide a cooling spray for the PRT.                                                            ________
3.4                ACCW is available if cooling the PRT with the RCDT Heat Exchanger.                                                          ________
Printed January 14, 2014 at 15:17
 
Approved By                                                                            Procedure  Version M.C. Henry                            Vogtle Electric Generating Plant                13004-1      20 Effective Date                                                                        Page Number 04/18/2013                          PRESSURIZER RELIEF TANK OPERATION                        29 of 43 INITIALS 4.4.3              PRT Cooldown Using Spray And Drain (One Hour Cooldown)
NOTE Two methods for cooling the PRT exist. Cooling the PRT by spray and drain is designed to cool the PRT in 1 hour. This method uses makeup water and drains to the Waste Processing System. Cooling the PRT by recirculation through the RCDT Hx is designed to cool the PRT in 8 hours. This method minimizes makeup water use and waste processing of liquid. The time required to cool the PRT and water usage should be considered before deciding which method to use.
4.4.3.1            Establish communications between the Liquid Waste Processing System Panel (PLPP) and the Control Room.                            ________
4.4.3.2            Verify the PRT pressure less than or equal to 50 psig as indicated by PRESSURIZER RELIEF TANK 1-PI-0469 to prevent RCDT System over pressurization.                                          ________
4.4.3.3            Verify open WPSL RCDT PUMPS DISCH TO RECYC EVAP 1-1901-U6-327. (1AB-RA27)                                            ________
4.4.3.4            Realign RCDT Pump Suction to the PRT and initiate spray as follows:
: a.        Stop the running REACTOR COOLANT DRAIN TANK PUMP
                              #1 1HS-1003A (PLPP)                                        ________
                              #2 1HS-1003B (PLPP)                                        ________
CAUTION The RCDT level should be monitored to prevent tank flooding.
: b.        Place REACTOR COOLANT DRAIN TANK LEVEL 1-LC-1003 in MANUAL and open the valve (PLPP).            ________
: c.        Close REACTOR COOLANT DRAIN TANK RECIRCULATION VALVE 1-HV-7144 (PLPP).                      ________
Printed January 14, 2014 at 15:17
 
Approved By                                                                          Procedure  Version M.C. Henry                            Vogtle Electric Generating Plant                13004-1      20 Effective Date                                                                        Page Number 04/18/2013                          PRESSURIZER RELIEF TANK OPERATION                        32 of 43 INITIALS 4.4.4              PRT Cooldown Using RCDT Heat Exchanger (Eight Hour Cooldown)
NOTE Two methods for cooling the PRT exist. Cooling the PRT by spray and drain is designed to cool the PRT in 1 hour. This method uses makeup water and drains to the Waste Processing System. Cooling the PRT by recirculation through the RCDT Hx is designed to cool the PRT in 8 hours. This method minimizes makeup water use and waste processing of liquid. The time required to cool the PRT and water usage should be considered before deciding which method to use.
4.4.4.1            Establish communications between the Liquid Waste Processing System Panel (PLPP) and the Control Room.                          ________
4.4.4.2            Verify PRESSURIZER RELIEF TANK 1-PI-0469 indicates less than OR equal to 50 psig to prevent RCDT System over pressurization. (QMCB)                                              ________
4.4.4.3            Realign the RCDT System to recirc the PRT as follows:
: a.        Stop the running REACTOR COOLANT DRAIN TANK PUMP
                              #1 1HS-1003A (PLPP)                                      ________
                              #2 1HS-1003B (PLPP)                                      ________
CAUTION The RCDT level should be monitored to prevent tank flooding.
: b.        Close REACTOR COOLANT DRAIN TANK PUMP SUCTION VALVE 1-HV-7127 (PLPP).                          ________
: c.        Close REACTOR COOLANT DRAIN TANK RECIRCULATION VALVE 1-HV-7144 (PLPP).                    ________
: d.        Open REACTOR COOLANT DRAIN TANK PRESSURE RELIEF TANK VALVE 1-HV-7141 (PLPP).                      ________
Printed January 14, 2014 at 15:17
: 1. 007EA2.06 001/LOIT AND LOCT/RO/C/A 4.3/4.5/007EA2.06/LO-TA-37002/27009///
Initial condition:
          - Unit 1 is operating at 55% reactor power with a plant startup in progress.
Current conditions:
          - Both Unit 1 RATs de-energize.
          - Both DGs start and re-energize their respective bus.
          - No other switchyard components are affected.
Which one of the following completes the following statement?
With no operator action, one minute after the RATs de-energize, the Reactor Trip Breakers will be __(1)__,
and DRPI __(2)__ be available to check control rod positions.
__(1)__                                __(2)__
A.                    open                                      will B.                    open                                  will NOT C.                  closed                                      will D.                  closed                                  will NOT K/A 007              Reactor Trip - Stabilization - Recovery EA2.06          Ability to determine or interpret the following as they apply to a reactor trip:
Occurrence of a reactor trip.
K/A MATCH ANALYSIS The question presents the candidate with a loss of both RATs and then requires the candidate to determine if a reactor trip occurs as a result of the LOSP, and whether DRPI will be energized after the busses are reenergized by the EDGs.
EXPLANATION OF REQUIRED KNOWLEDGE Wednesday, February 26, 2014 9:42:11 AM                                                    1
 
The candidate is required to recognize that above 50% power, the non-1E 13.8 KV and 4160 VAC buses will remain energized from the UATs. The 1E 4160 VAC buses will de-energize and be re-energized by the Emergency Diesel Generators (EDGs). As such, no reactor trip signals will be generated directly or indirectly. Additionally, DRPI is normally powered from 'B' train 1E power (1BA03->1BBC->1NYC2).
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. As described in the Explanation of Required Knowledge above, all buses are energized after the loss of the RATs and no reactor trip signals are generated.
However, below 50% power the non-1E buses are aligned to the RATs. A candidate not familiar with when the non-1E buses are transferred could conclude that the non-1E buses de-energized. In this case, several reactor trip signals would be energized.
The second part is correct. 1BA03 was re-energized by the DG and the sequence re-energized 1BBC, which re-energized DRPI.
B. Incorrect. Plausible. The first part is incorrect. See part one of Choice A above.
The second part is incorrect. As described in the Explanation of Required Knowledge above, DRPI is ultimately powered from 1BA03, which will be re-energized by its EDG. However, DRPI is powered from a non-1E panel 1NYC2. If the candidate is not familiar with the one-line distribution for DRPI, it could be reasonable for a candidate, who concluded that non-1E buses de-energized in part one of this distractor, to also conclude that DRPI, powered from a non-1E 120VAC panel, would also be de-energized.
C. Correct.                  The first part is correct. As described in the Explanation of Required Knowledge above, all buses are energized after the loss of the RATs and no reactor trip signals are generated.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. As described in the Explanation of Required Knowledge above, DRPI is ultimately powered from 1BA03, which will be re-energized by its EDG. It is reasonable for a candidate without specific knowledge of DRPI power supplies to recongnize DRPI is de-energized when 1BA03 losses power during scenarios and conclude DRPI will not be automatically restored on a loss of power, because DRPI is non-safety related and there are no steps in the EOPs to ensure it is energized.
Wednesday, February 26, 2014 9:42:11 AM                                                              2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            007EA2.06 Importance Rating:              4.3 / 4.5 Technical
 
==Reference:==
13432-1 Rev 49 References provided:            None Learning Objective:              LO-PP-28103-03 List all reactor trip set points, coincidences, permissives, and blocks.
LO-PP-28103-05 List all ESF actuation signals with applicable set points, coincidences, permissives, blocks, and discuss the systems response to each ESF actuation signal.
LO-PP-27201-03 State the power supplies for the DRPI System.
LO-TA-37002        Respond to a Reactor Trip Without Safety Injection per 19000-C and 19001-C LO-TA-27009        Swap DRPI Power Supplies using 13432-1/2 Question origin:                BANK - FNP EXAM BANK Question # 007EA2.06 007 Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 / 45.5 / 45.6 Comments:
You have completed the test!
Wednesday, February 26, 2014 9:42:11 AM                                                          3
 
Approved By                                                                              Procedure      Version M.G. Brill                              Vogtle Electric Generating Plant                  18031-C          28 Effective Date                                                                            Page Number LOSS OF CLASS 1E ELECTRICAL SYSTEMS 05/24/2013                                                                                      9 of 34 A. LOSS OF POWER WITH DG FAILING TO TIE TO BUS ACTION/EXPECTED RESPONSE                              RESPONSE NOT OBTAINED A14 A14. Verify DRPI - ENERGIZED.                            A14. Perform the following:
A14.a
: a. Swap DRPI power supply using 13432, 120V AC NON 1E INSTRUMENT DISTRIBUTION SYSTEM.
NORMAL        ALTERNATE UNIT    LOCATION      SUPPLY        SUPPLY 1    1NYC2          1BBC-20        1ABC-20 (CB-B66) 2    2NYC2          2BBC-20        2ABC-20 (CB-B12)
A14.b
: b. IF DRPI can NOT be energized, THEN refer to the following technical specifications as applicable:
TS 3.0.3 TS 3.1.7 TR 13.1.8 TR 13.1.9 A15
        *A15. Check DC bus loads:                                  A15.
A15.a
: a. Verify 125V DC battery loads -                      a. Evaluate selective load stripping LESS THAN THE FOLLOWING                            using ATTACHMENT B, DC LIMITS:                                            LOADS TO EVALUATE FOR LOAD STRIPPING DURING AD1B 300 AMPS LOSS OF 1E BUS.
BD1B 300 AMPS CD1B 100 AMPS DD1B 80 AMPS
 
Step 15 continued on next page Printed January 15, 2014 at 09:34
 
Approved By                                                                              Procedure  Version C. H. Williams, Jr.            Vogtle Electric Generating Plant                        13432-1      49 Effective Date                                                                          Page Number 6/3/13                    120V AC NON 1E INSTRUMENT DISTRIBUTION SYSTEM                      55 of 116 INITIALS 4.2.3            Transferring Regulated Instrument Distribution Panel 1NYC2, 1NYJ, 1NYR, 1NYS or 1NYRS to Alternate Source NOTES The normal regulated transformer is the preferred power source for the instrument distribution panels.
Any TS LCO or TRM actions entered as a result of panels being de-energized should be evaluated for exit after panel is re-energized and loads verified operable.
CAUTION When power is transferred for 1NYC2, DPM 1RX-12444 may be damaged if not removed from service by Chemistry.
4.2.3.1          To transfer 1NYC2 to the alternate regulated source 1ABC, perform the following:
: a.      Request Chemistry remove power at DPM 1RX-12444 to support transfer.                                              ________
Critical
: b.      Close the Alternate Regulated Transformer Supply Breaker 1ABC-20. (CB-B76)                                              ________
________
CV
 
Approved By                                                                            Procedure  Version C. H. Williams, Jr.            Vogtle Electric Generating Plant                        13432-1      49 Effective Date                                                                        Page Number 6/3/13                  120V AC NON 1E INSTRUMENT DISTRIBUTION SYSTEM                      56 of 116 INITIALS NOTE Steps 4.2.3.1.c and 4.2.3.1.d should be performed as quickly as possible to minimize the time the panel is de-energized.
: c. Open the normal Instrument Distribution Panel Breaker 1NYC2-02. (CB-B66)                                              ________
: d. Close the alternate Instrument Distribution Panel Breaker 1NYC2-01. (CB-B66)                                              ________
: e. Verify proper operation of 1NYC2 by observing approximately 120V AC on the Instrument Distribution Panel Voltmeter.                                                ________
Critical
: f. WHEN 1NYC2 is transferred to the Alternate Source, open the Normal Regulated Supply Breaker 1BBC-20. (CB-61)            ________
________
CV
: g. Request Chemistry restore power to DPM 1RX-12444 following energization.                                        ________
 
Approved By                                                                              Procedure  Version C. H. Williams, Jr.            Vogtle Electric Generating Plant                        13432-1      49 Effective Date                                                                          Page Number 6/3/13                    120V AC NON 1E INSTRUMENT DISTRIBUTION SYSTEM                      70 of 116 INITIALS 4.2.4            Transferring Regulated Instrument Distribution Panel 1NYC2, 1NYJ, 1NYR, 1NYS, or 1NYRS to Normal Source NOTES The normal regulated transformer is the preferred power source for the instrument distribution panels.
Any TS LCO or TRM actions entered as a result of panels being de-energized should be evaluated for exit after panel is re-energized and loads verified operable.
CAUTION When power is transferred for 1NYC2, DPM 1RX-12444 may be damaged if not removed from service by Chemistry.
4.2.4.1          To transfer 1NYC2 to the normal regulated source 1BBC, perform the following:
: a.      Request Chemistry remove power at DPM 1RX-12444 to support transfer.                                              ________
Critical
: b.      Close the Normal Regulated Transformer Supply Breaker 1BBC-20.                                                        ________
________
CV NOTE Steps 4.2.4.1.c and 4.2.4.1.d should be performed as quickly as possible to minimize the time the panel is de-energized.
: c.      Open the alternate Instrument Distribution Panel Breaker 1NYC2-01.                                                      ________
 
Approved By                                                                  Procedure  Version C. H. Williams, Jr.          Vogtle Electric Generating Plant                13432-1      49 Effective Date                                                                Page Number 6/3/13                120V AC NON 1E INSTRUMENT DISTRIBUTION SYSTEM                71 of 116 INITIALS
: d. Close the normal Instrument Distribution Panel Breaker 1NYC2-02.                                                ________
: e. Verify proper operation of 1NYC2 by observing approximately 120V AC on the Instrument Distribution Panel Voltmeter.                                        ________
Critical
: f. WHEN 1NYC2 is transferred to the normal Source, open the alternate Regulated Supply Breaker 1ABC-20.          ________
________
CV
: g. Request Chemistry restore power at DPM 1RX-12444 following energization.                                  ________
: 1. 008A1.03 001/LOIT/RO/C/A 2.7/2.9/008A1.03/LO-TA-60026///
Initial conditions:
            -  Unit 1 is at 100% reactor power.
            -  CCW Train 'A' is in service.
            -  Multiple CCW Train 'A' low pressure and low flow alarms annunciate, then clear.
            -  No other alarms were received.
Current conditions:
            - CCW Train 'A' pressure and flow are stable.
            - 18020-C, "Loss of Component Cooling Water," is entered.
Which one of the following completes the following statement?
A CCW Train 'A' pump has experienced a __(1)__,
and the standby CCW Train 'A' pump __(2)__ auto start in response to CCW low header pressure.
__(1)__                                __(2)__
A.                locked rotor                                did B.                locked rotor                            did NOT C.              sheared shaft                                did D.              sheared shaft                            did NOT K/A 008              Component Cooling Water A1.03            Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCWS controls including:
CCW pressure K/A MATCH ANALYSIS The KA addresses the relationship between the CCW System and the operators ability to monitor the system, specifically including pressure, and make decisions on the proper response and actions. The question requires the candidate to evaluate various indications and observed system responses, and determine whether the system is responding correctly to the malfunction or if operator intervention is required.
Wednesday, February 26, 2014 9:43:51 AM                                                      1
 
Note: Unlike the typcial CCW system on most Westinghouse plants, the CCW system is split into two systems on Vogtle 1&2 - CCW and ACCW. CCW only contains SFP Cooling, RHR Hx, and RHR seal cooler. As such, both CCW and ACCW are very simple. The only active valves in the CCW system are the pump discharge check valves. The only control circuits are associated with pump starts and trips. ACCW, which only supports non-essential loads, additionally has active valves around the RCP thermal barriers but these are already addressed on a different question in this exam.
Therefore, this question focus is the best overall fit for the K/A.
EXPLANATION OF REQUIRED KNOWLEDGE If a shaft shear occurs on one of the two operating CCW pumps, CCW system flow and pressure decrease and bring in multiple system alarms. With at least one CCW handswitch in AUTO, the low pressure auto start is enabled. The standby CCW pump will start and restore system pressure and flow to normal, clearing all the annunciators.
All three CCW pump motors will be running. The pump with the sheared shaft will not trip.
In contrast, a CCW pump with a locked rotor will produce the same sequence of events, except the handswitch on the pump with the locked rotor will have amber and green lights lit. Additionally, a QEAB annunciator will alarm when the associated 4160V breaker trips. In this case, the standby pump will start when the 52 contact changes state and not on low header pressure. The 52 contact change is instantaneous, whereas the low header pressure start requires system pressure to decay over a period of time.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. The candidate could determine the low pressure alarms are consistent with the trip of a running pump and the system response is to auto start the standby pump, rather than the pressure decay causing the pump to start on low pressure. The candidate may not consider the absence of electrical alarms in this condition when making their decision.
The second part is correct. Per the Explanation of Required Knowledge above, the standy CCW pump would start on a low header pressure signal with a shaft shear on the affected pump.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per the Explanation of Required Knowledge above, the standy CCW pump will start on a low header pressure signal with the affected pump shaft sheared.
However, if the candidate determined the start signal was a result of the trip of the running pump, the standby pump would have started directly from the 52 contact prior to a low pressure condition existing.
Wednesday, February 26, 2014 9:43:51 AM                                                              2
 
C. Correct.                    The first part is correct. The annunciators came in and then cleared, informing the candidate that actions have taken place to correct the malfunction. In addition, the candidate should recognize the absence of QEAB alarms and the handswitch light indication as symptoms of a sheared shaft as opposed to a locked rotor.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Level:                            RO Tier # / Group #                  T2 / G1 K/A#                              008A1.03 Importance Rating:                2.7 / 2.9 Technical
 
==Reference:==
1X3D-BD-L01A Rev 12.0 ARP 17002-1 Rev 24.1 ARP 17036-1 Rev 21.0 References provided:              None Learning Objective:              LO-PP-10101-04 From memory, describe the expected system response and operator corrective actions for each of the following:
: a. SI
: b. LOSP
: c. SI with LOSP
: d. Surge Tank Low level
: e. Low header pressure
: f. Pump shaft shear/locked rotor
: g. Three pumps running LO-TA-60026        Respond to a Loss of CCW per 18020-C Question origin:                  MODIFIED - HL15 Audit Question # 038EA1.25 001 Cognitive Level:                  C/A 10 CFR Part 55 Content:          41.5 / 45.5 Comments:
You have completed the test!
Wednesday, February 26, 2014 9:43:51 AM                                                            3
: 1. 038EA1.25 001/1/1/SGTR - CCW AMPS/C/A - 2.6/NEW/RO/HL-14 AUDIT/TNT / RLM Initial Unit 1 conditions:
        - 19031-C, "ES-3.1 Post-SGTR Cooldown Using Backfill" is in use.
        - RHR pump A has been placed in service in shutdown cooling mode.
Current conditions:
        - Several CCW Train A low flow and pressure alarms annunciate, then clear.
        - All 3 CCW Train A pumps red lights are lit.
        - CCW Train A system has normal flows and pressures.
A CCW pump          (1)    has occurred.
You must monitor pump amps            (2)  in order to determine which pump to stop.
(1)                                (2)
A.              locked rotor                        on the QEAB B.              shaft shear                        on the QEAB C.              locked rotor                        locally at 1AA02 D.              shaft shear                        locally at 1AA02 Wednesday, January 15, 2014 10:04:30 AM                                                    1
 
Approved By                                                                                  Procedure    Version P.H. Burwinkel                      Vogtle Electric Generating Plant                      17002-1      24.1 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 02 ON PANEL            Page Number 07/27/2012                                              1A1 ON MCB                                    3 of 42 (1)                    (2)          (3)            (4)          (5)            (6)
NSCW TRAIN A          NSCW TRAIN A  NSCW TRAIN A    NSCW TRAIN A  CCW TRAIN A    CCW TRAIN A A      F-1 HI VIB            F-2 HI VIB    F-3 HI VIB      F-4 HI VIB    SURGE TK      LO HDR PRESS LO-LO LVL NSCW TRAIN A          NSCW TRAIN A                                  CCW TRAIN A  CCW TRAIN A B      LO HDR PRESS          TRANSF PMP                                    SURGE TK      LO FLOW LO DISCH PRESS                                HI/LO LVL NSCW TRAIN A                          NSCW TRAIN A    NSCW TRAIN A  CCW TRAIN A  CCW TRAIN A C      CLG TWR BASIN                        DG CLR          RHR PMP & MTR  SURGE TK      RHR HX HI/LO LVL                            LO FLOW        CLR LO FLOW    MAKE UP LVL  HI FLOW NSCW TRAIN A  NSCW INTERTIE                                CCW TRAIN A D                              CNMT CLR 1 & 2 TRN A TO TRN B                              RHR HX LO FLOW        HI FLOW                                      LO FLOW NSCW TRAIN A  NSCW TRAIN A    RMWST          CCW TRAIN A    PRIMARY E                              CNMT CLR 5 & 6 NORM/BYP VLV    VAC DEGASIFIER RHR PMP SEAL  EQUIPMENT LO FLOW        MISPOSITIONED  PNL ALARM      LO FLOW        HI TEMP NSCW TRN A RX  RX MAKE UP      RX MAKE UP F                              CVTY CLG COIL  STOR TK        STOR TK LOW FLOW      LO-LO LVL      HI/LO LVL Printed January 15, 2014 at 12:41
 
Approved By                                                                        Procedure    Version C.H. Williams                    Vogtle Electric Generating Plant                17036-1      21 Effective Date                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 36 ON EAB    Page Number 06/27/2013                                          PANEL                                  3 of 57 ALB 36 (01)                (02)              (03)            (04) 4160V SWGR              480V SWGR        480V SWGR A  1AA02 TROUBLE          1AB04 TROUBLE    1AB05 TROUBLE    SEQ A TROUBLE 4160V SWGR              480V SWGR        SEQ A PNL        SEQ A SAFETY B  1AA02 NEG PH            1AB15 TROUBLE    DOORS OPEN      EQUIP FAILED SEQ BUS PT                                                TO START 480V MCC 1ABA          480V MCC 1ABB    480V MCC 1ABC    RAT FDR BRKR C TROUBLE                  TROUBLE          TROUBLE          TRN A FAILED TO OPEN 480V MCC 1ABD          480V MCC 1ABE    480V MCC 1ABF    SEQ A IN D TROUBLE                  TROUBLE          TROUBLE          MANUAL TEST 4160/480V SWGR          BAT 1BD1B        BAT 1CD1B        BAT 1DD1B E  TRN A TRANSFER          BRKR OPEN        BRKR OPEN        BRKR OPEN SW ON LOCAL BAT 1AD1B        ISO DEVICE PNL  ISO DEVICE PNL F                          BRKR OPEN        TRN A QIP1      TRN C QIP3 TROUBLE          TROUBLE Printed January 15, 2014 at 12:43
 
Contacts for low pressure start.
Contacts for trip of the other pumps if running.                    Pressure switch activates Aux Relay
: 1. 009EK2.03 001/LOIT/RO/M/F 3.0/3.3/009EK2.03/LO-TA-37008///
Given the following:
          -  Unit 1 experienced a small break LOCA.
          -  19012-C, "Post-LOCA Cooldown and Depressurization," is in progress.
          -  RCS pressure is 1315 psig and stable.
          -  Containment pressure is 3.5 psig.
Which one of the following completes the following statement?
19012-C requires a minimum SG NR level of __(1)__,
and decay heat removal is accomplished by __(2)__.
__(1)__                                  __(2)__
A.                    10%                      natural circulation and break flow B.                    10%                                break flow ONLY C.                    32%                      natural circulation and break flow D.                    32%                                break flow ONLY K/A 009              Small Break LOCA EK2.03          Knowledge of the interrelations between the small break LOCA and the following:
S/Gs K/A MATCH ANALYSIS The question requires the candidate to recall the minimum required SG level during a small break LOCA and the reason this level is maintained. The cooling method is dependent on the SG level, which is interrelated with the specific LOCA conditions described in the stem.
EXPLANATION OF REQUIRED KNOWLEDGE Per the Westinghouse background for E.1, during a small break LOCA RCS pressure will remain elevated even though enough subcooling to allow SI termination cannot be achieved. In this event, the RCS is cooled down and RCS pressure lowered until subcooling is achieved. In this configuration, break flow alone is not sufficient to provide RCS cooling. Therefore, SG level of at least 10% NR must be maintained to Wednesday, February 26, 2014 9:46:06 AM                                                    1
 
provide a heat sink. This level ensures the SG tubes are covered on the secondary side. This level is increased to 32% NR in adverse containment conditions to compensate for instrument inaccuracies. The RCS is expected to be saturated or slightly subcooled. Therefore, SG tubes should remain covered.
ANSWER / DISTRACTOR ANALYSIS A. Correct.                  The first part is correct. Per 19012-C step 9, with containment pressure below the adverse containment value, 10% NR level is required to ensure SG tubes are covered in order to achieve adequate cooling.
The second part is correct. With RCPs stopped and a small break LOCA in progress, per Westinghouse background, break flow alone is insufficient for core cooling. Single or two phase cooling flow from the SGs is required.
B. Incorrect. Plausible. The first part is correct. See the first part of Choice A above.
The second part is incorrect. As discussed above, break flow alone is insufficient for core cooling during a small break LOCA.
However, break flow is a heat transfer method available for a larger LOCA where SG tubes are not filled. If a candidate is not able to properly assess the RCS conditions, this method could be assumed to be possible.
C. Incorrect. Plausible. The first part is incorrect. Containment pressure is <3.8 psig, which is below the adverse containment value. However, if containment pressure was above the adverse containment value, 32% NR narrow range SG level would be required.
The second part is correct. See the second part of Choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of Choice C above.
The second part is incorrect. See the second part of choice B above.
Wednesday, February 26, 2014 9:46:06 AM                                                              2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            009EK2.03 Importance Rating:              3.0 / 3.3 Technical
 
==Reference:==
E-1 Background Rev 2, 4/30/2005 19012-C Rev 33.3 References provided:            None Learning Objective:              LO-LP-37111-02 State the effect of various size breaks on the primary system with respect to temperatures and pressures.
LO-LP-34700-26 Describe the process of reflux boiling.
LO-LP-37112-01 Using EOP 19012 as a guide, briefly describe how each step is accomplished.
LO-TA-37008      Perform Post-LOCA Cooldown and Depressurization of the RCS per 19012-C Question origin:                BANK - Vogtle 2010 HL-15R NRC - Question # 009EK2.03 Cognitive Level:                M/F 10 CFR Part 55 Content:          41.7 / 45.7 Comments:
You have completed the test!
Wednesday, February 26, 2014 9:46:06 AM                                                        3
 
Approved By                                                                              Procedure Versi on J. B. Stanley                                    Vogtle Electric Generating Plant        19012-C 33.3 Effective Date                                                                            Page Number ES - 1.2 POST-LOCA COOLDOWN AND 05/01/2013                                          DEPRESSURIZATION                          8 of 43 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 7
: 7.      Check if RHR Pumps should be                      7.
stopped:
7.a
: a. RHR Pumps - ANY RUNNING                          a. Go to Step 9.
WITH SUCTION ALIGNED TO RWST.
7.b
: b. RCS pressure:                                    b. Go to Step 9.
7.b.1)
: 1)    Greater than 300 psig.                    1) 7.b.2)
: 2)    Stable or rising.                          2) 7.c
: c. Stop RHR Pumps taking suction                    c.
from the RWST.
8
        *8.      IF RCS pressure lowers in an                      8.
uncontrolled manner to less than 300 psig, THEN restart RHR pumps.
9
        *9.      Check intact SG levels:                          9.
9.a
: a. NR level - AT LEAST ONE                          a. IF all SGs NR levels less GREATER THAN 10%                                  than 10% [32% ADVERSE],
[32% ADVERSE].                                    THEN maintain total feed flow greater than 570 gpm.
9.b
: b. Maintain NR levels between 10%                    b.
[32% ADVERSE] and 65%.
9.c
: c. NR level - ANY RISING IN AN                      c. Go to Step 10.
UNCONTROLLED MANNER.
9.d
: d. Go to 19030-C, E-3 STEAM                          d.
GENERATOR TUBE RUPTURE.
 
S Printed October 25, 2013 at 13:12
 
Breaks  3/8" equivalent diameter hole Breaks in this range are considered to be leaks, rather than small LOCAs, since the normal charging system can maintain reactor coolant inventory so that RCS pressure and pressurizer level do not decrease.
Very slight system depressurization may occur but no automatic trip or safety injection signal would be generated. The core will remain fully covered provided that the steam generators are available to remove energy, and makeup flow is continuously delivered to the RCS.
If charging flow is not available, the RCS transient behavior would be similar to the response described for Category 2.
If the leak is within Technical Specification limits or it can be isolated, the plant could remain in power operation. If the leak is above Technical Specification limits and cannot be isolated, then the plant should go to a cold shutdown condition utilizing the normal shutdown procedures. During cooldown the charging system should maintain pressurizer level and the RCS depressurization should be controlled to conform to the normal cooldown limits.
Breaks 3/8" < diameter <~ 1", minimum safety injection, or Category 1 breaks above with no charging flow assumed For these break sizes the normal makeup system cannot maintain level and pressure. The RCS will depressurize and an automatic reactor trip and safety injection signal will be generated. Provided that a secondary side heat sink exists, the RCS will reach an equilibrium pressure which corresponds to the pressure at which the liquid phase break flow equals the high pressure pumped safety injection flow. It has been verified that this equilibrium pressure condition will be established for plants with charging/SI pumps. This effect is described here by the presentation of a specific plant analysis for break sizes within this range. A general description of system behavior applicable to the sample transient is provided first, then specific comments concerning the sample analysis are provided.
E1 Background                  7              HPRev. 2, 4/30/2005 HE1BG.doc
 
Early in the transient a loss of subcooled liquid in the RCS occurs which results in a moderate depressurization to the pressure which corresponds to saturation pressure in the core and hot legs. At this point the upper head, upper plenum, hot legs, and core begin to experience some slight voiding, but more than enough liquid flow exists through the core to keep it covered and cooled. During this period of voiding, however, RCS depressurization occurs at a much slower rate than during the time when the entire system was subcooled.
Eventually the RCS depressurizes to the point of the reactor trip signal. Immediately following reactor trip, the RCS rapidly depressurizes, since only a fraction of the heat previous to trip is now being transferred to the primary fluid. Due to this rapid depressurization following reactor trip, a safety injection signal is quickly generated. Within a few minutes of the reactor trip time, an equilibrium pressure is established which is above the steam generator pressure. The fluid conditions in the RCS at the time of equilibrium pressure establishment may be characterized by slight voiding in the core and upper plenum and hot legs, and saturated or slightly subcooled liquid in the cold legs. Core heat is removed through the steam generators by continuous single or twophase natural circulation.
The primary mixture level in the steam generators does not drain for breaks of this size, and the core remains covered throughout the entire transient provided that SI is not interrupted. Once equilibrium pressure is established there is no further net loss of liquid volume in the RCS. The natural circulation heat removal mode continues until the time that the break can remove all the decay heat
( 1 day for a 1" break). Prior to this time, auxiliary feedwater is required to maintain the heat sink. Since the equilibrium pressure established is determined by means of a volume balance of SI flow and break flow, the P and T from primary to secondary side, together with the cold safety injection water, may provide a total heat sink greater than the decay heat generated and a cooling of the primary fluid can occur.
E1 Background                  8              HPRev. 2, 4/30/2005 HE1BG.doc
 
This effect is evidenced by the sample transient to be described below in which the fluid in the primary becomes subcooled after a long period of time at equilibrium pressure. The RCS response would be similar to that described here regardless of the break location since the RCS will not undergo a draining, and break flow is much less than the loop flow.
Abnormal indications should be present in the containment for this category of LOCA although the response will be slower and milder than for larger break size LOCAs. For example, containment pressure will probably not reach the containment High1 pressure of approximately 10 percent of containment design pressure (45 psig).
Case A    Standard 4loop type plant, oneinch equivalent diameter break in the cold leg. The shutoff head of the SI system is 2100 psia due to spilling line flow loss. Minimum safeguards safety injection is assumed, and loss of offsite power is assumed to occur at the reactor trip time.
Therefore, the only means of venting steam on the secondary side is through the steam generator safety valves. Minimum auxiliary feedwater is assumed available one minute after the reactor trip time. The analytical model and all other analysis assumptions are in conformance with Appendix K criteria.
Case A presents a sample analysis for a break near the larger end of the break size spectrum of this behavior mode category. Figure 1 shows the RCS pressure transient for this case. The RCS pressure stabilizes slightly above the steam generator safety valve set pressure. Figures 2 and 3 show the safety injection and break flows which are both stabilized at a flowrate of approximately 45 1b/sec
(~340 gpm). Figure 4 shows that the pressurizer empties at approximately 10 minutes and does not refill. The system remains in a stable condition with the core covered and decay heat being adequately removed. In this analysis, at approximately 14 hours into the transient, hot leg flow becomes subcooled, indicating a gradual cooling trend of the RCS primary fluid. It should be noted that the subcooling of the RCS for breaks in this category would be increased substantially by increased safety injection flow, e.g., if both trains E1 Background                  9              HPRev. 2, 4/30/2005 HE1BG.doc
 
were operating.
E1 Background  10 HPRev. 2, 4/30/2005 HE1BG.doc
: 1. 010A3.02 001/LOIT AND LOCT/RO/C/A 3.6/3.5/010A3.02/LO-TA-60050///
Initial condition:
              - Unit 1 is at 100% reactor power.
Current conditions:
              - PORV-455A starts leaking by.
              - OATC observes pressurizer pressure indicating 2205 psig and slowly lowering.
Which one of the following completes the following statement?
With no operator action, the pressurizer backup heaters __(1)__ be energized, and if pressure lowers to 2185 psig __(2)__ receive a close signal.
__(1)__                                __(2)__
A.                      will                            PORVs ONLY B.                      will                    both PORVs and block valves C.                  will NOT                            PORVs ONLY D.                  will NOT                    both PORVs and block valves K/A 010              Pressurizer Pressure Control A3.02            Ability to monitor automatic operation of the PZR PCS, including:
Pressurizer pressure K/A MATCH ANALYSIS The question tests the candidate's ability to monitor automatic operation of the pressurizer pressure control circuit, and determine if the heaters, PORVs, and block valves are operating properly for the given conditions.
EXPLANATION OF REQUIRED KNOWLEDGE With PZR pressure at 2235 psig, PZR Master Controller demand is 25%. When 1PV-455B, Loop 4 Pressurizer Spray Valve, starts leaking by, PZR pressure will lower.
As pressure lowers, the demand on the PZR Master Controller will decrease resulting in increased output from the proportional heaters (spray valves are already closed). PZR Master Controller demand will continue to decrease to 9.4%, when all PZR backup Friday, February 28, 2014 1:50:37 PM                                                          1
 
heaters with handswitches in AUTO will energize. This occurs at a pressurizer pressure of 2210 psig since the system was at normal equalibrium before the tansient.
Reference 18000-C Figure 1 for a pictoral representation of the control circuit. (Note:
Figure 1 depicts a steady state circuit. As transients occur, the entire scale "slides".
Depending on the transient, functions driven out of the master controller will occur at pressures different from those depicted on Figure 1. However, the conditions of this failure have been chosen to coincide with Figure 1 to ensure predictability by the candidate.)
ALB11-D02 PRZR CONTROL LO PRESS AND HEATERS ON is driven from of the PZR Master Controller output and corresponds to 9.4% demand (2210 psig). This alarm alerts the operator that PZR backup heaters should have energized. ALB12-D03 PRZR PRESS LO PORV BLOCK alerts the operator that PZR pressure is <2185 psig and therefore, PZR Master Controller demand should be at 0%.
Per ARP 17012-C and LOGIC drawings 1X6AA02-00230 & 00494, when 2/4 pressurizer pressure transmitter sense <2185 psig, both PORVs and both Block valves recieve a direct signal to close. Additionally, PORVs require a pilot pressure to open.
With block valves closed, PORVs are incapable of opening even if a demand signal is recieved from another part of the control circuit. Block valves can be manually opened using their handswitch with pressurizer pressure <2185 psig, however the valves will close again as soon as the handswitch is released. COPS must be armed to keep the block valves open with pressurizer pressure <2185 psig.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. As described in the Explanation of Required Knowledge above, as pressurizer pressure lowers to less than 2210 psig, pressurizer backup heaters will energize.
The second part is incorrect. As described in the Explanation of Required Knowledge above, when pressurizer pressure lowers to <2185 psig on 2 of 4 transmitters, both PORVs and both block valves will receive a direct signal to close. However, a candidate with insufficient knowledge of the pressurizer control circuit would find it reasonable for ONLY the PORVs to close and block valves remain open to ensure PORV would remain available if needed for manual operation.
B. Correct.                The first part is correct. See the first part of choice A above.
The second part is correct. As described in the Explanation of Required Knowledge above, when pressurizer pressure lowers to <2185 psig on 2 of 4 transmitters, both PORVs and both block valves will receive a direct signal to close.
C. Incorrect. Plausible. The first part is incorrect. As described in the Explanation of Required Knowledge above, as pressurizer pressure lowers to less than 2210 psig, pressurizer backup heaters will energize.
However, a candidate with insufficient knowledge of the PZR pressure control circiut may not recognize the backup heater Friday, February 28, 2014 1:50:38 PM                                                                  2
 
setpoint and determine that only proportional heaters should be energized.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Level:                        RO Tier # / Group #              T2 / G1 K/A#                          010A3.02 Importance Rating:            3.6 / 3.5 Technical
 
==Reference:==
18000-C Rev 5, page 5 17011-1 Rev 16.0, pages 31 & 32 17012-C Rev 21.0, pages 23 & 24 LOGIC 1X6AA02-00230, Rev 8.0 LOGIC 1X6AA02-00235, Rev 9.0 LOGIC 1X6AA02-00236, Rev 7.0 LOGIC 1X6AA02-00494, Rev 2.0 References provided:          None Learning Objective:            LO-PP-16301-02 Describe the purpose of the following pressurizer components or auxiliaries:
: a. Variable Heaters
: b. Backup Heaters
: c. Spray Valves
: d. Bypass Spray Valves
: e. PORVs
: f. PORV Block Valves LO-PP-16303-02 Describe how the response of pressurizer pressure control to the following failures:
: e. stuck open spray valve LO-TA-60050        Respond to a stuck Open PRZR Spray valve using AOP 18000-C Question origin:              NEW Cognitive Level:              C/A 10 CFR Part 55 Content:        41.7 / 45.5 Comments:
You have completed the test!
Friday, February 28, 2014 1:50:38 PM                                                                3
 
Approved By                                                              Procedure Number Rev C. S. Waldrup                        Vogtle Electric Generating Plant  18000-C          5 Date Approved                                                            Page Number PRESSURIZER SPRAY, SAFETY, OR RELIEF 2/3/09                                    VALVE MALFUNCTION                      5 of 5 FIGURE 1              Sheet 1 of 1 PRESSURIZER PRESSURE CONTROLLER BAND Printed October 28, 2013 at 10:09
 
Approved By                                                                                    Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                          17011-1      16 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 11 ON PANEL              Page Number 07/4/2011                                            1C1 ON MCB                                      31 of 52 WINDOW D02 ORIGIN                            SETPOINT PRZR CONTROL 1-PIC-0445A                        2210 psig                    LO PRESS AND output from                                                    HEATERS ON selected channel:
1-PT-0455 OR 1-PT-0457 1.0                PROBABLE CAUSE
: 1.        Pressurizer Pressure Control System malfunction.
: 2.        Pressurizer Spray or Relief Valve malfunction.
2.0                AUTOMATIC ACTIONS Pressurize Backup Heaters will energize.
3.0                INITIAL OPERATOR ACTIONS Check pressurizer pressure indications:
IF an instrument failure is indicated, initiate 18001-C, "Primary Systems Instrumentation Malfunction".
IF a failed PRZR Spray Valve, Safety Valve or PORV is indicated, initiate 18000-C Pressurizer Spray, Safety Or Relief Valve Malfunction.
AT 1965 psig and lowering trip RX and go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.
Printed January 15, 2014 at 16:02
 
Approved By                                                                                  Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                          17011-1      16 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 11 ON PANEL              Page Number 07/4/2011                                            1C1 ON MCB                                    32 of 52 WINDOW D02 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS NOTE A large load increase or transient may cause pressurizer pressure to decrease temporarily below the alarm setpoint.
: 1.        IF no instrument failure is indicated and no other reason for pressure decrease is evident, initiate 18004-C, "Reactor Coolant System Leakage".
: 2.        Refer to Technical Specification LCO 3.4.1.
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB112, 1X6AX01-106, 1X6AU01-182, 168, PLS RER 92-190, 93-213 Printed January 15, 2014 at 16:02
 
Approved By                                                                                  Procedure    Version W.R. Dunn                          Vogtle Electric Generating Plant                        17012-1    21 Effective Date                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON                  Page Number 05/06/2013                                        PANEL 1C1 ON MCB                                  23 of 52 WINDOW D03 ORIGIN                            SETPOINT PRZR PRESS Any 2 of the                      2185 psig                  LO PORV following:                                                    BLOCK 1-PT-0455 1-PT-0456 1-PT-0457 1-PT-0458 1.0                PROBABLE CAUSE
: 1.        RCS pressure transient during plant startup or shutdown.
: 2.        RCS Pressure Control Malfunction.
2.0                AUTOMATIC ACTIONS PRZR PORV 455A and 456A BLOCK VALVES 1-HV-8000A and 1-HV-8000B are blocked from opening, and will close if open with their handswitches in the auto position.
3.0                INITIAL OPERATOR ACTIONS NONE 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        Verify Block Valves are closed.
: 2.        Verify Pressurizer Pressure Control System is responding to restore pressure.
: 3.        IF a failure has occurred in the Pressurizer Pressure Control System, refer to 18001-C, "Primary Systems Instrumentation Malfunction."
: 4.        IF RCS pressure decrease is due to excessive steam demand, check Steam Dumps and Atmospheric Relief Valves closed.
: 5.        IF equipment failure is indicated, initiate maintenance as required.
Printed January 15, 2014 at 16:01
 
Approved By                                                                    Procedure    Version W.R. Dunn                        Vogtle Electric Generating Plant            17012-1    21 Effective Date                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON    Page Number 05/06/2013                                  PANEL 1C1 ON MCB                          24 of 52 WINDOW D03 (Continued) 5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB112, 1X6AA02-230, 1X6AU01-168; PLS Printed January 15, 2014 at 16:01
 
2185 PSIG interlock
 
2210 psig to turn on all heaters
 
2210 psig signal 2185 PSIG Loss of interlock results in a CLOSE  PORV demand for both the PORV and Block valve Block Vlv
: 1. 010K6.04 001/LOIT/RO/C/A 2.9/3.2/010K6.04/LO-PP-16301-06//HL-15 NRC/
Given the following Unit 1 conditions:
          - 1PV-455A, Pressurizer PORV, is stuck slightly open.
          - 1HV-8000A, PORV Block Valve, will not close.
          - 1PV-455A tail pipe temperature is reading ~280oF.
          - RCS pressure is stable at 1920 psig.
          - Pressurizer vapor space temperature is 630oF.
          - PRT pressure is 35 psig and slowly rising.
          - Containment pressure is 0 psig.
Which one of the following completes the following statement?
The 1PV-455A tail pipe temperature indication __(1)__ reading correctly, and with 1PV-455A partially open, its tail pipe temperature will rise to approximately
__(2)__.
REFERENCE PROVIDED
__(1)__                              __(2)__
A.                is                      630oF and then stabilize B.                is                338oF and then lower to ~212oF C.            is NOT                    630oF and then stabilize D.            is NOT                338oF and then lower to ~212oF K/A 010              Pressurizer Pressure Control K6.04            Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS:
                        - PRT K/A MATCH ANALYSIS This question requires the candidate to evaluate the tailpipe temperature of a stuck open PORV. The candidate is then required to determine the effect the rupture of the PRT rupture disk would have on the PORV tailpipe temperature.
EXPLANATION OF REQUIRED KNOWLEDGE Wednesday, February 26, 2014 12:29:07 PM                                                    1
 
The candidate is required to convert PSIG to PSIA for use on a REFERENCE PROVIDED steam table/Mollier Diagram and compare the temperature of the current pressurizer and PRT conditions against the indicated tailpipe temperatures. Then, the candidate is required to recognize that as the event progresses, PRT pressure will slowly rise until the PRT rupture disk fails at 100 psig. At this time, PRT tailpipe discharge pressure will quickly lower to Containment Pressure.
Per the steam stables, 1935 psia and 630F correspond to a saturated system with an enthalpy of approximately 1140. With a PRT pressure of 50 psia, enthalpy of 1140 is saturated with a temperature of 281F. Therefore the indication is correct. As PRT pressure rises to approximately 115 psia, the system remains saturated and temperature will rise to 338F. As soon as the disk ruptures, pressure will quickly drop to 15 psia and remains saturated with a temperature of 212F.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. Per the Explanation of Required Knwoledge above, the given plant conditions will result in a tailpipe temperature of 281F.
The second part is incorrect. As discussed in the Explanation of Required Knowledge, the tailpipe temperature will rise to 338F and then quickly lower to 212F when the PRT ruptures.
If the candidate does not understand the characteristics of the throttling process, it is reasonable to assume tailpipe temperature would eventually rise to equal the RCS temperature and remain there.
B. Correct.                    The first part is correct. See the first part of choice A above.
The second part is correct. Per the Explanation of Required Knwoledge above, the given plant conditions will result in PRT temperature rising to 338F and then lowering to 212F after the rupture disk blows.
C. Incorrect. Plausible. The first part is incorrect. As discussed in the Explanation of Required Knowledge above, the tailpipe temperature is expected to be 281F for the conditions listed. However, a candidate that does not understand throttling processes would find it reasonable for tailpipe temperature to be much higher than 281F.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 12:29:07 PM                                                              2
 
Level:                            RO Tier # / Group #                  T2 / G1 K/A#                              010K6.04 Importance Rating:                2.9 /3.2 Technical
 
==Reference:==
2000 ASME Steam Tables References provided:              2000 ASME Steam Tables Learning Objective:              LO-LP-34300-22 Explain the reduction of process pressure from throttling using an enthalpy-entropy (h-s) diagram or temperature-entropy (T-s) diagram.
LO-PP-16301-06 Determine the expected tail pipe temperature for a leaking or open PORV.
Question origin:                  MODIFIED - HL15 NRC Question # 007K1.01 Cognitive Level:                  C/A 10 CFR Part 55 Content:          41.7 / 45.7 Comments:
You have completed the test!
Wednesday, February 26, 2014 12:29:07 PM                                                        3
: 1. 011EK2.02 001/LOIT/RO/C/A 2.6/2.7/011EK2.02/LO-TA-37009///
Initial condition:
            - Unit 1 experienced a reactor trip and SI.
Current conditions:
            - The OATC aligns ECCS for Cold Leg Recirculation.
            - RCS pressure is 1450 psig.
            - 19010-C, "Loss of Reactor or Secondary Coolant," is in progress.
Which one of the following completes the following statements?
Prior to the LOCA, each RHR pump is aligned for injection into __(1)__ Cold Legs.
Following alignment to Cold Leg Recirculation, ECCS injection flow to the core is being provided by __(2)__.
__(1)__                                  __(2)__
A.                      2                                  CCPs ONLY B.                      2                                CCPs and SIPs C.                      4                                  CCPs ONLY D.                      4                                CCPs and SIPs K/A 011              Large Break LOCA EK2.02          Knowledge of the interrelations between the Pumps and the Large Break LOCA.
K/A MATCH ANALYSIS The question requires the candidate to recall how many Cold Legs each RHR pump is capable of injecting into in standby alignment. Additionally, the candidate is required to determine which ECCS pumps will be injecting water into the core during the LOCA based on plant conditions.
EXPLANATION OF REQUIRED KNOWLEDGE In standby alignment, each RHR pump is capable of injecting into 4 Cold Legs. The individual RHR pumps inject into 2 cold legs. However, the cross-connect valves HV8716A&B are open which allows each pump to inject into the other train cold legs also. During cold leg recirculation, HV-8716A&B are closed to prevent pump runout.
Reference 1X4DB121 and 1X4DB122.
Wednesday, February 26, 2014 2:17:08 PM                                                        1
 
The approximate shutoff head of the ECCS pumps are as follows:
RHR - 300 psig SI - 1625 psig CCPs - 2700 psig With RCS pressure at 1450 psig, RHR will be operating on mini-flow and the SI pumps and the CCPs will be injecting into the core.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Each RHR pump is capable of injecting into all 4 cold legs. However, each RHR feeds only two cold legs and the other train cold legs are fed via the cross-connect valves. A candidate with insufficient knowledge of RHR standby alignment who assumes HV-8716A&B are CLOSED will believe this answer correct.
The second part is incorrect. With RCS pressure at 1450 psig, the SIPs are operating below shutoff head and are injecting into the core along with the CCPs. A candidate unfamiliar with shutoff head pressure for an SIP may assume that the SIP is not injecting at this pressure.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. With RCS pressure at 1450 psig, the SIPs are operating below shutoff head and are injecting into the core along with the CCPs.
C. Incorrect. Plausible. The first part is correct. Each RHR pump is capable of injecting into all 4 cold legs - 2 cold legs from it's own piping and 2 cold legs via the cross-connect valves.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 2:17:08 PM                                                                2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            011EK2.02 Importance Rating:              2.6 / 2.7 Technical
 
==Reference:==
EOP 19013-C Rev 29.2, page 5 19010-C Rev 34.3 1X4DB121, Rev 42.0 1X4DB121, Rev 51.0 References provided:            None Learning Objective:              LO-PP-13101-06 State the following parameters for the ECCS pumps:
: a. shutoff head (pressure)
: b. rated flow (gpm)
: c. design features to prevent runout and provide pump miniflow LO-PP-13101-08 Explain the operation of ECCS in each of the three modes of operation.
LO-PP-13101-09 Describe the normal standby alignment of the ECCS.
LO-TA-13009      Manually align ECCS for Cold Leg Recirculation Phase using EOP 19013-C.
LO-TA-37009      Respond to a Large Break Loss of Primary Coolant per 19010-C.
Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 / 45.7 Comments:
You have completed the test!
Wednesday, February 26, 2014 2:17:08 PM                                                          3
 
Approved By                                                                        Procedure    Version J.B. STANLEY                          Vogtle Electric Generating Plant            19013-C      29.2 Effective Date                                                                      Page Number ES-1.3 TRANSFER TO COLD LEG 7/25/12                                      RECIRCULATION                              5 of 20 ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED 3
: 3.      Initiate ATTACHMENT A to align              3.
ECCS Pumps to the Cold Leg Recirculation flowpath and continue with Step 4.
4
: 4.      Notify Health Physics that radiation        4.
levels in the Auxiliary Building will change when Cold Leg Recirculation is established.
5
: 5.      Make a page announcement to clear            5.
personnel from the Auxiliary Building prior to initiating Cold Leg Recirculation.
6
: 6.      Initiate Continuous Actions Page.            6.
7
        *7.      Check RWST level - GREATER                  *7. Stop any ECCS Pumps taking THAN 8%.                                          suction from the RWST.
8
        *8.      Check if SI pumps should be                  8.
stopped.
8.a
: a. RCS pressure - GREATER                        a. IF RCS pressure rises to THAN 1625 PSIG.                                greater than 1625 psig, THEN stop SI Pumps.
Go To Step 9.
8.b
: b. Stop SI Pumps.                                b.
9
: 9.      Check ATTACHMENT A -                        9. Do NOT continue with this COMPLETE.                                        procedure until ATTACHMENT A has been COMPLETED.
 
S Printed January 28, 2014 at 13:15
: 1. 011K6.03 001/LOIT AND LOCT/RO/C/A 2.9/3.3/011K6.03/LO-TA-60030///
Initial conditions:
          -  Unit 1 is at 50% reactor power.
          -  RCS pressure is 2235 psig.
          -  Pressurizer level control is selected to CH 459 / 461.
          -  All pressurizer heaters are energized.
Current condition:
          - Pressurizer level transmitter, 1LT-461, fails LOW.
          - NO operator action has been taken.
Which one of the following completes the following statement?
1LV-460, Letdown Isolation Valve, __(1)__ close, and the pressurizer backup heaters are __(2)__.
__(1)__                                  __(2)__
A.            will                                  energized B.            will                                de-energized C.          will NOT                                  energized D.          will NOT                                de-energized K/A 011              Pressurizer Level Control K6.03            Knowledge of the effect of a loss or malfunction on the following will have on the PZR LCS:
                        - Relationship between PZR level and PZR heater control circuit.
K/A MATCH ANALYSIS The question tests the candidates knowledge of the impact on the pressurizer level control system from a control circuit failure - PZR level. From this the candidate must predict PZR heater system and spray valve response.
EXPLANATION OF REQUIRED KNOWLEDGE Wednesday, February 26, 2014 2:19:48 PM                                                        1
 
Pressurizer level control uses primary and secondary control circuits. The level transmitters that feed the control circuits are selectable - as the question is written, LT-459 is primary and LT-461 is secondary. The primary channel uses the level input to control pressurizer level. There are interlocks with the pressurizer pressure control circuit to energize heaters on a 5% level deviation and to de-energize heaters and isolate letdown if level is <17%. The secondary channel is a backup to the primary and will also de-energize heaters and isolate letdown if level is <17%. The primary channel closes LV-459 and letdown orifice valves. The secondary channel closes LV-460 and letdown orifice valves.
If LT-461 (secondary) fails low, heaters are de-energized and letdown isolates. As pressurizer level increases due to loss of letdown, the primary channel will sense a level rise on LT-459 and lower charging flow. With charging in automatic, FIC-121 will close until charging flow is 42 gpm. The control circuit is limited to ensure minimum seal injection flow can be maintained with some flow remaining through the charging nozzle to prevent thermal transients on the nozzle welds (42 gpm total charging flow with 32 gpm seal injection flow and 10 gpm flow through the charging nozzle).
Pressurizer level control also utilizes auctioneered high Tavg to provide a program level. Pressurizer level program is varied from 25% to 60% from 557F to 588.4F.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. With 459/461 selected for control, LT-461 is the input for the secondary channel. When the secondary channel indicates <17%, letdown will isolate. LV-460 will close and LV-459 will remain open. The letdown orifice isolation valves will close.
The second part is incorrect. With 459/461 selected for control, LT-461 is the input for the secondary channel. When the secondary channel indicates <17%, pressurizer heaters de-energize. However, pressurizer pressure control operates the pressuirzer heaters only from the primary channel. A candidate with insufficient knowledge of the PRZ level control circuit may assume only the primary channel controls the pressurizer heaters. As such, heaters would assume to be unaffected by LV-461. Therefore, this distractor is plausible.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. With 459/461 selected for control, LT-461 is the input for the secondary channel. When the secondary channel indicates <17%, pressurizer heaters de-energize.
C. Incorrect. Plausible. The first part is incorrect. With 459/461 selected for control, LT-461 is the input for the secondary channel. When the secondary channel indicates <17%, letdown will isolate. LV-460 will close and LV-459 will remain open. However, the primary channel is always an odd numbered instrument. Since LT-462 is calibrated for cold-cal use, LT-461 is used for both primary Wednesday, February 26, 2014 2:19:48 PM                                                              2
 
and secondary control functions depending on selector switch position. As such, it is possible for a candidate to assume that LV-459 and LV-460 are controlled by LT-459 and LT-460 and not the selected primary and secondary circuits and conclude that LV-460 would remain open.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Level:                          RO Tier # / Group #                T2 / G2 K/A#                            011K6.03 Importance Rating:              2.9 / 3.3 Technical
 
==Reference:==
ARP 17011-1 Rev 16, pages 14&15 LOGIC 1X6AA02-00236, Rev 7.0 LOGIC 1X6AA02-00496, Rev 1.0 References provided:            None Learning Objective:              LO-PP-16302-02 Describe how the response of pressurizer level control to the following failures:
: a. controlling (primary & secondary) channel fails low
: b. controlling (primary & secondary) channel fails high
: c. controller high or low failure
: d. controlling channel fails as is LO-TA-60030          Respond to a Failure of Pressurizer Level Instrumentation per 18001-C Question origin:                BANK Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 /45.7 Comments:
You have completed the test!
Wednesday, February 26, 2014 2:19:48 PM                                                              3
 
Approved By                                                                                Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                      17011-1      16 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 11 ON PANEL          Page Number 07/4/2011                                          1C1 ON MCB                                  14 of 52 WINDOW B01 ORIGIN                            SETPOINT PRZR LO LEVEL 1-LT-0459                        17%                        HTR CNTL OFF 1-LT-0460                                                    LTDN SECURED 1-LT-0461 1.0                PROBABLE CAUSE
: 1.        Pressurizer level Control System Malfunction.
: 2.        Charging - Letdown System Malfunction.
: 3.        RCS cooldown.
: 4.        Reactor Coolant System leak.
2.0                AUTOMATIC ACTIONS
: 1.        All Pressurizer Heaters turn off.
: 2.        Letdown isolation.
3.0                INITIAL OPERATOR ACTIONS
: 1.        Check pressurizer level instrumentation.
: 2.        IF instrument malfunction is indicated Go To 18001-C, "Primary Systems Instrumentation Malfunctions".
: 3.        IF a Reactor Coolant System leak is indicated, Go To 18004-C, "Reactor Coolant System Leakage".
Printed January 31, 2014 at 15:48
 
Approved By                                                                                Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                        17011-1      16 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 11 ON PANEL            Page Number 07/4/2011                                            1C1 ON MCB                                  15 of 52 WINDOW B01 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        IF actual pressurizer level is low,
: a. Verify level is returning to normal, and
: b. IF necessary, verify Charging Line Flow Control Valve is open and start additional Charging Pumps as needed to increase level to normal.
: 2.        After alarm clears, restore heaters and letdown in accordance with 13006-1, "CVCS Startup And Normal Operation".
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB112, 116-1, 1X6AU01-167, 183, PLS Printed January 31, 2014 at 15:48
 
Closes LV-460 Turns heates off
: 1. 012K5.01 001/LOIT/RO/M/F 3.3/3.8/012K5.01/LO-TA-60014///
Given the following:
          - Unit 1 is at 100% reactor power.
Which one of the following completes the following statement?
The __(1)__ Reactor Trip protects against Departure from Nucleate Boiling (DNB),
and the Trip Setpoint __(2)__ in response to lowering pressurizer pressure.
__(1)__                                  __(2)__
A.                OT delta T                            remains unchanged B.                OT delta T                                decreases C.                OP delta T                            remains unchanged D.                OP delta T                                decreases K/A 001              Reactor Protection K5.01            Knowledge of the operational implications of the following concepts as the apply to the RPS:
                        - DNB K/A MATCH ANALYSIS The question requires the candidate to understand the operational impact of lowering pressurizer pressure on Overtemperature Delta-T (OTdT) and the interrelationship between this reactor trip setpoint and DNB.
EXPLANATION OF REQUIRED KNOWLEDGE Reactor trip for Overtemperature Delta-T (OTdT) utilizes narrow range RCS Tavg, pressurizer pressure, and delta-NI power to calculate a reactor trip setpoint, which is compared to the actual delta-T for each RCS loop. As RCS pressure lowers, subcooling lowers and the RCS approaches saturation and gets closer to DNB.
Therefore, the OTdT setpoint lowers, getting closer to actual delta-T. If 2/4 loop OTdT setpoints are exceeded, an automatic reactor trip is initiated by SSPS. The calculation for the OTdT setpoint is listed in TS 3.3.1, Table 3.3.1-1 page 7 of 9. Candidates are required to understand the general effect of RCS parameter changes on the OTdT setpoint. Candidates are NOT required to calculate the setpoints. Per TS Bases for Wednesday, February 26, 2014 2:21:25 PM                                                        1
 
3.3.1 Fu 6, the OTdT trip function is provided to ensure that the design limit DNBR is met.
Correspondingly, the Overpower Delta T (OPdT) reactor trip utilizes narrow range Tavg and the rate of change of Tavg to determine the setpoint (Delta-NI is nulled out). This trip protects against overpower in the fuel (Kw/ft). These two trips are routinely confused by candidates because the names are so similar. The trip bases and setpoint inputs are commonly jumbled and mismatched by candidates.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. Per TS Bases 3.3.1 Fu 6, the Overtemperature Delta T trip is designed to ensure DNBR is met.
The second part is incorrect. As RCS pressure lowers, subcooling lowers and the RCS approaches saturation and gets closer to DNB. Therefore, the OTdT setpoint lowers.
However, candidates routinely confuse OPdT and OTdT. If the candidate did confuse OPdT and OTdT, the candidate would not expect the setpoint to change because pressure is not an input to OPdT.
B. Correct.                    The first part is correct. See the first part of choice A above.
The second part is correct. As RCS pressure lowers, subcooling lowers and the RCS approaches saturation and gets closer to DNB. Therefore, the OTdT setpoint lowers.
C. Incorrect. Plausible. The first part is incorrect. Per TS Bases 3.3.1 Fu 6, the Overtemperature Delta T trip is designed to ensure DNBR is met. However, candidates routinely confuse OTdT and OPdT (which protects from overpower in the fuel (Kw/ft)).
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 2:21:26 PM                                                                2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            012K5.01 Importance Rating:              3.3 / 3.8 Technical
 
==Reference:==
TS 3.3.1, Ammendment No. 165, pages 3.3.1-20&21 TS Bases 3.3.1, Rev 1-8/02, pages B 3.3.1-16 thru 20 References provided:            None Learning Objective:              LO-PP-16303-03 State the reactor trips and SI actuation signals, including set points and coincidences associated with pressurizer pressure protection channels.
LO-PP-28103-03 List all reactor trip set points, coincidences, permissives, and blocks.
LO-PP-28103-04 Discuss the bases for each reactor trip signal.
LO-PP-56101-19 Describe what happens on a reactor trip and what is being protected by the trip.
LO-TA-60014      Respond to Reactor Coolant System Leakage per 18004-C LO-TA-60029      Respond to a Failure of Pressurizer Pressure Instrumentation per 18001-C Question origin:                BANK Cognitive Level:                M/F 10 CFR Part 55 Content:          41.5 / 45.7 Comments:
You have completed the test!
Wednesday, February 26, 2014 2:21:26 PM                                                          3
 
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 9)
Reactor Trip System Instrumentation Note 1: Overtemperature Delta-T The Allowable Value of each input to the Overtemperature Delta-T function as defined by the equation below shall not exceed its as-left value by more than the following:
(1)    0.5% T span for the T channel (2)    0.5% T span for the Tavg channel (3)    0.5% T span for the pressurizer pressure channel (4)    0.5% T span for the f1(AFD) channel T    {1 + 1s}        1      &#xba;            {1 +  4 s}    1            &#xba;  (p)                          &#xba;
                      <<100                            >>  << K1 - K 2            << T          - T >>      - K 3{P - P} - f 1 (AFD) >>
                      &#xac; T0  {1 +  2 s} {1 + 3 s} 1/4  &#xac;          {1 + 5 s} &#xac; {1 +  6 s}      1/4                                1/4 Where:        T          measured loop specific RCS differential temperature, degrees F T0        indicated loop specific RCS differential at RTP, degrees F 1+1s      lead-lag compensator on measured differential temperature 1+2s 1, 2      time constants utilized in lead-lag compensator for differential temperature: 1 = 0 seconds, 2 = 0 seconds 1
1+3s      lag compensator on measured differential temperature 3          time constant utilized in lag compensator for differential temperature,  6 seconds K1          fundamental setpoint,  114.9% RTP K2          modifier for temperature, = 2.24% RTP per degree F 1+4s 1+5s      lead-lag compensator on dynamic temperature compensation 4, 5      time constants utilized in lead-lag compensator for temperature compensation: 4  28 seconds, 5  4 seconds T          measured loop specific RCS average temperature, degrees F 1
1+6s      lag compensator on measured average temperature 6          time constant utilized in lag compensator for average temperature,  6 seconds T          indicated loop specific RCS average temperature at RTP,  588.4 degrees F K3          modifier for pressure, = 0.177% RTP per psig P          measured RCS pressurizer pressure, psig P          reference pressure,  2235 psig s          Laplace transform variable, inverse seconds Vogtle Units 1 and 2                                            3.3.1-20                            Amendment No. 165 (Unit 1)
Amendment No. 147 (Unit 2)
 
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of 9)
Reactor Trip System Instrumentation Note 1: Overtemperature Delta-T (continued) f1(AFD)      modifier for Axial Flux Difference (AFD):
: 1. for AFD between -23% and +10%, = 0% RTP
: 2. for each % AFD is below -23%, the trip setpoint shall be reduced by 3.3% RTP
: 3. for each % AFD is above +10%, the trip setpoint shall be reduced by 1.95% RTP
{1 +  s}
4            1          &#xba; (p) The compensated temperature difference                      <<T {1 + s}    - T >> shall be no more negative than 3 degrees F.
{1 + 5 s}    &#xac;        6        1/4 Note 2: Overpower Delta-T The Allowable Value of each input to the Overpower Delta-T function as defined by the equation below shall not exceed its as-left value by more than the following:
(1)  0.5% T span for the T channel (2)  0.5% T span for the Tavg channel T    {1 + 1s}      1      &#xba;                { s}        1      &#xba;            1        &#xba;            &#xba;
                    <<100 T                          >>  << K 4 - <<K 5      7 T >> - K 6 <<T          - T>> - f 2 (AFD) >>
                    &#xac;        {1 +  2 s} {1 +  3 s} 1/4 <<        <<    {1 +  s}  {1 +  s}    >>1/4      <<&#xac; {1 +  }      >>1/4            >>1/4 0                            &#xac;      &#xac;          7          6                      6 Where:      T            measured loop specific RCS differential temperature, degrees F T0          indicated loop specific RCS differential at RTP, degrees F 1+1s        lead-lag compensator on measured differential temperature 1+2s 1, 2        time constants utilized in lead-lag compensator for differential temperature: 1 = 0 seconds, 2 = 0 seconds 1
1+3s        lag compensator on measured differential temperature 3            time constant utilized in lag compensator for differential temperature,  6 seconds K4            fundamental setpoint,  110% RTP K5            modifier for temperature change:  2% RTP per degree F for increasing temperature,  0% RTP per degree F for decreasing temperature 7s 1+7s        rate-lag compensator on dynamic temperature compensation 7            time constant utilized in rate-lag compensator for temperature compensation,  10 seconds T            measured loop specific RCS average temperature, degrees F 1
1+6s        lag compensator on measured average temperature Vogtle Units 1 and 2                                              3.3.1-21                        Amendment No. 165 (Unit 1)
Amendment No. 147 (Unit 2)
 
RTS Instrumentation B 3.3.1 BASES APPLICABLE          7. Overpower 'T SAFETY ANALYSES, LCO, and                The Overpower 'T trip Function (TDI-0411B, TDI-0421B, APPLICABILITY          TDI-0431B, TDI-0441B, TDI-0411A, TDI-0421A, TDI-0431A, (continued)          TDI-0441A) 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 Flux  High Setpoint trip. The Overpower 'T trip Function ensures that the allowable heat 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:
x    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 x    rate of change of reactor coolant average temperature including dynamic compensation for RTD response time delays.
The Overpower 'T trip Function is calculated for each loop as per Note 2 of Table 3.3.1-1. Trip occurs if Overpower 'T is indicated in two loops. Since the temperature signals are used for other control 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 remaining channels providing the protection function actuation. This results in a two-out-of-four trip logic. Section 7.2.2.3 of Reference 1 discusses control and protection system interactions for this function. Note that this Function also provides a signal to generate a turbine runback prior to reaching the Allowable Value. 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.
(continued)
Vogtle Units 1 and 2                  B 3.3.1-20                                Revision No. 0
 
RTS Instrumentation B 3.3.1 BASES APPLICABLE          5. Source Range Neutron Flux (continued)
SAFETY ANALYSES, LCO, and                subcritical, boron dilution (see LCO 3.3.8) and control APPLICABILITY          rod ejection events. The Function also provides visual neutron flux indication in the control room.
In MODE 2 when below the P-6 setpoint during a reactor startup, the Source Range Neutron Flux trip must be OPERABLE. Above the P-6 setpoint, the Intermediate Range Neutron Flux trip and the Power Range Neutron Flux  Low Setpoint trip will provide core protection for reactivity accidents.
Above the P-6 setpoint, the Source Range Neutron Flux trip is blocked.
In MODE 3, 4, or 5 with the reactor shut down, the Source Range Neutron Flux trip Function must also be OPERABLE. If the Rod Control 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 Rod Control System is not capable of rod withdrawal, the source range detectors are not required to trip the reactor. Source range detectors also function to monitor for high flux at shutdown. This function is addressed in Specification 3.3.8. Requirements for the source range detectors in MODE 6 are addressed in LCO 3.9.3.
: 6. Overtemperature T The Overtemperature 'T trip Function (TDI-0411C, TDI-0421C, TDI-0431C, TDI-0441C, TDI-0411A, TDI-0421A, TDI-0431A, TDI-0441A) 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 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 (continued)
Vogtle Units 1 and 2                B 3.3.1-16                                  Rev. 1-3/99
 
RTS Instrumentation B 3.3.1 BASES APPLICABLE          6. Overtemperature 'T (continued)
SAFETY ANALYSES, LCO, and                has the same effect on 'T as a power increase. The APPLICABILITY          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:
x    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; x    pressurizer pressure  the Trip Setpoint is varied to correct for changes in system pressure; and x    axial power distribution  f(AFD)x, the f(AFD) Function is used in the calculation of the Overtemperature 'T trip. It is a function of the indicated difference between the upper and lower NIS power range detectors. This Function measures the axial power distribution. The Overtemperature 'T 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 RTD response time delays.
The Overtemperature 'T trip Function is calculated for each loop as described in Note 1 of Table 3.3.1-1. A trip occurs if Overtemperature 'T is indicated in two loops. Since the pressure and temperature signals are used for other control 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.
(continued)
Vogtle Units 1 and 2                B 3.3.1-17                                  Rev. 1-8/02
 
RTS Instrumentation B 3.3.1 BASES APPLICABLE          6. Overtemperature 'T (continued)
SAFETY ANALYSES, LCO, and                This results in a two-out-of-four trip logic. Section 7.2.2.3 of APPLICABILITY          Reference 1 discusses control and protection system interactions for this function. 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 Overtemperature 'T condition and may prevent a reactor trip.
Delta-T0, as used in the overtemperature and overpower 'T trips, represents the 100% RTP value as measured for each loop. This normalizes each loop's 'T trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses. These differences in RCS loop 'T can be due to several factors, e.g., differences in RCS loop flows 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 values.
Therefore, loop specific 'T0 values are measured as needed to ensure they represent actual core conditions.
The parameter K1 is the principal setpoint gain, since it defines the function offset. The parameters K2 and K3 define the temperature gain and pressure gain, respectively. The values for T' and P' are key reference parameters corresponding directly to plant safety analyses initial conditions assumptions for the Overtemperature 'T function. For the purposes of performing a CHANNEL CALIBRATION, the values for K1, K2, K3, T', and P' are utilized in the safety analyses without explicit tolerances, but should be considered as nominal values for instrument settings. That is, while an exact setting is not expected, a setting as close as reasonably possible is desired.
Note that for T', the value for the hottest RCS loop will be set (continued)
Vogtle Units 1 and 2                B 3.3.1-18                                  Rev. 1-6/98
 
RTS Instrumentation B 3.3.1 BASES APPLICABLE          6. Overtemperature 'T (continued)
SAFETY ANALYSES, LCO, and                as close as possible to 588.4&#xba; F. The instrument uncertainty APPLICABLITY            calculations and safety analyses, in combination, have accounted for loop variation in loop specific, full power, indicated 'T and Tavg. With respect to Tavg, a value for T' common to all four loops is permissible within the limits identified in the uncertainty calculations. Outside of those limits, the value of T' will be set appropriately to reflect indicated, loop specific, full power values. In the case of decreasing temperature, the compensated temperature difference shall be no more negative than 3 &#xba;F to limit the increase in the setpoint during cooldown transients. The engineering scaling calculations use each of the referenced parameters as an exact gain or reference value. Tolerances are not applied to the individual gain or reference parameters.
Tolerances are applied to each calibration module and the overall string calibration. In order to ensure that the Overtemperature 'T instrument channel is performing in a manner consistent with the assumptions of the safety analyses, it is necessary to verify during the CHANNEL OPERATIONAL TEST that the magnitude of instrument drift from the as-left condition is within limits, and that the input parameters to the trip function are within the appropriate calibration tolerances for the defined calibration conditions (Ref. 7).
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.
(continued)
Vogtle Units 1 and 2                B 3.3.1-19                                    Rev. 4-9/06
: 1. 013K6.01 001/LOIT/RO/M/F 2.7/3.1/013K6.01/LO-TA-28013//HL18 NRC/016K1.09 Initial condition:
          - Unit 1 is at 100% reactor power.
Current conditions:
          - The bistable(s) for Containment Pressure Channel I (1PT-937) are de-energized.
          - No Tech Spec actions have been taken.
Which one of the following completes the following statement?
The MINIMUM number of ADDITIONAL channels required to initiate an actuation signal on High-1 is __(1)__,
and the MINIMUM number of ADDITIONAL channels required to initiate an actuation signal on High-3 is __(2)__.
__(1)__                                  __(2)__
A.                      1                                        1 B.                      1                                        2 C.                      2                                        1 D.                      2                                        2 K/A 013            Engineered Safety Features Actuation System (ESFAS)
K6.01          Knowledge of the effect of a loss or malfunction of the following will have on the ESFAS:
                        - Sensors and detectors K/A MATCH ANALYSIS:
The question addresses a de-energized Containment pressure channel. The candidate must determine how many of the remaining OPERABLE channels are required for High-1 and High-3 actuations to occur.
EXPLANATION OF REQUIRED KNOWLEDGE ESFAS actuations associated with Containment Pressure are HI-1 for Safety Injection, HI-2 for Steam Line Isolation, and HI-3 for Containment Spray. HI-1 and HI-2 are Wednesday, February 26, 2014 2:23:01 PM                                                    1
 
de-energize to actuation bistable. HI-3 is an energize to actuate bistable. HI-1 and HI-2 utilize a 2 of 3 logic utilizing Channel 2, 3, & 4 only. HI-3 is a 2 of 4 logic.
With Channel 1 PT-937 bistables de-energized, neither HI-1 nor HI-3 have any bistables tripped, HI-1 because Channel 1 is not used and HI-3 because they are energize to actuate bistables. Therefore, both HI-1 and HI-3 would require two additional bistables to trip before an actuation would occur.
The bullet stating "No Tech Spec actions have been taken" is required because Tech Spec actions would alter the state of the HI-3 Channel 1 bistable.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Channel 1 is not utilized for HI-1, and two channels are required for an actuation to occur. However, a candidate who does not have specific knowledge of which three containment pressure channels are utilized may find it reasonable for channels 1, 2, & 3 to be used. In that case, only one additional channel is required.
The second part is incorrect. HI-3 bistables are energize to actuate and two channels are required for an actuation to occur.
However, a candidate without specific knowledge of the bistable behavior may think that the bistables are de-energize to actuate and determine that only one additional channel is required.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. HI-3 bistables are energize to actuate and two channels are required for an actuation to occur.
C. Incorrect. Plausible. The first part is correct. Channel 1 is not utilized for HI-1 and 2 channels are required for an actuation to occur.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The first part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 2:23:01 PM                                                              2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            013K6.01 Importance Rating:              2.7 / 3.1 Technical
 
==Reference:==
TS 3.3.2, Amendment No. 165, pages 3.3.2-2 thru 3
                                                                                  & 9 thru 12 TS Bases 3.3.2, Rev 20, pages B 3.3.2-12, 13, 17, & 18 LOGIC 1X6AA02-00232, Rev 17.0 Picture of TSLB 4 References provided:            None Learning Objective:              LO-LP-39207-02 Given a set of Technical Specification and the Bases, determine for a specific set of plant conditions, equipment availability, and operational mode.
: a. Whether any Tech Spec LCOs of section 3.3 are exceeded.
: b. The required actions for any section 3.3 LCOs.
LO-LP-28103-05 List all ESFAS actuation signals with all applicable set points, coincidences, permissives, blocks, and discuss the system response to each ESF actuation signal.
LO-TA-28013        Trip Protection System Bistable LO-TA-28014        Use the BTI panel to bypass a protection channel Question origin:                BANK - HL18 NRC Question # 013K6.01 Cognitive Level:                M/F 10 CFR Part 55 Content:          41.7 / 45.5 to 45.8 Comments:
You have completed the test!
Wednesday, February 26, 2014 2:23:01 PM                                                              3
 
ESFAS Instrumentation 3.3.2 ACTIONS (continued)
CONDITION                    REQUIRED ACTION                        COMPLETION TIME C. One train inoperable.  --------------------NOTE-------------------
One train may be bypassed for up to 4 hours for surveillance testing provided the other train is OPERABLE.
                              ------------------------------------------------
C.1          Restore train to                    24 hours OPERABLE status.
OR C.2.1        Be in MODE 3.                        30 hours AND C.2.2        Be in MODE 5.                        60 hours D. One channel inoperable. --------------------NOTE-------------------
A channel may be bypassed for up to 12 hours for surveillance testing.
                              ------------------------------------------------
D.1          Place channel in trip.              72 hours OR D.2.1        Be in MODE 3.                        78 hours AND D.2.2        Be in MODE 4.                        84 hours (continued)
Vogtle Units 1 and 2                      3.3.2-2                      Amendment No. 116 (Unit 1)
Amendment No. 94 (Unit 2)
 
ESFAS Instrumentation 3.3.2 ACTIONS (continued)
CONDITION                    REQUIRED ACTION                        COMPLETION TIME E. One Containment        ---------------------NOTE------------------
Pressure High-3 channel One additional channel may be inoperable.            bypassed for up to 12 hours for surveillance testing.
                              ------------------------------------------------
E.1          Place channel in bypass.            72 hours OR E.2.1        Be in MODE 3.                        78 hours AND E.2.2        Be in MODE 4.                        84 hours F. One channel inoperable. F.1          Restore channel to                  48 hours OPERABLE status.
OR F.2.1        Be in MODE 3.                        54 hours AND F.2.2        Be in MODE 4.                        60 hours (continued)
Vogtle Units 1 and 2                      3.3.2-3                      Amendment No. 116 (Unit 1)
Amendment No. 94 (Unit 2)
 
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 7)
Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER                                                                                            NOMINAL SPECIFIED          REQUIRED                            SURVEILLANCE        ALLOWABLE                TRIP FUNCTION                CONDITIONS          CHANNELS        CONDITIONS        REQUIREMENTS              VALUE            SETPOINT
: 1. Safety Injection
: a. Manual                  1,2,3,4                2                B          SR 3.3.2.6                  NA                  NA Initiation
: b. Automatic              1,2,3,4                2                C          SR 3.3.2.2                  NA                  NA Actuation Logic                                                            SR 3.3.2.3 and Actuation                                                              SR 3.3.2.5 Relays
: c. Containment              1,2,3                3                D          SR    3.3.2.1              4.4 psig          3.8 psig (i)(j)
Pressure -                                                                  SR    3.3.2.4 (i)(j)
High 1                                                                      SR    3.3.2.7 SR    3.3.2.8
: d. Pressurizer            1,2,3(a)              4                D          SR    3.3.2.1            1856 psig        1870 psig (i)(j)
Pressure - Low                                                              SR    3.3.2.4 (i)(j)
SR    3.3.2.7 SR    3.3.2.8
: e. Steam Line              1,2,3(a)        3 per steam            D          SR    3.3.2.1            570(b) psig      585(b) psig (i)(j)
Pressure - Low                              line                          SR    3.3.2.4 (i)(j)
SR    3.3.2.7 SR    3.3.2.8 (continued)
(a) Above the P-11 (Pressurizer Pressure) interlock.
(b) Time constants used in the lead/lag controller are t1  50 seconds and t2  5 seconds.
(i) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.
(j) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and the as-left tolerances are specified in NMP-ES-033-006, Vogtle Setpoint Uncertainty Methodology and Scaling Instructions.
Vogtle Units 1 and 2                                            3.3.2-9                          Amendment No. 165 (Unit 1)
Amendment No. 147 (Unit 2)
 
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 7)
Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER                                                                                          NOMINAL SPECIFIED          REQUIRED                            SURVEILLANCE        ALLOWABLE                TRIP FUNCTION                CONDITIONS          CHANNELS        CONDITIONS          REQUIREMENTS            VALUE            SETPOINT
: 2. Containment Spray
: a. Manual                  1,2,3,4                2                B          SR 3.3.2.6                  NA                    NA Initiation
: b. Automatic              1,2,3,4                2                C          SR 3.3.2.2                  NA                    NA Actuation Logic                                                            SR 3.3.2.3 and Actuation                                                              SR 3.3.2.5 Relays
: c. Containment Pressure High - 3                  1,2,3                4                E          SR    3.3.2.1            22.4 psig          21.5 psig SR    3.3.2.4(i)(j)
(i)(j)
SR    3.3.2.7 SR    3.3.2.8 (continued)
(i) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.
(j) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and the as-left tolerances are specified in NMP-ES-033-006, Vogtle Setpoint Uncertainty Methodology and Scaling Instructions.
Vogtle Units 1 and 2                                          3.3.2-10                          Amendment No. 165 (Unit 1)
Amendment No. 147 (Unit 2)
 
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 3 of 7)
Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER                                                                                    NOMINAL SPECIFIED          REQUIRED                                SURVEILLANCE          ALLOWABLE    TRIP FUNCTION              CONDITIONS          CHANNELS        CONDITIONS            REQUIREMENTS            VALUE  SETPOINT
: 3. Phase A Containment Isolation (a)  Manual                1,2,3,4              2                B              SR 3.3.2.6                  NA        NA Initiation (b)  Automatic              1,2,3,4          2 trains              C              SR 3.3.2.2                  NA        NA Actuation Logic                                                              SR 3.3.2.3 and Actuation                                                                SR 3.3.2.5 Relays (c)  Safety Injection        Refer to Function 1 (Safety Injection) for all initiation functions and requirements.
: 4. Steam Line Isolation
: a. Manual              1,2(c),3(c)            2                F              SR 3.3.2.6                  NA        NA Initiation
: b. Automatic            1,2(c),3(c)            2                G              SR 3.3.2.2                  NA        NA Actuation Logic                                                              SR 3.3.2.3 and Actuation                                                                SR 3.3.2.5 Relays (continued)
(c) Except when one main steam isolation valve and associated bypass isolation valve per steam line is closed.
Vogtle Units 1 and 2                                      3.3.2-11                            Amendment No. 165 (Unit 1)
Amendment No. 147 (Unit 2)
 
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 7)
Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER                                                                                            NOMINAL SPECIFIED          REQUIRED                            SURVEILLANCE        ALLOWABLE                TRIP FUNCTION                CONDITIONS            CHANNELS          CONDITIONS        REQUIREMENTS            VALUE            SETPOINT
: 4. Steam Line Isolation (continued)
: c. Containment                                      3                D          SR  3.3.2.1                                14.5 psig 1,2(c),                                                                      15.4 psig Pressure -                                                                    SR  3.3.2.4(i)(j) 3(c)
(i)(j)
High 2                                                                        SR  3.3.2.7 SR  3.3.2.8
: d. Steam Line Pressure (1) Low                1,2(c),          3 per steam            D          SR  3.3.2.1          570 (b) psig      585 (b) psig line                          SR  3.3.2.4(i)(j) 3(a)(c)                                                          (i)(j)
SR  3.3.2.7 SR  3.3.2.8 (2) Negative            3(d)(c)          3 per steam            D          SR  3.3.2.1                                  100 (e)
Rate -                                    line                                      (i)(j)      125 (e)
SR  3.3.2.4 High                                                                                  (i)(j)      psi/sec              psi/sec SR  3.3.2.7 SR  3.3.2.8 (continued)
(a) Above the P-11 (Pressurizer Pressure) interlock.
(b) Time constants used in the lead/lag controller are t1  50 seconds and t2  5 seconds.
(c) Except when one main steam isolation valve and associated bypass isolation valve per steam line is closed.
(d) Below the P-11 (Pressurizer Pressure) interlock.
(e) Time constant utilized in the rate/lag controller is  50 seconds.
(i) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.
(j) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and the as-left tolerances are specified in NMP-ES-033-006, Vogtle Setpoint Uncertainty Methodology and Scaling Instructions.
Vogtle Units 1 and 2                                            3.3.2-12                        Amendment No. 165 (Unit 1)
Amendment No. 147 (Unit 2)
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE          b. Safety Injection - Automatic Actuation Logic and SAFETY ANALYSES,        Actuation Relays (continued)
LCO, and APPLICABILITY          consequences of an abnormal condition or accident. Unit pressure and temperature are very low and many ESF components are administratively locked out or otherwise prevented from actuating to prevent inadvertent overpressurization of unit systems.
: c. Safety Injection - Containment Pressure  High 1 (PI-0934, PI-0935, PI-0936)
NOTE: Containment pressure channels are also required OPERABLE by the Post Accident Monitoring Technical Specification.
PI-0937 is not This signal provides protection against the following included.                  accidents:
* SLB inside containment;
* LOCA; and
* Feed line break inside containment.
Containment Pressure  High 1 provides no input to any control functions. Thus, three OPERABLE channels are sufficient to satisfy protective requirements with a two-out-of-three logic. The transmitters (d/p cells) and electronics are located outside of containment with the sensing line (high pressure side of the transmitter) located inside containment.
Thus, the high pressure Function will not experience any adverse environmental conditions and the NTSP reflects only steady state instrument uncertainties. Containment Pressure  High 1 must be OPERABLE in MODES 1, 2, and 3 when there is sufficient energy in the primary and secondary systems to pressurize the containment following a pipe break. In MODES 4, 5, and 6, there is insufficient energy in the primary or secondary systems to pressurize the containment.
(continued)
Vogtle Units 1 and 2          B 3.3.2-12                              REVISION 20
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE          d. Safety Injection - Pressurizer Pressure  Low SAFETY ANALYSES, LCO, and                This signal (PI-0455A, B, & C, PI-0456, PI-0456A, PI-0457, APPLICABILITY          PI-0457A, PI-0458 & PI-0458A) provides protection (continued)          against the following accidents:
* Inadvertent opening of a steam generator (SG) relief or safety valve;
* SLB;
* A spectrum of rod cluster control assembly ejection accidents (rod ejection);
* Inadvertent opening of a pressurizer relief or safety valve;
* LOCAs; and
* SG Tube Rupture.
Pressurizer pressure provides both control and protection functions: input to the Pressurizer Pressure Control System, reactor trip, and SI. Therefore, the actuation logic must be able to withstand both an input failure to control system, which may then require the protection function actuation, and a single failure in the other channels providing the protection function actuation. Thus, four OPERABLE channels are required to satisfy the requirements with a two-out-of-four logic.
The transmitters are located inside containment, with the taps in the vapor space region of the pressurizer, and thus possibly experiencing adverse environmental conditions (LOCA, SLB inside containment, rod ejection). Therefore, the NTSP reflects the inclusion of both steady state and adverse environmental instrument uncertainties.
This Function must be OPERABLE in MODES 1, 2, and 3 (above P-11) to mitigate the consequences of an HELB inside containment. This signal may (continued)
Vogtle Units 1 and 2          B 3.3.2-13                                REVISION 20
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE          b. Containment Spray - Automatic Actuation Logic and SAFETY ANALYSES,        Actuation Relays (continued)
LCO, and APPLICABILITY          this MODE, adequate time is available to manually actuate required components in the event of a DBA. However, because of the large number of components actuated on a containment spray, actuation is simplified by the use of the manual actuation handswitches. Automatic actuation logic and actuation relays must be OPERABLE in MODE 4 to support system level manual initiation. In MODES 5 and 6, there is insufficient energy in the primary and secondary systems to result in containment overpressure. In MODES 5 and 6, there is also adequate time for the operators to evaluate unit conditions and respond, to mitigate the consequences of abnormal conditions by manually starting individual components.
: c. Containment Spray - Containment Pressure High  3 (PI-0934, PI-0935, PI-0936, PI-0937)
NOTE: Containment Pressure Channels are also required OPERABLE by the Post Accident Monitoring Technical Specification.
This signal provides protection against a LOCA or an SLB inside containment. The transmitters (d/p cells and electronics) are located outside of containment with the sensing line (high pressure side of the transmitter) located inside containment. Thus, they will not experience any adverse environmental conditions and the NTSP reflects only steady state instrument uncertainties.
This Function requires the bistable output to energize to perform its required action. It is not desirable to have a loss of power actuate containment spray, since the consequences of an inadvertent actuation of containment spray could be serious. Note that this Function also has the inoperable channel placed in bypass rather than trip to decrease the probability of an inadvertent actuation.
(continued)
Vogtle Units 1 and 2          B 3.3.2-17                                REVISION 20
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE          c. Containment Spray - Containment Pressure High  3 SAFETY ANALYSES,          (continued)
LCO, and APPLICABILITY          The Containment Pressure High-3 instrument Function consists of four channels in a two-out-of-four logic configuration. Since containment pressure is not used for control, this arrangement exceeds the minimum redundancy requirements. Additional redundancy is warranted because this Function is energize to trip.
Containment Pressure  High 3 must be OPERABLE in MODES 1, 2, and 3 when there is sufficient energy in the primary and secondary sides to pressurize the containment following a pipe break. In MODES 4, 5, and 6, there is insufficient energy in the primary and secondary sides to pressurize the containment and reach the Containment Pressure  High 3 setpoints.
: 3. Phase A Containment Isolation Phase A containment isolation is actuated automatically by SI, or manually via the automatic actuation logic.
: a. Phase A Isolation  Manual Initiation Manual Phase A Containment Isolation is actuated by either of two switches in the control room. Either switch actuates both trains. Note that manual initiation of Phase A Containment Isolation also actuates Containment Ventilation Isolation.
: b. Phase A Isolation  Automatic Actuation Logic and Actuation Relays Automatic Actuation Logic and Actuation Relays consist of the same features and operate in the same manner as described for ESFAS Function 1.b.
Under specific conditions, a single inoperable actuation relay does not require that the affected automatic actuation logic function be (continued)
Vogtle Units 1 and 2          B 3.3.2-18                              REVISION 20
: 1. 014A1.02 001/LOIT/RO/M/F 3.2/3.6/014A1.02/LO-PP-27201///
Initial condition:
            - Unit 1 is at 100% reactor power and stable.
The following DRPI alarms and indications are observed:
            -  ALB10-D05 RPI URGENT ALARM is lit.
            -  ALB10-C05 RPI NON URGENT ALARM is lit.
            -  ALB10-E05 ROD AT BOTTOM is lit.
            -  Control rod H2 Rod Bottom LED is lit.
            -  Control rod H2 General Warning LED is flashing.
            -  Data A and Data B Failure LEDs are flashing.
Which one of the following completes the following statement?
Control rod H2 __(1)__,
and automatic and manual rod motion __(2)__ inhibited.
__(1)__                              __(2)__
A.                  dropped                                  is B.                  dropped                                is NOT C.              did NOT drop                                is D.              did NOT drop                              is NOT K/A 014              Rod Position Indication System (RPIS)
A1.02            Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RPIS controls, including:
                        - Control rod position indication on control room panels.
K/A MATCH ANALYSIS The question tests the candidate's ability to address multiple DRPI alarms and indications and determine if these conditions are indicative of an actual dropped rod or an instrumentation failure. In addition, the candidate must determine if there is any impact to the rod control system.
Wednesday, February 26, 2014 2:26:51 PM                                                      1
 
EXPLANATION OF REQUIRED KNOWLEDGE All indications correspond to a simutaneous Data A and Data B failure for rod H2. The distinction between an actual dropped rod and a RPI failure are denoted by the presence of the General Warning LED and the Data A and Data B failure LEDs.
Additionally, rod H2 is a low-worth control rod located on the outer edges of the core. If this rod dropped, little to no change in reactor power would be noticed. ALB10-E05 ROD AT BOTTOM is driven from the RPI position, and would alarm because of the indicated position of 0 steps on RPI.
The RPI system does not interface with the rod control system and therefore can not restrict rod motion. The RPI and Rod Control annunciators are commonly confused.
The rod control urgent failure would send an inhibit signal to the PULSER/
OSCILLATOR and restrict rod motion in both auto and manual.
Reference ARP 17010-1 for specifics associated with expected plant conditions and responses.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. The presence of the General Warning and Data A and Data B LEDs are symptoms of simulatenous RPI Data A and Data B failure and not a dropped rod.
However, a candidate with insufficient knowledge of the RPI system would see the RPI rod at bottom LED and annunciator ALB10-E05 ROD AT BOTTOM and find it reasonable that the rod had actually dropped. These two conditions are entry symptoms for AOP 18003-C section A for a dropped rod in Mode 1. This is a common point of confusion among LOIT candidates.
The second part is incorrect. The RPI system does not interface with the rod control system and therefore can not restrict rod motion. However, the RPI and rod control alarms are commonly confused. If a candidate swaps the two systems and is thinking about a rod control urgent failure, then this answer would be correct.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. The RPI system does not interface with the rod control system and therefore can not restrict rod motion.
C. Incorrect. Plausible. The first part is correct. The presence of the General Warning and Data A and Data B LEDs are symptoms of simultaneous RPI Data A and Data B failure and not a dropped rod.
The second part is incorrect. See the second part of choice A above.
Wednesday, February 26, 2014 2:26:51 PM                                                            2
 
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Level:                          RO Tier # / Group #                T2 / G2 K/A#                            014A1.02 Importance Rating:              3.2 / 3.6 Technical
 
==Reference:==
ARP 17010-1, Rev 50.0, pages 3, 12-15, 21-24, 34-35, 44-46 AOP 18003-C, Rev 26.4, page 1 References provided:            None Learning Objective:              LO-PP-27101-20 Describe what occurs upon receipt of a Rod Control System urgent failure; include how rod motion is inhibited.
LO-PP-27201-02 Describe the principle of operation of the DRPI system.
LO-PP-27201-05 Describe how the DRPI system responds to a loss of Data A or B.
LO-PP-27201-06 State the conditons which will cause the following:
: a. RPI Urgent Failure LEDs
: b. RPI Urgent Failure annunciator
: c. General Warning (GW) LED(s)
: d. RPI Non-Urgent Failure annunciator
: e. CCC Failure LEDs
: f. Rod at bottom annunciator
: g. Two or more rods at bottom annunciator
: h. Rod Deviation annunicator
: i. DRPI indication for a rod moves from 12 steps withdrawn to 18 steps withdrawn Question origin:                BANK Cognitive Level:                M/F 10 CFR Part 55 Content:          41.6 / 41.7 / 45.5 Comments:
You have completed the test!
Wednesday, February 26, 2014 2:26:51 PM                                                              3
 
Approved By                                                                        Procedure    Version J. B. Stanley                          Vogtle Electric Generating Plant            18003-C        26.4 Effective Date                                                                      Page Number 6/14/13 ROD CONTROL SYSTEM MALFUNCTION                          1 of 30 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure provides instructions for malfunctions of the Rod Control System resulting in uncontrolled rod motion, dropped or misaligned rods.
SYMPTOMS SECTION A, DROPPED RODS IN MODE 1 ALB10-E5 ROD AT BOTTOM ALB10-F2 POWER RANGE HI NEUTRON FLX RATE ALERT ALB10-C2 POWER RANGE CHANNEL DEVIATION Rod bottom LED on digital rod position indication.
Rod misaligned greater than 110 steps from demand position Tavg dropping.
SECTION B, UNCONTROLLED CONTINUOUS ROD MOTION IN ALL MODES Rod motion with invalid demand from the Automatic Rod Control System.
Failure of rods to stop moving when the Rod Motion Switch is released.
SECTION C, MISALIGNED RODS IN MODE 1 ALB10-C2 POWER RANGE CHANNEL DEVIATION ALB10-D2 POWER RANGE UP DET HI FLX DEV ALB10-E2 POWER RANGE LWR DET HI FLX DEV Failure of ALB10-C4 ROD BANK LO LIMIT or ALB10-D4 ROD BANK LO-LO LIMIT to reset during rod withdrawal.
Rod misaligned greater than 12 steps and less than or equal to 110 steps from demand position.
Quadrant power tilt ratio calculation exceeds 1.02.
Printed February 4, 2014 at 09:03
 
Approved By                                                                                    Procedure Number Rev J.B. Stanley                          Vogtle Electric Generating Plant                        17010-1        50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL                Page Number 08/16/2011                                              1C1 ON MCB                                      3 of 66 ALB 10 (1)                    (2)            (3)                (4)          (5)            (6)
A    SR/IR                  NIS SOURCE AND  POWER RANGE HI    REACTOR BYPASS REACTOR BYPASS ROD CONTROL SIG PROCESSOR          INTMD RANGE    NEUTRON FLX HI    BRKR BYA      BRKR BYA      NON URGENT TROUBLE                TRIP BYPASS    SETPOINT ALERT    IN-OPERATE    CLOSE          FAILURE B    SOURCE RNG HI                          POWER RANGE      REACTOR BYPASS REACTOR BYPASS ROD CONTROL SHUTDOWN FLUX                          HI NEUTRON FLX    BRKR BYB      BRKR BYB      URGENT FAILURE ALARM BLOCKED                          LOW SETPOINT      IN-OPERATE    CLOSE C    SOURCE RANGE          POWER RANGE    OVERPOWER T      ROD BANK      RPI            NIS CHANNEL HI FLUX LEVEL          CHANNEL        ROD BLOCK AND    LO LIMIT      NON URGENT    ON TEST AT SHUTDOWN            DEVIATION      RUNBACK ALERT                    ALARM D    INTMD RANGE            PWR RANGE UP    OVERPOWER        ROD BANK      RPI            ROD DEV HI FLUX                DET HI FLX DEV  ROD STOP          LO-LO LIMIT    URGENT ALARM LEVEL ROD STOP E    SR/IR REMOTE          PWR RANGE LWR  OVERTEMP T                      ROD AT BOTTOM  RADIAL TILT SIG PROCESSOR          DET HI FLX DEV  ROD BLOCK AND DPU-B TROUBLE                          RUNBACK ALERT F    SR/IR                  POWER RANGE                      ROD DRIVE M-G  TWO OR MORE    DELTA FLUX AMPLIFIER              HI NEUTRON FLX                    SET TROUBLE    RODS AT BOTTOM DEVIATION TROUBLE                RATE ALERT Printed February 3, 2014 at 16:00
 
Approved By                                                                                  Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                          17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL            Page Number 08/16/2011                                        1C1 ON MCB                                      12 of 66 WINDOW A06 ORIGIN                          SETPOINT ROD CONTROL Power Cabinet                  Not Applicable              NON URGENT Logic Cabinet                                                FAILURE 1.0                PROBABLE CAUSE
: 1.      Power Cabinet Non Urgent Failure:
Loss of +28V DC power supply PS1 or PS2, or -24V DC power supply PS3 or PS4 due to low line voltage, blown fuse or failure of module's AC supply train.
: 2.      Logic Cabinet Non Urgent Failure:
: a. Loss of 16.5V DC power supplies PS1, PS2, PS4, or PS5,
: b. Loss of 100V DC power supplies PS3 or PS6.
2.0                AUTOMATIC ACTIONS NONE Printed February 3, 2014 at 16:00
 
Approved By                                                                                  Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                          17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL              Page Number 08/16/2011                                            1C1 ON MCB                                  13 of 66 WINDOW A06 (Continued) 3.0                INITIAL OPERATOR ACTIONS NOTE This procedure should be continued even if the reactor is tripped.
: 1.      IF the reactor is NOT tripped AND within one hour of receiving the alarm, check the PS1 and PS2 Power On status lights on the outside of listed cabinets located in the Control Building in room RB 71.
1-1606-U3-FLR-001            Power Cabinet 1BD 1-1606-U3-FLR-002            Power Cabinet 1AC 1-1606-U3-FLR-003            Power Cabinet 2BD 1-1606-U3-FLR-004            Power Cabinet 2AC 1-1606-U3-FLR-005            Power Cabinet SCDE
: a. Perform 13502-1, Section 4.4.6 to test backup batteries for PS1 and PS2 in each power cabinet.
: b. IF both PS1 and PS2 power supply lights are lit AND backup batteries test good for all Power Cabinets, Go To step 3.6.
: c. IF only one power supply light (either PS1 or PS2) for a Power Cabinet is lit, perform the following:
(1)    Reset the respective power supply:
Open the cabinet door and locate switches S6 and S7 in the lower left-hand area of the monitoring test cabinet.
Depress switch S6 to reset PS1 or S7 to reset PS2. Check the status light for PS1 or PS2 ON.
(2)    IF the power supply will NOT reset, pull and caution tag the 120V AC supply fuses to the failed power supply and timer to prevent possible damage to the Reset Timer relays:
Power Supply PS1            FU1A and FU1B Power Supply PS2            FU2A and FU2B (3)    Go to step 3.6.
Printed February 3, 2014 at 16:00
 
Approved By                                                                                    Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                          17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL              Page Number 08/16/2011                                            1C1 ON MCB                                    14 of 66 WINDOW A06 (Continued)
: d. IF neither PS1 and PS2 power supply lights are lit, attempt to reset at least one power supply by opening the cabinet door and locate switches S6 and S7 in the lower left hand area of the monitoring test cabinet. Depress switch S6 to reset PS1 or S7 to reset PS2; Check the status light for PS1 or PS2 ON.
NOTE Communications should be established between the control room and the rod control cabinets prior to continuing with this procedure.
CAUTION IF PS1 or PS2 are not reset, power to the holding coils is supplied from the back-up batteries. IF PS1 or PS2 are not reset within one hour, rod drops may occur due to low voltage on the holding coil.
: 2.      IF no power supply will reset, notify I&C to immediately initiate action to restore at least one power supply in each power cabinet and to monitor voltage at test point E1 to neutral.
: 3.      IF voltage at test point E1 has degraded to less than 19.5V DC or rods drop, trip the reactor and Go To 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION, while continuing actions in this procedure.
: 4.      For any power supply that will NOT reset, pull and caution tag the 120V AC supply fuses to the failed power supply and timer to prevent possible damage to the Reset Timer relays:
Power Supply PS1              FU1A and FU1B Power Supply PS2              FU2A and FU2B
: 5.      WHEN or IF the reactor is tripped AND within two hours of receiving the alarm, pull and caution tag the battery fuse to both power supply back-up batteries to prevent possible draining the battery.
Battery for Power Supply PS1      F5A Battery for Power Supply PS2      F6A Printed February 3, 2014 at 16:00
 
Approved By                                                                                Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                      17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL          Page Number 08/16/2011                                          1C1 ON MCB                                  15 of 66 WINDOW A06 (Continued)
: 6.      Notify I&C to determine which power supply has failed by checking the output voltage of all Power Cabinet and Logic Cabinet power supplies and repair as needed.
: 7.      Check, replace fuses and remove caution tags when power is ready to be restored to the power supplies and batteries.
4.0                SUBSEQUENT OPERATOR ACTIONS NOTE The Non Urgent Failure alarm will automatically reset when the malfunction is corrected.
Notify I&C personnel to investigate and correct the cause of the alarm.
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCE:==
1X6AT01-573, 1X6AT01-574, 1X6AT01-575, 1X6AT01-576, 1X6AT01-605, 1X3D-BD-R01C Printed February 3, 2014 at 16:00
 
Approved By                                                                                  Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                          17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL              Page Number 08/16/2011                                            1C1 ON MCB                                    21 of 66 WINDOW B06 ORIGIN                            SETPOINT ROD CONTROL Power Cabinet                    Not Applicable                URGENT FAILURE Logic Cabinet 1.0                PROBABLE CAUSE
: 1.      Power Cabinet Urgent Failure:
: a. Phase fault - voltage to coils has excessive ripple due to a blown fuse or thyristor that has lost gate control.
: b. Regulation failure - the coil current does not match current order within a preset time or full current is on too long.
: c. Multiplexing failure - power is being supplied to a movable or lift coil when movement of that rod has not been commanded.
: d. Logic failure - simultaneous zero current orders to stationary and movable grippers.
: e. Loose Card - loose or removed printed circuit card.
: 2.      Logic Cabinet Urgent Failure:
: a. Pulser fails to generate pulses when signaled.
: b. Slave Cycler receives "Go" order signal before completing previous step.
: c. Loose circuit card.
Printed February 3, 2014 at 16:00
 
Approved By                                                                                    Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                            17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL              Page Number 08/16/2011                                              1C1 ON MCB                                  22 of 66 WINDOW B06 (Continued) 2.0                AUTOMATIC ACTIONS Prevents automatic and manual rod motion by performing the following:
POWER CABINET URGENT FAILURE CAUTIONS Rods powered from the unaffected Power Cabinets can be moved in INDIVIDUAL BANK SELECT. However, IF the cause of the alarm is a loss of current to the stationary gripper (regulation failure) THEN moving the bank selector switch may cause the affected group of rods to drop.
IF the cause of the alarm is a logic failure in the Power Cabinet, THEN resetting the alarm from the QMCB or locally may cause ratcheting of the rods.
: 1.      Sends an inhibit signal to the PULSER/OSCILLATOR when the affected group is selected to move.
: 2.      Supplies holding current to the movable and stationary grippers and no current to the lift coils for the affected group.
: 3.      Sends an inhibit signal to the group step counter for the affected group.
Printed February 3, 2014 at 16:00
 
Approved By                                                                                  Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                          17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL              Page Number 08/16/2011                                          1C1 ON MCB                                    23 of 66 WINDOW B06 (Continued)
LOGIC CABINET URGENT FAILURE NOTES An Urgent Failure in the Logic Cabinet main circuits will prevent all rod motion for all of the control banks and shutdown banks A and B only.
Shutdown banks C, D, and E will not be affected unless the alarm is caused by a loose or missing card in which case Shutdown banks C, D, and E will not be allowed to move either.
An Urgent Failure in the Shutdown Banks C, D, and E portion of the Logic Cabinet will prevent all rod motion for Shutdown banks C, D, and E only.
The control banks and shutdown banks A and B will not be affected unless the alarm is cause by a loose or missing card in which case the control banks and shutdown banks a and B will not be allowed to move either.
: 1.      Sends an inhibit signal to the PULSER/OSCILLATOR 3.0                INITIAL OPERATOR ACTIONS IF all rod motion has NOT stopped, Go To 18003-C, "Rod Control System Malfunction".
4.0                SUBSEQUENT OPERATOR ACTIONS NOTES The Rod Control Urgent Failure alarm seals in and must be reset using the Rod Control Alarm Reset Handswitch, 1-HS-40039, when the condition causing the alarm has cleared.
Use of 1-HS-40039 resets the alarm circuits, demands full latching current and resets the MASTER CYCLER.
: 1.      Stabilize Tavg, using turbine load and boration or dilution.
: 2.      Notify appropriate plant personnel to investigate and correct the cause of the alarm.
: 3.      Refer To 13502-1 to reset rod control components.
Printed February 3, 2014 at 16:00
 
Approved By                                                                        Procedure Number Rev J.B. Stanley                    Vogtle Electric Generating Plant                  17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL    Page Number 08/16/2011                                      1C1 ON MCB                                24 of 66 WINDOW B06 (Continued) 5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCE:==
1X6AT01-573, 1X6AT01-574, 1X6AT01-575, and 1X6AT01-576 Printed February 3, 2014 at 16:00
 
Approved By                                                                                Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                      17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL          Page Number 08/16/2011                                            1C1 ON MCB                                34 of 66 WINDOW C05 ORIGIN                            SETPOINT RPI Control Board                    Not Applicable            NON URGENT Display Panel                                              ALARM 1.0                PROBABLE CAUSE
: 1.      Card failure or removal.
: 2.      Detector coil or cable failure.
: 3.      Power supply failure.
2.0                AUTOMATIC ACTIONS
: 1.      Data A Failure or Data B Failure LEDs flash depending on the cause and location of problem.
: 2.      The General Warning LEDs flash for the affected rod(s).
NOTE DRPI DATA A FAILURE results in system accuracy of +10, -4 steps. DRPI DATA B FAILURE results in system accuracy of +4, -10 steps.
: 3.      DRPI goes to Half Accuracy Mode for the affected rod(s).
3.0                INITIAL OPERATOR ACTIONS IF Dynamic Rod Worth measurement is in progress for Physics Testing, follow the Physics Testing Procedures Guidance for Operator actions in response to a DRPI Non-Urgent Failure Alarm.
IF Physics Testing is NOT in progress AND IF this alarm is coming in on an intermittent basis, THEN Refer To 13502-1, Control Rod Drive And Position Indication System, section for operation with DRPI at half accuracy.
Printed February 3, 2014 at 16:00
 
Approved By                                                                                  Procedure Number Rev J.B. Stanley                        Vogtle Electric Generating Plant                      17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL            Page Number 08/16/2011                                            1C1 ON MCB                                  35 of 66 WINDOW C05 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.      Notify I&C personnel to investigate and correct the cause of the alarm.
: 2.      IF it is determined that rod(s) are misaligned, Go To 18003-C, "Rod Control System Malfunction".
: 3.      Refer To 13502-1, Control Rod Drive And Position Indication System, section for operation with DRPI at half accuracy.
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCE:==
1X6AT02-187 Printed February 3, 2014 at 16:00
 
Approved By                                                                            Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                    17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL        Page Number 08/16/2011                                            1C1 ON MCB                              44 of 66 WINDOW D05 ORIGIN                            SETPOINT RPI Control Board                    Not Applicable        URGENT ALARM Display Panel 1.0                PROBABLE CAUSE
: 1.      Simultaneous failure of DATA A and DATA B.
: 2.      Greater than 1 BIT difference between DATA A and DATA B - DRPI sees the affected rod(s) at two different positions.
: 3.      Sum of DATA A and DATA B is greater than 38 - DRPI sees the affected rod(s) at greater than 228 steps withdrawn.
2.0                AUTOMATIC ACTIONS
: 1.      DRPI Control Board Displays:
: a. URGENT ALARM 1, 2, 3 LEDs flash.
: b. DATA A and DATA B FAILURE LEDs.
: c. GENERAL WARNING LED(s) for the affected rod(s)
: d. ROD BOTTOM LED(s) for the affected rod(s)
: 2.      QMCB alarms
: a. ROD AT BOTTOM and/or TWO OR MORE RODS AT BOTTOM.
Printed February 3, 2014 at 16:00
 
Approved By                                                                                Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                        17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL            Page Number 08/16/2011                                          1C1 ON MCB                                    45 of 66 WINDOW D05 (Continued) 3.0                INITIAL OPERATOR ACTIONS IF Dynamic Rod Worth measurement is in progress for Physics Testing, follow the Physics Testing Procedures Guidance for operator actions in response to a DRPI Urgent Failure Alarm.
IF Dynamic Rod Worth measurement is NOT in progress AND IF any rod is NOT fully inserted:
MODES 1 & 2
: 1.      Immediately place DRPI ACCURACY MODE switch to the A ONLY or B ONLY position.
: 2.      IF the RPI URGENT ALARM does NOT clear in either the A ONLY or B ONLY position, place rods in MANUAL and minimize rod motion.
: 3.      IF the RPI URGENT ALARM clears in either the A ONLY or B ONLY position, operate DRPI using the half accuracy mode per 13502-1, Control Rod Drive and Position Indication System.
MODES 3, 4, & 5
: 1.      Immediately place DRPI ACCURACY MODE switch to the A ONLY or B ONLY position.
: 2.      IF the RPI URGENT ALARM does NOT clear when the DRPI ACCURACY MODE switch is in either the A ONLY or B ONLY position:
: a. Immediately open the reactor trip breakers per TR 13.1.8 or 13.1.9.
: b. Return the DRPI ACCURACY MODE switch to DATA A+B per 13502-1, as determined by the SS.
IF this alarm is coming in on an intermittent basis, THEN Refer To 13502-1, Control Rod Drive And Position Indication System, section for operation with DRPI at half accuracy.
Printed February 3, 2014 at 16:00
 
Approved By                                                                                  Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                        17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL            Page Number 08/16/2011                                          1C1 ON MCB                                    46 of 66 WINDOW D05 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.      Refer To Technical Specification LCO 3.1.7 and Technical Requirements TR 13.1.8 and 13.1.9.
: 2.      Notify Maintenance (I&C) to determine the actual rod position of rod(s) in question and begin recording rod positions in accordance with 14915-1, "Special Condition Surveillance Logs".
: 3.      WHEN the DRPI problems have been resolved, return DRPI to normal per 13502-1.
5.0                COMPENSATORY OPERATOR ACTIONS
: 1.      Initiate Data Sheet 4 of 14915-1, "Special Conditions Surveillance Log".
: 2.      Log corrective actions to repair the disabled annunciator or reasons for no action on 10018-C, "Annunciator Control", Figure 2.
: 3.      Log compensatory actions on 10018-C, "Annunciator Control", Figure 5.
END OF SUB-PROCEDURE
 
==REFERENCE:==
1X6AT02-187 Printed February 3, 2014 at 16:00
: 1. 015A2.04 001/LOIT/RO/M/F 3.3/3.8/015A2.04/LO-LP-39206-06///
Initial condition:
          - Unit 1 is at 100% reactor power.
Current conditions:
          - A spurious turbine runback occurs and is terminated at 80% reactor power.
          - ALB10-F06 DELTA FLUX DEVIATION is received.
Which one of the following completes the following statement?
Per Tech Spec 3.2.3, "Axial Flux Difference (AFD)," the AFD is considered outside limits when a minimum of __(1)__ Power Range NI channels indicate outside AFD limits, and with AFD outside of the Tech Spec 3.2.3 limits, thermal power must be reduced to
__(2)__ within 30 minutes.
__(1)__                                __(2)__
A.                      1                                    < 50%
B.                      1                                    < 75%
C.                      2                                    < 50%
D.                      2                                    < 75%
K/A 015              Nuclear Instrumentation A2.04            Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
                        - Effects on axial flux density of control rod alignment and sequencing, xenon production and decay, and boron vs. control rod reactivity changes K/A MATCH ANALYSIS The question establishes a scenario where rod motion has resulted in an AFD transient.
Wednesday, February 26, 2014 2:28:18 PM                                                    1
 
Utilizing NI's, the candidate is required to predict when AFD will be outside Tech Spec required limits. Additionally, the candidate is required to utilize Tech Spec guidance to mitigate the consequences of the AFD transient.
EXPLANATION OF REQUIRED KNOWLEDGE The described turbine runback will result in auto rod insertion. Depending on the magnitude of the runback, control rods may insert well below the Rod Insertion Limit.
This large rod transient will affect AFD. Per the Note above the line in TS 3.2.3, AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD is outside limits.
With AFD outside limits, TS 3.2.3 Cond A states that reactor power must be reduced to
      <50% RTP within 30 minutes. Since this is a <1hr Tech Spec action, it is RO required knowledge. Additionally, ARP 17010-C for window F06 also directs lowering power to
      <50% with AFD outside limits.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per the Note above the line in TS 3.2.3, AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD is outside limits.
However, most Tech Spec limits are exceeded when a single instrument is outside the required value. It is reasonable for a candidate without specific knowledge of the Note for TS 3.2.3 to apply this generic knowledge concept and believe AFD is out of spec based on a single NI channel.
The second part is correct. With AFD outside limits, TS 3.2.3 Cond A states that reactor power must be reduced to <50%
RTP within 30 minutes.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. With AFD outside limits, TS 3.2.3 Cond A states that reactor power must be reduced to <50%
RTP within 30 minutes. However, other specs have power reduced below 75% as required action. For example, TS 3.2.4 with one Power Range Nuclear instrument inoperable, the QPTR action would not be applicable if Reactor Power was reduced below 75%. Additionally, TS 3.1.4 would required thermal power reduced to <75% if rod alignments are not within limits and SDM cannot be verified.
C. Correct.                  The first part is correct. Per the Note above the line in TS 3.2.3, AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD is outside limits The second part is correct. See the second part of choice A above.
Wednesday, February 26, 2014 2:28:18 PM                                                              2
 
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Level:                          RO Tier # / Group #                T2 / G2 K/A#                            015A2.04 Importance Rating:              3.3 / 3.8 Technical
 
==Reference:==
ARP 17010-C, Rev 50.0, pages 65 & 66 TS 3.1.4, Amendment No. 96, pages 3.1.4-1 & 2 TS 3.2.3, Amendment No. 158, page 3.2.3-1 TS 3.2.4, Amendment No. 96, pages 3.2.4-1 thru 4 References provided:            None Learning Objective:              LO-LP-39206-06 State the action required for being outside the band at various power levels.
LO-LP-39206-01 For any item in section 3.2 of Tech Specs, be able to:
: a. State the LCO.
: b. State any one hour or less required actions.
Question origin:                NEW Cognitive Level:                M/F 10 CFR Part 55 Content:          41.1 / 41.5 / 41.10 Comments:
You have completed the test!
Wednesday, February 26, 2014 2:28:18 PM                                                            3
 
Rod Group Alignment Limits 3.1.4 3.1  REACTIVITY CONTROL SYSTEMS 3.1.4  Rod Group Alignment Limits LCO 3.1.4            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:      MODES 1 and 2.
ACTIONS CONDITION                      REQUIRED ACTION                  COMPLETION TIME A. One or more rod(s)          A.1.1    Verify SDM is  the limit      1 hour untrippable.                          specified in the COLR.
OR A.1.2    Initiate boration to restore  1 hour SDM to within limit.
AND A.2      Be in MODE 3.                  6 hours B. One rod not within          B.1.1    Verify SDM is  the limit      1 hour alignment limits.                    specified in the COLR.
OR (continued)
Vogtle Units 1 and 2                        3.1.4-1                  Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
 
Rod Group Alignment Limits 3.1.4 ACTIONS CONDITION        REQUIRED ACTION                  COMPLETION TIME B.  (continued)    B.1.2  Initiate boration to restore  1 hour SDM to within limit.
AND B.2    Reduce THERMAL                2 hours POWER to  75% RTP.
AND B.3    Verify SDM is  the limit      Once per specified in the COLR.        12 hours AND B.4    Perform SR 3.2.1.1.            72 hours AND B.5    Perform SR 3.2.2.1.            72 hours AND B.6    Reevaluate safety              5 days analyses and confirm results remain valid for duration of operation under these conditions.
(continued)
Vogtle Units 1 and 2          3.1.4-2                  Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
 
AFD (RAOC Methodology) 3.2.3 3.2  POWER DISTRIBUTION LIMITS 3.2.3  AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology)
LCO 3.2.3            The AFD 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                        30 minutes POWER to < 50% RTP.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.2.3.1        Verify AFD within limits for each OPERABLE                          In accordance with excore channel.                                                      the Surveillance Frequency Control Program AND Once within 1 hour and every 1 hour thereafter with the AFD monitor alarm inoperable Vogtle Units 1 and 2                                3.2.3-1                      Amendment No. 158 (Unit 1)
Amendment No. 140 (Unit 2)
 
QPTR 3.2.4 3.2  POWER DISTRIBUTION LIMITS 3.2.4  QUADRANT POWER TILT RATIO (QPTR)
LCO 3.2.4                The QPTR shall be d 1.02.
APPLICABILITY:            MODE 1 with THERMAL POWER > 50% RTP.
ACTIONS CONDITION                        REQUIRED ACTION              COMPLETION TIME A.  -----------NOTE-------------    A.1    Limit THERMAL POWER      2 hours Required Action A.6                      to t 3% below RTP for must be completed                        each 1% of QPTR > 1.00.
whenever Required Action A.5 is                    AND implemented.
      -------------------------------- A.2.1  Perform SR 3.2.4.1.      Once per 12 hours QPTR not within limit.          AND A.2.2  Limit THERMAL POWER      -----------NOTE----------
to t 3% below RTP for    For performances of each 1% QPTR > 1.00. Required Action A.2.2 the Completion Time is measured from the completion of SR 3.2.4.1.
                                                                        -----------------------------
2 hours AND A.3    Perform SR 3.2.1.1 and  Within 24 hours after SR 3.2.2.1.              achieving equilibrium conditions with THERMAL POWER limited by Required Actions A.1 and A.2.2 (continued)
Vogtle Units 1 and 2                            3.2.4-1              Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
 
QPTR 3.2.4 ACTIONS CONDITION      REQUIRED ACTION                        COMPLETION TIME A.  (continued)                                              AND Once per 7 days thereafter AND A.4  Reevaluate safety                    Prior to increasing analyses and confirm                THERMAL POWER results remain valid for            above the limit of duration of operation                Required Action under this condition.                A.1 and A.2.2 AND A.5  -------------NOTE-------------
Perform Required Action A.5 only after Required Action A.4 is completed.
                          ----------------------------------
Calibrate excore detectors          Prior to increasing to show QPTR = 1.00.                THERMAL POWER above the limit of Required Action A.1 and A.2.2 AND (continued)
Vogtle Units 1 and 2      3.2.4-2                          Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
 
QPTR 3.2.4 ACTIONS CONDITION              REQUIRED ACTION                      COMPLETION TIME A.  (continued)          A.6  -------------NOTE-------------
Perform Required Action A.6 only after Required Action A.5 is completed.
                                ----------------------------------
Perform SR 3.2.1.1 and              -----------NOTE----------
SR 3.2.2.1.                        Only one of the following Completion Times, whichever becomes applicable first, must be met.
                                                                    -----------------------------
Within 24 hours after reaching RTP OR Within 48 hours after increasing THERMAL POWER above the limit of Required Action A.1 and A.2.2 B. Required Action and  B.1  Reduce THERMAL                      4 hours associated Completion      POWER to  50% RTP.
Time not met.
Vogtle Units 1 and 2              3.2.4-3                          Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
 
QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.2.4.1        ---------------------------NOTE------------------------------
With one power range channel inoperable, the remaining three power range channels can be used for calculating QPTR.
                    -----------------------------------------------------------------
Verify QPTR is within limit by calculation.                        In accordance with the Surveillance Frequency Control Program AND Once within 12 hours and every 12 hours thereafter with the QPTR alarm inoperable SR 3.2.4.2        ----------------------------NOTE-----------------------------
Only required to be performed if input to QPTR from one or more Power Range Neutron Flux channels is inoperable with THERMAL POWER 75% RTP.
                    -----------------------------------------------------------------
Confirm that the normalized symmetric power                        Once within 12 hours distribution is consistent with QPTR.
AND In accordance with the Surveillance Frequency Control Program Vogtle Units 1 and 2                                3.2.4-4                      Amendment No. 158 (Unit 1)
Amendment No. 140 (Unit 2)
 
Approved By                                                                            Procedure Number Rev J.B. Stanley                      Vogtle Electric Generating Plant                  17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL      Page Number 08/16/2011                                          1C1 ON MCB                              65 of 66 WINDOW F06 ORIGIN                          SETPOINT DELTA FLUX YC-1140                        Plant Technical        DEVIATION Data Book, Tab 6 1.0                PROBABLE CAUSE
: 1.        Xenon transient
: 2.        Control rod motion
: 3.        Thermal power transient
: 4.        Loss of Power Range Detector voltage.
: 5.        IPC Failure 2.0                AUTOMATIC ACTIONS NONE 3.0                INITIAL OPERATOR ACTIONS NOTE The Delta Flux Deviation Program satisfies the requirement for Verify AFD within limits for each OPERABLE excore channel in Technical Specifications SR 3.2.3.1.
NONE Printed February 4, 2014 at 9:22
 
Approved By                                                                                      Procedure Number Rev J.B. Stanley                        Vogtle Electric Generating Plant                            17010-1      50 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL                Page Number 08/16/2011                                            1C1 ON MCB                                      66 of 66 WINDOW F06 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS NOTE This annunciator comes in immediately upon exceeding the limits for acceptable operation.
: 1.        With Reactor Power greater than 50% check differential flux indications and if two or more are outside the limits, perform the following:
: a.      Restore the indicated AFD to within the limits, or
: b.      Reduce Thermal Power to less than 50% of Rated Thermal Power within 30 minutes.
: 2.        Thermal Power shall not be increased above 50% of Rated Thermal Power until the indicated AFD is within the limits specified by the COLR.
: 3.        IF loss of Power Range Detector voltage is determined, Go To 18002-C, "Nuclear Instrumentation System Malfunction".
: 4.        Refer To Technical Specification LCO 3.2.3.
: 5.        IF alarm is inoperable, begin recording differential flux in accordance with 14915-1, "Special Condition Surveillance Logs" (Technical Specifications SR 3.2.3.1).
: 6.        Refer To Plant Computer alarm summary display for additional information relating to this alarm.
5.0                COMPENSATORY OPERATOR ACTIONS Initiate Data Sheet 6 of 14915-1, "Special Condition Surveillance Logs" END OF SUB-PROCEDURE
 
==REFERENCES:==
Technical Specification Section LCO 3.2.3 Printed February 4, 2014 at 9:22
: 1. 015AA1.23 001/LOIT/RO/M/F 3.1/3.2/015AA1.23/LO-TA-16007///
Initial condition:
          - Unit 1 is at 100% reactor power.
Current condition:
          - ALB08-E03 RCP 1 VIBRATION ALERT is received.
Which one of the following completes the following statement?
Per the applicable Annunciator Response Procedure, the operators are directed to monitor the RCP vibration readings on the __(1)__,
and per 13003-1, "Reactor Coolant Pump Operation," the RCP maximum operating limit for frame vibration is __(2)__ mils.
__(1)__                                __(2)__
A.              plant computer                                5 B.              plant computer                                20 C. local vibration monitoring panel                        5 D. local vibration monitoring panel                        20 K/A 015              RCP Malfunctions A1.04            Ability to operate and / or monitor the following as they apply to the Reactor Coolant Pump Malfunctions:
                        - RCP vibration K/A MATCH ANALYSIS The question tests the candidates ability to monitor RCP vibration by selecting the correct monitoring location and operate (ie, shut down) the RCP as needed based on vibration readings.
EXPLANATION OF REQUIRED KNOWLEDGE RCP vibration is monitored using Control Room annunciators ALB07-E03, E04, F03, and F04. These alarms come in at a FRAME vibration of 3 mils or a SHAFT vibration of 15 mils. ARP 17008-1 directs dispatching an operator to the local panel Wednesday, February 26, 2014 2:29:46 PM                                                      1
 
1-1201-P5-VMP to determine actual vibration levels. The IPC has a screen to monitor most of the RCP support parameters; however, RCP vibration cannot be monitored in the Control Room.
Per SOP 13003-1 Limitation 2.2.10, an RCP shall be stopped if SHAFT vibration exceeds 20 mils or FRAME vibration exceeds 5 mils. ARP 17008-1 supports these limits and directs the operator to SOP 13003-1. ARP 17008-1 contains additional requirements to shutdown the RCP if the rate of vibration increase exceeds 1 mil/hr SHAFT or .2 mils/hr FRAME.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. RCP vibration readings can only be obtained locally at panel 1-1201-P5-VMP. However, all other RCP operating parameters are displayed on the IPC. A candidate without specific knowledge of where and how to obtain vibration readings could assume that the IPC would display vibration data also.
The second part is correct. Per SOP 13003-1 Limitation 2.2.10, an RCP shall be stopped if the FRAME vibration exceeds 5 mils.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per SOP 13003-1 Limitation 2.2.10, an RCP shall be stopped if the FRAME vibration exceeds 5 mils. It also states that an RCP shall be stopped if SHAFT vibration exceeds 20 mils. It is common for operators to confuse the frame and shaft limits.
C. Correct.                  The first part is correct. RCP vibration readings can only be obtained locally at panel 1-1201-P5-VMP.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Wednesday, February 26, 2014 2:29:46 PM                                                            2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            015AA1.23 Importance Rating:              3.1 / 3.2 Technical
 
==Reference:==
SOP 13003-1, Rev 47.1, page 7 ARP 17008-1, Rev 18.0, pages 36 thru 38 References provided:            None Learning Objective:              LO-PP-16401, Rev 5.4, slides 30 & 31 LO-TA-16007 Obtain RCP vibration data in response to RCP vibration alarms per ARP 17008-1 LO-TA-16001 Start a RCP using 13003-1 Question origin:                BANK Cognitive Level:                M/F 10 CFR Part 55 Content:          41.7 / 41.10 Comments:
You have completed the test!
Wednesday, February 26, 2014 2:29:46 PM                                                      3
 
Approved By                                                                                  Procedure    Version M.G. Brill                        Vogtle Electric Generating Plant                          13003-1        47.1 Effective Date                                                                                Page Number 06/12/2013                        REACTOR COOLANT PUMP OPERATION                                      7 of 42 INITIALS 2.2.8            The following starting duty cycle for the RCP should be observed:                ________
Only one RCP shall be started at any one time.
Two successive starts are permitted, provided the motor is permitted to coast to a stop between starts.
A third start may be made when the winding and core have cooled by running for a period of 20 minutes, or by standing idle for a period of 45 minutes.
2.2.9            During RCS filling and venting, RCS pressure must be greater than 325 psig prior to starting an RCP to verify adequate seal D/P is maintained throughout RCS fill and vent. If necessary, the RCP should be stopped prior to seal D/P dropping less than 200 psid. If the seal D/P goes below 200 psid during pump operation or coast down, the RCP should be evaluated before restarting the RCP.                    ________
2.2.10          An RCP shall be stopped IF any of the following conditions exist:                ________
Motor bearing temperature exceeds 195&deg;F.
Motor stator winding temperature exceeds 311&deg;F.
Seal water inlet temperature exceeds 230&deg;F Total loss of ACCW for a duration of 10 minutes.
RCP shaft vibration of 20 mils or greater.
RCP frame vibration of 5 mils or greater.
Differential pressure across the number 1 seal of less than 200 psid.
2.2.11          If a loss of RCP seal cooling (Seal Injection and/or ACCW to Thermal barrier) occurs, resulting in RCP shutdown due to exceeding operating limits, then the unit should be cooled down to Mode 5 to facilitate recovery. Upon reaching Mode 5, ACCW to the Thermal barrier should be restored. Seal injection should then be returned to service.
This sequence should prevent seal damage, RCP shaft bowing, ACCW System damage, etc. due to excessive thermal stresses.                      ________
Printed November 11, 2013 at 14:58
 
Approved By                                                                          Procedure Number Rev J.B. Stanely                        Vogtle Electric Generating Plant                  17008-1        18 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON          Page Number 07/08/11                                        PANEL 1A2 ON MCB                            36 of 55 WINDOW E03 ORIGIN                          SETPOINT RCP 1 1-XE-0471A,B                    3 MILS FRAME            VIBRATION 1-XE-0471C,D                    15 MILS SHAFT          ALERT 1.0                PROBABLE CAUSE
: 1.      Pump Bearing failure.
: 2.      Pump Impeller - shaft assembly out-of-balance.
: 3.      Misalignment between Pump Shaft and Motor Shaft.
: 4.      RCS operating temperature below 500&deg;F.
2.0                AUTOMATIC ACTIONS NONE 3.0                INITIAL OPERATOR ACTIONS NONE Printed February 4, 2014 at 10:48
 
Approved By                                                                                    Procedure Number Rev J.B. Stanely                      Vogtle Electric Generating Plant                            17008-1        18 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON                    Page Number 07/08/11                                          PANEL 1A2 ON MCB                                    37 of 55 WINDOW E03 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS NOTE The Vibration Monitoring Panel displays auctioneered high vibration levels.
: 1.      Dispatch an operator to the Vibration Monitoring Panel 1-1201-P5-VMP to:
: a. Check both vibration channels and alarm setpoints for shaft and frame of RCP 1 (4 points in all) to verify no obvious vibration monitoring equipment problems exist.
: b. Notify maintenance to verify alarm condition.
: c. Log any RCP Vibration LEDs illuminated and any elevated vibration readings in Control Room Electronic Log.
NOTE If alarming condition has cleared, holding master reset in the depressed position for 2-3 seconds will clear all alarms for all RCPs.
CAUTION IF alarming condition has not cleared, system engineer should be contacted prior to approving resetting any alarms.
: d. When SS directs, attempt to reset alarm by pressing the Black Master Reset Button located on left side of bottom Card Panel above key switch.
: e. A condition report should be written to capture this event.
: 2.      Continue operation of affected RCP 1 and frequently monitor vibration.
: 3.      Refer to 13003-1, "Reactor Coolant Pump Operation" and shut down RCP 1 if rate of frame vibration increase exceeds .2 MILS/hour.
: 4.      Refer to 13003-1, "Reactor Coolant Pump Operation" and shut down RCP 1 if rate of shaft vibration increase exceeds 1 MIL/hour.
Printed February 4, 2014 at 10:48
 
Approved By                                                                    Procedure Number Rev J.B. Stanely                      Vogtle Electric Generating Plant            17008-1        18 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON    Page Number 07/08/11                                    PANEL 1A2 ON MCB                        38 of 55 WINDOW E03 (Continued) 5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB113, 1X6AB09-119, 1X3D-BD-M01A, 1X3D-CD-M10A, 1X6AB09-88, CX5DT101-176A, CX5DT101-176B Printed February 4, 2014 at 10:48
: 1. 022AK3.01 001/LOIT/RO/C/A 2.7/3.1/022AK3.01/LO-TA-16009/LO-PP-09///
Initial condition:
              - Unit 1 is at 100% reactor power.
Current condition:
              - CVCS makeup capability is lost.
Which one of the following completes the following statement?
As VCT level begins to slowly lower, VCT pressure is __(1)__ maintained.
If VCT pressure were to lower from 25 to 18 psig, RCP seal #1 leak-off flow rates would __(2)__.
__(1)__                                  __(2)__
A.              automatically                              decrease B.              automatically                              increase C.                  manually                                decrease D.                  manually                                increase K/A 022              Loss of Reactor Coolant Makeup AK3.01          Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup:
                          - Adjustment of RCP seal backpressure regulator valve to obtain normal flow K/A MATCH ANALYSIS The question examines the candidates knowledge of the operation of the VCT Hydrogen pressure regulator. VCT pressure establishes the backpressure on RCP seal leakoff. The question addresses the type of regulator (manual or automatic) and the reason for maintaining constant pressure.
EXPLANATION OF REQUIRED KNOWLEDGE At power, VCT Hydrogen pressure is established by Chemistry to regulate RCS Hydrogen concentration. The allowable pressure band is 18-45 psig. The lower limit is bounded by RCP seal #1 leakoff and the upper limit by boric acid emergency boration flow. Any change in VCT pressure affects RCP seal #1 leakoff, which in turn affects Friday, March 07, 2014 9:28:50 AM                                                            1
 
RCP seal #2 flow.
As VCT level lowers, the hydrogen regulator valve will open and automatically bring hydrogen pressure back to setpoint. As VCT level increases, hydrogen pressure increases and more gas flows into the waste gas system, which eventually brings pressure back down. If VCT level changes quickly, a change in hydrogen pressure will be observed because the control system response lags. If VCT level is changed slowly, hydrogen pressure will remain essentially constant.
VCT pressure creates a back pressure for RCP seal #1 leakoff. This backpressure is required to force water from the #1 seal up and through the #2 and #3 seals. As VCT pressure lowers, seal #1 leakoff increases and the flow to seal #2 decreases.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. As VCT level decreases, the hydrogen regulator valve will open and automatically bring hydrogen pressure back to setpoint.
The second part is incorrect and assumes the candidate does not understand the system configuration and relationship between RCP seal leakoff and the VCT. If the candidate does not know that VCT pressure provides backpressure to seal leakoff, then the candidate may find it reasonable to assume that as VCT pressure lowers, flow would also lower. In addition, the candidate may confuse the #1 and #2 seal response, and assume that as VCT pressure lowers, #1 seal flow decreases.
B. Correct.              The first part is correct. See the first part of choice A above.
The second part is correct. As VCT pressure lowers, back pressure on the #1 seal leakoff also lowers, resulting in an increase in seal #1 leakoff flow.
C. Incorrect. Plausible. The first part is incorrect. PCV-8156 is a regulator valve which controls VCT hydrogren pressure. However, a candidate unfamiliar with hydrogen makeup to the VCT could reasonably assume this regulator is a manual valve, since raising VCT pressure during normal operation requires manual adjustment of the regulator.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Friday, March 07, 2014 9:28:50 AM                                                                  2
 
Level:                    RO Tier # / Group #          T1 / G1 K/A#                      022AK3.01 Importance Rating:        2.7 / 3.1 Technical
 
==Reference:==
ARP 17007-1 Rev 29.1 SOP 13007-1, Rev 34.5, page 7 V-LO-PP-16401, Rev 5.4, slides 12 thru 14 References provided:      None Learning Objective:        LO-PP-09200-12 Describe how the following affects seal injection and seal return flow:
: a. RCS pressure changes
: b. VCT pressure changes
: c. Safe injection signal LO-TA-16009        Respond to abnormal RCP seal per 13003-1/2 Question origin:          NEW Cognitive Level:          C/A 10 CFR Part 55 Content:    41.5 / 41.10 / 45.6 / 45.13 Comments:
You have completed the test!
Friday, March 07, 2014 9:28:50 AM                                                          3
 
Approved By                                                                            Procedure    Version S. E. Prewitt                    Vogtle Electric Generating Plant                    13007-1      34.5 Effective Date                                                                        Page Number 02/21/2013                      VCT GAS CONTROL AND RCS CHEMICAL ADDITION                      7 of 64 INITIALS 4.0                INSTRUCTIONS 4.1                ALIGNING VCT HYDROGEN PURGE - NORMAL OPERATION 4.1.1              Request Chemistry to verify, by sample analysis, the nitrogen content in the VCT gas space.                                        ________
4.1.2              IF a nitrogen atmosphere exists in the VCT, align the tank for hydrogen purge operation per Section 4.3.                            ________
4.1.3              Request Chemistry to verify, by sample analysis, that the oxygen concentration in the VCT gas space is less than 2% by volume.        ________
4.1.4              IF the VCT oxygen concentration limit is approached, lower the oxygen content using Section 4.4.                                    ________
4.1.5              Verify a hydrogen atmosphere exists in the VCT as follows:
: a.        Open 1-2406-U4-001 HYDROGEN SUPPLY TO CVCS VCT ISOLATION AB-A24.                                      ________
: b.        Open 1-1208-U4-107 VCT Hydrogen Manifold Isolation, AB A24.                                                    ________
: c.        Close 1-1208-U4-108 VCT Nitrogen Manifold Isolation, AB A24.                                                    ________
: d.        Close 1-1208-U4-352 Waste Gas Decay Shutdown Tank Supply To VCT, AB A47.                                    ________
4.1.6              Check that VCT Hydrogen Regulator 1-PCV-8156 is set to maintain 18 psig or greater AB PIPE CHASE ROOM A24.                  ________
4.1.7              IF the Hydrogen Regulator requires adjustment, loosen set screw and adjust point to raise or lower pressure to maintain 18 psig or greater.                                                            ________
4.1.8              Verify the Gaseous Waste Processing System in operation, AND aligned to a Normal Gas Decay Tank, per 13201-1, "Gaseous Waste Processing System."                                            ________
4.1.9              Verify that the VCT Purge Flow Controller 1-HIC-1094 (1-PGPP),
is set at zero.                                                      ________
Printed January 28, 2014 at 12:34
 
Approved By                                                                                  Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                        17007-1      29.1 Effective  Date            ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 07 ON PANEL              Page Number 07/25/2012                                          1A2 ON MCB                                    50 of 51 WINDOW F05 ORIGIN                            SETPOINT VCT 1-PT-0115                        Hi: 45 psig                HI/LO PRESS Low: 18 psig 1.0              PROBABLE CAUSE
: 1.        High pressure:
: a.      High Volume Control Tank (VCT) level,
: b.      Hydrogen or Nitrogen Pressure Regulator malfunction.
(1-PCV-8156 or 1-PCV-8155).
: 2.        Low pressure:
: a.      Hydrogen or Nitrogen Pressure Regulator malfunction (1-PCV-8156 or 1-PCV-8155),
: b.      Open or leaking Vent Valve,
: c.      System leak.
2.0              AUTOMATIC ACTIONS                                      NOTE: Low VCT level is not listed as On low VCT pressure, 1-PV-0115 closes.                  a symptom for the low pressure 3.0              INITIAL OPERATOR ACTIONS                                condition. VCT pressure is NONE                                                    expected to be maintained automatically on a slow event like lowering level.
Printed September 20, 2013 at 11:05
 
Approved By                                                                                Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                        17007-1      29.1 Effective  Date            ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 07 ON PANEL            Page Number 07/25/2012                                          1A2 ON MCB                                  51 of 51 WINDOW F05 (Continued) 4.0              SUBSEQUENT OPERATOR ACTIONS
: 1.        Monitor VCT pressure and level using 1-PI-0115 and 1-LI-0185.
: 2.        IF VCT level is high:
: a.      Divert letdown flow to the Recycle Holdup Tank (HUT position) using 1-HS-0112A on the QMCB,
: b.      Operate makeup per 13009-1, "CVCS Reactor Makeup Control System."
: c.      WHEN desired pressure/level is obtained, place 1-HS-0112A to the AUTO position
: 3.        IF VCT level is normal, adjust VCT pressure per 13007-1, "VCT Gas Control And RCS Chemical Addition."
: 4.        IF a system leak is suspected, dispatch personnel to locate and isolate the leak.
: 5.        Return VCT pressure to normal as soon as possible.
: 6.        VCT pressure should be maintained below 45 psig for the BAT pumps to be relied upon as a boration flowpath.
: 7.        IF equipment failure is indicated, initiate maintenance as required.
5.0              COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB116-1, 1X4DB128, PLS Printed September 20, 2013 at 11:05
 
Objective 4 Demonstrate using the seal package model.
The seal package consist of three seals
: 1)  Number 1 seal (film riding) a) The primary seal b) Seal is accomplished with a hydrostatic film between the shaft runner and seal ring.
c) No mechanical contact between seal ring and shaft runner (must keep P >200 psid)
: 2) Number 2 seal (face rubbing) a) Provides back up for #1 seal b) Consist of carbon graphite (face rubbing seal) c) Graphite makes contact with runner which rotates with shaft d) If #1 seal fails , #2 seal converts to a film riding seal if #1 seal leak off valve is closed and seal is exposed to full RCS pressure. #2 seal designed to allow plant shutdown and should last approximately 24 hours.
e) Placing #2 seal in service with the RCP shaft still rotating will tend to score the shaft at the #2 seal area. This can require extensive repairs before placing the RCP back in service. Vogtle chooses to remove RCP from service and allow its shaft to come to a standstill before closing the #1 seal leak off valve to avoid this problem.
: 3) Number 3 seal (face rubbing) a) Prevents the leakage of liquid and gases from the RCS into containment.
b) Consist of carbon graphite seal which makes contact with runner (face rubbing) c) The runner is around the shaft and rotates with it.
d) The seal is actually two graphite sealing surfaces called dams.
V-LO-PP-16401                                                                                                            12
 
Objective 1d
: 2) RCP Motor Auxiliaries A) Motor Cooler
: 1) Containment Air is drawn into the motor by fan blades on motors rotor
: 2) It is then exhausted through the motor cooler
: 3) ACCW is the cooling medium used in the cooler (cools the outgoing air)
: 4) This arrangement limits containment air temperature rise and in turn limits motor temperature.
: 3) Flywheel A) Addressed in tech spec administrative section B) Stores rotational energy of the pump and motor while running then releases energy by maintaining pump motion to slowly reduce core flow following loss of power for core protection.
: 4) RCP motor space heater A) Each RCP motor has a electric resistance heater.
B) Used to prevent moisture accumulation in windings when motor is shutdown.
C) Not needed when motor is in operation because of heat generation from motor windings.
D) Heaters are automatically energized when either of the RCP motor breakers are opened.
E) Heaters are supplied from 480 V MCCs RCP motor burned up at Vogtle that was attributed to moisture from space heater breaker being open when pump was shutdown. The space heater did not energize therefore moisture accumulated while the pump was shutdown during the outage. Upon restart the motor windings shorted out. Motor rebuild was required.
V-LO-PP-16401                                                                                                      13
 
Objective 2 RCP Seal Injection a) Provided from CVCS b) 8 gpm per pump c) 5 gpm is directed through to lower pump radial bearing and into the RCS loop.
d) The remaining 3 gpm supplies #1 and #2 seals e) #3 seal injection is from small tanks called Standpipes. (Gravity Fed) f)  Flow path
: 1) 8 gpm from CVCS enters RCP at 2250 psig
: 2) 5 gpm passes through the lower pump bearing lubricating and cooling it.
: 3) Seal injection at 2250 psig prevents RCS water from escaping the loop.
: 4) 3 gpm is directed through #1 seal
: 5) A pressure drop at 2220 psid across the #1 seal occurs.
: 6) Approximately 3 gph (0.05 gpm) leak off from #1 seal is used as seal injection to #2 seal.
: 7) The remainder of #1 seal leak off is directed back to the VCT via seal water return.
: 8) 3 gph passes through #2 seal and the leak off is directed to the RCDT (~5-6 psig)
: 9) 800 cc/hr seal injection for #3 seal is provided by standpipe (~10 psig)
: 10) The standpipes are located at a higher elevation than the RCP and gravity feeds #3 seal; standpipes Auto fill from RMWST.
: 11) #3 seal injection is injected between the two dams and sealing surfaces.
: 12) #3 seal injection pressure is slightly higher than #2 seal injection leak off.
: 13) This prevents RCS liquids or gases from escaping to the containment environment.
: 14) #3 seal has two leak off paths a) The outer dam leak off (400 cc/hr) combines with #2 seal leak off and is routed to RCDT b) The inner dam leak off (400 cc/hr) is directed to the containment sump.
SMART - Solid Knowledge. If the #1 seal leakoff was isolated, the #2 seal would become a film riding seal due to increased pressure across the #2 seal facing.
V-LO-PP-16401                                                                                                            14
: 1. 022K3.02 001/LOIT AND LOCT/RO/C/A 3.0/3.3/022K3.02/LO-LP-36104-01///
Given the following:
          - Unit 1 is in Mode 3 following a steam line break in containment.
          - Only Train 'A' containment cooling units are available.
          - Containment temperature is 241&deg;F and slowly rising.
          - Containment pressure is 11 psig and slowly rising.
Which one of the following completes the following statement?
SG NR level instruments on the QMCB will indicate __(1)__ than actual level, and the instrument inaccuracies are a direct result of changes in containment __(2)__.
__(1)__                                  __(2)__
A.                    lower                                  pressure B.                    lower                                temperature C.                    higher                                  pressure D.                    higher                                temperature K/A 022              Containment Cooling System (CCS)
K3.02            Knowledge of the effect that a loss or malfunction of the CCS will have on the following:
                        - Containment instrumentation readings.
K/A MATCH ANALYSIS The question presents the candidate with a valid scenario in which a Secondary LOCA has occurred that affects containment temperature and pressure. The candidate is required to determine the effect on SG level indication as a result of elevated containment pressure and temperature originating from the design basis accident in conjuction with loss of Containment Coolers.
EXPLANATION OF REQUIRED KNOWLEDGE Wednesday, February 26, 2014 3:02:18 PM                                                    1
 
SG level transmitters utilize a closed reference leg with a condensing pot. As containment temperature increases, the reference leg density lowers, resulting in SG level indication reading higher than actual. EOPs require normal SG levels to be >10%
NR to ensure a heat sink is maintained. Per the WOG EOP Setpoint Documents, >19%
level must be added to compensate for impacts stemming from containment temperature under ADVERSE condtions. (Reference attached Westinghouse letter WWA5247.) Since the level transmitters utilize a closed reference leg, containment pressure has no direct impact on level indication.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. As containment temperature increases, the reference leg density lowers, resulting in SG level indication reading higher than actual. However, a candidate may either invert the density effect in the reference and variable legs or reverse the dP impacts between the two.
The second part is incorrect. Since the level transmitters utilize a closed reference leg, containment pressure has no direct impact on level indication. However, if the candidate believes SG level instruments use an open reference leg, this answer would be correct.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. As containment temperature increases, the reference leg density lowers, resulting in SG level indication reading higher than actual.
C. Incorrect. Plausible. The first part is correct. As containment temperature increases, the reference leg density lowers, resulting in SG level indication reading higher than actual.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 3:02:18 PM                                                              2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            022K3.02 Importance Rating:              3.0 / 3.3 Technical
 
==Reference:==
WOG Background Documents for Adverse Containment, Westinghouse letter WWA5247 References provided:            None Learning Objective:              LO-PP-29101-08 Describe routine actions taken to adjust Containment pressure and temperature.
LO-PP-29101-09 State the likely sources of Containment pressure increase during normal operations.
LO-LP-36104-01 List and describe four adverse environmental conditions that affect the reliability of instrumentation associated with critical plant parameters.
Question origin:                BANK - San Onofre 2006 NRC Question # 13 022K3.02.
Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 / 41.14 Comments:
You have completed the test!
Wednesday, February 26, 2014 3:02:18 PM                                                            3
 
As compared to ADVERSE numbers on the following sheets.
: 1. 024AG2.2.22 001/LOIT/RO/M/F 4.0/4.7/024AG2.2.22/LO-TA-09029///
Given the following:
          - Unit 1 requires Emergency Boration.
          - Shift Supervisor directs Emergency Boration through the BIT flow path.
Which one of the following completes the following statement?
For the selected Emergency Boration flow path, 13009-1, "CVCS Reactor Makeup Control System," directs the operator to establish a minimum flow rate to the RCS greater than __(1)__ gpm, and the boron concentration for the source used above is required to be between
__(2)__ ppm.
__(1)__                                    __(2)__
A.                      30                                    2400 - 2600 B.                      30                                    7000 - 7700 C.                    87.5                                    2400 - 2600 D.                    87.5                                    7000 - 7700 K/A 024              Emergency Boration G2.2.22          Knowledge of limiting conditions for operations and safety limits.
K/A MATCH ANALYSIS The question sets up a valid scenario in which an emergency boration is directed by the Shift Supervisor, requiring the candidate to determine the suction source, boron concentration, and required flow rate.
EXPLANATION OF REQUIRED KNOWLEDGE The Shift Supervisor directs Emergency Boration through the BIT flow path. This boration will be performed per SOP 13009-1 section 4.9.4. As such, the RWST will be aligned as the boration source and a minimum flow rate of 87.5 gpm is required. Per TS SR 3.5.4.3, RWST boron concentration must be between 2400 and 2600 ppm.
Conversely, if the SS directs boration through HV-8104 or FV-0110, the suction would Wednesday, February 26, 2014 3:03:46 PM                                                    1
 
be aligned to the BAST. Per SOP 13009-1, a flow rate of 30 gpm is required. Per TRM TRS 13.1.7.3, BAST concentration must be between 7,000 and 7,700 ppm.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Emergency boration through the BIT will be made with the suction aligned to the RWST and a flow rate of 87.5 gpm is required. However, if the candidate is unfamilar with the emergency boration flow paths and believes the CCP suctions will be aligned to the BAST, 30 gpm is be the correct answer.
The second part is correct. Per TS SR 3.5.4.3, RWST boron concentration must be between 2400 and 2600 ppm.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per TS SR 3.5.4.3, RWST boron concentration must be between 2400 and 2600 ppm. However, if the candidate believes boration is occuring from the BAST, a boron concentration of 7,000 - 7,700 ppm would be correct per TRM TRS 13.1.7.3.
C. Correct.                  The first part is correct. Emergency boration through the BIT will be made with the suction aligned to the RWST, and a flow rate of 87.5 gpm is required.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Wednesday, February 26, 2014 3:03:46 PM                                                            2
 
Level:                          RO Tier # / Group #                T1 / G2 K/A#                            024G2.2.22 Importance Rating:              4.0 / 4.7 Technical
 
==Reference:==
SOP 13009-1, Rev 49.0, pages 47 & 69 TS SR 3.5.4.3, Amendment No. 158, page 3.5.4-2 TRM TRS 13.1.7.3, Rev 1-10/22/98, page 13.1-12 References provided:            None Learning Objective:              LO-PP-60327-10 List the methods (flow paths/sources) available to Emergency borate the RCS.
LO-LP-39209-01 For any given item in section 3.5 of Tech Specs, be able to:
: a. State the LCO.
: b. State any one hour or less required actions.
LO-PP-09300-13 State the Technical Requirement, applicability, and any one hour or less actions for each of the following:
: f. TR 13.1.7, Borated Water Sources -
Operating LO-TA-09029      Perform Emergency Boration using 13009-1/2 Question origin:                NEW Cognitive Level:                M/F 10 CFR Part 55 Content:          41.5 / 43.2 /45.2 Comments:
You have completed the test!
Wednesday, February 26, 2014 3:03:46 PM                                                          3
 
RWST 3.5.4 ACTIONS (continued)
CONDITION                            REQUIRED ACTION                      COMPLETION TIME E. Required Action and            E.1          Be in MODE 3.                      6 hours associated Completion Time of Condition A or D        AND not met.
E.2          Be in MODE 5.                      36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                              FREQUENCY SR 3.5.4.1        --------------------------NOTE-------------------------------
Only required to be performed when ambient air temperature is < 40&deg;F.
                    -----------------------------------------------------------------
Verify RWST borated water temperature is                          In accordance with 44&deg;F and  116&deg;F.                                                the Surveillance Frequency Control Program SR 3.5.4.2        Verify RWST borated water volume is  686,000                      In accordance with gallons.                                                          the Surveillance Frequency Control Program SR 3.5.4.3        Verify RWST boron concentration is  2400 ppm                      In accordance with and  2600 ppm.                                                    the Surveillance Frequency Control Program SR 3.5.4.4        Verify each sludge mixing pump isolation valve                    In accordance with automatically closes on an actual or simulated                    the Surveillance RWST Low-Level signal.                                            Frequency Control Program Vogtle Units 1 and 2                                3.5.4-2                      Amendment No. 158 (Unit 1)
Amendment No. 140 (Unit 2)
 
Approved By                                                                          Procedure  Version J. B. Stanley                      Vogtle Electric Generating Plant                  13009-1      49 Effective Date                                                                        Page Number 08/09/2012                        CVCS REACTOR MAKEUP CONTROL SYSTEM                        47 of 69 INITIALS 4.9.4            Emergency Boration From The RWST Through The BIT Isolation Valves 4.9.4.1          Verify one (1) Charging Pump is running and supplied with cooling water.                                                        ________
4.9.4.2          Open the following Charging Pump Suctions from the RWST:
1-LV-0112D                                                      ________
1-LV-0112E                                                      ________
4.9.4.3          Close the following VCT Outlet Isolations:
1-LV-0112B                                                      ________
1-LV-0112C                                                      ________
4.9.4.4          Place 1-LV-0112A to the HUT position.                                ________
4.9.4.5          Open the following BIT DISCH ISOLATION valves:
1-HV-8801A                                                      ________
1-HV-8801B                                                      ________
4.9.4.6          Verify BIT Flow (1-FI-0917A), plus total seal injection flow, minus total seal return flow is greater than 87.5 gpm.                      ________
4.9.4.7          Adjust Charging Line Flow Controller 1-FIC-0121 as necessary to maintain RCP seal injection flow at maximum flow less than 13 gpm per pump.                                                      ________
4.9.4.8          IF required for RCS inventory control, place an additional letdown orifice in service per 13006-1.                                      ________
4.9.4.9          Operate the Pressurizer Backup Heaters as necessary to equalize boron concentrations between the RCS and the Pressurizer.                                                          ________
Printed November 11, 2013 at 14:41
 
Approved By                                                                                                                  Procedure    Version J. B. Stanley                                    Vogtle Electric Generating Plant                                            13009-1        49 Effective Date                                                                                                                Page Number 08/09/2012                                      CVCS REACTOR MAKEUP CONTROL SYSTEM                                                  69 of 69 TABLE 1 EMERGENCY BORATION FLOW PATH ALTERNATIVES Valve    Other Pump Flow path          BATP                                        Flows                    Flow                    Note Alignments  Required At least OPEN      Any charging 42 GPM                      30 GPM HV8104                                                                                              Operate heaters one  1HV-8104  pump        1FI-0121C                    1FI-0183A OPEN      Any charging Charging          At least                        42 GPM                      30 GPM 1FV-0110A  pump                                                          Operate heaters Flow path            one                          1FI-0121C                    1FI-0110A 1FV-0110B OPEN 1LV-0112D 1LV-0112E 8 to 13 GPM RWST to                    CLOSE      Any charging 100 GPM NA                                                        seal injection flow  Operate heaters Regen Hx                    1LV-0112B  pump        1FI-0121C 1HV-0182 1LV-0112C HUT 1LV-0112A OPEN 1LV-0112D 1LV-0112E 1HV-8801A              BIT flow (1FI-0917A) + total 1HV-8801B  Any charging seal flow                    Adjust 1FIC-0121C to RWST to BIT            NA                                                                              Operate heaters CLOSE      pump        - seal return flow          13 GPM per RCP 1LV-0112B              87.5 GPM 1LV-0112C HUT 1LV-0112A OPEN      RHR other RHR                                                                                              Establish water removal path to NA    HV-8812A/B than S/D    >100 gpm                    See Proc.
(Mode 6)                                                                                            prevent vessel overflow HV-8809A/B Cooling OPEN SI                    HV-8923A/B                                                                Establish water removal path to NA                    SI    >100 gpm                    See Proc.
(Mode 6)                  HV-8821A/B                                                                prevent vessel or cavity overflow HV-8835 Printed February 4, 2014 at 14:07
: 1. 026K1.01 001/LOIT/RO/M/F 4.2/4.2/026K1.01/LO-TA-13009///
Given the following:
            - Unit 1 reactor trip and SI occurred due to a LOCA.
            - Containment Spray (CS) actuated.
Which one of the following completes the following statement?
During the ECCS injection phase, CS pumps and ECCS pumps __(1)__ suction header(s) penetrating the RWST, and the CS pumps' sump suction valves __(2)__ automatically open on LO-LO RWST level.
__(1)__                          __(2)__
A.                  share a common                        will B.                  share a common                      will NOT C.                  have separate                        will D.                  have separate                      will NOT K/A 026              Containment Spray System (CSS)
K1.01            Knowledge of the physical connections and/or cause effect relationships between the CSS and the following systems:
                          - ECCS K/A MATCH ANALYSIS Question asks about the physical relationships between ECCS and the Containment Spray suction header from the RWST. Both systems share the same suction header from the RWST. The second part of the question is related to the design of the CSS when swapping suction sources and whether this is a manual or automatic swap.
EXPLANATION OF REQUIRED KNOWLEDGE CS and ECCS pumps share a common 24" suction pipe that penetrates the RWST.
This common header can be isolated by a single manual valve, 1-1204-U4-207. Once inside the Auxiliary Building, the common suction header branches off to supply the ECCS and CS pumps. Ref P&ID 1X4DB121.
Friday, March 07, 2014 1:01:43 PM                                                            1
 
ECCS containment sump suction valves automatically open on receipt of an RWST Lo-Lo Level signal following an SI actuation. The RWST semi-automatic swap over has a separate retentive circuit which allows reset of SI without resetting the RWST swap over function. During this swap over, the containment sump suction valves automatically open, and the RWST suctions remain open. The operator must close the RWST suctions with the handswitch. Ref Elementary 1X3D-BD-E03F for an example of this logic.
CS sump suction valves do not have automatic opening circuitry of any kind. These valves must be opened using the valve handswitch. Ref Elementary 1X3D-BD-J02G.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. The ECCS and CS pumps share a common 24" suction header that penetrates the RWST.
The second part is incorrect. The CS sump suction valves never automatically open. These valves open only when the handswitch is taken to the open position. However, the ECCS sump suction valves do automatically open. A candidate without specific knowledge of the CS sump suction valves could find it reasonable that the CS suction valves would behave the same as the ECCS valves to prevent the CS pumps from losing suction.
B. Correct.              The first part is correct. See the first part of choice A above.
The second part is correct. The CS sump suction valves never automatically open. These valves open only when the handswitch is taken to the open position.
C. Incorrect. Plausible. The first part is incorrect. The ECCS and CS pumps share a common 24" suction header that penetrates the RWST.
However, the common suction header branches off after entering the Auxiliary Building. The branches are not common between any of the ECCS or CS pumps. A candidate without specific knowledge of ECCS and CS piping could find it reasonable that the ECCS and CS pumps have separate piping to ensure train separation. Plant walkdowns can reinforce this incorrect assumption because all the branching occurs well before the individual pump rooms.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Friday, March 07, 2014 1:01:43 PM                                                                  2
 
Level:                    RO Tier # / Group #          T2 / G1 K/A#                      026K1.01 Importance Rating:        4.2 / 4.2 Technical
 
==Reference:==
P&ID 1X4DB121, Rev 42.0 Elementary 1X3D-BD-E03F, Rev 9.0 Elementary 1X3D-BD-J02G, Rev 6.0 References provided:      None Learning Objective:        V-LO-PP-15101 Containment Spray System LO-LP-37113-02 Using EOP 19013 as a guide, briefly describe how each step is accomplished.
LO-PP-15101-04 List all components that receive a Containment Spray Actuation signal and their change in status.
LO-PP-15101-02 Describe what will actuate the Containment Spray System, including coincidence and set point.
LO-TA-13013        Draw functional diagram of ECCS LO-TA-15005        Draw the Containment Spray System LO-TA-13009        Manually align ECCS for Cold Leg Recirculation Phase using EOP 19013-C Question origin:          BANK - Farely 2010 NRC Question # 026K1.01 Cognitive Level:          M/F 10 CFR Part 55 Content:    41.2 to 41.9 / 45.7 to 45.8 Comments:
You have completed the test!
Friday, March 07, 2014 1:01:43 PM                                                            3
 
Common suction piping that penetrates the RWST.
 
Open circuitry does not contain any Sump Suction HS SSPS contacts.
does not have an Manual open only.
AUTO position.
 
RWST LO-LO level contacts from SSPS
: 1. 026K2.02 001/LOIT/RO/C/A 2.7/2.9/026K2.02/LO-PP-11101-55///062K1.02 Given the following:
          - An SI occurred and has NOT been reset.
          - An LOSP occurs a few minutes later.
          - 1AA02 is powered from DG1A.
          - 1BA03 is powered from DG1B.
While the DGs are operating, an electrical perturbation results in the following:
          - DG1A 186A lockout relay energizes (Generator Differential).
          - DG1B 186B lockout relay energizes (Phase Overcurrent).
Which one of the following describes the current condition of the Containment Spray discharge valves?
1HV-9001A, Containment Spray Pump 'A' Discharge Isolation, is __(1)__,
and 1HV-9001B, Containment Spray Pump 'B' Discharge Isolation, is __(2)__.
__(1)__                                    __(2)__
A.                energized                                    energized B.                de-energized                                  energized C.                energized                                  de-energized T
D.                de-energized                                de-energized K/A 026              Containment Spray K2.02            Knowledge of bus power supplies to the following:
                        - MOVs K/A MATCH ANALYSIS The question tests the candidate's knowledge of power supplies which feed the Containment Spray MOVs by requiring the candidate to determine if the MOV will be energized following 186 protective relay actuations concurrent with SI.
Wednesday, February 26, 2014 3:07:09 PM                                                    1
 
EXPLANATION OF REQUIRED KNOWLEDGE Per SOP 13145A/B-1/2 Precautions and Limitations, with SI present and a 186A relay actuation, the associated DG output breaker will open and the DG will shutdown. With SI present and a 186B relay actuation, the associated DG output breaker will remain closed and the DG will continue to operate. Therefore, 1AA02 will de-energize and 1BA03 will remain energized.
The one-line power for 1HV9001A is 4160V SWGR 1AA02 -> 480V SWGR 1AB15 ->
480V MCC 1ABD.
The one-line power for 1HV9001B is 4160V SWGR 1BA03 -> 480V SWGR 1BB16 ->
480V MCC 1BBD.
[Note - The valve titles used are from 13115 not the QMCB (they are not worded the same).]
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. The Generator Differential (186A) lockout relay will emergency trip DG1A under all conditions, de-energizing bus 1AA02, and removing power from 1HV-9001A.. However, 186 B and C lockout relays also exist, each causing the DG and output breaker to behave differently based on the presence of an LOSP and/or a SI signal. These lockouts are easily confused. This distractor would be correct for either 186 B or C lockout relay.
The second part is correct. With SI present and a 186B relay actuation, the 1B DG output breaker will remain closed and the DG will continue to operate.
B. Correct.                  The first part is correct. With SI present and a 186A relay actuation, the 1A DG output breaker will open and the DG will shut down.
The second part is correct. See the second part of choice A above.
C. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Phase Overcurrent (186B) would not trip DG1B or its output breaker because this trip is not active during SI conditions. However, 186 A and C lockout relays also exist, each causing the DG and output breaker to behave differently based on the presence of an LOSP and/or a SI signal. These lockouts are easily confused. This distractor would be correct for a 186A lockout relay.
D. Incorrect. Plausible. The first part is correct. See the first part of choice B above.
Wednesday, February 26, 2014 3:07:09 PM                                                              2
 
The second part is incorrect. See the second part of choice C above.
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            026K2.02 Importance Rating:              2.7 / 2.9 Technical
 
==Reference:==
13145A-1 Rev 6, page 13 1X3D-AA-F12A Rev 18.0 1X3D-AA-F12B Rev 6.0 1X3D-AA-E17A Rev 12.0 1X3D-AA-F11A Rev 24.0 1X3D-AA-F11B Rev 10.0 1X3D-AA-E16A Rev 9.0 References provided:            None Learning Objective:              LO-PP-11101-55 Identify the primary relays which will actuate each of the following lockout relays and how the diesel generator will respond to each normal start from SI, UV, Local Emergency Start, and Normal Start.
: a. 186A lockout relay
: b. 186B lockout relay LO-TA-01023        Draw the electrical distribution system Question origin:                MODIFIED - HL18 NRC Question # 062K1.02 Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 Comments:                        Question appears to match the KA. Answer choices could be cleaned up by asking:
1HV-9001A is __(1)___ and 1HV-9001B is __(2)___.
Then all choices can toggle between energized/de-energized. Question appears to be okay.
                                        -JAT 12/19/2013 (Editorial)
Comment incorporated. SAT
                                        -JAT 2/4/14 You have completed the test!
Wednesday, February 26, 2014 3:07:10 PM                                                            3
: 1. 062K1.02 001/2/1/AC DIST - EDG/C/A - 2.5/MODIFIED/R/NRC RO/TNT / RLM Given the following:
          - An SI has occurred and is NOT reset.
          - An LOSP occurs a few minutes later.
          - 1AA02 is powered from DG1A
          - 1BA03 is powered from DG1B While the DG's are operating, an electrical perturbation results in the following;
          - DG1A 186A lockout relay energizes (Generator Differential)
          - DG1B 186B lockout relay energizes (Phase Overcurrent)
Which ONE of the following is CORRECT with respect to the status of power to the 4160 1E Emergency Buses at this time?
A. Both 4160 1E emergency buses would be energized.
B. Both 4160 1E emergency buses would be de-energized.
C. 1AA02 would be energized, 1BA03 would be de-energized.
D. 1BA03 would be energized, 1AA02 would be de-energized.
Monday, December 02, 2013 8:11:24 AM                                                      1
 
Approved By                                                                              Procedure    Version C.H. Williams                      Vogtle Electric Generating Plant                      13145A-1      6 Effective Date                                                                            Page Number 06/25/2013                              DIESEL GENERATOR TRAIN A                                13 of 85 2.2.18            The following table lists the DG Lockout Relays and the related functions:
LOCKOUT PRIMARY RELAYS          BREAKER STATUS            ENGINE STATUS RELAYS Generator            Trips Open              Shuts Down 186A Differential            Always                  Always Trips Open on Normal Start,          Shuts Down Local Emergency            on Normal Start Phase Overcurrent Start, or LOSP 186B                    or Loss of Field Remains Running Remains Closed on LOSP, Local Emergency on SI Start, or SI Reverse Power or Trips Open 186C          Negative Phase                                  Remains Running in Parallel Mode Only Sequence
________
Printed December 2, 2013 at 8:18
 
1HV9001A is breaker #48 on drawing 1X3D-AA-F11B
 
1HV9001B is breaker #48 on drawing 1X3D-AA-F12B.
: 1. 027AK2.03 001/LOIT/RO/C/A 2.6/2.8/027K2.03////
Initial conditions:
          - Unit 1 is at 100% reactor power.
          - One group of pressurizer backup heaters is energized in MANUAL.
          - Pressurizer pressure control is selected to CH 455 / 456.
Current condition:
          - Pressurizer pressure transmitter, 1PT-455, fails LOW.
Which one of the following completes the following statement?
In response to the failure, the remaining pressurizer backup heaters will __(1)__,
and the pressurizer spray valve controllers will demand full __(2)__ position.
__(1)__                                __(2)__
A.                  energize                                open B.                  energize                                closed C.                remain off                                open D.                remain off                              closed K/A 027              Pressurizer Pressure Control System Malfunction K2.03            Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following:
                        - Controllers and positioners K/A MATCH ANALYSIS The question requires the candidate to determine the response of the pressurizer heaters and the spray controller demand to a loss of a pressurizer pressure transmitter.
EXPLANATION OF REQUIRED KNOWLEDGE The pressurizer pressure control circuit utilizes primary and secondary selectable control loops. With 455 / 456 selected, PT-455 is the input to the primary control loop.
PT-455 failing low will result in the pressurizer master controller demand lowering, resulting in all heaters on. The associated pressurizer spray slave controller demand Wednesday, February 26, 2014 3:08:50 PM                                                      1
 
will also lower, demanding full close on the spray valves. Reference 18000-C, Figure 1, for a diagram of the heater control circuit.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. PT-455 is the primary controlling channel which inputs to the heaters and the spray valves to maintain a constant 2235 psig. When PT-455 fails low, the control circuit will sense a low pressure condition and energize the heaters in response.
The second part is incorrect. PT-455 is the primary controlling channel which controls the spray valves to maintain a constant 2235 psig. When PT-455 fails low the control circuit will sense a low pressure condition and energize the heaters in response.
However, a candidate without specific knowledge of the spray control circuit may assume that heaters and sprays are controlled by the primary and secondary loops respectively. In this case, PT-456 would see increasing pressure and call for an increase in spray flow.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. PT-455 is the primary controlling channel, which inputs to the spray valve controllers to maintain a constant 2235 psig. When PT-455 fails low the master controller will send a signal to the slave controllers to close the spray valves.
C. Incorrect. Plausible. The first part is incorrect. PT-455 is the primary controlling channel which inputs to the heaters and the spray valves to maintain a constant 2235 psig. When PT-455 fails low, the control circuit will sense the low pressure condition and energize the heaters in response. However, the candidate may confuse which channel inputs to the primary circuit and believe PT-456 is controlling. In this case, no change in heater condition is expected.
The second part is incorrect. When PT-455 fails low the control circuit will sense the low pressure condition and energize the heaters in response. However, a candidate may believe that, because the backup heater is energized in manual, the resulting failure will result in an actual high pressure condition. If the candidate confused which channel inputs to the primary circuit and believes PT-456 is controlling, he could determine that this would cause the spray valves to open to mitigate the pressure increase.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B Wednesday, February 26, 2014 3:08:50 PM                                                                2
 
above.
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            027AK2.03 Importance Rating:              2.6 / 2.8 Technical
 
==Reference:==
LOGIC 1X6AA02-235, Rev 9.0 LOGIC 1X6AA02-236, Rev 7.0 AOP 18000-C, Rev 5.0, page 5 References provided:            None Learning Objective:              LO-LP-60301-10 Given that the channel selector switch is in the NORMAL position (455/456),
describe how and why the plant will respond to the following pressurizer pressure instrument failures. Consider each separately and include effects on the Pressurizer Pressure Control System response, alarms, RPS, and ESF actuations.
: a. 455 fails high
: b. 455 fails low
: c. 456 fails high Question origin:                MODIFIED, HL-15 NRC Question # 027AK2.03 Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 / 41.10 / 45.3 Comments:
You have completed the test!
Wednesday, February 26, 2014 3:08:50 PM                                                            3
: 1. 027AK2.03 001/1/1/PZR PRESS CNTRL/2.6/2.8 C/A/LORQ BANK/RO/SRO/NRC/GCW Given the following:
          - The unit is at 100% power.
          - All control systems are in their normal alignments.
          - The Pressurizer Master Pressure Controller output demand fails LOW.
Assuming no action has been taken by the crew, which ONE of the following describes the effect on the Pressurizer heaters, and the resulting effect on the plant?
A. PZR heaters energize.
PZR pressure rise is controlled by PZR PORV operation.
B. PZR heaters energize.
PZR pressure rise is controlled by PZR spray valve operation.
C. PZR heaters de-energize.
ONLY the PZR spray valves open, reactor trips on low PZR pressure.
D. PZR heaters de-energize.
PZR spray valves and one PZR PORV open, reactor trips on low PZR pressure.
Wednesday, February 05, 2014 1:35:53 PM                                                  1
 
Approved By                                                              Procedure Number Rev C. S. Waldrup                        Vogtle Electric Generating Plant  18000-C          5 Date Approved                                                            Page Number PRESSURIZER SPRAY, SAFETY, OR RELIEF 2/3/09                                    VALVE MALFUNCTION                      5 of 5 FIGURE 1              Sheet 1 of 1 PRESSURIZER PRESSURE CONTROLLER BAND Printed February 5, 2014 at 13:29
: 1. 028K1.01 001/LOIT/RO/C/A 2.5/2.5/028K1.01/LO-TA-29013///
Initial conditions:
          - Unit 1 is at 100% reactor power.
          - Containment Spray pump 'A' is tagged out.
Current conditions:
          -  Large break LOCA occurs.
          -  19010-C, "Loss of Reactor or Secondary Coolant," is in progress.
          -  Containment Spray pump 'B' trips.
          -  Containment pressure is 22 psig.
          -  Containment hydrogen concentration is 5.1%.
Which one of the following completes the following statement?
The preferred method for reducing containment hydrogen concentration is __(1)__
containment atmosphere, and a __(2)__ path entry condition exists for 19251-C, "Response to High Containment Pressure."
__(1)__                              __(2)__
A.                  diluting                              red B.                  diluting                            orange C.                  purging                                red D.                  purging                              orange K/A 028              Hydrogen Recombiner and Purge Control System K1.01            Knowledge of the physical connections and/or cause-effect relationships between the HRPS and the following systems:
                        - Containment annulus ventilation system (including pressure limits).
K/A MATCH ANALYSIS The question presents the candidate with a plausible scenario in which conditions are provided in the stem that the candidate must use to determine the proper action based on his knowledge of the procedure steps and limitations, including specific alignments Wednesday, February 26, 2014 3:49:04 PM                                                      1
 
allowed per the SOP. This also includes pressure limits allowed to perform the process and which method is preferred. In addition, the candidate must use the information provided to evaluate the challenge to the containment barrier and determine the proper procedural response.
EXPLANATION OF REQUIRED KNOWLEDGE With containment hydrogen concentration at 5.1%, hydrogen concentratin must be reduced. Per SOP 13130-1, two methods are available to accomplish this task. The "preferred" method dilutes containment atmosphere with service air, and is the preferred method since it does not create an emergency release. However, this method is not to be performed if containment pressure is >40 psig, unless exceptional circumstances exist. The second method is also governed by SOP 13130-1 and requires Emergency Director approval prior to initiation because it results in an emergency exposure to the general public and the plant population.
Critical Safety Function Status Tree (CSFST) priority of operator action is based on the following colors in descending priority order: Red, Orange, Yellow, and Green. The Containment CSFST distinguishes between red and orange paths based on containment pressure above or below 52 psig. The distinction between orange and yellow paths is based on containment pressure above or below 21.5 psig, and whether a CSP is running. Based on containment pressure at 22 psig and neither CSP available, an Orange path is present per EOP 19200-C.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. Per SOP 13130-1, with containment pressure <40 psig, diluting containment with service air is the preferred method to prevent an emergency release.
The second part is incorrect. Per 19200-C, with containment pressure at 22 psig and no CS pumps available, an ORANGE path exists on the CSFST. However, if the candidate confused the first and second level of the containment pressure decision tree and believes the first decision is based on >21.5 psig, then a RED path would be correct.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. Per 19200-C, with containment pressure at 22 psig and no CS pumps available, an ORANGE path exists on the CSFST.
C. Incorrect. Plausible. The first part is incorrect. Per SOP 13130-1, with containment pressure <40 psig, diluting containment with service air is the preferred method to prevent an emergency release. However, a candidate without specific knowledge of the 40 psig limit may find it non-conservative to increase containment pressure with service air when it is currently >21.5 psig.
The second part is incorrect. See the second part of choice A Wednesday, February 26, 2014 3:49:04 PM                                                              2
 
above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Level:                          RO Tier # / Group #                T2 / G2 K/A#                            028K1.01 Importance Rating:              2.5 / 2.5 Technical
 
==Reference:==
SOP 13130-1, Rev 20.0, pages 11 & 14 EOP 19200-C, Rev 24.2, page 9 References provided:            None Learning Objective:              LO-LP-36107-03 State the means available to measure and control the containment hydrogen concentration LO-PP-29101-03 List the systems that are designed to control and mitigate hydrogen gas buildup in Containment.
LO-PP-29101-18 State the upper and lower limits for an explosive mixture of hydrogen.
LO-TA-29013        Perform a Dilution of Containment Hydrogen Concentration Using the Service Air System per 13130-1/2 LO-TA-29012        Operate the Post-LOCA Containment Hydrogen Purge system per 13130-1/2 Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.2 to 41.9 Comments:
You have completed the test!
Wednesday, February 26, 2014 3:49:04 PM                                                          3
 
Approved By                                                                Procedure    Version J. B. Stanley                          Vogtle Electric Generating Plant    19200-C        24.2 Effective Date                                                              Page Number F-0 CRITICAL SAFETY FUNCTION STATUS TREES 7/25/12                                                                            9 of 11 If candidate        Sheet 1 of 1 confuses the 52 psig tree with the 21.5 psig tree.
Printed February 5, 2014 at 14:30
 
Approved By                                                                            Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                    13130-1 20 Effective Date                                                                          Page Number 07/19/2012                          POST-ACCIDENT HYDROGEN CONTROL                            11 of 22 INITIALS 4.3                SHUTDOWN NONE 4.4                NON PERIODIC OPERATION 4.4.1              Deleted 4.4.2              Diluting Containment Hydrogen Concentration Using The Service Air System NOTES Containment design pressure is 52 psig.
CAUTION Do not perform this section if containment pressure is greater than 40 psig unless so directed by the Emergency Director.
4.4.2.1            Reset CIA by taking the following handswitches to RESET and observe ALB06-E06 extinguished:
1HS-40120                                                  ________
1HS-40122                                                  ________
4.4.2.2            Open SERVICE AIR CNMT HDR ISOL 1-HV-9385 as follows:
: a.      Place 1-HS-9385A on Main Control Room Panel QPCP to OPEN.                                                      ________
: b.      Hold 1-HS-9385B on Panel QPCP in OPEN until 1-HV-9385 is fully open.                                    ________
Printed February 5, 2014 at 14:27
 
Approved By                                                                              Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                      13130-1 20 Effective Date                                                                          Page Number 07/19/2012                          POST-ACCIDENT HYDROGEN CONTROL                              14 of 22 INITIALS 4.4.3              Post-LOCA Containment Hydrogen Purge System Operation CAUTIONS The Post-LOCA Containment Hydrogen Purge System is to be operated only if the containment hydrogen concentration cannot be maintained below 4% by other means.
The Post-LOCA Containment Hydrogen Purge System is designed to operate with a maximum pressure of 3 psi downstream of CNMT POST LOCA PURGE EXH DUCT CONTROL VLV 1-FV-2693.
Service air header pressure should be maintained greater than 80 psig to prevent header isolation while performing this section 4.4.3.1            Obtain Emergency Director approval to perform this section:          ________
4.4.3.2            Verify the Service Air System is operating.                          ________
4.4.3.3            Place disconnect switch at local Heater Control Panel 1-1508-N7-001-H01 to ON.                                              ________
4.4.3.4            Push RESET button at local Heater Control Panel 1-1508-N7-001-H01 and verify that reset red light is ON.              ________
Critical 4.4.3.5            Due to high radiation area potential, verify Containment Inside Isolation Valves 1-HV-2624A and 1-HV-2624B are closed and remain closed during the performance of the next step and until personnel have exited the area.                                      ________
________
CV 4.4.3.6            Unlock and open POST LOCA PURGE CTB ISO VALVE 1-1508-U4-012. (KEY# 1OP3-381)[Equip. Bldg. roof (Dog House)]        ________
4.4.3.7            Reset CVI by placing the following handswitches to the RESET position:
1HS-40121                                                    ________
1HS-40123                                                    ________
Printed February 5, 2014 at 14:27
: 1. 029EG2.4.34 001/LOIT AND LOCT/RO/C/A 4.2/4.1/029EG2.4.34/LO-TA-37014///
Initial condition:
            - Unit 1 is at 100% reactor power.
Current condition:
            - ATWT is in progress.
            - The reactor trip and bypass breakers will NOT open locally.
Which one of the following completes the following statement?
Per 19211-C, "Response to Nuclear Power Generation / ATWT," the Control Rod Drive MG Set __(1)__ breakers will be opened, and P-4 __(2)__ be generated when the MG set breakers are locally opened.
__(1)__                                  __(2)__
A.                    input                                    will B.                    input                                  will NOT C.                    output                                    will D.                    output                                  will NOT K/A 029              ATWS G2.4.34          Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.
K/A MATCH ANALYSIS The question tests the candidate's knowledge of the correct method to locally trip the reactor during an ATWS, and the resultant operational impact on turbine trip / reactor trip due to the P-4 signal not being generated because the reactor trip breakers remained closed.
EXPLANATION OF REQUIRED KNOWLEDGE Per EOP 19211-C step 8, if the reactor has not tripped, the operator is directed to open the reactor trip and bypass breakers locally. If this action is unsuccessful, the operator is directed to open the MG set output breakers locally. The breakers can only be Wednesday, February 26, 2014 3:52:59 PM                                                        1
 
operated locally and are immediately upstream of the RTBs. Opening these breakers will result in an immediate removal of power to the rod control cabinets and the reactor will trip. Conversely, if the MG set input breakers are opened, power is not immediately removed. Instead, the MG set will slowly coast down due to the large fly-wheel. As voltage decays, the stationary grippers will lose power and open, and the rods will fall.
Since the voltage threshold can differ for each gripper, industry experience has shown that the rods will fall randomly and can result in significant local power peaking and fuel damage. (Reference V-LO-PP-27101 for a simplified diagram of Rod Control power.)
Since the reactor trip breakers are not open, a P-4 signal will not be generated. As a result, a turbine trip signal will not be generated on the reactor "trip" when the MG set output breakers are open. A manual turbine trip is required to ensure an excessive RCS cooldown does not occur when the reactor trips. (Reference 1X6AA02-00226 &
240)
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per EOP 19211-C step 8, the operator is directed to locally open the MG set output breakers.
However, the MG set input breakers can be operated from the control room and are obviously labeled on the QEAB panel (see attached pictures of handswitches). Historically, prior to the industry OE, Vogtle's ATWS EOP directed opening the MG set input breakers.
The second part is incorrect. Reactor trip breakers remain closed resulting in neither train P-4 signal (which inserts the Turbine trip) being generated. The MG set breakers do not feed into the P-4 circuit. However, candidates normally think of a turbine trip resulting from a reactor trip. Since the reactor is tripped, it is reasonable for a candidate who does not know how P-4 is generated to assume P-4 would exist simply because the control cabinets are de-energzied.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. Reactor trip breakers remain closed resulting in neither train P-4 being generated. The MG set breakers do not feed into the P-4 circuit or directly cause a turbine trip.
C. Incorrect. Plausible. The first part is correct. Per EOP 19211-C step 8, the operator is directed to locally open the MG set output breakers.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 3:52:59 PM                                                                2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            029EG2.4.34 Importance Rating:              4.2 / 4.1 Technical
 
==Reference:==
EOP 19211-C, Rev , page LOGIC 1X6AA02-00226, Rev 9.0 LOGIC 1X6AA02-00240, Rev 9.0 Lesson Plan V-LO-PP-27101, Rev 2.0, slide 32 Pictures of HS-1NB0801 and HS-1NB0901 References provided:            None Learning Objective:              LO-PP-28103-02 List all permissives with applicable set points, coincidences, and functions.
LO-PP-28103-03 List all reactor trip set points, coincidences, permissives, and blocks.
LO-PP-27101-02 State the power supplies for the Rod Control System.
LO-TA-37014      Respond to a Nuclear Power Generation/
ATWT Condition per 19211-C LO-TA-27008      Draw and label a one-line diagram of the Control Rod Drive Power Supply Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 / 41.10 Comments:
You have completed the test!
Wednesday, February 26, 2014 3:52:59 PM                                                        3
 
Approved By                                                                          Procedure Number Rev J. D. Williams                        Vogtle Electric Generating Plant                19211-C          20.5 Date Approved                                                                          Page Number FR-S.1 RESPONSE TO NUCLEAR POWER 1-23-2007                                  GENERATION/ATWT                                  6 of 20 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 7
      *7.      Check for SI:                              7.
7.a
__a. SI signal - EXISTS OR                        __a. IF an SI signal is actuated ACTUATED.                                      during this procedure, THEN initiate ATTACHMENT A.
__    Go to Step 8 7.b
: b. Initiate ATTACHMENT A.
8
: 8.      Check the following trips have              8.
occurred:
8.a
__a. Reactor trip.                                __a. Locally trip the Reactor trip and Bypass breakers.
__    IF the trip breakers will NOT open, THEN trip the Control Rod Drive MG Set output breakers at the Reactor Trip Switchgear.
8.b
__b. Turbine trip.                                __b. Dispatch operator to trip turbine at the HP Turbine front standard.
9
      *9.      Check Reactor power:                        9.
9.a
__a. LESS THAN 5%.                                __a. Go to Step 10.
9.b
__b. IR SUR - LESS THAN 0 DPM.                    __b. Go to Step 10.
9.c
__c. Go to Step 24.                                  c.
 
S Printed February 6, 2014 at 10:55
 
MG Set Output Breakers (ND31-09)                    (ND31-04)
CONTROL POWER - SSPS                    CONTROL POWER - SSPS CHARGING SPRING - AD11-09                CHARGING SPRING - BD11-09 NB09-10                                                                    RTA      RTB MG #1 260vac                        BYA      BYB                        DC HOLD NB08-10                                                                                                        CABINET 1CB (ND32-01)          MG #2 (ND32-06)                                                                          S1 150vac/120vac MG Set Input                                                                                                                    125vdc 70vdc Breakers                              SURGE PROTECTOR fuses FU1a&b fuses FU3a&b 100vdc    +16.5 vdc -16.5 vdc battery PS1                +28vdc PS1
                                                                                                  -24vdc Auto/manual                          PS3 fuse F7A reset LOGIC CABINET fuse F5A CARD FRAME 100vdc    +16.5 vdc -16.5 vdc fuse F6A                                                OFF LATCH battery PS2                                                          HOLD fuse F8A        -24vdc PS4
                                                                                +28vdc PS2 SURGE PROTECTOR                              Auto/manual                                                        POWER reset                          fuses FU4a&b NYS 27 fuses FU2a&b CABINET LIFT COIL MOVABLE                      STATIONARY GRIPPER                      GRIPPER V-LO-PP-27101 Rev-2.0                                                                                                                32
: 1. 033AG2.2.44 001/LOIT/RO/C/A 4.2/4.4/033AG2.2.44/LO-TA-60035///
Initial conditions:
          - Unit 1 reactor startup is in progress.
          - Critical data is being collected per 12003-C, "Reactor Startup (Mode 3 to Mode 2)."
Current condition:
          - Intermediate Range N35 fails bottom of scale.
Which one of the following completes the following statement?
Based on the current conditions, the SR BLOCK PERMISSIVE P-6 light on the BPLB
__(1)__ remain lit, and placing the affected channel's Level Trip switch in BYPASS per 18002-C, "Nuclear Instrumentation System Malfunction," aligns NI __(2)__ power to the SSPS input relay.
__(1)__                                    __(2)__
A.                      will                                    control B.                      will                                  instrument C.                  will NOT                                    control D.                  will NOT                                  instrument K/A 033              Loss of Intermediate Range NI G2.2.44          Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
K/A MATCH ANALYSIS The question has the candidate determine the status of P-6 on the BPLB based on the plant conditions. The question then requires the candidate to demonstrate understanding of the effect of placing the NI channel in BYPASS during the recovery actions.
EXPLANATION OF REQUIRED KNOWLEDGE At the point of taking critical data, reactor power is 2 x 10-3. P-6 permissive has been Wednesday, February 26, 2014 3:55:36 PM                                                        1
 
met and SR Hi Flux trips have been blocked. When N35 fails low, one of two P-6 bistables de-energizes. However, one of two bistables remains and maintains the P-6 permissive. Therefore, the P-6 BPLB remains lit.
Per 18002-C direction, the operating crew will place the affected channel's LEVEL TRIP switch to the BYPASS position. This aligns control power directly to the SSPS input bay contacts, preventing the NI channel from causing a tripped input and allowing I&C to troubleshoot the instrument.
There are two fuses on the NI drawers - control and instrument. The control power fuse will de-energize the instrument power as well as all the bistables in the NI drawer and allow SSPS to sense a tripped condition. The instrument power fuse will cause the instrument output to fail downscale low and the NI bistables to trip. However, the input bay of SSPS is unchanged.
ANSWER / DISTRACTOR ANALYSIS A. Correct.                  The first part is correct. When N35 fails low, one of two P-6 bistables de-energizes. However, one of two bistables remains and maintains the P-6 permissive. Therefore, the P-6 BPLB remains lit.
The second part is correct. The control power fuse will de-energize the instrument power as well as all the bistables in the NI drawer and allow SSPS to sense a tripped condition.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. Control power feeds the bistables in the NI drawers. However, there are two power sources to the NI drawers - control and instrument. These power sources are frequently confused by intial candidates as to how they function.
C. Incorrect. Plausible. The first part is incorrect. When one of two IR detectors reaches 2 x10-5, the P-6 BPLB illuminates. Losing one of two IR detectors leaves one of two IR detectors available to meet the permissive. Therefore, the bistable does not change state.
On lowering power two of two TSLBs for P-6 must extinguish for the P-6 BPLB to extinguish and automatically unblock the SR Hi Flux Trips. Candidates frequently confuse the order and coincidence of these light indications.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See second part of choice B above.
Wednesday, February 26, 2014 3:55:36 PM                                                            2
 
Level:                          RO Tier # / Group #                T1 / G2 K/A#                            033AG2.2.44 Importance Rating:              4.2 / 4.4 Technical
 
==Reference:==
1X6AA02-00227 Rev 9.0 1X6AA02-00227 Rev 8.0 18002-C Rev 20.1 V-LO-PP-17201 Rev 3.0 References provided:            None Learning Objective:              V-LO-PP-17201-01 Discuss the operation of the Source &
Intermediate Range Detectors to include:
: d. All Reactor Trip signals
: e. All Permissives & Interlocks
: g. Power supplies (also including the effects on loss of instrument or control power)
LO-TA-60035        Respond to Nuclear Instrumentation System Malfunction per 18002-C LO-TA-61001        Reactor Startup using 12003-C LO-TA-61003        Reactor Shutdown to Hot Standby using 12005-C Question origin:                MODIFIED - LOIT Question 033AA2.03-1 Cognitive Level:                C/A 10 CFR Part 55 Content:          41.5 / 43.5 / 45.12 Comments:
You have completed the test!
Wednesday, February 26, 2014 3:55:36 PM                                                          3
 
Name: ________________________________                                        Export Temp Test Form: 0 Version: 0
: 1. 033AA2.03 001/1/2/LOSS OF IRNI/C/A-3.1/NEW/SRO/HL-15R AUDIT/DNS/
(SRO ONLY)
Original Question Initial conditions:
            - The unit is at 35% power.
            - All equipment is aligned for automatic control.
Current conditions:
            - SR & IR NI channels N31 & N35 level meters drop to zero.
            - SUR indications for SR & IR SUR channels N31 & N35 remain on zero.
            - TSLB Channel 1 SR & IR high flux trip bistables illuminate.
            - TSLB Channel 1 P6 bistable goes dark.
            - All BPLB lights remain unchanged.
Which one of the following describes the correct diagnosis for the 'A' Train SR/IR Control Room Signal Processor, and the actions to take per 18002-C, Section A, "Source / Intermediate Range Channel Malfunction"?
A. Loss of instrument power.
Restore inoperable Intermediate Range Channel prior to raising power. (TS 3.3.1)
B. Loss of control power.
Restore inoperable Intermediate Range Channel prior to raising power. (TS 3.3.1)
C. Loss of instrument power.
Place LEVEL TRIP switch in BYPASS on the affected Source/Intermediate Range Drawer.
D. Loss of control power.
Place LEVEL TRIP switch in BYPASS on the affected Source/Intermediate Range Drawer.
Friday, September 27, 2013 7:56:08 AM                                                            1
 
Approved By                                                                      Procedure    Version J.B. Stanley                          Vogtle Electric Generating Plant          18002-C      20.1 Effective Date                                                                    Page Number NUCLEAR INSTRUMENTATION SYSTEM 01/11/2013                                    MALFUNCTION                              5 of 11 A. SOURCE / INTERMEDIATE RANGE CHANNEL MALFUNCTION ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED
 
CAUTION Two redundant drawer assemblies should not be placed in test at the same time.
A18 A27. Perform the following on the affected            A18.
Source/Intermediate Range Drawer:
A18.a
: a. Place LEVEL TRIP switch in                      a.
BYPASS.
A18.b
: b. Check LEVEL TRIP BYPASS                        b.
light - ON.
A18.c
: c. Block SR High Flux At Shutdown                  c.
alarm.
A18.d
: d. Select audio count rate to                      d.
operating channel, if required.
A18.e
: e. Select startup rate to operating                e.
channel, if appropriate.
A19 A28. Notify I&C to initiate repairs.                  A19.
A20 A29. Bypass the affected channels by                  A20.
initiating 13509-C, BYPASS TEST INSTRUMENTATION (BTI) PANEL OPERATION, if desired.
A21 A30. Initiate 13501, NUCLEAR                          A21.
INSTRUMENTATION SYSTEM when repairs are complete.
 
S Printed September 26, 2013 at 09:42
 
Source & Intermediate Range NIS If either channel fails HIGH or exceeds 105 cps, a reactor trip signal will be generated unless it has been manually blocked above P-6 or power is above the P-10 setpoint.
The manual block allows operation of the plant during startup conditions above the level trip set point. This is done after the Intermediate Range Nuclear Instruments have confirmed operation by coming on-scale at the proper flux level and have reached a preset Permissive set point (1/2 Intermediate Range Channels >
2x10-5 % Reactor Power). This is also called P-6.
(Objective V-LO-PP-17201-01d)
(Objective V-LO-PP-17201-01e)
V-LO-PP-17201 Rev-03                                                                        45
 
Source & Intermediate Range NIS The most important point to get from this slide is that in BYPASS, the input to SSPS is energized from control power (instrument power ONLY supplies indication when the switch is in bypass.
Questions (assume power is < P-10):
What happens on a loss of instrument power with LEVEL TRIP in NORMAL?
              -Channel indication is LOST and the reactor TRIPS.
What happens on a loss of instrument power with LEVEL TRIP in BYPASS?
              -Channel indication is LOST and the reactor does NOT trip.
What happens on a loss of control power?
              -Channel indication is AVAILABLE and the reactor TRIPS.
(OBJECTIVE V-LO-PP-17201-01g)
Remember that there is only one set of breakers for both SR and IR, and the effect of the loss of power to BOTH instruments must be considered. Also, the status of P-6, P-10, and whether SR or IR trips has been blocked may be very important to properly answering the question.
V-LO-PP-17201 Rev-03                                                                          61
: 1. 034A4.01 001/LOIT/RO/M/F 3.3/3.7/034A4.01/LO-PP-32101-08///
Given the following:
          - Unit 1 is in Mode 6.
          - Preparation for fuel movement is in progess in the Fuel Handling Building.
          - The following radiation monitors are in service:
ARE-2532A/B, Fuel Handling Building Effluent Radiogas Monitors 1RE-008, Fuel Handling Building Area Monitor Which one of the following completes the following statement?
ARE-2532A/B __(1)__ be monitored and the alarm setpoints adjusted from the SRDC panel, and 1RE-008 __(2)__ be monitored and the alarm setpoints adjusted from the SRDC panel.
__(1)__                                __(2)__
A.                    can                                    can B.                    can                                  can NOT C.                  can NOT                                    can D.                  can NOT                                can NOT K/A 034              Fuel Handling Equipment A4.01            Ability to manually operate and/or monitor in the control room:
                        - Radiation levels.
K/A MATCH ANALYSIS The KA addresses the relationship between fuel handling activities and the ability to monitor radiation levels and make required setpoint adjustments. The question has all the required elements to include the activity taking place in the fuel handling building and where the control room staff can monitor and adjust the setpoints for the radiation monitors.
EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17102-1, rad monitor A-RE-2532A/B is located on the SRDC in the control Wednesday, February 26, 2014 4:02:28 PM                                                        1
 
room, where it can be monitored and its setpoints changed. 1RE-0008 can be monitored from the control room on either the IPC or Perms Console. However, since it is not located on the SRDC, its setpoint cannot be changed from the control room.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. A-RE-2532A/B is one of the radiation monitors that is operated from the main control room SRDC panel.
The second part is incorrect. 1RE-0008 is not one of the radiation monitors that is operated from the main control room SRDC panel. However, ARE-2532A/B and 1RE-008 are both listed in ARP 17100-C. The ARE-2532A/B entry sends you to ARP 17102-1 for the actions. A candidate without specific knowledge of rad monitor locations may remember seeing both listed and assume that both are on the SRDC.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. RE-0008 is not one of the radiation monitors that located on the main control room SRDC panel.
C. Incorrect. Plausible. The first part is incorrect. A-RE-2532A/B is one of the radiation monitors that is operated from the main control room SRDC panel. However, a candidate without specific knowledge of rad monitor locations may assume that since this is an "A" or common rad monitor, it would not be located on the Unit 1 SRDC panel.
The second part is incorrect. 1RE-0008 is not one of the radiation monitors that is operated from the main control room SRDC panel. However, if the candidate used the logic described in the first part of choice C above, it could be reasonable to assume 1RE-008 is located on the Unit 1 SRDC.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 4:02:28 PM                                                              2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            034A4.01 Importance Rating:              3.3 / 3.7 Technical
 
==Reference:==
ARP 17100-1, Rev 26.2, pages 6-8 ARP 17102-1, Rev 20.2, page 4 References provided:            None Learning Objective:              LO-PP-32101-08 List all safety-related radiation monitors by tag number and name. Describe those automatic actions that occur for each of the following safety-related monitors when its high alarm setpoint is exceeded:
: b. fuel handling building effluent (ARE-2532A, B and ARE 2533A, B)
Question origin:                NEW Cognitive Level:                M/F 10 CFR Part 55 Content:          41.13 Comments:
You have completed the test!
Wednesday, February 26, 2014 4:02:28 PM                                                          3
 
Approved By                                                                    Procedure    Version S. E. Prewitt                      Vogtle Electric Generating Plant            17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)          6 of 88 ANNUNCIATOR RESPONSE INDEX RADIATION MONITOR INDEX DETECTOR NO.                USE                                                  PAGE 1-RE-0001                    Control Room                                        9 1-RE-0002                    Containment - Low Range
* 1-RE-0003                    Containment - Low Range
* 1-RE-0004                    Containment Hatch                                    10 1-RE-0005                    Containment - High Range
* 1-RE-0006                    Containment - High Range
* A-RE-0007A                  Rad Chem Lab                                        **
A-RE-0007B                  Sample Room                                          12 1-RE-0008                    Fuel Handling Bldg                                  14 A-RE-0009A                  Decon Station (Large Parts)                          **
A-RE-0009B                  Decon Station (Small Parts)                          **
A-RE-0009C                  Decon Station (Instruments)                          **
1-RE-0011                    Seal Table Room                                      16 1-RE-0013                    Waste Gas Processing                                18 A-RE-0014                    Waste Gas Processing                                20 A-RE-0016                    Boron Recycle Liquid                                **
1-RE-0017A                  CCW Train A                                          22 1-RE-0017B                  CCW Train B                                          24 1-RE-0018                    Waste Liquid                                        26 1-RE-0019                    SG Sample Liquid                                    28 1-RE-0020A                  NSCW Train A                                        31 1-RE-0020B                  NSCW Train B                                        33 1-RE-0021                    SG Blowdown Liquid                                  35 1-RE-0024A                  Selected Cubical-Air Particulate                    **
1-RE-0024B                  Selected Cubical-Radiogas                            **
A-RE-0025                    Aux Steam Condensate Return Liquid                  **
* Safety Related. Go to 17102-1, "ARP For The SRDC QRM2".
** Not Functional - Detectors Removed.
*** Passive collector. No electronic components.
Printed February 6, 2014 at 14:59
 
Approved By                                                                            Procedure    Version S. E. Prewitt                      Vogtle Electric Generating Plant                  17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND        Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                  7 of 88 ANNUNCIATOR RESPONSE INDEX RADIATION MONITOR INDEX DETECTOR NO.                USE                                                          PAGE 1-RE-0039A                  Waste Gas Decay Tank - Radiogas                              38 1-RE-0039B                  Waste Gas Compressor & Catalytic Recombiner - Radiogas      40 1-RE-0724                    N16 Rad Monitor                                              42 1-RE-0810                    SJAE Exhaust Rad Monitor                                    45 1-RE-0848                    Turbine Bldg Drains - Liquid                                48 1-RE-1950                    ACCW - Liquid                                                50 A-RE-2532A                  FHB - Radiogas
* A-RE-2532B                  FHB - Radiogas
* A-RE-2533A                  FHB - Radiogas
* A-RE-2533B                  FHB - Radiogas
* 1-RE-2562A                  Containment - Air Particulate
* 1-RE-2562B                  Containment - Passive Iodine Cartridge                      ***
1-RE-2562C                  Containment Air - Radiogas
* 1-RE-2565A                  Containment Vent - Particulate                              54 1-RE-2565B                  Containment Vent - Iodine                                    56 1-RE-2565C                  Containment Vent - Radiogas                                  58 1-RE-12116                  Control Room Air In - Train A
* 1-RE-12117                  Control Room Air In - Train B
* 1-RE-12442A                  Plant Vent Air Particulate (Low Range)                      60 1-RE-12442B                  Plant Vent Iodine Particulate (Low Range)                    62 1-RE-12442C                  Plant Vent Radiogas Particulate (Low Range)                  64
* Safety Related. Go to 17102-1, "ARP For The SRDC QRM2".
** Not Functional - Detectors Removed.
*** Passive collector. No electronic components.
Printed February 6, 2014 at 14:59
 
Approved By                                                                                  Procedure    Version S. E. Prewitt                      Vogtle Electric Generating Plant                          17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND              Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                        8 of 88 ANNUNCIATOR RESPONSE INDEX RADIATION MONITOR INDEX DETECTOR NO.                USE                                                                PAGE 1-RE-12444A                  Plant Vent Air Particulates - (High Range) - Passive Detector      ***
1-RE-12444B                  Plant Vent Passive Iodine Cartridge(High Range)                    ***
1-RE-12444C                  Plant Vent Wide Range Radiogas(Low Range)                          66 1-RE-12444D                  Plant Vent Wide Range Radiogas (Mid Range)                          68 1-RE-12444E                  Plant Vent Wide Range Radiogas (High Range)                        70 1-RE-12839A                  SJAE - Passive Particulates Detector                                ***
1-RE-12839B                  SJAE - Passive Iodine Cartridge                                    ***
1-RE-12839C                  SJAE - Wide Range Radiogas (Low Range)                              72 1-RE-12839D                  SJAE - Wide Range Radiogas (Mid Range)                              75 1-RE-12839E                  SJAE - Wide Range Radiogas (High Range)                            77 1-RE-13119                  MSL Loop 4
* 1-RE-13120                  MSL Loop 1
* 1-RE-13121                  MSL Loop 2
* 1-RE-13122                  MSL Loop 3
* A-RE-16971                  RPF HIC Area A-RE-16972                  RPF Demineralizer Area A-RE-16973                  RPF Dress-out Area A-RE-16980                  RPF Vent Particulate 1-RE-17646                  Control Building Sump Effluent                                      **
1-RE-48000                  CVCS Letdown                                                        87 A-RE-50002A                  TSC Work Area                                                      **
A-RE-50002B                  TSC CRT Room                                                        **
A-RE-50003                  TSC Air Intake                                                      **
* Safety Related. Go to 17102-1, "ARP For The SRDC QRM2".
** Not Functional - Detectors Removed.
*** Passive collector. No electronic components.
Printed February 6, 2014 at 14:59
 
Approved By                                                                    Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant          17102-1      20.2 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY Page Number 6/5/13                                    RELATED DISPLAY CONSOLE QRM2                  4 of 42 SAFETY RELATED DISPLAY CONSOLE A1 A2 A3 A4 A5                      1-RE-2562A A6                      1-RE-2562C A7 A8 B1                      1-RE-13119 B2                      1-RE-13120 B3                      1-RE-0002 B4                      1-RE-0005 B5                      1-RE-12116 B6                      A-RE-2532A B7                      A-RE-2532B B8 C1                      1-RE-13121 C2                      1-RE-13122 C3                      1-RE-0003 C4                      1-RE-0006 C5                      1-RE-12117 C6                      A-43-2533A C7                      A-RE-2533B C8 D1 D2 D3 D4 D5 D6 D7 D8 SRDC CHANNEL DISPLAYS Printed February 6, 2014 at 14:39
: 1. 038EA2.05 001/LOIT/RO/2.8/2.9/038EA2.05/LO-TA-37011///
Given the following:
          -  Unit 1 experienced a SGTR.
          -  19030-C, "Steam Generator Tube Rupture," is in progress.
          -  1NAA is de-energized due to a bus fault.
          -  Crew is preparing to perform a maximum rate cool down.
Which one of the following completes the following statement?
In accordance with 19030-C, intact SG AFW flow rates are __(1)__,
and then the cooldown is initiated per Step 17 using __(2)__.
A. (1) raised to prevent AFW re-initiation (2) intact SG ARVs B. (1) raised to prevent AFW re-initiation (2) steam dumps C. (1) lowered to prevent MSL isolation upon initiation of cooldown (2) intact SG ARVs D. (1) lowered to prevent MSL isolation upon initiation of cooldown (2) steam dumps Wednesday, February 26, 2014 4:04:44 PM                                    1
 
K/A 0038            Steam Generator Tube Rupture A2.05            Ability to determine or interpret the following as they apply to a SGTR:
                        - Causes and consequences of shrink and swell in S/Gs K/A MATCH ANALYSIS The question requires the candidate to determine if AFW flow to SGs should be raised or lowered in 19030-C prior to the max rate cooldown and the reason for the flow change. The reason is related to the consequences of shrink and swell associated with the increase steam flow. The candidate is also required to determine the component used to control steam flow - ARVs or steam dumps.
EXPLANATION OF REQUIRED KNOWLEDGE Following a trip from higher power level, SG levels will shrink to less than the AFW low level actuation setpoint. AFW is then throttled to slowly restore SG levels to approximately 65% during the initial operator actions of 19000-C. With a tube rupture present, AFW to the ruptured SG is isolated as soon as possible using either the approved early operator actions of 10020-C or the isolation steps of 19030-C. Once ruptured SG isolation is complete in 19030-C, a max rate cooldown is performed to establish required subcooling for the subsequent depressurization, which will minimize RCS to SG deltaP. If SG levels are near the AFW low level actuation setpoint when the max rate cooldown commences, the resulting swell and subsequent shrink of SG levels can result in an AFW actuation. The isolated ruptured SG can have AFW flow re-initiated to it, decreasing margin to release to the public. In order to prevent this scenario, 19030-C step 14 directs raising intact SG levels prior to the max rate cooldown to establish levels well above the AFW low level actuation setpoint. Step 14 is a converted note from the original WOG version and the background and bases for this operator knowledge/action is stated in the WOG E-3 background document.
With 1NAA de-energized, only 1 Circ Water Pump is available. In order to utilize steam dumps, which is preferred to minimize release to the public, C-9 must be present and instrument air available. C-9 is determined by one of two Circ Water Pump breakers closed, voltage on the associated pump bus, and condenser vacuum available (14.92" Hg vacuum on one of two instruments on three of three condenser hoods). This interlock is met, so steam dumps will be utilized. Since only the 13.8 KV bus 1NAA was affected and not the RAT, no change to air compressor status is expected - 1NA01 and 1NA05 are energized. Therefore, steam dumps remain available. Per step 17 of 19030-C, ARVs will be used only if steam dumps are unavailable.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. Per 19030-C step 14 and the associated WOG background, intact SG levels are raised by increasing AFW flow in advance of the max rate cooldown to Wednesday, February 26, 2014 4:05:31 PM                                                          1
 
prevent re-initiation of AFW flow to the ruptured SG due to the swell and shrink observed during the max rate cooldown.
The second part is incorrect. As discussed in the Explanation of Required Knowledge above, air compressors and C-9 are present and steam dumps are available and will be utilized per step 17 of 19030-C. However, a candidate with insufficient knowledge of plant configuration on a loss of 1NAA could determine that insufficient air compressors or Circ Water pumps are available, resulting in a loss of steam dumps. In this case, ARVs would be utilized.
B. Correct.                    The first part is correct. See the first part of choice A above.
The second part is correct. As discussed in the Explanation of Required Knowledge above, air compressors and C-9 are present. Steam dumps are available and will be utilized per step 17 of 19030-C.
C. Incorrect. Plausible. The first part is incorrect. Per 19030-C step 14 and the associated WOG background, intact SG levels are raised by increasing AFW flow in advance of the max rate cooldown to prevent re-initiation of AFW flow to the ruptured SG due to the swell and shrink observed during the max rate cooldown.
However, a candidate with insufficient knowledge of plant response or of the reason for step 14 could rationalize that AFW flow would need to be lowered to prevent SG level from swelling during the max rate cooldown and causing water flow into the main steam lines, which would require a manual steam line isolation to prevent water hammer.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Level:                            RO Tier # / Group #                  T1 / G1 K/A#                              038EA2.05 Importance Rating:                2.8 / 2.9 Technical
 
==Reference:==
SOP 13503A-1, Rev 7.2, page 34 EOP 19030-C Rev 39.2, pages 10-15 E-3 WOG Background Rev 2, 4/30/2005 P&ID 1X3D-AA-C02A, Rev 16.0 References provided:              None Wednesday, February 26, 2014 4:05:31 PM                                                                2
 
Learning Objective:              LO-LP-37311-07 Using EOP 19030-C as a guide, briefly describe how each step is accomplished.
LO-LP-37311-11 Given a NOTE or CAUTION statement from the EOP, state the bases for that NOTE or CAUTION statement.
LO-PP-18101-24 Discuss the "Shrink" and Swell" in the Steam Generators to include:
: a. The conditions that lead to each
: b. The possible adverse consequences LO-PP-28103-05 List all ESF actuation signals with applicable set points, coincidences, permissives, blocks, and discuss the systems response to each ESF actuation signal.
LO-TA-37011        Respond to a Steam Generator Tube Rupture per 19030-C Question origin:                NEW Cognitive Level:                M/F 10 CFR Part 55 Content:          43.5 / 45.13 Comments:                        Question appears to match the KA.
Need to make sure that EOP Background information is RO knowledge.
The question itself should reference the document being used to support the correct answer:
                                            "In accordance with <19030-C> intact SG AFW flow rates" and "in accordance with <E-3 WOG Background Document>, the reason" There appears to be 2 non-plausible distractors: why would someone choose B (raise AFW flow to intact SGs due to excessive swell) or C (lower AFW flow to intact SGs due to excessive shrink)?
The answer choices do not need to explicitly call out shrink and swell to match the KA. Testing understanding of AFW reinitiation vs. MSL isolation tests the same concept without leading. One possible idea is to ask:
                                        "In accordance with 19030-C step 14, intact SG AFW flow rates are ___(1)____ and then cooldown is initiated per step 17 using ___(2)___."
A. Raised to prevent AFW reinitiation/intact SG ARVs B. Raised to prevent AFW reinitiation/steam dumps C. Lowered to prevent MSL isolation upon initiation of the cooldown/intact SG ARVs D. Lowered to prevent MSL isolation upon initiation of the cooldown/steam dumps The stem will likely need some more information to bolster the correct answer (i.e., preferred cooldown method given Wednesday, February 26, 2014 4:05:31 PM                                                            3
 
the conditions of the stem), but the KA is met with the first part of the question and the distractors are all plausible.
                                        -JAT 12/19/2013 (Unsat/Editorial)
New question incorporated above suggestion.
                                        -JAT 2/4/2014 You have completed the test!
Wednesday, February 26, 2014 4:05:31 PM                                                              4
 
Approved By                                                                          Procedure    Version J.B. Stanley                            Vogtle Electric Generating Plant            19030-C      39.2 Effective Date                                                                        Page Number E-3 STEAM GENERATOR TUBE RUPTURE 05/01/2013                                                                                  12 of 57 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 13
: 13. Align Steam Dumps for RCS                      13.
cooldown:
13.a
: a. IF Steam Dumps are in T AVG                      a.
mode, THEN 13.a.
: 1)    Match demand on SG Header Pressure Controller PIC-507 and SD demand meter UI-500.
13.a.2
: 2)    Transfer Steam Dumps to STM PRESS mode using HS-500C 13.b
: b. RCS temperature - GREATER                        b. Momentarily place THAN 550&deg;F.                                  HS-0500A and HS-0500B in the BYPASS INTERLOCK position.
13.c
: c. As RCS cooldown is initiated,                    b.
hold HS-0500A and HS-0500B in the BYPASS INTERLOCK position until RCS temperature is less than 550&deg;F.
14
: 14.        Raise intact SG levels prior to                14.
maximum rate cooldown.
15
: 15.        Check at least one RCP - RUNNING.              15. Perform the following:
15.a
: a. Suspend monitoring ruptured loop T-Cold indication on RCS Integrity Status Tree.
15.b
: b. Resume monitoring ruptured loop T-Cold indication on RCS Integrity Status Tree if this procedure is exited.
 
S Printed September 27, 2013 at 14:39
 
Approved By                                                                            Procedure    Version J.B. Stanley                            Vogtle Electric Generating Plant              19030-C      39.2 Effective Date                                                                          Page Number E-3 STEAM GENERATOR TUBE RUPTURE 05/01/2013                                                                                  14 of 57 ACTION/EXPECTED RESPONSE                            RESPONSE NOT OBTAINED 17
: 17. Initiate RCS cooldown:                        17.
17.a
: a.      Dump steam to Condenser from                a. Dump steam at maximum intact SG(s) at maximum rate                      rate from intact SG ARV(s).
using Steam Dumps by slowly raising demand on PIC-507.
IF no intact SG is available, THEN perform the following:
Use faulted SG.
                                                                                  -OR-Go to 19131-C, ECA-3.1 SGTR WITH LOSS OF REACTOR COOLANT:
SUBCOOLED RECOVERY DESIRED.
18
        *18. Check if RCS cooldown should be                18.
stopped:
18.a
: a.      Core Exit TCs - LESS THAN                  a. WHEN Core Exits are less REQUIRED TEMPERATURE.                            than required, THEN perform Steps 18.b and 18.c.
Go to Step 19.
18.b
: b.      Stop RCS cooldown.                            b.
18.c
: c.      Maintain Core Exit TCs - LESS                  c.
THAN REQUIRED TEMPERATURE.
 
S Printed September 27, 2013 at 14:39
 
Approved By                                                                                                  Procedure  Version J.B. Stanley                          Vogtle Electric Generating Plant                                      13503A-1 7.2 Effective Date                TRAIN A REACTOR CONTROL SOLID-STATE PROTECTION                                Page Number 6/21/13                                                        SYSTEM                                            34 of 38 ATTACHMENT C                                        Sheet 2 of 6 PERMISSIVES, CONTROL INTERLOCKS, REACTOR TRIPS AND ESF ACTUATIONS CONTROL INTERLOCKS Control Interlock                Setpoint/Coincidence                  Function C-1 IR Rod Stop          1/2 IR NIS  20% Rx Power            Auto/manual Rod Stop*
May be manually blocked above P-10 C-2 PR Hi                1/4 PR NIS  105 % Rx power          Auto/manual rod stop*
Rod Stop C-3 OTT Rod              2/4 T loops w/in 3% of Rx trip      Auto/manual rod stop*
Stop / Runback            setpoint Initiates turbine runback C-4 OPT Rod              2/4 T loops w/in 3% of Rx trip      Auto/manual rod stop*
Stop/Runback              setpoint Initiates turbine runback C-5 15% Turbine          PT-505  15%                          Blocks auto rod withdrawal*
Power C-7                        10% in 2 minutes load decrease as  Arms Steam Dumps Load Rejection            sensed on PT-506 C-9 Condenser            1/2 CW pumps running with 2/2 13.8    Indicates condenser available for Steam Dumps Available                kV bus UV relays not energized AND 1/2 condenser vacuum sensors 24.92" Hg Vacuum on 3/3 condensers (LPT hoods A, B, & C)
C-11 CBD Full            CBD at 220 steps Blocks auto rod withdrawal
* Withdrawal C-16 Stop                Auct Lo NR Tavg  553 F              Stops any increase in turbine load Turbine Loading OR Auct Lo Tavg 20 F below Tref C-20 AMSAC                2/2 Turbine Impulse press  40%      Automatically enables AMSAC Enabled
* AUTO Rod Withdrawal defeated Printed January 22, 2014 at 14:42
 
STEP DESCRIPTION TABLE FOR E3          Step  7 STEP:      Check Intact SG Levels PURPOSE: o To control feed flow to the intact steam generators to prevent excessive RCS cooldown and steam generator overfill.
o To maintain an adequate secondary side heat sink.
o To identify a previously undetected steam generator tube failure which could potentially result in steam generator overfill BASIS:
In most cases, feed flow will exceed steam flow from the intact steam generators resulting in an accumulation of water in the steam generators. This excess feed flow will also result in a cooldown of the RCS at a rate dependent upon the feed flow rate and heat generation rate in the primary system. Consequently, feed flow must be adjusted to control steam generator level and reactor coolant temperature. This step also provides for monitoring level in the intact steam generators to detect multiple or subsequent tube failures. In that case, the operator is directed to stop any RCS cooldown in progress and return to Step 1 to isolate the affected steam generator and repeat the recovery actions.
If reactor trip occurs from a high power level, the water level may shrink below the narrow range so that temporarily no reliable indication of steam generator water level is available. During this time, feed flow should be maintained greater than (S.02) to ensure an adequate secondary side heat sink. This minimum feed flow requirement satisfies the feed flow requirement of the Heat Sink Status Tree until level in at least one steam generator is restored in the narrow range.
Narrow range level is reestablished in all intact steam generators to maintain symmetric cooling of the RCS. Once intact SG level has been reestablished in the narrow range, the operator is directed to establish a control band between the AFW actuation setpoint and 50%.
This control range ensures an adequate inventory will be maintained close to the typical SG level control band and prevent the actuation of the AFW signal. Actuation of the AFW signal could result in potential releases from the ruptured SG through the opened steam E3 Background                    92            HPRev. 2, 4/30/2005 HE3BG.doc
 
supply valves to the turbinedriven AFW pump if the ruptured SG contained the steam supply tap.
19030-C step 14 does not have a direct step or note listed in the WOG EOP background. It is instead part of the operator knowledge for maintaining SG levels post trip to prevent reinitiation of feed flow once the ruptured SG has been isolated.
Due to operator tendency to forget to prepare for the shrink and swell seen during the max rate cooldown, Vogtle placed a NOTE just prior to the cooldown step to remind operators to raise INTACT SG levels. This prevented SG levels from initially swelling above the low level actuation setpoint and subsequently falling below the setpoint resulting in an re-actuation of AFW and reinitiation of feed flow to the ruptured SG(s). Instead, levels are raised in advance and AFW flow maintained at a relatively high rate during the max rate cooldown.
The NOTE was later converted to step 14 during a HU procedure re-write effort.
E3 Background                            93                HPRev. 2, 4/30/2005 HE3BG.doc
: 1. 039K4.02 001/LOIT AND LOCT/RO/C/A 3.1/3.2/039K4.02/LO-TA-37002/LP-37011///
Initial condition:
          - Unit 1 is at 100% reactor power.
Current conditions:
          - Unit 1 reactor trips.
          - RTB 'B' fails to open.
Which one of the following completes the following statement?
RCS temperature will be controlled at approximately __(1)__ by the __(2)__.
__(1)__                                    __(2)__
A.          557&deg; F                                steam dumps B.          557&deg; F                                    ARVs C.          559&deg; F                                steam dumps D.          559&deg; F                                    ARVs K/A 039              Main and reheat steam K4.02            Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the following:
                        - Utilization of T-avg program control when steam dumping through atmospheric relief/dump valves, including T-avg limits.
K/A MATCH ANALYSIS The question tests the candidate's knowlege of the design and interlocks associated with Steam Dumps, which are part of the Main Steam System, including the response in Tavg mode with a failure of P-4 and the corresponding steady state RCS temperature expected.
EXPLANATION OF REQUIRED KNOWLEDGE Per 12004-C step 4.1.41.f, ARVs are adjusted to a setpoint of 7.47 during startup. This setpoint corresponds to a RCS temperature of 560F based on saturation. However, each potentiometer has "slop" in the potentiometer and does not control exactly. As such, RCS temperature will normally stabilize between 559F-561F following a reactor trip where the Steam dumps do not arm.
Wednesday, February 26, 2014 4:07:14 PM                                                    1
 
The steam dumps require C-9 and an arming signal from either PT-506 or 'A' Train P-4 to open. Since RTB 'A' opened, the steam dumps received an arming signal. Steam dumps either utilize the Plant Trip or Load Reject controller in Tavg mode. The load reject controller has a 2F dead band and is designed to replace the loss of load while at power by using Tref for a demand signal. When the reactor trips, Tref goes to 557F.
The 2F deadband will cause the RCS temperature to stabilize at approximately 559F.
The Plant Trip controller has a setpoint of 557F. The Load Reject controller is used when Train 'B' P-4 is not present. Steam dumps transfer to the Plant Trip controller when the RTB 'B' opens.
PT-505 feeds Tref. PT-506 feeds the Steam Dump arming circuit. Both instruments are 1st Stage Turbine Pressure and are side by side on the control board. Both instruments feed into AMSAC.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. RTB 'B' did not open, leaving the steam dumps to control on the Load Reject controller at 559F.
However, a candidate that confuses which RTB enables the Plant Trip Controller may believe that steam dumps are controlling on the Plant Trip Controller at 557F.
The second part is correct. Steam dumps are armed by both RTB 'A' and PT-506 lowering on the turbine trip. A candidate who reverses which RTB arms the steam dumps could recognize steam dumps are armed by PT-506 alone and conclude steam dumps are still armed.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Steam dumps were armed by both RTB 'A' and PT-506. ARV's are set to 7.47 and control at 560F and therefore should not open since Steam Dumps should control at a slightly lower temperature. However, candidates often confuse which RTB feeds which function. If the candidate reversed the RTBs and the functions they enable, and the PT-506 arming circuit is overlooked, the candidate could believe that the Steam Dumps did not receive an arming signal and ARVs would open to control temperature.
C. Correct.                  The first part is correct. RTB 'B' did not open causing the steam dumps to control on the Load Reject controller at 559F.
The second part is correct. See the first part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Wednesday, February 26, 2014 4:07:14 PM                                                              2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            039K4.02 Importance Rating:              3.1 / 3.2 Technical
 
==Reference:==
AOP 18001-C, Rev 35.0, page 43 UOP 12004-C, Rev 107.1, page 35 LOGIC 1X6AA02-00234, Rev 9.0 References provided:            None Learning Objective:              LO-LP-37011-02 State how the following control systems are employed to automatically stabilize the plant after a reactor trip:
: a. steam dumps LO-PP-21101-10 Discuss the following concerning the "Atmospheric Relief valves" (ARV):
: a. Why we have them
: b. Basic description of how they operate (automatic and manual)
LO-TA-37002    Respond to a Reactor Trip Without Safety Injection per 19000-C and 19001-C Question origin:                MODIFIED - HL17 Question # 041K3.02 Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 Comments:
You have completed the test!
Wednesday, February 26, 2014 4:07:15 PM                                                        3
: 1. 041K3.02 001/2/2/STEAM DUMPS-RCS/H 3.8/3.9/NEW/H-17 NRC/RO/SRO/EMT/GCW Unit 1 initial conditions:
Original Question
          - Power is 100%.
          - Steam Dumps are in the Tavg mode.
Current conditions:
          - Main Turbine automatically trips.
          - Reactor Trip Breaker "A" opens.
          - Reactor Trip Breaker "B" is CLOSED and cannot be opened.
Based on the current conditions, which one of the following correctly completes the following statement, if no other operator actions are performed?
RCS Tavg will be controlled at __(1)__ due to Steam Dumps controlling on the __(2)__
controller.
__(1)__                __(2)__
A. 559&deg;F                plant trip B. 559&deg;F                load reject C. 557&deg;F                plant trip D. 557&deg;F                load reject Friday, February 07, 2014 9:49:37 AM                                                        1
 
Approved By                                                                    Procedure    Version J.B. Stanley                        Vogtle Electric Generating Plant          18001-C        35 Effective Date                                                                  Page Number SYSTEMS INSTRUMENTATION MALFUNCTION 03/25/2013                                                                            43 of 45 H. FAILURE OF TURBINE IMPULSE PRESSURE INSTRUMENTATION ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED
 
IMMEDIATE OPERATOR ACTIONS H1 H1. Check - NO ROD MOTION.                    H1. Place ROD BANK SELECTOR SWITCH in MAN position.
SUBSEQUENT OPERATOR ACTIONS H2 H2. Restore TAVG to program band.              H2.
H3 H3. Perform the following:                    H3.
H3.a
: a. Verify PIC-507 STEAM DUMP                    a.
CONTROL set at 1092 psig (approximately 7.28).
Corresponds to a                        H3.b
: b. Verify PIC-507 in AUTO.                saturation temperature of                          H3.c
: c. Place HS-500C STEAM DUMP                557F.b.
CONTROL MODE SELECT in STM PRESS.
H4 H4. Check P-7 and P-13 status lights          H4.
indicate correctly for plant condition within one hour. (TS 3.3.1)
H5 H5. Initiate the applicable actions of        H5.
Technical Specification 3.3.1.
H6 H6. Notify I&C to initiate repairs.            H6.
H7 H7. Initiate the Continuous Actions Page.      H7.
 
S Printed February 6, 2014 at 16:42          2
 
Approved By                                                                              Procedure  Version C.E.H. Williams                    Vogtle Electric Generating Plant                      12004-C 107.1 Effective Date                                                                            Page Number 05/30/2013                              POWER OPERATION (Mode 1)                              35 of 114 INITIALS NOTE If the Unit will be held at approximately 30% reactor power, the steam dumps should remain in the steam pressure mode until reactor power is increased.
: e.      Transfer Steam Dumps to Tavg Mode per 13601, Main Steam System.                                                ________
: f.      Align each of the Steam Generator Atmospheric Relief Valves for standby by adjusting potentiometers to 7.47 and placing controllers in AUTO:
SG1 ARV          PIC-3000A        Pot @ 7.47 and in AUTO      ________
SG2 ARV          PIC-3010A        Pot @ 7.47 and in AUTO      ________
SG3 ARV          PIC-3020A        Pot @ 7.47 and in AUTO      ________
SG4 ARV          PIC-3030A        Pot @ 7.47 and in AUTO      ________
: g.      Transfer sealing steam supply from Auxiliary Steam to Main Steam per 13825, "Turbine Steam Seal System."            ________
: h.      Transfer the SJAE Steam Supply from the Auxiliary Steam System to the Main Steam System per 13620, "Condenser Air Ejector System."                                          ________
: i.      Perform calorimetric calibration of Nuclear Instrumentation per 14030, Nuclear Instrument Calorimetric Calibration."
14030:___________/_______                        ________*
Date        Time At setpoint of 7.47 corresponds to a pressure of 1120 psig, which corresponds to approximately 559-561F depending on the amount of "slop" in the potentiometer.
Printed February 6, 2014 at 15:49
 
RTB A arming              controls at Tref +
input                    2F deadband controls at 557F PT-506 Arming Input Train B P-4 not present Stm Dmp Arming                                  Load Reject signals and C-9                                  controller still active since 'B' P-4 not present.
: 1. 040AG2.2.36 001/LOIT AND LOCT/RO/C/A 3.9/4.0/040AG2.2.36/LO-TA-60019///
Initial conditions:
          - Unit 1 is at 100% reactor power.
          - Train 'A' sequencer simulated software sequence test is in progress per 13540A-1, "Safety Features Sequencer System - Train A."
          - Containment Coolers #1, 2, 5, and 6 are running in high speed.
Current conditions:
          -  A steam line rupture occurs on SGs #1 and 2.
          -  SG pressures lower from 900 psig to 600 psig in 3 seconds.
          -  Containment pressure is 14 psig.
          -  Reactor trip, Safety Injection, and Steam Line Isolation occur.
          - NO operator action has been taken.
Which one of the following completes the following statement?
Based on the given conditions, the Steam Line Isolation was initiated by __(1)__,
and Train 'A' Containment Coolers will be running in __(2)__ speed.
__(1)__                                  __(2)__
A.          containment High-2                                high B.          containment High-2                                low C.            SG low pressure                                  high D.            SG low pressure                                  low Wednesday, February 26, 2014 4:20:39 PM                                                  1
 
K/A 040              Steam line rupture - excessive heat transfer G2.2.36          Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions of operations.
K/A MATCH ANALYSIS The question presents a plausible scenario in which conditions are provided in the stem requiring the candidate to determine the system response, taking into account the maintenance activity in progress. The event includes the steam line rupture and has the candidate determine which actuation signal resulted in the equipment lineup stated.
EXPLANATION OF REQUIRED KNOWLEDGE With 13540A-1 in progress, Sequencer 'A' is in TEST. Some or all actuations are blocked depending on the progression of the test. When the SI signal is received, the Sequencer will automatically exit the TEST mode and run the SI sequence.
Containment Coolers #1, 2, 5, and 6 are running in high speed. These are all Train 'A' components. When the SI sequence runs, the High Speed breaker will trip open and the Low Speed breaker will be closed by the sequencer.
The initiating event for the SI is a steam line rupture on SGs #1 and 2. During this event SG pressures lower from 900 psig (approximately normal pressure for 100%
RTP) to 600 psig in 3 seconds. The SI/SLI setpoint is 585 psig on 2/3 SG pressure channels on 1/4 SGs. This circuit is also rate compensated and resulted in a SLI above the setpoint.
Simutaneously, containment pressure rises to 14 psig due to the mass and heat addition. The SLI setpoint is 14.5 psig on 2/3 containment pressure channels. This circuit IS NOT rate compensated and therefore will not result in a SLI.
Note: This scenario was run on the simulator to ensure it was plausbile and the conditions were able to be duplicated exactly as stated.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Containment pressure rises to 14 psig due to the mass and heat addition. The SLI setpoint is 14.5 psig on 2/3 containment pressure channels. This circuit IS NOT rate compensated and therfore will not result in a SLI.
However, the candidate may believe this circuit is rate compensated or mistake the HI-1 setpoint of 3.8 psig with the HI-2setpoint and believe a SLI would have occured.
The second part is incorrect. With 13540A-1 in progress, Wednesday, February 26, 2014 4:24:54 PM                                                            1
 
Sequencer 'A' is in TEST. Some or all actuations are blocked depending on the progression of the test. When the SI signal is recieved, the Sequencer will automatically exit the TEST mode and run the SI sequence. However, SSPS will not automatically exit the TEST mode. A candidate who confuses SSPS and the Sequencer may believe that the Train 'A' Containment Coolers would continue running in High Speed.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. With 13540A-1 in progress, Sequencer 'A' is in TEST. Some or all actuations are blocked depending on the progression of the test. When the SI signal is recieved, the Sequencer will automatically exit the TEST mode and run the SI sequence.
C. Incorrect. Plausible. The first part is correct. The SI/SLI setpoint is 585 psig on 2/3 SG pressure channels on 1/4 SGs. This circuit is also rate compensated and will result in a SLI above the setpoint.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
You have completed the test!
Wednesday, February 26, 2014 4:24:54 PM                                                              2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            040AG2.2.36 Importance Rating:              3.1 / 4.2 Technical
 
==Reference:==
SOP 13540A-1, Rev 2.2, page 13 LOGIC 1X6AA02-00231, Rev 8.0 LOGIC 1X6AA02-00232, Rev 17.0 References provided:            None Learning Objective:              LO-PP-21101-06 Describe the operation of the Main Steamline Pressure instruments to include:
: a. Where they can be read at (Main Control Room, Shutdown Panels)
: b. The number of channels per Main Steamline used by the Reactor Protection System
: c. How they function as part of the Reactor Protection System above and below P-11 LO-PP-28201-05 Discuss the sequencer testing operations to include:
: a. Simulated Software Testing
: b. Selected Output Relay Testing
: c. Receipt of a SI or UV signal if testing in progress LO-LP-37121-05 Describe the plant response to the following conditions:
: a. steam line break vs feed line break
: b. break at end of life vs break at beginning of life core
: c. steam break at full power initially vs zero power
: d. feed break inside last check valve vs feed break outside last check valve
: e. steam break between SG and first MSIV vs steam break between SG and outside last MSIV LO-TA-37010        Isolate a Faulted Steam Generator per 19020-C as BOP LO-TA-60019        Respond to a Loss of Secondary Coolant per 18008-C Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.4 / 41.5 / 41.7 Wednesday, February 26, 2014 4:25:54 PM                                                              1
 
Comments:
You have completed the test!
Wednesday, February 26, 2014 4:25:54 PM                            2
 
Approved By                                                                                  Procedure    Version C.S. Waldrup                      Vogtle Electric Generating Plant                          13540A-1        2.2 Effective Date                                                                                Page Number 04/18/2013                    SAFETY FEATURES SEQUENCER SYSTEM - TRAIN A                              13 of 28 INITIALS 4.3                SEQUENCER TESTS NOTES If a manual test is in progress, a valid SI or U/V signal will override the test and initiate the accident sequence; therefore, performance of these tests will not render the sequencer inoperable.
An SI signal will override a U/V signal.
The following tests may be used to functionally test the Train A Sequencer for return to service following maintenance. Maintenance and/or Engineering should be consulted in determining required testing for returning the sequencer to an operable condition.
All steps in the following tests are performed at the Safety Features Sequencer Train A Manual Test Panel, unless otherwise specified.
4.3.1              Simulated Software Sequence Test (SI)
NOTES The SI SEQUENCE test verifies the proper operation of the SEQUENCER SI timing logic for sequencing SI loads.
Any alarms received during sequencer testing may be reset by depressing the ALARM ACKNOWLEDGE button on the NORMAL OPERATION OVERVIEW page.
4.3.1.1            Verify the Sequencer Train A Timing Logic as follows:
: a.      On the DIRECTORY page, verify the TEST PERMISSIVE yellow indicator is lit.                                            ________
: b.      Place the FUNCTION ENABLE keylock switch to the ACTIVE position (Key 1OP3-41).                                      ________
ALB36-D04 SEQ A IN MANUAL TEST
: c.      Momentarily depress the ENTER TEST MODE button.                    ________
Printed February 7, 2014 at 10:03
: 1. 045K5.17 001/LOIT/RO/C/A 2.5/2.7/045K5.17/LO-TA-61010/LP-61101///
Given the following:
          - Unit 1 is at 25% reactor power.
          - Power ascension is in progress.
Which one of the following completes the following statement?
As turbine load is increased, main steam header pressure will __(1)__,
and the method used to maintain Tavg on program during the power increase that results in the MOST negative MTC is __(2)__.
A. (1) decrease (2) control rod withdrawal with boron concentration held constant B. (1) decrease (2) boron concentration reduction with control rod position held constant C. (1) increase (2) control rod withdrawal with boron concentration held constant D. (1) increase (2) boron concentration reduction with control rod position held constant K/A 045              Main Turbine Generator K5.17            Knowledge of the operational implications of the following concepts as they apply to the MT/G system:
                        - Relationship between MTC and boron concentration in RCS as load increases.
K/A MATCH ANALYSIS The KA is addresses the relationship between changing main turbine load and the expected parameter responses and how operator actions can impact reactor core conditions, specifically how rod and boron changes affect MTC.
EXPLANATION OF REQUIRED KNOWLEDGE Wednesday, February 26, 2014 4:28:07 PM                                                  1
 
As reactor power increases, RCS Tavg increases due to a ramped program Tavg. The increased RCS Tavg results in a higher SG saturation pressure. However, as the turbine control valves are opened to increase steam flow, SG pressure lowers. The pressure reduction due to increased steam flow is of a larger magnitude than the increase associated with saturation pressure. SG pressure at no load Tavg of 557F is approximiately 1100 psig. SG pressure at full load Tavg of 586.4F is approximately 900 psig.
If the power increase is performed utilizing rods and leaving boron concentration the same, then the boron concentration would be higher at a given power level than if the increase was performed using boron and rod position was unchanged. Per Fundamental Reactor Theory, as boron concentration increases, MTC becomes more positive. Holding rods constant and diluting to power would create the most NEGATIVE MTC scenario.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. As reactor power is increased, steam flow is increased. SG pressure is reduced more by the increase in steam flow than it is increased by raising Tavg. The net affect is a reduction in SG pressure as reactor power increases.
The second part is incorrect. If control rods are withdrawn and boron is held constant, MTC would become more POSITIVE.
However, when comparing the change in both rods and boron, it is easy to confuse which one is changing, and think that adjusting boron is actually increasing boron instead of decreasing born concentration.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. If control rods are held constant and boron is changed, MTC would become more NEGATIVE, due to the dilution of the RCS.
C. Incorrect. Plausible. The first part is incorrect. As reactor power is increased, steam flow is increased. SG pressure is reduced more by the increase in steam flow than it is increased by raising Tavg. The net affect is a reduction in SG pressure as reactor power increases.
However, if a candidate only considers the change in saturation pressure of the SG and neglects the velocity pressure component, then this answer could be considered correct.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Wednesday, February 26, 2014 4:28:07 PM                                                              2
 
Level:                          RO Tier # / Group #                T2 / G2 K/A#                            045K5.17 Importance Rating:              2.5 / 2.7 Technical
 
==Reference:==
None References provided:            None Learning Objective:              LO-LP-39205-06 State the reason for limitations on MTC.
LO-LP-61101-02 Describe the parameters that change on a load decrease, to include:
: a. Turbine power
: b. Control valve/CIV positions
: c. Steam pressure
: d. Steam flow
: e. Steam generator level
: f. Feed flow
: g. Tave
: h. Reactor power
: i. Rod position (auto mode)
LO-TA-61010        Perform Power Ascent During Low Power Operations using 12004-C Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.1 / 41.5 / 41.14 Comments:
You have completed the test!
Wednesday, February 26, 2014 4:28:07 PM                                                        3
: 1. 054AA1.02 001/LOIT AND LOCT/RO/C/A 4.4/4.4/054AA1.02/LO-PP-20101-04///
At time 1100:
            - Unit 1 is at 11% reactor power.
            - The Control Room is evacuated due to a fire per 18038-1, "Operation From Remote Shutdown Panels."
            - MFP 'A' is tripped prior to exiting the Control Room.
At time 1105:
            - Personnel established local control of all components per 18038-1.
            - SG NR levels have lowered to 55%.
            - Shift Supervisor directs start of all AFW pumps.
Which one of the following completes the following statement?
The MDAFW Pumps __(1)__ require manual start, and the TDAFW Pump __(2)__ require manual start.
__(1)__                                __(2)__
A.                      will                                    will B.                      will                                will NOT C.                  will NOT                                  will D.                  will NOT                                will NOT K/A 054              Loss of Main Feedwater A1.02            Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW):
                        - Manual startup of electric and steam-driven AFW pumps.
K/A MATCH ANALYSIS The question tests the candidate's ability to determine if manual starts of the MDAFW and TDAFW pumps is required based on a sequence of events and current plant condtions associated with a Loss of Main Feedwater.
EXPLANATION OF REQUIRED KNOWLEDGE Thursday, February 27, 2014 9:24:11 AM                                                        1
 
MDAFW pump automatic start signals are:
              - SI
              - Trip of both MFPs
              - AMSAC
              - Lo SG level on 2 of 4 transmitters on 1 of 4 SGs
              - LOSP on its associated bus TDAFW pump automatic start signals are:
              - AMSAC
              - Lo SG level on 2 of 4 transmitters on 2 of 4 SGs
              - LOSP on either bus Based on plant conditions at 11% RTP, only one MFP is in service. Trip of both MFP actuation circuit is required to be operable prior to raising power above 5% RTP. This is the only MDAFW start signal present. Both MDAFW pumps should be running and would not have to be manually started.
Based on plant conditions, AMSAC is bypassed <40% 1st stage trubine pressure. SG levels are all above 38% NR. No LOSP conditions exist. TDAFW pump should NOT be running and would have to be manually started by the operator.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. The MDAFW pumps received an auto start signal from the trip of both MFPs and will NOT require a manual start.
The second part is correct. The TDAFW pump has NOT received an auto start signal and will require a manual start.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. The TDAFW pump has NOT received an auto start signal and will require a manual start.
C.Correct.                    The first part is correct. The MDAFW pumps received an auto start signal from the trip of both MFPs.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Thursday, February 27, 2014 9:24:11 AM                                                              2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            054AA1.02 Importance Rating:              4.4 / 4.4 Technical
 
==Reference:==
LOGIC 1X6AA02-00231, Rev 8.0 LOGIC 1X6AA02-00231, Rev 11.0 Lesson Plan V-LO-PP-20101, Rev 3.3, slides 45 & 49 References provided:            None Learning Objective:              LO-PP-20101-04 List the AFW system automatic start signals and component actuations.
LO-PP-20101-18 Describe the differences between control room and remote shutdown panel operation of the AFW system.
LO-TA-20016        Start the TDAFW Pump from Shutdown Panel "C" using 18038-1/2 Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 / 45.5 / 45.6 Comments:
You have completed the test!
Thursday, February 27, 2014 9:24:11 AM                                                        3
 
AMSAC
 
AFW Actuation Signal AFW is aligned to be placed in service automatically Motor Driven AFW actuation signal
* SI signal (associated train)
* Trip of BOTH (2/2) MFPTs
* AMSAC signal
* Loss of offsite power (blackout) associated train
* Lo-Lo Level in SG - 2/4 detectors on 1 SG V-LO-PP-20101 Rev-03.3                                                    45
 
TDAFW Actuation Signal Actuating signals AMSAC signal (From AMSAC Panel)
Loss of offsite power (either train) (From Sequencer Panel)
Lo-Lo SG water level - 38% - 2/4 on 2/4 steam generators (From SSPS) 49
: 1. 055EK1.02 001/LOIT AND LOCT/RO/C/A 4.1/4.4/055EK1.02/LO-TA-37018//HL-18 NRC/056AK3.02 Given the following:
            - 19100-C, "Loss of All AC Power," is in progress on Unit 1.
            - The crew is performing Step 29, "Depressurize intact SGs to 300 psig."
            - UO could NOT stop depressurizing SGs at 300 psig.
            - All SGs reach 175 psig before the depressurization is stopped.
Which one of the following completes the following statement?
A potential operational implication that could result from the excessive SG depressurization is _____________.
A. an undesired automatic SI signal can occur, which complicates other recovery actions that are in progress B. nitrogen injection from the accumulators can occur, which disrupts natural circulation flow in the RCS C. voiding in the reactor vessel can occur, which requires re-pressurization of the RCS before continuing in 19100-C D. an excessive RCS cool down can occur, which requires transition to 19241-C, "Response to Imminent Pressurized Thermal Shock Condition" 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 K/A MATCH ANALYSIS The KA addresses the relationship between loss of all AC power and the operational implications of exceeding procedure limits during the natural circulation RCS cooldown.
EXPLANATION OF REQUIRED KNOWLEDGE Per EOP 19100-C step 29, SGs are depressurized to 300 psig at the maximum rate to mitigate a Loss of All AC. Per the WOG Background ECA-0.0, this action is performed for two reasons. One reason is to minimize the DP across the RCP seals. By lowering RCS pressure, less designed leakage will flow through the seals, delaying seal failure and minimizing inventory loss. Secondly, by rapidly depressurizing the SGs to 300 Thursday, February 27, 2014 9:37:29 AM                                                        1
 
psig, the SI Accumulators will inject, maximizing RCS inventory. Combining the strategies maximizes mitigation time to allow for recovery of the electrical system.
The max rate depressurization is performed using manual ARV hand pumps. For this reason, stopping the depressurization quickly at 300 psig can be extremely challenging.
EOP 19100-C sets a lower limit of 200 psig. This limits ensures the Accumulators will fully inject without injecting nitrogen into the RCS. Nitrogen injection will result in voiding of the SG tubes, disrupting natural circulation and resulting in a loss of Core Cooling.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. 19100-C contains a Continuous Action step to reset a Safety Injection signal that is processed while in this procedure. This step is repeated immediately following the max rate cooldown.
The candidate may assume that this is one of the adverse consequences of the maximum rate cooldown that must be avoided. 300 psig is the shutoff head of the RHR pumps and is a pressure commonly associated with ensuring injection flow.
B. Correct.                    Per EOP 19100-C WOG Background ECA-0.0, if the minimum Steam Generator pressure of 200 psig was exceeded, it could result in loss of natural circulation cooling due to the blocked flowpath as nitrogen voids the SG U-tubes.
C. Incorrect. Plausible. 19100-C has five different criteria for immediately stopping the RCS cooldown, and some of the EOPs dealing with natural circulation has the candidate re-pressurize the RCS to prevent possible interruption of natural circulation flow. (Ref 19002-C step 23 as an example.)
D. Incorrect. Plausible. 19100-C step 30 has criteria to stop RCS cooldown to prevent a PTS challenge; however, this is based on cold leg temperatures instead of Steam Generator pressure. Per WOG Background ECA-0.0 page 141, SG pressure at 300 psig should not challenge this limit. (Note: Tsat for 300 psig is 422F)
Thursday, February 27, 2014 9:37:29 AM                                                                  2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            055EK1.02 Importance Rating:              4.1 / 4.4 Technical
 
==Reference:==
EOP 19100-C, Rev 38.1, pages 12, 20 thru 22 EOP 19002-C, Rev 24.0, page 13 WOG EOP Background ECA-0.0, Rev 2, 4/30/2005, pages 24 & 141 References provided:            None Learning Objective:              LO-LP-37031-08 Using EOP 19100-C as a guide, briefly describe how each step is accomplished.
LO-LP-34700-22 Explain the conditions which must exist to establish natural circulation.
LO-LP-34700-23 Describe the means by which the operator can enhance natural circulation.
LO-LP-34700-25 Describe how gas binding affects natural circulation LO-LP-36101-09 State how the formation of noncondensible gases and/or steam can result in degradation of natural circulation flow .
LO-TA-37018        Respond to a Loss of All AC Power per 19100-C Question origin:                BANK - HL18 NRC Question # 056AK3.02 Cognitive Level:                C/A 10 CFR Part 55 Content:          41.2 / 41.8 / 41.10 Comments:
You have completed the test!
Thursday, February 27, 2014 9:37:29 AM                                                                  3
: 1. 056AK3.02 002/1/1/LOSP - EOP ACTIONS/MEM - 4.4/4.7/NEW/HL-18 NRC/RO/SRO/KAJ Given the following:                                Original Question
        - A loss of all AC occurred and 19100-C, "Loss of All AC Power," is entered.
        - A depressurization of all SGs at the maximum rate is in progress.
Which ONE of the following completes the following statement?
The reason for stopping the SG depressurization at 300 psig is to prevent ________.
A. a steam bubble from forming in the Reactor Vessel Head B. N2 injection into the RCS from the ECCS Accumulators C. challenging the integrity critical safety function D. a rapid loss of pressurizer level Monday, February 10, 2014 8:58:41 AM                                                        1
 
ECA - 0.0 Loss of all AC Power                                            19100-C VOGTLE                      Version 38.1 Unit C                  Page 12 of 50 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                    RESPONSE NOT OBTAINED
: 10. Initiate 13427A/B 4160V AC Bus          10.
AA02/BA03 1E Electrical Distribution System to energize at least one AC Emergency Bus using any available power supply:
IF offsite power available to either RAT, THEN use normal or emergency incoming feeder breaker.
              -OR-IF power available to SAT, THEN initiate 13418A(B) Standby Auxiliary Transformer Unit 1(2) Train A(B) Operations.
              -OR-IF a Diesel Generator extended AOT is in Check for SI and progress AND a Wilson Black Start is necessary,                                            reset is repeated at THEN initiate 13419-C Diesel Generator                step 34 following Extended AOT Step 4.3.6.                              the the max rate cooldown
      *11. Check for SI:                            11.            termination steps.
: a. SI signal - ACTUATED.                    a. IF an SI signal exists or is actuated during this procedure, THEN reset SI.
IF SI will NOT reset, THEN initiate ATTACHMENT 9.
Go to Step 12.
: b. Reset SI.                                b. IF SI will NOT reset, THEN initiate ATTACHMENT 9.
Printed February 10, 2014 at 09:04
 
ECA - 0.0 Loss of all AC Power                                                19100-C VOGTLE                    Version 38.1 Unit C                  Page 20 of 50 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                            RESPONSE NOT OBTAINED
: b. As time permits, perform the following:
Evaluate securing unnecessary battery loads using ATTACHMENT 1.
Initiate 18032-1, 18032-2 Loss of 120V AC Instrument Power and 18034-1, 18034-2 Loss of Class 1E 125V DC Power if the following criteria is met:
Any Inverter must be shut down.
                            -OR-Any battery breaker must be opened due to battery overload or low DC Bus voltage.
NOTES The SGs should be depressurized at maximum rate to minimize RCS inventory loss.
PRZR level may be lost and Reactor Vessel Upper Head voiding may occur due to depressurization of the SGs. Depressurization should not be stopped to prevent these occurrences.
      *29. Depressurize intact SGs to 300 psig:            29.
: a. Check SG NR levels - GREATER                    a. Perform the following:
THAN 10% [32% ADVERSE] IN AT LEAST ONE SG.                                    (1) IF all SG NR levels less than 10% [32% ADVERSE],
THEN maintain maximum TDAFW flow.
(2) WHEN NR level in at least one SG greater than 10% [32% ADVERSE],
THEN go to Step 29.b.
Go to Step 33.
Printed February 10, 2014 at 09:04
 
ECA - 0.0 Loss of all AC Power                                          19100-C VOGTLE                    Version 38.1 Unit C                  Page 21 of 50 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                    RESPONSE NOT OBTAINED
: b. Locally dump steam using SG ARVs        b.
at maximum rate:
PV-3000 (South Main Steam Valve Room)
PV-3010 (North Main Steam Valve Room)
PV-3020 (North Main Steam Valve Room)
PV-3030 (South Main Steam Valve Room)
: c. Maintain the following during          c.
depressurization:
SG pressures - GREATER THAN 200 PSIG.
SG NR level - GREATER THAN            Stop depressurization until level is 10% [32% ADVERSE] IN AT                restored in at least one SG.
LEAST ONE INTACT SG
      *30. Check RCS WR cold leg temperatures -    *30. Perform the following:
GREATER THAN 280&deg;F [295&deg;F ADVERSE].
: a. Control SG ARVs to stop SG depressurization.
: b. Go to Step 33.
      *31. Check SG pressure - LESS THAN          *31. WHEN SG pressures lower to less than 300 PSIG.                                  300 psig, THEN go to Step 32.
Go to Step 33.
      *32. Locally control SG ARVs to maintain SG  32.
pressures at 300 psig.
Printed February 10, 2014 at 09:04
 
ECA - 0.0 Loss of all AC Power                                          19100-C VOGTLE                  Version 38.1 Unit C                Page 22 of 50 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                    RESPONSE NOT OBTAINED
      *33. Check Reactor - SUBCRITICAL:            *33. Control SG ARVs to stop SG depressurization and allow RCS to heat IR channels - ZERO OR NEGATIVE            up.
STARTUP RATE SR channels - ZERO OR NEGATIVE STARTUP RATE
      *34. Check SI signal status:                  34.
: a. SI - ACTUATED.                          a. IF SI actuates during SG depressurization, THEN reset SI.
IF SI will NOT reset, THEN initiate ATTACHMENT 9.
Go to Step 35.
: b. Reset SI.                                b. IF SI will NOT reset, THEN initiate ATTACHMENT 9.
: 35. Check Containment Isolation Phase A      35. Actuate Phase A.
using ATTACHMENT 2:
Computer Points                            IF valves do NOT close, THEN locally close at least one valve at
              -OR-                                      each penetration.
Handswitch Indication                      Locally close any open valve as time permits.
: 36. Check Containment Ventilation Isolation  36. Close dampers and valves.
using ATTACHMENT 3:
Computer Points                            IF dampers and valves can NOT be closed,
              -OR-                                      THEN locally close.
Handswitch Indication Printed February 10, 2014 at 09:04
 
Approved By                                                                                  Procedure    Version M.G. Brill                              Vogtle Electric Generating Plant                    19002-C        24 Effective Date                                                                                Page Number ES-0.2 NATURAL CIRCULATION COOLDOWN 05/01/2013                                                                                        13 of 24 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 23
: 23. Check that steam void in Reactor                    23. Repressurize RCS within limits of Vessel does NOT exist:                                  Technical Specification LCO 3.4.3 (PTLR) to collapse potential voids in PRZR level - NO UNEXPECTED                        system and continue cooldown.
LARGE VARIATIONS.
IF RCS depressurization must RVLIS Upper Range - GREATER                        continue, THAN 94%.                                          THEN go to 19003-C, ES-0.3 NATURAL CIRCULATION COOL DOWN WITH VOID IN VESSEL (WITH RVLIS).
24
        *24. Check if ECCS should be locked                      24.
out:
24.a
: a. Check RCS WR pressure - LESS                          a. WHEN RCS WR pressure is THAN 950 PSIG.                                        less than 950 psig, THEN go to Step 25.
Go to Step 29.
25
: 25. Isolate SI Accumulators:                            25.
25.a
: a. Dispatch an operator to close                        a.
ACCUM ISO VLV MOV breakers:
UNIT 1              UNIT 2 MOV CB ROOM          CB ROOM HV-8808A            1ABE-19 (B79)    2ABE-19 (B01)
HV-8808B            1BBC-19 (B61)    2BBC-19 (B18)
HV-8808C            1ABC-19 (B76)    2ABC-19 (B04)
HV-8808D            1BBE-19 (A77)    2BBE-19 (A79)
 
Step 25 continued on next page Printed February 10, 2014 at 08:52
 
A second limitation to controlled cooldown without ac power is related to the possibility of introducing noncondensible gases into the RCS.
Under normal plant shutdown conditions with ac power available, the accumulator injection lines are isolated prior to reducing RCS pressure below 1000 psig. This is done to prevent the accumulator contents from entering the reactor coolant system. In most cases, isolation of the injection lines will not be possible without ac power. Thus, following a total loss of ac event, depressurization to a pressure low enough to allow complete purging of the accumulators must be avoided. In general the operator's only means of doing this will be through controlling the amount of steam being released from the steam generators. Even so, if seal leakage from the RCPs becomes very large, the operator may not be able to control RCS pressure and avoid introducing nitrogen into the system.
The net effect of the limitations on cooldown following a total loss of ac power event is that the operator should be aware of the core criticality concern and how it is affected by fuel burnup. He should understand the desirability of cooling and depressurizing the RCS to reduce the potential for RCP seal failure and to replace lost RCS inventory with accumulator water.
Finally, he should be aware of the limiting low pressure necessary to prevent introduction of noncondensibles from the accumulators.
Understanding these considerations, he will be able to depressurize and control secondary pressure to minimize RCS inventory loss while preventing introduction of nitrogen into the RCS and return of the core to a critical condition.
2.4. Transient Analysis Quantitative Results In order to illustrate quantitatively the system responses to a loss of all ac power as described in the preceding subsection, several computer generated transients representing various scenarios are presented in this subsection. The results presented are for the high pressure (HP) reference plant (i.e., standard 4loop, 3425 MWt NSSS design); however, standard 2 and 3loop designs have also been analyzed and determined to exhibit responses similar to those presented herein. Where it is informative, results for the other ECA0.0 Background              24            HPRev. 2, 4/30/2005 HECA00BG.doc
 
STEP DESCRIPTION TABLE FOR ECA0.0        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 ECA0.0 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 (P1), initial nitrogen gas volume (V1), and final nitrogen gas volume (V2). The final nitrogen gas volume should be equivalent to the total accumulator tank volume.
The RCS pressure at empty tank conditions (P2) is determined from:
P1V1  = P2V2 where  = 1.25 for ideal gas expansion. The steam generator pressure limit is then determined by subtracting the RCS to SG delta p from P2 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.
ECA0.0 Background              141            HPRev. 2, 4/30/2005 HECA00BG.doc
: 1. 056AK1.01 001/LOIT/RO/M/F 3.7/4.2/056AK1.01/LO-TA-37006///
Initial condition:
            - Loss of off-site power has occurred.
Current condition:
            - 19002-C, "Natural Circulation Cooldown," is in progress.
Which one of the following completes the following statement?
Loop deltaT is expected to __(1)__ as natural circulation flow is FIRST being established, and with only 1 available CRDM fan, RCS subcooling is required to be maintained greater than a minimum of __(2)__ during the natural circulation cooldown.
__(1)__                                    __(2)__
A.                      rise                                      74&deg;F B.                      rise                                    124&deg;F C.                    lower                                      74&deg;F D.                    lower                                    124&deg;F K/A 056              Loss of Off-Site Power AK1.01          Knowledge of the operational implications of the following concepts as they apply to Loss of Off-Site Power:
                        - Principle of cooling by natural convection.
K/A MATCH ANALYSIS The KA addresses the relationship between LOSP and natural circulation cooldown. In addition, it refers to the 'principle' of natural circulation, which implies the fundamentals of natural circulation and the things that influence the process. The question has the candidate predict the change in loop delta-T power based on increasing natural circulation flow, and determine, based on the number of CRDM fans available, the amount of RCS subcooling required to prevent Reactor Head voiding and possible interruption of flow. All these are included in the principle of the process and the types of things that change or influence the actions.
Thursday, February 27, 2014 9:46:57 AM                                                              1
 
EXPLANATION OF REQUIRED KNOWLEDGE Per WOG Background NAT CIRC page 4, an additional indication of established natural circulation cooling is hot-to-cold leg temperature differences (ie delta T power) approximately equal to full-power forced convection temperature differences.
Per EOP 19002-C step 21, if at least two CRDM fans are running, the RCS subcooling requirement is >74F. If one or no CRDM fans are running, the RCS subcooling requirement is increased to >124F.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. In order to establish the driving head for natural circulation flow, a temperature difference is required between the heat source and heat sink. Per WOG Background NAT CIRC page 4, delta T power should approximately equal full-power forced convection delta T power.
The second part is incorrect. Per EOP 19002-C step 21, if at least two CRDM fans are running, required RCS subcooling is
                                      >74F. However, a candidate without specific knowledge of the consequences of inadequate head cooling may assume that the decision between 74F vs 124F is made based on any CRDM fan operating, and not recognize a mimum of two fans is required.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. Per EOP 19002-C step 21, if at least two CRDM fans are running, required RCS subcooling is
                                      >74F.
C. Incorrect. Plausible. The first part is incorrect. In order to establish the driving head for natural circulation flow, a difference in temperature is required between the heat source and heat sink. Per WOG Background NAT CIRC page 4, delta T power should approximately equal full-power forced convection delta T power.
However, the candidate may determine that the rising delta T indication implies inadequate heat removal by the secondary and the absence of natural circulation, as seen in a LOHS scenario.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, February 27, 2014 9:46:57 AM                                                                2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            056AK1.01 Importance Rating:              3.7 / 4.2 Technical
 
==Reference:==
EOP 19002-C, Rev 24.0, page 12 WOG Background NATCIRC, Rev 2, 4/30/2005, page 4 References provided:            None Learning Objective:              LO-LP-37012-05 Using 19002-C, 19003-C, and 19004-C as guides, summarize the actions of these emergency procedures which guide operator response in a natural circulation condition.
LO-LP-37012-15 State the limitations on subcooling and cooldown rate associated with natural circulation cooldown. Include the bases for any variations. (commitment)
LO-TA-37006        Conduct a Natural Circulation Cooldown per 19002-C Question origin:                NEW Cognitive Level:                M/F 10 CFR Part 55 Content:          41.8 / 41.10 / 45.3 Comments:
You have completed the test!
Thursday, February 27, 2014 9:46:57 AM                                                                3
 
Approved By                                                                              Procedure    Version M.G. Brill                              Vogtle Electric Generating Plant                19002-C        24 Effective Date                                                                            Page Number 05/01/2013 ES-0.2 NATURAL CIRCULATION COOLDOWN                            2 of 24 CONTINUOUS ACTIONS Step                              Actions 2                  - Monitor for SI actuation to go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.
3                  - Continue attempts to start an RCP to go to appropriate procedure.
4                  - Continue attempts to start CRDM fans as they come available for additional reactor vessel head cooling.
5                  - Maintain required boron concentration during cooldown.
10                - Maintain RCS cooldown rate less than 50&deg;F/Hr, SG levels at 65% and RCS temperature and pressure within required limits during cooldown.
18                - Maintain RCS pressure at 1950 psig and PRZR level at 25% after depressurization.
19                - Monitor RCS cooldown.
20                - IFcooldown rate must exceed 50&#xba;F/hour, go to 19003-C, ES-0.3 NATURAL CIRCULATION COOL DOWN WITH VOID IN VESSEL (WITH RVLIS).
21                - Maintain RCS subcooling greater than 74&deg;F (greater than 124&deg;F with less than two CRDM fans) during RCS depressurization.
22                - Maintain RCS cooldown rate less than 50&deg;F/Hr and subcooling greater than 74&deg;F (greater than 124&deg;F with less than two CRDM fans) while continuing RCS cooldown and depressurization.
24                - Monitor RCS pressure less than 950 psig to isolate SI Accumulators.
26                - Monitor RCS WR temperatures less than 350&deg;F to lockout ECCS Pumps.
29                - Maintain letdown flow as necessary.
30                - Maintain seal injection flow to all RCPs 8 to 13 gpm.
32                - Monitor RCS WR cold leg temperatures any less than 220&deg;F to arm COPS.
Printed February 10, 2014 at 10:23
 
Approved By                                                                            Procedure    Version M.G. Brill                              Vogtle Electric Generating Plant              19002-C        24 Effective Date                                                                          Page Number ES-0.2 NATURAL CIRCULATION COOLDOWN 05/01/2013                                                                                  12 of 24 ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED 21
        *21. Initiate RCS depressurization:                21.
21.a
: a. Verify CRDM Fans - AT LEAST                      a. Maintain RCS subcooling -
TWO RUNNING.                                      GREATER THAN 124&deg;F.
Go to Step 21.c 21.b
: b. Maintain RCS subcooling -                        b.
GREATER THAN 74&deg;F.
21.c
: c. Verify letdown - IN SERVICE.                  c. Depressurize RCS using one PRZR PORV.
Go to Step 22.
21.d
: d. Depressurize RCS using                        d.
Auxiliary Spray.
22
        *22. Continue RCS cooldown and                  22.
depressurization:
22.a
: a. Maintain cooldown rate in RCS                a.
Cold Legs - LESS THAN 50&deg;F/Hr.
22.b
: b. Maintain RCS subcooling -                    b. Stop depressurization.
GREATER THAN 74&deg;F (124&deg;F with less than two CRDM Fans                      Re-establish subcooling.
running).
22.c
: c. Maintain RCS temperature and                  c.
pressure - WITHIN LIMITS OF TECHNICAL SPECIFICATION LCO 3.4.3 (PTLR):
Use 60&deg;F/HR curve and RCS Cold Leg temperature.
 
S Printed February 10, 2014 at 10:23
 
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: 1. 057AK3.01 001/LOIT AND LOCT/RO/M/F 4.1/4/4/057AK3.01/LO-TA-37018///055EK1.01 Initial condition:
            - Unit 1 is at 100% reactor power.
Current conditions:
            - An LOSP occurs.
            - 19100-C, Loss of All AC Power," is in progress.
            - 1DD1 bus voltage is lowering.
Which one of the following completes the following statement?
Per 19100-C, as bus voltage lowers to less than __(1)__, the operator is required to remove 1DD1 from service using 13431-1, "120 VAC Vital Instrument Distribution System,"
and the reason for removing 1DD1 from service is to prevent __(2)__.
__(1)__                                __(2)__
A.                  100 VDC                    battery damage from cell reversal B.                  100 VDC                an explosion due to hydrogen production C.                  105 VDC                    battery damage from cell reversal D.                  105 VDC                an explosion due to hydrogen production K/A 057              Loss of Vital AC Instrument Bus:
AK3.01          Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC instrument Bus:
                        - Actions contained in EOP for Loss of Vital AC Instrument Bus.
K/A MATCH ANALYSIS The question tests the candidate's knowledge of the reason for removing a battery from service when the bus voltage drops below 105VDC during a Loss of All AC.
EXPLANATION OF REQUIRED KNOWLEDGE Per EOP 19100-C, Step 28, operators monitor 1E bus voltages to ensure they are >105 VDC. If battery voltage drops below 105 VDC, the inverter is removed from service and Thursday, February 27, 2014 9:50:17 AM                                                      1
 
the associated battery breaker opened. Per lesson plan V-LO-PP-41201, Slide 18, if battery voltage drops below 90 VDC, cell reversal can occur and permanently damage the cells. (Reference SER 3-99, Vendor Document AX3AD01-00025)
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per EOP 19100-C, Step 28, operators monitor 1E bus voltages to ensure they are >105 VDC. If a battery drops below 105 VDC, the inverter is removed from service and the associated battery breaker opened. However, Lesson Plan V-LO-PP-41201, Slide 7, discusses that the 1E DC system is designed to operate from 140-100 VDC. An exception to this is the inverters, which have a minimum safe voltage of 105 VDC. A candidate without specific knowledge of the procedure step could find both numbers reasonable and, with 100 VDC being 83% of rated voltage, may assume this value is high enough to support the required loads.
The second part is correct. Per lesson plan V-LO-PP-41201, Slide 18, if battery voltage drops below 90 VDC, cell reversal can occur and permanently damage the cells.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per lesson plan V-LO-PP-41201, Slide 18, if battery voltage drops below 90 VDC, cell reversal can occur and permanently damage the cells. However, EOP 19100-C, Step 27, states that all 1E SWGR and inverter room doors must be propped open within 30 minutes to prevent damage due to overheating. Lesson Plan V-LO-PP-41201, Slide 11, discusses how hydrogen gas is produced during the charging process. Battery discharge does not produce as much hydrogen gas, but this concept is commonly reversed by students to justify propping open battery room doors as well as 1E SWGR/Inverter room doors.
C. Correct.                  The first part is correct. Per EOP 19100-C, Step 28, operators monitor 1E bus voltages to ensure they are >105 VDC. If a DC bus drops below 105 VDC, the inverter is removed from service and the associated battery breaker opened.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Thursday, February 27, 2014 9:50:17 AM                                                              2
 
Level:                          RO Tier # / Group #                T1 / G1 K/A#                            057AK3.01 Importance Rating:              4.1 / 4.4 Technical
 
==Reference:==
EOP 19100-C, Rev 37.5, page 19 Lesson Plan V-LO-PP-41201, Rev 1.1, slides 7, 11, & 18 Vendor Manual AX3AD01-00025, Rev 6.0, page 30 INPO SER 3-99, page 8 & 9 References provided:            None Learning Objective:              LO-LP-60330-11 Describe why inverters are shutdown when associated battery bus voltage drops to 105 VDC.
LO-TA-37018    Respond to a Loss of All AC Power per 19100-C Question origin:                BANK Cognitive Level:                M/F 10 CFR Part 55 Content:          41.8 / 41.10 Comments:
You have completed the test!
Thursday, February 27, 2014 9:50:17 AM                                                          3
 
Approved By                                                                              Procedure    Version J.B. Stanley                          Vogtle Electric Generating Plant                  19100-C      37.5 Effective Date                                                                            Page Number ECA-0.0 LOSS OF ALL AC POWER 6/13/13                                                                                        19 of 54 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED
 
CAUTIONS Equipment failures and loss of control power may occur if doors are not opened within 30 minutes of onset of loss of AC power.
Room B76 on Unit 1 and room B04 on Unit 2 have two doors with installed door stops.
27
: 27. Open all doors that have installed                  27.
door stops in the following affected unit's Control Building electrical equipment rooms:
UNIT 1 B47, B48, B52, B55, B61, B76, B63 UNIT 2 B26, B29, B31, B36, B04, B18, B30 28
        *28. Check DC Bus loads:                                28.
28.a
: a. Monitor all 1E Battery Bus                            a. IF any 1E Inverter Battery voltages - GREATER THAN                                Bus voltage drops to 105V DC.                                              105V DC or less, THEN perform the following:
28.a.
: 1)  Verify associated Inverter(s) shutdown per 13431, 120V AC 1E VITAL INSTRUMENT DISTRIBUTION SYSTEM.
28.a.2
: 2)  Open associated battery breaker after Inverter(s) shutdown.
 
Step 28 continued on next page Printed November 12, 2013 at 09:13
 
Storage Batteries and Chargers
        - DC systems utilized as a reliable source of continuous power for control and instrumentation circuits.
        - (Review from Electrical Distribution) There are four safety features 125 VDC Trains (designated A, B, C, and D). Each Train has a Battery bank, switchgear, two Battery Chargers, one Inverter (A and B have two), and 125 VDC distribution panels.
        -Trains "A", "B", and "C" each have a 125 VDC Motor Control Center for motor operated valves.
        -The Batteries provide Emergency power to the DC buses if the Battery Chargers fail.
Each Battery is separately housed in a ventilated room (to limit the buildup of Hydrogen) apart from its Chargers and Distribution System and are sized to ensure that all Battery voltages are maintained > 106.2 Volts at the last minute of the 2.75 hour LOCA/LOSP discharge cycle or at the last minute of the 4-hour Station Black Out duty profile.
        -If a DC over voltage condition is sensed by a Battery Charger, the "Battery Charger Trouble" alarm is annunciated in the Main Control Room.
        - All equipment connected to the DC Power System is designed to operate at 140 VDC that exists during the 12-hour period that the Batteries are equalized. All equipment is also designed to operate at 100 VDC except the Inverter Systems that are designed to operate at 105 VDC.
        - Control power for many pump breakers will be lost with the DC bus being de-energized and the breakers cannot be remotely operated and protective trips will not open the breaker.
        - A Reactor Trip will occur if power is lost to AD1 or BD1 due to the closure of MSIVs and MFIVs.
        - Both chargers normally aligned to the bus and load share. Operating experience showed that with no battery on the bus the chargers will both try to carry the loads and will start developing oscillations. If the electrical bus must be energized by the battery chargers alone (without battery breaker closed in), only one charger should be energized to supply the bus.
V-LO-PP-41201                                                                                  7
 
Storage Batteries and Chargers (Objective - 1)
          - As a lead-acid battery is charged, the lead sulfate (PbSO4) is driven off the plates and back into the electrolyte (H2SO4). The return of acid to the electrolyte will reduce the sulfate on the plates and increase the specific gravity of the electrolyte. This will continue to happen until all of the lead sulfate is driven from the plates and back into the electrolyte.
charge 2PbSO4 + 2H2O PbO2 + Pb + 2H2SO4
          - A lead-acid battery cannot absorb all the energy from the charging source when the battery is nearing the completion of the charge. This excess energy dissociates water by way of electrolysis into hydrogen and oxygen. Oxygen is produced by the positive plate, and hydrogen is produced by the negative plate.
This action (gassing) occurs because the charging current is usually raised in an attempt to drive the remaining lead sulfate off the plates. The excess current also ionizes the water (H2O) in the electrolyte.
          - Excessive gassing is undesirable because it raises the acidity of the electrolyte which could damage the plates, produces an explosive gas, and if the gassing action is violent enough, it can dislodge active material from the grid structure.
          - Gassing is first noticed when cell voltage reaches 2.30-2.35 volts per cell and increases as the charge progresses. At full charge, the amount of hydrogen produced is about one cubic foot per cell for each 63 ampere-hours input. If gassing occurs and the gases are allowed to collect, an explosive mixture of hydrogen and oxygen can be readily produced. It is necessary, therefore, to ensure that the area in the vicinity of the battery being charged is well ventilated and that it remains free of any open flames or spark-producing equipment.
          - As long as battery voltage is greater than 2.30 volts per cell, gassing will occur and cannot be prevented entirely. To reduce the amount of gassing, charging voltages above 2.30 volts per cell should be minimized (e.g., 13.8 volts for a 12 volt battery).
V-LO-PP-41201                                                                                  11
 
Storage Batteries and Chargers
          - 125 VDC batteries response to heavy loads.
          - On a loss of AC power (battery chargers), the:
                    *1E batteries are designed to operate at full load for 23/4 hours for a LOCA/LOSP and 4 hours for a SBO (Station Blackout).
                    *Non-1E batteries are designed to operate at full load for 2 hours.
          - The 1E batteries consist of 59 lead-calcium cells.
          - The expected minimum cell voltages for the last minute of the SBO 4-hour profile are as follows:
                    *Train A and B - 109.7 battery volts (1.86 volts/cell)
                    *Train C - 108.3 battery volts (1.835 volts/cell)
                    *Train D - 106.2 battery volts (1.80 volts/cell)
          - The longer a battery discharges, the quicker the discharge rate (voltage decay) becomes. The effect is exponential, not linear.
          - As loads are stripped battery voltage will take an initial rise and as they are loaded onto the bus the voltage will drop from the present.
          - Cell reversal (reverse polarity) will be seen by quick (step change) voltage drops (not caused by load changes). Industry guidance suggests that cell reversal (reverse polarity) may occur when battery voltage drops below 90 VDC.
          - Think ahead and be prepared to disconnect the battery from the DC bus by the time battery voltage drops to 100 volts.
          - The inverters for AY1A, BY1B, CY1A, and DY1B will not trip until electrical perturbations most likely cause inverter fuse blowing to occur.
          - For long-term discharge concerns, a battery can be damaged if voltage is allowed to go down to low.
          - It will take about 12 weeks for replacements to be manufactured and delivered.
V-LO-PP-41201                                                                            18
 
SNC VER. 6.0
`                                    SIGNIFICANT EVENT REPORT                          SER 3-99 The initial slow drop in voltage led station      managers, were new to their positions at Indian personnel to believe that the battery would        Point Unit 2 and had not received training on continue to discharge at a linear rate. Station    their roles and responsibilities during personnel were unaware that individual cell        emergencies.
voltage was a critical parameter for assessing the battery conditions.                            Consequences Emergency Preparedness                            Annunciators Unavailable t 9:55 p.m., voltage on the No. 24 T
he station declared a Notification of Unusual Event (NOUE) at 9:55 p.m.,
after losing power to the control room annunciators when the battery discharged. The A      battery had decreased to approximately 105 VDC and the static inverter supplying power to the No. 24 instrument bus wording in the emergency action level (EAL)        shut down, deenergizing the bus.
procedure for a loss of off-site power was unclear. With the station blackout relays                        Contingency Planning tripped, the unaffected 480-volt ESF buses                  No formal contingency plans could not be realigned to off-site power. The              were prepared for the eventual emergency procedures did not adequately                    loss of the instrument bus and address this situation, nor did the EAL table in            annunciators the emergency preparedness procedures. The EAL procedure was incomplete because it only      Most of the control room annunciators lost specified declaring an NOUE if all three          power at this time. The reactor coolant system sources of off-site power were unavailable.        was still at no-load temperature and pressure, The EAL procedure did not indicate that the        and operations management decided to delay basis for the declaration was the inability to    cooling down until the 6A bus and No. 24 provide off-site power to the 480-volt            instrument bus were restored to service.
engineered safety features buses.                  Station personnel were reluctant to initiate a Consequently, operators did not declare an        cooldown with the control room annunciators NOUE at 2:50 p.m. when off-site power had          unavailable because of the difficulties that been unavailable to the 480 volt ESF buses for    would be encountered in monitoring plant greater than 15 minutes.                          conditions. Initiating a plant cooldown with the annunciators out of service would have met Missed Opportunity                  the conditions for declaring an Alert.
Operators did not recognize that, although off-site power              Degradation of the No. 24 Battery was available, it could not be aligned to the 480-v Class 1E buses, and plant conditions met the technical basis for T    he No. 24 battery was a 125 VDC, 58-cell, lead-calcium battery with an 8-hour rating of 462 amps-hours to 1.81 volts per cell.
declaring an Unusual Event.                Technicians began monitoring battery terminal voltage three hours after the battery began Some senior reactor operators did not believe      discharging. Monitoring did not include that their job responsibilities included          determination of individual cell voltages, evaluating plant conditions with respect to        although industry guidance suggests that cell implementation of the emergency plan. Other        reversal may occur when battery voltage drops station personnel, and particularly senior        below 90 VDC.
7
 
`                                  SIGNIFICANT EVENT REPORT                            SER 3-99 coolant system feed-and-bleed cooling using Battery Left On-Line                  the pressurizer PORVs following a total loss of The battery was allowed to                  auxiliary feedwater.
continue discharging after the AC instrument bus lost power                Auxiliary Feedwater System When the No. 24 instrument bus static inverter                ne motor-driven AFW pump was tripped at 9:55 p.m., the battery had been discharging for 7.5 hours, and battery voltage was less than 100 volts. With the inverter out O        already unavailable because of the 6A bus outage; however, the one remaining motor-driven AFW pump and the TD-of service, there was no reason to leave the No. AFWP operated satisfactorily until the No. 24 24 battery connected to its DC bus, but the          instrument bus lost power. When the bus battery remained in service an additional three      deenergized, the flow control valve from the hours. Cell reversal occurred at approximately      TD-AFWP to the No. 24 steam generator failed 57 VDC, and battery voltage was 35 VDC              full open, as designed.
when it was disconnected.
Additional Risk Extended Shutdown                                          Operators challenged the turbine-driven AFW pump by periodically starting and I
nvestigation into the incorrectly set overcurrent relays revealed that other                  stopping the pump to maintain nonsafety-related breakers at the station also          levels in two of the steam might have incorrect trip setpoints. As a result,          generators.
a 45-day unplanned outage was required to inspect and test the plant electrical system.        Operators responded by operating the TD-AFWP in a batch mode, starting and stopping Effects on Important Systems and                    the pump as necessary to maintain adequate level in the No. 23 and 24 steam generators.
Components for Mitigating Core This posed an additional challenge on the AFW Damage                                              system because the TD-AFWP was more likely he Indian Point Unit 2 probabilistic risk    to fail when being restarted than it would if T      assessment (PRA) identifies systems and components based on their relative importance to preventing core damage operated continuously. The control room crew did not dispatch an operator to manually operate the failed open flow control valve.
during postulated accidents. The PRA also identifies risk significant operator actions that    Pressurizer Power-Operated Relief Valves may need to be performed to mitigate an accident.
The two most important systems for mitigating events at the station are the auxiliary feedwater I  ndian Point 2 operates with both pressurizer power-operated relief valve (PORV) block valves normally closed. Power to one of the block valves was lost when the 6A bus system and the pressurizer power-operated            deenergized.
relief valves. The two most important operator actions in these events are: starting and aligning the turbine-driven auxiliary feedwater pump (TD-AFWP) following a loss of the motor-driven AFW pumps and initiating reactor 8
: 1. 059A4.03 001/LOIT/RO/C/A 2.9/2.9/059A4.03/LO-TA-18018///
Initial conditions:
            - Unit 1 is at 30% reactor power.
            - MFP 'A' is in service.
Current condition:
            - Power is ramping up to 50% at 8% per hour.
Which one of the following completes the following statement?
As power is ramped up, MFP 'A' speed will __(1)__,
and the BFRVs will continue to throttle open until __(2)__.
A. (1) remain constant (2) fully open, then remain open B. (1) remain constant (2) reaching a prescribed steam flow, then fully close C. (1) increase (2) fully open, then remain open D. (1) increase (2) reaching a prescribed steam flow, then fully close K/A 059              Main Feedwater A4.03            Ability to manually operate and monitor in the control room:
                        - Feedwater control during power increase and decrease K/A MATCH ANALYSIS The KA addresses the relationship between Reactor Power changes and the feedwater system response, including the expected indications for main feedwater pump speed and feedwater control valve response.
EXPLANATION OF REQUIRED KNOWLEDGE Thursday, February 27, 2014 9:54:14 AM                                                    1
 
Per SOP 13506-C Attachment A step 1.5, the Main Feedwater Pump speed will increase as flow demand rises during the power ramp. The feed pump speed demand is calculated as a function of feedwater flow demand. The feed pump speed program accomplishes the following:
(1) Provides adequate pump head to ensure flow to the steam generators is maintained during an expected transient.
(2) Controls feed pump speed and pump head to optimize the position and throttling affect of the feedwater control valves.
(3) Four independent feedwater flow demands (one demand per loop) are calculated by the system. The loop demanding the greatest amount of flow will determine the pump speed demand.
Per SOP 13506-C Attachment A step 1.3, as reactor power rises the BFRVs throttle open to maintain SG levels on program. When the BFRVs reach a demand signal of 50.28 percent the MFRVs receive a signal to throttle open and the MFRVs along with the BFRVs work together to control Narrow Range Level.
When Steam Generator Steam Flow reaches 40 % flow rate the BFRVs receive a signal to start closing and will ramp close over an extended period of time (approximately 0.5% per minute of valve demand). The MFRVs will control Narrow Range level from there to 100% power.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per UOP 12004-C, prior to exceeding 5% RTP, the individual slave controller for the inservice MFP and the Master controller for the feed pumps are placed in AUTO. MFP speed will increase as a function of total feedwater demand signal. However, MFPs are in manual while feedwater is transitioned from AFW to the BFRVs. A candidate without sufficient knowledge of MFP operation may assume MFPs are left in manual until a higher power to prevent changes in MFP speed from adversely affecting SG level. This was the case prior to the installation of DFW.
The second part is incorrect. Per SOP 13506-C, BFRVs open until 40% steam flow is reached and the fully close at a rate of 0.5%/min. However, the BFIVs are maitained open after the MFRVs and MFIVs are placed in service. A candidate with insufficient knowledge of the BFRV control circuit may assume that BFRVs are maintained open once MFRVs are placed in service to keep flow through the lines and prevent thermal shock, just like the BFIVs.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. Per SOP 13506-C, BFRVs open until 40% steam flow is reached and then fully close at a rate of 0.5%/min.
C. Incorrect. Plausible. The first part is correct. Per UOP 12004-C, prior to exceeding Thursday, February 27, 2014 9:54:14 AM                                                                2
 
5% RTP, the individual slave controller for the inservice MFP and the Master controller for the feed pumps are placed in AUTO. MFP speed will increase as a function of total feedwater demand.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, February 27, 2014 9:54:14 AM                                                                3
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            059A4.03 Importance Rating:              2.9 / 2.9 Technical
 
==Reference:==
SOP 13506-C, Rev 6.0, pages 68 & 70 UOP 12004-C, Rev 108.0, pages 16-20 & 28-29 References provided:            None Learning Objective:              LO-LP-61202-16 Describe the steps to transfer feedwater control from the bypass feedwater regulating valves to the main feedwater regulating valves (Include operator concerns during this transfer).
LO-PP-18101-15 Discuss the operation of the Main Feedwater Pump Turbine SPEED control to include:
: a. How the General Electric (G.E.)
potentiometer is used to control pump speed
: b. How the Westinghouse Controller is used to control pump speed
: c. How the Master Controller is used to control pump speed LO-TA-18017        Shift SG Level control from AFW to the Bypass Feed Regulating Valves using 12004-C LO-TA-18018        Shift SG Level control from the Bypass Feed Regulating Valves to the Main Feed Regulating Valves Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 / 45.5 to 45.8 Comments:
You have completed the test!
Thursday, February 27, 2014 9:54:14 AM                                                              4
 
Approved By                                                                                  Procedure Version R C.E. Williams                      Vogtle Electric Generating Plant                          13506-C Effective Date                                                                              Page Number 06/03/2013                          DIGITAL FEEDWATER CONTROL SYSTEM                            64 of 69 ATTACHMENT A                                Sheet 2 of 7 The sum of the NR Level Regulator and the Flow Regulator (wide range level feed-forward signal) is used to generate a compensated flow demand for the Valve Lift Calculator.
High Power Mode The High Power Mode uses Steam Flow/Feedwater Flow Rate-Lag Compensation.
This compares the rate at which the flow signals are changing instead of the difference between the signals. The error from the steam flow feed flow rate lag compensation is summed with the NR level error. Since 3 signals are used to create the error this is referred to as 3 element control. The combined error is used for PID control.
Power Mode Transition The transition between the Low Power Mode and the High Power Mode is based on the measured loop feedwater flow exceeding a predefined threshold.
During power ascension the transition to high power occurs at 17%
measured feed water flow.
On decreasing power the transition to low power occurs at 14%
measured feed water flow.
Both the High Power Mode and the Low Power Mode Track one another which enables a bumpless transfer from one mode to the other.
1.3        Design program for BFRVs and MFRVs during power ascent and descent As reactor power ascends the BFRVs throttle open to maintain SG levels on program. When the BFRVs reach a demand signal of 50.28 percent the MFRVs receive a signal to throttle open and the MFRVs along with the BFRVs work together to control Narrow Range Level. When Steam Generator Steam Flow reaches 40 %
flow rate the BFRVs receive a signal to start closing and will ramp close over an extended period of time (approximately 0.5% per minute of valve demand). The MFRVs will control Narrow Range level from there to 100% power.
During Power Decent the MFRVs will throttle to maintain SG Narrow Range level on program until Steam Generator Steam Flow lowers to < 30% Steam Flow. The BFRVs will throttle open and both the BRFVs along with the MFRVs will control together until the BFRV control signal lowers to 30.28% Demand.
Due to this being a low power level the MFRVs will be manually closed by the operator prior to going below 12 % power to comply with the FSAR 10.4.7.2.3.1.
RTYPE 0006
 
Approved By                                                                                    Procedure Version R C.E. Williams                      Vogtle Electric Generating Plant                            13506-C Effective Date                                                                                  Page Number 06/03/2013                          DIGITAL FEEDWATER CONTROL SYSTEM                                66 of 69 ATTACHMENT A                                    Sheet 4 of 7 The loss of controllers, force total responsibility of SG Level control for the failed loop(s) onto the operator. Once a controller is restored the M/A Stations return to Manual mode without operator action (either soft or M/A can then be used for control).
As previously stated each M/A Station communicates directly with its RLI I/O module. If a loss of communication between the M/A Station and the RLI I/O occurs the LED indicators on the M/A Station continually flash on and off until communication is restored. When this condition occurs the operator can still use the soft controls at the Operator Work Station to control the affected component.
1.5        Feed pump speed demand The feed pump speed demand is calculated as a function of feedwater flow demand. The feed pump speed program accomplishes the following:
Provides adequate pump head to ensure flow to the steam generators is maintained during expected transients.
Controls feed pump speed and pump head to optimize the position and throttling affect of the feedwater control valves.
Four independent feedwater flow demands (one demand per loop) are calculated by the system. The loop demanding the greatest amount of flow will determine the pump speed demand.
RTYPE 0006
 
Approved By                                                                            Procedure  Version Ronald M. Brown                    Vogtle Electric Generating Plant                    12004-C 108 Effective Date                                                                          Page Number 01/16/2014                              POWER OPERATION (Mode 1)                              16 of 119 INITIALS NOTES MFRVs and BFRVs may leakby as the BFIV opens, SG levels should be monitored and AFW flow adjusted to maintain SG levels constant.
The following step should be completed prior to continuing with Step 4.1.8.
: g.        Open the Bypass Feed Isolation Valve for all SGs one at a time. (1985303297, 1985305760, 1985304988, 1991321521)
SG 1      HV-15196                                  ________
SG 2      HV-15197                                  ________
SG 3      HV-15198                                  ________
SG 4      HV-15199                                  ________
Printed February 10, 2014 at 14:41
 
Approved By                                                                          Procedure  Version Ronald M. Brown                    Vogtle Electric Generating Plant                12004-C 108 Effective Date                                                                      Page Number 01/16/2014                              POWER OPERATION (Mode 1)                          17 of 119 INITIALS CAUTION During AFW forward flow operations of less than 150 gpm, correct mini-flow valve positions must be maintained. The mini-flow should be checked frequently.
4.1.8              Transfer from Auxiliary Feed Water Flow Control Valves to the Steam Generator BFRVs, one Steam Generator at a time, by performing the following: (1985303296 applies to Steps a-h)
: a.        Using the IPC or OWS, monitor flow as control is being transferred on each loop, IPC              OWS SG 1          UF-5404      DFW1FWFLOW            ________
SG 2          UF-5424      DFW2FWFLOW            ________
SG 3          UF-5444      DFW3FWFLOW            ________
SG 4          UF-5464      DFW4FWFLOW            ________
NOTE About 63% or 64% is the preferred SG level when transferring from AFW to the BFRVs.
: b.        Stabilize and maintain the SG NR level(s) between 60%
and 70% while transferring to BFRVs. (1985306829)        ________
Printed February 10, 2014 at 14:41
 
Approved By                                                                                  Procedure  Version Ronald M. Brown                    Vogtle Electric Generating Plant                        12004-C 108 Effective Date                                                                              Page Number 01/16/2014                              POWER OPERATION (Mode 1)                                  18 of 119 INITIALS NOTES The following step is performed after MFPT control has been transferred from the GE potentiometer to the M/A station per 13615, Condensate and Feedwater Systems.
By design the 0% to 100% range on the Ovation Speed Demand Signal correlates to a range from 3557 rpm to 6417 rpm.
When Process Variable is selected, the pump speed or rpm displayed in the digital window will not be accurate until actual pump speed is 3200 rpm or higher.
The display select button is used to select BIAS, Process Variable or Output in the digital window (BIAS, PV or OUT).
PV or Process Variable is displayed in RPM and OUT is displayed in percent, or PCT of setpoint demand.
The display select button may be depressed at any time and as often as necessary to switch parameters in the display window.
: c.        Raise speed of the Feed Pump using the up arrowhead
() on MFPT-A(B) controller, SIC-0509B(SIC-509C), to 17.4% output, which is approximately 4050 RPM.                  ________
NOTES Allowing SG level to stabilize slightly below 65% will increase the open output signal to the BFRV causing it to throttle open when placed in AUTO.
AFW Miniflow valves should be monitored for proper operation when throttling AFW flow.
: d.        For the SG to be transferred, adjust AFW flow until level is maintaining between 63% to 65%.
SG 1        HV-5139                                      ________
SG 2        HV-5132                                      ________
SG 3        HV-5134                                      ________
SG 4        HV-5137                                      ________
Printed February 10, 2014 at 14:41
 
Approved By                                                                          Procedure  Version Ronald M. Brown                    Vogtle Electric Generating Plant                  12004-C 108 Effective Date                                                                      Page Number 01/16/2014                              POWER OPERATION (Mode 1)                          19 of 119 INITIALS
: e.        Place the BFRV controller in AUTO mode and check the BFRV open Output signal increases.
SG 1        LIC-550                              ________
SG 2        LIC-560                              ________
SG 3        LIC-570                              ________
SG 4        LIC-580                              ________
NOTE Evidence of the BFRV beginning to throttle open should be anticipated between 10 and 20% output signal on the M/A Controller.
: f.        Close the Auxiliary Feed Water Supply Valve and check the associated BFRV throttles open:
AFW Supply Valve      BFRV SG 1        HV-5139        LV-5243              ________
SG 2        HV-5132        LV-5244              ________
SG 3        HV-5134        LV-5245              ________
SG 4        HV-5137        LV-5242              ________
: g.        Verify SG NR level is maintained at program.
SG 1                                              ________
SG 2                                              ________
SG 3                                              ________
SG 4                                              ________
Printed February 10, 2014 at 14:41
 
Approved By                                                                            Procedure  Version Ronald M. Brown                    Vogtle Electric Generating Plant                    12004-C 108 Effective Date                                                                          Page Number 01/16/2014                              POWER OPERATION (Mode 1)                            20 of 119 INITIALS
: h.        Repeat Steps d. through g. for the remaining Steam Generators:
SG 1      transferred to Main Feedwater            ________
SG 2      transferred to Main Feedwater            ________
SG 3      transferred to Main Feedwater            ________
SG 4      transferred to Main Feedwater            ________
NOTES To place the MFP speed controller in auto requires at least one BFRV or MFRV in AUTO.
ALL steam generator levels should be on program before placing the Main Feedpump in auto to prevent a rapid change in feedpump speed.
The MFP Output signal should be approximately 17.4% prior to placing its controller in Auto.
: i.        When ALL steam generator levels are approximately 65%, place the running MFP controller SIC-509B /
SIC-509C in AUTO.                                        ________
NOTES Whether in Auto or Manual, the Master Feed Pump Soft Controller will track feed pump slave output which allows a "bumpless transfer."
Master Feedpump Soft Control is accessed from OWS Graphic 10000.
: j.        At the Operator Work Station, place the MASTER FEEDPMP SPEED Control in AUTO.                              ________
: k.        Check DFWCS is operating properly and maintaining SG levels at approximately 65%.                                ________
Printed February 10, 2014 at 14:41
 
Approved By                                                                            Procedure  Version Ronald M. Brown                    Vogtle Electric Generating Plant                    12004-C 108 Effective Date                                                                          Page Number 01/16/2014                              POWER OPERATION (Mode 1)                            28 of 119 INITIALS 4.1.22            At greater than 12% Reactor Power, place the MFRVs in service as follows:
NOTES MFRVs may leak by as the MFIV opens, SG levels should be monitored and feedwater flow adjusted if required, to maintain SG levels.
The following annunciator(s) may actuate due to a temperature deviation caused by the cooler water entering the Steam Generators when the MFIVs are opened, ALB13-F03(4,5,6);
STM GEN 1 (2,3,4) DIGITAL FW CONTROL SYS TROUBLE All four MFIVs should be open prior to placing any MFRV in AUTO.
The following annunciator(s) may actuate while opening MSIVS, ALB16-A6,ALB16-B6,ALB16-C6,ALB16-D6, MFIV Accum Gas Lo Press.
: a.        Verify feed flow is stable and Steam Generator NR levels are stable between 63% and 67%.                            ________
: b.        Open the Main Feed Water Isolation Valves for all SGs one at a time: (1985306829, 1984301705)
SG 1          HV-5227                              ________
SG 2          HV-5228                              ________
SG 3          HV-5229                              ________
SG 4          HV-5230                              ________
: c.        One at a time place MFRV Flow Controllers in AUTO and verify feed flow and SG levels remain stable:
SG 1          FIC-510                              ________
SG 2          FIC-520                              ________
SG 3          FIC-530                              ________
SG 4          FIC-540                              ________
Printed February 10, 2014 at 14:41
 
Approved By                                                                              Procedure  Version Ronald M. Brown                    Vogtle Electric Generating Plant                      12004-C 108 Effective Date                                                                            Page Number 01/16/2014                              POWER OPERATION (Mode 1)                                29 of 119 INITIALS NOTES MFRVs should automatically begin to throttle open when the BFRV setpoint (SP) signal reaches 50.28%.
The display select button may be depressed at any time and as often as necessary to select SP, PV or OUT in the digital display window.
: d.        When BFRVs setpoint (SP) signal reaches 50.28%, check associated MFRV begins to throttles open:
SG-1    MFRV FV-510                                    ________
SG-2    MFRV FV-520                                    ________
SG-3    MFRV FV-530                                    ________
SG-4    MFRV FV-540                                    ________
4.1.23            Verify Main Turbine Warming is complete per 13800, "Main Turbine Operation" and continue with preparations for Main Turbine Roll.                                                          ________
CAUTION Ensure control of the EX2100 COI touchscreen from the Control Room is established and M1 is selected on the EX2100 COI touchscreen prior to Main Turbine Roll to prevent any delay in generator sync which can lead to high turbine vibrations.
4.1.24            Verify Generator/Exciter preparation for startup is complete per 13830, "Main Generator Operation."                                      ________
4.1.25            IF not previously performed, stop Sparging Steam to the Condenser per 13615, Condensate And Feed Water Systems."                ________
Printed February 10, 2014 at 14:41
: 1. 059AK2.02 001/LOIT/RO/M/F 2.7/2.7/059AK2.02/LO-TA-32007///071A1.06 004 Given the following:
            - Unit 1 CVCS mixed bed demin resin transfer is in progress.
            - A valve failure causes a liquid release on 'C' level of the Auxiliary Building, which results in airborne radioactivity.
Which one of the following completes the following statement?
The airborne radioactivity release will be monitored by __(1)__,
and the running Auxiliary Building HVAC units __(2)__ automatically trip.
A. (1) ARE-0014, Waste Gas Processing System Effluent Monitor (2) will NOT B. (1) 1RE-12442C, Plant Vent Radiogas Particulate (Low Range)
(2) will NOT C. (1) ARE-0014, Waste Gas Processing System Effluent Monitor (2) will D. (1) 1RE-12442C, Plant Vent Radiogas Particulate (Low Range)
(2) will Thursday, February 27, 2014 9:57:36 AM                                                            1
 
K/A 059              Accidental Liquid RadWaste Release AK2.02          Knowledge of the interrelations between the Accidental Liquid Radwaste Release and the following:
                        - Radioactive - gas monitors.
K/A MATCH ANALYSIS The questions tests the candidate's knowledge of Rad Gas monitors and the relationship of these rad monitors to the building HVAC during an accidental gaseous release.
EXPLANATION OF REQUIRED KNOWLEDGE During normal operation, 2 Aux Building Supply and 3 Aux Building Exhaust Filtration units are in service maintaining negative pressure in the Aux Building envelope. These units are nonsafety-related. During an accidental gaseous release, the effluent is filtered and the exhaust monitored by Plant Vent Rad Monitors RE-12442 and 12444.
These monitors are downstream of ARE-12442 and 12444 and have no automatic actuations.
ARE-0014 monitors the release pathway during discharge from the Gaseous Rad Waste system. During a planned gaseous release, ARE-0014 would actuate on receipt of a high rad signal and isolate the release path by closing ARV-0014.
Aux Building Normal Supply and Exhaust units use the Piping Pen Filtration units ductwork during normal operation. On receipt of a CVI signal, the Piping Pen units start and the duct work common to both the Piping Pen and Aux Building Normal systems isolates. Additionally, the Aux Building Normal HVAC units receive a direct trip signal from CVI to ensure these units do not pressurize the Aux Buidling due to the partial ductwork isolation.
CVI is generated by rad monitors RE-002, RE-003, and RE-2565.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. ARE-0014 monitors the release pathway during discharge from the Gaseous Rad Waste system. However, the candidate may determine that, since this is a gas 'release' and A-RE-014 monitors the Auxiliary Building HVAC ducting, it would detect this condition.
The second part is correct. The running Auxiliary Building Exhaust Units will NOT trip as a result of this condition. Aux Building Normal HVAC units receive a direct trip signal from CVI. CVI is generated by rad monitors RE-002, RE-003, and RE-2565.
Thursday, February 27, 2014 9:58:20 AM                                                              1
 
B. Correct.                  The first part is correct. RE-12442C is the low range detector that would be used to monitor this type of release.
The second part is correct. See the second part of choice A above.
C. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above..
The second part is incorrect. The running Auxiliary Building Exhaust Units will NOT trip as a result of this condition. Aux Building Normal HVAC units receive a direct trip signal from CVI. CVI is generated by rad monitors RE-002, RE-003, and RE-2565. However, A-RE-0014 does have automatic trip functions which isolate the release flowpath. It is reasonable for a candidate to assume that tripping the exhaust units downstream of this rad monitor might be a suitable secondary measure.
D. Incorrect. Plausible. The first part is correct. See the first part of choice B above.
The second part is incorrect. See the scond part of choice C above.
You have completed the test!
Thursday, February 27, 2014 9:58:20 AM                                                              2
 
Level:                        RO Tier # / Group #              T1 / G2 K/A#                          059AK2.02 Importance Rating:            2.7 / 2.7 Technical
 
==Reference:==
P&ID 1X4DB129, Rev 43.0 P&ID 1X4DB203, Rev 19.0 P&ID 1X4DB208-1, Rev 27.0 P&ID 1X4DB208-2, Rev 17.0 ELEMENTARY 1X3D BG-D02C, Rev 7.0 ELEMENTARY 1X3D BG-D02L, Rev 9.0 SOP 13503A-1, Rev 7.2, page 37 References provided:          None Learning Objective:            LO-PP-23101-02 Describe how the Auxiliary Building Normal HVAC System is operated to perform the following:
: a. Minimize the release of radioactive materials to the environment
: c. Prevent the buildup of airborne activity level in the Auxiliary Building LO-PP-23101-03 Describe the response of the Piping Penetration Area Filtration and Exhaust System upon receipt of the CVI signal when initiated by Safety Injection or by Containment High radiation.
LO-PP-23101-04 Describe how the Piping Penetration Area Filtration and Exhaust System is operated to perform the following:
: a. Minimize the release of radioactive materials to the environment
: b. Prevent the buildup of airborne activity levels in the Piping Penetration Areas and Rooms
: d. Maintain a negative pressure in the Piping Penetration Areas LO-PP-23101-05 Briefly describe how the Piping Penetration Area Filtration and Exhaust System interfaces with the Auxiliary Building Normal Supply and Exhaust System.
LO-PP-29101-21 State any auto actions that occur in the systems listed as a result of the following signals: SI, High Rad, and CVI.
: c. Preaccess (normal) Purge
: d. Mini Purge LO-PP-46101-11 State the events that require immediate termination of a gaseous release.
LO-PP-47101-08 Describe the major steps required for Thursday, February 27, 2014 9:59:06 AM                                                            1
 
Operations to release a WMT.
LO-PP-29101-12 Describe the importance of RE-12442C and RE-12444C on Containment Mini-Purge.
LO-PP-32101-09 Describe those automatic actions that occur for each of the following non-safety monitors when its high alarm setpoint is exceeded:
: a. containment vent effluent (RE-2565A, B, C)
: b. Waste Gas Processing System Effluent (ARE-0014)
LO-PP-46101-03 State the purposes of the following Gaseous Radwaste System components:
: e. trip valve RV-014
: f. rad monitors RE-013 and RE-014 LO-TA-32007        Verify Proper Automatic actions to high radiation alarms Question origin:                BANK - LOIT Question # 071A1.06 004 (HL-15R)
Cognitive Level:                M/F 10 CFR Part 55 Content:          41.7 / 41.11 / 41.13 Comments:
You have completed the test!
Thursday, February 27, 2014 9:59:06 AM                                                              2
 
Approved By                                                                                                          Procedure    Version J.B. Stanley                          Vogtle Electric Generating Plant                                              13503A-1 7.2 Effective Date                TRAIN A REACTOR CONTROL SOLID-STATE PROTECTION                                      Page Number 6/21/13                                                        SYSTEM                                                      37 of 38 ATTACHMENT C                                              Sheet 5 of 6 PERMISSIVES, CONTROL INTERLOCKS, REACTOR TRIPS AND ESF ACTUATIONS ESF ACTUATIONS Actuation            Setpoint                                  Function/Reset/Bases AFW                  MDAFWP          Lo Lo Level on 1SG      Ensures adequate heat sink.
Actuation                              SI (Train related)      MDAFWP Actuation will also close SGBD isolation Trip of Both MFPs      valves and SG sample valves.
LOSP (Train related)
AMSAC                  TDAFWP and MDAFWP actuation sends open signal to discharge MOVs.
TDAFWP          Lo Lo Lvl on 2 SGs      TDAFWP actuation sends open signal to HV-5106.
LOSP (either Train)    No reset switches but discharge MOVS may be AMSAC                  overridden to reduce flow (override indicated by white HS light).
Manual actuation consist of manually starting pumps and aligning valves CIA                  SI or 1/2 manual actuation HSs          Isolates non-essential Cnmt penetrations to prevent escape of radioactive material during post-accident conditions.
Manual actuation will also result in CVI.
May be reset anytime after actuation using Reset HSs (i.e. SI does not have to be reset in order to reset CIA).
CVI                  SI                                      Isolates Cnmt ventilation to prevent escape of Manual CIA                              radioactive material during post-accident conditions.
Manual CS                                If initiating signal is SI, then CVI may be reset using HSs on QMCB-A, even if SI has not been reset.
Hi Rad on RE-002 or 003 (1/2 logic)
If initiating signal was Hi Rad, then HSs on QMCB-C OR                                      may be used to reset CVI, even if Hi Rad still exist.
Either set of Reset HSs may be used if initiating CVI Hi Rad on RE-2565 (1/3 channels)        signal has cleared or has been removed/blocked.
Hi Rad from RE-002/003 is de-energize to actuate Hi Rad from RE-2565 is an energize to actuate function.
CRI                  SI                                      Pressurizes CR atmosphere and recirculates air through Hi Rad on RE-12116 or RE-12117          HEPA filters in order to maintain CR habitable Manual on 1/2 HSs on QHVC                May be reset using Reset HSs on QHVC after initiating signal is cleared.
Train B Filter Unit is lead with Train A as lag. Train A will start after time delay if Train B fails to start.
Rad monitor input may be manually blocked in QESF.
ESF Chilled Water pump start delayed (120 secs on Unit 1, 60 secs on Unit 2) to allow NSCW pressure interlock to be satisfied.
ESF Chiller starts after 500 gpm chilled water interlock satisfied.
Printed February 11, 2014 at 8:36
 
ARE-0014 and ARV-0014 to isolate tank release flow path
 
RE-12442 and 12444 skids on plant vent downstream of ARE-0014
 
2 Supply and 3 Exhaust units are normally in service.
Gas from duct work on previous sheets
 
Tank discharge line ties into ductwork in the Tank room        Flow path goes across several sheets to the Exhaust units
 
CVI contacts that trip the supply fan Trip Coil for supply breaker
 
CVI signals trip the exhaust fans
: 1. 059K4.19 001/LOIT/RO/M/F 3.2/3.4/059K4.19/LO-TA-28018///
Initial condition:
            - Unit 1 is at 100% reactor power.
Current conditions:
            - Reactor trips due to a feedwater transient.
            - ALB13-E04 FWI SI OR P14 SG HI-HI LVL is lit.
Which one of the following completes the following statement?
The Main Feedwater System will respond by __(1)__,
and in order to restore Main Feedwater capability, the reactor trip breakers __(2)__ required to be cycled.
A. (1) closing all FWI valves ONLY (2) are B. (1) closing all FWI valves ONLY (2) are NOT C. (1) closing all FWI valves and tripping both MFPTs (2) are D. (1) closing all FWI valves and tripping both MFPTs (2) are NOT 059              Main Feedwater (MFW) System K4.19            Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following:
                        - Automatic feedwater isolation of MFW K/A MATCH ANALYSIS The question tests the candidate's knowledge of the P-14 and P-4 interlocks associated with automatic FWI and how the isolation is reset.
EXPLANATION OF REQUIRED KNOWLEDGE Thursday, February 27, 2014 10:04:40 AM                                                        1
 
Per SSPS Logic 1X6AA02-00237, P-14 will initiate a full Feed Water Isolation (FWI) -
Main Feed Pumps trip and all main and bypass feed water isolation and regulating valves close.
P-4 has two interlocks associated with FWI isolation. First, P-4 in conjunction with Lo Tavg (<564F) closes all FW valves. Second, P-4 seals in an SI or P-14 signal if present.
Resetting FWI isolation requires that the SI signal be cleared or blocked and the P-14 signal must be cleared. Reactor trip breakers must be cycled to break the seal in. Lo Tavg FWI must be reset. EOP 19231-C step 22 depicts this sequence well.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. With P-14 present, MFPs will trip and all FWI valves will close. However, FWI isolation from P-4 and Lo Tavg will only close the FWI valves. A candidate without detailed knowledge of this circuit and interlocks may not realize P-14 also trips the MFPs and attribute pump trips only to SI.
The second part is correct. P-4 seals in the P-14 FWI isolation signal. Once P-14 is cleared, RTBs must be cycled to break the seal-in.
B. Correct.                  The first part is incorrect. See the first part of choice A above.
The second part is incorrect. P-4 seals in the P-14 FWI isolation signal. Once P-14 is cleared, RTBs must be cycled to break the seal-in. However, if the candidate does not realize that P-14 is sealed in by P-4, then it would be reasonable to believe that resetting FWI would be the only action required.
Furthermore, if the reactor were at 4% power and a P-14 were to occur, cycling the RTBs would not be necessary since a manual reactor trip would not be required due to being on AFW.
In this situation, clearing P-14 and resetting FWI would be all that was necessary.
C. Incorrect. Plausible. The first part is correct. With P-14 present, MFPs will trip and all FWI valves will close.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the wecond part of choice B above.
Thursday, February 27, 2014 10:04:40 AM                                                                2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            059K4.19 Importance Rating:              3.2 / 3.4 Technical
 
==Reference:==
1X6AA02-00237, Rev 10.0 EOP 19231-C, Rev 34.0, page 14 References provided:            None Learning Objective:              LO-PP-28103-05 List all ESF actuation signals with applicable set points, coincidences, permissives, blocks, and discuss the systems response to each ESF actuation signal.
LO-PP-16101-07 State the set point, coincidence, and the purpose for FWI on low Tavg.
LO-PP-20101-06 Given a set of plant conditions, analyze the data to determine if a secondary heat sink is properly established and any actions you would take to establish the heat sink if necessary.
LO-TA-37051    Respond to a Loss of Secondary Heat Sink per 19231-C LO-TA-28018    Reset Feedwater Isolation Question origin:                BANK - HL16 059K4.19 Cognitive Level:                M/F 10 CFR Part 55 Content:          41.7 Comments:
You have completed the test!
Thursday, February 27, 2014 10:04:40 AM                                                          3
 
FR - H 1 Response to Loss of Secondary Heat Sink                                                    19231-C VOGTLE                    Version 34 Unit C                Page 14 of 57 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                                  RESPONSE NOT OBTAINED CAUTIONS If offsite power is lost after SI reset, action is required to restart the following ESF equipment if plant conditions require their operation:
RHR Pumps SI Pumps Post-LOCA Cavity Purge Units Containment Coolers in low speed (Started in high speed on a UV signal).
ESF Chilled Water Pumps (If CRI is reset).
: 22. Perform the following:                                22.
: a. Reset SI.                                              a. IF SI will NOT reset, THEN initiate Attachment 5.
: b. Close RTBs.                                            b.
: c. Reset FW Isolation.                                    c.
: d. Energize Stub Busses by performing                    d.
the following as necessary:
NB01                      NB10
: 1) Open breaker            1) Open breaker NB01-01                    NB10-01
: 2) Close breaker            2) Close breaker AA02-22                    BA03-18
: 3) Close breaker            3) Close breaker NB01-01                    NB10-01 Printed 02/11/2014 at 10:54:00
 
P-4 seals in FWI.
Cycle RTBs to clear once P-14 is clear.
MFPTs trip and FW valves close
: 1. 061AA2.03 001/LOIT/RO/M/F 3.0/3.3/061AA2.03/LO-TA-32007///072A4.01 Radiation monitors are as follows:
            - 1RE-002, Containment Low Range Area Monitor
            - 1RE-003, Containment Low Range Area Monitor
            - 1RE-2565, Containment Vent Monitor Initial conditions:
            - Unit 1 is in Mode 6.
            - Core offload is in progress.
Current condition:
            - ALB06-E01 CNMT VENT ISO ACTUATION is received.
Which one of the following completes the following statement?
A HIGH radiation level of 15 mrem/hour has been exceeded on __(1)__,
and the HIGH radiation condition __(2)__ latch in on the SRDC panel requiring manual reset.
__(1)__                                  __(2)__
A.            1RE-002/1RE-003                                  does B.            1RE-002/1RE-003                                does NOT C.                  1RE-2565                                    does D.                  1RE-2565                                  does NOT K/A 061              ARM System Alarms AA2.03            Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms:
                        - Setpoints for alert and high alarms.
K/A MATCH ANALYSIS Thursday, February 27, 2014 10:06:30 AM                                                        1
 
The question tests the candidate's ability to recall the setpoint of the high alarms for the CVI area rad monitors RE-002 and 003 in comparison to CVI effluent rad monitor RE-2565.
EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17006-1 for ALB06-E01 CNMT VENT ISO ACTUATION and ARP 17100-1, a CVI can be generated by either rad monitors RE-002 and RE-003 or RE-2565A, B, & C.
Per TS 3.3.6 and ARP 17102-1, RE-002 and RE-003 high alarm setpoints are set to 15mR/hr during CORE ALTERATIONS. These two rad monitors are normally set at 50 times background. RE-2565 only reads in uCi/cc and monitors the Containment Purge Exhaust effluent.
Per ARP 17102-1 page 5, a high alarm on the SRDC will latch and require a manual reset. RE-002 and RE003 are located on the SRDC, RE-2565 is not.
ANSWER / DISTRACTOR ANALYSIS A. Correct.                  The first part is correct. Per ARP 17102-1, RE-002 and RE-003 high alarms are set to 15mR/hr during CORE ALTERATIONS.
The second part is correct. Per ARP 17102-1, any high alarm on the SRDC latches in and requires a manual reset.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. Per ARP 17102-1, any high alarm on the SRDC latches in and requires a manual reset. However, RE-2565A, B, & C are not on the SRDC. A candidate with insufficient knowledge of rad monitor locations may conclude that RE-002 and RE-003 are also not on the SRDC.
C. Incorrect. Plausible. The first part is incorrect. Per ARP 17102-1, RE-002 and RE-003 high alarms are set to 15mR/hr during CORE ALTERATIONS. However, RE-2565 reads in uCi/cc and does not monitor containment area radiation, but containment effluent. A candidate without sufficient knowledge of rad monitor setpoints could easily confuse RE-002/003 and RE-2565A, B, & C and believe RE-2565 reads in mR/hr.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Thursday, February 27, 2014 10:06:30 AM                                                            2
 
Level:                          RO Tier # / Group #                T1 / G2 K/A#                            061AA2.03 Importance Rating:              3.0 / 3.3 Technical
 
==Reference:==
TS 3.3.6, Amendment No. 158, page 3.3.6-6 ARP 17006-1, Rev 33.1, pages 44-46 ARP 17100-1, Rev 26.2, pages 54-59 ARP 17102-1, Rev 20.3, pages 5, 20, 21, 33, & 34 References provided:            None Learning Objective:              LO-LP-61208-04 State the various evolutions performed during Post Refueling Operations that are considered core alterations.
LO-LP-39207-01 For any given item in section 3.3 of Tech Specs, be able to:
: a. State the LCO.
: b. State any one hour or less required actions.
LO-TA-32007    Verify Proper Automatic actions to high radiation alarms Question origin:                BANK - HL18 Audit 072A4.01 Cognitive Level:                M/F 10 CFR Part 55 Content:          41.11 / 41.13 Comments:
You have completed the test!
Thursday, February 27, 2014 10:06:30 AM                                                          3
 
Containment Ventilation Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1)
Containment Ventilation Isolation Instrumentation APPLICABLE MODES OR OTHER                      REQUIRED                SURVEILLANCE FUNCTION                          SPECIFIED                    CHANNELS                REQUIREMENTS              TRIP SETPOINT CONDITIONS
: 1. Manual Initiation                        1,2,3,4                          2                    SR 3.3.6.6                    NA
: 2. Automatic Actuation Logic                                                                      SR 3.3.6.2                  NA and Actuation Relays                      1,2,3,4                          2                    SR 3.3.6.3 SR 3.3.6.5
: 3. Containment Radiation                                                                          SR 3.3.6.1 SR 3.3.6.4 1,2,3,4,6(c)
(a) 2 SR 3.3.6.7 SR 3.3.6.8 (b)
: a. Gaseous (RE-2565C)
(b)
: b. Particulate (RE-2565A)
(b)
: c. Iodine (RE-2565B) 15 mr/h(c)
: d. Area Low Range 50x background(d)
(RE-0002, RE-0003)
: 4. Safety Injection(d)                      1,2,3,4                Refer to LCO 3.3.2, "ESFAS Instrumentation,"Function 1, for all initiation functions and requirements.
(a)    Containment ventilation radiation (RE-2565) is treated as one channel and is considered OPERABLE if the particulate (RE-2565A) and iodine monitors (RE-2565B) are OPERABLE or the noble gas monitor (RE-2565C) is OPERABLE.
(b)    Setpoints will not exceed the limits of Specifications 5.5.4.h and 5.5.4.i of the Radioactive Effluent Controls Program.
(c)    During CORE ALTERATIONS and movement of irradiated fuel assemblies within containment.
(d)    During MODES 1, 2, 3, and 4.
Vogtle Units 1 and 2                                              3.3.6-6                      Amendment No. 158 (Unit 1)
Amendment No. 140 (Unit 2)
 
Approved By                                                                          Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                  17006-1      33.1 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 06 ON PANEL      Page Number 07/23/2013                                          1A2 ON MCB                            44 of 59 WINDOW E01 ORIGIN                          SETPOINT CNMT VENT ISO See Probable Cause.              Not Applicable          ACTUATION 1.0                PROBABLE CAUSE
: 1.        Manual CIA actuation.
NOTE
              *BPLB, 2.7, CVI RAD, and the applicable TSLB-4 CNMT HIGH RAD status light will be lit if caused by radiation relay.
: 2.        Any automatic CVI actuation including:
: a. RE-0002, SSPS Input Relay K164, Train A or B.*
: b. RE-0003, SSPS Input Relay K264, Train A or B.*
: c. RE-2565, SSPS Input Relay K711, Train A or B.*
: d. CVI SSPS Slave Relay K746.
Printed February 11, 2014 at 13:53
 
Approved By                                                                              Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                      17006-1      33.1 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 06 ON PANEL          Page Number 07/23/2013                                          1A2 ON MCB                                45 of 59 WINDOW E01 (Continued) 2.0                AUTOMATIC ACTIONS Automatic isolation of the following valves:
Computer Point          Valve                Description ZD9044                  HV-12975              CNMT AIR RAD MONITOR INL ZD9046                  HV-12976              CNMT AIR RAD MONITOR INL ZD9048                  HV-12977              CNMT AIR RAD MONITOR OUT ZD9050                  HV-12978              CNMT AIR RAD MONITOR OUT ZD9204                  HV-2626A              CNMT PREACCESS PURGE SUPPLY ZD9208                  HV-2627A              CNMT PREACCESS PURGE SUPPLY ZD9206                  HV-2626B              CNMT MINI PURGE SUPPLY ZD9210                  HV-2627B              CNMT MINI PURGE SUPPLY ZD9212                  HV-2628A              CNMT PREACCESS PURGE EXH ZD9216                  HV-2629A              CNMT PREACCESS PURGE EXH ZD9214                  HV-2628B              CNMT MINI PURGE EXH ZD9218                  HV-2629B              CNMT MINI PURGE EXH ZD9236                  HV-2624A              CTB POST LOCA PURGE EXH ZD9238                  HV-2624B              CTB POST LOCA PURGE EXH ZD9583                  HV-12604              AUX BLDG VENT SYS SUPPLY ZD9587                  HV-12605              AUX BLDG VENT SYS RETURN ZD9589                  HV-12606              AUX BLDG VENT SYS RETURN ZD9585                  HV-12607              AUX BLDG VENT SYS SUPPLY NONE                    HV-12596              RECYCLE HOLDUP TANK NONE                    HV-12597              RECYCLE HOLDUP TANK 3.0                INITIAL OPERATOR ACTIONS
: 1.        Check that valves and dampers align per automatic actions.
: 2.        Determine cause of CVI:
: a. Check CNMT HIGH RAD trip status lights on TSLB-4, 16.1 RE-0002 CNMT HIGH RAD, 16.2 RE-0003 CNMT HIGH RAD, 16.3 RE-2565 CNMT HIGH RAD,
: b. check for high radiation levels on the DRMS Communications Console and the Integrated Plant Computer.
Printed February 11, 2014 at 13:53
 
Approved By                                                                                    Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                          17006-1      33.1 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 06 ON PANEL              Page Number 07/23/2013                                            1A2 ON MCB                                    46 of 59 WINDOW E01 (Continued)
: 3.        Notify HP and plant management of CVI actuation due to high radiation.
: 4.        Take actions to mitigate cause of high radiation.
4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        IF equipment failure is indicated, initiate maintenance.
: 2.        WHEN the cause of the CVI has been corrected and it has been determined that the CVI actuated equipment is ready to return to service, initiate CVI recovery action per 11886-1, "Recovery From ESF Actuations."
: 3.        Refer to Technical Specification 3.4.15 due to the isolation of the containment atmosphere rad monitor, 1RE-2562.
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X6AA02-232, 2X6AX01-305, -383 Printed February 11, 2014 at 13:53
 
Approved By                                                                                  Procedure  Version S. E. Prewitt                      Vogtle Electric Generating Plant                        17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND              Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                      54 of 88 ORIGIN                            SETPOINT 1-RE-2565A Moving Paper                      As determined by            (High)
Airborne                          Chemistry Department Particulate Effluent Monitor NOTES This Moving Paper Particulate Monitor has 1 hour time constant so indicator changes should be slow.
For other than HIGH conditions see Pages 4 and 5.
1.0                PROBABLE CAUSE High level of radiation from airborne Particulates in Containment Purge Vent.
2.0                AUTOMATIC ACTIONS Initiates Containment Ventilation Isolation.
3.0                INITIAL OPERATOR ACTIONS
: 1.        Initiate evacuation of Containment IF the alarm is due to unexpected or unexplained radiation increases, OR IF appropriate HP controls are NOT in place for the radiological conditions indicated.
: 2.        IF the alarm is due to expected radiation increases from preplanned evolutions AND appropriate HP controls are in place, THEN request HP and Chemistry to investigate the cause of alarm. IF required, initiate evacuation of Containment.
4.0                SUBSEQUENT OPERATOR ACTIONS NOTE Exhaust gasses are monitored at the plant vent by 1-RE-12442 A, B and C.
: 1.        Verify Containment Ventilation Isolation by observing MLB Lights.
: 2.        Account for all personnel in the Containment.
: 3.        Notify Health Physics to survey and determine the source of radioactivity.
Printed February 11, 2014 at 13:57
 
Approved By                                                                                  Procedure  Version S. E. Prewitt                      Vogtle Electric Generating Plant                          17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND              Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                      55 of 88 1-RE-2565A (Continued)
: 4.        Check for increased level of radioactivity indicated on 1-RE-12442A.
: 5.        Refer to NMP-EP-110, "Emergency Classification And Implementing Instructions".
: 6.        Obtain detector trend data per 13508-1, "Radiation Monitoring Systems".
: 7.        Monitor the channel for further changes.
: 8.        IF the cause was a spurious alarm AND WHEN conditions permit, have Chemistry reset and return the monitor to normal.
: 9.        IF sampling and analysis determine the channel has malfunctioned:
: a. Comply with Technical Specification LCO 3.3.6.
Critical
: b. Unlock CVI BLOCK PANEL 1-1609-P5-CB3 (Equipment Bldg R-117), and place 1-HS-13261 to BLOCK,                                                    ________
Initial
________
CV Initial
: c. Request Chemistry to investigate and take corrective action.
: 10.      WHEN conditions permit, reset CVI per 11886-1, "Recovery From ESF Actuations".
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB213-1, 1X5DX4151 Printed February 11, 2014 at 13:57
 
Approved By                                                                                  Procedure  Version S. E. Prewitt                      Vogtle Electric Generating Plant                        17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND              Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                      56 of 88 ORIGIN                            SETPOINT 1-RE-2565B Effluent Iodine                    As determined by            (High)
Monitor                            Chemistry Department NOTE For other than HIGH conditions see Pages 4 and 5.
1.0                PROBABLE CAUSE High concentration of gaseous radioactive iodine in the Containment Purge Vent.
2.0                AUTOMATIC ACTIONS Initiates Containment Ventilation Isolation.
3.0                INITIAL OPERATOR ACTIONS
: 1.        Initiate evacuation of Containment IF the alarm is due to unexpected or unexplained radiation increases, OR IF appropriate HP controls are NOT in place for the radiological conditions indicated.
: 2.        IF the alarm is due to expected radiation increases from preplanned evolutions AND appropriate HP controls are in place, THEN request HP and Chemistry to investigate the cause of alarm. IF required, initiate evacuation of Containment.
4.0                SUBSEQUENT OPERATOR ACTIONS NOTE Exhaust gasses are monitored at the plant vent by 1-RE-12442A, B and C.
: 1.        Verify Containment Ventilation Isolation by observing MLB Lights.
: 2.        Account for all personnel in the containment.
: 3.        Notify Health Physics to survey and determine the source of radioactivity.
: 4.        Check for increased level of radioactivity indicated on 1-RE-12442B.
Printed February 11, 2014 at 13:57
 
Approved By                                                                                Procedure  Version S. E. Prewitt                      Vogtle Electric Generating Plant                      17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND            Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                    57 of 88 1-RE-2565B (Continued)
: 5.        Refer to NMP-EP-110, "Emergency Classification And Implementing Instructions".
: 6.        Obtain detector trend data per 13508-1, "Radiation Monitoring Systems".
: 7.        Monitor the channel for further changes.
: 8.        IF the cause was a spurious alarm AND WHEN conditions permit, have Chemistry reset and return the monitor to normal.
: 9.        IF sampling and analysis determine the channel has malfunctioned:
: a. Comply with Technical Specifications LCO 3.3.6.
Critical
: b. Unlock CVI BLOCK PANEL 1-1609-P5-CB3 (Equipment Bldg R-117), and place 1-HS-13261 to BLOCK,                                                  ________
Initial
________
CV Initial
: c. Request Chemistry to investigate and take corrective action.
: 10.      WHEN conditions permit, reset CVI per 11886-1, "Recovery From ESF Actuations".
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB213-1, 1X5DX4151 Printed February 11, 2014 at 13:57
 
Approved By                                                                                  Procedure  Version S. E. Prewitt                      Vogtle Electric Generating Plant                        17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND              Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                      58 of 88 ORIGIN                            SETPOINT 1-RE-2565C Effluent Radiogas                  As determined by            (High)
Monitor                            Chemistry Department NOTE For other than HIGH conditions see Pages 4 and 5.
1.0                PROBABLE CAUSE Increase in concentration of radioactive gas in the Containment Purge Vent.
2.0                AUTOMATIC ACTIONS Initiates Containment Ventilation Isolation.
3.0                INITIAL OPERATOR ACTIONS
: 1.        Initiate evacuation of Containment IF the alarm is due to unexpected or unexplained radiation increases, OR IF appropriate HP controls are NOT in place for the radiological conditions indicated.
: 2.        IF the alarm is due to expected radiation increases from preplanned evolutions AND appropriate HP controls are in place, THEN request HP and Chemistry to investigate the cause of alarm. IF required, initiate evacuation of Containment.
4.0                SUBSEQUENT OPERATOR ACTIONS NOTE Exhaust gasses are monitored at the plant vent by 1-RE-12442A, B and C.
: 1.        Verify Containment Ventilation Isolation by observing MLB Lights.
: 2.        Account for all personnel in the containment.
: 3.        Notify Health Physics to survey and determine the source of radioactivity.
: 4.        Check for increased level of radioactivity indicated on 1-RE-12442C.
Printed February 11, 2014 at 13:57
 
Approved By                                                                                Procedure  Version S. E. Prewitt                      Vogtle Electric Generating Plant                      17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND            Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                    59 of 88 1-RE-2565C (Continued)
: 5.        Refer to NMP-EP-110, "Emergency Classification And Implementing Instructions".
: 6.        Obtain detector trend data per 13508-1, "Radiation Monitoring Systems".
: 7.        Monitor the channel for further changes.
: 8.        IF the cause was a spurious alarm AND WHEN conditions permit, have Chemistry reset and return the monitor to normal.
: 9.        IF sampling and analysis determine the channel has malfunctioned:
: a. Comply with Technical Specifications LCO 3.3.6.
Critical
: b. Unlock CVI BLOCK PANEL 1-1609-P5-CB3 (Equipment Bldg R-117), and place 1-HS-13261 to BLOCK,                                                  ________
Initial
________
CV Initial
: c. Request Chemistry to investigate and take corrective action.
: 10.      WHEN conditions permit, reset CVI per 11886-1, "Recovery From ESF Actuations".
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB213-1, 1X5DX4151 Printed February 11, 2014 at 13:57
 
Approved By                                                                                  Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                          17102-1      20.3 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY                Page Number 11/4/13                                  RELATED DISPLAY CONSOLE QRM2                                5 of 42 Each channel on the SRDC has a separate display. Normally each display reads the radiation activity level being monitored in 3 digits and an exponent. Units vary from channel to channel. Each channel has an Alert, High and Equipment Trouble alarm display and an indicator that the SRDC is bypassed:
RED        YELLOW      BLUE        AMBER HIGH        ALERT    BYPASS      TROUBLE On detecting a high radiation level, the audible alarm on ALB 05 sounds and the red HIGH indicator lamp on the SRDC channel lights. The alarm is also indicated in the TSC and the Health Physics and Chemistry Labs, by displaying the channel identification number on their CRT in red. The alarm is also displayed at the Communications Console (QRM1). A loud horn and a strobe light may announce the high alarm close to the detector.
Alert, Bypass and Equipment Trouble indications do not sound audible alarms.
For very high radiation levels, the TOP OF SCALE, the EQUIPMENT TROUBLE and the HIGH alarms will all light and the sections of the digital display go to "9999999". This causes the alarm to latch, so it will not automatically clear when the radiation level drops. The TOP OF SCALE must be manually reset at the Channel Display and Control Area (CDCA) on the SRDC.
A high alarm will also latch and require a manual resetting.
The Bypass indicates that the channel has been put in the local control mode at the Data Processing Module (DPM) and for 1-RE-0002 or 1-RE-0003 the Containment Ventilation Isolation Block Switches have been placed in the BLOCK position.
Printed February 11, 2014 at 13:55
 
Approved By                                                                                  Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                          17102-1      20.3 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY                Page Number 11/4/13                                    RELATED DISPLAY CONSOLE QRM2                            19 of 42 WINDOW CDCA B3 ORIGIN                            SETPOINT 1-RE-0002 Containment Low                    As determined              (RED LAMP LIT)
Range Area Monitor                by Chemistry                (HIGH) 1-RE-0002                          (15 mR/hr during refueling)
NOTE For other than HIGH conditions see Pages 5 and 6.
1.0                PROBABLE CAUSE NOTES During refueling operations indicates a fuel drop accident.
During power operation indicates possible loss of coolant accident.
High radiation in the Containment Building.
2.0                AUTOMATIC ACTIONS Initiates Containment Building Ventilation Isolation.
3.0                INITIAL OPERATOR ACTIONS
: 1.        Initiate evacuation of Containment IF the alarm is due to unexpected or unexplained radiation increases, OR IF appropriate HP controls are NOT in place for the radiological conditions indicated.
: 2.        IF the alarm is due to expected radiation increases from preplanned evolutions AND appropriate HP controls are in place, THEN request HP and Chemistry to investigate the cause of alarm. If required, initiate evacuation of Containment.
Printed February 11, 2014 at 13:55
 
Approved By                                                                                    Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                          17102-1      20.3 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY                Page Number 11/4/13                                    RELATED DISPLAY CONSOLE QRM2                              20 of 42 WINDOW CDCA B3 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        Verify Containment Ventilation Isolation.
: 2.        If required, verify that Containment has been evacuated and all personnel accounted for.
: 3.        Notify Chemistry to independently determine radiation level in the Containment.
: 4.        IF the channel has malfunctioned:
: a.      Comply with Technical Specifications LCO 3.3.6.
: b.      Unlock panel and place 1-HS-13259 on CVI BLOCK PANEL 1-1609-P5-CB1 to BLOCK (Cont. Bldg RB-70).
: c.      Request Chemistry to investigate and take corrective action.
: d.      Reset CVI per 11886-1, "Recovery From ESF Actuations."
: 5.        IF the alarm is an actuation resulting from Fuel Handling, initiate 18006-C, Fuel Handling Event, as appropriate.
: 6.        IF the channel has not malfunctioned, initiate 18004-C, RCS Leakage.
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB213-2, 1X5DS4C02 Printed February 11, 2014 at 13:55
 
Approved By                                                                                  Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                          17102-1      20.3 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY                Page Number 11/4/13                                    RELATED DISPLAY CONSOLE QRM2                            33 of 42 WINDOW CDCA C3 ORIGIN                            SETPOINT 1-RE-0003 Containment Low                    As determined              (RED LAMP LIT)
Range Area Monitor                by Chemistry                (HIGH) 1-RE-0003                          (15 mR/hr during refueling)
NOTE For other than HIGH conditions see Pages 5 and 6.
1.0                PROBABLE CAUSE NOTES During refueling operations indicates a fuel drop accident.
During power operation indicates possible loss of coolant accident.
High radiation in the Containment Building.
2.0                AUTOMATIC ACTIONS Initiates Containment Building Ventilation isolation.
3.0                INITIAL OPERATOR ACTIONS
: 1.        Initiate evacuation of Containment IF the alarm is due to unexpected or unexplained radiation increases, OR IF appropriate HP controls are NOT in place for the radiological conditions indicated.
: 2.        IF the alarm is due to expected radiation increases from preplanned evolutions AND appropriate HP controls are in place, THEN request HP and Chemistry to investigate the cause of alarm. If required, initiate evacuation of Containment.
Printed February 11, 2014 at 13:55
 
Approved By                                                                                    Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                          17102-1      20.3 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY                Page Number 11/4/13                                    RELATED DISPLAY CONSOLE QRM2                              34 of 42 WINDOW CDCA C3 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        Verify Containment Ventilation Isolation.
: 2.        If required, verify that the Containment has been evacuated and all personnel accounted for.
: 3.        Notify Chemistry to independently determine radiation level in the Containment.
: 4.        IF the channel has malfunctioned:
: a.      Comply with Technical Specifications LCO 3.3.6.
: b.      Unlock panel and place 1-HS-13260 on CVI BLOCK PANEL 1-1609-P5-CB2 to BLOCK (Cont. Bldg RB-38).
: c.      Request Chemistry to investigate and take corrective action.
: d.      Reset CVI per 11886-1, "Recovery From ESF Actuations."
: 5.        IF the alarm is an actuation resulting from Fuel Handling, initiate 18006-C, Fuel Handling Event, as appropriate.
: 6.        IF the channel has not malfunctioned, initiate 18004-C, RCS Leakage.
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB213-2, 1X5DS4A02 Printed February 11, 2014 at 13:55
: 1. 061K2.01 001/LOIT AND LOCT/RO/M/F 3.2/3.3/061K2.01/LO-PP-20101-09///
Initial condition:
            - Unit 1 is at 100% reactor power.
Current conditions:
            - An electrical fault results in the following indications:
                  - 1HV-5106, TDAFW Steam Admission Valve, red and green handswitch lights are out.
                  - 1FV-5132 and 5134, MDAFW SGs #2 and 3 Discharge Throttle Valves, red, green, and white lights are out.
Which one of the following completes the following statement?
A loss of Train 'C' __(1)__,
and a loss of Train 'B' __(2)__ power have occurred.
__(1)__                                  __(2)__
A.                  125 VDC                                  480 VAC B.                  125 VDC                                  125 VDC C.                  480 VAC                                  480 VAC D.                  480 VAC                                  125 VDC K/A 061              Auxiliary / Emergency Feedwater (AFW) System K2.01            Knowledge of bus power supplies to the following:
                        - AFW system MOVs K/A MATCH ANALYSIS The question tests the candidates knowledge of the the AFW system MOV bus power supplies by requiring the candidate to select the specific voltage supplied to two different MOVs - one associated with TDAFW and the other with MDAFW.
EXPLANATION OF REQUIRED KNOWLEDGE Thursday, February 27, 2014 10:08:32 AM                                                    1
 
AFW is separated into 3 trains. The two motor driven trains are supplied from its train 4160VAC SWGR and the MOV from its train 480VAC MCC. The TDAFW train is steam driven off loops 1&2. The steam supply isolations from each loop are A or B train powered 125VDC. All other MOVs in the TDAFW system are train 'C' 125VDC powered.
The absence of light indication on the individual MOV handswitches denotes a loss of control power. The control power for the MDAFW pump discharge MOVs comes from a 480VAC/120VAC control power transformer inside the MCC bucket. (Ref ELEMENTARY 1X3D-BC-F08A&B) The control power for the TDAFW MOVs comes from the same 125VDC power that feeds the MOV itself inside the bucket. (Ref ELEMENTARY 1X3D-BC-F02A)
ANSWER / DISTRACTOR ANALYSIS A. Correct.                  The first part is correct. 1HV-5106 is powered from 125VDC MCC 1CD1M.
The second part is correct. 1FV-5132 is powered from 480VAC MCC 1BBB.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. 1FV-5132 is powered from 480VAC MCC 1BBB. However, the train 'B' steam supply from loop 1 to the TDAFW pump, 1HV-3009 is powered from 125VDC MCC 1BD1M. It is reasonable for a candidate without specific knowledge of bus power supplies to assume that all the AFW MOVs are powered from the associated train 125VDC.
C. Incorrect. Plausible. The first part is incorrect. 1HV-5106 is powered from 125VDC MCC 1CD1M. However, the vast majority of MOVs are 480VAC powered and 19100-C step 4 RNO directs the operator to locally operate HV5106 using an attachment in AOP 18034-1/2. Manual operation is a sign-off for both licensed and non-licensed operators. It is reasonable for an operator without specific knowledge of bus power supplies to assume that this guidance exists due to a loss of 480VAC power to HV5106 during a Loss of All AC. Additionally, RHR loop suction valves HV8701B and HV8702A are 480VAC MOVs that are powered from a 125VDC/480VAC inverter. It is reasonable for an operator to believe that AFW has a similar arrangement.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Thursday, February 27, 2014 10:08:32 AM                                                            2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            061K2.01 Importance Rating:              3.2 / 3.3 Technical
 
==Reference:==
EOP 19100-C, Rev 38.1, page 8 AOP 18034-1, Rev 13.1, page 77 & 78 ELEMENTARY 1X3D-BC-F02A, Rev 11.0 ELEMENTARY 1X3D-BC-F08A, Rev 9.0 ELEMENTARY 1X3D-BC-F08B, Rev 10.0 P&ID 1X4DB161-2, Rev 28.0 P&ID 1X4DB161-3, Rev 42.0 References provided:            None Learning Objective:              LO-PP-20101-09 Determine the impact to AFW system operation and the overall integrated plant operations to the following types of power supply failures:
: a. U/V condition on either AA02 or BA03 with the bus being re-energized from the EDG while at 100% power.
: b. U/V condition on either AA02 or BA03 with the bus remaining de-energized while at 100% power.
: c. Loss of a 120 VAC 1E vital instrument bus.
: d. Loss of a 125 VDC 1E bus
: e. Loss of All AC Power LO-TA-20009    TDAFW Pump Local Manual Control using HV-5106 and SOP 13610-1/2 Question origin:                BANK - HL16 NRC Question # 061K2.01 Cognitive Level:                M/F 10 CFR Part 55 Content:          41.7 You have completed the test!
Thursday, February 27, 2014 10:08:32 AM                                                          3
 
ECA - 0.0 Loss of all AC Power                                                  19100-C VOGTLE                    Version 38.1 Unit C                  Page 8 of 50 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED
: 4. Verify AFW flow - GREATER THAN                4. Perform the following:
570 GPM.
: a. Verify TDAFW Pump is running:
PDIC-5180 - INCREASED/ MAX DEMAND HV-5106 TDAFW - OPEN HV-3009 LP-1 MS SPLY TO AUX FW TD PMP OPEN
                                                                          -OR-HV-3019 LP-2 MS SPLY TO AUX FW TD PMP OPEN
: b. IF TDAFW Pump can NOT be operated normally due to governor or DC power failure, THEN dispatch Operator to attempt local manual control by initiating 18034-1, 18034-2 Loss of Class 1E 125V DC Power.
: c. Verify AFW Throttle Valves open.
: 5. Trip all RCPs.                                5.
: 6. Trip the NCP.                                  6.
: 7. Initiate the following:                        7.
Continuous Action Page.
NMP-EP-110 Emergency Classification Determination and Initial Action.
Printed February 17, 2014 at 09:28
 
Approved By                                                                              Procedure Number Rev J. Thomas                              Vogtle Electric Generating Plant                  18034-1        13.1 Date Approved                                                                              Page Number LOSS OF CLASS 1E 125V DC POWER 3/16/12                                                                                        77 of 84 ATTACHMENT C2                          Sheet 1 of 2 TURBINE DRIVEN AFW PUMP LOCAL MANUAL CONTROL WITHOUT DC POWER
 
NOTE This attachment gives instructions to operate the TDAFW Pump if DC bus 1CD1 is de-energized. In this case the TDAFW Pump governor valve will fail full open. Steam to the turbine must be manually throttled to prevent overspeed of the turbine and overpressurization of the piping. Some of the local instrumentation including turbine speed and pump discharge pressure will be unavailable.
CAUTION This attachment shall be used only with the permission of the SS.
1
: 1.      Establish communications between the Main Control Room and the TDAFW Pump Room.
2
: 2.      Locally verify closed TDAFW Pump Steam Supply 1-HV-5106.
3
: 3.      Locally verify open the TDAFW Pump Trip & Throttle valve 1-PV-15129.
 
NOTE Without power there is no direct indication of TDAFW pump speed. Miniflow indicator 1-FI-15100 should be used to ascertain TDAFW Pump performance. 140 gpm recirculation flow corresponds to a speed slightly less than normal speed of 4230 rpm with AFW throttle valves shut. If throttle valves are opened, 140 gpm corresponds to a speed above the normal speed of 4230 rpm but still less than the overspeed setpoint of 4830 rpm.
4
: 4.      Throttle open TDAFW Pump Steam Supply 1-HV-5106 while observing 1-FI-15100.
Continue to open valve until approximately 140 gpm is observed on 1-FI-15100.
 
S Printed February 17, 2014 at 09:44  2
 
Approved By                                                                              Procedure Number Rev J. Thomas                              Vogtle Electric Generating Plant                  18034-1        13.1 Date Approved                                                                            Page Number LOSS OF CLASS 1E 125V DC POWER 3/16/12                                                                                        78 of 84 ATTACHMENT C2                        Sheet 2 of 2 TURBINE DRIVEN AFW PUMP LOCAL MANUAL CONTROL WITHOUT DC POWER NOTE If TDAFW Pump trips, it should be reset using 13610, AUXILIARY FEEDWATER SYSTEM.
5
: 5.      Adjust steam supply valve position as necessary to control pump speed and flow:
5.a
: a. Speed indication (strobe) available - Throttle steam flow to limit speed to 4230 rpm.
5.b
: b. Speed indication not available - Limit recirculation flow indicated on 1-FI-15100 to 140 gpm or discharge pressure to 1700 psig.
6
: 6.      When operation of the TDAFW pump is no longer required:
6.a
: a. Trip pump using the Manual Overspeed Trip Mechanism Thumb Lever.
6.b
: b. Restore pump to standby per 13610, AUXILIARY FEEDWATER SYSTEM.
END OF ATTACHMENT C2
 
END OF ATTACHMENT C2 Printed February 17, 2014 at 09:44  2
: 1. 062K3.02 001/LOIT AND LOCT/RO/C/A 4.1/4.4/063K3.02/LO-TA-60012///063K3.01 Initial conditions:
            - Unit 1 at 100% reactor power.
            - 1AY2A is de-energized.
Current condition:
            - Safety Injection is actuated.
Which one of the following completes the following statement?
DG1A __(1)__ automatically start, and the Train A sequencer __(2)__ run the Safety Injection load sequence.
__(1)__                                  __(2)__
A.              will                                    will B.              will                                will NOT C.          will NOT                                    will D.          will NOT                                will NOT K/A 062              AC Electrical Distribution K3.02            Knowledge of the effect that a loss or malfunction of the AC distribution system will have on the following:
                        - ED/G K/A MATCH ANALYSIS The question tests the candidates knowledge of the effect that a loss of 120VAC Vital AC bus will have on the operation of the ED/G and its associated sequencer during a Safety Injection actuation.
EXPLANATION OF REQUIRED KNOWLEDGE During a loss of 1AY2A, the train 'A' sequencer is de-energized. Therefore, when a Safety Injection actuation occurs, the SI sequence will not run and no equipment will change state. During an SI actuation, the ED/G receives a start signal from the Thursday, February 27, 2014 10:10:07 AM                                                      1
 
sequencer. With 1AY2A de-energized, this start signal will not be generated. However, the ED/G also receives a start signal from SSPS during an SI acutation. This signal will be received since 1AY1A remains energized. The DG will start and run unloaded with its output breaker open.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. The ED/G will receives a start signal from SSPS during an SI actuation, independent of the sequencer.
The second part is incorrect. During a loss of 1AY2A, the train
                                      'A' sequencer is de-energized. Therefore, when a Safety Injection actuation occurs, the SI sequence will not run and no equipment will change state. However, candidates frequently confuse the power supply to SSPS slave relays (AY1A) and the power supply to the sequencer (AY2A). If the candidate confuses the power supplies, they may conclude that the sequencer will run the SI sequence.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. During a loss of 1AY2A, the train 'A' sequencer is de-energized. Therefore, when a Safety Injection actuation occurs, the SI sequence will not run and no equipment will change state.
C. Incorrect. Plausible. The first part is incorrect. The ED/G receives a start signal from SSPS during an SI actuation, independent of the sequencer.
However, candidates frequently confuse the power supply to SSPS slave relays (AY1A) and the power supply to the sequencer (AY2A). If a candidate reversed the two and believed the ED/G only received a start signal from SSPS during an SI, then it would be reasonable for the candidate to assume the ED/G would not start, but the sequencer would run the SI sequence.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, February 27, 2014 10:10:08 AM                                                                2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            063K3.02 Importance Rating:              4.1 / 4.4 Technical
 
==Reference:==
ELEMENTARY 1X3D-BH-G03C, Rev 7.0 ONE-LINE 1X3D-AA-G02A, Rev 28.0 ONE-LINE 1X3D-AA-G02C, Rev 14.0 References provided:            None Learning Objective:              LO-LP-60324-04 State the effect on SSPS and Sequencer operation on loss of the following:
: a. 1AY1A
: b. 1AY2A LO-TA-60012    Respond to a Loss of Vital Instrument Power per 18032-1/2 Question origin:                MODIFIED - HL17 NRC # 063K3.01 Cognitive Level:                C/A 10 CFR Part 55 Content:          41.7 Comments:
You have completed the test!
Thursday, February 27, 2014 10:10:08 AM                                                      3
: 1. 063K3.01 001/2/1/LOSS DC DG STARTING/F 3.7/4.1/NEW/H-17 NRC/RO/SRO/EMT/GCW Unit 1 is at 100% power.                        Original Question
          - A loss of 125V DC distribution panel 1AD11 occurs.
Which one of the following correctly completes the following statement?
DG1A __(1)__ be started, and if DG1A was running prior to the loss of 1AD11, the stop pushbuttons on the QEAB
__(2)__ stop the DG.
__(1)__              __(2)__
A. can                  will NOT B. can                  will C. can NOT              will NOT D. can NOT              will Monday, February 17, 2014 11:05:22 AM                                                    1
 
SSPS Power supply
 
Sequencer power supply
 
ED/G start signals from SSPS and Sequencer
: 1. 062K4.07 001/LOIT/RO/C/A 2.7/3.1/062K4.07/LO-TA-60012///
Initial conditions:
            - Unit 1 is at 100% reactor power.
            - DG1A is tagged out for maintenance.
Current condition:
            - Unit 1 experiences an LOSP.
Which one of the following completes the following statement?
1CY1A is normally powered by a(an) __(1)__,
and 1CY1A will be __(2)__ after the LOSP occurs.
__(1)__                              __(2)__
A.                  inverter                            energized B.                  inverter                            de-energized C.          regulated transformer                        energized D.          regulated transformer                        de-energized Thursday, February 27, 2014 10:12:43 AM                                      1
 
K/A 062              AC Electrical Distribution K4.07            Knowledge of AC distribution system design feature(s) and/or interlock(s) which provide for the following:
                        - One-line diagram of 4 kVAC to 480 VAC distribution, including sources of normal and alternative power.
K/A MATCH ANALYSIS The question tests the candidates knowledge of AC distribution design features to include how the 125Vdc bus powers 1CY1A via an inverter. Alternately, 1CY1A can be powered from a 480VAC transformer. Both alignments are used in the stem to test the knowledge of the candidate related to the given scenario and the impact on the AC electrical distribution system.
EXPLANATION OF REQUIRED KNOWLEDGE During a LOSP with the 1A DG tagged out, 1CY1A will energize or de-energize based on the source from which it is powered. Normally, 1CY1A is energized by a DC inverter fed from the 125VDC battery backed bus 1CD1. During the LOSP, 1CD1 will remain energized from its battery bank even thought the two chargers on the bus are de-energized. In this configuration, 1CY1A will remain energized for several hours until battery capacity is depleted. The design basis for battery depletion is 2.75 hours for SI and 4 hours for station blackout.
Alternately, 1CY1A can be energized by a 480VAC regulated transformer. This transformer is 1E powered by a 480V MCC, via a 480V Switchgear, from the 4160V 1E bus 1AA02. During the LOSP, the regulated transformer will de-energize resulting in a loss of power to 1CY1A also. The sequencer will sense the UV condition and perform a load shed. The bus will remain de-energized since the 1A DG is tagged out. Normally, the DG will tie to the bus and the 480V MCC will be loaded back at the first step of the sequence. Sufficient fuel is stored on site to allow the DG to operate for 7 days.
ANSWER / DISTRACTOR ANALYSIS A. Correct.                  The first part is correct. 1CY1A is normally supplied by an inverter from 1CD1.
The second part is correct. The loss of the 4160V bus will not immediately effect 1CD1 and 1CY1A will remain powered through the inverter from 1CD1.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. The loss of the 4160V bus will not immediately effect 1CD1 and 1CY1A will remain powered through the inverter off 1CD1. However, a candidate with Thursday, February 27, 2014 10:13:26 AM                                                            1
 
insufficient knowledge of electrical distribution could remember that the bus is normally aligned to the inverter, but believe the inverter will load shed as part of the LOSP sequence and not re-energize because the DG is tagged out.
C. Incorrect. Plausible. The first part is incorrect. 1CY1A is normally supplied by an inverter from 1CD1. However, a candidate with insufficient knowledge of the distribution system who realizes the bus can be powered from both sources may believe that the regulating transformer is the normal source, and that it would auto transfer to the inverter during a loss of power. Component power sources auto transferring to a battery backup is a common design feature for important systems.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
You have completed the test!
Thursday, February 27, 2014 10:13:26 AM                                                                2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            062K4.07 Importance Rating:              2.7 / 3.1 Technical
 
==Reference:==
SOP 13431-1, Rev 31.1, pages 12 & 15 One-line 1X3D-AA-H04A One-line 1X3D-AA-G02A References provided:            None Learning Objective:            LO-PP-01101-05 Describe how a failure of DC control power affects the electrical distribution system and its components.
LO-LP-60324-02 Given that a loss of 120VAC instrument power has occurred to any of the following panels, and given the appropriate plant procedures, describe the operator actions required and why these actions are taken.
: e. 1CY1A LO-LP-60329-02 Given the appropriate drawings, logics, and/or procedures, describe how the plant will respond to a loss of the following 125V DC vital buses:
: c. 1CD1 LO-TA-01023        Draw the electrical distribution system LO-TA-01025        Energizing A Vital Instrument Distribution Panel LO-TA-60012        Respond to a Loss of Vital Instrument Power per 18032-1/2 Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:        41.7 Comments:                      Question appears to match the KA.
The last paragraph of "explanation of required knowledge" seems to be missing something between the end of that page and the middle of the sentence that starts on the next page.
It seems implausible that a candidate would not know that 1CY1A cannot simultaneously have alternate and normal supply breakers closed - are there many examples of 120VAC distribution panels at Vogtle where this is the case?
If not, there are two implausible distractors.
The first question asked appears to be somewhat backwards logic - I would recommend stating what the Thursday, February 27, 2014 10:14:17 AM                                                              1
 
source is for the power (normal or alternate, rather than inverter vs regulating transformer), and then ask if panel 1CY1A is energized or deenergized. This way, the applicant would have to know which source (inverter or regulating transformer) is the normal and which is the alternate source.
Additionally, it gets away from teaching in the question (i.e.,
the way the question is framed, they know that one of the sources of power keeps the panel energized, and if I had to take a guess not knowing anything about the system, I would have guessed the inverter).
                                        -JAT 12/19/2013 (U/E)
The new question incorporates the above suggestion, however, more information in the stem may be required.
                                        - JAT 2/4/2014 You have completed the test!
Thursday, February 27, 2014 10:14:17 AM                                                              2
 
Approved By                                                                                  Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                      13431-1 31.1 Effective Date                                                                              Page Number 8/2/13                    120V AC 1E VITAL INSTRUMENT DISTRIBUTION SYSTEM                        12 of 60 INITIALS 4.2                SYSTEM OPERATION 4.2.1              Transferring Vital Instrument Distribution Panel To Alternate Source CAUTIONS A Vital Panel should not be transferred to its alternate source if the battery is not connected to the 125V DC bus.
1-HV-15214 will close when 1DY1B is swapped to the alternate source. 1-HV-8160 will close when 1CY1A is swapped to its alternate source. Closure is permissible during solid plant conditions for short duration since PSV-8117 provides pressure relief for low pressure letdown piping. 1-HV-15214 and 1-HV-8160 should be re-opened after their associated bus is re-energized.
A reactor trip signal will be generated from Source range and Intermediate Range instrumentation if 1AY1A or 1BY1B is transferred below P-10.
4.2.1.1            To preclude any unwanted actuations from opening Instrument Distribution Panel AC Breaker, review the appropriate attachment of 18032-1 Loss of 120 Volt AC Instrument Power and the 120V AC Load Database for the Vital Instrument Panel to be transferred, to determine what equipment will be impacted and any contingency actions required.                                          ________
4.2.1.2            If transferring 1AY1A to alternate source, perform attachment 1 as directed.                                                                  ________
4.2.1.3            If transferring 1BY1B to alternate source, perform attachment 2 as directed.                                                                  ________
4.2.1.4            IF transferring 1DY1B, perform either a or b as directed by SS:
: a.        IF it is desired to keep CVCS letdown in service, perform Attachment 3.                                                    ________
: b.        IF CVCS Letdown will be removed from service, continue to step 4.2.1.6.                                                ________
Printed February 17, 2014 at 11:51
 
Approved By                                                                          Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                13431-1 31.1 Effective Date                                                                        Page Number 8/2/13                    120V AC 1E VITAL INSTRUMENT DISTRIBUTION SYSTEM                  15 of 60 INITIALS 4.2.1.12          IF transferring 1CY1A OR 1DY1B, remove SGBD and CVCS Letdown from service to prevent auto isolation from occurring:
SGBD removed from service                                ________
CVCS Letdown removed from service                        ________
NOTE Steps 4.2.1.13 and 4.2.1.14 should be performed as quickly as possible to minimize the time the Instrument Distribution Panel is de-energized.
4.2.1.13          Open Instrument Distribution Panel AC Breaker:                      ________
PANEL      PANEL AC BKR 1AY1A      1AY1A-02 1AY2A      1AY2A-02 1BY1B      1BY1B-02 1BY2B      1BY2B-02 1CY1A      1CY1A-02 1DY1B      1DY1B-02 4.2.1.14          Close the Instrument Distribution Panel AC Breaker from the regulated source:                                                  ________
PANEL      PANEL AC BKR 1AY1A      1AY1A-01 1AY2A      1AY2A-01 1BY1B      1BY1B-01 1BY2B      1BY2B-01 1CY1A      1CY1A-01 1DY1B      1DY1B-01 Printed February 17, 2014 at 11:51
: 1. 063A1.01 001/LOIT/RO/C/A 2.5/3.3/063A1.01/LO-TA-37018///
Initial condition:
            - Unit 1 is at 100% reactor power.
Current conditions:
            - Unit 1 reactor is tripped.
            - 19100-C, "Loss of All AC Power," is in progress.
Which one of the following completes the following statement?
Based on the given conditions, 1CD1B batteries have sufficient capacity to supply the required loads for __(1)__ hours, and in order to minimize drain on the batteries, control of the TDAFW feed flow to the SGs is done using __(2)__.
__(1)__                                      __(2)__
A.              4                1-PDIC-5180A, TDAFW pump speed controller B.              4                      TDAFW pump discharge throttle valves C.            6.75                1-PDIC-5180A, TDAFW pump speed controller D.            6.75                    TDAFW pump discharge throttle valves K/A 063              DC Electrical Distribution A1.01            Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including:
                          - Battery capacity as it is affected by discharge rate.
K/A MATCH ANALYSIS The question addresses how battery load and operator action determine discharge rates. This includes the procedural actions associated with TDAFW valve operation that were incorporated to extend the life on the batteries.
EXPLANATION OF REQUIRED KNOWLEDGE Per TS 3.8.4 BASES, the 1E batteries are sized to have sufficient capacity to supply the Thursday, February 27, 2014 10:16:37 AM                                                      1
 
required loads for 2.75 hours during a LOSP with a LOCA and for 4.0 hrs during a station blackout. In order to meet this requirement, non-required loads are stripped from the bus and operation of DC components is minimized. Per SOP 13610-1 Limitation 2.1.8, if it is necessary to minimize drain on the C-Train batteries during emergency conditions, TDAFW pump speed may be adjusted in lieu of, or in addition to, using the pump discharge throttle valves to control AFW flow to the SGs.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. The 1E batteries are sized to have sufficient capacity to supply the required loads for 4.0 hrs during a station blackout.
The second part is correct. If it is necessary to minimize drain on the C-Train batteries during emergency conditions, AFW flow to the SGs is controlled by adjusting TDAFW pump speed in lieu of using the pump discharge throttle valves.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is incorrect. If it is necessary to minimize drain on the C-Train batteries during emergency conditions, AFW flow to the SGs is controlled by adjusting TDAFW pump speed in lieu of using the pump discharge throttle valves. However, the throttling of discharge valves is the normal means for adjusting flow. EOP 19100-C steps 21 and 24 even discuss closing the throttle valves "one at a time". Therefore, it is reasonable that a candidate with insufficient knowledge of the best method to conserve battery power would be reluctant to reduce TDAFW speed and would in turn adjust each throttle valve.
C. Incorrect. Plausible. The first part is incorrect. The 1E batteries are sized to have sufficient capacity to supply the required loads for 2.75 hours during a LOSP with a LOCA and for 4.0 hrs during a station blackout. However, a candidate with insufficient knowledge of battery capacities could combine the 2.75 hr capacity for LOSP with a LOCA, and the 4 hr station blackout capacity, not realizing the 2.75 hr capacity accounted for both events.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Thursday, February 27, 2014 10:16:37 AM                                                                2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            063A1.01 Importance Rating:              2.5 / 3.3 Technical
 
==Reference:==
EOP 19100-C, Rev 38.1, pages 15 & 17 SOP 13610-1, Rev 50.4, page 9 TS 3.8.4 Bases, Rev 2-3/05, page B 3.8.4-2 References provided:            None Learning Objective:              LO-LP-41201-01 Briefly describe the principle of operation of a lead acid wet cell. Include both charging and discharging.
LO-LP-39212-02 Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode:
: a. Whether any Tech Spec LCOs of section 3.8 are exceeded.
: b. The required actions for all section 3.8 LCOs.
LO-LP-37031-08 Using EOP 19100-C as a guide, briefly describe how each step is accomplished.
LO-PP-20101-09 Determine the impact to AFW system operation and the overall integrated plant operations to the following types of power supply failures:
: a. U/V condition on either AA02 or BA03 with the bus being re-energized from the EDG while at 100% power.
: b. U/V condition on either AA02 or BA03 with the bus remaining de-energized while at 100% power.
: c. Loss of a 120 VAC 1E vital instrument bus.
: d. Loss of a 125 VDC 1E bus
: e. Loss of All AC Power LO-TA-37018      Respond to a Loss of All AC Power per 19100-C Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:          41.8 / 41.10 Comments:
You have completed the test!
Thursday, February 27, 2014 10:16:38 AM                                                            3
 
DC Sources  Operating B 3.8.4 BASES BACKGROUND          Batteries are sized in accordance with IEEE 485 (Ref. 3) to have (continued)      sufficient capacity to supply the required loads for a loss of coolant/
loss of offsite power (LOCA/LOSP) duration of 2 3/4 hours and a station blackout (SBO) duration of 4 hours. For LOSP/LOCA, they are sized at a minimum temperature of 70&deg;F; their initial capacity was increased by 10% for load growth and 25% for aging. The required final (end of duty cycle and end of life) battery cell voltages for each load group have been analyzed to demonstrate that adequate voltage is provided to the loads. The battery voltage specifications are discussed in detail for each load group in FSAR, Chapter 8 (Ref. 4).
The battery cells are of flooded lead acid construction with a nominal specific gravity of 1.215. This specific gravity corresponds to an open circuit battery voltage of approximately 121.8 V for a 59 cell battery (i.e., cell voltage of 2.065 volts per cell (Vpc)). The open circuit voltage is the voltage maintained when there is no charging or discharging. Once fully charged with its open circuit voltage  2.065 Vpc, the battery cell will maintain its capacity for 30 days without further charging per manufacturer's instructions. Optimal long term performance however, is obtained by maintaining a float voltage 2.20 to 2.25 Vpc. This provides adequate over-potential, which limits the formation of lead sulfate and self discharge. The nominal float voltage of 2.23 Vpc corresponds to a total float voltage output of 131.6 V for a 59 cell battery as discussed in the FSAR, Chapter 8 (Ref. 4).
Each 125 VDC battery is provided with two battery chargers, each of which is sized to supply the continuous (long term) demand on its associated DC system while providing sufficient power to replace 110% of the equivalent ampere-hours removed from the battery during a design basis battery discharge cycle within a 12 hour period after charger input power is restored. Normally, both battery chargers are on line with load sharing circuitry to ensure that the DC load is properly shared between the two chargers. Only one charger is required OPERABLE to support the associated DC power system.
The sizing of each battery charger meets the requirements of IEEE 308 (Ref. 1) and Regulatory Guide 1.32 (Ref. 5).
The battery chargers are normally in the float-charge mode. Float-charge is the condition in which the charger is supplying the connected loads and the battery cells are receiving adequate current to optimally charge the battery. This assures the internal losses of a battery are overcome and the battery is maintained in a fully charged state.
(continued)
Vogtle Units 1 and 2                      B 3.8.4-2                                Rev. 2-3/05
 
ECA - 0.0 Loss of all AC Power                                19100-C VOGTLE            Version 38.1 Unit C        Page 15 of 50 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                    RESPONSE NOT OBTAINED
: 20. Check for faulted SG(s):                20. Go to Step 23.
ANY SG PRESSURE LOWERING IN AN UNCONTROLLED MANNER.
              -OR-ANY SG COMPLETELY DEPRESSURIZED.
: 21. Isolate faulted SG(s):                  21.
: a. Close the TDAFW Throttle Valves on affected SG(s) one at a time:
HV-5122 (SG 1)
HV-5125 (SG 2)
HV-5127 (SG 3)
HV-5120 (SG 4)
: b. Close only one TDAFW Pump Steam Supply Valve from affected SG(s):
HV-3009 (SG 1)
                      -OR-HV-3019 (SG 2)
: c. Verify affected SG ARV(s) closed:
PV-3000 (SG 1)
PV-3010 (SG 2)
PV-3020 (SG 3)
PV-3030 (SG 4)
Printed February 17, 2014 at 12:51
 
ECA - 0.0 Loss of all AC Power                                  19100-C VOGTLE            Version 38.1 Unit C          Page 17 of 50 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                      RESPONSE NOT OBTAINED
      *24. Isolate ruptured SG(s):                    24.
: a. Close the TDAFW Throttle Valves on affected SG(s) one at a time:
HV-5122 (SG 1)
HV-5125 (SG 2)
HV-5127 (SG 3)
HV-5120 (SG 4)
: b. Close only one TDAFW Steam Supply Valve from affected SG(s):
HV-3009 (SG 1)
                      -OR-HV-3019 (SG 2)
: c. Limit ruptured SG ARV(s) operation:
Adjust controller setpoint to 1160 psig (pot setting 7.73).
WHEN SG pressure is less than 1160 psig, THEN verify SG ARV(s) closed.
IF SG ARV(s) can NOT be closed, THEN locally isolate SG ARV(s).
: d. Locally close the MDAFW Throttle Valves on affected SG(s) using ATTACHMENT 6.
: e. Verify ruptured SG(s) remains isolated during subsequent recovery actions unless needed for RCS cooldown.
Printed February 17, 2014 at 12:51
 
Approved By                                                                            Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                      13610-1      50.4 Effective Date                                                                          Page Number 6/21/13                            AUXILIARY FEEDWATER SYSTEM                                9 of 109 INITIALS 2.1.6            Manual reset of the 186 Lockout Relay is required on the MDAFW Pump Motor Feeder Breakers on phase or ground overcurrent.
Also, the System Status Monitor Panel will indicate BYPASSED until the lockout relay is reset.                                      ________
2.1.7            TDAFW and MDAFW Pumps are normally aligned to CST-1. If CST-1 is rendered inoperable, CST supply and mini-flow valves should be aligned to CST-2. (SNC12990,1987311237)                      ________
2.1.8            During emergency conditions if it is necessary to minimize drain on the C-Train Batteries, TDAFW Pump speed may be adjusted in lieu of, or in addition to, using the pump discharge throttle valves to control AFW flow to the SGs. In this case, TDAFW Pump discharge pressure should be adjusted to ensure the required feed flow is supplied to the SG with the highest pressure.            ________
2.1.9            Thirty (30) seconds after TDAFW Pump Steam Supply 1-HV-5106 begins to open, turbine speed less than 175 rpm will cause the governor controller to go into TRIP MODE. This will cause ALB16-E03 AFW TURB TROUBLE to illuminate, and the controller will not provide startup functions. 1-HV-5106 must be closed and then the Trip and Throttle Valve must be cycled open to reset the controller.                                              ________
2.1.10            Handswitches and indicators mentioned in this procedure are located on Panel QMCB unless otherwise stated.                        ________
2.1.11            It is not desirable to remove the TDAFW pump from service on either unit when work is in progress in the Low Voltage Switchyard or on any Reserve Auxiliary Transformer.                              ________
2.1.12            MDAFW Pump mini-flow valve operation should be periodically monitored when operating throttle valves.                              ________
Printed November 6, 2013 at 10:50
: 1. 064G2.1.31 001/LOIT AND LOCT/RO/M/F 4.6/4.3/064G2.1.31/LO-TA-11015///
Given the following:
            - Unit 1 is at 100% reactor power.
            - DG1A is in its normal standby alignment.
Which one of the following completes the following statement?
The DSL GEN 1A UNIT/PARALLEL switch, 1HS-4414B, on the QEAB is in the __(1)__
mode of operation, and its associated blue indicating light is __(2)__.
__(1)__                                      __(2)__
A.            unit                                          lit B.            unit                                        NOT lit C.          parallel                                        lit D.          parallel                                      NOT lit K/A 064              Emergency Diesel Generator G2.1.31          Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.
K/A MATCH ANALYSIS The questions tests the candidate's ability to locate the control room switch and light indication for the Emergency D/G unit selector switch and determine the correct standby position and light indication.
EXPLANATION OF REQUIRED KNOWLEDGE Per Control Room Rounds 11874-1, the Emergency D/G mode selector switch must be in the UNIT position during standby alignment. Per SOP 13145A-1 step 4.3.1.1, during shutdown of the D/G, the mode selector switch is momentarily placed in the UNIT position and the blue indicator light is verified LIT.
ANSWER / DISTRACTOR ANALYSIS A. Correct.                  The first part is correct. The Emergency D/G mode selector switch must be in the UNIT position during standby alignment.
Thursday, February 27, 2014 10:18:35 AM                                                          1
 
The second part is correct. With the mode selector switch momentarily placed in the UNIT position, the blue indicator light is LIT.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. With the mode selector switch momentarily placed in the UNIT position, the blue indicator light is LIT. However, candidates frequently confuse the two DG modes and which one the blue light LIT represents.
Additionally, the original Vogtle design had the blue light LIT for parallel mode. It was subsequently changed to LIT in unit mode due to difficulty in verifying standby alignment versus a blown bulb.
C. Incorrect. Plausible. The first part is incorrect. The Emergency D/G mode selector switch must be in the UNIT position during standby alignment.
However, candidates frequently confuse the two DG modes.
Additionally, an LOSP signal places the DG in unit mode. The candidate may believe that the normal alignment would be parallel, since that is what mode the DG is most frequently operated in during survelliances and that it will shift to unit mode during and LOSP.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Thursday, February 27, 2014 10:18:35 AM                                                                2
 
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            064G2.1.31 Importance Rating:              4.6 / 4.3 Technical
 
==Reference:==
SOP 13145A-1, Rev 6.1, page 32 LINEUP 11874-1, Rev 77.0, page 6 References provided:            None Learning Objective:              LO-TA-01022    Transfer From RAT to SAT per 13418-C LO-TA-11007    Parallel Normal Incoming Source to 4160V Bus Being Supplied from Diesel Generator using 13427A/B-1/2 LO-TA-11014A    Perform emergency Diesel Generator operability test using 14980A/B-1/2 LO-TA-11015    Prepare Diesel Generator for Automatic Operation using 13145-1/2 LO-PP-11101-49 For the following controls on the Generator Control Panel and/or the QEAB, describe the response of the diesel generator to the selection of each position:
: a. Local/Remote switch
: b. Speed RAISE/LOWER switch (pushbutton)
: c. Exciter Enable pushbutton
: d. Emergency Shutdown pushbutton
: e. Delete
: f. Field Flash pushbutton
: g. Voltage Control RAISE/LOWER switch (pushbutton)
: h. Unit/Parallel switch
: i. Exciter Permissive Switch
: j. Load Pot Question origin:                NEW Cognitive Level:                M/F 10 CFR Part 55 Content:          41.10 / 45.12 Comments:
You have completed the test!
Thursday, February 27, 2014 10:18:35 AM                                                          3
 
Approved By                                                                            Procedure    Version C.H. Williams                      Vogtle Electric Generating Plant                    13145A-1      6.1 Effective Date                                                                          Page Number 12/05/2013                            DIESEL GENERATOR TRAIN A                              32 of 86 INITIALS 4.3                SHUTDOWN 4.3.1              Stopping Train A Diesel Generator 4.3.1.1            IF stopping DG1A from the QEAB, perform the following:
: a.        At the Generator Control Panel, verify LOCAL-REMOTE Switch 1HS-4516 to REMOTE.                                  ________
: b.        Momentarily place DSL GEN 1A UNIT/PARALLEL switch 1HS-4414B to UNIT and check the blue DSL GEN 1A UNIT MODE FAST START light is lit.                          ________
NOTE The DG must idle for 30 seconds after the UNIT/PARALLEL switch is placed in UNIT to verify that the Governor Slow Start timer can time out and thus permit the DG to Fast Start after shutdown.
ALB35 F06 DG1A SWITCH NOT IN AUTO
: c.        At MCC 1NBI place Lube Oil Circulating Pump handswitch 1HS-24006 in the OFF position.                              ________
: d.        WHEN the DG has idled greater than 30 seconds, depress STOP pushbutton 1HS-4571B.                                  ________
: e.        At PDG1, check generator voltage drops to zero on FIELD VOLTMETER 1EI-40139.                                        ________
: f.        At Generator Control Panel PDG1, check EXCITER PERMISSIVE switch 1HS-4913 is in the NORMAL position.      ________
Printed February 17, 2014 at 14:24
 
Approved By                                                                                                  Procedure    Version R.M. Brown                                Vogtle Electric Generating Plant                                    11874-1      77 Effective Date                                                                                              Page Number 10/22/2013                                        CONTROL ROOM ROUNDS SHEETS                                        6 of 20 Sheet 3 of 5 FIGURE 1 UNIT 1 EMERGENCY SAFEGUARDS EQUIPMENT CHECKLIST (APPLICABILITY - MODES 1 THROUGH 4)
S H I F  T  COMMENT COMPONENT                                NUMBER              POSITION    STATUS        DAY      NIGHT NUMBER RWST TO CCP A & B SUCTION                1-HS-0112D          AUTO        CLOSED RWST TO CCP A & B SUCTION                1-HS-0112E          AUTO        CLOSED VCT OUTLET ISOLATION                    1-HS-0112B          AUTO        OPEN VCT OUTLET ISOLATION                    1-HS-0112C          AUTO        OPEN BIT DISCH ISOLATION                      1-HS-8801A          AUTO        CLOSED BIT DISCH ISOLATION                      1-HS-8801B          AUTO        CLOSED NUCLEAR RECORDER                        1-NR-0045 (30)      N/A          NO ALARMS TDAFW                                    1-HS-5106A          AUTO (3)    OPEN /
CLOSED MDAFW-B                                  1-HS-5130A          AUTO (3)    ON-OFF MDAFW-A                                  1-HS-5131A          AUTO (3)    ON-OFF LP-1 MS SPLY TO AUX FW TD PMP-1                                    1-HS-3009          N/A          OPEN (3)
LP-2 MS SPLY TO AUX FW TD PMP-1                                    1-HS-3019          N/A          OPEN (3)
SG-1 FROM TDAFW                          1-HS-5122A          AUTO        OPEN (3)
SG-2 FROM TDAFW                          1-HS-5125A          AUTO        OPEN (3)
SG-3 FROM TDAFW                          1-HS-5127A          AUTO        OPEN (3)
SG-4 FROM TDAFW                          1-HS-5120A          AUTO        OPEN (3)
SG-2 FROM MDAFW PMP-B                    1-HS-5132A          AUTO        OPEN (3,4)
SG-3 FROM MDAFW PMP-B                    1-HS-5134A          AUTO        OPEN (3,4)
SG-1 FROM MDAFW PMP-A                    1-HS-5139A          AUTO        OPEN (3,4)
SG-4 FROM MDAFW PMP-A                    1-HS-5137A          AUTO        OPEN (3,4)
AFP DIFF PRESS                          1-PDIC-5180A        IN SPEED    OPERABLE CONTROL AT 100% (3)
DG 1A UNIT/PARALLEL                      1-HS-4414B          UNIT        N/A DG 1B UNIT/PARALLEL                      1-HS-4452B          UNIT        N/A POWER QHVC                                    ALL ZLBs          N/A          AVAILABLE (5)
RX CAVITY CONCRETE TEMP                  1-TJI-12270 (30)    N/A          NO ALARMS INDICATOR CTB CLG UNIT FAN-1 LOW SPEED            1-HS-12582A        AUTO        ON-OFF CTB CLG UNIT FAN-1 HIGH SPEED            1-HS-12582D        AUTO        ON-OFF CTB CLG UNIT FAN-2 LOW SPEED            1-HS-2582A          AUTO        ON-OFF CTB CLG UNIT FAN-2 HIGH SPEED            1-HS-2582D          AUTO        ON-OFF CTB CLG UNIT FAN-3 LOW SPEED            1-HS-12583A        AUTO        ON-OFF CTB CLG UNIT FAN-3 HIGH SPEED            1-HS-12583D        AUTO        ON-OFF (3) NOT REQUIRED IN MODE 4 (4) POSITION MAY BE THROTTLED WHEN RUNNING ASSOCIATED AFW PUMP (5) ZLBs 18, 19, 36, 37, 38, 43 AND 44 ARE NORMALLY DE-ENERGIZED AND THUS ARE EXCLUDED.
(30) IF RECORDER DISPLAYS NOT ENOUGH FREE SPACE ON MEDIA MEDIA CARD SHOULD BE CHANGED PER 10001-C SECTION 5.2.
COMMENTS:
Printed February 17, 2014 at 14:22
: 1. 065AA2.08 001/LOIT/RO/C/A 2.9/3.3/065AA2.08/LO-TA-60007///
Initial condition:
            - Unit 1 is at 7% reactor power.
Current condition:
            - Instrument air line to 1FV-0121, CVCS Charging Flow Control Valve, is severed.
Which one of the following completes the following statement?
With no operator action, seal injection flow to the RCPs will __(1)__,
and a reactor trip __(2)__ occur on pressurizer level.
__(1)__                                __(2)__
A.          increase                                    will B.          increase                                will NOT C.        decrease                                      will D.        decrease                                  will NOT K/A 065              Loss of Instrument Air AA2.08          Ability to determine and interpret the following as they apply to the Loss of Instrument Air:
                        - Failure modes of air-operated equipment.
K/A MATCH ANALYSIS The question tests the candidate's ability to identify the failure mode of AOV FV-0121 due to a loss of instrument air. The candidate will demonstrate the ability to determine the integrated system response by predicting how seal injection flow will react, and whether a reactor trip will occur due to pressurizer level response.
EXPLANATION OF REQUIRED KNOWLEDGE Per P&ID 1X4DB116-1, FV-0121 is a fail-open AOV. The question establishes a scenario where the instrument air line to FV-0121 fails, resulting in maximum charging flow. Backpressure valve HV-182 is down stream of FV-121 and is a manually controlled AOV. As total charging flow increases without HV-182 position changing, Thursday, March 06, 2014 11:36:51 AM                                                          1
 
more flow will be directed to the RCP seals. Additionally, an increase in flow through HV-182 to the normal charing nozzle will also occur. The net result is a continuous rise in Pressurizer Level. With reactor power < P-7 (2/4 Power range NI's >10%), a reactor trip will not occur when pressurizer level is >92%.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. When instrument air is lost to FV-0121 the valve will fail open. This action will increase the total charging flow and with HV-0182 throttled to a set position it would force more flow into the RCP seal injection line.
The second part is incorrect. With reactor power < P-7 (2 of 4 Power range NI's >10%), a reactor trip will not occur when pressurizer level is >92%. However, a candidate may not recognize, or may have forgotten, the P-7 interlock and assume the reactor trips on pressurizer level >92%.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. With reactor power < P-7 (2 of 4 Power range NI's >10%), a reactor trip will not occur when pressurizer level is >92%.
C. Incorrect. Plausible. The first part is incorrect. When instrument air is lost to FV-0121 the valve will fail open. This action will increase the total charging flow and with HV-0182 throttled to a set position it will force more flow to the RCP seal injection line. However, if the candidate assumes that HV-182 also loses instrument air pressure, then HV-182 would fail open and seal injection flow would lower. The candidate could also reverse the order of FV-0121 and HV-0182 in the flow path and conclude that seal injection flow would lower.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, March 06, 2014 11:36:51 AM                                                                  2
 
Level:                        RO Tier # / Group #              T1 / G1 K/A#                          065AA2.08 Importance Rating:            2.9 / 3.3 Technical
 
==Reference:==
SOP 13503A-1, Rev 7.2, pages 33 & 35 P&ID 1X4DB116-1, Rev 50.0 References provided:          None Learning Objective:            LO-PP-09100-07 Given that a partial or complete loss of instrument air has occurred, determine how CVCS letdown system will respond and describe the steps to control RCS inventory.
LO-PP-09200-06 Given that a partial or complete loss of instrument air has occurred, determine how the CVCS charging system will respond and describe the steps that are required to control RCS inventory and seal injection.
LO-TA-60007    Respond to a Loss of Instrument Air per 18028-C.
Question origin:              NEW Cognitive Level:              C/A 10 CFR Part 55 Content:        43.5 / 45.13 Comments:
You have completed the test!
Thursday, March 06, 2014 11:36:51 AM                                                          3
 
Approved By                                                                                                    Procedure    Version J.B. Stanley                          Vogtle Electric Generating Plant                                        13503A-1 7.2 Effective Date                TRAIN A REACTOR CONTROL SOLID-STATE PROTECTION                                  Page Number 6/21/13                                                      SYSTEM                                                  33 of 38 ATTACHMENT C                                        Sheet 1 of 6 PERMISSIVES, CONTROL INTERLOCKS, REACTOR TRIPS AND ESF ACTUATIONS PERMISSIVES Permissive                Setpoint/Coincidence              Function P-4                        Train related Rx trip &          Trips Main Turbine Bypass breaker open                        Train A - mechanical Train B - electrical FWI if Lo Tavg (2/4  564 F) present Seals in FWI if caused by SI or P-14 (Hi Hi Level)
Must be present to block auto SI after SI reset.
P-4 Train A arms Steam Dumps P-4 Train B swaps Steam Dumps to plant trip controller P-6                        1/2 IR Detectors  2.0 E -5      Allows manual block of SR High  trip
                                % Rx Power P-7                        P-10 (2/4 PR NIs  10%            Unblocks "At Power" Trips Rx power) or                              Przr Low Pressure P-13 (PT-505 or 506                      Przr High Level 10% turbine power)                        RCS Two Loop Low Flow RCP UF RCP UV P-8                        2/4 PR NIS  48% Rx power        Enables Single Loop Low Flow Rx Trip P-9                        2/4 PR NIS  40% Rx power        Enables Turbine trip Rx trip P-10                      2/4 PR NIS  10% Rx power        Auto block of SR High  trip Enables P-7 Allows manual block of IR rod stop and Hi  trip Allows manual block of PR Hi  trip Lo Setpoint P-11                      2/3 Przr Pressure channels        Auto enables Lo Przr Press SI & Lo Steamline Press 2000 psig                      SI/SLI & sends signal to open Accum Isolation Valves when P-11 resets. P-11 allows operator to block PRZR & Steamline low pressure SI & SLI. Also activates "Not Full Open" annunciators for Accumulator MOVs (ALB16; A5, B5, C5 &
D5) and HV-8806; (ALB16 E03).
P-12                      2/4 NR Tavg  550 F              Interlocks Steam Dumps closed ( Cooldown Dump Valves PV-507A, B & C) may be reopened by use of Bypass Interlock switches)
P-13                      1/2 Turbine Impulse channels    Enables P-7 10%
P-14                      2/4 NR SG Level channels  82%    Actuates FWI Actuates MFP and Main Turbine trip Printed February 17, 2014 at 15:40
 
Approved By                                                                                                      Procedure    Version J.B. Stanley                          Vogtle Electric Generating Plant                                          13503A-1 7.2 Effective Date                TRAIN A REACTOR CONTROL SOLID-STATE PROTECTION                                    Page Number 6/21/13                                                        SYSTEM                                                  35 of 38 ATTACHMENT C                                          Sheet 3 of 6 PERMISSIVES, CONTROL INTERLOCKS, REACTOR TRIPS AND ESF ACTUATIONS REACTOR TRIPS Reactor Trip        Setpoint/Permissive/Blocks                Bases Manual                1/2 handswitches on QMCB                Ensures capability of operator to manually trip Rx SR Hi                1/2 SR Channels  105 cps may be        Protection against uncontrolled rod withdrawal from manually blocked above P-6 and auto      subcritical conditions blocked by P-10 IR Hi                1/2 IR Channels  25% Power, may be      Protection against uncontrolled rod withdrawal from manually blocked above P-10              subcritical conditions PR Hi  Low          2/4 PR Channels  25% power, may        Protection against uncontrolled rod withdrawal Setpoint              be manually blocked above P-10          from subcritical or low power conditions PR Hi  High          2/4 PR Channels  109% power            Protection against power excursions from all power Setpoint                                                      levels PR Positive          2/4 PR Channels  +5% in 2 seconds      Protection against rod ejection accident Rate OTT                  2/4 T Channels  T Setpoint            Protection against DNB OPT                  2/4 T Channels  T Setpoint            Protection against exceeding kW/Ft limitations Przr Hi Press        2/4 Przr Press Channels  2385 psig      Ensures RCS integrity (prevents overpressurization)
Przr Lo Press        2/4 Przr Press Channels  1960 psig      Protection against DNB (rate compensated) auto blocked below P-7 Przr Hi Level        2/3 Przr Level Channels  92%            Prevents water relief through Przr Code Safety Valves auto blocked below P-7 Single Loop          2/3 RCS Flow Channels  90% on 1/4      DNB protection in the event of loss of flow condition Low Flow              loops auto blocked below P-8 Two Loop              2/3 RCS Flow Channels  90% on 2/4      DNB protection in the event of loss of flow condition Low Flow              loops auto blocked below P-7 RCP UV                1/2 13.8 busses  70% nominal voltage    DNB protection in the event of a loss of forced flow auto blocked below P-7 RCP UF                1/2 13.8 busses  57.3 Hz auto blocked  DNB protection in the event of a loss of forced flow below P-7, will also trip all RCPs Turbine Trip /        4/4 Turbine SVs not full open or 2/3    Ensures Rx trip upon load reduction outside of design Rx Trip              ETS pressure switches  580 psig auto    limits (that which can be handled by steam dumps &
blocked below P-9                        rod control)
SG Lo Lo Level        2/4 NR SG Level Channels  38%          Protects against loss of heat sink with sufficient water level to allow for starting delays of AFW Safety Injection      Any SI, auto or manual                  Assures subcriticality during accident conditions General              GW condition on 2/2 SSPS Trains          Protection against potential loss of auto trip capability Warning Printed February 17, 2014 at 15:40
 
Seal Injection valve HV-0182 position does not change, causing more flow to the charging nozzle and seal inj FV-0121 fails open due to loss of instrument air
: 1. 069AA2.02 001/LOIT AND LOCT/RO/M/F 3.9/4.4/069AA2.02/LO-LP-39210-01///
Initial condition:
            - Unit 1 is at 100% reactor power.
Current conditions:
            - Containment entry is in progress.
            - Entry crew reports the inner air lock door seal is damaged.
            - The inner air lock door is declared inoperable.
Which one of the following completes the following statement?
To comply with the required action statement of Tech Spec 3.6.2, "Containment Air Locks," __(1)__ air lock door(s) must be verified closed, and the Tech Spec 3.6.2 required action must be completed __(2)__.
__(1)__                                __(2)__
A.            both                                immediately B.            both                              within 1 hour C.      only the outer                          immediately D.      only the outer                          within 1 hour K/A 069              Loss of Containment Integrity AA2.02          Ability to determine and interpret the following as they apply to the Loss of Containment Integrity:
                        - Verification of automatic and manual means of restoring integrity.
K/A MATCH ANALYSIS The question tests the candidates ability to determine the status of the containment air lock based on reported damage, and the ability to restore integrity by manual means as directed by Tech Spec 3.6.2.
EXPLANATION OF REQUIRED KNOWLEDGE Thursday, March 06, 2014 11:37:56 AM                                                          1
 
Per TS 3.6.2 Cond A, if one air lock door is inoperable, the OPERABLE door must be verified CLOSED within 1 hour and locked CLOSED within 24 hours. The candidate must determine that ONLY the outer door is required to be closed to comply with this RAS.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per TS 3.6.2 Cond A, the outer door must be verified closed within 1 hr. However, the normal alignment would have both doors closed. It is reasonable to assume the candidate may conclude the correct action is to shut both doors even if one is not OPERABLE. The candidate may also consider the situation where the inoperable door is the inner door and in order to perform maintenance on the door the outer must be opened, and that both the doors would need to be shut to maintain integrity.
The second part is incorrect. Per TS 3.6.2 Cond A, the outer door must be verified closed within 1 hr. However, per TS 3.9.4, if the same type of condition were discovered with the plant in lower MODES of operation, the action time would be immediate. Additonally, TS 3.6.2 Cond C for two airlock doors in the same airlock has an action time of immediate.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. One Containment airlock door inoperable per Tech Spec 3.6.2, Containment Air Locks, Condition A.1 requires the OPERABLE door be shut within 1 hour.
C. Incorrect. Plausible. The first part is correct. One Containment airlock door inoperable per Tech Spec 3.6.2, Containment Air Locks, Condition A.1 requires the OPERABLE door only to be shut, which is the outer door.
The second part is incorrect. See the second part of choice A above.
D. Correct                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, March 06, 2014 11:37:56 AM                                                                  2
 
Level:                        RO Tier # / Group #              T1 / G2 K/A#                          069AA2.02 Importance Rating:            3.9 / 4.4 Technical
 
==Reference:==
TS 3.6.2, Amendment No. 96, page 3.6.2-1 TS 3.9.4, Amendment No. 115, page 3.9.4-1 References provided:          None Learning Objective:            LO-LP-61209-13 State when containment integrity is required.
LO-LP-39210-01 For any given item in section 3.6 of Tech Specs, be able to:
: a. State the LCO.
: b. State any one hour or less required actions.
Question origin:              BANK Cognitive Level:              M/F 10 CFR Part 55 Content:        43.9 / 45.13 Comments:
You have completed the test!
Thursday, March 06, 2014 11:37:56 AM                                                          3
 
Containment Air Locks 3.6.2 3.6 CONTAINMENT SYSTEMS 3.6.2 Containment Air Locks LCO 3.6.2                  Two containment air locks shall be OPERABLE.
APPLICABILITY:              MODES 1, 2, 3, and 4.
ACTIONS
-----------------------------------------------------------NOTES----------------------------------------------------------
: 1. Entry and exit are permissible to perform repairs on the affected air lock components.
: 2. Separate Condition entry is allowed for each air lock.
: 3. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when air lock leakage results in exceeding the overall containment leakage rate.
-------------------------------------------------------------------------------------------------------------------------------
CONDITION                                REQUIRED ACTION                          COMPLETION TIME A. One or more                          -----------------NOTES--------------------
containment air locks                1. Required Actions A.1, A.2, and with one containment air                    A.3 are not applicable if both lock door inoperable.                      doors in the same air lock are inoperable and Condition C is entered.
: 2. Entry and exit are permissible for 7 days under administrative controls if both air locks are inoperable.
                                            ------------------------------------------------
(continued)
Vogtle Units 1 and 2                                      3.6.2-1                        Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
 
Containment Air Locks 3.6.2 ACTIONS CONDITION      REQUIRED ACTION                      COMPLETION TIME A. (continued)      A.1  Verify the OPERABLE                1 hour door is closed in the affected air lock.
AND A.2  Lock the OPERABLE                  24 hours door closed in the affected air lock.
AND A.3  ------------NOTE--------------
Air lock doors in high radiation areas may be verified locked closed by administrative means.
                          ----------------------------------
Verify the OPERABLE                Once per 31 days door is locked closed in the affected air lock.
(continued)
Vogtle Units 1 and 2      3.6.2-2                        Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
 
Containment Penetrations 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Containment Penetrations LCO 3.9.4            The containment penetrations shall be in the following status:
: a. The equipment hatch is capable of being closed and held in place by four bolts;
: b. The emergency and personnel air locks are isolated by at least one air lock door, or if open, the emergency and personnel air locks are isolable by at least one air lock door with a designated individual available to close the open air lock door(s); and
: c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
: 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or
: 2. capable of being closed by at least two OPERABLE Containment Ventilation Isolation valves APPLICABILITY:      During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.
ACTIONS CONDITION                        REQUIRED ACTION                COMPLETION TIME A. One or more containment        A.1      Suspend CORE                Immediately penetrations not in                      ALTERATIONS.
required status.
AND A.2      Suspend movement of        Immediately irradiated fuel assemblies within containment.
Vogtle Units 1 and 2                      3.9.4-1                    Amendment No. 115 (Unit 1)
Amendment No. 93 (Unit 2)
 
Containment Penetrations 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Containment Penetrations LCO 3.9.4            The containment penetrations shall be in the following status:
: a. The equipment hatch is capable of being closed and held in place by four bolts;
: b. The emergency and personnel air locks are isolated by at least one air lock door, or if open, the emergency and personnel air locks are isolable by at least one air lock door with a designated individual available to close the open air lock door(s); and
: c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
: 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or
: 2. capable of being closed by at least two OPERABLE Containment Ventilation Isolation valves APPLICABILITY:      During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.
ACTIONS CONDITION                        REQUIRED ACTION                COMPLETION TIME A. One or more containment        A.1      Suspend CORE                Immediately penetrations not in                      ALTERATIONS.
required status.
AND A.2      Suspend movement of        Immediately irradiated fuel assemblies within containment.
Vogtle Units 1 and 2                      3.9.4-1                    Amendment No. 115 (Unit 1)
Amendment No. 93 (Unit 2)
: 1. 071A3.03 001/LOIT AND LOCT/RO/M/F 3.6/3.8/071A3.03/LO-TA-32007//HL-15 AUDIT/
Initial condition:
            - Unit 1 Waste Gas Decay Tank release is in progress.
Current condition:
            - ARE-0014, Waste Gas Processing Effluent Monitor, is in Intermediate alarm.
Which one of the following completes the following statement?
ARV-0014, Plant Vent Radwaste Gas, isolation valve is __(1)__,
and the position of ARV-0014 is to be verified __(2)__.
A. (1) open (2) on the IPC in the Control Room B. (1) open (2) locally on the Gaseous Waste Panel C. (1) closed (2) on the IPC in the Control Room D. (1) closed (2) locally on the Gaseous Waste Panel K/A 071              Waste Gas Disposal System (WGDS):
A3.03            Ability to monitor automatic operation of the Waste Gas Disposal System, including:
                        - Radiation monitoring system alarm and actuating signals.
K/A MATCH ANALYSIS The question tests the candidates ability to monitor the Waste Gas Disposal System release path by determining if an actuation signal has been generated based on the effluent rad monitor system alarm.
EXPLANATION OF REQUIRED KNOWLEDGE Thursday, March 06, 2014 11:38:56 AM                                                      1
 
Per SOP 13202-1 step 4.2.2 and ARP 17216-1, a HIGH radiation alarm on ARE-014 produces an actuation signal, which closes ARV-0014 and terminates the release.
Since ARE-0014 is in INTERMEDIATE alarm, ARV-0014 should remain open.
ARV-0014 position is not available in the main control room. ARV-0014 is monitored and controlled only from the local Waste Gas Panel on 'D' Level of the Aux Bldg. (Ref P&ID 1X4DB129)
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. ARV-0014 will remain open following the intermediate radiation alarm,but will automatically isolate on a high alarm condition.
The second part is incorrect. ARV-0014 is monitored and controlled only from the local Waste Gas Panel on 'D' Level of the Aux Bldg. However, many important valves and some rad monitor flow transmitters can be monitored from the IPC.
ARV-0014 and AFT-0014 are not among these.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. ARV-0014 is monitored and controlled only from the local Waste Gas Panel on 'D' Level of the Aux Bldg.
C. Incorrect. Plausible. The first part is incorrect. ARV-0014 will remain open following the intermediate radiation alarm, but will automatically isolate on a high alarm condition. However, the candidate may believe that any alarm, intermediate or high, would result in an isolation signal. Additionally, some procedures associated with releases will direct the operator to isolate the flowpath if an intermediate alarm is recieved. Containment purge is an example of this -
reference SOP 13125-1 precaution 2.1.7.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, March 06, 2014 11:38:56 AM                                                                  2
 
Level:                        RO Tier # / Group #              T2 / G2 K/A#                          071A3.03 Importance Rating:            3.6 / 3.8 Technical
 
==Reference:==
ARP 17216-1, Rev 2.3, pages 18 & 19 SOP 13125-1, Rev 54.0, pages 4 SOP 13202-1, Rev 20.1, pages 8-14 P&ID 1X4DB129, Rev 43.0 References provided:          None Learning Objective:            LO-PP-32101-09 Describe those automatic actions that occur for each of the following non-safety monitors when its high alarm setpoint is exceeded:
: b. Waste Gas Processing System Effluent (ARE-0014)
LO-TA-32007        Verify Proper Automatic actions to high radiation alarms Question origin:              BANK Cognitive Level:              M/F 10 CFR Part 55 Content:        41.7 / 41.13 / 45.5 Comments:
You have completed the test!
Thursday, March 06, 2014 11:38:56 AM                                                              3
 
Approved By                                                                              Procedure  Version M.G. Brill                        Vogtle Electric Generating Plant                        13125-1      54 Effective Date                                                                            Page Number 04/09/2013                                CONTAINMENT PURGE SYSTEM                                4 of 37 NOTE A containment release permit is considered a batch release. Batch pre-release permits are calculated based upon the radionuclide mix in the sample, the estimated volume of the release and the maximum release rate of the release. These pre-release bounding parameters ensure compliance with the ODCM. Releases therefore must remain within these bounding parameters for the pre-release calculations to be valid. Stopping a containment release before the Release may not continue beyond (Date/Time) indicated on 36022-C Data Sheet 1 ensures the volume to be released does not exceed the pre-release permit volume.
2.1.3              If a Purge and/or Vent is started within the allotted time limit, the release/vent may continue until the Release may not continue beyond (Date/Time) indicated on 36022-C Data Sheet 1.                            ________
2.1.4              Containment purges may be stopped and subsequently restarted without any sampling and analysis being performed and without Effluent Permit closure following review and approval of chemistry department personnel. Releases may not continue beyond the Release may not continue beyond (Date/Time) indicated on 36022-C Data Sheet 1.                                                              ________
2.1.5              To ensure accurate sampling by 1RE-2565, prior to placing CTB Mini-Purge in service, sampling flow valves 1-1609-U4-052 should be verified open and 1-1609-U4-054 closed.                                    ________
2.1.6              To ensure accurate sampling by 1RE-2565, prior to placing CTB Main Purge in service, sampling flow valves 1-1609-U4-054 should be verified open and 1-1609-U4-052 closed.                                    ________
2.1.7              If a valid Alert is received on 1RE-2565C while a Containment Purge or Pressure Relief is ongoing, the release/vent must be terminated immediately.                                                              ________
2.1.8              If radiation monitor 1RE-2565 turns magenta color on the Coms Console, it may be inoperable. Reference Precaution 2.1.9.
IF CVI operability is required when this occurs and 1RE-002 AND 1RE-003 are not BOTH operable, immediately terminate any release or vent in progress, verify Containment Dampers closed and notify chemistry.                                                                ________
Printed February 28, 2014 at 9:56
 
Approved By                                                                                    Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                            13202-1      20.1 Effective Date                                                                                  Page Number 11/08/2012                                  GASEOUS RELEASES                                            8 of 36 INITIALS 4.2                RELEASE CAUTION No part of the release should be performed until the approved Gaseous Effluent Permit is received in the Control Room.
4.2.1              Verify that NO additions have been made to the tank since the sample collection Date/Time listed on the release permit by checking the pressure of the GDT to be released, recorded on Data Sheet 1.                                                                ________
4.2.2              Verify Chemistry has performed a source AND channel check of Waste Gas Processing System Effluent Monitor A-RE-0014. IF A-RE-0014 is operable, perform a pulse check A-RE-0014 as follows:                                                                      ________
: a.        Notify the Control Room to expect an alarm from A-RE-0014 on the Digital Radiation Monitor System.                  ________
NOTE A-HS-0014 (trip indication) white light will remain illuminated until full open.
: b.        Open WASTE GAS DISCHARGE VALVE A-RV-0014 by performing the following:
(1)    Set A-HIC-0014 to 0% DEMAND.                                ________
(2)    Place A-HS-0014 in OPEN.                                    ________
(3)    Set A-HIC-0014 to 100%.                                      ________
(4)    Verify A-RV-0014 OPENS.                                      ________
: c.        Request Chemistry to activate the pulse test on channel A-RE-0014.                                                          ________
Printed February 28, 2014 at 9:51
 
Approved By                                                                            Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                    13202-1      20.1 Effective Date                                                                        Page Number 11/08/2012                                    GASEOUS RELEASES                                  9 of 36 INITIALS
: d.        Verify the following:
WASTE GAS DISCHARGE VALVE A-RV-0014 CLOSES.                                            ________
AND Hi Radiation alarm in Control Room annunciates. ________
: e.        Place A-HS-0014 in the OPEN position and check that A-RV-0014 remains in the closed position.                  ________
NOTE After performing a pulse check on A-RV-0014 it is normal that A-RV-0014 will not reopen until Chemistry restores channel A-RV-0014, this action proves the operability of A-RE-0014.
: f.        Notify Chemistry that A-RV-0014 has closed and will NOT reopen.                                                    ________
: g.        Request Chemistry to restore channel A-RE-0014 to normal.                                                    ________
4.2.3              IF operable, verify channel check of A-FT-0014 has been performed by Chemistry.                                              ________
Printed February 28, 2014 at 9:51
 
Approved By                                                                          Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                  13202-1      20.1 Effective Date                                                                      Page Number 11/08/2012                                GASEOUS RELEASES                                10 of 36 INITIALS 4.2.4              Place WASTE GAS DISCHARGE CONTROL VALVE A-HS-0014 in CLOSE (1-PGPP) (RD-56).                                        ________
4.2.5              Set WASTE GAS DISCHARGE CONTROL A-HIC-0014 to 0%
demand (1-PGPP) (RD-56).                                          ________
NOTE IF the Radioactive Gaseous Monitoring Instrument ARE-0014 OR Flow Transmitter AFT-0014 is determined to be inoperable PRIOR to start of the release, the release may be performed IF the requirements of ODCM Section 3.1.1 are met.
Critical 4.2.6              Perform the appropriate Checklist to align the Gaseous Waste Processing System for the tank to be released: (CV REQUIRED)
Gas Decay Tank #1-Checklist #1                          ________
Gas Decay Tank #2-Checklist #2                          ________
Gas Decay Tank #3-Checklist #3                          ________
Gas Decay Tank #4-Checklist #4                          ________
Gas Decay Tank #5-Checklist #5                          ________
Gas Decay Tank #6-Checklist #6                          ________
Gas Decay Tank #7-Checklist #7                          ________
Shutdown Gas Decay Tank #9-Checklist #8                  ________
Shutdown Gas Decay Tank #10-Checklist #9                ________
Printed February 28, 2014 at 9:51
 
Approved By                                                                            Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                    13202-1      20.1 Effective Date                                                                          Page Number 11/08/2012                                GASEOUS RELEASES                                    11 of 36 INITIALS Critical 4.2.7              Perform the appropriate Checklist to align the Gaseous Release Header for the tank to be released:
(CV REQUIRED)
Gas Decay Tank #1-Checklist #1                              ________
Gas Decay Tank #2-Checklist #2                              ________
Gas Decay Tank #3-Checklist #3                              ________
Gas Decay Tank #4-Checklist #4                              ________
Gas Decay Tank #5-Checklist #5                              ________
Gas Decay Tank #6-Checklist #6                              ________
Gas Decay Tank #7-Checklist #7                              ________
Shutdown Gas Decay Tank #9-Checklist #8                    ________
Shutdown Gas Decay Tank #10-Checklist #9                    ________
4.2.8              Verify all conditions of the Gaseous Effluent Permit that MUST be satisfied PRIOR to the release are met.                              ________
Critical 4.2.9              Verify the tank aligned for release is the same tank for which the Gaseous Effluent Permit was issued.                                  ________
________
CV 4.2.10            Note the maximum allowable release flow rate and A-RE-0014 setpoint given on the Gaseous Effluent Permit.                        ________
4.2.11            Notify the Unit 1 Control Room that the release is starting so they can monitor the flow and radiation data. (IPC points F6416 and R6253)                                                                ________
4.2.12            Place A-HS-0014 (1-PGPP) (RD-56) in OPEN.                            ________
Printed February 28, 2014 at 9:51
 
Approved By                                                                              Procedure    Version J.B. Stanley                    Vogtle Electric Generating Plant                      13202-1      20.1 Effective Date                                                                          Page Number 11/08/2012                                GASEOUS RELEASES                                      12 of 36 INITIALS CAUTION All Unit 1 and Unit 2 GDTs and SDTs should be monitored to verify that only the GDT or SDT being released is decreasing in pressure.
4.2.13            IF a GDT or SDT NOT being released decreases in pressure, immediately stop the release and notify the SS.                        ________
4.2.14            Continuously monitor all Gas Decay Tanks and Shutdown Decay Tanks pressures during the first hour of the release, AND THEN check all pressures hourly until the release is complete.
Document on Data Sheet 1.                                              ________
CAUTION Do not exceed the maximum allowable release rate or A-RE-0014 setpoint stated on the release permit. If at any time during the release, the allowable release rate or A-RE-0014 setpoint is exceeded, the release should be stopped and the SS notified.
4.2.15            IF AFT-0014 is OPERABLE, monitor A-RI-0014 and release flow rate while adjusting A-HIC-0014 to obtain the required release rate.                                                                  ________
Printed February 28, 2014 at 9:51
 
Approved By                                                                                    Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                          13202-1      20.1 Effective Date                                                                                Page Number 11/08/2012                                      GASEOUS RELEASES                                      13 of 36 INITIALS 4.2.16            IF AFT-0014 is INOPERABLE, the following steps must be performed to comply with the action step of the ODCM. Verify Flowrate by performing the following:
NOTE A-HS-0014 (trip indication) white light will remain illuminated until full open.
CAUTION The Release Permit flow rate should be immediately verified, IF the (trip indication) white light does not remain illuminated until HIC-0014 has fully opened.
: a.        Adjust A-HIC-0014 to 20% OPEN.                                    ________
: b.        Log the start time and initial pressure of the GDT being released in the ABO logbook.                                      ________
: c.        WHEN the initial pressure has decreased by 2 psig, perform the following to verify the initial flowrate is within the limits of the release permit:
(1)    Subtract the present pressure reading from the initial pressure reading then divide the result by 14.7.                                                      ________
(2)    Multiply the result of 4.2.16c(1) by 600.                  ________
(3)    Divide the result of 4.2.16c(2) by the number of minutes the release has been occurring. The resultant number is the flowrate in standard cubic feet per minute.                                            ________
: d.        Adjust A-HIC-0014 as needed to comply with the permit release rate.                                                      ________
: e.        Every 4 hours, verify the release rate by performing steps 4.2.16.c(1) thru 4.2.16.c(3).                                      ________
4.2.17            Record the start parameters on the Gaseous Effluent Permit.                  ________
Printed February 28, 2014 at 9:51
 
Approved By                                                                              Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                      13202-1      20.1 Effective Date                                                                          Page Number 11/08/2012                                  GASEOUS RELEASES                                    14 of 36 INITIALS 4.2.18            WHEN tank pressure falls below 20 psig, switch the associated pressure indicator on 1-PGPP (RD-56) to the LOW RANGE position.                                                              ________
4.2.19            WHEN tank pressure falls to 10 psig, or the required amount is released, terminate the release as follows:
: a.        Place A-HS-0014 (1-PGPP)(RD-56)in CLOSE.                    ________
________
IV
: b.        Set A-HIC-0014 (1-PGPP)(RD-56) to 0%.                        ________
________
IV
: c.        Record the stop time, date, AND final tank pressure on the Gaseous Effluent Permit.                                    ________
Printed February 28, 2014 at 9:51
 
Approved By                                                                      Procedure    Version T.E. Tynan                        Vogtle Electric Generating Plant              17216-1      2.3 Effective Date                ANNUNCIATOR RESPONSE PROCEDURE FOR ALB ON WASTE    Page Number 09/14/2012                                  PROCESSING SYSTEM-GAS PANEL                18 of 48 WINDOW A08 ORIGIN                              SETPOINT PLANT VENT A-RE-0014                            Variable        MONITOR HI RAD.
1.0              PROBABLE CAUSE
: 1.        High radiation level in gas being released.
: 2.        Radiation from safety valve release.
2.0              AUTOMATIC ACTIONS Plant Vent Radwaste Gas, A-RV-0014, closes.
Critical 3.0              INITIAL OPERATOR ACTIONS Verify A-RV-0014 is closed Printed January 24, 2014 at 8:47
 
Approved By                                                                                          Procedure    Version T.E. Tynan                          Vogtle Electric Generating Plant                                17216-1      2.3 Effective Date                ANNUNCIATOR RESPONSE PROCEDURE FOR ALB ON WASTE                        Page Number 09/14/2012                                  PROCESSING SYSTEM-GAS PANEL                                    19 of 48 WINDOW A08 (Continued) 4.0              SUBSEQUENT OPERATOR ACTIONS
: 1.        Determine the gas discharge line radiation using A-RI-0014 on Panel PGPP.
: 2.        Verify A-1902-U4-004 closed.                                          ________
________
IV
: 3.        Notify the Control Room of the alarm and system status prior to performing the remaining actions.
: 4.        IF planned release was in progress:
: a.      Request Health Physics to sample waste gas decay shutdown tank being released.
: b.      Check release conditions and rate.
: 5.        IF release is not planned:
: a.      Check waste gas decay shutdown tank pressure using A-PIS-1056 and A-PIS-1057 on Panel PGPP.
: b.      Check status of A-PSV-7884A and A-PSV-7884B.
: c.      IF relief valves have lifted at 100 psig, isolate the source.
: 6.        Refer to the requirements of the ODCM Manual.
: 7.        IF equipment failure is indicated, initiate maintenance as required.
5.0              COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE Printed January 24, 2014 at 8:47
 
AHS-0014 is located on panel PGPP. The first "P" denotes a field panel. A "Q" would denote the main control room.
: 1. 072K3.02 001/LOIT/RO/M/F 3.1/3.5/072K3.02/LO-TA-32007///072K1.03 Initial conditions:
            - Fuel movement is in progress in Unit 1 spent fuel pool.
            - ARE-2533A/B, Fuel Handling Building Effluent Radiogas Monitor, is OOS.
            - ARE-2532A/B, Fuel Handling Building Effluent Radiogas Monitor, is in service.
Current conditions:
            - 1RE-008, Fuel Handling Building Area Monitor, detector fails high.
            - Chemistry has removed 1RE-008 from service.
Which one of the following completes the following statement?
The personnel in the SFP area were alerted to the failed detector by a __(1)__,
and for movement of irradiated fuel in the spent fuel pool, __(2)__ is required to be in service.
__(1)__                                    __(2)__
A.            flashing light ONLY                            ARE-2532A/B B.            flashing light ONLY                                1RE-008 C. flashing light and audible alarm                      ARE-2532A/B D. flashing light and audible alarm                          1RE-008 K/A 072              Area Radiation Monitoring K3.02            Knowledge of the effect that a loss or malfunction of the ARM system will have on the following:
                        - Fuel handling operations.
K/A MATCH ANALYSIS The question tests the candidate's knowledge of how the personnel in the fuel handling building would be alerted to local area monitor RE-008 alarm and the impacts to fuel handling operations.
EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17100-1, if 1RE-0008 is in high alarm, an audible horn alarm will sound and a Thursday, March 06, 2014 11:40:27 AM                                                        1
 
strobe light will illuminate and the Fuel Handling Buidling is required to be evacuated.
(Note: Per ARP 17102-1 page 5, not all rad monitors have both a horn and a strobe light.)
Per TRM 13.3.6 Table 13.3.6-1, only 1 Channel of FHB Rad monitor actuation logic is required to be FUNCTIONAL. Since ARE-2532A/B remains in service, the LCO is met.
Personnel will be required to leave the FHB when the alarm comes in. Once the issue is addressed and 1RE-0008 removed from service and the alarm silenced, fuel movement may be resumed. There are no requirements to have ARM 1RE-0008 in service during fuel movement. 1RE-0008 is not listed in Tech Specs, TRM, or ODCM.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per ARP 17100-1, if 1RE-0008 is in high alarm, an audible horn alarm will sound and a strobe light will illuminate. However, the fuel handling building is at times a high noise area when the area circulator fans are in service, and other devices such as the CAS crane and fuel handling machine have audible alarms tha could be confused with an alarm horn. It is reasonable for a candidate with insufficient knowledge of 1RE-0008 to conclude that only a strobe light would be suitable for this area.
The second part is correct. Since ARE-2532A/B remains in service, LCOs 13.3.6 AND 13.9.5 are met. There are no requirements in Tech Specs, TRM, or ODCM for 1RE-008 to be functional and therefore it does not impact fuel handing operations in the fuel handling building.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Since ARE-2532A/B remains in service, LCOs 13.3.6 AND 13.9.5 are met. There are no requirements in Tech Specs, TRM, or ODCM for 1RE-008 to be functional and therefore it does not impact fuel handing operations in the fuel handling building. However, TRM 13.3.6 does direct actions to be taken for TRM 13.9.5 if the required channel of actuation logic is nonfunctional. These actions would include either placing a FHB filter in operation or suspending fuel movement. Therefore, it the candidate believes 1RE-008 is part of TRM 13.3.6, it is reasonable for them to believe that fuel movement could not be resumed with it OOS.
C.Correct.                  The first part is correct. Per ARP 17100-1, if 1RE-0008 is in high alarm, an audible horn alarm will sound and a strobe light will illuminate.
The second part is correct. See the second part of choice A above.
Thursday, March 06, 2014 11:40:27 AM                                                                  2
 
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Level:                          RO Tier # / Group #                T2 / G2 K/A#                            072K3.02 Importance Rating:              3.1 / 3.5 Technical
 
==Reference:==
TRM 13.3.6, Rev 31.0, pages 13.3-19 & 20 TRM 13.9.5, Rev 3.0 9/18/03, pages 13.9-6 & 7 ARP 17100-1, Rev 26.2, page 14 ARP 17102-1, Rev 20.3, page 5 References provided:            None Learning Objective:            LO-LP-39207-03 For any given item in section 13.3 of the Technical Requirements Manual, be able to:
: a. State the Technical Requirement (TR) for operation.
: b. State any one hour or less required actions.
LO-TA-32007        Verify Proper Automatic actions to high radiation alarms Question origin:                MODIFIED - HL18 072K1.03 Cognitive Level:                M/F 10 CFR Part 55 Content:        41.7 / 45.6 Comments:
You have completed the test!
Thursday, March 06, 2014 11:40:27 AM                                                              3
: 1. 072K1.03 001/2/2/ARM - FHBI/C/A - 3.6/3.7/MOD - HL-15R AUDIT/HL-18 NRC/RO/SRO/KAJ A dropped spent fuel assembly in the Unit 1 Spent Fuel Pool has resulted in the following radiation monitor alarms:
Orignal Question RE-0008, FHB Area Monitor, indicates HIGH.
          - A-RE-2532A(B) and A-RE-2533A(B), FHB Effluent Monitors, indicate ALERT.
          - The crew is implementing 18006-C, "Fuel Handling Event".
For the given conditions, which ONE of the following completes the following statement?
1-RE-0008 ___(1)___ provide audible and visual indications of the alarm in the Unit 1 SFP area, and the FHB Post-Accident Filtration Units ___(2)___ automatically start.
A. (1) will (2) will B. (1) will (2) will NOT C. (1) will NOT (2) will D. (1) will NOT (2) will NOT Friday, February 28, 2014 1:36:24 PM                                                          1
 
Fuel Handling Building Post Accident Ventilation Actuation Instrumentation TR 13.3.6 13.3  Instrumentation TR 13.3.6      Fuel Handling Building Post Accident Ventilation Actuation Instrumentation (common system).
TR 13.3.6      The fuel handling building (FHB) post accident ventilation actuation instrumentation identified in Table 13.3.6-1 shall be OPERABLE.
APPLICABILITY:        Whenever irradiated fuel is in either storage pool.
ACTIONS CONDITION                          REQUIRED ACTION                COMPLETION TIME A. One or more required          A.1      Apply Required Actions of      In accordance with FHB ventilation actuation              TR 13.9.5.                    TR 13.9.5.
instruments inoperable.
TECHNICAL REQUIREMENT SURVEILLANCES SURVEILLANCE                                        FREQUENCY TRS 13.3.6.1    Perform CHANNEL CHECK                                      12 hours TRS 13.3.6.2    Perform COT                                                18 months TRS 13.3.6.3    Perform ACTUATION LOGIC TEST                                31 days on a STAGGERED TEST BASIS TRS 13.3.6.4    Perform CHANNEL CALIBRATION                                18 months TRS 13.3.6.5    Perform TADOT                                              18 months Vogtle Units 1 and 2                    13.3 - 19                                  REVISION 31 Technical Requirement
 
Fuel Handling Building Post Accident Ventilation Actuation Instrumentation TR 13.3.6 Table 13.3.6-1 FHB Post Accident Ventilation Actuation Instrumentation Surveillance Instruments          Required Channels          Requirements            Trip Setpoint
: 1. Manual Initiation                1                  TRS 13.3.6.5                  NA
: 2. FHB Exhaust Duct                1                  TRS 13.3.6.1                  (a)
Radiation Signal                                    TRS 13.3.6.2 (ARE-2532 A&B                                      TRS 13.3.6.4 ARE-2533 A&B)
: 3. Automatic                        1                  TRS 13.3.6.3                  NA Actuation Logic and Actuation Relays (a)    Setpoints will not exceed the limits of TS 5.5.4.g.
Vogtle Units 1 and 2                      13.3 - 20                                REVISION 31 Technical Requirement
 
Approved By                                                                                    Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                            17102-1      20.3 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY                  Page Number 11/4/13                                  RELATED DISPLAY CONSOLE QRM2                                  5 of 42 Each channel on the SRDC has a separate display. Normally each display reads the radiation activity level being monitored in 3 digits and an exponent. Units vary from channel to channel. Each channel has an Alert, High and Equipment Trouble alarm display and an indicator that the SRDC is bypassed:
RED        YELLOW        BLUE        AMBER HIGH        ALERT      BYPASS      TROUBLE On detecting a high radiation level, the audible alarm on ALB 05 sounds and the red HIGH indicator lamp on the SRDC channel lights. The alarm is also indicated in the TSC and the Health Physics and Chemistry Labs, by displaying the channel identification number on their CRT in red. The alarm is also displayed at the Communications Console (QRM1). A loud horn and a strobe light may announce the high alarm close to the detector.
Alert, Bypass and Equipment Trouble indications do not sound audible alarms.
For very high radiation levels, the TOP OF SCALE, the EQUIPMENT TROUBLE and the HIGH alarms will all light and the sections of the digital display go to "9999999". This causes the alarm to latch, so it will not automatically clear when the radiation level drops. The TOP OF SCALE must be manually reset at the Channel Display and Control Area (CDCA) on the SRDC.
A high alarm will also latch and require a manual resetting.
The Bypass indicates that the channel has been put in the local control mode at the Data Processing Module (DPM) and for 1-RE-0002 or 1-RE-0003 the Containment Ventilation Isolation Block Switches have been placed in the BLOCK position.
Not all rad monitors have audiable horn and strobe light Printed February 10, 2014 at 13:54
 
Approved By                                                                                Procedure  Version S. E. Prewitt                    Vogtle Electric Generating Plant                        17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND            Page Number 12/9/12                          EFFLUENT RADIATION MONITORING SYSTEM (RMS)                    14 of 88 ORIGIN                          SETPOINT 1-RE-0008 Area Monitor                    As determined by            (High)
Chemistry Department NOTE For other than HIGH conditions see Pages 4 and 5.
1.0                PROBABLE CAUSE Increase in radiation level near Unit 1 Spent Fuel Pool in the Fuel Handling Building.
2.0                AUTOMATIC ACTIONS On the south wall of the Fuel Handling Building Spent Fuel Pool Room near the door:
: a.        Alarm horn on 1-RA-0008 sounds.
: b.        Strobe light on 1-RA-0008 blinks.
3.0                INITIAL OPERATOR ACTIONS Evacuate the Fuel Handling Building.
Printed October 22, 2013 at 12:46
: 1. 073K1.01 001/LOIT AND LOCT/RO/C/A 3.6/3.9/073K1.01/LO-PP-04101-03///
Initial condition:
            - Unit 1 is at 100% reactor power.
Current conditions:
            - 1RE-1950, ACCW - Liquid, radiation level trends up and then stabilizes.
            - ACCW surge tank level increases from 50% to 55% and then stabilizes.
Which one of the following completes the following statement?
A __(1)__ heat exchanger tube leak has occurred, and the ACCW surge tank level stabilized due to an automatic isolation of the flow path due to high __(2)__.
__(1)__                                  __(2)__
A.                  letdown                                    flow B.                  letdown                                temperature C.              thermal barrier                                  flow D.              thermal barrier                            temperature K/A 073              Process Radiation Monitoring K1.01            Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems:
                        - Those systems served by PRMs K/A MATCH ANALYSIS The question tests the candidate's knowledge of the physical connection between ACCW and systems that interface with the RCS (specifically the letdown heat exchange and RCP thermal barrier) and the associated responses of the process rad monitors and system response.
EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17100-1, the source of leakage causing 1RE-1950 to alarm could come from either a letdown heat exchanger or RCP thermal barrier leak. Both RE-1950 and surge Thursday, March 06, 2014 11:42:23 AM                                                        1
 
tank level trending up and then stabilizing are symptoms of the leak being isolated.
Letdown does not have automatic isolation signals from either RE-1950 or the ACCW system. RCP thermal barriers have both individual outlet auto isolations as well as a common header outlet isolation valves. Per 17004-1 ALB A05, B05, C05, and D05, the individual RCP thermal barrier outlet isolation valves will close at a flow rate of 65 gpm.
The ARP directs trending RE-1950 and ACCW surge tank level to validate the alarm, and verify HV-19053 is CLOSED. Per ARP17004-1 B06, the common thermal barrier outlet isolation valve closes on a pressure of 155 psig in the header. (Reference P&ID 1X4DB138-2 for thermal barrier isolation valve and flow transmitter.)
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. As described above in the Explanation of Required Knowledge, the stem gives symptoms of a thermal barrier tube leak and subsequent automatic isolation. However, a letdown heat exchanger leak is also a possible source of leakage per ARP 17100-1 and would exhibit some of the same symptoms (ie. a letdown heat exchanger leak would result in the indication provided in the stem, except indications would contine to rise rather than stablize).
The second part is correct. Per ARP 17004-1, the individual thermal barrier return lines isolate automatically on high flow.
The common isolation will isolate on high flow or pressure.
This distractor is plausible since the letdown flow path has automatic isolations based on temperature and has several alarms associated with flow and temperature.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per ARP 17004-1, the individual thermal barrier return lines isolate automatically on high flow.
However, letdown does have an isolation on high temperature.
(Reference P&ID 1X4DB114 for valves HV15214 and HV8160.)
C. Correct.                  The first part is correct. As described above in the Explanation of Required Knowledge, the stem gives symptoms of a thermal barrier tube leak and subsequent automatic isolation.
The second part is correct. See the first part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the first part of choice B above.
Thursday, March 06, 2014 11:42:23 AM                                                                  2
 
Level:                        RO Tier # / Group #              T/G K/A#                          073K1.01 Importance Rating:              3.6 / 3.9 Technical
 
==Reference:==
ARP 17004-1, Rev 24.0, pages 3, 21 thru 26 ARP 17100-1, Rev 26.2, pages 50 thru 52 ELEMENTARY 1X3D-BD-C01R, Rev 3.0 P&ID 1X4DB114, Rev 41.0 P&ID 1X4DB138-2, Rev 19.0 References provided:          None Learning Objective:            LO-PP-04101-04 From memory describe the expected system response and operator corrective actions for each of the following:
: g. Thermal barrier heat exchanger leak
: i. CVCS letdown heat exchanger leak LO-PP-04101-03 Describe how ACCW surge tank level and RE-1950 are used to determine source of in-leakage and when the in-leakage is isolated.
Question origin:              BANK Cognitive Level:              C/A 10 CFR Part 55 Content:        41.2 to 41.9 / 45.7 to 45.8 Comments:
You have completed the test!
Thursday, March 06, 2014 11:42:23 AM                                                              3
 
Approved By                                                                                    Procedure Number Rev P. H. Burwinkel                      Vogtle Electric Generating Plant                          17004-1        24 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 04 ON PANEL                Page Number 3/11/12                                                  1A1 ON MCB                                      3 of 57 ALB-04 (1)                  (2)            (3)                (4)            (5)            (6)
ACCW SURGE TK        ACCW            ACCW RCP 1 CLR    ACCW RCP 1 CLR ACCW RCP 1 A    HI/LO LVL            LO HDR PRESS    LO FLOW            OUTLET HI TEMP THRM BARRIER HX HI FLOW BOP                  ACCW RX        ACCW RCP 2 CLR    ACCW RCP 2 CLR ACCW RCP 2    ACCW RCP B    PROT GR I            COOLANT DRN TK  LO FLOW            OUTLET HI TEMP THRM BARRIER  THRM BARRIER BYPASS                HX LO FLOW                                        HX HI FLOW    HI PRESS BOP                  ACCW EXCESS    ACCW RCP 3 CLR    ACCW RCP 3 CLR ACCW RCP 3 C    PROT GR II            LTDN HX        LO FLOW            OUTLET HI TEMP THRM BARRIER BYPASS                LO FLOW                                          HX HI FLOW BOP                  ACCW RTN HDR    ACCW RCP 4 CLR    ACCW RCP 4 CLR ACCW RCP 4 D    PROT GR III          FROM RCP        LO FLOW            OUTLET HI TEMP THRM BARRIER BYPASS                LO FLOW                                          HX HI FLOW TRAIN A              TRAIN B                            BOP PROCESS    BOP PROCESS    BOP PCS CABS E    SYS STATUS            SYS STATUS                        PROT CABINET  PROT DOOR OPEN PWR SUPPLY MON PNL ALERT        MON PNL ALERT                      DOORS OPEN    >1 CABINET    FAILURE TRAIN C                              TRAIN A            TRAIN B        TRAIN C        ASIS F    SYS STATUS                            SHUTDOWN PNL      SHUTDOWN PNL  SHUTDOWN PNL  TROUBLE MON PNL ALERT                        ON LOCAL CNTL      ON LOCAL CNTL  ON LOCAL CNTL Printed February 28, 2014 at 14:12
 
Approved By                                                                                      Procedure Number Rev P. H. Burwinkel                    Vogtle Electric Generating Plant                              17004-1        24 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 04 ON PANEL                  Page Number 3/11/12                                                1A1 ON MCB                                      21 of 57 WINDOW B05 ORIGIN                              SETPOINT 1-FSH-19054B                        65 gpm                        ACCW RCP 2 THRM BARRIER HX HI FLOW 1.0                PROBABLE CAUSE
: 1.        Leak into Auxiliary Component Cooling Water (ACCW) from Reactor Coolant Pump (RCP) 2 Thermal Barrier Heat Exchanger (HX).
: 2.        Leak downstream of 1-FE-19054.
2.0                AUTOMATIC ACTIONS RCP 2 Thermal Barrier HX Outlet Isolation Valve 1-HV-19053 closes on high flow of 69 gpm.
3.0                INITIAL OPERATOR ACTIONS NONE 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        Verify RCP 2 seal injection and return flows are normal.
: 2.        Check Radiation Monitor 1-RE-1950 for signs of radiation leakage into ACCW.
: 3.        Check computer point L2700 for increasing Surge Tank level.
: a. IF ACCW Surge tank level was NOT rising, (1)    IF closure was spurious (for example closed during ACCW Pump swap, power supply fluctuation or maintenance activities) AND leakage is NOT suspected) monitor Surge tank level AND attempt to open 1-HV-19053.
(2)    IF closure reason is unknown OR leakage is suspected, maintain 1-HV-19053 closed, continue with this procedure AND initiate maintenance.
: b. IF ACCW Surge tank level was OR is rising, verify 1-HV-19053 CLOSED.
Printed February 28, 2014 at 14:12
 
Approved By                                                                                        Procedure Number Rev P. H. Burwinkel                    Vogtle Electric Generating Plant                              17004-1        24 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 04 ON PANEL                    Page Number 3/11/12                                                  1A1 ON MCB                                      22 of 57 WINDOW B05 (Continued)
: 4.        IF RCP 2 temperatures and seal flows are normal and there is no indication of radiation in the ACCW System, initiate maintenance as required to correct cause of the alarm.
: 5.        IF 1-HV-19053 closes, verify the RCP is operated within the limits established in 13003-1, "Reactor Coolant Pump Operation" and ensure actions of Technical Requirement (TR) 13.7.4 are met.
: 6.        IF 1-HV-2041 closed, restore flow to the intact thermal barriers as follows:
: a.      Check computer trend (IPC Point L2700) to determine if ACCW Surge tank level was rising prior to isolation.
: b.      Check ACCW Surge tank level is now stable (leak is isolated).
: c.      Check leaking thermal barrier isolated (1-HV-19053 closed).
: d.      Check 1-HV-2041 has not been closed for greater than 30 minutes to preclude the possible formation of steam in the thermal barrier.
(1)    IF 1-HV-2041 has been closed greater than 30 minutes, contact engineering to evaluate opening 1-HV-2041 PRIOR to proceeding with this step.
: e.      Monitor ACCW Surge tank for level rise in the following step:
(1)    Open 1-HV-2041.
(2)    IF a level rise is observed, close 1-HV-2041.
(3)    Check Surge Tank level stable.
Printed February 28, 2014 at 14:12
 
Approved By                                                                                        Procedure Number Rev P. H. Burwinkel                    Vogtle Electric Generating Plant                              17004-1        24 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 04 ON PANEL                    Page Number 3/11/12                                                  1A1 ON MCB                                      23 of 57 WINDOW B05
: 7.        IF it becomes necessary to stop the pump and:
: a.      Reactor power is above 15%:
(1)    Trip the reactor and initiate 19000-C, "E-0 Reactor Trip Or Safety Injection,"
(2)    Stop the pump.
: b.      Reactor power is 15% or less, stop the pump and initiate 18005-C, "Partial Loss Of Flow."
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB138-2, 1X3D-BD-L03L, CX5DT101-130 Printed February 28, 2014 at 14:12
 
Approved By                                                                                      Procedure Number Rev P. H. Burwinkel                    Vogtle Electric Generating Plant                              17004-1        24 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 04 ON PANEL                  Page Number 3/11/12                                                1A1 ON MCB                                      24 of 57 WINDOW B06 ORIGIN                              SETPOINT 1-PSH-2041B                          155 psig                      ACCW RCP THRM BARRIER HI PRESS 1.0                PROBABLE CAUSE Leak into Auxiliary Component Cooling Water (ACCW) from a Reactor Coolant Pump (RCP) Thermal Barrier Heat Exchanger (HX).
2.0                AUTOMATIC ACTIONS ACCW RCP Thermal Barrier Outlet Header Isolation Valve 1-HV-2041 closes on high header pressure and/or flow.
3.0                INITIAL OPERATOR ACTIONS NONE 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        Verify valve 1-HV-2041 is closed.
: 2.        Verify all RCP seal injection and return flows are normal.
: 3.        Check Radiation Monitor 1-RE-1950 for signs of radiation leakage into ACCW.
: 4.        Verify the RCPs are operated within the limits established in 13003-1, "Reactor Coolant Pump Operation."
Printed February 28, 2014 at 14:12
 
Approved By                                                                                        Procedure Number Rev P. H. Burwinkel                    Vogtle Electric Generating Plant                                17004-1        24 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 04 ON PANEL                    Page Number 3/11/12                                                1A1 ON MCB                                        25 of 57 WINDOW B06 (Continued)
: 5.        IF 1-HV-2041 closed, restore flow to the intact thermal barriers as follows:
: a. Check computer trend (IPC Point L2700) to determine if ACCW Surge tank level was rising prior to isolation.
(1)  IF ACCW Surge tank level was not rising, THEN perform (a) OR (b):
(a)    IF closure was spurious (for example closed during ACCW Pump swap, power supply fluctuation or maintenance activities) AND leakage is NOT suspected) monitor Surge tank level AND attempt to open 1-HV-2041.
(b)  IF closure reason is unknown, maintain 1-HV-2041 closed, initiate maintenance AND go to step 5.b.
(2)    IF ACCW Surge tank level was rising, continue with next step.
: b. Check ACCW Surge tank level is now stable.
: c. Check 1-HV-2041 has not been closed for greater than 30 minutes to preclude the possible formation of steam in the thermal barrier.
(1)    IF 1-HV-2041 has been closed greater than 30 minutes, contact engineering to evaluate opening 1-HV-2041 PRIOR to proceeding with this step.
: d. Check ANY individual RCP Isolation valve CLOSED to isolate thermal barrier leaking.
1-HV-19051 for RCP-1 1-HV-19053 for RCP-2 1-HV-19055 for RCP-3 1-HV-19057 for RCP-4
: e. Monitor ACCW Surge tank for level rise in the following step:
(1)    Open 1-HV-2041.
(2)    IF a level rise is observed, close 1-HV-2041.
(3)    Check Surge Tank level stable.
Printed February 28, 2014 at 14:12
 
Approved By                                                                                        Procedure Number Rev P. H. Burwinkel                    Vogtle Electric Generating Plant                              17004-1        24 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 04 ON PANEL                    Page Number 3/11/12                                                1A1 ON MCB                                        26 of 57 WINDOW B06 (Continued)
: 6.        IF it becomes necessary to stop the pump and:
: a.      Reactor power is above 15%:
(1)    Trip the reactor, (2)    Initiate 19000-C, "E-0 Reactor Trip Or Safety Injection,"
(3)    Stop affected RCP, and (4)    Verify Thermal Barrier Outlet Isolation Valve of affected RCP is closed.
1-HV-19051 for RCP-1 1-HV-19053 for RCP-2 1-HV-19055 for RCP-3 1-HV-19057 for RCP-4
: b.      Reactor power is 15% or less:
(1)    Stop affected RCP, (2)    Verify Thermal Barrier Outlet Isolation Valve of affected RCP is closed.
(3)    Initiate 18005-C, "Partial Loss Of Flow."
: 7.        IF equipment failure is indicated, initiate maintenance as required and ensure actions of Technical Requirement (TR) 13.7.4 are met.
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB138-2, 1X3D-BD-L03P, CX5DT101-129 Printed February 28, 2014 at 14:12
 
Approved By                                                                                  Procedure  Version S. E. Prewitt                      Vogtle Electric Generating Plant                        17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND              Page Number 12/9/12                          EFFLUENT RADIATION MONITORING SYSTEM (RMS)                      50 of 88 ORIGIN                            SETPOINT 1-RE-1950 Skid mounted                      As determined by            (High)
Liquid Process                    Chemistry Department Monitor NOTE For other than HIGH conditions see Pages 4 and 5.
1.0                PROBABLE CAUSE High radiation level in the ACCW from inleakage of radioactive water.
2.0                AUTOMATIC ACTIONS NONE 3.0                INITIAL OPERATOR ACTIONS NONE 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        Request Chemistry to sample and analyze the ACCW.
: 2.        Notify Health Physics of the alarm.
: 3.        Locate the source of inleakage.
: a. Check IPC points T0145, P0135 and F0134 (IPC Group 21) for changes, in an attempt to determine if a Letdown HX tube leak.
: b. Check IPC points T2714, T2716, T2718 and T2720 (ICP Group 242) for changes, in an attempt to determine if leakage is from a RCP thermal barrier.
: 4.        Isolate the source if possible.
Printed February 28, 2014 at 14:05
 
Approved By                                                                              Procedure  Version S. E. Prewitt                      Vogtle Electric Generating Plant                      17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND          Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                  51 of 88 1-RE-1950 (Continued)
: 5.        IF 1-RE-1950 is reading high due to Letdown Heat Exchanger Tube leakage:
Critical
: a. Place LETDOWN TO DEMIN/VCT 1-TV-0129 to the VCT position using 1-HS-0129.                          ________
Initial
________
CV Initial Critical (1)    Verify 1-TV-0129 aligns to the VCT.            ________
Initial
________
CV Initial Critical
: b. Place VCT HUT LETDOWN DIVERT 1-LV-0112A to the HUT position using 1-HS-0112A.                    ________
Initial
________
CV Initial Critical (1)    Verify 1-LV-0112A aligns to the RHUT.          ________
Initial
________
CV Initial Printed February 28, 2014 at 14:05
 
Approved By                                                                                  Procedure  Version S. E. Prewitt                      Vogtle Electric Generating Plant                        17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND              Page Number 12/9/12                            EFFLUENT RADIATION MONITORING SYSTEM (RMS)                      52 of 88 1-RE-1950 (Continued)
Critical
: c. Isolate letdown. Verify closed:                          ________
Initial
________
CV Initial (1)    1-HV-8149A, B, C.
(2)    1-LV-0459.
(3)    1-LV-0460.
(4)    1-HV-8152.
(5)    1-HV-8160.
(6)    1-PV-0131,-set to max pressure.
(7)    1-TV-0130,-set to max temperature.
: d. Shut Letdown Heat Exchanger manual valves:
(1)    (AB-A08) 1-1208-U6-041.
(2)    (AB-A17) 1-1217-U4-126.
(3)    (AB-108) 1-1217-U4-129.
: e. Notify Chemistry,
: f. Initiate 18007-C, "CVCS Malfunction" to deal with the loss of letdown.
: 6.        Obtain detector trend from the IPC computer.
: 7.        Monitor the channel for further changes.
: 8.        IF sampling and analysis determine the channel has malfunctioned, request Chemistry to deactivate the channel.
Printed February 28, 2014 at 14:05
: 1. 076K2.04 001/LOIT/RO/C/A 2.5/2.6/076K2.04///003K2.02/
Initial conditions:
            - Unit 1 is at 100% reactor power.
            - ACCW pump #1 is running.
            - ACCW pump #2 is in standby.
Current conditions:
            - An SI occurred and has been reset.
            - 5 minutes later, RAT '1A' experiences a fault.
            - DG1A re-energizes 1AA02 and completes the load sequence.
Which one of the following completes the following statement?
Based on the given sequence of events, ACCW pump #1 __(1)__ be running, and ACCW pump #2 __(2)__ be running.
__(1)__                              __(2)__
A.                      will                                will B.                      will                              will NOT C.                  will NOT                                will D.                  will NOT                              will NOT K/A 076              Service Water K2.04            Knowledge of bus power supplies to the following:
                        - Reactor building closed cooling water.
K/A MATCH ANALYSIS The question tests the candidate knowledge of which ACCW pump(s) will be in service following various events which change the power status of both 1E 4160V buses.
(Note: Vogtle's bus scheme is extremely simple with just a single 4160V bus powered by a RAT or a DG. Interaction with the Sequencer has been incorporated to increase LOD above a 1.0.)
EXPLANATION OF REQUIRED KNOWLEDGE Thursday, March 06, 2014 1:25:32 PM                                                        1
 
Per ELEMENTARY 1X3D-BD-L03B, if an SI actuation occurs, SSPS energizes a slave relay contact in the ACCW pump start circuit that prevents all autostarts. This block will remain until SI has been RESET. The SI actuation does not change the status of either pump beyond inserting this block. Since the SI signal has been reset in this question, the block is no longer inserted and the ACCW pumps will behave normally. When the LOSP occurs on 1AA02, the ACCW header pressure lowers and ACCW Pump #2 will auto start. In parallel with this, the 'A' train sequencer performs a load shed, the already running DG (which started on the SI signal and continues to run until manually shut down) output breaker closes, the bus re-energizes, and the LOSP sequence is run.
ACCW Pump #1 is started as part of the LOSP sequence. Therefore, ACCW Pumps
        #1 and #2 will be running. (Reference ONELINE 1X3D-AA-K02A)
ANSWER / DISTRACTOR ANALYSIS A. Correct                  The first part is correct. ACCW Pump #1 will restart on the LOSP sequence since the SI signal has been reset.
The second part is correct. ACCW Pump #2 will autostart on low header pressure since it is not blocked by the SI signal when ACCW Pump #1 loses power.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. ACCW Pump #2 will autostart on low header pressure since it is not blocked by the SI signal when ACCW Pump #1 loses power. However, if the candidate does not remember the low header pressure autostart, they will not expect ACCW Pump #2 to be running, but will expect ACCW Pump #1 to be started by the LOSP sequence.
C. Incorrect. Plausible. The first part is incorrect. ACCW Pump #1 will restart on on the LOSP sequence since the SI signal has been reset. However, if the candidate misses that the SI has been reset, and believes the low pressure autostart is not affected by the SI signal then they will not expect ACCW Pump #1 to start as part of the SI Sequence, but will expect ACCW Pump #2 to have started when ACCW Pump #1 lost power.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. Both parts are incorrect. ACCW Pump #1 will restart on the LOSP sequence since the SI signal has been reset. ACCW Pump #2 will autostart on low header pressure since it is not blocked by the SI signal when ACCW Pump #1 loses power.
However, if the candidate misses the SI reset and believes the SI signal is still present, then neither ACCW pump will be running and this would be the correct answer.
Thursday, March 06, 2014 1:25:32 PM                                                                2
 
Level:                        RO Tier # / Group #              T2 / G1 K/A#                          076K2.04 Importance Rating:            2.5 / 2.6 Technical
 
==Reference:==
ELEMENTARY 1X3D-BD-L03B, Rev 10.0 ONELINE 1X3D-AA-K02A, Rev 15.0 References provided:          None Learning Objective:          LO-LP-60318-01 Describe how the ACCW pumps are affected by a simultaneous loss of offsite power and safety injection.
LO-PP-11101-54 Describe the Operation of the sequencer relative to the EDG.
LO-TA-60005    Respond to a Loss of ACCW per 18022-C LO-TA-60009A  Respond to a Loss of Class 1E Electrical Systems per 18031-1/2 Question origin:              MODIFIED - HL17 NRC # 003K2.02 Cognitive Level:              C/A 10 CFR Part 55 Content:      41.7 Comments:
You have completed the test!
Thursday, March 06, 2014 1:25:32 PM                                                          3
: 1. 003K2.02 001/2/1/RCP-POWER ACCW/H-2.5*/2.6*/NEW/HL17 NRC/RO/SRO/TNT/GCW Initial conditions:
          - ACCW pump # 1 is running.                      Original Question
          - ACCW pump # 2 is in standby.
Current sequence of events:
          - Safety Injection occurs.
          - 2 minutes later, RAT "1A" experiences a fault.
          - DG1A energizes 1AA02.
          - SI has NOT been reset.
Based on the current conditions, which one of the following correctly describes the status of the ACCW pumps after the DG1A load sequence is complete, if no operator actions occur?
ACCW Pump # 1                ACCW Pump # 2 A.      OFF                          OFF B.      RUNNING                      OFF C.      OFF                          RUNNING D.      RUNNING                      RUNNING Monday, March 03, 2014 8:48:51 AM                                                          1
 
"M" denotes a manual start.
: 1. 077AK3.02 001/LOCT/RO/M/F 3.6/3.9/077AK3.02/LO-TA-60043///
Initial condition:
            - Unit 1 is at 100% reactor power.
Current condition:
            - 18017-C, "Abnormal Grid Disturbances / Loss of Grid," Section 'A' for "Degraded Grid Conditions," is in progress.
Which one of the following completes the following statement?
The DGs are maintained in standby alignment to __(1)__,
and the Main Generator is operated within the acceptable region of the reactive capability curve to prevent damage to the __(2)__ due to overheating.
A. (1) prevent a grid disturbance from impacting availability (2) Main Generator B. (1) prevent a grid disturbance from impacting availability (2) reactive loads C. (1) comply with the required Tech Spec alignment for operability (2) Main Generator D. (1) comply with the required Tech Spec alignment for operability (2) reactive loads K/A 077              Generator Voltage and Electric Grid Disturbances AK3.02          Knowledge of the reasons for the following responses as they apply to the Generator Voltage and Electric Grid Disturbances:
                        - Actions contained in abnormal operating procedure for voltage and grid disturbances.
K/A MATCH ANALYSIS The question tests the candidate's knowledge of the reasons for placing the EDG's in standby alignment as directed by the grid distrubance AOP and the reason for ensuring Thursday, March 06, 2014 11:45:05 AM                                                          1
 
the main generator is operated within the capacity curve during a grid disturbance.
EXPLANATION OF REQUIRED KNOWLEDGE Per AOP 18017-C step A1, DG's are checked "IN STANDBY". The RNO to this step is to "restore DG's to operable status". The RNO is somewhat misleading in that a DG is OPERABLE when paralleled to the Grid. AOP 18017-C does not have a background document, however the intent of this step is to place the DG's in a standby alignment, not just for them to be OPERABLE. Lesson plan V-LO-LP-60330 states "it is not desirable to go ahead and start the engines and run unloaded or even parallel to the bus". The lesson plan cites issues with cylinder loading. Additionally, IN 84-69 "SUPPLEMENT 1: OPERATION OF EMERGENCY DIESEL GENERATORS," states that EDGs should not be paralleled to the Grid during times of instability to prevent the Grid from damaging the EDG and rendering both AC sources to the bus incapable of supplying emergency power. The safest configuration for the EDGs is in standby.
Per AOP 18017-C step A5, the Main Generator is to be maintained within the capability curve of Figure 1. Figure 1 depicts the generator operational limits associated with the bounding curves. Additonally, lesson plan V-LO-LP-60330 states that if the main generator cannot be maintained within the capability curve, the reactor and generator are tripped to prevent damage to the generator windings due to frequency/voltage swings.
ANSWER / DISTRACTOR ANALYSIS A. Correct.                  The first part is correct. The reason for maintaining the diesel generators in standby alignment is to eliminate or reduce the potential for a grid disturbance impacting both the normal and emergency sources of power at the same time.
The second part is correct. The reason for maintaining the main generator operations within the prescribed region of the reactive curve is to prevent overheating of the generator components themselves to include the stator winding, rotor, and conductors.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. The candidate must consider the potential damage to components due to excessive VAR loading output by the main generator. Being outside the Tech Spec required voltage (either above or below limits) can damage components. In a degraded grid condition, however, the main generator would actually respond to attempt to maintain the grid voltage within limits, not drive voltage outside the limits. This automatic control action exposes the main generator to potential damage and the stem focuses on the main generator, not the impacted loads.
C. Incorrect. Plausible. The first part is incorrect. The reason for maintaining the diesel Thursday, March 06, 2014 11:45:06 AM                                                                  2
 
generators in standby alignment is to eliminate or reduce the potential for a grid disturbance impacting both the normal and emergency sources of power at the same time. However, the candidate must consider the Tech Spec (TS) required alignment. Tech Specs refer to the diesel generator as the standby source which implies a required alignment. The candidate may not remember that the diesel generator is considered OPERABLE either in standby or parallel condition and assume the EDG is only OPERABLE in standby. 18017-C step A1 tends to back up this false logic in that the step states "Check... in standby" and the RNO states, "restore... to operable status."
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Thursday, March 06, 2014 11:45:06 AM                                                                  3
 
Level:                        RO Tier # / Group #              T1 / G1 K/A#                          077AK3.02 Importance Rating:            3.6 / 3.9 Technical
 
==Reference:==
18017-C, Rev 9.5, pages 3 & 7 Lesson Plan V-LO-LP-60330, Rev 4.0, pages 7 & 8 INPO IN 84-69, Supplement 1, February 24, 1986 References provided:          None Learning Objective:            LO-LP-60330-03 State the reasons for tripping the reactor in AOP-18017-C.
LO-LP-60330-06 State the control room indications for a loss of grid.
LO-LP-60330-07 State the consequences of operating the 1E 4160 buses outside the voltage limits listed in AOP-18017.
LO-LP-60330-09 Given a set of control room indications/notifications, determine if entry conditions for either section of 18017 is met.
LO-LP-60330-10 Given that a loss of grid has occurred from 100% power, describe the expected flowpath through the EOPs and AOP 18017-C.
LO-LP-60104-05 State why, when anticipating an LOP, the Control Room Operator is instructed NOT to start the diesel generator(s).
LO-TA-60043        Respond to an Abnormal Grid Disturbance per 18017-C Question origin:              NEW Cognitive Level:              M/F 10 CFR Part 55 Content:        41.4 / 41.5/ 41.7/ 41.10 / 45.8 Comments:
You have completed the test!
Thursday, March 06, 2014 11:45:06 AM                                                                4
 
Approved By                                                                        Procedure    Version J.B. Stanley                          Vogtle Electric Generating Plant            18017-C        9.5 Effective Date                                                                      Page Number ABNORMAL GRID DISTURBANCES/LOSS OF 05/07/2013                                          GRID                                  3 of 53 A. DEGRADED GRID CONDITIONS ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED A1 A1.      Check Diesel Generators - IN              A1. Restore Diesel Generators to STANDBY.                                      operable status.
A2 A2.      Terminate maintenance or testing          A2.
activities on critical electrical distribution components.
A3 A3.      Check Main Generator Power System          A3. Perform actions of TABLE 1, as Stabilizer on COI - PSS ENABLED.              necessary.
A4 A4.      Initiate the Continuous Actions Page.      A4.
A5
        *A5. Maintain Main Generator -                  A5. IF Main Generator can NOT be OPERATING WITHIN THE                            maintained within the capability REACTIVE CAPABILITY CURVE OF                    curve, FIGURE 1.                                      THEN trip the reactor and initiate 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.
A6 A6.      Place the following on alternate          A6.
power supply using 13800, MAIN TURBINE OPERATION:
Main Turbine Turning Gear Turning Gear Oil Pump A7 A7.      Verify Turning Gear Oil Pump:              A7.
With turbine on line - IN AUTO.
                        -OR-With turbine on turning gear - IN              Verify Main Turbine Auxiliary OPERATION.                                    Emergency DC Lube Oil Pump is operating.
 
S Printed January 23, 2014 at 15:03
 
Approved By                                                          Procedure    Version J.B. Stanley                        Vogtle Electric Generating Plant 18017-C        9.5 Effective Date                                                        Page Number ABNORMAL GRID DISTURBANCES/LOSS OF 05/07/2013                                      GRID                      7 of 53 FIGURE 1            Sheet 1 of 1 REACTIVE CAPABILITY CURVE Printed March 3, 2014 at 10:19
 
V-LO-LP-60330 III. LESSON OUTLINE
: 2) A DG paralleled to the grid during degraded voltage conditions may trip if conditions worsen just when it will be called upon to start. Local operator actions may be necessary to return the DG to service. The        13145 DG could also sustain damage due to the transient.
Precaution 2.1.10 The SOPs 13145 have the following precaution:
The Diesel Generators should NOT be operated in parallel with the offsite grid for prolonged periods of time. This is to keep disturbances in the grid from affecting the Diesel Generators.
The OSPs 14980 have the following precautions:
During surveillance testing, only one DG shall be paralleled at a time to the off site power source.                                              14980 To serve as a dependable backup power source, a DG should be              Precautions 3.11 kept separate from the offsite source if it is the only OPERABLE diesel. The DG should remain in standby and only be paralleled with        and 3.18 an offsite source to meet surveillance requirements. Parallel operations may be conducted as a part of a preplanned activity if a supporting risk assessment has been completed.
: 3. Any maintenance in progress on critical electrical equipment should be stopped.
a    This equipment should be restored to service if possible.
b    If the maintenance is being performed due to equipment inoperability, the SS should evaluate the need to continue work in an attempt to restore the component to service.
: 4. Make sure Main Generator Power System Stabilizer in service                        Review Table 1
: a. PSS increases the EX2100 AC regulator control stability by negating the regulators inherent lag response to a problem which left to itself tends to destabilize generator output during relatively small transients.
: b. The action of the PSS is to increase stability and prevent the voltage Regulator from driving the system into increasing amplitude oscillations.
: 5. Maintain proper Main Generator operation within the Reactive Capability            Objective 3 Curve of 13830, Main Generator Operation.                                          Review Capability Curve Useage
: a. If it cannot be maintained within the capability curve, the Reactor Is          Objective 3, 4, 5, 8 required to be tripped to prevent damage to generator windings due to frequency/voltage swings and Emergency Procedures entered.
: 6. Inform shift personnel of the grid condition and stress the potential for a loss of site power.
: a. The intent is to ensure everyone has a chance to prepare and implement contingency actions should the loss of power occur.
: b. These actions may include the call in of additional personnel or at least placing them on call to more readily respond if needed.
8
 
V-LO-LP-60330 III. LESSON OUTLINE
: 2. Notification from the Power Control Center that the distribution center is "one Entry into Section A contingency away" from being unable to maintain system voltage between 230 and 242 kV.
: a. The statement that the distribution center is "one contingency away" from being able to maintain grid voltage within limits means that one more system fault will cause a degraded system to exist.
: 3. Annunciator Response Procedures 17036 and 17037, SEQ TROUBLE                    Entry into Section A windows, if a degraded grid condition exists.
: 4. Section B is entered when All Offsite Power are Deenergized                  Objective 9
 
==C. PROCEDURE==
USAGE-SECTION A DEGRADED GRID VOLTAGE
: 1. Purpose of Section A - The intent is to ensure the plants vital equipment is available for a Loss of Offsite Power.
: 2. The Diesel Generators are verified to be operable and in standby.
: a. The DGs are the backup emergency power supplies for the 1E bus.
: b. It is important to note that it may be desirable to station someone at the DG to ensure it starts and runs properly in the event a degraded voltage condition eventually develops.
: c. It is not desirable to go ahead and start the engines and run unloaded or even parallel to the bus.
: 1) Extended unloaded or low load operation results in the buildup of        SMART - Ensuring combustion products in the engine exhausts due to incomplete fuel    DG availability burn in the cylinders.
a) These products can eventually cause fouling of cylinder valves b)  preventing proper operation and reduction in the load carrying capacity of the DG.
7
 
V-LO-LP-60330 III. LESSON OUTLINE
: 2) A DG paralleled to the grid during degraded voltage conditions may trip if conditions worsen just when it will be called upon to start. Local operator actions may be necessary to return the DG to service. The        13145 DG could also sustain damage due to the transient.
Precaution 2.1.10 The SOPs 13145 have the following precaution:
The Diesel Generators should NOT be operated in parallel with the offsite grid for prolonged periods of time. This is to keep disturbances in the grid from affecting the Diesel Generators.
The OSPs 14980 have the following precautions:
During surveillance testing, only one DG shall be paralleled at a time to the off site power source.                                              14980 To serve as a dependable backup power source, a DG should be              Precautions 3.11 kept separate from the offsite source if it is the only OPERABLE diesel. The DG should remain in standby and only be paralleled with        and 3.18 an offsite source to meet surveillance requirements. Parallel operations may be conducted as a part of a preplanned activity if a supporting risk assessment has been completed.
: 3. Any maintenance in progress on critical electrical equipment should be stopped.
a    This equipment should be restored to service if possible.
b    If the maintenance is being performed due to equipment inoperability, the SS should evaluate the need to continue work in an attempt to restore the component to service.
: 4. Make sure Main Generator Power System Stabilizer in service                        Review Table 1
: a. PSS increases the EX2100 AC regulator control stability by negating the regulators inherent lag response to a problem which left to itself tends to destabilize generator output during relatively small transients.
: b. The action of the PSS is to increase stability and prevent the voltage Regulator from driving the system into increasing amplitude oscillations.
: 5. Maintain proper Main Generator operation within the Reactive Capability            Objective 3 Curve of 13830, Main Generator Operation.                                          Review Capability Curve Useage
: a. If it cannot be maintained within the capability curve, the Reactor Is          Objective 3, 4, 5, 8 required to be tripped to prevent damage to generator windings due to frequency/voltage swings and Emergency Procedures entered.
: 6. Inform shift personnel of the grid condition and stress the potential for a loss of site power.
: a. The intent is to ensure everyone has a chance to prepare and implement contingency actions should the loss of power occur.
: b. These actions may include the call in of additional personnel or at least placing them on call to more readily respond if needed.
8
 
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 February 24, 1986 Information Notice No. 84-69, SUPPLEMENT 1: OPERATION OF EMERGENCY DIESEL GENERATORS Addresses:
All nuclear power reactor facilities holding an operating license (OL) or a construction permit (CP).
Purpose:
Information Notice 84-69, issued on August 29, 1984, was provided to alert recipients of potentially significant safety problems that can arise when one or more emergency diesel generators (EDGs) are operated in modes other than the prescribed standby service mode, such as loaded on non-emergency buses parallel with offsite power sources. The purpose of this supplement is to reemphasize the need for licensees to review the information provided in IN 84-69, in addition to the information contained herein, for applicability to their facilities and consider actions, if appropriate, to preclude similar problems at their facilities. However, suggestions contained in this supplement do not constitute NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances:
Following a 10 CFR 50.72 report made to the NRC Headquarters Operations Center on August 12, 1985, it was discovered that Crystal River Unit 3 was continuously running the one operable EDG loaded in parallel with the grid while the other EDG was declared inoperable. Crystal River Technical Specifications require fast starting of the operable EDG (i.e., verifying that the diesel starts from ambient conditions and accelerates to the required speed within a required period of time) within 1 hour after the declaration of an inoperable EDG and every 8 hours thereafter. Because of a concern about increased EDG wear and reduced overall EDG reliability, the licensee chose to keep the EDG running loaded parallel to the offsite grid rather than fast starting the EDG every 8 hours.
The licensee believed that continuous running was an acceptable alternative to the test starts required by the Technical Specifications and that the EDG was operable per Technical Specifications while running in parallel with the offsite power system. The licensee indicated also that it was aware of IN 84-69 and had implemented procedures that prohibited operating the EDG parallel to the grid during inclement weather (e.g. , lightning, heavy
 
winds).
.
IN 84-69, Supplement 1 February 24, 1986 Page 2 of 3 Discussion:
When an EDG is operated connected to offsite or nonvital loads, the emergency power system is not independent of disturbances on the nonvital and offsite power systems that can adversely affect emergency power availability. The situation is of particular concern when the onsite emergency power system is already in a degraded condition due to an EDG being inoperable and the operable EDG is loaded on non-emergency loads. In this condition, a disturbance in the non-emergency power system could result in both a loss of offsite power and a disabling of the remaining emergency power source. Although the events described in IN 84-69 occurred due to weather conditions, the concerns of the IN apply to parallel operation of EDGs with non-emergency loads at all times.
If a fault develops while the EDG is connected to non-emergency buses, EDG availability for subsequent emergency demands may be affected. In some design configurations, the EDG would trip as a result of overcurrent or reverse power, actuate a lockout device, and require local operator action to reset the lockout. In such cases, the EDG is recoverable, but the timeliness of its availability is not comparable to that of having the EDG in its normal standby service.
In other design configurations the EDG may not trip, but the operation of the load sequencer may be adversely affected. The load sequencer timers are often linked with the closing of the EDG output breaker or with detection of loss of voltage on the bus. If the EDG does not trip, conditions are not proper for the designed operation of the load sequencers. Consequently, the EDG cannot perform automatically in a manner comparable to that of having the EDG in its normal standby mode.
Another potential concern deals with the vulnerability of the EDG to trip signals which are bypassed for emergency demands but are operable for manual starts and during running for test purposes. The EDG would be more vulnerable to such trips.
The licensee's concern regarding excessive test starts is valid. In this particular case, the licensee was encouraged to address that concern more directly by submitting changes to the plant Technical Specifications. Such changes were approved for North Anna Unit 2 on April 25, 1985.
.
IN 84-69, Supplement 1
 
February 24, 1986 Page 3 of 3 No specific action or written response is required by this information notice. If you have any questions about this matter, please contact the Regional Administrator of the appropriate regional office or this office.
Edward L. Jordan, Director Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical Contacts: Joseph G. Giitter, IE (301) 492-9001 J. T. Beard, NRR (301) 492-7465
 
==Attachment:==
List of Recently Issued Information Notices
: 1. 078A4.01 001/LOIT AND LOCT/RO/C/A 3.1/3.1/078A4.01/LO-TA-02013///079A4.01 Initial conditions:
            - Unit 1 is at 100% reactor power.
            - Air Compressors #1 and #3 are in service.
Current conditions:
            - ALB01-C06 SERVICE AIR HDR LO PRESS is received.
            - 1PI-9377, Service Air Header Pressure, on the QMCB lowered to 92 psig and is now stable at 105 psig.
            - NO operator action has been taken.
Which one of the following completes the following statement?
All air compressors in AUTO-PTL __(1)__ currently running, and 1-PV-9375, Service Air Dryer Inlet Isolation Valve, __(2)__ closed.
__(1)__                                  __(2)__
A.                      are                                      is B.                      are                                  is NOT C.                  are NOT                                      is D.                  are NOT                                  is NOT K/A 078              Instrument Air A4.01            Ability to manually operate and/or monitor in the control room:
                        - Pressure gauges.
K/A MATCH ANALYSIS The question matches the KA by testing the candidate's abillity to determine the expected system response as the instrument air system pressure indication indication lowers in the main control room. The candidate is required to monitor air pressure and determine how the instrument air system will respond.
Thursday, March 06, 2014 11:47:02 AM                                                        1
 
EXPLANATION OF REQUIRED KNOWLEDGE AOP 18028-C requires the operator to verify that numerous automatic actions occur on lowering Instrument Air header pressure. Per SOP 13710-1 (Note prior to section 4.2.1) and 17001-C ALB01-C06, all air compressors with handswitches in AUTO PTL will start at a pressure of 100 psig as sensed at its local controller. This is a silent action because no control room annunciator alerts the operator that the setpoint has been reached. All air compressors that auto start continue to run until manually stopped by the operator.
Per ARP 17001-1, ALB01-C06 SERVICE AIR HDR LO PRESS will alarm at 95 psig.
There are no equipment actuations associated with this setpoint. If pressure continues to lower, service air isolation valve PV-9375 will close at a header pressure of 80 psig.
PV-9375 requires two conditions to reopen: First, PI-19380 must read >97 psig.
Second, PSL-9375 must be reset locally.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. The standby air compressors automatically start at 100 psig on lowering header pressure, as sensed at its local controller.
The second part is incorrect. The alarm setpoint for ALB01-C06 is 95 psig. No equipment actuations occur at this setpoint - it provides an alarm only. Service air isolation valve PV-9375 will close at a header pressure of 80 psig. The lowest pressure observed was 92 psig. Therefore, PV-9375 is open. However, candidates often confuse the annunciators associated with lowering instrument air pressure and it is common for candidates to encounter ALB01-C06 and believe that it alerts the operator to the closure of PV-9375.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. The alarm setpoint for ALB01-C06 is 95 psig. No equipment actuation occur at this setpoint - it provides an alarm only. Service air isolation valve PV-9375 will close at a header pressure of 80 psig. The lowest pressure observed was 92 psig. Therefore, PV-9375 is open.
C. Incorrect. Plausible. The first part is incorrect. The standby air compressors automatically start at 100 psig on lowering header pressure, as sensed at its local controller. However, it is a common misconception that air compressors will auto stop if header pressure rises above 100 psig. This was true for the original air compressor controllers, but is not part of the current control system. In addition, other components like the RMWST and boric acid pumps do auto stop when there is no longer a demand signal present.
Thursday, March 06, 2014 11:47:02 AM                                                                  2
 
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Level:                          RO Tier # / Group #                T2 / G1 K/A#                            078A4.01 Importance Rating:              3.1 / 3.1 Technical
 
==Reference:==
ARP 17001-1, Rev 31.1, pages 23, 33, & 34 SOP 13710-1, Rev 39.0, page 15 References provided:            None Learning Objective:            LO-LP-60321-06 Describe the operator actions required during normal full power operation when instrument air header pressure fails below 80 psig and/or below 70 psig.
LO-PP-02101-09 List the sequence of major events on a decreasing instrument air pressure condition.
LO-TA-02013        Respond to a loss of instrument air using 18028-C Question origin:                BANK Cognitive Level:                C/A 10 CFR Part 55 Content:        41.7 / 45.5 to 45.8 Comments:
You have completed the test!
Thursday, March 06, 2014 11:47:02 AM                                                                3
 
Question Number:            53 K/A:      078 A4.01 Ability to manually operate and/or monitor in the control room: Pressure gauges Tier:        2      RO Imp:                RO Exam:          53      Cognitive Level:      Low Group:      1      SRO Imp:      n/a      SRO Exam:          53      Source:                WBN Bank Applicable 10CFR55 Section:            (CFR: 41.7 / 45.5 to 45.8)
Learning Objective:          3-OT-SYS032A, Obj. 16, List the events and their corresponding set points that take place on decreasing control air pressure.
 
==References:==
AOI-10, "Loss of Control Air," Rev. 38 Question: 53 Which ONE of the following completes the sentence below for a lowering control air system pressure?
The setpoint at which the Auxiliary Air Compressors start is __(1)__ psig. If pressure continues to lower to ___(2)___
psig, air to the Reactor Building will automatically isolate.
(1)              (2)
: a.        83                75
: b.        83                70
: c.        79.5              75
: d.        79.5              70 DISTRACTOR ANALYSIS
: a. Incorrect. Plausible since the Auxiliary Air Compressor starts at 83 psig, but the essential and non-essential air systems do not isolate at 75 psig. 75 psig is plausible since this is the required pressure for reopening air to containment valve.
: b. CORRECT. Auxiliary Air Compressor starts at `83 psig, and the essential and non-essential air systems isolate at 70 psig.
: c. Incorrect. Plausible since the auxiliary air system isolates at 79.5 psig, but the essential and non-essential air systems do not isolate at 75 psig. 75 psig is plausible since this is the required pressure for reopening air to containment valve.
: d. Incorrect. Plausible since the auxiliary air system isolates at 79.5 psig, but the essential and non-essential air systems isolate at 70 psig.
 
Approved By                                                                              Procedure    Version J.B. Stanley                        Vogtle Electric Generating Plant                    13710-1      39 Effective Date                                                                          Page Number 08/13/2012                                      SERVICE AIR SYSTEM                              15 of 84 INITIALS 4.2                SYSTEM OPERATION NOTE During normal operation, one or more compressors should be operated in manual with their respective handswitches in Normal After Start. The remaining available compressor(s) should be in Standby with their handswitch(es) in AUTO PTL. If the system load increases such that the base load compressor(s) cannot supply the demand, the Standby compressor(s) starts at 100 psig as sensed at its local controller.
4.2.1              Placing An Air Compressor In Standby 4.2.1.1            Verify the following for the compressor to be placed in standby:
TPCCW (TPCCW or Utility Water for rotary only) is available to the compressor.                                ________
Compressor lube oil level is normal.                        ________
CAUTION To allow the residual voltage restart relays to reset, the compressor should be stopped per Section 4.3.1 or 4.3.2 and allowed to remain in STOP for at least three seconds before placing the handswitch in AUTO PTL position.
4.2.1.2            IF the air compressor is running, shut down compressor per Section 4.3.                                                          ________
4.2.1.3            On QMCB, place the desired compressor handswitch(es) in AUTO PTL position:
COMPRESSOR                      HANDSWITCH 1-2401-C4-501                    1HS-19338                            ________
1-2401-C4-502                    1HS-9383                              ________
1-2401-C4-503                    1HS-9382                              ________
A-2401-C4-504                    1HS-9381                              ________
Printed October 21, 2013 at 16:12
 
Approved By                                                                                    Procedure Number Rev S. E. Prewitt                      Vogtle Electric Generating Plant                            17001-1      31.1 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 01 ON PANEL                Page Number 08/16/2010                                            1A1 ON MCB                                      23 of 48 WINDOW B06 ORIGIN                            SETPOINT INSTR AIR 1-PSL-19414                      70 psig                        EQUIP LO PRESS 1.0                PROBABLE CAUSE
: 1.        Instrument Air Dryer, Prefilter or Afterfilter clogged.
: 2.        System piping leak.
: 3.        System valve misalignment.
: 4.        Loss of all Air Compressors.
2.0                AUTOMATIC ACTIONS NONE 3.0                INITIAL OPERATOR ACTIONS Go To 18028-C, "Loss Of Instrument Air."
4.0                SUBSEQUENT OPERATOR ACTIONS NONE 5.0                COMPENSATORYOPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X3D-BH-R50L, 1X4DB175-2, CX5DT1101-95B Printed October 21, 2013 at 16:28
 
Approved By                                                                                  Procedure Number Rev S. E. Prewitt                      Vogtle Electric Generating Plant                          17001-1      31.1 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 01 ON PANEL            Page Number 08/16/2010                                            1A1 ON MCB                                  33 of 48 WINDOW C06 ORIGIN                            SETPOINT SERVICE AIR 1-PSL-9375                        95 psig                    HDR LO PRESS 1.0                PROBABLE CAUSE
: 1.        Excessive service air demand.
: 2.        Air Compressor trip.
: 3.        System leak.
: 4.        Standby compressor failed to start.
2.0                AUTOMATIC ACTIONS
: 1.        Service Air Dryer Inlet Isolation Valve 1-PV-9375 closes at a service air pressure of 80 psig.
: 2.        Any standby air compressor with its handswitch in AUTO-PTL position will auto start at a discharge pressure of 100 psig decreasing.
3.0                INITIAL OPERATOR ACTIONS NONE Printed January 23, 2014 at 14:38
 
Approved By                                                                                  Procedure Number Rev S. E. Prewitt                      Vogtle Electric Generating Plant                        17001-1      31.1 Date Approved                ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 01 ON PANEL            Page Number 08/16/2010                                            1A1 ON MCB                                  34 of 48 WINDOW C06 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        Check QMCB indications and start a standby Air Compressor if necessary to maintain service air header pressure above 100 psig.
: 2.        Dispatch an operator to check for system leaks or excessive air usage.
: 3.        IF pressure continues to fall and CANNOT be restored, refer to 18028-C, "Loss Of Instrument Air".
: 4.        Refer to 13710-1, "Service Air System" and verify Air Compressors are operating properly.
: 5.        IF equipment failure is indicated, initiate maintenance as required.
CAUTION Procedure 13710-1 Service Air System should be referenced prior to performing the following step if service air has isolated due to low pressure.
: 6.        WHEN service air header pressure is greater than 97 psig as read on 1-PI-19380 on panel PMEC, reset 1-PSL-9375. Switch is located on instrument rack 15 (1-1624-P5-R15) on Turbine Building level 1 near Powdex vessels.
5.0                COMPENSATORYOPERATOR ACTIONS
: 1.        Trend the Service Air System pressure on the Plant Computer.
: 2.        Check the Service Air System pressure greater than or equal to 100 psig once per hour, and initiate the appropriate Subsequent Operator Actions IF pressure is low.
: 3.        Log corrective actions to repair the disabled annunciator or reasons for no action on 10018-C, "Annunciator Control", Figure 2.
: 4.        Log compensatory actions on 10018-C, "Annunciator Control", Figure 5.
END OF SUB-PROCEDURE
 
==REFERENCES:==
1X3D-BH-R50L, 1X4DB175-2, CX5DT1101-95A Printed October 21, 2013 at 16:21
: 1. 086K4.02 001/LOIT/RO/M/F 3.0/3.4/086K4.02/LO-TA-43003///
Initial condition:
            - Unit 1 at 100% reactor power.
Current condition:
            - The running fire water jockey pump trips.
Which one of the following completes the following statement?
The standby fire water jockey pump __(1)__ automatically start as fire header pressure lowers, and the electric fire pump's automatic start setpoint is __(2)__ psig.
__(1)__                                __(2)__
A.                      will                                  95 B.                      will                                  110 C.                  will NOT                                  95 D.                  will NOT                                  110 K/A 086              Fire Protection K4.02            Knowledge of design feature(s) and/or interlock(s) which provide for the following:
                        - Maintenance of fire header pressure.
K/A MATCH ANALYSIS The question tests the candiate's knowledge of fire protection design features by asking if jockey pumps auto start on a trip of a pump and the auto start setpoint for the electric fire pump.
EXPLANATION OF REQUIRED KNOWLEDGE The fire header pressure is normally maintained around 140 psig by operating 1 or 2 jockey pumps, depending on system leakage. Per ELEMENTARY CX3D-BH-F50C, the jockey pumps are manual start and stop only. If a leak in the fire system header or a sprinkler system actuation causes fire header pressure to drop, per SOP 13903-C Thursday, March 06, 2014 12:25:36 PM                                                            1
 
Limitation 2.2.9, the electric fire pump would start at 110 psig, Diesel fire pump #1 at 95 psig, and Diesel fire pump #2 at 85 psig. These three fire pumps have handswitches on the QPCP panel in the main control room that allow for remote starting and run status, but the pumps can only be stopped locally.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. The jockey pumps are manual start and stop only. The remaining pumps in the system are designed to automatically start under the stated condition, and most of the pumps in the plant have automatic start features. It is reasonable to believe that the standby jockey pump would be designed to automatically start to maintain fire header pressure.
The second part is incorrect. The electric fire pump starts at 110 psig, Diesel fire pump #1 at 95 psig, and Diesel fire pump
                                    #2 at 85 psig. However, candidates routinely confuse the fire pump start setpoints.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. The electric fire pump would start at 110 psig.
C. Incorrect. Plausible. The first part is correct. The jockey pumps are manual start and stop only. Therefore, when the running jockey pump trips, the remaining jockey pump must be manually started.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, March 06, 2014 12:25:36 PM                                                                  2
 
Level:                        RO Tier # / Group #              T2 / G2 K/A#                          086K4.01 Importance Rating:            3.0 / 3.4 Technical
 
==Reference:==
SOP 13903-C, Rev 44.0, page 6 ELEMENTARY CX3D-BH-F50C, Rev 3.0 References provided:          None Learning Objective:            LO-PP-43101-05 Discuss the system response to an auto sprinkler actuation or fire hose operation which lowers system header pressure.
LO-TA-43003    Abnormal and Emergency Starting Of A Diesel Fire Pump using 13903-C Question origin:              BANK Cognitive Level:              M/F 10 CFR Part 55 Content:        41.7 Comments:
You have completed the test!
Thursday, March 06, 2014 12:25:36 PM                                                          3
 
Approved By                                                                              Procedure      Version C.H. Williams                      Vogtle Electric Generating Plant                      13903-C        44 Effective Date                                                                            Page Number 07/26/2013                          FIRE PROTECTION SYSTEM OPERATION                              6 of 92 INITIALS 2.2.6              Per DOEJ-SM-C070400401-001, the Portable B.5.b Pump cannot be considered a backup fire suppression system as required by FP LCO 4.3 Cond A or B. The Portable B.5.b Pump is a defense in depth contingency for the interim period between total loss of suppression capability and arrival of Burke County EMA. A Burke County EMA pumper truck is the credited backup.                        ________
2.2.7 Per DOEJ-SM-C070400401-001, all hot work shall be suspended and hourly Fire Watches shall be established for 1A, 1B, 2A, and 2B Cable Spreading Rooms if a backup suppression system has to be established in either FP LCO 4.3 Cond A or B.                                      ________
2.2.8 A Condition Report should be generated anytime two jockey pumps are required running to maintain Fire Protection System pressure.                                  ________
2.2.9              Fire pumps auto start on decreasing header pressure at:
110 psig Electric fire pump starts                            ________
95 psig #1 Diesel fire pump starts                            ________
85 psig #2 Diesel fire pump starts                            ________
2.2.10            Attachment 1 provides guidance for developing troubleshooting steps in the event a low header pressure condition is encountered with no obvious explanation. A troubleshooting plan in accordance with NMP-AD-002 "Problem Solving and Troubleshooting Guidelines" must still be completed using these guidelines. (LVL 3 AI 2009201678)          ________
3.0                PREREQUISITES 3.1      One or two Jockey Pump(s) is/are running to maintain or attempting to maintain the Fire Protection System pressure. ________
3.2      The Fire Pump House No. 1 and/or Fire Pump House No. 2 Heating, Ventilation, and Air Conditioning System (HVAC) has been aligned per 13330-C, "Outside Area Buildings HVAC System."                                                ________
Printed October 21, 2013 at 15:29
 
Manual start only
: 1. 103A2.03 001/LOIT/RO/M/F 3.5/3.8/103A2.03/LO-TA-15003///
Initial conditions:
            -  LOCA has occurred on Unit 1.
            -  19000-C, "Reactor Trip or Safety Injection," is entered.
            -  Verification of Immediate Operator Actions is complete.
            -  Containment pressure is 23.8 psig.
Current conditions:
            - ALB06-D06 CNMT SPRAY ACTUATION is NOT LIT.
            - ALB06-E06 CNMT ISO PHASE A ACTUATION is NOT LIT.
Which one of the following completes the following statement?
The first required action in 19000-C to be performed by the crew is to actuate __(1)__,
and this action will be taken using __(2)__ direction.
__(1)__                                __(2)__
A.                    CIA                                Foldout Page B.                    CIA                              OATC Initial Actions C.            Containment Spray                            Foldout Page D.            Containment Spray                          OATC Initial Actions K/A 103              Containment System A2.03            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.
                        - Phase A and B isolation.
K/A MATCH ANALYSIS The question tests the candidate's ability to predict the impact of high containment pressure on containment systems. The candidate is required to determine if CIA or Containment Spray actuation is the higher priority based on containment pressure, and, based on the priority, determine which procedure direction will accomplish the mitigating actions.
Thursday, March 06, 2014 12:26:58 PM                                                        1
 
EXPLANATION OF REQUIRED KNOWLEDGE Based on a containment pressure of 23.8 psig, both Containment Spray and CIA acutations signals should have been generated. Per 19000-C step 6 & 7, the Foldout Page actions are a higher priority than OATC/UO Initial Actions. EOP/AOP Writer's Guide 10020-C states that the Foldout Page is a continuous action page that is applicable to the entire procedure for which it is included. Therefore, any Foldout Page action is applicable and has higher priority than the OATC/UO Initial Actions.
Containment Spray Actuation is listed on Foldout Page step 3 and OATC Initial Actions step 8. CIA is not on the Foldout Page and is OATC Initial Action step 2. Therefore, with both actuations present, Containment Spray should be manually aligned first using the Foldout Page guidance.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Containment Spray would be manually aligned using the Foldout Page as the first action.
However a candidate not familiar with the contents of the Foldout Page of 19000-C might believe CIA is also contained on the Foldout Page. As such, it would be reasonable to assume that CIA would be a higher priority than Containment Spray to ensure release to the public is minimized. This is the order of the items in the OATC Initial Actions.
The second part is correct. The Containment Spray actuation direction would be utilized from the Foldout Page.
B. Incorrect. Plausible. The both the first and second parts are incorrect. Containment Spray would be manually aligned using the Foldout Page as the first action. However a candidate who does not recognize the Containment Spray acutation setpoint has been exceeded and only believes the CIA actuation is present or who forgets about Containment Spray actuation being on the Foldout Page would believe that CIA would be the first system to align using step 2 of the OATC Initial Actions since CIA is not on the Foldout Page. Containment Spray would be verified at step 8.
C. Correct.                  The first part is correct. Containment Spray would be manually aligned using the Foldout Page as the first action.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. Containment Spray would be manually aligned using the Foldout Page. However, a candidate unfamiliar with 19000-C may recognize both Containment Spray and CIA actuations present but believe Containment Spray is listed first in the OATC Initail Actions and Thursday, March 06, 2014 12:26:58 PM                                                                2
 
believe alignment would be addressed using the body of the procedure instead of the Foldout Page or not remember Containment Spray being on the foldout page.
Level:                          RO Tier # / Group #                T2/G1 K/A#                            103A2.03 Importance Rating:              3.5/3.8 Technical
 
==Reference:==
EOP 19000-C, Rev 37.1, pages 5, 18, 20 & 35 ADMIN 10020-C, Rev 9.0, page 9 ARP 17006-1, Rev 33.1, pages 3, 42, 43, 52 References provided:            None Learning Objective:              LO-LP-37002-01 State how each of the following EOP format elements are used to guide the operator in proper performance of the steps of the procedure.
: e. Foldout page LO-TA-15003          Manually Actuate and Align Containment Spray for Operation per 19000-C LO-TA-28016          Manually actuate CIA / CVI Question origin:                BANK Cognitive Level:                M/F 10 CFR Part 55 Content:          41.5 / 41.7 / 43.5 / 45.3 / 45.13 Comments:
You have completed the test!
Thursday, March 06, 2014 12:26:58 PM                                                                3
 
Approved By                                                                        Procedure    Version M.G. Brill                            Vogtle Electric Generating Plant            19000-C        37.1 Effective Date                                                                      Page Number E-0 REACTOR TRIP OR SAFETY INJECTION 7-5-13                                                                                    5 of 35 ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED 4
: 4.      Check if SI is actuated:                  4. Check if SI is required:
IF one or more of the following Any SI annunciator - LIT.                conditions has occurred:
SI ACTUATED BPLB window -                    PRZR pressure less than or LIT.                                          equal to 1870 psig.
Go to Step 6.
Steam line pressure less than or equal to 585 psig.
Containment pressure greater than or equal to 3.8 psig.
Automatic alignment of ECCS equipment to injection phase.
THEN actuate SI and go to Step 6.
SUBSEQUENT OPERATOR ACTIONS 5
: 5.      Perform the following to limit RCS        5.
cooldown:
5.a
: a. Check NR level in at least one              a. Maintain AFW flow greater than SG greater than 10%.                          570 gpm and go to 19001-C, ES-0.1 REACTOR TRIP RESPONSE.
5.b
: b. Reduce AFW flow.
5.c
: c. Go to 19001-C, ES-0.1 REACTOR TRIP RESPONSE.
6
: 6.      Initiate the Foldout Page.                6.
7
: 7.      Perform the following:                    7.
OATC Initial Actions Page.
UO Initial Actions Page.
 
S Printed January 23, 2014 at 10:21
 
Approved By                                                                          Procedure      Version M.G. Brill                            Vogtle Electric Generating Plant              19000-C        37.1 Effective Date                                                                        Page Number E-0 REACTOR TRIP OR SAFETY INJECTION 7-5-13                                                                                    18 of 35 OATC INITIAL ACTIONS                    Sheet 2 of 4 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 1
: 1.      Check both trains of ECCS                  1. Actuate SI.
equipment - ALIGNING FOR INJECTION PHASE:
MLB indication.
2
: 2.      Check Containment Isolation Phase          2. Actuate CIA.
A - ACTUATED:
IF valves do NOT close, CIA MLB indication.                      THEN close valves.
3
: 3.      Check ECCS Pumps and NCP                  3.
status:
3.a
: a. CCPs - RUNNING.                              a. Perform the following for available CCP(s):
3.a.1)
: 1)    Place alternate miniflow valve handswitch in ENABLE PTL:
HS-8508A HS-8508B 2.2)
: 2)    Start CCP(s).
3.b
: b. SI Pumps - RUNNING.                          b. Start Pumps.
3.c
: c. RHR Pumps - RUNNING.                          c. Start Pumps.
3.d
: d. NCP - TRIPPED.                                d. Stop the NCP.
4
: 4.      Verify CCW Pumps - ONLY TWO                4.
RUNNING EACH TRAIN.
 
S Printed January 23, 2014 at 10:27      4
 
Approved By                                                                          Procedure      Version M.G. Brill                            Vogtle Electric Generating Plant              19000-C        37.1 Effective Date                                                                        Page Number E-0 REACTOR TRIP OR SAFETY INJECTION 7-5-13                                                                                    20 of 35 OATC INITIAL ACTIONS                    Sheet 4 of 4 ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED 8
: 8.      Check Containment pressure -              8. Verify the following:
REMAINED LESS THAN 21.5 PSIG.                                                              2.a
: a. Containment Spray actuated.
2.b
: b. Containment Spray Pump discharge valves open.
2.c
: c. Containment Spray Pumps running.
9
: 9.      Check ECCS flows:                          9.
9.a
: a. BIT flow.                                    a. Align Valves using ATTACHMENT B.
9.b
: b. RCS pressure - LESS THAN                    b. Go to Step 10.
1625 PSIG.
9.c
: c. SI Pump flow.                                c. Align Valves using ATTACHMENT C.
9.d
: d. RCS pressure - LESS THAN                    d. Go to Step 10.
300 PSIG.
9.e
: e. RHR Pump flow.                              e. Align Valves using ATTACHMENT D.
10
: 10.      Check ECCS Valve alignment -              10. Align valves using ATTACHMENT B, PROPER INJECTION LINEUP                          ATTACHMENT C, and INDICATED ON MLBs.                              ATTACHMENT D as necessary.
 
S Printed January 23, 2014 at 10:27      4
 
Approved By                                                                              Procedure    Version M.G. Brill                            Vogtle Electric Generating Plant                    19000-C        37.1 Effective Date                                                                            Page Number E-0 REACTOR TRIP OR SAFETY INJECTION 7-5-13                                                                                          35 of 35 FOLDOUT
: 1.      RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below occur:
: a. CCPs or SI pumps - AT LEAST ONE RUNNING.
: b. RCP Trip Parameter - RCS PRESSURE LESS THAN 1375 PSIG.
: 2.      AFW SUPPLY SWITCHOVER CRITERION Switch to alternate CST by initiating 13610, AUXILIARY FEEDWATER SYSTEM when CST level lowers to less than 15%.
: 3.      CNMT SPRAY ACTUATION CRITERIA Verify the following when CNMT pressure is greater than or equal to 21.5 psig:
: a. CNMT Spray actuated.
: b. CNMT Spray Pump discharge valves open.
: c. CNMT Spray Pumps running.
: 4.      Monitor SPENT FUEL POOL COOLING conditions:
Verify annunciators ALB05-A6, SPENT FUEL PIT HI TEMP and ALB05-E2, SPENT FUEL PIT LOW LEVEL are both clear.
IF alarms are NOT CLEAR, THEN initiate 18030-C, LOSS OF SPENT FUEL POOL LEVEL OR COOLING.
IF SPENT FUEL POOL LEVEL OR COOLING ALARMS are not available, THEN dispatch operator to start 2 HR interval local checking that level > 217 ft and temperature < 130&deg;F.
IF either parameter is exceeded, THEN initiate 18030-C, LOSS OF SPENT FUEL POOL LEVEL OR COOLING IF applicable, Using PTDB TAB 26, determine time to restore SFP LEVEL OR COOLING is < time to reach 200&deg;F in Spent Fuel Pool.
IF NOT initiate 18030-C, LOSS OF SPENT FUEL POOL LEVEL OR COOLING
 
S Printed January 23, 2014 at 10:22      1
 
Approved By                                                                                    Procedure    Version J.B. Stanley                          Vogtle Electric Generating Plant                          17006-1      33.1 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 06 ON PANEL                Page Number 07/23/2013                                                1A2 ON MCB                                    3 of 59 (1)                  (2)            (3)              (4)          (5)            (6)
A      RHR PMP 1                            ACCUM TANK 1    ACCUM TANK 1 ACCUM TANK 1  CNMT HI-1 DISCH HI PRESS                      HI/LO LEVEL      HI/LO PRESS  ISO VLV 8808A  PRESS ALERT NOT FULLY OPEN ADVERSE CNMT B      RHR PMP-2            RCS MIDLOOP    ACCUM TANK 2      ACCUM TANK 2 ACCUM TANK 2  CNMT HI-2 DISCH HI PRESS      HI LEVEL        HI/LO LEVEL      HI/LO PRESS  ISO VLV 8808B  PRESS ALERT NOT FULLY OPEN C      RHR PMP              RCS MIDLOOP    ACCUM TANK 3      ACCUM TANK 3 ACCUM TANK 3  CNMT HI-3 OVERLOAD TRIP        LO LEVEL        HI/LO LEVEL      HI/LO PRESS  ISO VLV 8808C  PRESS ALERT NOT FULLY OPEN RHR HL D      VLV OPEN                            ACCUM TANK 4      ACCUM TANK 4 ACCUM TANK 4  CNMT SPRAY AND HI RCS                          HI/LO LEVEL      HI/LO PRESS  ISO VLV 8808D  ACTUATION PRESS                                                              NOT FULLY OPEN CNMT VENT ISO E      ACTUATION                            RWST TO SI PMP    RWST        RWST          CNMT ISO ISO VLV 8806      LO LEVEL    EMPTY LEVEL    PHASE A NOT FULLY OPEN                                ACTUATION F      CSFST                                RWST              RWST        RWST          SI PMP TROUBLE                              HI LEVEL          LO-LO LEVEL  LO-LO LEVEL    OVERLOAD TRIP ALERT Printed January 23, 2014 at 11:02
 
Approved By                                                                              Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                      17006-1      33.1 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 06 ON PANEL        Page Number 07/23/2013                                            1A2 ON MCB                              42 of 59 WINDOW D06 ORIGIN                            SETPOINT CNMT SPRAY 2 out of 4                          21.5 psig              ACTUATION 1-PT-0934A                      (2/4 channels) 1-PT-0935A                      (relay K643) 1-PT-0936A 1-PT-0937A or both                          Not Applicable 1-HS-40010 1-HS-40011 or both                          Not Applicable 1-HS-40004 1-HS-40005 1.0                PROBABLE CAUSE
: 1.        Manual actuation of the Containment Spray System.
: 2.        Containment HI-3 setpoint reached on 2 or more Containment pressure channels.
2.0                AUTOMATIC ACTIONS
: 1.        Containment Spray Pumps start.
: 2.        Containment Spray Isolation Valves 1-HV-9001A and 1-HV-9001B open.
3.0                INITIAL OPERATOR ACTIONS NOTE Actions for a containment spray actuation are contained in Emergency Operating Procedures.
Printed January 23, 2014 at 11:01
 
Approved By                                                                                Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                        17006-1      33.1 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 06 ON PANEL            Page Number 07/23/2013                                            1A2 ON MCB                                  43 of 59 WINDOW D06 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS IF a spurious Containment Spray actuation has occurred:
: 1.        Reset CS signal.
: 2.        Stop CS pumps.
: 3.        Shut CNMT SPRAY ISO VLVS:
: a. 1-HV-9001A
: b. 1-HV-9001B
: 4.        IF Containment Spray is actuated and terminated prior to recirculation, a controlled cleanup and inspection of equipment in containment should begin within five days of the event 5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB131, 1X6AA02-232, PLS, 1X6AU01-178, 1X6AX01-322, 1X6AX01-409, 1X6AX01428, 1X6AV01-242 Printed January 23, 2014 at 11:01
 
Approved By                                                                                Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                        17006-1      33.1 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 06 ON PANEL            Page Number 07/23/2013                                            1A2 ON MCB                                  52 of 59 WINDOW E06 ORIGIN                            SETPOINT CNMT ISO Safety Injection                  Not Applicable            PHASE A OR                                                        ACTUATION 1-HS-40006 or                      Not Applicable 1-HS-40009 1.0                PROBABLE CAUSE
: 1.        Safety Injection Actuation          >3.8 psig
: 2.        Manual Actuation.
2.0                AUTOMATIC ACTIONS Initiates Containment Phase A Isolation.
3.0                INITIAL OPERATOR ACTIONS If a Safety Injection has occurred, initiate 19000-C, "E-O Reactor Trip Or Safety Injection."
4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        IF an inadvertent Phase A Isolation has occurred in Modes 1, 2 or 3, then perform the following:
: a. Reset Phase A by placing both 1-HS-40120 and 1-HS-40122 to RESET position,
: b. Open Instrument Air to containment 1-HV-9378 using both 1-HS-9378A and 1-HS-9378B,
: c. Restore normal letdown/charging per 13006-1, "Chemical And Volume Control System,"
: d. Open RCP Seal Return 1-HV-8100 and 1-HV-8112 using 1-HS-8100 and 1-HS-8112,
: e. Reset Containment Ventilation Isolation by placing both 1-HS-40121 and 1-HS-40123 to RESET position.
: 2.        Complete the applicable portions of 11886-1, "Recovery From ESF Actuations," for CIA and CVI.
Printed January 23, 2014 at 10:58
 
Approved By                                                                                  Procedure Number Rev C.S. WALDRUP                    Vogtle Electric Generating Plant                            10020-C        9 Date Approved                                                                                Page Number 01/26/2011                          EOP AND AOP RULES OF USAGE                                        9 of 27 3.5                CONTINUOUS ACTION STEPS 3.5.1              Continuous Action steps are marked in EOPs and AOPs by bolding the AER step and by use of an asterisk (*) immediately preceding the AER step number (or RNO step number if there is one). Only the high level step number is marked with the asterisk (*) indicating that there is a continuous action in the AER or RNO text, the associated Note, or Caution.
3.5.1.1            Continuous action steps that are skipped by procedure step transitions, are still in affect and should be applied.
3.5.1.2            Continuous Action Pages will be posted in a separate book in the control room for EOPS and AOPs that have continuous actions and updated as new revisions are issued.
3.5.2              NOTES and CAUTIONS may contain continuous action steps if written passively.
3.5.3              Logical terms such as WHEN, IF, THEN are considered continuous operator actions.
3.5.4              Action verbs such as control, limit, match, maintain, and monitor are continuous action verbs. See TABLE 2 for definitions of these verbs.
3.5.5              NOTEs and CAUTIONs prior to step 1 of EOPs and AOPs apply to the entire procedure.
3.5.6              Foldout Page is a continuous action page that is applicable to the entire procedure for which it is included.
3.5.6.1            Foldout Pages will be posted in a separate book (the same book as the continuous actions) in the control room for EOPS that have foldout pages.
3.5.6.2            The foldout page should be copied on the back of each page in the EOP except for the foldout page itself.
3.5.7              Continuous action steps are no longer applicable when exiting a procedure unless restated in the next procedure entered, or they direct return to the procedure in which they are stated, e.g. 19013-C where you are directed to return and perform Containment Spray Pump suction swap once conditions are met.
Printed March 3, 2014 at 15:18
: 1. 103K1.08 001/LOIT/RO/M/F 3.6/3.8/103K1.08/LO-PP-28103-06///
Given the following:
            - A LOCA with SI actuation occurred on Unit 1.
            - SI can NOT be reset.
Which one of the following completes the following statement?
The CIA signal __(1)__ be reset, and the CVI signal __(2)__ be reset.
__(1)__                                __(2)__
A.                    can                                    can B.                    can                                  can NOT C.                  can NOT                                    can D.                  can NOT                                can NOT K/A 103              Containment System K1.08            Knowledge of the physical connections and/or cause effect relationships between the containment system and the following systems:
                        - SIS, including action of safety injection reset K/A MATCH ANALYSIS Question requires the candidate to determine which interlocks are required to be cleared or reset in order to reset CIA and CVI, and is based on RO knowledge level requirements. In both cases the candidate must determine if the logic device is "retentive memory" or "retentive memory with actuation block" when making a determination of system response.
EXPLANATION OF REQUIRED KNOWLEDGE In the EOP network, the reset of either CIA or CVI signal is always preceeded with a step that resets SI. (See 19012-C step 2 and 3, for example.) The RNO for resetting SI directs the operator to perform an Attachment that overrides SSPS and manually rolls the slave relays and effectively resets SI. Per LOGIC 1X6AA02-00232, both CVI Thursday, March 06, 2014 12:28:40 PM                                                        1
 
and CIA can be reset with the SI actuation signal present. Once blocked, the block will hold and prevent any subsequent actuation signal from being processed. Therefore, the inability to reset SI will not prevent the reset of either CIA or CVI and does not impede the progress of the procedure. The inability to reset SI may complicate other recovery actions later and therefore it is desirable to reset SI using the Attachment.
(Note: This RNO is a relatively new addtion to the EOPs based on an industry event in which an inadvertant SI was encountered and could not be reset, resulting in challenges to the operating crew. Even without SI reset, the plant is capable of being maintained. However, knowledge space decisions are required. This RNO puts operation in rule space and allows normal and expected responses to EOP steps.)
ANSWER / DISTRACTOR ANALYSIS A. Correct.                  The first part is correct. The CIA reset logic device is a "retentive memory with actuation block" which is designed to allow the actuation signal to be overriden. This knowledge is important because procedures are designed in general to reset the initiating signal (such as SI) first. However, if the ability to reset SI has failed, the operator must know that this does not impede the ability to perform reset of CIA.
The second part is correct. The CVI reset logic device is a "retentive memory with actuation block" which is designed to allow the actuation signal to be overriden. This knowledge is important because procedures are designed in general to reset the initiating signal (such as SI) first. However, if the ability the reset SI has failed, the operator must know that this does not impede the ability to reset of CVI.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. The CVI reset logic device is a "retentive memory with actuation block" which is designed to allow the actuation signal to be overriden. However, there is no direction within the E-1/ES-1 series of procedures that directs the resetting of CVI. Therefore, it is reasonable for a candidate to conclude that "retentive memory with actuation block" would not be necessary for this actuation signal and SI must be reset prior to CVI being reset.
C. IncorrectPlausible.      The first part is incorrect. The CIA reset logic device is a, "retentive memory with actuation block" which is designed to allow the actuation signal to be overriden. However the candidate may determine the CIA actuation signal uses the standard "retentive memory" which has no actuation block capability and SI must be reset first. This logic would be erroneously re-enforced by the procedural sequence and RNO as described above in the Explanation of Required Knowledge.
The second part is correct. See the second part of choice A Thursday, March 06, 2014 12:28:41 PM                                                                  2
 
above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Level:                        RO Tier # / Group #              T2 / G1 K/A#                          103K1.08 Importance Rating:            3.6 / 3.8 Technical
 
==Reference:==
LOGIC 1X6AA02-00232, Rev EOP 19012-C, Rev 33.3, pages 3 & 39 References provided:          None Learning Objective:            LO-PP-28103-06 Determine when ESF actuation signal can be reset and describe actions required to reset the signal.
LO-PP-28103-07 Discuss SI reset to include:
: a. Time delay
: b. SI reset with P-4
: c. SI reset without P-4
: d. Auto and Manual actuation capabilities following reset Question origin:              NEW Cognitive Level:              M/F 10 CFR Part 55 Content:        41.2 to 41.9 / 45.7 to 45.8 Comments:
You have completed the test!
Thursday, March 06, 2014 12:28:41 PM                                                                3
 
Approved By                                                                                  Procedure    Version J. B. Stanley                          Vogtle Electric Generating Plant                      19012-C        33.3 Effective Date                                                                                Page Number ES - 1.2 POST-LOCA COOLDOWN AND 05/01/2013                                    DEPRESSURIZATION                                      3 of 43 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 1
: 1.      Initiate the following:                          1.
Continuous Actions and Foldout Page.
Critical Safety Function Status Trees per 19200-C, F-O CRITICAL SAFETY FUNCTION STATUS TREE.
CAUTION If offsite power is lost after SI reset, action is required to restart the following ESF equipment if plant conditions require their operation:
RHR Pumps SI Pumps Post-LOCA Cavity Purge Units Containment Coolers in low speed (Started in high speed on a UV signal).
ESF Chilled Water Pumps (If CRI is reset).
2
: 2.      Reset SI.                                          2. IF SI will NOT reset, THEN initiate ATTACHMENT D.
CAUTION Repositioning Phase A Isolation Valves may cause radiation problems throughout the plant.
3
: 3.      Reset Containment Isolation                        3.
Phase A.
 
S Printed March 4, 2014 at 08:55
 
Approved By                                                                              Procedure      Version J. B. Stanley                          Vogtle Electric Generating Plant                  19012-C        33.3 Effective Date                                                                            Page Number ES - 1.2 POST-LOCA COOLDOWN AND 05/01/2013                                    DEPRESSURIZATION                                39 of 43 ATTACHMENT D                            Sheet 1 of 1 RESPONSE TO INADVERTENT SI AND INABILITY TO RESET OR BLOCK SI 1
: 1.      Identify the affected train.      Circle:      A Train    B Train NOTE De-energizing the two 48 VDC power supplies to a train of SSPS will result in the following:
General Warning ALB05-E06 or ALB05-F06 will illuminate Undervoltage Driver output de-energizes Reactor Trip condition (Reactor Trip Breaker OPEN) on the affected train (already initiated from the Turbine Trip) 48 VDC is removed from all master relays 2
: 2.      At the affected train SSPS Logic Cabinet, de-energize both 48 VDC power supplies (Located in the upper 2 sections) by placing the ON/OFF switch to the OFF position.
3
: 3.      At the affected train Safeguards Test Cabinet (STC) #1, reset SSPS Slave Relays by momentarily turning TEST RESET SWITCH S-821 to the RESET position.
4
: 4.      At the affected train Safeguards Test Cabinet (STC) #2, reset SSPS Slave Relays by momentarily turning TEST RESET SWITCH S-921 to the RESET position.
5
: 5.      At the affected train, locate and open the Output Cabinet and place the MODE SELECTOR Switch in the TEST position and check the OPERATE lamp NOT lit.
6
: 6.      Notify I&C to investigate the affected train SSPS to determine the source of the SI signal.
 
END OF ATTACHMENT D Printed March 4, 2014 at 08:55
 
Retentive Memory Logic Retentive Memory Logic
: 1. G2.1.14 001/LOIT/RO/M/F 2.7/4.1/G2.1.14/LO-LP-05110-01//HL-15R/
Which one of the following completes the following statement?
Plant announcements are required per UOP guidance for reactor __(1)__, and per EOP guidance when realigning ECCS for __(2)__ leg recirculation.
__(1)__                                      __(2)__
A.                  startups                                      hot B.                  startups                                      cold C.                shutdowns                                        hot D.                shutdowns                                      cold K/A G2.1.14          Generic Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes,etc.
K/A MATCH ANALYSIS The question requires the student to correctly identify which plant evolutions require a Plant PA anouncement (notifying internal organizations).
EXPLANATION OF REQUIRED KNOWLEDGE Per UOP 12003-C step 4.2.12, announcement of the reactor startup to plant personnel is required prior to commencement of rod withdrawal or dilution to criticality in steps 4.2.13 and 4.2.14. In contrast, UOP 12005-C does not contain any direction to make page announcements during a reactor shutdown.
Per EOP 19013-C step 5, operators are directed to make a page announcement to clear personnel from the Auxiliary Building prior to initiating Cold Leg Recirc. In contrast, EOP 19014-C does not contain any diretion to make page announcements prior to transition to Hot Leg Recirc.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. Per UOP 12003-C step 4.2.12, announcement of the reactor startup to plant personnel is required prior to commencement of rod withdrawal or dilution to criticality in steps 4.2.13 and 4.2.14.
The second part is incorrect. EOP 19014-C does not contain Thursday, March 06, 2014 12:33:55 PM                                                              1
 
any direction to make page announcements prior to transition to Hot Leg Recirc. However, EOP 19013-C step 5 directs operators to make a page announcement to clear personnel from the Auxiliary Building prior to initiating Cold Leg Recirc. It is reasonable for a candidate to confuse the two procedures and believe 19013-C is for Hot Leg Recirc.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. EOP 19013-C step 5 directs operators to make a page announcement to clear personnel form the Auxiliary Building prior to initiating Cold Leg Recirc.
C. Incorrect. Plausible. The first part is incorrect. UOP 12005-C does not contain any direction to make page announcements during a reactor shutdown. However, candidates are trained to make a page announcement of an unplanned Reactor Trip and/or Safety Injection. This is not a procedural requirement, but a good practice. It is reasonable for a candidate to assume that since an unplanned Reactor Trip is announced, any reactor trip is required to be announced.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, March 06, 2014 12:33:55 PM                                                                  2
 
Level:                        RO Tier # / Group #              T3 K/A#                          G2.1.14 Importance Rating:            3.1 / 3.1 Technical
 
==Reference:==
UOP 12003-C, Rev 51.2, pages 25 thru 27 UOP 12005-C, Rev 28.0 EOP 19013-C, Rev 29.2, page 5 EOP 19014-C, Rev 15.2 References provided:          None Learning Objective:            LO-LP-05110-01 Describe the functions of the following and explain how and when each of the following are used:
: a. Telephone/Page System LO-TA-13009        Manually align ECCS for Cold Leg Recirculation Phase using EOP 19013-C LO-TA-13012        Manually align ECCS for Hot Leg Recirculation Phase using EOP 19014-C LO-TA-61001        Reactor Startup using 12003-C LO-TA-61003        Reactor Shutdown to Hot Standby using 12005-C Question origin:              BANK Cognitive Level:              M/F 10 CFR Part 55 Content:        41.10 / 43.5 / 45.12 Comments:
You have completed the test!
Thursday, March 06, 2014 12:33:55 PM                                                          3
 
Approved By                                                                        Procedure    Version J.B. STANLEY                          Vogtle Electric Generating Plant            19013-C      29.2 Effective Date                                                                      Page Number ES-1.3 TRANSFER TO COLD LEG 7/25/12                                      RECIRCULATION                              5 of 20 ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED 3
: 3.      Initiate ATTACHMENT A to align              3.
ECCS Pumps to the Cold Leg Recirculation flowpath and continue with Step 4.
4
: 4.      Notify Health Physics that radiation        4.
levels in the Auxiliary Building will change when Cold Leg Recirculation is established.
5
: 5.      Make a page announcement to clear            5.
personnel from the Auxiliary Building prior to initiating Cold Leg Recirculation.
6
: 6.      Initiate Continuous Actions Page.            6.
7
        *7.      Check RWST level - GREATER                  *7. Stop any ECCS Pumps taking THAN 8%.                                          suction from the RWST.
8
        *8.      Check if SI pumps should be                  8.
stopped.
8.a
: a. RCS pressure - GREATER                        a. IF RCS pressure rises to THAN 1625 PSIG.                                greater than 1625 psig, THEN stop SI Pumps.
Go To Step 9.
8.b
: b. Stop SI Pumps.                                b.
9
: 9.      Check ATTACHMENT A -                        9. Do NOT continue with this COMPLETE.                                        procedure until ATTACHMENT A has been COMPLETED.
 
S Printed March 4, 2014 at 11:46
 
Approved By                                                                                Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                        12003-C 51.2 Effective Date                                                                              Page Number 02/14/2013                      REACTOR STARTUP (MODE 3 TO MODE 2)                                25 of 42 INITIALS NOTE Step 4.2.11 is not applicable if the conditions of Step 4.2.10 have been applied.
4.2.11          Initiate an Inverse Count Rate Ratio (ICRR) plot per Data Sheet
: 1. (1987213952, 1988214388, 1988214835, 1989215543)                        ________
CAUTIONS All pertinent indications should be monitored during approach to criticality such as flux level, SUR, recorders, and count rate instrumentation.(19843001331, 1990318409)
Conservative action should be taken, (i.e., stop startup and insert control rods), whenever unexpected situations arise, with respect to reactivity, criticality, power level, or any other anomalous behavior of the reactor core.
Activities should be avoided during reactor startup that could distract operators and supervisors involved with the startup such as shift turnovers and surveillance testing. (1990318409) 4.2.12          Announce the reactor startup to plant personnel.                            ________
Printed November 14, 2013 at 13:58
 
Approved By                                                                                  Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                          12003-C 51.2 Effective Date                                                                                Page Number 02/14/2013                      REACTOR STARTUP (MODE 3 TO MODE 2)                                  26 of 42 INITIALS NOTES IF the startup has been delayed since completing Step 4.2.9, repeat Step 4.2.9 prior to commencing the startup per Step 4.2.13.
IF Reactor startup is suspended for shift turnover, Step 4.2.13 should be reperformed, prior to proceeding.
IF this startup is a dilution to criticality for LPPT, Step 4.2.13 should be performed in conjunction with LPPT-GAE/GBE-01.
CAUTIONS Criticality SHALL be anticipated any time positive reactivity is being added, including but NOT limited to when control rods are being withdrawn or RCS dilution in progress.
A sustained SUR of 1.0 dpm should not be exceeded.
During approach to criticality, two positive reactivity additions will NOT be performed simultaneously.
4.2.13          Commence Rod Withdrawal:
: a.        Verify shutdown banks withdrawn greater than or equal to the insertion limit specified in the COLR. (TS 3.1.5)
(1987212898)                                                      ________
: b.        Verify Rod Bank Selector Switch in MANUAL.                        ________
________
IV
: c.        Withdraw rods in 50 step increments, or as directed by the Reactor Engineer and approved by the SS.                          ________
Printed November 14, 2013 at 13:58
 
Approved By                                                                                Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                        12003-C 51.2 Effective Date                                                                            Page Number 02/14/2013                      REACTOR STARTUP (MODE 3 TO MODE 2)                                27 of 42 INITIALS NOTE For startups following a refueling outage where initial criticality will be achieved by dilution to critical for Low Power Physics testing, Mode 2 shall be declared when dilution commences.
4.2.14          WHEN Control Bank withdrawal commences, OR dilution to Criticality for LPPT commences at Step 4.2.21.b, whichever is applicable, perform the following (1984300231, 19853003137):
: a.        Log Mode 2 entry into the Control Room Log Record Time: ________                                          ________
: b.        Update IPC to reflect Mode as follows:
(1)    On Top Right of IPC Screen click on MODE.                ________
(2)    Click on the desired set mode button.                    ________
(3)    Verify current mode changes to the desired mode.        ________
4.2.15          WHEN Control Bank A reaches 12 steps as indicated by DRPI, verify the following annunciators reset:
: a.        ROD AT BOTTOM (ALB10E05),                                      ________
: b.        TWO OR MORE RODS AT BOTTOM (ALB10F05).                          ________
4.2.16          WHEN Control Bank A reaches 115 steps, verify Control Bank B begins withdrawing. (TS SR 3.1.6.3) (1995330520)                          ________
4.2.17          WHEN Control Bank B reaches 115 steps, verify Control Bank C begins withdrawing. (TS SR 3.1.6.3) (1995330520)                          ________
Printed November 14, 2013 at 14:09
: 1. G2.1.28 001/LOIT/RO/M/F 4.1/4.1/G2.1.28/LO-PP-23101-09///
Initial condition:
            - Unit 1 is at 100% reactor power.
Current conditions:
            - ARE-2532A, FHB - Effluent Radiogas, fails high.
            - ALB05-C03 HIGH RADIATION ALARM is received.
Which one of the following completes the following statement?
When ALB05-C03 was received, __(1)__ FHB Post Accident Filter Unit(s) automatically started, and when the FHB Isolation Reset handswitch is taken to RESET, the white light illuminating alerts the operator that the actuation signal __(2)__ present.
__(1)__                                __(2)__
A.                ONLY one                                      is B.                ONLY one                                  is NOT C.                    BOTH                                      is D.                    BOTH                                    is NOT K/A G2.1.28          Knowledge of the purpose and function of major system components and controls.
K/A MATCH ANALYSIS The KA addresses the purpose of major components and controls. The question has identified the Fuel Handling Building (FHB) HVAC system as the major component and the white handswitch lights as a system function designed to allow the operator to analyze system response. In addition, the candidate is given a specific event and must determine system response and control indications in order to make a correct decision.
EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17102-1, ARE-2532A in HIGH alarm would switch the Normal Fuel Handling Building Ventilation to Accident Mode Ventilation. Per SOP 13320-1 step 4.2.2, when a FHB actuation is received from either train, BOTH FHB Post-Accident Filter Fans start.
Thursday, March 06, 2014 12:35:52 PM                                                        1
 
Per SOP 13320-1 Precaution 2.1.4, if a FHB Isolation signal is still present and either train's hand switch is taken to RESET, then that train's isolation logic is rendered inoperable and the corresponding WHITE light will be LIT on AHS-2532B/2533B.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per SOP 13320-1 step 4.2.2, when a FHB actuation is received for either train, BOTH FHB Post-Accidnet Filter Fans start. However, the candidate may determine that since only the Train A powered radiation monitor went into alarm only the Train A unit will automatically start.
The second part is correct. Per SOP 13320-1 Precaution 2.1.4, if a FHB Isolation signal is still present and either train's hand switch is takent to RESET, then that train's isolation logic is rendered inoperable and the corresponding WHITE light will be LIT on AHS-2532B/2533B.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per SOP 13320-1 Precaution 2.1.4, if a FHB Isolation signal is still present and either train's hand switch is takent to RESET, then that train's isolation logic is rendered inoperable and the corresponding WHITE light will be LIT on AHS-2532B/2533B. However, the candidate may determine that the white light indicates the signal has been reset and manual alignment is allowed because the signal is no longer present.
C. Correct.                  The first part is correct. Per SOP 13320-1 step 4.2.2, when a FHB actuation is received for either train, BOTH FHB Post-Accident Filter Fans start.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Thursday, March 06, 2014 12:35:52 PM                                                                      2
 
Level:                        RO Tier # / Group #              T3 K/A#                          G2.1.28 Importance Rating:            4.1 / 4.1 Technical
 
==Reference:==
SOP 13320-C, Rev 33.1, pages 4 & 16 ARP 17102-1, Rev 20.2, page 25 References provided:          None Learning Objective:            LO-PP-23101-09 Explain how the Fuel Handling Building HVAC System responds to a Fuel Handling Building Isolation Signal.
LO-PP-23101-07 Describe the Fuel Handling Building HVAC System flow path for both Normal and Post Accident Conditions.
Question origin:              MODIFIED - HL18 NRC # 072K1.03 Cognitive Level:              M/F 10 CFR Part 55 Content:        41.7 / 41.13 Comments:
You have completed the test!
Thursday, March 06, 2014 12:35:52 PM                                                        3
: 1. 072K1.03 001/2/2/ARM - FHBI/C/A - 3.6/3.7/MOD - HL-15R AUDIT/HL-18 NRC/RO/SRO/KAJ A dropped spent fuel assembly in the Unit 1 Spent Fuel Pool has resulted in the following radiation monitor alarms:
RE-0008, FHB Area Monitor, indicates HIGH.
          - A-RE-2532A(B) and A-RE-2533A(B), FHB Effluent Monitors, indicate ALERT.
          - The crew is implementing 18006-C, "Fuel Handling Event".
For the given conditions, which ONE of the following completes the following statement?
1-RE-0008 ___(1)___ provide audible and visual indications of the alarm in the Unit 1 SFP area, and the FHB Post-Accident Filtration Units ___(2)___ automatically start.
A. (1) will (2) will B. (1) will (2) will NOT C. (1) will NOT (2) will D. (1) will NOT (2) will NOT Original Question Tuesday, March 04, 2014 2:09:39 PM                                                          1
 
Approved By                                                                                Procedure    Version P. H. Burwinkel                  Vogtle Electric Generating Plant                        13320-C      33.1 Effective Date                                                                              Page Number 08/13/2012                        FUEL HANDLING BUILDING HVAC SYSTEM                                4 of 55 INITIALS 2.0                PRECAUTIONS AND LIMITATIONS 2.1                PRECAUTIONS 2.1.1              The Emergency Filtration Units automatically start on high radiation in the Exhaust Header.                                                    ________
2.1.2              Outside Air Intake Low Temperature Cutout Control Switch, A-TSL-2520 for A-1541-A7-001 or A-TSL-2521 for A-1541-A7-002 will trip the FHB Normal Supply Fan if it senses a temperature of less than or equal to 37F. The switch must be manually reset, but will not reset until sensing bulb temperature increases to 49F. The reset switches are located on the side of the corresponding HVAC supply unit.                                                              ________
2.1.3              One train of FHB Post Accident Filter Unit should be placed in service if the normal supply units will not operate due to low outside temperature as discussed in Precaution Step 2.1.2.                        ________
2.1.4              If a FHB Isolation Signal is still present and either trains hand switch is taken to reset, then that Trains Isolation logic is rendered inoperable. A corresponding White Light will be LIT on AHS-2532B(AHS-2533B), and a corresponding alarm is received on the Unit One and Unit Two SSMP Panels and bring in annunciators on Both Unit One and Unit Two (1/2ALB04-E01(E02).            ________
2.1.5              Sudden changes in the HVAC configuration such as a CVI or FHB Isolation can cause rapid changes in Spent Fuel Pool (SFP) and Refueling Cavity water level when the two are interconnected. SFP and Refueling Cavity levels should be checked following any such change to prevent overflow. (1992224401) _______
2.2                LIMITATIONS 2.2.1              Two independent FHB Post Accident Exhaust Systems shall be operable whenever irradiated fuel is in either Spent Fuel Pool. (TR 13.9.5)                                                                    ________
2.2.2              The Fuel Handling building Post Accident Ventilation Actuation Instrumentation shall be OPERABLE. (TR 13.3.6, Table 13.3.6-1)            ________
Printed October 29, 2013 at 9:50
 
Approved By                                                                        Procedure    Version P. H. Burwinkel                  Vogtle Electric Generating Plant                13320-C      33.1 Effective Date                                                                      Page Number 08/13/2012                        FUEL HANDLING BUILDING HVAC SYSTEM                      16 of 55 INITIALS CAUTION The Train B Post Accident Filter Unit and the normal exhaust HVAC system discharge to the same exhaust header. They should not be aligned to discharge to their common exhaust stack at the same time.
4.2.1              Actuate FHB ISOLATION by momentarily placing either trains handswitch to ACTUATE position.
FHB ISOLATION MANUAL ACTUATION                        AHS-2532A (A54)      ________
FHB ISOLATION MANUAL ACTUATION                        AHS-2533A (A55)      ________
4.2.2              Verify FHB Isolation:
: a.        FHB Isolation actuated:
Red Light at FHB ISOLATION MANUAL ACTUATION, AHS-2532A (A54) LIT                    ________
Red Light at FHB ISOLATION MANUAL ACTUATION, AHS-2533A (A55) LIT                    ________
: b.        POST ACCIDENT FILT/EXH FANS:
Train A: A-1542-N7-001 (C54)        RUNNING      ________
Train B: A-1542-N7-002 (C55)        RUNNING      ________
Printed January 23, 2014 at 7:52
 
Approved By                                                                                  Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                        17102-1      20.2 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY              Page Number 6/5/13                                    RELATED DISPLAY CONSOLE QRM2                              25 of 42 WINDOW CDCA B6 ORIGIN                            SETPOINT A-RE-2532A Fuel Handling                    As determined                (RED LAMP LIT)
Building Effluent                by Chemistry                  (HIGH)
Radiogas Monitor                  Department ARE-2532A NOTE For other than HIGH conditions see Pages 5 and 6.
1.0                PROBABLE CAUSE
: 1.        High airborne radioactivity in the Fuel Handling Building.
: 2.        Equipment malfunction.
2.0                AUTOMATIC ACTIONS Switches the Normal Fuel Handling Building Ventilation to Accident Mode Ventilation.
3.0                INITIAL OPERATOR ACTIONS Evacuate the Fuel Handling Building.
Printed January 23, 2014 at 7:50
: 1. G2.1.8 001/LOIT/RO/M/F 3.4/4.1/G2.1.8/LO-TA-60047//H-17 NRC/
Given the following:
            - Unit 1 Control Room is being evacuated due to a fire.
            - The operating crew is fully staffed.
Which one of the following completes the following statement?
In accordance with 18038-1, "Operation From Remote Shutdown Panels," the __(1)__
will be dispatched to Shutdown Panel 'C' (TDAFW Pump Room),
and the PREFERRED method of communications to coordinate in-plant activities with personnel outside the control room is via __(2)__.
__(1)__                                    __(2)__
A.              Unit Operator                        sound powered telephones (red box)
B.              Unit Operator                      bridge phone extension 3145, codes 123# or 234#
C.            System Operator                        sound powered telephones (red box)
D.            System Operator                      bridge phone extension 3145, codes 123# or 234#
K/A G.2.1.8          Ability to coordinate personnel activities outside the control room.
K/A MATCH ANALYSIS The question presents a plausible scenario in which a Control Room evacuation is in progress. The candidate must determine which operator is required to report to Shutdown Panel "C" (TDAFW panel) and the preferred method of communications to co-ordinate activities with personnel outside the Control Room. Knowing where to report and how to communicate meets the KA.
EXPLANATION OF REQUIRED KNOWLEDGE Per AOP 18038-1 step 7, the System Operator goes to Shutdown Panel C (TDAFW Pump Room). The Unit Operator goes to Shutdown Panel B, unless he is the ENN Communicator, in which case he would go to the TSC.
Thursday, March 06, 2014 12:31:45 PM                                                          1
 
Per AOP 18038-1 step 14, communication is to be established between all stations, preferably using sound powered telephone remote shutdown channel, red box. The RNO of this step is to utilize either Bridge Phone Ext 3145, Page, or Radio.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. The System Operator is to report to Shutdown Panel C (TDAFW Pump Room). However, the report position of the UO is not commonly asked. It is reasonable for a candidate to believe that the UO would report to Shutdown Panel C since a System Operator has little experience and training in controlling SG levels, and this is normally the responsibility of the UO in the Main Control Room.
The second part is correct. Per AOP 18038-1 step 14, communication is to be established between all stations preferably using sound powered telephone remote shutdown channel, red box.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per AOP 18038-1 step 14, communication is to be established between all stations preferably using sound powered telephone remote shutdown channel, red box. However, the RNO of this step is to utilize either Bridge Phone Ext 3145, Page, or Radio. It is reasonable for a candidate without specific knowledge of the procedural requirement to conclude that the bridge phone would be a "better" option due to sound quality issues and need for headsets.
C. Correct.                  The first part is correct. Per 18038-1 step 7, the System Operator is to report to Shutdown Panel C (TDAFW Pump Room).
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Thursday, March 06, 2014 12:31:45 PM                                                              2
 
Level:                        RO Tier # / Group #              T3 K/A#                          G2.1.8 Importance Rating:            3.4 / 4.1 Technical
 
==Reference:==
18038-1, Rev 33.7, pages 10 & 15 References provided:          None Learning Objective:            LO-PP-60327-13 State the locations where the following operators will be stationed during operation from Remote Shutdown Panels.
: b. Reactor Operator (OATC)
: g. Extra shift personnel LO-TA-20016        Start the TDAFW Pump from Shutdown Panel "C" using 18038-1/2 LO-TA-60047        Establish control from remote shutdown panels Question origin:              BANK - Direct Reuse of HL17 NRC G2.1.8 Cognitive Level:              M/F 10 CFR Part 55 Content:        41.10 / 45.5 / 45.12 / 45.13 Comments:
You have completed the test!
Thursday, March 06, 2014 12:31:45 PM                                                            3
 
Approved By                                                                                  Procedure    Version J. THOMAS                                Vogtle Electric Generating Plant                    18038-1      33.7 Effective Date                                                                                Page Number OPERATION FROM REMOTE SHUTDOWN 6/5/13                                                  PANELS                                      10 of 124 ACTION/EXPECTED RESPONSE                                  RESPONSE NOT OBTAINED
: 2.          NOTE ATTACHMENT H is to be used at SS discretion as an aid for operators dispatched to perform local actions.
CAUTION Fire event qualified instrumentation is only available on Shutdown Panel B and marked in red.
7
__7.        Make a page announcement that the                      7.
Control Room is being evacuated and perform the following:
7.a
: a. Dispatch Operators to the                                a. IF insufficient personnel are following locations:                                    available, THEN use the following priority:
: 1)    Shutdown Panel B                                                                    7.a.1)
(CB-A43):                                        1)  Shutdown Panel B.
Shift Supervisor                                                                7.a.2)
: 2)  Shutdown Panel A.
Extra Shift Personnel 7.a.2)
: 2)    Shutdown Panel A                                  3)  Shutdown Panel C (CB-A75):                                              (TDAFW Pump Room).
7.a.4)
Reactor Operator                            4)  TSC - Plant Computer Terminal.
: 3)    Shutdown Panel C (TDAFW                                                              7.a.3)
Pump Room):
System Operator
 
Step 7 continued on next page Printed February 10, 2014 at 14:33
 
Approved By                                                                          Procedure    Version J. THOMAS                            Vogtle Electric Generating Plant                18038-1      33.7 Effective Date                                                                        Page Number OPERATION FROM REMOTE SHUTDOWN 6/5/13                                          PANELS                                    15 of 124 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 14
__14.      Establish communications between          __14. All stations on same channel or all stations (preferably sound                      line:
powered telephones remote                        __ Bridge Phone Ext 3145, codes shutdown channel, red box).                          123# or 234#.
__ Page
__ Radio 15
__15.      Initiate the Continuous Actions Page.        15.
16
__*16.      Monitor for Control Room                  __*16. WHEN Control Room conditions habitability.                                      become habitable, THEN go to Step 76 to continue recovery actions.
17
__17.      Place all transfer switches on                17.
1AA02-00 (CB-A48) to LOCAL.
18
__18.      Place all transfer switches on                18.
1BA03-00 (CB-A50) to LOCAL.
19
        *19. Check at least one ACCW Pump                *19. Perform the following:
RUNNING (approximately 62 amps):
__a. Stop all RCPs.                      19.a
__ 1AA02-15
__b. Isolate letdown by closing          19.b
                        -OR-                                              LETDOWN ISOLATION VLV UPSTREAM 1-LV-460 (Shutdown Panel A) and
__ 1BA03-20                                                LETDOWN ISOLATION VLV DOWNSTREAM 1-LV-459 (Shutdown Panel A.)
 
S Printed February 10, 2014 at 14:30
: 1. G2.2.15 001/LOIT AND LOCT/RO/C/A 3.9/4.3/G2.2.15/LO-TA-63007///
Initial conditions:
            - Unit 1 is at 45% reactor power.
            - EHC Pump A discharge filter #8 is in service.
            - EHC Pump A discharge filter #4 is tagged out for replacement.
Current conditions:
            - The following EHC Pump A discharge filter #4 valve tags are ready for release:
1-1615-U4-592, EHC HYD PUMP A DISCH FILTER #4 INLET ISO 1-1615-U4-593, EHC HYD PUMP A DISCH FILTER #4 OUTLET ISO
            - EHC Pump A discharge filter #8 is to remain in service.
Which one of the following completes the following statement?
Both valves will be __(1)__ after the tagout is released to standby alignment, and if EHC pressure were to lower to 1000 psig, the reactor __(2)__ trip.
REFERENCE PROVIDED
__(1)__                                    __(2)__
A.            open                                      would B.            open                                    would NOT C.          closed                                      would D.          closed                                  would NOT K/A G2.2.15          Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.
K/A MATCH ANALYSIS The KA addresses the relationship between plant documents and component status control to include asking the candidate to determine how to align components based on plant conditions. In addition, the candidate must determine which condition would cause a main turbine trip if EHC pressure lowered during the clearance release.
Thursday, March 06, 2014 12:39:43 PM                                                          1
 
EXPLANATION OF REQUIRED KNOWLEDGE Per lineup 11840-1, both valves have a required position of OPEN with a note that modifies the required position if filter 8 is in service, stating that the valve should be CLOSED in that case. Additionally, P&ID 1X4DB194 shows both valves as normally open. Both of these documents are given as references. Candidates are then required to reconcile the contradiction between the documents. P&IDs are utilized during tagout release preparation, but the lineup takes priority. All discrepancies must be carefully scrutinized, and compared to the operator's knowledge of system design and current plant configuration.
Note: There have been several instances in plant OE of inexperienced licensed operators releasing tagouts based on P&ID position alone. They did not reference the lineups or resolve discepancies between the two, reslulting in misposition events.
With EHC pressure lowered to less than 1100 psig, the turbine will trip. Since reactor power is above the P-9 setpoint of 40%, the reactor will also trip.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per lineup 11840-1, these valves should be released CLOSED since filter 8 is in service.
However, a candidate may believe the P&ID takes priority, or fail to reference the lineup and conclude the required position is OPEN.
The second part is correct. Per ALB20-D01, with EHC pressure lowered to less than 1100 psig, the turbine will trip. Since reactor power is above the P-9 setpoint of 40%, the reactor will also trip.
B.Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. Per ALB20-D01, with EHC pressure lowered to less than 1100 psig, the turbine will trip.
Since reactor power is above the P-9 setpoint of 40%, the reactor will also trip. However, a candidate can easily confuse the P-8 setpoint of 48% with the P-9 setpoint of 40% and assume the turbine will trip and the reactor will not. Or, a candidate could confuse the alarm setpoint of 1500 psig and the EHC pump autostart setpoint of 1400 psig with the turbine trip setpoint of 1100 psig and believe that the turbine will not trip.
C. Correct.                  The first part is correct. Per lineup 11840-1, these valves should be released CLOSED since filter 8 is in service.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
Thursday, March 06, 2014 12:39:43 PM                                                                  2
 
The second part is incorrect. See the second part of choice B above.
Level:                          RO Tier # / Group #                T3 K/A#                            G2.2.15 Importance Rating:              3.9 / 4.3 Technical
 
==Reference:==
LINEUP 11840-1 "EHC System Alignment' page 4, Rev 14.1 ARP 17020-1, Rev 53.2, pages 41, 42, & 68 SOP 13503A-1, Rev 7.2, page 33 P&ID 1X4DB194, Rev 29.0 References provided:            LINEUP 11840-1 "EHC System Alignment' page 4, Rev 14.1 P&ID 1X4DB194, Rev 29.0 Learning Objective:            LO-PP-30103-04 Describe EHC pumps normal operating pressure.
LO-PP-30201-12 State the signals that will generate a turbine trip; excluding specific generator trips.
LO-TA-63007          Tagout Review in accordance with NMP-AD-003 Question origin:                NEW Cognitive Level:                C/A 10 CFR Part 55 Content:        41.10 / 43.3 / 45.13 Comments:                      Question appears to match the KA.
All of question (2) is LOD=1 - I would expect that all operators should already know the appropriate way to perform a check of a valve, whether it is open or closed.
Question (1) is a LOD=1 - the answer for part (1) is a direct lookup from the provided lineup, thereby making all of the other distractors non plausible.
Recommend replacing this question. Need to make sure that the selected reference doesn't lend itself to a direct lookup -
i.e., you could use a reference for one system where the way another system is aligned impacts the correct answer for the given system. The applicant would need to have plant specific knowledge of the secondary system to know which answer is correct.
                                        -JAT 12/19/2013 Question is improved, but need to ensure there is only one correct answer. Is "standby alignment" defined anywhere?
Is it possible that someone could successfully argue that Thursday, March 06, 2014 12:39:43 PM                                                                  3
 
returning Filter #4 to service and taking filter #8 out of service meets "standby alignment" (thereby possibly having two correct answers to this question)?
                                      - JAT 2/4/14 A bullet was added to current conditions that states "EHC Pump A discharge filter #8 is to remain in service" to eliminate the possible contention basis.
                                      - JCC 4/3/14 You have completed the test!
Thursday, March 06, 2014 12:39:43 PM                                                              4
 
Approved By                                                                                      Procedure Number Rev R. K. Pope                                      Vogtle Electric Generating Plant                11840-1      14.1 Date Approved                                                                                    Page Number 11/18/2003                        MAIN TURBINE ELECTRO-HYDRAULIC CONTROL (EHC) SYSTEM ALIGNMENT          4 of 11 CONDITION  LINEUP COMPONENT                      DESCRIPTION                                            REQUIRED    (INITIALS)
LOCATED AT THE HYDRAULIC POWER UNIT - TURBINE BLDG LEVEL 1 1-1615-U4-590                  EHC HYD PUMP A DISCHARGE ISOLATION                      OPEN        __________
1-1615-U4-592                  EHC HYD PUMP A DISCH FILTER #4 INLET ISO                OPEN (1)    __________
1-1615-U4-593                  EHC HYD PUMP A DISCH FILTER #4 OUTLET ISO              OPEN (1)    __________
1-1615-U4-594                  EHC HYD PUMP A DISCH FILTER #8 INLET ISO                CLOSED (2)  __________
1-1615-U4-595                  EHC HYD PUMP A DISCH FILTER #8 OUTLET ISO              CLOSED (2)  __________
1-1615-X4-807                  EHC HYD PUMP A DISCH FILTER #4 VENT                    CLOSED      __________
1-1615-X4-808                  EHC HYD PUMP A DISCH FILTER #8 VENT                    CLOSED      __________
1-1615-X4-801                  EHC HYD PUMP A DISCH FILTER PDI-6496 HI SIDE ISOLATION  OPEN        __________
1-1615-X4-802                  EHC HYD PUMP A DISCH FILTER PDI-6496 LOW SIDE ISOLATION OPEN        __________
1-1615-X4-803                  EHC HYD PUMP A DISCH FILTER PDI-6496 BYPASS            CLOSED      __________
1-1615-U4-591                  EHC HYD PUMP B DISCHARGE ISOLATION                      OPEN        __________
1-1615-U4-596                  EHC HYD PUMP B DISCH FILTER #5 INLET ISO                OPEN (3)    __________
1-1615-U4-597                  EHC HYD PUMP B DISCH FILTER #5 OUTLET ISO              OPEN (3)    __________
1-1615-U4-598                  EHC HYD PUMP B DISCH FILTER #9 INLET ISO                CLOSED (4)  __________
(1)      CLOSED IF FILTER 8 SELECTED FOR SERVICE (2)      OPEN IF FILTER 8 SELECTED FOR SERVICE (3)      CLOSED IF FILTER 9 SELECTED FOR SERVICE (4)      OPEN IF FILTER 9 SELECTED FOR SERVICE Printed November 27, 2013 at 13:35
 
Approved By                                                                                                    Procedure    Version J.B. Stanley                          Vogtle Electric Generating Plant                                        13503A-1 7.2 Effective Date                TRAIN A REACTOR CONTROL SOLID-STATE PROTECTION                                  Page Number 6/21/13                                                      SYSTEM                                                  33 of 38 ATTACHMENT C                                        Sheet 1 of 6 PERMISSIVES, CONTROL INTERLOCKS, REACTOR TRIPS AND ESF ACTUATIONS PERMISSIVES Permissive                Setpoint/Coincidence              Function P-4                        Train related Rx trip &          Trips Main Turbine Bypass breaker open                        Train A - mechanical Train B - electrical FWI if Lo Tavg (2/4  564 F) present Seals in FWI if caused by SI or P-14 (Hi Hi Level)
Must be present to block auto SI after SI reset.
P-4 Train A arms Steam Dumps P-4 Train B swaps Steam Dumps to plant trip controller P-6                        1/2 IR Detectors  2.0 E -5      Allows manual block of SR High  trip
                                % Rx Power P-7                        P-10 (2/4 PR NIs  10%            Unblocks "At Power" Trips Rx power) or                              Przr Low Pressure P-13 (PT-505 or 506                      Przr High Level 10% turbine power)                        RCS Two Loop Low Flow RCP UF RCP UV P-8                        2/4 PR NIS  48% Rx power        Enables Single Loop Low Flow Rx Trip P-9                        2/4 PR NIS  40% Rx power        Enables Turbine trip Rx trip P-10                      2/4 PR NIS  10% Rx power        Auto block of SR High  trip Enables P-7 Allows manual block of IR rod stop and Hi  trip Allows manual block of PR Hi  trip Lo Setpoint P-11                      2/3 Przr Pressure channels        Auto enables Lo Przr Press SI & Lo Steamline Press 2000 psig                      SI/SLI & sends signal to open Accum Isolation Valves when P-11 resets. P-11 allows operator to block PRZR & Steamline low pressure SI & SLI. Also activates "Not Full Open" annunciators for Accumulator MOVs (ALB16; A5, B5, C5 &
D5) and HV-8806; (ALB16 E03).
P-12                      2/4 NR Tavg  550 F              Interlocks Steam Dumps closed ( Cooldown Dump Valves PV-507A, B & C) may be reopened by use of Bypass Interlock switches)
P-13                      1/2 Turbine Impulse channels    Enables P-7 10%
P-14                      2/4 NR SG Level channels  82%    Actuates FWI Actuates MFP and Main Turbine trip Printed January 23, 2014 at 13:35
 
Approved By                                                                                    Procedure    Version J.B. Stanley                        Vogtle Electric Generating Plant                            17020-1      53.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 20 ON PANEL                Page Number 05/08/2013                                            1B2 ON MCB                                      41 of 80 WINDOW D05 ORIGIN                              SETPOINT HYD FLUID 1-PISL-6338A                        1500 psig                  LO PRESS 1-PISL-6338B 1.0                PROBABLE CAUSE
: 1.        Failure of Electrohydraulic Control (EHC) Fluid Pumps.
: 2.        Clogged strainers and filters in pump suction or discharge.
: 3.        EHC Fluid System leak.
2.0                AUTOMATIC ACTIONS
: 1.        If pressure drops below 1400 psig, the standby EHC Fluid Pump will start.
: 2.        If pressure continues to drop to 1100 psig, the Turbine will trip.
3.0                INITIAL OPERATOR ACTIONS
: 1.        IF a reactor trip occurs, Go To 19000-C, "E-0 Reactor Trip Or Safety Injection."
: 2.        Verify standby EHC Fluid Pump is on, if needed.
4.0                SUBSEQUENT OPERATOR ACTIONS CAUTION EHC fluid is a fire resistant fluid that may be harmful to personnel. Observe proper safety precautions when in contact with this fluid.
: 1.        Dispatch an operator to the Hydraulic Power Unit to check for system leaks or pump failure.
: 2.        IF equipment failure is indicated, initiate maintenance as required.
Printed January 23, 2014 at 14:21
 
Approved By                                                                        Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                  17020-1      53.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 20 ON PANEL    Page Number 05/08/2013                                      1B2 ON MCB                                42 of 80 WINDOW D05 (Continued) 5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB194, 1X3D-BC-Q56B, 1X4AA01-280-0, CX5DT1101-17A Printed January 23, 2014 at 14:21
 
Approved By                                                                                Procedure    Version J.B. Stanley                      Vogtle Electric Generating Plant                        17020-1      53.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 20 ON PANEL            Page Number 05/08/2013                                            1B2 ON MCB                                  68 of 80 WINDOW E05 ORIGIN                            SETPOINT HYD FLUID 2 out of 2:                        1100 psig                  LO PRESS 1-PISL-6338A                      with a 3-sec                TURB TRIP 1-PISL-6338B                      time delay 1.0                PROBABLE CAUSE
: 1.        Failure of Electrohydraulic Control (EHC) Fluid Pumps.
: 2.        Clogged strainers and filters in pump suction or discharge.
: 3.        EHC Fluid System leak.
2.0                AUTOMATIC ACTIONS Turbine trip.
3.0                INITIAL OPERATOR ACTIONS
: 1.        IF a reactor trip has occurred, Go To 19000-C, "E-0 Reactor Trip Or Safety Injection."
: 2.        IF a reactor trip has NOT occurred, Go To 18011-C, "Turbine Trip Below P-9."
4.0                SUBSEQUENT OPERATOR ACTIONS NONE 5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
1X4DB194, 1X3D-BC-Q56B, 1X4AA01-280-0, 1X5DN203-1, CX5DT1101-17A Printed January 23, 2014 at 14:21
 
Approved By                                                                                      Procedure Number Rev R. K. Pope                                      Vogtle Electric Generating Plant                11840-1      14.1 Date Approved                                                                                    Page Number 11/18/2003                        MAIN TURBINE ELECTRO-HYDRAULIC CONTROL (EHC) SYSTEM ALIGNMENT          4 of 11 CONDITION  LINEUP COMPONENT                      DESCRIPTION                                            REQUIRED    (INITIALS)
LOCATED AT THE HYDRAULIC POWER UNIT - TURBINE BLDG LEVEL 1 1-1615-U4-590                  EHC HYD PUMP A DISCHARGE ISOLATION                      OPEN        __________
1-1615-U4-592                  EHC HYD PUMP A DISCH FILTER #4 INLET ISO                OPEN (1)    __________
1-1615-U4-593                  EHC HYD PUMP A DISCH FILTER #4 OUTLET ISO              OPEN (1)    __________
1-1615-U4-594                  EHC HYD PUMP A DISCH FILTER #8 INLET ISO                CLOSED (2)  __________
1-1615-U4-595                  EHC HYD PUMP A DISCH FILTER #8 OUTLET ISO              CLOSED (2)  __________
1-1615-X4-807                  EHC HYD PUMP A DISCH FILTER #4 VENT                    CLOSED      __________
1-1615-X4-808                  EHC HYD PUMP A DISCH FILTER #8 VENT                    CLOSED      __________
1-1615-X4-801                  EHC HYD PUMP A DISCH FILTER PDI-6496 HI SIDE ISOLATION  OPEN        __________
1-1615-X4-802                  EHC HYD PUMP A DISCH FILTER PDI-6496 LOW SIDE ISOLATION OPEN        __________
1-1615-X4-803                  EHC HYD PUMP A DISCH FILTER PDI-6496 BYPASS            CLOSED      __________
1-1615-U4-591                  EHC HYD PUMP B DISCHARGE ISOLATION                      OPEN        __________
1-1615-U4-596                  EHC HYD PUMP B DISCH FILTER #5 INLET ISO                OPEN (3)    __________
1-1615-U4-597                  EHC HYD PUMP B DISCH FILTER #5 OUTLET ISO              OPEN (3)    __________
1-1615-U4-598                  EHC HYD PUMP B DISCH FILTER #9 INLET ISO                CLOSED (4)  __________
(1)      CLOSED IF FILTER 8 SELECTED FOR SERVICE (2)      OPEN IF FILTER 8 SELECTED FOR SERVICE (3)      CLOSED IF FILTER 9 SELECTED FOR SERVICE (4)      OPEN IF FILTER 9 SELECTED FOR SERVICE Printed November 27, 2013 at 13:35
 
Depicted as OPEN, which contradicts the Lineup.
: 1. G2.2.23 001/LOIT/RO/C/A 3.1/4.6/G2.2.23/LO-LP-61202-06///
Procedure title as follows:
            - 14807A-1, "Train 'A' Motor Driven Auxiliary Feedwater Pump and Check Valve Inservice and Response Time Test" At time 1000:
            - Unit 1 is at 7% reactor power with a startup in progress.
            - MDAFW pump 'A' discharge valves, 1HV-5137 and 1HV-5139, to the SGs are closed by the UO to perform a surveillance per 14807A-1.
At time 1005:
            - MDAFW pump 'A' is started.
At time 1010:
            - MDAFW pump 'A' is stopped.
At time 1015:
            - Train 'A' MDAFW system is returned to normal standby alignment.
Which one of the following completes the following statement?
During performance of the surveillance and per Tech Spec 3.7.5, "Auxiliary Feedwater (AFW) System," the UO would track an LCO not met (out-of-service) time of __(1)__
minutes, and if a Train 'A' MDAFW actuation signal were received during the surveillance when the discharge valves were closed, the valves would __(2)__.
__(1)__                                  __(2)__
A.                      5                                automatically open B.                      5                                  remain closed C.                      15                                automatically open D.                      15                                  remain closed Thursday, March 06, 2014 12:41:42 PM                                                        1
 
K/A G2.2.23          Ability to track Technical Specification limiting conditions for operations.
K/A MATCH ANALYSIS The question tests the candidate's ability to track Technical Specifications, and, using procedure guidance, to correctly track safety related equipment OOS time associated with an AFW surveillance.
EXPLANATION OF REQUIRED KNOWLEDGE Per OSP 14807A-1 CAUTION prior to step 5.1.3, the Train A MDAFW pump is inoperable as soon as either discharge valve is closed. Therefore, the LCO would not be met starting at time 10:00 and would remain not met until 10:15 when the AFW system was restored to standby alignment. The logged out-of-service time would be 15 minutes.
Per ELEMENTARY 1X3D-BC-F08C, the discharge valve handswitches are spring return to auto. Once the discharge valves were stroked closed, the handswitch would be in AUTO. Therefore, when the AFW Actuation signal was processed, both discharge valves would stroke fully open. Even though AFW discharge valves will automatically stroke open, per commitments 1984301714 and 1984301715, AFW will be considered inoperable with discharge valves closed and reactor power >5% RTP.
NOTE: The first part of the question is associated with AFW OPERABILITY. It is part of the normal RO job function to notify the SS when an LCO is not met as part of a surveillance as well as when the inoperable condition is restored. The inoperability call is explicitely called out in the surveillance in a CAUTION. The OPERABILITY requirement associated with AFW discharge valves is taught as part of systems training. Therefore, this is an RO level function.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per OSP 14807A-1 CAUTION prior to step 5.1.3, the Train A MDAFW pump is inoperable as soon as either discharge valve is closed. Therefore, the LCO would not be met starting at time 10:00 and would remain not met until 10:15, when the AFW system was restored to standby alignment. The logged out-of-service time would be 15 minutes. However, the candidate may determine that since the discharge valves would stroke open on an automatic signal, the LCO time would only be associated with the time from when the AFW pump was stopped until standby alignment was achieved, or from when the discharge valves were closed until the pumps were started. Either of these would be recorded as 5 minute durations.
The second part is correct. The discharge valve handswitches Thursday, March 06, 2014 12:42:07 PM                                                              1
 
are spring return to auto. Once the discharge valves were stroked closed, the handswitch would be in AUTO. Therefore, if an AFW Actuation signal was received, both discharge valves would stroke fully open.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. The discharge valve handswitches are spring return to auto. Once the discharge valves were stroked closed, the handswitch would be in AUTO.
Therefore, if an AFW Actuation signal was received, both discharge valves would stroke fully open. However, the candidate may determine that since the required alignment is open, the valves do not receive an automatic signal. In addition, the knowledge that the open signal can be overridden and the valves not automatically open makes this selection plausible.
C. Correct.                  The first part is correct. Per OSP 14807A-1 CAUTION prior to step 5.1.3, the Train A MDAFW pump is inoperable as soon as either discharge valve is closed. Therefore, the LCO would not be met starting at time 10:00 and would remain not met until 10:15 when the AFW system was restored to standby alignment. The logged out-of-service time would be 15 minutes.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
You have completed the test!
Thursday, March 06, 2014 12:42:07 PM                                                              2
 
Level:                      RO Tier # / Group #            T3 K/A#                        G2.2.23 Importance Rating:          3.1 / 4.6 Technical
 
==Reference:==
UOP 12004-C, Rev 108.0, page 25 SOP 13610-1, Rev 50.4, page 10 & 96 OSP 14870A-1, Rev 5.0, page 10 LESSON PLAN V-LO-PP-20101, Rev 3.2, slide 14 ELEMENTARY 1X3D-BC-F08C, Rev 8.0 References provided:        None Learning Objective:          LO-LP-61202-06 Describe the basic steps required to place the AFW System in standby readiness.
LO-PP-20101-01 Describe when the AFW system is used to support normal power operations and what the maximum power AFW operation is used for.
LO-PP-20101-02 Describe the normal at power standby alignment of the AFW system.
LO-PP-20101-04 List the AFW system automatic start signals and component actuations.
LO-LP-63404-02 Describe the following as applicable to the surveillance test program:
: a. purpose of surveillance work orders
: b. where the procedure number to be used for performance of surveillance tests are identified
: c. who must authorize the performance of tests that manipulate or affect plant equipment
: d. who reviews the test results to confirm that they satisfy the acceptance criteria
: e. purpose of surveillance test
: f. failure of surveillance tests g.duties and responsibilities of surveillance test performer if a test fails or cannot be completed within the specified time LO-TA-61010        Perform Power Ascent During Low Power Operations using 12004-C Question origin:            NEW Cognitive Level:            C/A 10 CFR Part 55 Content:      41.4 / 41.7 / 41.10 Thursday, March 06, 2014 12:42:26 PM                                                                1
 
Comments:
You have completed the test!
Thursday, March 06, 2014 12:42:26 PM                              2
 
Approved By                                                                              Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                      13610-1      50.4 Effective Date                                                                            Page Number 6/21/13                              AUXILIARY FEEDWATER SYSTEM                                10 of 109 INITIALS 2.2              LIMITATIONS 2.2.1            Technical Specification LCO 3.7.5 requires that three independent Auxiliary Feedwater trains be operable in MODES 1, 2, or 3. TS SR 3.7.5.1 requires that each automatic valve in the discharge flowpath must be in the fully open position for standby readiness.        ________
Testing of TDAFW Pump in accordance with Technical Specification SR 3.7.5.2 is not required to be performed until 24 hours after steam generator pressure is greater than or equal to 900 psig.                                                                ________
2.2.2            The TDAFW Pump Turbine must be coupled to the pump prior to entry into MODE 3 per Technical Specification LCO 3.7.5.                  ________
2.2.3            Technical Specification LCO 3.7.6 requires that one Condensate Storage Tank (CST) be operable, with a safety-related volume of greater than or equal to 340,000 gallons (66% of span) when in Modes 1, 2, or 3.
(SNC16060, 1996332946, (SNC16059, 1996332945)                              ________
2.2.4            To prevent pump degradation and abnormal wear, the following minimum flow requirements have been established for the MDAFW and TDAFW Pumps:
: a.        For short periods of operation not to exceed three (3) hours in a 24-hour period:; (SNC15010, 1990319699)
(1)    MDAFW Pump greater than or equal to 150 gpm.            ________
(2)    TDAFW Pump greater than or equal to 53 gpm
                                    @ 1535 rpm varying linearly to 145 gpm @ 4230 rpm (see Figure 1).                                    ________
: b.        For continuous pump operation greater than three (3) hours:
(1)    MDAFW Pump greater than or equal to 265 gpm.            ________
(2)    TDAFW Pump greater than or equal to 175 gpm
                                    @ 1535 rpm varying linearly to 450 gpm @ 4230 rpm (see Figure 2                                      ________
Printed November 15, 2013 at 9:25
 
Approved By                                                                                            Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                                    13610-1      50.4 Effective Date                                                                                          Page Number 6/21/13                                AUXILIARY FEEDWATER SYSTEM                                            96 of 109 CHECKLIST 2 -            AUXILIARY FEEDWATER SYSTEM ALIGNMENT                                          Sheet 1 of 3 FOR STANDBY READINESS NOTES This checklist should be performed on non-operating equipment only. Any equipment which is currently in operation should be marked N/A and noted in the comment section of this checklist.
The AFW System is OPERABLE in MODES 1-3 with the MDAFW Pump Miniflow Isolation Valves (1-1302-U4-769 and 770) either open or closed. The preferred position for standby readiness is closed because the system is more tolerant of potential failure of 1-FV-5154 and 5155 (and associated controls) when 1-1302-U4-769 and 770 are closed.
Any of the following sections in this checklist should be marked N/A if not used.
I. Motor Driven Auxiliary Feedwater Pump A Alignment:
CONDITION          LINEUP        VERIFICATION COMPONENT            DESCRIPTION                          REQUIRED      (INITIALS)      (INITIALS)
(1)  1HS-5139A            SG-1 FROM MDAFW PMP-A                OPEN          __________      ____________
(2)  1HS-5137A            SG-4 FROM MDAFW PMP-A                OPEN          __________      ____________
(3)  1HS-5131A            MDAFW-A 1-1302-P4-003                AUTO          __________      ____________
(4)  OVERCURRENT LOR MDAFW-A 1-1302-P4-003                    RESET        __________      ____________
(At breaker 1AA02-17)
(5) FV-5155              AFW P-3 MINI FLOW                    OPEN          _________        ___________
(ZLB-4 INDICATION LIT)
(6)  1-1302-U4-769 AFW MDAFW PUMP A                            CLOSED        __________      ____________
MINI-FLOW ISO VALVE (7)  MDAFW PUMP A suction & discharge piping                  NOT hot
            .                                                        to touch      __________
Printed November 15, 2013 at 9:26
 
Approved By                                                                              Procedure  Version Ronald M. Brown                    Vogtle Electric Generating Plant                      12004-C 108 Effective Date                                                                          Page Number 01/16/2014                              POWER OPERATION (Mode 1)                              25 of 119 INITIALS
: o.        Outage management has reviewed the outage schedule and determined that no impacts to mode change exist.        ________
Outage Management Representative
_____________________________
(Signature)
: p.        Engineering review of 00309-C, Control of Unattended Temporary Material In Containment In Modes 1-4 has been completed for Mode 1 entry and any restrictions are known and resolved.                                          ________
: q.        Notify SM that conditions are met to change status from Mode 2 to Mode 1. (SM to document authorization to change modes on Checklist 1.)                                ________
4.1.16            PRIOR to exceeding 5% reactor power, as read on the highest reading PR NIS or highest reading Loop T, verify the AFW has been aligned for Standby per Checklist 2 of 13610, Auxiliary Feed Water System." (1984301714, 1984301715)                          ________*
NOTE The purpose of this step is to ensure no obstructions are present in steam lines or feed lines.
4.1.17            Perform a channel check of NIS and Delta-Ts and initiate periodic monitoring to ensure all four loops track power increase together.                                                              ________
Printed March 4, 2014 at 15:21
 
Approved By                                                                        Procedure  Version T. A. Bussiere                    Vogtle Electric Generating Plant                  14807A-1      5 Effective Date          TRAIN A MOTOR DRIVEN AUXILIARY FEEDWATER PUMP AND CHECK    Page Number 1-22-13                            VALVE INSERVICE AND RESPONSE TIME TEST                10 of 23 INITIALS 5.1              TEST OF MDAFW PUMP A 5.1.1            Check Train A MDAFW Pump suction pressure on 1-PI-5129A is 15 psig or greater and record pressure.
________psig                                        ________
5.1.2            IF performing response time test,
: a.        Calculate the pressure at which the stop watch will be stopped for measuring response time to discharge pressure.                                                ________
Discharge Pressure = Suction Pressure + Minimum P Discharge Pressure = ___________          +  1625 psid Discharge Pressure = ___________ psig
: b.        Record target discharge pressure in Step 5.1.11.2 a. ________
CAUTION The following step will render Train A MDAFW Pump INOPERABLE.
(Technical Specifications LCO 3.7.5) 5.1.3            Close the following valves:
Critical SG-4 FROM MDAFW PMP-A            1-HV-5137              ________
________
CV Critical SG-1 FROM MDAFW PMP-A            1-HV-5139              ________
________
CV Printed November 15, 2013 at 9:28
 
Auxiliary Feedwater System Objectives 1 & 2
* Review functions of system - ESF heat sink/feedwater source when MFW not in service.
* Review major flowpaths from each CST through pumps into each SG.
* Discuss power supplies for each pump and MOV
* Steam supplies for TDAFW pump from A(3009) & B(3019) Train DC power.
* All other TDAFW MOVs (and speed controller) are Train C DC power.
* MDAFW pumps and MOVs are Train related 4160/480VAC power.
* Show control room indications and controls
* Discuss normal standby alignment of system modes 1, 2, 3.
AFW is used for startup (up to 4% power) for maintaining SG level.
It is also designed for SG level maintenance following a reactor trip.
(V-LO-PP-20101-01)
After swapping to MFPs in Mode 2 (before transition to Mode 1,) AFW is put in a standby alignment:
      -TDAFW may already be aligned but MDAFW cannot be aligned when being used for SGWL control.
      -MDAFW pump switches in AUTO.
      -TDAFW HV-5106 is shut with the Trip and Throttle valve latched and open.
      -Both TDAFW steam supply valves are OPEN.
      -TDAFW speed controller demand at maximum.
      -All TDAFW and MDAFW discharge valves FULLY OPEN.
V-LO-PP-20101 Rev-03.2                                                                              14
 
Auxiliary Feedwater System (V-LO-PP-20101-02)
V-LO-PP-20101 Rev-03.2        14
: 1. G2.3.13 001/LOIT/RO/C/A 3.4/3.8/G2.3.13/LO-LP-63308-02///
Initial conditions:
            - Unit 1 is in Mode 5.
            - RCS level has been lowered to 192 feet to remove reactor head.
            - ALARA briefing in progress for closing Equipment Hatch.
Current conditions:
              - The Unit Operator is performing a surveillance that specifies an Independent Verification (IV) of a manual valve inside containment.
            - The area where the valve is located has the following conditions:
                  - Airborne contamination is 0.1 DAC.
                  - The operator is expected to receive 15 mrem of dose while performing the IV.
Per NMP-GM-005-002, "Human Performance Tools Instruction," which one of the following identifies the reason for waiving the IV?
A. Plant Mode of operation.
B. RCS at reduced inventory.
C. A significant radiation exposure.
D. Airborne contamination in excess of limits.
K/A G2.3.13          Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
K/A MATCH ANALYSIS The question tests the candidate's knowledge of containment entry requirements and radiological safety procedures associated with the Independent Verification process as it applies to excessive dose while performing IVs.
EXPLANATION OF REQUIRED KNOWLEDGE NMP-GM-005-002 allows for an exemption to the Independent Verification requirement associated with return to service of safety related components based on significant radiological exposure. Step 5.2.2.a.b(2) states IV's may be waived for the following reasons:
                  - In cases that involve significant radiation exposure.
Thursday, March 06, 2014 12:45:22 PM                                                          1
 
                - In cases that involve containment entry, while containment integrity is established.
NMP-GM-005-002 Definition 3.7 describes Significant Radiation Exposure as greater than or equal to 10mrem whole body dose or airborne contamination in excess of ALARA guidelines. HP ADMIN procedure 00930-C Definition 2.1.b defines areas with an airborne concentration of >0.3 DAC as being in excess of ALARA guidelines for general entries.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. NMP-GM-005-002 makes no provisions for exemption of IVs based on Mode. However, prior to NMP-GM-005-002, the Vogtle specific procedure had an exemption for Modes 1-4 with containment integrity established. A candidate without specific knowledge of the exemption requirements may remember an association with Modes and find it reasonable that IVs would not be required in a lower Mode. In general, as the plant enters Mode 5, many of the safety-related systems are no long required to be maintained OPERABLE. Therefore, this assumption could appear to be justified by Tech Specs.
B. Incorrect. Plausible. NMP-GM-005-002 makes no provisions for exemption of IVs based on RCS inventory. Access to containment and allowed work activities are restricted during these times due to the increased nuclear risk. Additionally, non essential personnel are restricted from access to areas of containment during a head lift due to dose rates. A candidate without specific knowledge of the exemptions found in NMP-GM-005-002 may find it reasonable to exempt the IVs under these circumstances.
However, in these cases, the IV would not be exempted it would be delayed and performed later, when conditions were more favorable.
C. Correct.                  Per NMP-GM-005-002, an IV may be exempted if a dose of greater than or equal to 10 mrem whole body is anticipated to be received.
D. Incorrect. Plausible. NMP-GM-005-002 makes a provision for exemption of IVs based on airborne contamination as a significant exposure.
However, airborne contamination would have to be in excess of 0.3 DAC before this would be considered significant. However, a candidate without specific knowledge of IV exemptions may conclude that any airborne contamination would be significant and therefore an exemption would be allowed.
Thursday, March 06, 2014 12:45:22 PM                                                              2
 
Level:                        RO Tier # / Group #              T3 K/A#                          G2.3.13 Importance Rating:            3.4 / 3.8 Technical
 
==Reference:==
00930-C, Rev 26.0, page 2 NMP-GM-005-002, Rev 2.0, pages 4, 16, & 17 References provided:          None Learning Objective:            LO-LP-63308-01 Briefly describe the independent verification policy. Include a discussion of the types of verification that are available including concurrent verification.
LO-LP-63308-02 With regards to independent verification, describe the following:
: a. conditions which warrant independent verification
: b. components/systems that require independent verification (and exceptions)
: c. safety-related, as applicable to independent verification Question origin:              BANK Cognitive Level:              C/A 10 CFR Part 55 Content:        41.12 Comments:
You have completed the test!
Thursday, March 06, 2014 12:45:22 PM                                                              3
 
Human Performance Tools Instruction                                            NMP-GM-005-002 SNC                      Version 2.0 Unit S                  Page 4 of 33 1.0        PURPOSE The purpose of this instruction is to identify the core set of Human Performance error reduction tools that will be applied at all Southern Nuclear Operating Company (SNC) sites. It identifies when the Human Performance tools will be used, the error precursors that may lead up to the event, the consequences of not using the human performance tools, and at risk practices to avoid.
2.0        APPLICABILITY This procedure applies to the entire SNC workforce, site and corporate, as well as contractors and vendors who perform work at SNC facilities.
3.0        DEFINITIONS 3.1        At-risk Practice - A behavior, belief, assumption, or condition that tends to diminish the effectiveness of the tool.
3.2        Critical Activity -An activity, if performed improperly, will cause irreversible harm to plant equipment or people or will significantly impact plant operations.
3.3        Critical Step - A procedure step, series of steps, or action that will cause irreversible harm to plant equipment or people or will significantly impact plant operation if performed improperly.
3.4        Error-Likely Situation - A work situation in which there is greater opportunity for error when a specified action or task is performed due to the presence of error precursors.
3.5        Error-Precursors (Risk Factors) - Task related conditions related to a specific activity or task that provoke human error, increase the chance of a technical error or an adverse consequence.
3.6        Latent Error - An error, act, or decision that unknowingly creates an undesired condition(s). It may be embedded in a process, culture, plant configuration of systems, structures, or components or the design bases, or reduces equipment reliability that remains undetected until revealed by subsequent operational activities.
3.7        Significant Radiation Exposure - As applicable to activities described by this procedure, greater than or equal to 10 mrem whole body dose or airborne contamination in excess of ALARA guidelines.
0.3 DAC required per 00930-C, RADIATION AND CONTAMINATION 4.0        RESPONSIBILITIES CONTROL Responsibilities for this instruction listed in NMP-GM-005 RESPONSIBILITIES section.
Printed December 16, 2013 at 11:51
 
Human Performance Tools Instruction                                                NMP-GM-005-002 SNC                    Version 2.0 Unit S                Page 16 of 33 5.2.1      Pre-Job Brief (continued)
: 3.      (continued)
: g.      During the Pre-Job Briefing, a determination for a post-job review should be determined and communicated by the job supervisor. The communication will include the time and location of the Post-Job Review as applicable. Note that a Post-Job Review may also be required by NMP-DP-001, Operational Risk Awareness. See Attachment 4 for items to be covered in a Post-Job Review.
5.2.2      Verification Practices Verification practices refers broadly to three toolsconcurrent verification, independent verification, and peer-checkingthat involve a second person to confirm the actions and results achieved by a performer. While peer-checking (PC) focuses on preventing a mistake by the performer, independent verification (IV) and concurrent verification (CV) focus more on confirming the correct configuration, or status, of equipment.
: 1.      Independent Verification Independent verification is a series of actions by two individuals working independently to confirm the condition of a component after the original act that placed it in that condition.
: a.      The Purpose of the Independent Verification (IV) tool:
The IV process confirms the condition of equipment required to be in a particular condition to maintain the plants physical configuration required for safe operation. Otherwise, adverse consequences could result later if the improper condition remains undetected. IV can only be used when an immediate, adverse consequence of a mistake by the performer cannot occur, because IV catches errors after they have been made, not before or during.
NOTE Independent Verification is performed by an individual who has basic knowledge of the type of component involved (valve, breaker, etc.). The individual need not be trained on the activity or system involved.
: b.      When to perform Independent Verification:
(1)    Independent Verification is required for restoration of:
Safety related systems or components.
Valve positions in liquid or gaseous radioactive waste systems that if mis-positioned could lead to unintended or unmonitored radio activity release.
Printed December 16, 2013 at 11:51
 
Human Performance Tools Instruction                                                NMP-GM-005-002 SNC                  Version 2.0 Unit S              Page 17 of 33 5.2.2      Verification Practices (continued)
: 1.      b.      (1)    (continued)
Other component positioning as determined necessary by the Operations Director (2)    Exemption - IV may be waived for the following reasons:
In cases that involve significant radiation exposure.
In cases that involve containment entry (PWR) or drywell entry (BWR), while containment integrity is established.
NOTE Performer and verifier should be dispatched separately.
: c.      How to perform Independent Verification (1)    The performer of the component manipulation is SEPARATED from the verifier by time and distance.
(2)    The performer shall, with the use of the controlling document:
(a)    LOCATE the component and identify each unique identifier on the component label.
(b)    PERFORM the intended action.
(3)    The verifier shall, with use of the controlling document:
(a)    LOCATE the component and identify each unique identifier on the component label.
(b)    CONFIRM the completed action.
NOTE Concurrent Verification is performed by an individual qualified for the activity, systems, and/or components involved and the relationship of these activities, components, and systems to plant safety.
: 2.      Concurrent Verification Concurrent verification (CV) is a series of actions by two individuals working together at the same time and place to separately confirm the condition of a component before, during, and after an action, when the consequences of an incorrect action would lead to immediate and possibly irreversible harm to the plant or personnel.
Printed December 16, 2013 at 11:51
 
Approved By                                                                                Procedure Number Rev C.R.Dedrickson                    Vogtle Electric Generating Plant                      00930-C          26 Date Approved                                                                              Page Number 08/05/2009                        RADIATION AND CONTAMINATION CONTROL                            2 of 28 1.0              PURPOSE(1984301253) (1985303534)
This procedure establishes requirements and responsibilities for monitoring and controlling exposure to radiation and contamination. It includes criteria for Radiation Controlled Areas, the Radiation Work Permit (RWP) system, sampling and surveys, shielding, and the Self Monitoring process as follows:
4.0        REQUIREMENTS 5.0        PROCEDURE 5.1        POSTING 5.2        RADIATION WORK PERMIT 5.3        CONTAMINATION CONTROLS 5.4        PERSONNEL MONITORING 5.6        SURVEYS AND SAMPLING 5.7        SHIELDING 5.8        SELF MONITORING 2.0              DEFINITIONS 2.1              AIRBORNE RADIOACTIVITY AREA (1985306088) (1993327366)
A room, enclosure, or area in which airborne radioactive materials, composed wholly or partly of licensed material, exist in concentrations:
: a.        To such a degree that an individual present in the area without respiratory protection equipment could exceed, during the hours an individual is present in a week, an intake of 0.6 percent of the annual limit on intake (ALI) or 12 Derived Air Concentrations (DAC)-hours.
OR
: b.        An area should be considered an Airborne Radioactivity Area when the airborne activity reaches 0.3 DACs for isotopes which have a classification other than submersion, and 100% of the DAC limits for isotopes which have a classification of submersion. DAC limits are specified in 10CFR20 Appendix B, Table I, Column 3.
2.2              ANNUAL LIMIT ON INTAKE (ALI)
The derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI values for intake by ingestion and by inhalation of selected radionuclides are given in 10CFR20, Appendix B, Table 1, Columns 1 and 2. (1993327366)
Printed December 16, 2013 at 11:54
: 1. G2.3.15 001/LOIT/RO/C/A 2.9/2.9/G2.3.15/LO-LP-60309-08///
Initial condition:
            - Unit 1 is at 11% reactor power with a startup in progress.
Current condition:
            - ALB15-C01 HIGH RADIATION ALARM is received due to SG tube leakage.
Which one of the following completes the following statement?
1RE-0724, Steam Line Radiation, __(1)__ available to monitor SG activity, and the monitor is designed to detect __(2)__ radiation.
__(1)__                                __(2)__
A.                      is                                  noble gas B.                      is                                nitrogen-16 C.                  is NOT                                noble gas D.                  is NOT                                nitrogen-16 K/A G2.3.15          Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
K/A MATCH ANALYSIS The KA addresses the relationship between fixed radiation monitors and their response to alarms. This question is generic in nature because it addresses the operation of a rad monitor that is common to both units.
EXPLANATION OF REQUIRED KNOWLEDGE Tube Leakage Rad Monitor RE-724 utilizes N-16 to detect and quantify primary to secondary leakage through the SG tubes. N-16 is produced by the fission process at approximately 16% RTP and above. The power level at which N-16 is produced is not exactly 16% RTP, however the detector has proven reliable at greater than or equal to 16% RTP. The rad monitor will indicate at power levels lower than 16%, but will be inaccurate. For this reason, UOP 12004-C step 4.2.28 directs the removal of RE-724 from service at less than 16% power. In contrast, RE-810 and RE-12839 monitor noble gases and are available at lower power levels, although they are not as accurate as Thursday, March 06, 2014 12:46:51 PM                                                          1
 
RE-724, . RE-810 monitors SG steam flow and RE-12839 monitors the steam packing and steam jet air ejector flows.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. RE-724 is not accurate and therefore removed from service with reactor power <16% RTP. However, RE-810 is commonly confused with RE-724 and would be available and accurate at 11% RTP.
The second part is incorrect. RE-724 utilizes N-16 for detection. However, other secondary rad monitors utilize noble gas, including RE-0810 and RE-12839. RE-724 and RE-810 are commonly confused by candidates.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. RE-724 utilizes N-16 for detection.
C. Incorrect. Plausible. The first part is correct. RE-724 is not accurate and therefore removed from service with reactor power <16% RTP.
The second part is incorrect. See the second part of choice A above.
D. Correct.                  The first part is correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, March 06, 2014 12:46:51 PM                                                                  2
 
Level:                        RO Tier # / Group #              T3 K/A#                          G2.3.15 Importance Rating:            2.9 / 2.9 Technical
 
==Reference:==
ARP 17100-1, Rev 26.2, page 42 UOP 12004-C, Rev 107.1, page 56 References provided:          None Learning Objective:            LO-LP-60309-08 Describe how to obtain plant computer readings that provide trend displays that are useful to determine leak rate and rate of leak changes.
LO-PP-21101-16 Discuss why and where radiation monitors are, install on the Main Steam System.
Question origin:              BANK Cognitive Level:              C/A 10 CFR Part 55 Content:        41.11 / 43.4 / 45.9 Comments:
You have completed the test!
Thursday, March 06, 2014 12:46:51 PM                                                              3
 
Approved By                                                                              Procedure  Version C.E.H. Williams                    Vogtle Electric Generating Plant                      12004-C 107.1 Effective Date                                                                            Page Number 05/30/2013                                POWER OPERATION (Mode 1)                              56 of 114 INITIALS NOTE Failure of the Main Generator output breakers to open should not be considered an abnormal response requiring initiation of 19000-C, "E-0 Reactor Trip or Safety Injection."
(4)    Trip the Turbine per 13800, "Main Turbine operation."                                            ________
Verify turbine tripped per 13800, "Main Turbine operation".                            ________
Verify the Main Generator PCBs are OPEN.
per 13800, "Main Turbine operation".          ________
4.2.27            At recorder NR-45:
Verify all Power Range channels indicating properly.          ________
Verify all Intermediate Range channels indicating properly. ________
4.2.28            IF reactor power will be less than 16% for at least a week, notify chemistry to remove N-16 Rad Monitor RE-724 from service.              ________
NOTE Step 4.2.29 should only be performed with Shift Manager authorization.
4.2.29            IF authorized by Shift Manager, place the Auxiliary Feedwater System in service on mini-flow, per 13610, Auxiliary Feedwater System.                                                                ________
NOTE Shutting down reactor by manually driving All Rods IN should not be performed IF the ARO position was NOT verified to be 228 in step 4.2.4.
4.2.30            IF the reactor shall be shut down by manually driving ALL RODS IN, go directly to step 4.2.34.                                ________
4.2.31            IF the reactor shall be shut down by performing a manual reactor trip, continue with step 4.2.32.                                ________
Printed January 22, 2014 at 11:46
 
Approved By                                                                              Procedure  Version S. E. Prewitt                    Vogtle Electric Generating Plant                      17100-1      26.2 Effective Date              ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND          Page Number 12/9/12                          EFFLUENT RADIATION MONITORING SYSTEM (RMS)                    42 of 88 ORIGIN                      SETPOINT 1-RE-0724 Steam Line                  As determined by              (High)
(N16 monitor)              Chemistry Department NOTES For other than HIGH conditions see Pages 4 and 5.
This detector monitors secondary activity at power and indicates primary to secondary leakage in GPD and GPD/HR.
1.0                PROBABLE CAUSE Steam Generator Tube leakage 2.0                AUTOMATIC ACTIONS NONE 3.0                INITIAL OPERATOR ACTIONS NONE Printed January 22, 2014 at 11:39
: 1. G2.3.5 001/LOIT/RO/M/F 2.9/2.9/G2.3.5/LO-TA-23005///013K1.13 Initial condition:
            - Unit 1 and Unit 2 are at 100% reactor power.
Current conditions:
            - Unit 1, ALB05-C03 HIGH RADIATION ALARM is received.
            - 1RE-12116, Control Room Air Intake Radiogas Monitor (Train 'A'), RED indication light is illuminated on the SRDC.
            - Unit 2 has NO radiation monitor in alarm.
Which one of the following completes the following statement?
The Unit 1, Train 'B' CREFS unit __(1)__ automatically start, and the Unit 2, Train 'B' CREFS unit __(2)__ automatically start.
__(1)__                    __(2)__
A.                    will                        will B.                    will                    will NOT C.                  will NOT                      will D.                  will NOT                  will NOT K/A Generic G2.3.5            Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
K/A MATCH ANALYSIS The question tests the candidate's ability to use Control Room Intake Rad Monitors 1/2RE-12116/12117 to determine expected operation/response.
EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17102-1, 1RE-12116 red lamp indicates a HIGH alarm. The Control Room Ventilation is expected to switch to the Post Accident Mode. Per step 4.1, if the switch did not occur automatically, a manual actuation is to be performed using SOP 13301-C.
Per SOP 13301-C section 4.4.1.1, placing either CRI handswitch to ACTUATE will start Thursday, March 06, 2014 12:43:59 PM                                                          1
 
the (LEAD) Train B CREFs filter unit on Unit 1 ONLY. Section 4.4.1.2 must be performed to actuate the Unit 2 CREFs system if necessary. Additionally, P&IDs AX4DB206-1 &3 also show that 1/2RE-12116/12117 feed separate Train and separate Unit CRI Logic circuits. Therefore, 1RE-12116 in HIGH alarm will only start the Unit 1 Train B CREFs unit.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. Per ARP 17102-1 and SOP 13301-C section 4.4.1.1, 1RE-12116 in HIGH alarm is expected to result in an automatic start of the Unit 1 Train B CREFs unit.
The second part is incorrect. Per ARP 17102-2, SOP 13301-C section 4.4.1.1 and P&ID AX4DB206-1 &3, 1RE-12116 in HIGH alarm will NOT result in an automatic start of the Unit 2 Train B CREFs unit. Actuation of the Unit 2 B Train CREFs unit would come from either 2RE-12116 or 2RE-12117 in HIGH alarm.
However, since the Unit 1 and Unit 2 Control Rooms share a common envelope, it is reasonable to believe that an actuation of one Unit's intake rad monitor would actuate CRI on both Units.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. Per ARP 17102-2, SOP 13301-C section 4.4.1.1 and P&ID AX4DB206-1 &3, 1RE-12116 in HIGH alarm is NOT expected to result in an automatic start of the Unit 2 Train B CREFs unit. Actuation of the Unit 2 B Train CREFs unit would come from either 2RE-12116 or 2RE-12117 in HIGH alarm.
C. Incorrect. Plausible. The first part is incorrect. Per ARP 17102-1 and SOP 13301-C section 4.4.1.1, 1RE-12116 in HIGH alarm is expected to result in an automatic start of the Unit 1 Train B CREFs unit.
However, a candidate could find it reasonable that 1RE-12116 would actuate Train A CREFs and not Train B CREFs.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.
You have completed the test!
Thursday, March 06, 2014 12:43:59 PM                                                                  2
 
Level:                          RO Tier # / Group #                T3 K/A#                            G2.3.5 Importance Rating:              2.9 / 2.9 Technical
 
==Reference:==
P&ID AX4DB206-1, Rev 31.0 P&ID AX4DB206-3, Rev 29.0 SOP 13301-C, Rev 30.0, pages 27-34 ARP 17102-1, Rev 20.2, pages 23 & 24 References provided:            None Learning Objective:            LO-TA-23005        Manual actuation of Control Room Isolation using 13301-C LO-TA-32007        Verify Proper Automatic actions to high radiation alarms LO-PP-23301-03 Describe the Emergency and Isolation mode of operation for the Control Room HVAC system including flow paths and interlocks.
LO-PP-23301-05 List the actuating signals for CRI including indications and automatic actions.
Question origin:                BANK Cognitive Level:                M/F 10 CFR Part 55 Content:        41.7 / 41.11 / 41.12 Comments:        Question appears to match the KA, but is not plant-specific and is not at the RO level. Anyone qualified to work in the RCA should know the answer to this question.
I recommend replacing this question. Given that the KA is generic and broad (ability to use radiation monitoring systems pretty much lends itself to any detector used to measure radiation), I would ask a question where a radiation monitor is alarming (during fuel handling, during operations, while shut down, etc) and ask the required actions (can be out of an AOP or annunciator response) or ask if a particular AOP is required to be entered (entry conditions for AOPs and EOPs is RO knowledge). This question can be similar to any questions that may also fit a process radiation monitor question. The key is to keep it plant specific and at the level of an operator.
                          -JAT 12/19/2013 Question is plant specific and appears to match the KA. One thing to note is that there are four distinct answers just with the first part of the question - the part regarding the Essential chillers does not appear needed to answer the question. Additionally, since choice A is not Thursday, March 06, 2014 12:44:19 PM                                                                  1
 
specific as to which train of CREFS starts, it is an outlier. Really, what youre asking is which train of CREFS starts and on which units. That can be done by asking:
Unit 1 Train A CREFS [does/does not] start and Unit 2 Train A CREFS
[does/does not] start.
Or something similar i.e., make it a 2x2, soliciting the answer to two pieces of information, since thats what the question ultimately boils down to. It doesnt have to be asked like the example I gave; thats just a starting point. This way, question psychometrics cant be used to whittle the choices down, and you also dont use the word ONLY, which can be problematic.
                          -JAT 2/4/14 You have completed the test!
Thursday, March 06, 2014 12:44:19 PM                                                                2
 
Each Rad Monitor is associated with a specific Unit's and Train's CRI Logic.
 
Each Rad Monitor is associated with a specific Unit's and Train's CRI Logic.
 
Approved By                                                                              Procedure    Version R.M. Brown                        Vogtle Electric Generating Plant                      13301-C 30 Effective Date                                                                          Page Number 10/02/20123              CBCR NORMAL HVAC AND EMERGENCY FILTRATION SYSTEM                      27 of 47 INITIALS 4.4                NON-PERIODIC OPERATIONS 4.4.1              Manual Actuation Of Control Room Isolation NOTES This section is written using Unit 1, Unit 2 and Common component designations. Some Unit 2 designations are shown in parenthesis.
If the TRAIN B CR FLTR UNIT SUPPLY FAN fails to start on actuation, the Train A Fan will start after a 30 second time delay.
The TSC Air Filtration System will automatically start on manual initiation of Control Room Isolation.
ALB05-D05 GROUP 4 MONITOR LIGHT COMP OFF NORM ALB39-D05 480V SWGR ANB30 TROUBLE ALB50-B03 CR HI/LO DIFF PRESS 4.4.1.1            Manually initiate Control Room Isolation on Unit One:
: a.        Place either CR ISO MANUAL ACTUATION Switch in ACTUATE:
1-HS-12195A [A4] (TRAIN A) to ACTUATE.                ________
1-HS-12196A [A6] (TRAIN B) to ACTUATE.                ________
: b.        Verify that TRAIN B CR FLTR UNIT SUPPLY FAN (LEAD),
1-1531-N7-002 [B10] starts.                                  ________
: c.        Verify that TRAIN A CR FLTR UNIT SUPPLY FAN, 1-1531-N7-001 [B8] (STANDBY) does NOT start:                ________
Placing either CRI actuation handswitch on Unit 1 to ACTUATE starts the B Train CREFs units on Unit 1 only. Unit 2 must be actuated separately by section 4.4.1.2.
Printed March 4, 2014 at 16:27
 
Approved By                                                                        Procedure    Version R.M. Brown                        Vogtle Electric Generating Plant                13301-C 30 Effective Date                                                                    Page Number 10/02/20123              CBCR NORMAL HVAC AND EMERGENCY FILTRATION SYSTEM                28 of 47 INITIALS
: d.        Verify that both KIT TOIL + CONF RM EXH ISO DMPRs close:
A-HV-12162 [D6], TRAIN A, CLOSED                      ________
A-HV-12163 [D7], TRAIN B, CLOSED                      ________
NOTES All positions in Step 4.4.1.1.e may be verified in any order.
Unit Two dampers are in bold print to help in identification.
: e.        Verify the following damper positions; (1)    CR NORM AIR SUPPLY ISO DMPRs, 1-HV-12146 [C6], TRAIN A CLOSED                ________
2-HV-12146 [C6], TRAIN A CLOSED                ________
1-HV-12147 [C7], TRAIN B CLOSED                ________
2-HV-12147 [C7],, TRAIN B CLOSED                ________
(2)    CR NORM AIR RTN ISO DMPRs, 1-HV-12149 [E6] TRAIN A CLOSED                  ________
2-HV-12149 [E6] TRAIN A CLOSED                  ________
1-HV-12148 [E7] TRAIN B CLOSED                  ________
2-HV-12148 [E7] TRAIN B CLOSED                  ________
Printed March 4, 2014 at 16:27
 
Approved By                                                                              Procedure    Version R.M. Brown                        Vogtle Electric Generating Plant                      13301-C 30 Effective Date                                                                            Page Number 10/02/20123              CBCR NORMAL HVAC AND EMERGENCY FILTRATION SYSTEM                      29 of 47 INITIALS
: f.        Verify the following for the Filter Unit that started; (1)    IF Train B Filter Unit, 1-1531-N7-002 [B10] started, (a)    Verify that the Train B CR FILTER UNIT OUTLET AIR DMPR, 1-HV-12129 [C11] OPEN.        ________
(b)    Verify that the Train B CR RTN FAN INLET AIR DMPR on the running train, 1-HV-12131 [D10] OPEN.                          ________
(c)    Verify that the Train B CR NORMAL HVAC UNIT INTAKE ISO DMPR on the running train, A-HV-12152 [B7] CLOSED.                        ________
(2)    IF Train A Filter Unit, 1-1531-N7-001 [B8] started, (a)    Verify that the Train A CR FILTER UNIT OUTLET AIR DMPR, 1-HV-12128 [C9] OPEN.          ________
(b)    Verify that the Train A CR RTN FAN INLET AIR DMPR on the running train, 1-HV-12130 [D8] OPEN.                          ________
(c)    Verify that the Train A CR NORMAL HVAC UNIT INTAKE ISO DMPR on the running train, A-HV-12153 [B6] CLOSED.                        ________
: g.        Verify that the CR NORM AC UNIT SUPPLY FANs, shut down.
A-1531-A7-001 [C4], STOPPED.                            ________
A-1531-A7-002 [C5], STOPPED.                            ________
: h.        Verify that the CR NORM AC UNIT EXH FANs, shut down.
A-1531-B7-009 [D4] STOPPED.                            ________
A-1531-B7-010 [D5], STOPPED.                            _______
: i.        Verify that the KITCH TOILET AND CONF RM EXH FAN, A-HS-12164 in the Shift AA's Office, stops.                  _______
Printed March 4, 2014 at 16:27
 
Approved By                                                                                Procedure    Version R.M. Brown                          Vogtle Electric Generating Plant                        13301-C 30 Effective Date                                                                              Page Number 10/02/20123              CBCR NORMAL HVAC AND EMERGENCY FILTRATION SYSTEM                        30 of 47 INITIALS NOTE If it is necessary to isolate outside air to the Control Room in the next step due to smoke or toxic gas intake, both the Unit 1 and Unit 2 dampers should be shut.
: j.        IF Control Room outside air is restricted for Control Room habitability due to smoke or toxic gas intake, THEN close the CR Outside Air Supply Dampers for BOTH Units:
UNIT 1 1-HS-12114 [E8]                                                ________
1-HS-12115 [E10]                                                ________
UNIT 2 2-HS-12114 [E8]                                                ________
2-HS-12115 [E10]                                                ________
k        Verify proper operation of the TSC Air Filtration System per 13303-C, "Technical Support Center And Central Alarm Station HVAC Systems."                                          ________
: l.        Verify proper Essential Chiller operation.                      ________
Printed March 4, 2014 at 16:27
 
Approved By                                                                              Procedure    Version R.M. Brown                        Vogtle Electric Generating Plant                      13301-C 30 Effective Date                                                                          Page Number 10/02/20123              CBCR NORMAL HVAC AND EMERGENCY FILTRATION SYSTEM                      31 of 47 INITIALS NOTES This section is written using Unit 1, Unit 2 and Common component designations. Some Unit 2 designations are shown in parenthesis.
If the TRAIN B CR FLTR UNIT SUPPLY FAN fails to start on actuation, the Train A Fan will start after a 30 second time delay.
The TSC Air Filtration System will automatically start on manual initiation of Control Room Isolation.
ALB05-D05 GROUP 4 MONITOR LIGHT COMP OFF NORM ALB39-D05 480V SWGR ANB30 TROUBLE ALB50-B03 CR HI/LO DIFF PRESS 4.4.1.2            Manually initiate Control Room Isolation on Unit Two:
: a.        Place either CR ISO MANUAL ACTUATION Switch in ACTUATE:
2-HS-12195A [A4] (TRAIN A) to ACTUATE.                ________
2-HS-12196A [A6] (TRAIN B) to ACTUATE.                ________
: b.        Verify that TRAIN B CR FLTR UNIT SUPPLY FAN (LEAD),
2-1531-N7-002 [B10] starts.                                  ________
: c.        Verify that TRAIN A CR FLTR UNIT SUPPLY FAN, 2-1531-N7-001 [B8] (STANDBY) does NOT start:                ________
: d.        Verify that both KIT TOIL + CONF RM EXH ISO DMPRs close:
A-HV-12162 [D6], TRAIN A, CLOSED                            ________
A-HV-12163 [D7], TRAIN B, CLOSED                            ________
Printed March 4, 2014 at 16:27
 
Approved By                                                                              Procedure    Version R.M. Brown                        Vogtle Electric Generating Plant                      13301-C 30 Effective Date                                                                            Page Number 10/02/20123              CBCR NORMAL HVAC AND EMERGENCY FILTRATION SYSTEM                      32 of 47 INITIALS NOTES All positions in Step 4.4.1.2.e may be verified in any order.
Unit Two dampers are in bold print to help in identification.
: e.        Verify the following damper positions; (1)    CR NORM AIR SUPPLY ISO DMPRs, 1-HV-12146 [C6], TRAIN A CLOSED                        ________
2-HV-12146 [C6], TRAIN A CLOSED                        ________
1-HV-12147 [C7], TRAIN B CLOSED                        ________
2-HV-12147 [C7],, TRAIN B CLOSED                      ________
(2)    CR NORM AIR RTN ISO DMPRs, 1-HV-12149 [E6] TRAIN A CLOSED                        ________
2-HV-12149 [E6] TRAIN A CLOSED                        ________
1-HV-12148 [E7] TRAIN B CLOSED                        ________
2-HV-12148 [E7] TRAIN B CLOSED                        ________
: f.        Verify the following for the Filter Unit that started; (1)    IF Train B Filter Unit, 2-1531-N7-002 [B10] started, (a)  Verify that the Train B CR FILTER UNIT OUTLET AIR DMPR, 2-HV-12129 [C11] OPEN.          ________
(b)  Verify that the Train B CR RTN FAN INLET AIR DMPR on the running train, 2-HV-12131 [D10] OPEN.                          ________
(c)  Verify that the Train B CR NORMAL HVAC UNIT INTAKE ISO DMPR on the running train, A-HV-12152 [B7] CLOSED.                          ________
Printed March 4, 2014 at 16:27
 
Approved By                                                                              Procedure    Version R.M. Brown                        Vogtle Electric Generating Plant                      13301-C 30 Effective Date                                                                          Page Number 10/02/20123              CBCR NORMAL HVAC AND EMERGENCY FILTRATION SYSTEM                      33 of 47 INITIALS (2)    IF Train A Filter Unit, 2-1531-N7-001 [B8] started, (a)    Verify that the Train A CR FILTER UNIT OUTLET AIR DMPR, 2-HV-12128 [C9] OPEN.        ________
(b)    Verify that the Train A CR RTN FAN INLET AIR DMPR on the running train, 2-HV-12130 [D8] OPEN.                          ________
(c)    Verify that the Train A CR NORMAL HVAC UNIT INTAKE ISO DMPR on the running train, A-HV-12153 [B6] CLOSED.                        ________
: g.        Verify that the CR NORM AC UNIT SUPPLY FANs, shut down.
A-1531-A7-001 [C4], STOPPED.                          ________
A-1531-A7-002 [C5], STOPPED.                          ________
: h.        Verify that the CR NORM AC UNIT EXH FANs, shut down.
A-1531-B7-009 [D4] STOPPED.                            ________
A-1531-B7-010 [D5], STOPPED.                          _______
: i.        Verify that the KITCH TOILET AND CONF RM EXH FAN, A-HS-12164 in the Shift AA's Office, stops.                  _______
Printed March 4, 2014 at 16:27
 
Approved By                                                                                Procedure    Version R.M. Brown                          Vogtle Electric Generating Plant                        13301-C 30 Effective Date                                                                              Page Number 10/02/20123              CBCR NORMAL HVAC AND EMERGENCY FILTRATION SYSTEM                        34 of 47 INITIALS NOTE If it is necessary to isolate outside air to the Control Room in the next step due to smoke or toxic gas intake, both the Unit 1 and Unit 2 dampers should be shut.
: j.        IF Control Room outside air is restricted for Control Room habitability due to smoke or toxic gas intake, THEN close the CR Outside Air Supply Dampers for BOTH Units::
UNIT 1 1-HS-12114 [E8]                                                ________
1-HS-12115 [E10]                                                ________
UNIT 2 2-HS-12114 [E8]                                                ________
2-HS-12115 [E10]                                                ________
k        Verify proper operation of the TSC Air Filtration System per 13303-C, "Technical Support Center And Central Alarm Station HVAC Systems."                                          ________
: l.        Verify proper Essential Chiller operation.                      ________
Printed March 4, 2014 at 16:27
 
Approved By                                                                            Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                    17102-1      20.2 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY        Page Number 6/5/13                                  RELATED DISPLAY CONSOLE QRM2                        23 of 42 WINDOW CDCA B5 ORIGIN                            SETPOINT 1-RE-12116 Control Room Air                  As determined        (RED LAMP LIT)
Intake Process                    by Chemistry          (HIGH)
Radiogas Monitor                  Department 1-RE-12116 NOTE For other than HIGH conditions see Pages 5 and 6.
1.0                PROBABLE CAUSE
: 1.        Gaseous radioactivity in the incoming air.
: 2.        Equipment malfunction.
2.0                AUTOMATIC ACTIONS Control Room and Technical Support Center Ventilation switches to the Post Accident Mode.
3.0                INITIAL OPERATOR ACTIONS NONE Printed January 23, 2014 at 15:03
 
Approved By                                                                                    Procedure  Version J.B. Stanley                        Vogtle Electric Generating Plant                          17102-1      20.2 Effective Date                  ANNUNCIATOR RESPONSE PROCEDURES FOR THE SAFETY                Page Number 6/5/13                                      RELATED DISPLAY CONSOLE QRM2                              24 of 42 WINDOW CDCA B5 (Continued) 4.0                SUBSEQUENT OPERATOR ACTIONS
: 1.        Verify a Control Room Filtration Unit is running. If not, manually start per 13301-C CBCR Normal HVAC and Emergency Filtration System, Section for manually initiating a Control Room Isolation.
: 2.        Verify the Chilled Water System associated with the operating Emergency Control Room HVAC train has started per 13744A-1 Train A Essential Chilled Water System or 13744B-1 Train B Essential Chilled Water System as appropriate.
: 3.        Verify Technical Support Center Air Filtration Unit in service per 13303-C Technical Support Center And Central Alarm Station HVAC Systems.
: 4.        Refer to 91001-C, "Emergency Classification And Implementing Instructions."
: 5.        Notify Health Physics to locate and determine the cause of the contamination.
: 6.        IF sampling and analysis determine that the channel has malfunctioned:
: a. Comply with Technical Specifications LCO 3.3.7.
: b. Place 1-HS-12195C on QESF to TEST BLOCK CHAN I.
: c. Request Chemistry to investigate and take corrective action.
5.0                COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE
 
==REFERENCES:==
AX4DB206-3, 1X3D-BG-C03C, 1X5DX2101 Printed January 23, 2014 at 15:03
: 1. G2.4.25 001/LOIT/RO/M/F 3.3/3.7/G2.4.25/LO-PP-43101-02///
Given the following:
            - Both Unit 1 and Unit 2 are at 100% reactor power.
            - A fire occurs in the Fuel Handling Building as a result of a seismic event.
Which one of the following completes the following statement?
The __(1)__ pumps and piping are designed to provide firefighting water following the seismic event, and per 13903-C, "Fire Protection System Operation," the flow path __(2)__ automatically initiate water flow without operator action.
__(1)__                                __(2)__
A.                  NSCW                                      will B.                  NSCW                                    will NOT C.                Diesel fire                                  will D.                Diesel fire                              will NOT K/A G2.4.25          Knowledge of fire protection procedures.
K/A MATCH ANALYSIS The question addresses the operators knowledge of fire procedures related to a specific event, the fire protection system that is available, and the action required to place it in service.
EXPLANATION OF REQUIRED KNOWLEDGE Per the NOTE at the begining of procedure SOP 13150A-1, SOP 13903-C is utilized to place the Seismic Standpipe Fire Water Protection System in service. SOP 13903-C section 4.4 gives direction to unlock and open the manual valves in Table 1 that supply water to the various building seismic standpipes. Step 4.4.1 requires the candidate to determine the inservice NSCW train and use it to supply it's respective standpipe system.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. The pumps and piping available for fire Thursday, March 06, 2014 12:48:42 PM                                                            1
 
protection following a seismic event are in the NSCW system.
The second part is incorrect. SOP 13903-C section 4.4 gives direction to unlock and open the manual valves in Table 1 that supply water to the various building seismic standpipes.
However, the candidate may believe seismic standpipe system piping is integral to the normal fire water piping and is available without local action in the event of a fire, and will automatically actuate when a fire is present.
B. Correct.                  The first part is correct. See the first part of choice A above.
The second part is correct. SOP 13903-C section 4.4 gives direction to unlock and open the manual valves in Table 1 that supply water to the various building seismic standpipes.
C. Incorrect. Plausible. The first part is incorrect. The pumps and piping available for fire protection following a seismic event are in the NSCW system. However, the candidate may determine the fire system is designed to class 1 seismic standards since it is protecting buildings with safety related components. In addition, the fire protection system is designed to handle almost all other events, including a loss of offsite power via the diesel powered fire pumps. This could lead the candidate to believe it is available following a seismic event as well.
The second part is incorrect. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the second part of choice C above.
The second part is correct. See the second part of choice B above.
Thursday, March 06, 2014 12:48:42 PM                                                                    2
 
Level:                        RO Tier # / Group #              T/G K/A#                          G2.4.25 Importance Rating:            3.3 / 3.7 Technical
 
==Reference:==
SOP 13150A-1, Rev 8.2, page 4 SOP 13903-C, Rev 44.0, pages 38 & 78 thru 80 References provided:          None Learning Objective:            LO-PP-06101-13 Describe how the NSCW system may be used to provide fire protection water following a seismic event at VEGP.
LO-PP-43101-02 Describe how the Seismic I standpipe is placed in service and the areas protected.
Question origin:              NEW Cognitive Level:              M/F 10 CFR Part 55 Content:        41.4 / 41.8 / 41.10 / 43.5 /45.13 Comments:
You have completed the test!
Thursday, March 06, 2014 12:48:42 PM                                                              3
 
Approved By                                                                              Procedure  Version J.B. Stanley                      Vogtle Electric Generating Plant                      13150A-1 8.2 Effective Date                                                                            Page Number 04/09/2013                  TRAIN A NUCLEAR SERVICE COOLING WATER SYSTEM                        4 of 125 1.0                PURPOSE NOTE 13903-C, Fire Protection System Operation provides direction to place the Seismic Category 1 Standpipe Fire Water Protection System in service.
This procedure provides instructions for operation of the Nuclear Service Cooling Water (NSCW) System including operation of the NSCW Pumps, Cooling Towers, and Cooling Tower Makeup System. Operating instructions are provided in the following subsections:
4.1.1        Train A NSCW Startup to Standby 4.1.2        Train A NSCW Startup from Standby 4.1.3        Train A NSCW Cooling Tower Makeup System Startup 4.2.1        Shifting Train A NSCW Pumps 4.2.2        Shifting NSCW Cooling Tower Makeup Pumps 4.3.1        Train A NSCW Shutdown to Standby 4.4.1        Deicing Train A NSCW Cooling Tower when Operating 4.4.2        Idle NSCW Train Cooling Tower Deicing 4.4.3        Manual NSCW Cooling Tower Makeup to Train A 4.4.4        Alternate NSCW Cooling Tower Makeup to Train A 4.4.5        Makeup to NSCW Cooling Tower A from NSCW Cooling Tower B 4.4.6        Fill and Vent Train A using NSCW Cooling Tower Cross Pumping Provision 4.4.7        Fill and Vent of Train A Safety Related Pump Motor Coolers 4.4.8        Train A NSCW Single Pump Operation (Outage) 4.4.9        Train A NSCW Single Pump Operation (Abnormal)
Printed January 28, 2014 at 11:14
 
Approved By                                                                                  Procedure    Version C.H. Williams                      Vogtle Electric Generating Plant                          13903-C      44 Effective Date                                                                              Page Number 07/26/2013                          FIRE PROTECTION SYSTEM OPERATION                                38 of 92 INITIALS 4.3.4              WHEN directed, stop Jockey Pump C-2301-P4-004 as follows:
: a.        Check Jockey Pump C-2301-P4-001(in Electric Pump House) in operation.                                            ________
: b.        Place handswitch C-HS-7902 to OFF on PFH2.                      ________
4.4                CATEGORY I STANDPIPE OPERATION NOTES The Standpipe System should be used only if the normal Fire Protection System is incapable of fire water delivery.
CO1985306084 Standpipe connections back pack in the Primary Dress Out Locker (CB R104) has two 100 foot sections of 1 1/2 hose, two 1 1/2 nozzles, AND one gated Y.
For additional equipment in the Auxiliary Building, two 2-1/2 fire hoses, three adjustable spray nozzles, bolt cutters, zip ties, etc. can be found in a JOBOX located against the south wall of the BTRS chiller room (AB Rm 124)
CAUTION The design basis for the Standpipe System is two fire hoses operating for 30 minutes at 100 gal/min each. (6000 gal extracted from the Ultimate Heat Sink) 4.4.1              Determine which NSCW Train is in service to supply its respective Standpipe System.                                              ________
4.4.2              Unlock and open the in service NSCW train Standpipe System Supply Valve to the affected building per TABLE 1.                        ________
4.4.3              Open the Fire Hose Valve to the affected area per TABLE 2.
4.4.4              Notify the Control Room to monitor NSCW basin level.                      ________
4.4.5              Close all valves opened in Steps 4.4.2 and 4.4.3 WHEN officially secured from the fire event; (IV REQUIRED).                                ________
Printed January 28, 2014 at 11:17
 
Approved By                                                                  Procedure    Version C.H. Williams                      Vogtle Electric Generating Plant        13903-C      44 Effective Date                                                              Page Number 07/26/2013                          FIRE PROTECTION SYSTEM OPERATION                78 of 92 Sheet 1 of 3 TABLE 1 COMPONENT                    DESCRIPTION                            LOCATION AUX BLDG TRAIN A 1-1202-U4-089                NSCW AUX BLDG LEVEL B ISO FROM NSCW    R-A40A 2-1202-U4-089                NSCW TRAIN A SPLY TO AB FIRE HOSE CONN R-B92 TRAIN B A-1202-U4-153                NSCW NUC SERV CLG WTR TO AUX BLDG ISO VLV                                R-B08(FHB) 1-1202-U4-091                NSCW AUX BLDG LEVEL B ISO FROM NSCW    R-B28 2-1202-U4-091                NSCW TRAIN B SPLY TO AB FIRE HOSE CONN R-B92 CONTROL BLDG TRAIN A 1-1202-U4-004                NSCW AUX BLDG LEVEL B ISO FROM NSCW    1CB316 (IN OVERHEAD AT FREIGHT ELEVATOR) 2-1202-U4-004                NSCW TRAIN A SPLY TO CB FIRE HOSE CONN 2CB328 TRAIN B 1-1202-U4-003                NSCW SYS INTERTIE TO CB LEVEL 2 FIREWATER                              1CB316 (IN OVERHEAD AT FREIGHT ELEVATOR) 2-1202-U4-003                NSCW TRAIN B SPLY TO CB FIRE HOSE CONN 2CB308 Printed January 28, 2014 at 11:17
 
Approved By                                                                  Procedure    Version C.H. Williams                      Vogtle Electric Generating Plant          13903-C      44 Effective Date                                                                Page Number 07/26/2013                          FIRE PROTECTION SYSTEM OPERATION                79 of 92 Sheet 2 of 3 TABLE 1 COMPONENT                    DESCRIPTION                            LOCATION CONTAINMENT BLDG TRAIN A 1-1202-U4-002                NSCW TO FIRE PROT HOSE ADAPTOR ISOLATION                              EL 207' COL 11 2-1202-U4-002                NSCW TRAIN A SPLY TO CNMT FIRE HOSE CONN                                  EL 210' COL 11 TRAIN B 1-1202-U4-001                NSCW TO FIRE PROT HOSE ADAPTOR ISOLATION                              LEVEL 1 (7 FT RIGHT OF COLUMN 18) 2-1202-U4-001                NSCW TRAIN B SPLY TO CNMT FIRE HOSE CONN                                  EL 220' COL 19 Printed January 28, 2014 at 11:17
 
Approved By                                                                Procedure    Version C.H. Williams                    Vogtle Electric Generating Plant          13903-C      44 Effective Date                                                              Page Number 07/26/2013                          FIRE PROTECTION SYSTEM OPERATION                80 of 92 Sheet 3 of 3 TABLE 1 COMPONENT                    DESCRIPTION                          LOCATION DIESEL GENERATOR BLDG TRAIN A 1-1202-U4-028                NSCW SYS HOSE ISO TRAIN A DG BLDG    DG A R-103 2-1202-U4-028                NSCW TRAIN A SPLY TO DG A FIRE HOSE CONN                                  DG A R-101 TRAIN B 1-1202-U4-029                NSCW SYS HOSE ISO TRAIN A DG BLDG    DG B R-101 2-1202-U4-029                NSCW TRAIN B SPLY TO DG B FIRE HOSE CONN                                  DG B R-103 Printed January 28, 2014 at 11:17
: 1. G2.4.45 001/LOIT/RO/C/A 4.1/4.3/G2.4.45/LO-TA-60019///
Initial conditions:
            - Unit 1 is at 100% reactor power.
            - ALB13-A01 STM GEN 1 FLOW MISMATCH is received.
Current conditions:
            - RCS pressure is 2237 psig and rising.
            - Main generator output is 1225 MWe and rising.
Which one of the following completes the following statement?
A __(1)__ line break is in progress, and per 18008-C, "Secondary Coolant Leakage," a main steam line isolation __(2)__
required to be manually actuated following the manual reactor trip.
__(1)__                                __(2)__
A.                    feed                                      is B.                    feed                                  is NOT C.                    steam                                      is D.                    steam                                  is NOT K/A G2.4.45          Ability to prioritize and interpret the significance of each annunciator or alarm.
K/A MATCH ANALYSIS The questions tests the candidate's ability to interpret annunciator ALB13-A01 STM GEN 1 FLOW MISMATCH utilizing current plant condtions, and respond appropriately.
EXPLANATION OF REQUIRED KNOWLEDGE ALB13-A01 STM GEN 1 FLOW MISMATCH simply detects a mismatch between steam and feedwater flows in each loop. Plant conditions must be referenced and interpreted to distinguish between the various accidents and malfunctions. The stem gives RCS pressure slightly above normal and rising with Turbine load at 1211 MWe, which is slightly above normal and rising. Both of these conditions are symptoms of a heat up of the RCS, which would either be a decrease in feed water flow or a decrease in steam flow. Since a decrease in steam flow would also cause a reduction in Turbine Thursday, March 06, 2014 12:50:51 PM                                                          1
 
load, the event must be a reduction in feed flow by either a break or restriction.
Per 18008-C step 1, if leakage is hazardous to personnel or equipment, the reactor is tripped and a Main Steam Line isolation performed after verification of the reator trip.
Then EOP 19000-C, "Reactor Trip or Safety Injection" is entered. The determination of "hazardous" is based on the potential to cause damage or injury. Operators are trained that if a steam or feed leak can be detected using the control room instruments and there is no evidence that the leak is being directed through a tailpipe, then the leak is considered hazardous. No tailpipe flow paths exist for feedwater, therefore the crews' ability to see the response makes it hazardous by definition. This originated from industry OE at Oconee Nuclear Plant, where 8 personnel were killed due to an extraction steam leak in the Turbine Building with the leak not even noticeable in the Main Control Room.
Main steam lines are required to be isolated on a feedline break to prevent blowing down all steam generators once the faulted stream generator is blown down through the connecting steam line piping.
ANSWER / DISTRACTOR ANALYSIS A. Correct                  The first part is correct. ALB13-A01 STM GEN 1 FLOW MISMATCH in conjunction with elevated and rising pressurizer pressure and turbine load are symptoms of a feed line break.
The second part is correct. 18008-C step 1 requires tripping the reactor, verifying the reactor tripped, and isolating steam lines even for feedwater line breaks.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. 118008-C step 1 requires tripping the reactor, verifying the reactor tripped, and isolating steam lines even for feedwater line breaks. However, it is reaonsable for a candidate to not consider the back flow from the other SGs following blow down and conclude that a MSLI is not required.
No steam flow will be seen from the other SGs until the faulted SG blows down all its inventory and creates a pathway for reverse flow. Also, the candidate may believe that the resulting P-4/Lo Tavg automatic feedwater isolation resulting from the reactor trip will isolate the feed line break and a MSLI would not be required.
C. Incorrect. Plausible. The first part is incorrect. ALB13-A01 STM GEN 1 FLOW MISMATCH in conjunction with elevated and rising pressurizer pressure and turbine load are symptoms of a feed line break.
However, a candidate with insufficient diagnostic skills could conclude that an increase in steam flow will result in an increase in reactor power and erroneously conclude that the increase in reactor power will result in an increase in RCS temperature. Additionally, 18008-C is most commonly entered for steam line leaks during scenario training.
Thursday, March 06, 2014 12:50:51 PM                                                                  2
 
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
LLevel:                        RO Tier # / Group #                T3 K/A#                            G2.4.45 Importance Rating:              4.1 / 4.3 Technical
 
==Reference:==
AOP 18008-C, Rev 9.2, pages 1 & 3 ARP 17001-1, Rev 31.0, page 7 References provided:            None Learning Objective:            LO-LP-37121-05 Describe the plant response to the following conditions:
: a. steam line break vs feed line break
: b. break at end of life vs break at beginning of life core
: c. steam break at full power initially vs zero power
: d. feed break inside last check valve vs feed break outside last check valve
: e. steam break between SG and first MSIV vs steam break between SG and outside last MSIV LO-TA-60019        Respond to a Loss of Secondary Coolant per 18008-C Question origin:                MODIFIED - HL17 # 040AA1.20 Cognitive Level:                C/A 10 CFR Part 55 Content:        41.4 / 41.5 / 41.10 / 43.5 /45.3 / 45.12 Comments:
You have completed the test!
Thursday, March 06, 2014 12:50:51 PM                                                                  3
: 1. 040AA1.20 001/1/1/SLB-CNMT PRESS/TEMP/H-4.1/4.2/NEW/H-17 NRC/RO/SRO/TNT/GCW Initial conditions:
Original Question
        - Unit 1 is at 100% power.
Current conditions:
        - Reactor power is 100.4% and slowly rising.
        - RCS pressure is 2212 psig and slowly lowering.
        - Turbine load is 1200 MWe and lowering.
        - Containment pressure is 1.4 psig and rising.
        - Containment temperature is 117.5oF and rising.
Which one of the following correctly completes the following statement?
A ___(1)___ break is the event in progress and per 18008-C, "Secondary Coolant Leakage", the FIRST action the operators will perform is to ___(2)___ .
__(1)__                        __(2)__
A. steamline                  reduce turbine load B. steamline                  manually insert rods C. feedline                    reduce turbine load D. feedline                    manually insert rods Wednesday, March 05, 2014 12:24:23 PM                                                1
 
Approved By                                                                      Procedure      Version S.A. Phillips                      Vogtle Electric Generating Plant              18008-C          9.2 Effective Date                                                                    Page Number 08/14/2012 SECONDARY COOLANT LEAKAGE                                1 of 6 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure specifies operator actions for secondary leaks which do NOT actuate Engineered Safeguards Features.
SYMPTOMS ALB13-A01 (B01, C01, D01) STM GEN 1 (2, 3, 4) FLOW MISMATCH ALB13-A06 (B06, C06, D06) STM GEN 1 (2, 3, 4) HI/LO LVL DEVIATION High containment pressure, temperature, moisture, and sump levels WITHOUT radiation.
High condenser hotwell makeup rates.
High CST makeup rates.
Observed secondary leakage.
Unexplained rise in reactor power.
Reactor power significantly higher than turbine power.
MAJOR ACTIONS Evaluate and stabilize plant conditions.
Locate and isolate leakage.
Determine whether to continue operation or initiate plant shutdown.
__
Printed March 5, 2014 at 13:14
 
Approved By                                                                    Procedure      Version S.A. Phillips                            Vogtle Electric Generating Plant      18008-C          9.2 Effective Date                                                                  Page Number SECONDARY COOLANT LEAKAGE 08/14/2012                                                                              3 of 6 ACTION/EXPECTED RESPONSE                            RESPONSE NOT OBTAINED 1
: 1.      Perform the following as necessary:          1.
Reduce Turbine load if any of the following indications exceed 100% power:
UQ1118 (GREATER THAN 100% MWT)
NIs Ts Isolate the leak.
IF leakage is such that significant hazard to personnel or equipment exists OR leakage rate is unstable and is worsening, THEN:
1.1)
: 1)      Trip the reactor.                        1) 1.2)
: 2)      WHEN reactor trip is                    2) verified, THEN close MSIVs and BSIVs.
1.3)
: 3)      Go to 19000 - C, E - 0                  3)
REACTOR TRIP OR SAFETY INJECTION.
2
: 2.      Initiate the Continuous Actions Page.        2.
 
S Printed March 5, 2014 at 13:14
 
Approved By                                                                                Procedure  Version M. C. Henry                        Vogtle Electric Generating Plant                        17013-1      31 Effective Date              ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 13 ON PANEL            Page Number 12/23/13                                              1B1 ON MCB                                    7 of 77 WINDOW A01 DIGITAL FEEDWATER POINT NAME                SETPOINT DFW1FLOWMM                                  N/A            STM GEN 1 FLOW MISMATCH 1.0                PROBABLE CAUSE
: 1.        Steam Generator 1 Feedwater Level Control System malfunction.
: 2.        Main Feedwater System malfunction.
: 3.        Feedwater/Steam leak.
: 4.        Main Steam Isolation Valve malfunction.
2.0                AUTOMATIC ACTIONS IF turbine power is greater than C-20 (40% on PT505 and PT506), an AMSAC actuation will occur after a time delay IF the feedwater flow on any of three loops is less than 25% of normal 100% feedwater flow. The time delay ramps from 27.5 seconds at 100% turbine power to 230 seconds at 40% turbine power.
3.0                INITIAL OPERATOR ACTIONS
: 1.        Verify all MSIVs are open. IF any MSIV has failed shut, THEN trip the reactor AND Go To 19000-C, E-0 Reactor Trip or Safety Injection.
: 2.        Check Steam Generator 1 feedflow AND steamflow indications:
: a. IF leakage is indicated, Go To 18008-C, "Secondary Coolant Leakage".
: b. IF a Main Feedwater System malfunction is indicated, Go To 18016-C, "Condensate And Feedwater Malfunction".
: 3.        Verify proper operation of 1-FV-4486. IF malfunction is indicated, THEN take manual control of 1-FIC-4486 AND control as necessary to maintain MFPT suction pressure greater than 300 psig AND flow as indicated on 1-FIC-4486 greater than 7400 gpm.
Printed February 10, 2014 at 14:07
: 1. WE02EA1.03 001/LOIT AND LOCT/RO/C/A 3.8/4.0/WE02EA1.03/LO-TA-37009///
Initial condition:
            - 19010-C, "Loss of Reactor and Secondary Coolant," is in use.
Current conditions:
            - Step 11, "Check if ECCS flow should be reduced," is in progress with the following conditions:
                  - RCS pressure is 1335 psig and stable.
                  - Pressurizer level is 30% and stable.
                  - CETCs are 550&deg;F.
                  - WR THot is 540&deg;F.
                  - Containment pressure is 4.5 psig.
                  - All SG NR levels are <10% with 200 gpm AFW flow to each SG.
Which one of the following completes the following statement?
Based on the current conditions, ECCS flow __(1)__ be reduced because __(2)__.
A. (1) can (2) all termination criteria have been satisfied B. (1) can NOT (2) pressurizer level and RCS subcooling do not meet termination criteria C. (1) can NOT (2) RCS subcooling and secondary heat sink do not meet termination criteria D. (1) can NOT (2) pressurizer level is the only termination criterion not met Friday, March 07, 2014 8:01:51 AM                                                        1
 
K/A WE02            SI Termination EA1.03          Ability to operate and / or monitor the following as they apply to the (SI Termination):
                          - Desired operating results during abnormal and emergency situations.
K/A MATCH ANALYSIS The question test the candidate's ability to monitor current plant conditions and determine if SI Termination criteria are met, and if not, which condition(s) is(are) not met.
EXPLANATION OF REQUIRED KNOWLEDGE EOP 19010-C step 11 checks to see if ECCS flow should be reduced. The following criteria must be met:
                  - RCS Subcooling GREATER THAN 24 F [38 F ADVERSE].
                  - Total feed flow to intact SG(s) GREATER THAN 570 GPM.
                                          -OR-
                  - NR level in at least one intact SG GREATER THAN 10% [32% ADVERSE].
                  - RCS pressure STABLE OR RISING.
                  - PRZR level GREATER THAN 9% [37% ADVERSE].
If all these criteria are met, a transition to 19011-C will be made and ECCS flow will be reduced.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Both pressurizer level and RCS subcooling (32.4&deg;F) are below the required values. However, if the candidate failed to recognize containment pressure is above adverse value then all termination criteria would appear to be met and ECCS flow could be reduced.
B. Correct.                ECCS flow cannot be reduced. Both pressurizer level and RCS subcooling are below the required values. Since containment pressure is >3.8 psig, adverse values must be used.
Pressurizer level is required to be 37% and based on the stem is currently 30%. RCS subcooling is required to be 38&deg;F and based on stem conditions is calculated to be 32.4&deg;F. RCS pressure conditions and secondary heat sink are within termination limits.
C. Incorrect. Plausible. ECCS flow cannot be reduced. Both pressurizer level and RCS subcooling are below the required values. Since containment Friday, March 07, 2014 8:02:30 AM                                                                1
 
pressure is >3.8 psig, adverse values must be used.
Pressurizer level is required to be 37% and based on the stem is currently 30%. RCS subcooling is required to be 38&deg;F and based on stem conditions is calculated to be 32.4&deg;F. RCS pressure conditions and secondary heat sink are within termination limits. However, if the candidate remembers the pressurizer level value incorrectly and mis-reads the AFW flow as 200 gpm total instead of 200 gpm per SG, then the candidate could believe termination criteria are not met on heat sink and RCS subcooling.
D. Incorrect. Plausible. ECCS flow cannot be reduced. Both pressurizer level and RCS subcooling are below the required values. Since containment pressure is >3.8 psig, adverse values must be used.
Pressurizer level is required to be 37% and based on the stem is currently 30%. RCS subcooling is required to be 38&deg;F and based on stem conditions is calculated to be 32.4&deg;F. RCS pressure conditions and secondary heat sink are within termination limits. However, if the candidate mis-calculates subcooling by using WR THot insted of CETs, the subcooling would be calculated as 42.4&deg;F and only pressurizer level would not be met.
Friday, March 07, 2014 8:02:31 AM                                                              2
 
Level:                    RO Tier # / Group #          T1 / G2 K/A#                      WE02EA1.03 Importance Rating:        3.8 / 4.0 Technical
 
==Reference:==
19011-C, Rev. 34.3, page 9 Properties of Saturated and Superheated Steam, 2000 ASME STEAM TABLES References provided:      Properties of Saturated and Superheated Steam, 2000 ASME STEAM TABLES Learning Objective:        LO-LP-37311-08 State how termination/reduced ECCS flow is accomplished. State any differences in the methods used in 19030-C with the directions provided in 19011-C and 19012-C.
LO-LP-37111-08 Using 19010-C as a guide, briefly describe how each step is accomplished.
LO-TA-37015        Perform the Initial Recovery Actions for a small Loss of Reactor or Secondary Coolant per 19010-C LO-TA-37009        Respond to a Large Break Loss of Primary Coolant per 19010-C Question origin:          MODIFIED - HL18 # WE02EA2.1 Cognitive Level:          C/A 10 CFR Part 55 Content:    41.8 / 41.10 / 45.5 / 45.6 Comments:
You have completed the test!
Friday, March 07, 2014 8:02:31 AM                                                              3
: 1. WE02EA2.1 001/1/2/SI TERM - FUNCTIONS/C/A - 3.3/4.2/BANK-FARLEY/HL-18 NRC/RO/SRO/AML The following conditions exist on Unit 1:
Original Question
          - A LOCA is in progress.
          - Main Steam Line Isolation has occurred due to Containment pressure.
          - 19010-C, "Loss of Reactor or Secondary Coolant," is in progress.
The crew is at the step to, "Check if ECCS flow should be reduced," with plant parameters as follows:
          - RCS pressure is 1725 psig and stable.
          - CETCs indicate 570&deg;F.
          - Total available AFW flow is 580 gpm.
          - SG NR levels are all between 12 - 15%.
          - PZR level is 30% and slowly rising.
Based on the current conditions, which ONE of the following actions are the operators required to take?
A. Continue in 19010-C.
B. Transition to 19011-C, "SI Termination."
C. Transition to 19012-C, "Post-LOCA Cooldown and Depressurization."
D. Transition to 19231-C, "Response to Loss of Secondary Heat Sink."
Thursday, March 06, 2014 11:32:18 AM                                                        1
 
Approved By                                                                                Procedure    Version J. B. Stanley                          Vogtle Electric Generating Plant                    19010-C        34.3 Effective Date                                                                              Page Number E-1 LOSS OF REACTOR OR SECONDARY 7/25/12                                            COOLANT                                        9 of 27 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 11
        *11. Check if ECCS flow should be                      11.
reduced:
11.a
: a. RCS Subcooling - GREATER                            a. Go to Step 12.
THAN 24F [38F ADVERSE].
11.b
: b. Secondary Heat Sink:                                b. Go to Step 12.
Total feed flow to intact SG(s) -
GREATER THAN 570 GPM.
                                    -OR-NR level in at least one intact SG - GREATER THAN 10%
[32% ADVERSE].
11.c
: c. RCS pressure - STABLE OR                            c. Go to Step 12.
RISING.
11.d
: d. PRZR level - GREATER THAN                          d. Try to stabilize RCS pressure:
9% [37% ADVERSE].
Use Normal PRZR Spray if Instrument Air to Containment available.
Do NOT use PRZR PORVs to stabilize RCS pressure.
Go to Step 12.
11.e
: e. Go to 19011-C, ES-1.1 SI                            e.
TERMINATION.
12
        *12. Check if Containment Spray                        12.
should be stopped:
12.a
: a. CS Pumps - RUNNING.                                a. Go to Step 13.
 
Step 12 continued on next page Printed January 28, 2014 at 10:03
: 1. WE04EK1.03 001/LOIT/RO/M/F 3.5/3.9/WE04EK1.03/LO-TA-37020A//HL18 NRC/WE04EK2.01 Initial condition:
            - The crew is performing 19112-C, "LOCA Outside Containment."
Current condition:
            - RCS pressure is 1500 psig.
Which one of the following completes the following statement?
The FIRST system to be isolated from the RCS to attempt leak isolation is __(1)__,
and the instrument that will be used to determine isolation of the leak is __(2)__.
__(1)__                                __(2)__
A.                      SI                          pressurizer pressure B.                      SI                            RCS WR pressure C.                    RHR                          pressurizer pressure D.                    RHR                            RCS WR pressure K/A WE04            LOCA Outside Containment EK1.03          Knowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment):
                          - Annunciators and conditions indicating signals, and remedial actions associated with the LOCA Outside Containment K/A MATCH ANALYSIS The question requires the candidate to recall the operational implications associated with the order in which leakage is systematically located and isolated during a LOCA outside containment. The candidate must also recall the associated indications and conditions of successfully isolating the leak during remedial actions.
EXPLANATION OF REQUIRED KNOWLEDGE Per step 2 of 19112-C, RHR Hot Leg suction and discharge and SIP Hot Leg discharge valves are verified CLOSED in that order. These valves are expected to be CLOSED and therefore no action is required. However, direction is given to CLOSE the valves if Friday, March 07, 2014 8:03:55 AM                                                              1
 
necessary. Step 3 CLOSES RHR Cold Leg discharge valves one at a time. Step 4 CLOSES SI Cold Leg discharge valves one at a time. Therefore, whether the candidate considers the question to address the Hot Leg checks or the Cold Leg manipulations as the FIRST action, RHR is always first.
Following each of the isolation steps, RCS pressure is checked. Rising RCS pressure indicates the leak is isolated. RCS WR pressure is utilized since Pressurizer Pressure is off-scale low at 1700 psig. With plant pressure below 1700 psig, Pressurizer Pressure will continue to indicate 1700 psig.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. The RHR valves are the first to be checked. The SI valves are checked next.
The second part is incorrect. RCS WR pressure is checked to determine if the leak is isolated. Pressurizer pressure is offscale low and indicating 1700 psig at the current plant conditions. However, pressurizer pressure is normally monitored to determine RCS Pressure.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is correct. RCS WR pressure is monitored to determine if the leakage is isolated as described in the Explanation of Required Knowledge above.
C. Incorrect. Plausible. The first part is correct. The RHR valves are the first to be checked or operated as described in the Explanation of Required Knowledge above.
The second part is incorrect. See the second part of choice A above.
D. Correct.              The RHR pump valves are the first to be checked or operated as described in the Explanation of Required Knowledge above.
Addtionally, RCS WR pressure is monitored to determine if the leak is isolated.
Friday, March 07, 2014 8:03:55 AM                                                                  2
 
Level:                    RO Tier # / Group #          T1 / G1 K/A#                      WE04EK1.03 Importance Rating:        3.5 / 3.9 Technical
 
==Reference:==
EOP 19112-C, Rev 6.2, pages 2 thru 5 References provided:      None Learning Objective:        LO-PP-37116-02 Describe the steps taken to isolate a LOCA outside containment.
LO-PP-37116-03 Describe the indications used to confirm that a LOCA outside containment was successfully isolated.
LO-TA-37020A        Respond to a LOCA Outside Contament per 19112-C.
Question origin:          BANK - Direct Reuse of HL18 NRC Question # WE04EK2.01 Cognitive Level:          M/F 10 CFR Part 55 Content:    41.8 / 41.10 / 45.3 Comments:
You have completed the test!
Friday, March 07, 2014 8:03:55 AM                                                          3
 
Approved By                                                                              Procedure      Version J.B. STANLEY                          Vogtle Electric Generating Plant                    19112-C          6.2 Effective Date                                                                            Page Number ECA-1.2 LOCA OUTSIDE CONTAINMENT 7/25/12                                                                                            2 of 7 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 1
: 1. Verify SI reset.                                  1. IF SI will NOT reset, ,
THEN initiate ATTACHMENT A.
2
: 2. Verify proper valve alignment:                    2.
2.a
: a. RHR Pump suction from RCS -                          a.
CLOSED:
HV-8701A - RHR PMP-A DOWNSTREAM SUCTION FROM HOT LEG LOOP-1 HV-8701B - RHR PMP-A UPSTREAM SUCTION FROM HOT LEG LOOP-1 HV-8702A - RHR PMP-B DOWNSTREAM SUCTION FROM HOT LEG LOOP-4 HV-8702B - RHR PMP-B UPSTREAM SUCTION FROM HOT LEG LOOP-4 2.b
: b. RHR Pump Hot Leg injection                          b. Dispatch an Operator to close valve - CLOSED:                                      affected Unit valve:
HV-8840 - RHR TO HL ISO                        1-HV-8840 - RHR TO HL ISO VLV                                              VLV (AB-A13) 2-HV-8840 - RHR TO HL ISO VLV (AB-A18)
 
Step 2 continued on next page Printed November 14, 2013 at 14:29
 
Approved By                                                                              Procedure      Version J.B. STANLEY                          Vogtle Electric Generating Plant                  19112-C          6.2 Effective Date                                                                            Page Number ECA-1.2 LOCA OUTSIDE CONTAINMENT 7/25/12                                                                                          3 of 7 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 2.c
: c. SI Pump Hot Leg injection valves                    c. Dispatch an Operator to close
                      - CLOSED:                                            affected Unit valves:
HV-8802A - SI PMP-A TO                        1-HV-8802A - SI PMP-A TO HOT LEG 1&4 ISO VLV                            HOT LEG 1&4 ISO VLV (AB-A09)
HV-8802B - SI PMP-B TO 1-HV-8802B - SI PMP-B TO HOT LEG 2&3 ISO VLV HOT LEG 2&3 ISO VLV (FHB-A10) 2-HV-8802A - SI PMP-A TO HOT LEG 1&4 ISO VLV (AB-A103) 2-HV-8802B - SI PMP-B TO HOT LEG 2&3 ISO VLV (FHB-A01) 3
: 3. Try to identify and isolate RHR Cold              3.
Leg injection break:
3.a
: a. Close RHR PMP-A TO COLD                            a.
LEG 1&2 ISO VLV HV-8809A.
3.b
: b. Check RCS pressure - RISING.                        b. Open RHR PMP-A TO COLD LEG 1&2 ISO VLV HV-8809A.
Go to Step 3.d.
3.c
: c. Go to Step 3.f.                                    c.
3.d
: d. Close RHR PMP-B TO COLD                            d.
LEG 3&4 ISO VLV HV-8809B.
3.e
: e. Check RCS pressure - RISING.                        e. Open RHR PMP-B TO COLD LEG 3&4 ISO VLV HV-8809B.
Go to Step 4.
 
Step 3 continued on next page Printed November 14, 2013 at 14:29
 
Approved By                                                                              Procedure      Version J.B. STANLEY                          Vogtle Electric Generating Plant                  19112-C          6.2 Effective Date                                                                            Page Number ECA-1.2 LOCA OUTSIDE CONTAINMENT 7/25/12                                                                                          4 of 7 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 3.f
: f. Stop RHR Pump in train with leak                  f.
isolated.
3.g
: g. Go to Step 5.                                      g.
4
: 4. Try to identify and isolate SI Cold Leg          4.
injection break:
4.a
: a. Close SI PMP-A TO COLD LEG                          a.
ISO VLV HV-8821A.
4.b
: b. Check RCS pressure - RISING.                        b. Open SI PMP-A TO COLD LEG ISO VLV HV-8821A.
Go to Step 4.d.
4.c
: c. Go to Step 4.k.                                    c.
4.d
: d. Close SI PMP-B TO COLD LEG                          d.
ISO VLV HV-8821B.
4.e
: e. Check RCS pressure - RISING.                        e. Open SI PMP-B TO COLD LEG ISO VLV HV-8821B.
Go to Step 4.g.
4.f
: f. Go to Step 4.k.                                  f.
4.g
: g. Close COLD LEG INJECTION                            g.
FROM SIS HV-8835.
4.h
: h. Check RCS pressure - RISING.                        h. Open COLD LEG INJECTION FROM SIS HV-8835.
Go to Step 5.
4.i
: i. Stop both SI pumps.                              i.
4.j
: j. Go to Step 5.                                    j.
 
Step 4 continued on next page Printed November 14, 2013 at 14:29
 
Approved By                                                                      Procedure      Version J.B. STANLEY                          Vogtle Electric Generating Plant          19112-C          6.2 Effective Date                                                                    Page Number ECA-1.2 LOCA OUTSIDE CONTAINMENT 7/25/12                                                                                  5 of 7 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 4.k
: k. Stop SI Pump in train with leak              k.
isolated.
5
: 5. Check if break is isolated:                5.
5.a
: a. Check RCS pressure - RISING.                  a. Go to 19111-C, ECA-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION.
5.b
: b. Go to 19010-C, E-1 LOSS OF REACTOR OR SECONDARY COOLANT.
 
END OF PROCEDURE TEXT Printed November 14, 2013 at 14:29
: 1. WE05EA1.02 001/LOIT AND LOCT/RO/C/A 3.7/4.0/WE05EA1.02/LO-TA-37051//HL-18 NRC/054AA1.04 Initial conditions:
            -  Unit 1 reactor tripped.
            -  19231-C, "Response to Loss of Secondary Heat Sink," is in use.
            -  Bleed and Feed has been initiated.
            -  1AA02 is faulted.
            -  SIP 'B' is running.
            -  1PORV-455 is CLOSED.
            -  1PORV-456 is OPEN.
Current conditions:
            - ALB07-C06 CHARGING PUMP OVERLOAD TRIP is received.
            - ALB07-B06 CHARGING LINE HI/LO FLOW is received.
            - CCP 'B' handswitch green and amber lights are LIT.
Which one of the following completes the following statement?
Per 19231-C, the minimum requirement for the RCS Feed path __(1)__ met, and the Reactor Head vents must be __(2)__ to prevent core uncovery.
__(1)__                      __(2)__
A.                is                      opened B.                is                        closed C.            is NOT                      opened D.            is NOT                      closed K/A WE05            Loss of Secondary Heat Sink EA1.02          Ability to operate and / or monitor the following as they apply to the (Loss of Secondary Heat Sink):
                          - Operating behavior characteristics of the facility K/A MATCH ANALYSIS The question tests the candidates ability to monitor plant conditions to determine if adequate feed flow exists. Additionally, the question requires the candidate to relate Friday, March 07, 2014 8:27:57 AM                                                              1
 
these plant conditions to the need to initiate feed and bleed and open head vents when CCPs and both PORVs are not available respectively.
EXPLANATION OF REQUIRED KNOWLEDGE During a Loss of Secondary Heat Sink event, RCS bleed and feed must be initiated if 3 of 4 SG WR levels lower to les than 29%. This ensures that RCS pressure will be lowered sufficiently to ensure adequate ECCS injection flow is maintained. The PORVs at Vogtle are insufficiently sized to maintain this state indefinitely. If prompt operator action does not occur, RCS pressures will rise to near or above the shutoff head of the CCPs. Eventually, a Loss of Core Cooling (and subsquent core uncovery) will occur.
The event is worsened if a CCP or either PORV is not available. In the event both CCPs are lost, an immediate transition to bleed and feed is required to ensure RCS pressure is lowered below the shutoff head of the SIPs. If performed in a timely manner, the SIPs are an adequate feed source. If either PORV is unavailable, Reactor Head vents are opened to increase the bleed flow rate. However, one PORV and all head vents are not an adequate bleed path and introduction of low pressure water sources to a SG must be initiated immediately due to the increased risk for a Loss of Core Cooling and the reduced time available to implement this contingency action.
Per 19231-C the procedural flow path for the listed plant conditions is as follows:
Step 5            - determine CCPs not available. Stop all RCPs and go to step 33.
Step 34            - actuate SI, which will ensure all available ECCS pumps are running.
SIP B is available.
Step 36 & 37 - establish a bleed path through block valves and PORVS. OPEN Rx Head vents if either PORV is not available.
Step 40            - maintain RCS heat removal with SIP flow and PORVs and/or Rx Head vents ANSWER / DISTRACTOR ANALYSIS A. Correct.                The first part is correct. Per the current plant conditions, only SIP B is available. With NO CCPs available, bleed and feed must be initiate immediately to reduce RCS pressure below SIP shutoff head. This action establishes an adequate feed source as described in the procedure actions listed in the Explanation of Required Knowledge above.
The second part is correct. Since both PORVs could not be OPENED, the RNO of step 37 will OPEN all the Rx Head vents.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. Since both PORVs could not be OPENED, the RNO of step 37 will OPEN all the Rx Head vents.
However, when injection flow is lost in 19100-C, Rx Head vents are verified CLOSED to maximize the time to core uncovery.
Additionally, the candidate may not recognize (1) PORV is an inadequate bleed path and believe Rx Head vents should remain CLOSED.
Friday, March 07, 2014 8:27:57 AM                                                                      2
 
C. Incorrect. Plausible. The first part is incorrect. Per the current plant conditions, only SIP B is available. With NO CCPs available, bleed and feed must be initiate immediately to reduce RCS pressure below SIP shutoff head. This action establishes an adequate feed source as described in the procedure actions listed in the Explanation of Required Knowledge above. However, the candidate may believe that, since bleed and feed is immediately initiated with NO CCPs available, a single SIP alone would not constitute an adequate feed source.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Friday, March 07, 2014 8:27:57 AM                                                                  3
 
Level:                    RO Tier # / Group #          T1 / G1 K/A#                      WE05EA1.02 Importance Rating:        3.7 / 4.0 Technical
 
==Reference:==
WOG FR-H.1 Background Rev 2, 4/30/2005 WOG ECA 0.0 Background Rev 2, 4/30/2005, page 88 EOP 19100-C, Rev 38.1, pages 6 & 7 EOP 19231-C, Rev 34.0, pages 5 & 20 thru 22 References provided:      None Learning Objective:        LO-LP-37051-11 Define loss of secondary heat sink in accordance with 19231-C, "Response to Loss of Secondary Heat Sink," requiring immediate initiation of bleed and feed control.
LO-LP-37051-08 Using 19231-C as a guide, briefly describe how each major step is accomplished. Describe the bases for each.
LO-TA-37051        Respond to a Loss of Secondary Heat Sink per 19231-C.
Question origin:          MODIFIED - HL-18 NRC Question # 054AA1.04, HL17 NRC Question # WE05EK2.1 Cognitive Level:          C/A 10 CFR Part 55 Content:    41.7 / 41.10 / 45.5 / 45.6 Comments:
You have completed the test!
Friday, March 07, 2014 8:27:57 AM                                                              4
: 1. 054AA1.04 001/1/1/LOSS MFW - HPI/C/A - 4.4/4.5/MOD - HL17 NRC/HL-18 NRC/RO/SRO/KAJ Initial conditions:
          - 19231-C, "Response to Loss of Secondary Heat Sink", is in use.
          - Bleed and Feed has been initiated.
          - SIPs are NOT available.
          - Both CCPs are running.
          - PORV-455 is CLOSED.                              HL18 Original Question
          - PORV-456 is OPEN.
Current conditions:
          - ALB07-C06, CHARGING PUMP OVERLOAD TRIP illuminates.
          - 'B' CCP handswitch green and amber lights are LIT.
Based on the current conditions, which one of the following completes the following statement?
Per 19231-C, the minimum requirement for the RCS Feed path ___(1)___ met, and the minimum requirement for the RCS Bleed path ___(2)___ met.
__(1)__                      __(2)__
A.      is                        is B.      is                        is NOT C. is NOT                        is D. is NOT                        is NOT Wednesday, November 13, 2013 4:12:50 PM                                                    1
: 1. WE05EK2.1 001/1/1/LOHS - COMPONENTS/H-3.7/3.9/NEW/H-17 NRC/RO/SRO/TNT/GCW Initial conditions:
        - 19231-C, "FR-H.1 Response to Loss of Secondary Heat Sink" is in progress.
        - RCS Bleed and Feed has been initiated.
Current conditions:                                HL17 Original Question
        - One SIP is running.
        - The CCPs and other SIP are NOT available.
        - One PRZR PORV is open.
        - The other PORV is NOT available.
Per 19231-C, which one of the following correctly completes the following statement?
One SIP running is ___(1)___ for the RCS Feed path and one PRZR PORV open is ___(2)___ for the RCS Bleed path.
For the purposes of this question, "adequate" as defined by procedure 19231-C means that 19231-C will NOT direct further adjustments to the RCS Feed or Bleed path.
__(1)__                      __(2)__
A. adequate                      adequate B. adequate                      NOT adequate C. NOT adequate                  adequate D. NOT adequate                  NOT adequate Wednesday, November 13, 2013 4:11:07 PM                                                    1
 
ECA - 0.0 Loss of all AC Power                                              19100-C VOGTLE                    Version 38.1 Unit C                Page 6 of 50 IMMEDIATE OPERATOR ACTIONS ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED NOTE CSFSTs should be monitored for information only. Function Restoration Procedures (FRP) should not be implemented.
: 1. Verify Reactor trip:                        1. Trip Reactor using both Reactor trip handswitches.
Reactor Trip and Bypass Breakers -
OPEN.                                  IF Reactor NOT tripped, THEN dispatch an Operator to locally trip Neutron Flux - LOWERING.                    the Reactor Trip and Bypass Breakers.
: 2. Verify Turbine trip:                        2. Trip Turbine.
All Turbine Stop Valves - CLOSED.          IF Turbine will NOT trip, THEN run back Turbine.
IF Turbine can NOT be run back, THEN close Main Steamline Isolation and Bypass Valves.
SUBSEQUENT OPERATOR ACTIONS ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED
: 3. Check if RCS is isolated:                    3.
: a. PRZR PORVs - CLOSED.                        a. IF PRZR pressure is less than 2315 psig, THEN verify closed affected PRZR PORV(s).
Perform the following to isolate affected PORV as necessary:
Open affected PORV power supply breaker:
AD1M-04 (PV-455A)
BD1M-04 (PV-456A)
Printed March 7, 2014 at 08:14
 
ECA - 0.0 Loss of all AC Power                            19100-C VOGTLE            Version 38.1 Unit C          Page 7 of 50 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE              RESPONSE NOT OBTAINED
: b. Letdown Orifice Isolation Valves - b. Close valves.
CLOSED:
HV-8149A - LETDOWN ORIFICE 45 GPM HV-8149B - LETDOWN ORIFICE 75 GPM HV-8149C - LETDOWN ORIFICE 75 GPM
: c. Letdown Isolation Valves - CLOSED: c. Close valves.
LV-0459 - LETDOWN ISOLATION VLV DOWNSTREAM LV-0460 - LETDOWN ISOLATION VLV UPSTREAM
: d. Excess Letdown Isolation Valves -  d. Close valves.
CLOSED:
HV-8153 - EXCESS LETDOWN LINE ISO VLV HV-8154 - EXCESS LETDOWN LINE ISO VLV
: e. Reactor Vessel Head Vent Isolation e. Close valves.
Valves - CLOSED:
HV-8095A - RX HEAD VENT TO LETDOWN ISOLATION VLV HV-8095B - RX HEAD VENT TO LETDOWN ISOLATION VLV HV-8096A - RX HEAD VENT TO LETDOWN ISOLATION VLV HV-8096B - RX HEAD VENT TO LETDOWN ISOLATION VLV Printed March 7, 2014 at 08:14
 
FR - H 1 Response to Loss of Secondary Heat Sink                                            19231-C VOGTLE                      Version 34 Unit C                  Page 8 of 57 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED
        *5. Check CCP status - AT LEAST ONE            NO *5. Stop all RCPs.
AVAILABLE.
Go to Step 33.
        *6. Check if RCS bleed and feed is required:        6.
: a. Check the following:                          a. WHEN criteria for bleed and feed are met, WR level in any 3 SGs - LESS THAN                THEN perform Step 6.b and Step 6.c.
29% [44% ADVERSE].
Go to Step 7.
                      -OR-RCS pressure - GREATER THAN 2335 PSIG DUE TO LOSS OF SECONDARY HEAT SINK
: b. Trip all RCPs.                                b.
: c. Go to Step 33 and perform bleed and          c.
feed actions.
: 7. Place Containment Hydrogen Monitors in          7.
service by initiating 13130 Post-Accident Hydrogen Control.
        *8. Check CST level - GREATER THAN                *8. Swap to alternate CST by initiating 15%.                                            13610 Auxiliary Feedwater System.
: 9. Verify SG Blowdown isolated:                    9.
SG Blowdown Isolation Valves -
CLOSED WITH HANDSWITCHES IN CLOSE POSITION.
SG Sample Isolation Valves -
CLOSED.
Printed 03/07/2014 at 08:17:00
 
FR - H 1 Response to Loss of Secondary Heat Sink                                          19231-C VOGTLE                    Version 34 Unit C              Page 20 of 57 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED
: 32. Check for loss of secondary heat sink:        32. Return to Step 4.
WR level in any 3 SGs - LESS THAN 29% [44% ADVERSE].
                -OR-RCS pressure - GREATER THAN 2335 PSIG DUE TO LOSS OF SECONDARY HEAT SINK.
: 33. Initiate CONTINUOUS ACTIONS PAGE              33.
FOR AFTER ESTABLISHING BLEED AND FEED.
CAUTION Step 34 thru Step 37 should be performed quickly in order to establish RCS heat removal by RCS bleed and feed.
: 34. Verify SI actuated.                          34.
Printed 03/07/2014 at 08:17:00
 
FR - H 1 Response to Loss of Secondary Heat Sink                                            19231-C VOGTLE                      Version 34 Unit C                  Page 21 of 57 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                        RESPONSE NOT OBTAINED
: 35. Verify RCS feed path:                        35. Start pumps and align valves as necessary to establish injection flow using Attachment 1 or Attachment 2.
IF a feed path can NOT be established, THEN continue attempts to establish feed flow.
Return to Step 10.
: a. Verify ECCS Pump status:                    a.
CCPs - AT LEAST ONE RUNNING.
                      -OR-SI Pumps - AT LEAST ONE RUNNING.
: b. Verify ECCS valve alignment -              b.
PROPER INJECTION LINEUP INDICATED ON MLBs.
CAUTION During bleed and feed operation the PRT may rupture.
: 36. Establish RCS bleed path:                    36.
: a. Place all PRZR Heaters in OFF/PTL.          a.
: b. Check power to PRZR PORV Block              b. Restore power to block valves.
Valves - AVAILABLE.
: c. Arm COPS and check PRZR PORV                c. Open both PRZR PORV Block Valves.
Block Valves - BOTH OPEN.
: d. Open both PRZR PORVs.                      d.
Printed 03/07/2014 at 08:17:00
 
FR - H 1 Response to Loss of Secondary Heat Sink                                        19231-C VOGTLE                      Version 34 Unit C                  Page 22 of 57 SUBSEQUENT OPERATOR ACTIONS (continued)
ACTION/EXPECTED RESPONSE                      RESPONSE NOT OBTAINED
: 37. Verify adequate RCS bleed path:          37. Perform the following:
COPS - ARMED.
: a. Open Reactor Vessel Head Vent Valves:
PRZR PORV Block Valves - BOTH OPEN.                            NO HV-8095A - RX HEAD VENT TO LETDOWN ISOLATION VLV PRZR PORVs - BOTH OPEN.
HV-8095B - RX HEAD VENT TO LETDOWN ISOLATION VLV HV-8096A - RX HEAD VENT TO LETDOWN ISOLATION VLV HV-8096B RX HEAD VENT TO LETDOWN ISOLATION VLV HV-0442A - REACTOR HEAD VENT TO PRT HV-0442B - REACTOR HEAD VENT TO PRT
: b. Align an available low pressure water source to at least one intact SG by initiating Attachment 3.
: 38. Initiate Attachment 4 while continuing    38.
with this procedure.
: 39. Initiate CONTINUOUS ACTIONS AFTER        39.
ESTABLISHING BLEED AND FEED.
      *40. Maintain RCS heat removal:                40.
ECCS flow.
PRZR PORVs - BOTH OPEN.                    Maintain Reactor Vessel Head Vent Valves open.
Printed 03/07/2014 at 08:17:00
 
STEP DESCRIPTION TABLE FOR ECA0.0        Step  3 STEP:      Check If RCS Is Isolated PURPOSE:  To ensure all RCS outflow paths are isolated BASIS:
A check for RCS isolation is performed to ensure that RCS inventory loss is minimized. The valves itemized are those in major RCS outflow lines that could contribute to rapid depletion of RCS inventory. This step is written for plants which utilize air operated valves (AOVs) in the itemized locations. The step structure assumes that the AOVs fail closed on loss of all ac power (i.e., loss of air supply). The operator, therefore, checks that the valves are closed.
If any valve is open, the operator should attempt to close the valve.
Reasons for a valve remaining open are plant specific, for example the valves may have legitimate or spurious open signals and air pressure could be available due to air receivers or air bottles located in the air supply system. Plants with air receivers may take up to 30 minutes to lose air pressure. If nitrogen bottles are provided for specific valves, such as PORVs, pneumatic pressure may be available for more than 30 minutes.
The sequence for checking valves is based on capacity of the outflow lines and potential for RCS inventory loss:
: 1) The pressurizer PORVs are checked first. Since the turbinedriven AFW pump should be running, the secondary side is removing decay heat and RCS pressure should be under the pressurizer PORV setpoint.
: 2) The letdown line isolation valves adjacent to the RCS loop are checked second. These valves are normally open and receive a low pressurizer level isolation signal. If these valves, in conjunction with the letdown orifice isolation valves, remain open, a leak path to the pressurizer relief tank (PRT) via the letdown line relief valve may exist. These valves, including the letdown orifice isolation
: valves, if necessary, should be manually closed as soon as possible to ECA0.0 Background              88            HPRev. 2, 4/30/2005 HECA00BG.doc
 
about 40 lbm/sec (290 gpm), with both trains operating, at an RCS pressure of 2300 psig. Since makeup flow from the charging/SI pump system will not keep up with inventory lost out of the pressurizer PORVs, the RCS will eventually dry out enough to cause core uncovery.
In summary, the loss of all feedwater transient from a power condition without operator action will lead to a loss of secondary heat sink followed by a loss of RCS inventory through the pressurizer PORVs.
Core uncovery will result at an RCS pressure equal to or greater than the pressurizer PORV setpoint and charging/SI flow, if manually initiated late in the transient, will not be sufficient to prevent core uncovery.
2.2  RCS Bleed and Feed Heat Removal For a loss of all secondary heat sink, operator action to establish RCS bleed and feed heat removal can prevent or minimize core uncovery.
To establish RCS bleed and feed heat removal the operator must initiate and verify high pressure SI flow to feed subcooled fluid to the RCS and then manually open all pressurizer PORVs to bleed hot reactor coolant out of the RCS. To be certain that the bleed and feed heat removal path will be effective, typically at least two PORVs must open.
The effectiveness of RCS bleed and feed heat removal depends on four basic considerations. These are: 1) the timeliness of operator action to initiate bleed and feed following indications of the symptoms of loss of all secondary heat sink (see subsections 2.2.3 and 2.2.4), 2) the core decay heat at the time of RCS bleed and feed initiation, 3) the capacity of the pressurizer PORVs (i.e., number and size of valves), and 4) the capacity of the high pressure SI system (i.e., number, size, and shutoff head of the high pressure SI pumps).
These considerations govern the RCS depressurization, repressurization and pressure stabilization after RCS bleed and feed heat removal is established. The fourth consideration also governs the amount of SI flow delivered to the RCS at any RCS pressure. RCS bleed and feed effectiveness is maximized by a combination of these considerations which maximizes the initial RCS depressurization, minimizes the subsequent RCS repressurization and the pressure FRH.1 Background              11            HPRev. 2, 4/30/2005 HFRH1BG.doc
: 1. WE08EK1.02 001/LOIT/RO/M/F 3.4/4.0/WE08EK1.02/LO-TA-37022///
Initial condition:
            - A steam rupture inside containment occurred on Unit 1.
Current conditions:
            - RCS cold leg temperature is slowly lowering.
            - RCS pressure is slowly lowering.
            - 19241-C, "Response to Imminent Pressurized Thermal Shock Condition," is in progress.
Which one of the following completes the following statement?
Per 19241-C major operator actions, the in-progress RCS cooldown __(1)__ required to be stopped, and the in-progress RCS depressurization __(2)__ required to be stopped.
__(1)__                                __(2)__
A.                      is                                    is B.                      is                                  is NOT C.                  is NOT                                    is D.                  is NOT                                is NOT K/A WE08            RCS Overcooling - PTS EK1.02          Knowledge of the operational implications of the following concepts as they apply to the (Pressurized Thermal Shock):
                          - Normal, abnormal, and emergency operating procedures associated with Pressurized Thermal Shock.
K/A MATCH ANALYSIS This meets the KA since the question tests the operational implications of 19241-C.
The operational implications include having to decrease RCS pressure, stabilizing temperature and pressure, and performing a soak. Other procedures are allowed during soak as long as these parameters are not changed.
Friday, March 07, 2014 8:29:20 AM                                                            1
 
EXPLANATION OF REQUIRED KNOWLEDGE The bases behind this step is to prevent propagation of an existing flaw that could cause a through wall crack at the beltline region of the vessel. The KA does not ask for the reason or bases, but the operational implication (i.e., how we would operate the plant differently or be able to perform other actions while in this condition).
Per 19241-C step 6, if RCS WR Cold Leg temperatures are lowering, then the RNO attempts to stop the RCS cooldown by stopping all steam release, throttling RHR, and controlling feed water addition. Cooling the RCS increases the tensile stress on the vessel inner wall.
Per 19241-C step 25, the RCS is to be depressuirzed to lower RCS subcooling. The goal of this step is to reduce subcooling to the minimum value allowed in the EOPs, which minimizes RCS pressure. Lowering RCS pressure reduces the tensile stress on the vessel inner wall.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. RCS temperature must be stabilized to minimize the tensile stress on the reactor vessel inner wall due to the cooldown. However, it is reasonable to assume that a cooldown to <220F to ARM COPs would be an appropriate action for a cold overpressure event. This action is common in many of the EOPs.
The second part is correct. RCS pressure is required to be lowered to <125psig, or subcooling must be between 24F and 34F, or Przr Lvl >75%. With the steam leak present, RCS temperature will continue to lower and therefore RCS pressure will also continue to lowere due to this continuous action step.
B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.
The second part is incorrect. RCS pressure is required to be lowered to <125psig, or subcooling must be between 24F and 34F, or Przr Lvl >75%. With the steam leak present, RCS temperature will continue to lower and therefore RCS pressure will also continue to lowere due to this continuous action step.
However, in most EOPs, either temperature or pressure is changed at any one time. It is reasonable for a candidate to believe that if RCS temperature is being lowered, RCS pressure will be held stable.
C. Correct.              The first part is correct. Per step 6, the cooldown will be stopped (if possible) by isolating all steam (and feed) flow from (and to) the SG's.
The second part is correct. See the second part of choice A above.
Friday, March 07, 2014 8:29:20 AM                                                                  2
 
D. Incorrect. Plausible. The first part is correct. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Level:                      RO Tier # / Group #            T1 / G2 K/A#                        WE08EK1.02 Importance Rating:          3.4 / 4.0 Technical
 
==Reference:==
19241-C, Rev 25.3, pages 4, 16, & 17 References provided:        None Learning Objective:        LO-LP-37071-04 State the actions for preventing or mitigating the severity of overcooling and repressurizing transients.
LO-TA-37022          Respond to Imminent Pressurized Thermal Shock per 19241-C.
Question origin:            BANK Cognitive Level:            M/F 10 CFR Part 55 Content:    41.8 / 41.10 / 45.3 Comments:
You have completed the test!
Friday, March 07, 2014 8:29:20 AM                                                                  3
 
Approved By                                                                          Procedure    Version J. B. Stanley                          Vogtle Electric Generating Plant              19241-C        25.3 Effective Date                                                                        Page Number FR-P.1 RESPONSE TO IMMINENT 04/25/2013                  PRESSURIZED THERMAL SHOCK CONDITION                              4 of 35 ACTION/EXPECTED RESPONSE                            RESPONSE NOT OBTAINED 6
: 6. Check if RCS cooldown has stopped:            6.
6.a
: a. RCS WR Cold Leg temperatures                    a. Try to stop RCS cooldown:
                      - STABLE OR RISING.
6.a.1)
: 1)  Verify SG ARVs closed.
6.a.2)
: 2)  Verify Main Steamline Isolation and Bypass Valves closed.
6.a.3)
: 3)  IF RHR system in service, THEN stop any cooldown from RHR system.
6.a.4)
: 4)  Stop RCS cooldown to non-faulted SGs by performing the following:
Control feed flow.
Maintain total feed flow greater than 570 gpm until NR level greater than 10%
[32% ADVERSE] in at least one non-faulted SG.
7
: 7. Check SGs secondary pressure                  7.
boundaries:
7.a SG Pressures:                                a. Go to Step 9.
Any lowering in an uncontrolled manner.
                                      -OR-Any completely depressurized.
 
S Printed November 13, 2013 at 13:46
 
Approved By                                                                              Procedure    Version J. B. Stanley                          Vogtle Electric Generating Plant                19241-C        25.3 Effective Date                                                                            Page Number FR-P.1 RESPONSE TO IMMINENT 04/25/2013                  PRESSURIZED THERMAL SHOCK CONDITION                                16 of 35 ACTION/EXPECTED RESPONSE                                  RESPONSE NOT OBTAINED
 
NOTE The Upper Head region of the vessel may void during RCS depressurization if RCPs are NOT running. This will result in a rapidly rising PRZR level.
CAUTION RCS depressurization may result in RCP seal P lowering to less than 200 psid.
Shutdown of RCPs is required in this case.
25
: 25. Depressurize RCS to lower RCS                        25.
subcooling:
25.a
: a. Check if ANY of the following                          a. Go to Step 25.c.
conditions are satisfied:
RCS subcooling - 24F to 34&deg;F
[38F to 48&deg;F ADVERSE].
                                      -OR-PRZR level - GREATER THAN 75% [52% ADVERSE].
                                      -OR-RCS pressure - LESS THAN 125 PSIG.
25.b
: b. Go to Step 29.                                          b.
 
Step 25 continued on next page Printed November 13, 2013 at 13:46
 
Approved By                                                                            Procedure    Version J. B. Stanley                        Vogtle Electric Generating Plant                  19241-C        25.3 Effective Date                                                                          Page Number FR-P.1 RESPONSE TO IMMINENT 04/25/2013                  PRESSURIZED THERMAL SHOCK CONDITION                              17 of 35 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 25.c
: c. Check Normal PRZR Spray -                            c. IF letdown is in service, AVAILABLE.                                            THEN go to Step 26.
IF letdown is NOT in service, THEN use one PRZR PORV by performing the following:
25.c.1
: 1)    Arm one train of COPS and verify PRZR PORV Block Valve - OPEN.
25.c.2
: 2)    Open associated PRZR PORV.
25.c.3
: 3)    Go to Step 27.
IF RCS can NOT be depressurized using any PRZR PORV, THEN go to Step 26 even though letdown is NOT in service.
25.d
: d. Open Normal PRZR Spray                                d.
Valves.
25.e
: e. Go to Step 27.                                        e.
26
: 26. Establish Auxiliary Spray by                      26.
performing the following:
26.a
: a. Verify PRZR Heaters - OFF.                            a.
26.b
: b. Verify at least one CCP running.                      b.
 
Step 26 continued on next page Printed November 13, 2013 at 13:46
: 1. WE11G2.1.25 001/LOIT AND LOCT/RO/C/A 3.9/4.2/WE11G2.1.25/LO-TA-37020//HL-18 AUDIT/WE11EK1.2 Initial conditions:
              -  Large break LOCA occurred on Unit 1.
              -  Containment pressure is 23 psig.
              -  Both Containment Spray (CS) Pumps are running.
              -  RWST level is 26% and lowering.
              -  Four Containment Coolers are running in low speed.
Current conditions:
              - 19111-C, "Loss of Emergency Coolant Recirculation," is in progress.
              - The crew is at the step to "Determine Containment Spray requirements."
Per 19111-C, which of the following is the required operation of the Containment Spray Pumps?
REFERENCE PROVIDED A. Stop BOTH CS Pumps immediately and do not restart.
B. Continue to allow BOTH CS Pumps to run until RWST level lowers to < 8%, then stop BOTH CS Pumps.
C. Stop BOTH CS Pumps until suctions can be aligned to the Containment Sump, then restart the CS Pumps that were shut down.
D. Stop ONE CS Pump immediately. Stop the remaining CS pump when RWST level lowers to < 8% if unable to realign to the Containment Sump.
K/A WE11            Loss of Emergency Coolant Recirc.
G2.1.25          Ability to interpret reference materials, such as graphs, curves, tables, etc.
K/A MATCH ANALYSIS The candidate is required to assess current plant conditions and determine the number of Containment Spray Pumps (CSPs) required by interpreting the table in 19111-C
        'Loss of Emergency Coolant Recirculation' step 7, which is given as a reference.
EXPLANATION OF REQUIRED KNOWLEDGE 19111-C is entered when the ability to recirculate coolant from the emergency sumps through RHR is lost. The major priority of this procedure is to conserve RWST Friday, March 07, 2014 8:30:16 AM                                                                1
 
inventory to maximize the time to RWST depletion and subsequent stopping of all ECCS pumps. To this end, CSPs are stopped based on available Containment Coolers, Containment Pressure,and RWST level.
For the current plant conditions of 26% RWST level, 23 psig containment pressure, and 4 fans running in low speed, (1) one CSP is required to be in operation. Once the required number of CSPs is determined, pumps and discharge valves are operated as necessary.
In continuous action step 8, a check is made to see if alignment for recirculation should be made. If at least (1) one CSP is running with sufficient level in the sumps, the running CSP(s) will be aligned for recirculation. If no CSPs are running, this step is bypassed. The purpose of this step is to conserve RWST inventory.
ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Per the Explanation of Required Knowledge above, (1) one CSP is required to remain in service. However, the number of fans running in low speed is 4, which does not have a direct correlation in the table. It could be reasonable for a candidate to believe that since more than 3 fans are running in low speed, (0) zero CSPs are required and BOTH CSPs should be stopped.
B. Incorrect. Plausible. Per the Explanation of Required Knowledge above, (1) one CSP is required to remain in service. However, it is common for candidate to improperly navigate the table and cross up rows.
As such, it is reasonable for the candidate to believe (2) two CSPs are required and not stop either one. When RWST level lowers <8%, both CSPs would be stopped. This operation is consistent with the normal operation of the CS system.
C. Correct.              Per the Explanation of Required Knowledge above, (1) one CSP is required to remain in service. When RWST level lowers to <8%, the remain CSP would be stopped.
D. Incorrect. Plausible. Per the Explanation of Required Knowledge above, (1) one CSP is required to remain in service. However, 4 fans are running in low speed , which does not have a direct correlation in the table. As such, it would be reasonable for a candidate to believe that since more than 3 fans are in low speed, (0) zero CSPs are required and BOTH CSPs should be stopped. Step 8 checks for the need for recirculation alignment. Since BOTH CSPs would be stopped, this step would be bypassed.
However, the second part would be correct if (1) one or more CSPs were in operation.
Friday, March 07, 2014 8:30:16 AM                                                                2
 
Level:                    RO Tier # / Group #          T1 / G1 K/A#                      WE11G2.1.25 Importance Rating:        3.9 / 4.2 Technical
 
==Reference:==
EOP 19111-C Rev 33.2, page 7 References provided:      EOP 19111-C Rev 33.2, page 7 Learning Objective:        LO-PP-37115-02 Describe the actions taken to conserve RWST inventory for a loss of emergency coolant recirculation.
LO-PP-37115-06 Discuss the basis for controlling CS pumps and CNMT Cooler Fans during a loss of emergency coolant recirculation.
LO-PP-37115-10 Describe the actions taken to keep the core covered and protect ECCS equipment when RWST level drops below 8% during a loss of emergency coolant recirculation.
LO-TA-37020        Respond to a Loss of Emergency Coolant Recirculation Capability per 19111-C.
Question origin:          BANK Cognitive Level:          C/A 10 CFR Part 55 Content:    41.10 / 43.5 / 45.12 Comments:
You have completed the test!
Friday, March 07, 2014 8:30:16 AM                                                              3
 
Approved By                                                                        Procedure      Version C. S. Waldrup                            Vogtle Electric Generating Plant          19111-C        33.2 Effective Date                                                                      Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                      RECIRCULATION                              7 of 49 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 7.b
: b. Determine number of CS Pumps                      b.
required from Table:
RWST                      CONTAINMENT        FAN COOLERS    SPRAY PUMPS LEVEL                      PRESSURE            IN SLOW        REQUIRED GREATER THAN 52 PSIG          N/A              2 BETWEEN                        0                2 GREATER 21.5 PSIG and 52 PSIG          4                1 THAN 29%
8                0 LESS THAN 21.5 PSIG            N/A              0 GREATER THAN 52 PSIG          N/A              2 BETWEEN                    BETWEEN                        3                1 8% and 29%                  21.5 PSIG and 52 PSIG          6                0 LESS THAN 21.5 PSIG            N/A              0 LESS THAN 8%                N/A                            N/A              0 7.c
: c. Check CS Pumps running -                          c. Reset Containment Spray.
EQUAL TO NUMBER REQUIRED.                                          Operate CS Pumps and discharge valves as required.
 
S Printed November 13, 2013 at 08:47                27
 
Approved By                                                                          Procedure      Version C. S. Waldrup                            Vogtle Electric Generating Plant            19111-C        33.2 Effective Date                                                                        Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013                                      RECIRCULATION                              7 of 49 ACTION/EXPECTED RESPONSE                            RESPONSE NOT OBTAINED 7.b
: b. Determine number of CS Pumps                      b.
required from Table:
RWST                      CONTAINMENT          FAN COOLERS    SPRAY PUMPS LEVEL                      PRESSURE              IN SLOW        REQUIRED GREATER THAN 52 PSIG            N/A              2 BETWEEN                          0                2 GREATER 21.5 PSIG and 52 PSIG            4                1 THAN 29%
8                0 LESS THAN 21.5 PSIG            N/A              0 GREATER THAN 52 PSIG            N/A              2 BETWEEN                    BETWEEN 23 psig                  3 4              1 8% and 29%                  21.5 PSIG and 52 PSIG            6                0 26%                    LESS THAN 21.5 PSIG            N/A              0 LESS THAN 8%                N/A                            N/A              0 7.c
: c. Check CS Pumps running -              NO          c. Reset Containment Spray.
EQUAL TO NUMBER REQUIRED.                                          Operate CS Pumps and discharge valves as required.
 
S Printed November 13, 2013 at 08:47                27
: 1. WE16EK2.02 001/LOIT/RO/M/F 2.6/3.0/WE16EK2.02/LO-TA-37009/13009///
Given the following:
            - Large break LOCA is in progress on Unit 1.
            - 1RE-002, 1RE-003, 1RE-005, and 1RE-006 are in high alarm.
            - 19013-C, "Transfer to Cold Leg Recirculation," is in progress.
            - The crew just completed realigning Containment Spray suctions.
Which one of the following completes the following statement?
Per 19013-C, if dose rates will not allow reading local suction and discharge pressure indications, proper operation of Containment Spray shall be verified by observing containment __(1)__ lowering, and per 19010-C, "Loss of Reactor or Secondary Coolant," containment spray is required to remain in the recirculation mode for no less than __(2)__ hours.
__(1)__                                __(2)__
A.                  pressure                                  1.5 B.                  pressure                                    2 C.                temperature                                  1.5 D.                temperature                                    2 K/A WE16            High Containment Radiation EK2.02          Knowledge of the interrelations between the (High Containment Radiation) and the following:
                          - 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.
K/A MATCH ANALYSIS Containment Spray is a heat removal system employed during a large break LOCA.
Emergency coolant is recirculated via the containment emergency sumps and sprayed Friday, March 07, 2014 8:34:42 AM                                                              1
 
into the containment atmosphere to cool containment and lower overall containment tempertaure and pressure. There is an interrelationship between high containment radiation and expected dose rates at the CS Pumps rooms, which is the only location where suction and discharge pressures can be read. An additional iterrelationship exists between the minimum recirculation time requirements and the entrainment of Iodine inside containment.
NOTE: K/A category WE16 specifically addresses FR-Z.3, "Repsonse to High Containment Radiation Level". The specific knowledge item EK2.02 addresses an interralationship with heat removal. At Vogtle our FR-Z.3 (EOP 19253-C) does not have any direction association with any heat removal system. The closest interrelationship within the EOP network between High Containment Radiation and a Containment Heat Removal system would be Containment Spray on Sump Recirc during a LOCA event with High Containment Radiation. Therefore, this is the approach taken with this question.
EXPLANATION OF REQUIRED KNOWLEDGE With RE-005 and 006 in high alarm, fuel damage is expected. As such, dose rates in the area of the Containment Spray pump rooms will be too high for personnel to be dispatched to verify suction and discharge pressures, per 19013-C guidance. These indications are only available locally. Therefore, 19013-C has a CAUTION prior to step 17 that states dispatch will only be performed if radiation levels permit. Step 18.c. uses stable or lowering containment pressure as an alternte indication of proper operation.
Per step 12.d. of 19010-C, Containment Spray is required to be operated for 2 hours, 1.5 hours of which is in recirc mode if any of the containment rad monitors (RE-002, 003, 005, or 006) are in high alarm. The 2 hour operation requirement is for iodine scrubbing. During this period, iodine is "washed" out of the containment atmosphere and into solution in the emergency sumps. The 1.5 hours on recirculation is required to ensure proper mixing of the TSP located in baskets on the containment floor into the emergency sump fluid. The TSP raises the PH level of the water in the sump, minimizing corrosion on containment structures and equipment as the sump water is sprayed in containment.
ANSWER / DISTRACTOR ANALYSIS A. Correct.              The first part is correct. Per 19013-C CAUTION above step 17 and step 18.c, if radiation levels do not allow local verification of suction and discharge pressures, stable or lowering containment pressure is sufficient to alternately ensure proper operation.
The second part is correct. Per 19010-C step 12.d, at least 1.5 hours in recirculation mode is required.
B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect. Per 19010-C step 12.d, at least 1.5 hours in recirculation mode is required. However, step 12.d requires the Containment Spray Pumps to operate for a Friday, March 07, 2014 8:34:42 AM                                                                    2
 
minimum of 2 hours total for iodine scrubbing.
C. Incorrect. Plausible. The first part is incorrect. Per 19013-C step 18.c, if radiation levels do not allow local verification of suction and discharge pressures, stable or lowering containment pressure is sufficient to alternately ensure proper operation. However, since Containment Spray lowers pressure by lowering the bulk atmosphere temperature, monitoring temperature could be a reasonable response to ensure the CS system is operating properly.
The second part is correct. See the second part of choice A above.
D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is incorrect. See the second part of choice B above.
Friday, March 07, 2014 8:34:42 AM                                                                  3
 
Level:                    RO Tier # / Group #          T1 / G2 K/A#                      WE16EK2.02 Importance Rating:        2.6 / 3.0 Technical
 
==Reference:==
EOP 19010-C, Rev 34.3, pages 9 & 10 EOP 19013-C, Rev 29.2, pages 9 & 10 References provided:      None Learning Objective:        LO-LP-37113-02 Using EOP 19013-C as a guide, briefly describe how each step is accomplished.
LO-LP-37113-05 Given a NOTE or CAUTION statement from the EOP, state the bases for that NOTE or CAUTION statement.
LO-LP-37111-08 Using EOP 19010-C as a guide, briefly describe how each step is accomplished.
LO-TA-13009        Manually align ECCS for Cold Leg Recirculation Phase using EOP 19013-C.
LO-TA-37009        Respond to a Large Break Loss of Primary Coolant per 19010-C LO-PP-15101-06 State the reason for a minimum required time the Containment Spray system is left on recirculation following a LOCA.
Question origin:          NEW Cognitive Level:          M/F 10 CFR Part 55 Content:    41.7 / 41.9 / 41.10 / 45.7 Comments:
You have completed the test!
Friday, March 07, 2014 8:34:42 AM                                                              4
 
Approved By                                                                                Procedure    Version J. B. Stanley                          Vogtle Electric Generating Plant                    19010-C        34.3 Effective Date                                                                              Page Number E-1 LOSS OF REACTOR OR SECONDARY 7/25/12                                            COOLANT                                        9 of 27 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 11
        *11. Check if ECCS flow should be                      11.
reduced:
11.a
: a. RCS Subcooling - GREATER                            a. Go to Step 12.
THAN 24F [38F ADVERSE].
11.b
: b. Secondary Heat Sink:                                b. Go to Step 12.
Total feed flow to intact SG(s) -
GREATER THAN 570 GPM.
                                      -OR-NR level in at least one intact SG - GREATER THAN 10%
[32% ADVERSE].
11.c
: c. RCS pressure - STABLE OR                            c. Go to Step 12.
RISING.
11.d
: d. PRZR level - GREATER THAN                            d. Try to stabilize RCS pressure:
9% [37% ADVERSE].
Use Normal PRZR Spray if Instrument Air to Containment available.
Do NOT use PRZR PORVs to stabilize RCS pressure.
Go to Step 12.
11.e
: e. Go to 19011-C, ES-1.1 SI                            e.
TERMINATION.
12
        *12. Check if Containment Spray                        12.
should be stopped:
12.a
: a. CS Pumps - RUNNING.                                  a. Go to Step 13.
 
Step 12 continued on next page Printed November 12, 2013 at 14:40
 
Approved By                                                                          Procedure    Version J. B. Stanley                          Vogtle Electric Generating Plant              19010-C        34.3 Effective Date                                                                        Page Number E-1 LOSS OF REACTOR OR SECONDARY 7/25/12                                          COOLANT                                  10 of 27 ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED 12.b
: b. Containment pressure - LESS                    b. WHEN Containment pressure THAN 15 PSIG.                                  is less than 15 psig, THEN go to Step 12.c.
Go to Step 13.
12.c
: c. Any Containment radiation levels              c. Perform the following:
                      - INDICATE HIGH DUE TO                                                              12.c.1 PRIMARY LOCA:                                  1)    Reset Containment Spray signal.
RE-002                                                                            12.c.2
: 2)    Stop Containment Spray RE-003                                              Pumps.
12.c.3 RE-005                                        3)    Close CNMT SPRAY ISO VLV:
RE-006 HV-9001A HV-9001B Go to Step 13.
12.d
: d. Operate CS Pumps:                              d. WHEN CS Pumps have operated for at least 2 hours AND in the recirculation mode Minimum of 2 hours.                      for at least 1.5 hours, THEN perform Step 12.c RNO.
At least 1.5 hours in recirculation mode.
 
S Printed November 12, 2013 at 14:40
 
Approved By                                                                            Procedure    Version J.B. STANLEY                          Vogtle Electric Generating Plant                  19013-C      29.2 Effective Date                                                                          Page Number ES-1.3 TRANSFER TO COLD LEG 7/25/12                                      RECIRCULATION                                    9 of 20 ACTION/EXPECTED RESPONSE                              RESPONSE NOT OBTAINED
 
CAUTIONS The specified actions in Steps 17 through 19 should be promptly completed to avoid loss of CS Pump suction.
Local observation of CS Pump suction and discharge pressure gauges should only be performed if radiation levels permit.
UNIT 1 (AB D75)      UNIT 2 (AB D06) 17
: 17. Reset Containment Spray.                          17.
18
: 18. Align CS Pump A for recirculation:                18.
18.a
: a. Open CS Pump A suction valves                        a. Locally open:
from Containment Emergency Sump:                                                1-HV-9003A (AB-C134) 2-HV-9003A (AB-C124)
HV-9002A, CNMT SPRAY                          IF valves can NOT be PUMP A CNMT SUMP                                opened, SUCT IRC                                        THEN stop CS Pump A.
HV-9003A, CNMT SPRAY                          Go to Step 19.
PUMP A CNMT SUMP SUCT ORC 18.b
: b. Close CNMT SPRAY PUMP A                              b.
RWST SUCT ISO VLV:
HV-9017A
 
Step 18 continued on next page Printed November 12, 2013 at 14:29
 
Approved By                                                                              Procedure    Version J.B. STANLEY                          Vogtle Electric Generating Plant                  19013-C      29.2 Effective Date                                                                            Page Number ES-1.3 TRANSFER TO COLD LEG 7/25/12                                        RECIRCULATION                                  10 of 20 ACTION/EXPECTED RESPONSE                                RESPONSE NOT OBTAINED 18.c
: c. Check Train A CS proper                              c. Verify valve alignment operation using the following                          correct:
indications, if available:
HV-9002A - OPEN Pump suction pressure PI-0972                            HV-9003A - OPEN
                        - GREATER THAN 7 PSIG.                                  HV-9017A - CLOSED HV-9001A - OPEN Pump discharge pressure PI-0974 - APPROXIMATELY 185 PSIG ABOVE SUCTION PRESSURE.
Containment pressure - STABLE OR LOWERING.
19
: 19. Align CS Pump B for recirculation:                  19.
19.a
: a. Open CS Pump B suction valves                        a. Locally open:
from Containment Emergency Sump:                                                  1-HV-9003B (FHB-C08) 2-HV-9003B (FHB-C02)
HV-9002B, CNMT SPRAY                            IF valves can NOT be PUMP B CNMT SUMP                                opened, SUCT IRC                                        THEN stop CS Pump B.
HV-9003B, CNMT SPRAY                            Go to Step 20.
PUMP B CNMT SUMP SUCT ORC 19.b
: b. Close CNMT SPRAY PUMP B                              b.
RWST SUCT ISO VLV:
HV-9017B
 
Step 19 continued on next page Printed November 12, 2013 at 14:29
 
FR-Z.3 Response to High Containment Radiation Level                                            19253-C VOGTLE                        Version 9 Unit C                    Page 5 of 7 SUBSEQUENT OPERATOR ACTIONS ACTION/EXPECTED RESPONSE                          RESPONSE NOT OBTAINED
: 1. Verify Containment Ventilation Isolation:      1.
: a. Dampers and Valves - CLOSED:                  a.
CVI MLB indication
                      -OR-Reference ATTACHMENT 1 as necessary.
: 2. Check Piping Penetration Filtration and        2. Start fans by initiating 13305-1, 13305-2 Exhaust Units - BOTH RUNNING.                    Auxiliary Building HVAC System.
: 3. Place the Containment Preaccess Filter          3.
units in service by initiating 13125-1, 13125-2 Containment Purge System.
: 4. Notify TSC of Containment radiation level      4.
to obtain recommended action.
: 5. Return to procedure and step in effect.        5.
END OF PROCEDURE TEXT Printed 03/06/2014 at 10:23:00}}

Latest revision as of 15:35, 10 January 2025

301 Draft RO Written Examination
ML14338A056
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 12/03/2014
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