ML101520091

Vogtle 2007-301 Draft SRO Written Exam (Section 3)

ML101520091
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{{#Wiki_filter:() ES-401' Site-Specific SRO Written Examination Cover Sheet Form ES-401-8 u.s. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name: Date: Facility/Unit: YOG7Lk:. Region: I 0, II III 0 IV 0 Reactor Type: 0 BWDGED . , Start Time: Finish Time: Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours to complete the combined examination, and 3 hours if you are only taking the SRO portion. Applicant Certification All work done on this examination is my own. I have neither given nor received aid. Applicant's Signature Results RO/SRO*OnlylTotal Examination Values 7S" / 2,S"/ 100 Points --Applicant's Scores --/ --/ --Points Applicant's Grade --/ --/ --Percent ES-401, Page 31 of 33 ( l 1. 001AA2.01 001 The RO withdraws control rods 3 steps, upon release of the IN-HOLD-OUT switch, rods continue to withdraw and cannot be stopped. Which ONE of the following reactor trips is based on this event and would result in the reactor trip breakers opening to mitigate the event without any operator action? (Assume all appropriate procedural actions have been performed to this point) A. SR Hi Flux reactor trip, power at the point for taking critical data. IR Hi Flux reactor trip, power at 4% just prior to entering Mode 1 . c. PR Low Setpoint Hi Flux reactor trip, power at 18% power during swap to MFRVs. D. PR Positive Rate Trip, power near 100% power performing rod operability testing. Page: 10f48 6/6/2007 001 Continuous Rod Withdrawal AA2.01 Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal. Reactor tripped breaker indicator KIA MATCH ANALYSIS Question gives a plausible scenario where an uncontrolled rod motion occurs after the RO releases the reactor trip breakers. The candidate has to pick which Reactor Trip setpoint is based on this event and power level. Question meets 10CFR55.43(b) criteria item 2 -Facility operating limits in Tech Specs and their bases. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible the candidate may know the bases for SR Hi Flux is uncontrolled rod withdrawal but the trip is blocked above P-6 when critical data is taken. B. Correct. IR Hi Flux based on uncontrolled rod withdrawal and would not be blocked at this power level. C. Incorrect. Plausible the candidate may know the bases for PR Low Setpoint Hi Flux is for an uncontrolled rod withdrawal but this trip should be blocked per UOP actions when P-10 is received, far below the current power level stated in the question. D. Incorrect. Plausible the candidate may confuse the bases for PR High Positive Rate reactor trip which is an ejected rod event with uncontrolled rod motion. REFERENCES Technical Specifications 3.3.1 and Bases for Reactor Trip Instrumentation. VEGP learning objectives: LO-LP-39207-02, 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 3.3 are exceeded. Page: 2of48 6/6/2007 (, Number Text LO-LP-39206-08 Define nuclear enthalpy rise hot channel factor. LO-LP-39206-09 Define quadrant power tilt ratio. State the required action for exceeding the limit at various power levels. LO-LP-39207-01 For any given item in section 3.3 of Tech Specs, be able to: a.O State the LeO. b.O State anyone hour or less required actions. LO-LP-39207-02 Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a.O Whether any Tech Spec LeOs of section 3.3 are exceeded. b.O The required actions for all section 3.3 LeOs. LO-LP-39207-03 For any given item in section 13.3 of the Technical Requirements Manual, be able to: a.O State the Technical Requirement (TR) for operation. LO-LP-39207 -04 LO-LP-39207 -05 LO-LP-39207 -06 LO-LP-39208-01 b.O State anyone hour or less required actions. Describe the bases for any given Tech Spec in section 3.3. Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: a.D Whether any TR of section 13.3 has been exceeded. b.O The required actions for all section 13.3 TRs. State the values for the Limiting Safety System Settings. For any given item in section 3.4 of Tech Specs, be able to: a.D State the LeO. b.D State anyone hour or less required actions. LO-LP-39208-02 Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a.O Whether any Tech Spec LeOs of section 3.4 are exceeded. b.O The required actions for all section 3.4 LeOs. LO-LP-39208-03 For any given item in section 13.4 of the TRM, be able to: a.O State the TR for operation. b.O State anyone hour or less required actions. LO-LP-39208-04 Describe the bases for any given Tech Spec in section 3.4. LO-LP-39208-05 State why the mode 5 ReS loops and coolant circulation specifications for loops filled vs. loops not filled vary with regard to the number of RHR trains required and steam generator aVailability. LO-LP-39208-06 State the reason for limiting the ReS specific activity. Friday, June 01, 2007 Page 42 of 165 ReFer:vN'ct lZ:-lL 00 I !1-i/2,() I c- ____________________________________________________ _ APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

a. Power Range Neutron Flux -High b. The Power Range Neutron Flux -High trip Function ensures that protection is provided, from all power levels, against a positive reactivity excursion leading to DNB during power operations.

These can be caused by rod withdrawal or reductions in RCS temperature. The LCO requires all four of the Power Range Neutron Flux -High channels to be OPERABLE. In MODE 1 or 2, when a positive reactivity excursion could occur, the Power Range Neutron Flux -High trip must be OPERABLE. This Function will terminate the reactivity excursion and shut down the reactor prior to reaching a power level that could damage the fuel. In MODE 3, 4, 5, or 6, the NIS power range detectors cannot detect neutron levels in this range. In these MODES, the Power Range Neutron Flux -High does not have to be OPERABLE because the reactor is shut down and reactivity excursions into the power range are extremely unlikely. Other RTS Functions and administrative controls provide protection against reactivity additions when in MODE 3, 4, 5 , or 6. Power Range Neutron Flux -Low The LCO requirement for the Power Range Neutron Flux -Low trip Function ensures that protection is provided against a positive reactivity excursion from low power or subcritical conditions. The LCO requires all four of the Power Range Neutron Flux -Low channels to be OPERABLE. R. k) Ovt r In MODE 1, below the Power Range Neutron Flux (P-10 lit setpoint), and in MODE 2, the Power Range Neutron Flux -({ f) l r Low trip must be OPERABLE. This Function may be manually \...:..., blocked by the operator when two out of four power range channels are greater than approximately 10% RTP (P-10 Vogtle Units 1 and 2 B 3.3.1-12 Revision No. 0 M,-,rO I Dr..... f) J:: o f\.-, 00 ( RTS Instrumentation 1\ 1 c B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

b. Power Range Neutron Flux -Low (continued) setpoint).

This Function is automatically unblocked when three out of four power range channels are below the P-10 setpoint. Above the P-10 setpoint, positive reactivity additions are mitigated by the Power Range Neutron Flux -High trip Function. In MODE 3, 4, 5, or 6, the Power Range Neutron Low trip Function does not have to be OPERABLE because the reactor is shut down and the NIS power range detectors cannot detect neutron levels in this range. Other RTS trip Functions and administrative controls provide protection against positive reactivity additions or power excursions in MODE 3, 4, 5, or 6. 3. Power Range Neutron Flux -High Positive Rate \A(ej Oqr It D \ \ bV\ f-f ("-bl 'C The Power Range Neutron Flux -High Positive Rate trip uses the same channels as discussed for Function 2 above. The Power Range Neutron Flux -High Positive Rate trip Function ensures that protection is provided against rapid increases in neutron flux that are characteristic of an RCCA drive rod housing rupture and the accompanying ejection of the RCCA. This Function compliments the Power Range Neutron Flux -High and Low Setpoint trip Functions to ensure that the criteria are met for a rod ejection from the power range. The LCO requires all four of the Power Range Neutron Flux -High Positive Rate channels to be OPERABLE. In MODE 1 or 2, when there is a potential to add a large amount of positive reactivity from a rod ejection accident (REA), the Power Range Neutron Flux -High Positive Rate trip must be OPERABLE. In MODE 3, 4, 5, or 6, the Power Range Neutron Flux -High Positive Rate trip Function does not have to be felL-001 tht I . (continued) Vogtle Units 1 and 2 B 3.3.1-13 Revision No. 0 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

3. Power Range Neutron Flux -High Positive Rate (continued)

OPERABLE because other RTS trip Functions and administrative controls will provide protection against positive reactivity additions. In MODE 6, no rods are withdrawn and the SDM is increased during refueling operations. The reactor vessel head is also removed or the closure bolts are detensioned preventing any pressure buildup. In addition, the NIS power range detectors cannot detect neutron levels present in this mode. 4. Intermediate Range Neutron Flux If The Intermediate Range Neutron Flux (NI-035B, D, & E, NI-036B, D, & G) trip Function ensures that protection is 1\/., lNfJ tL-provided against an uncontrolled RCCA bank rod withdrawal (( 0 ( J accident from a subcritical condition during startup. This trip \ Function provides redundant protection to the Power Range Neutron Flux -Low Setpoint trip Function. The NIS intermediate range detectors are located external to the reactor vessel and measure neutrons leaking from the core. The NIS intermediate range detectors do not provide any input to control systems. Note that this Function also provides a signal to prevent automatic and manual rod withdrawal prior to initiating a reactor trip. Vogtle Units 1 and 2 The LCO requires two channels of Intermediate Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function. Because this trip Function is important only during startup, there is generally no need to disable channels for testing while the Function is required to be OPERABLE. Therefore, a third channel is unnecessary. In MODE 1 below the P-10 setpoint, and in MODE 2, when there is a potential for an uncontrolled RCCA bank rod withdrawal accident during reactor startup, the Intermediate Range Neutron Flux trip must be OPERABLE. B 3.3.1-14 Revision No. 0 ( BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

4. Intermediate Range Neutron Flux (continued)

Above the P-10 setpoint, the Power Range Neutron Flux -High Setpoint trip and the Power Range Neutron High Positive Rate trip provide core protection for a rod withdrawal accident. In MODE 3, 4, or 5, the Intermediate Range Neutron Flux trip does not have to be OPERABLE because the reactor cannot be started up in this condition. The core also has the required SDM to mitigate the consequences of a positive reactivity addition accident. In MODE 6, all rods are fully inserted and the core has a required increased SDM. Also, the NIS intermediate range indication is typically low off-scale in this MODE. 5. Source Range Neutron Flux ,at C( lit c, l \ b k... The LCO requirement for the Source Range Neutron Flux trip (NI-0031 B, 0, & E, NI-0032B, 0, & G) Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition during startup. This trip Function provides redundant protection to the Power Range Neutron Flux -Low Setpoint and Intermediate Range Neutron Flux trip Functions. In MODES 3, 4, and 5, administrative controls also prevent the uncontrolled withdrawal of rods. The NIS source range detectors are located external to the reactor vessel and measure neutrons leaking from the core. Vogtle Units 1 and 2 The NIS source range detectors do not provide any inputs to control systems. The source range trip is the only RTS automatic protection function required in MODES 3, 4, and 5. Therefore, the functional capability at the specified Trip Setpoint is assumed to be available. The LCO requires two channels of Source Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function. The LCO also requires two channels of the Source Range Neutron Flux to be OPERABLE in MODE 3, 4, or 5 with RTBs closed. The Source Range Neutron Flux Function provides protection for control rod withdrawal from B 3.3.1-15 Rev. 1-3/99 r * (;V l/flt ) ,z e f c-£ rJ' (L -f Cf\.../ RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and 5. Source Range Neutron Flux (continued) subcritical, boron dilution (see LCO 3.3.8) and control rod ejection events. The Function also provides visual neutron flux indication in the control room. APPLICABI LlTY /0 r A r c.. 0 J.-I n MODE 2 when below the P-6 setpoint during a reactor \ V" l L J L..t '\ startup, the Source Range Neutron Flux trip must be OPERABLE. Above the P-6 setpoint, the Intermediate Range ('tt \ 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. Vogtle Units 1 and 2 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 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 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) B 3.3.1-16 Rev. 1-3/99

2. 002A2.04 001 Given the following conditions:

r -Crew is performing the actions of 19231-C, "Loss of Heat Sink" due to a prolonged \.. _ loss of feedwater. -RCS Bleed and Feed has been initiated. -SG WR levels are all approximately 15%. -Containment pressure is 0.2 psig -RE-002 and RE-003 are in INTERMEDIATE alarm -AFW flow has just been restored to a single Steam Generator Which ONE of the following is CORRECT? A. Containment pressure will remain stable, immediately transition to 1901 O-C, Loss of Reactor or Secondary Coolant. B. Containment pressure will rise over time, immediately transition to 19010-C, Loss of Reactor or Secondary Coolant. C. Containment pressure will remain stable, remain in 19231-C, Loss of Secondary Heat Sink to perform further actions. Containment pressure will rise over time, remain in 19231-C, Loss of Secondary Heat Sink to perform further actions. Page: 3 of 48 6/6/2007 002 Reactor Coolant System (RCS). A2.04 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. Loss of heat sinks. KIA MATCH ANALYSIS Questions gives a scenario during a large break LOCA where the crew is required to transition to address RWST level lowering with inadequate containment sump level. The candidate must determine the correct transition flow path to address the problem. Question meets 1 OCFR55.43(b) criteria item # 5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions. Therefore, the question is SRO only. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible the candidate may not recall PORVs open causing containment pressure to rise but rad monitors indicate PRT ruptured. 19010 is improper due to PORVs must be sequentially closed before the transition. B. Incorrect.Plausible the candidate may recognize PORVs cause containment pressure to rise and think transition to 19010 is appropriate and not think of terminating feed and bleed. C. Incorrect. Plausible the candidate may not recall PORVs open causing containment pressure to rise but rad monitors indicate PRT ruptured. Remaining in 19231 to terminate SI is correct. D. Correct. Containment pressure should rise and remain in 19231 is correct. REFERENCES 19231-C, Loss of Secondary Heat Sink pages 21 -24, and 31 -36. VEGP learning objectives: LO-LP-37051-08, Using EOP 19231 as a guide, briefly describe how each major step is accomplished. Describe the bases for each. (commitment) LO-LP-37051-10, State all conditions when the procedure 19231, Response to Loss of ( .. ; Secondary Heat Sink, would be terminated. Page: 40f48 6/6/2007 r LO Cluster 37 Loss of Heat Sink Objectives LO-LP-37051-01 LO-LP-37051-02 LO-LP-37051-03 LO-LP-37051-04 LO-LP-37051-05 LO-LP-37051-06 LO-LP-37051-07 LO-LP-37051-08 LO-LP-37051-09 LO-LP-37051-10 LO-LP-3705 11 LO-LP-37051-12 LO-LP-37051-13 LO-LP-37051-14 Cite potential events that could lead to a loss of secondary heat sink. (commitment) State the reason for tripping RCPs earty in the transient. State RCS temperature and pressure response to a loss of secondary heat sink (LOSHS) with and without required operator actions. (commitment) State time considerations on initiation of bleed and feed. (commitment) State the precautions which should be taken in feeding a hot, dry steam generator following recovery from a loss of heat sink accident. Interpret the CSFST for a challenge to the heat sink safety function. (19200, F.0-3 -Heat Sink). State the intent of 19231, Response to Loss of Secondary Heat Sink. Using EOP 19231 as a guide, briefly describe how each major step is accomplished. Describe the bases for each. (commitment) Given a NOTE or CAUTION statement from the EOP , state the bases for that NOTE or CAUTION statement. State all conditions when the procedure 19231, Response to Loss of Secondary Heat Sink, would be terminated. Define loss of secondary heat sink in accordance with 19231, Response to Loss of Secondary Heat Sink, requiring immediate initiation of bleed and feed control. (commitment) Discuss SOER-86.001 with regard to failure of the AFW System and the potential for sustained loss of AFW capability due to equipment malfunction or operator error. (commitment) Describe how to operate the turbine-driven AFW pumps' turbine trip and throttle valve after a mechanical overspeed trip with or without electrical power to the valve motor. (commitment) State the consequences of failure to establish an alternate heat sink if AFW is lost. (commitment) Wednesday, June 06, 2007 Page 1 ofl PROCEDURE NO. VEGP 19231-C REVISION NO. ACTION/EXPECTED RESPONSE PAGE NO. 29 21 of 45 RESPONSE NOT OBTAINED CAUTION: Feed flow rates should be controlled to prevent excessive RCS cooldown.

• 47. Try to establish IIDAFW flow to at least one SG: a. Check MDAFW Pump -AVAILABLE: o -Power available o -Suction pressure o -Discharge pressure b. Select SG(s) to feed: o 1) All SG WR levels -LESS THAN 9% (31% ADVERSE) o c. Check Core Exit TCs -STABLE OR LOWERING d. Restore feed flow to selected SG -BETWEEN 30 GPM AND 100 GPM: o -IPC Point -UF5403 o e. Check Dry SG WR level -GREATER THAN 9% (31% ADVERSE) Of. Raise feed flow to restore NR level grea e r than 10% (32% ADVERSE) and go to Step 68. a. Perform the following:

0-Initiate actions to restore a MDAFW Pump. 0-WHEN MDAFW Pump is started, THEN go to Step 47b. 0-Go to Step 51. b. Perform the following: o -Restore feed flow to Non-Dry SG(s) by going to Step 48. o c. Do NOT limit feed flow to the selected SG if Core Exit TCs are rising and go to step 47f. De. WHEN Dry SG WR level is greater than 9% (31% ADVERSE) THEN raise feed flow to restore NR level greater than 10% (32% ADVERSE) . DGo to Step 68. PROCEDURE NO. VEGP REVISION NO. 19231-C PAGE NO. 29 22 of 45 ,ZefvLflVCC fGlL-001.,..,fr 0 U ACTION/EXPECTED RESPONSE RESPONSE NOT 48. Verify MDAFW Pump throttle 48. valves open for selected SG (s) : D HV-5139 MDAFW Pump A to SG 1 D HV-5137 MDAFW Pump A to SG 4 D HV-5132 MDAFW Pump B to SG 2 D HV-5134 MDAFW Pump B to SG 3 D 49. Verify adequate feed flow to 49. raise SG levels. D 50. Go to Step 68. Perform the following as necessary to establish MDAFW feed flow: -Open MDAFW Pump crosstie valves: Da. 1302-U4-055 Db. 1302-U4-056 D-Limit flow rate to avoid pump runout -LESS THAN 600 GPM IF feed flow to at least one SG verified, THEN perform the following: D a. Maintain flow to restore NR level to greater than 10% [32% ADVERSE] . Db. Go to Step 68. D IF feed flow to at least one SG can NOT be verified, THEN go to Step 51. ( ( PROCEDURE NO. VEGP 19231-C REVISION NO. ACTION/EXPECTED RESPONSE PAGE NO. 29 23 of 45 RESPONSE NOT OBTAINED CAUTION: Feed flow rates should be controlled to prevent excessive RCS cooldown.

• 51. Try to establish TDAFW flow to at least one SG: a. Check TDAFW Pump -AVAILABLE: O-0-Steam admission valve HV-5106 -OPEN Trip & Throttle valve PV-15129 -OPEN (HS-15111)

Governor valve SV-15133 -OPERATING PROPERLY (PDIC-5180A)

b. Select SG(s) to feed: o 1) All SG WR levels -LESS THAN 9% (31% ADVERSE) Dc. Check Core Exit TCs -STABLE OR LOWERING d. Restore feed flow to selected SG -BETWEEN 30 GPM AND 100 GPM: o -IPC Point -UF5403 De. Check Dry SG WR level -GREATER THAN 9% (31% ADVERSE) D f. Raise feed flow to restore NR level greate than 10% (32% ADVERSE) and go to Step 68. a. Perform the following:

0-0-Initiate 13610, AUXILIARY FEEDWATER SYSTEM to operate TDAFW Pump as necessary. WHEN TDAFW Pump is started, THEN go to Step SIb. o -Go to Step 55. b. Perform the following: b c. o -Restore feed flow to Non-Dry SG(s) by going to Step 52. Do NOT limit feed flow to the selected SG if Core Exit TCs are rising and go to 51f. De. WHEN Dry SG WR level is greater than 9% (31% ADVERSE) THEN raise feed flow to restore NR level greater than 10% (32% ADVERSE) . DGo to Step 68. {' ( ( PROCEDURE NO. VEGP 19231-C REVISION NO. ACTION/EXPECTED RESPONSE 52. Verify TDAFW pump throttle valves open for selected SG (s) : D HV-5122 TDAFW Pump to 0 HV-5125 TDAFW Pump to 0 HV-5127 TDAFW Pump to D HV-5120 TDAFW Pump to SG SG SG SG D 53. Verify adequate feed flow to raise SG levels. D 54. Go to Step 68. 29 1 2 3 4 *55. Try to establish main FW flow to at least one SG: D a. Check condensate system -IN SERVICE 56. Verify the following: D. MFRVs CLOSED AND CONTROLLERS AT 0% DEMAND IN MANUAL D. BFRVs CLOSED AND CONTROLLERS AT 0% DEMAND IN MANUAL PAGE NO. 24 of 45 RESPONSE NOT OBTAINED 53. IF feed flow to at least one SG verified, THEN perform the following: D a. Maintain flow to restore NR level to greater than 10% [32% ADVERSE] . Db. Go to Step 68. D IF feed flow to at least one SG can NOT be verified, THEN go to Step 55. D a. Place condensate system in service by initiating 13615, CONDENSATE AND FEEDWATER SYSTEM. o WHEN Condensate system in service, THEN go to Step 56. DReturn to Step 47. ( ( PROCEDURE 110. VEGP 19231-C REVISION NO. ACTION/EXPECTED RESPONSE 64. Depressurize selected SG to less than 550 psig: D a. Check MSIVs and BSIVs -ANY OPEN Db. Close all MSIVs and BSIVs except on selected SG. D c. Depressurize selected SG using Steam Dumps. D 65. Open Main Feed Pump discharge valves. D 66. Open BFIV for selected SG. D 67. Slowly open BFRV for the selected SG to establish feed flow. 68. Check for adequate secondary heat sink: o a. NR level in at least one SG -GREATER THAN 10% [32% ADVERSE] PAGE NO. 29 31 of 45 RESPONSE NOT OBTAINED 064. Actuate Main Steamline Isolation. DDump steam using selected SG ARV. o IF unable to dump steam, THEN return to Step 47. o 65. IF discharge valves can NOT be opened, THEN locally open MFP bypass valve 1305-U4-655. (TB-Lvl 2) D 66. Open MFIV on selected SG. D IF MFIV will NOT open, THEN dispatch an operator to locally open the selected BFIV. o IF neither BFIV or MFIV will open, THEN return to Step 47. D 67. Open MFRV for the selected SG. o IF MFRV will NOT open, THEN dispatch an operator to locally open the BFRV. o IF neither BFRV or MFRV will open, THEN return to Step 47. o a. IF feed flow to at least one SG verified, THEN do NOT continue until NR level is restored to greater than 10% [32% ADVERSE] . ( ( PROCEDURE NO. VEGP 19231-C REVISION NO. ACTION/EXPECTED RESPONSE PAGE NO. 29 32 of 45 RESPONSE NOT OBTAINED 69. Check RCS temperatures: 069. Return to Step 47. ( o -Core exit TCs -LOWERING o -RCS WR hot leg temperatures -LOWERING 70. Verify Reactor Head Vent Valves -CLOSED: 0-HV-8095A -RX HEAD VENT LETDOWN ISOLATION VLV 0-HV-8095B -RX HEAD VENT LETDOWN ISOLATION VLV 0-HV-8096A -RX HEAD VENT LETDOWN ISOLATION VLV 0-HV-8096B -RX HEAD VENT LETDOWN ISOLATION VLV 0-HV-0442A -REACTOR HEAD VENT TO PRT 0-HV-0442B -REACTOR HEAD VENT TO PRT TO TO TO TO NOTE: The following step will prevent an unwanted SI actuation from occurring when securing RCS bleed and feed. o 71. Check SG pressures -GREATER THAN 585 PSIG 71. Bypass the SG LOW PRESS inputs for 2 channels of any depressurized SG by initiating 13509-C, BYPASS TEST INSTRUMENTATION {BTI} PANEL OPERATION: o SG1:PB514A, PB515A, PB516A o SG2:PB524A, PB525A, PB526A o SG3:PB534A, PB535A, PB536A o SG4:PB544A, PB545A, PB546A PROCEDURE 110. VEGP 19231-C REVISION NO. PAGE NO. 29 33 of 45 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 72. 73. 74. NOTE: After closing a PRZR PORV, it may be necessary to wait for RCS pressure to rise before determining if ECCS flow can be terminated. Check if ECCS flow can be terminated, D a. RCS subcooling -GREATER D a. Go to Step 73. THAN 24°F [3S0F ADVERSE] Db. Check RVLIS full range indication -GREATER THAN 62% Dc. Go to Step 74. r I , Check RCS bleed path status: D a. PRZR PORVs and associated block valves -ANY BLEED PATH OPEN b. Close ONE PRZR PORV: D l} Block one train of COPS. D 2} Place associated PRZR PORV in AUTO. D 3} Verify proper operation of PORV. Dc. Return to Step 72. Stop ECCS Pumps: D-SI Pumps All but one cCP r: Db. Go to Step 73. D a. Go to 19010-C, E-l LOSS OF REACTOR OR SECONDARY COOLANT. Db. Close its associated block valve. D IF block valve can NOT be closed, THEN go to 19010-C, E-l LOSS OF REACTOR OR SECONDARY COOLANT. ;1!l .. P (Cj LA. '> i/o (e... ... +-yO"t 'Ok"" +-po R. (/5 J-/vcp w7v ( PROCEDURE NO. VEGP 19231-C REVISION NO. ACTION/EXPECTED RESPONSE 75. Check RCS bleed path status: \V D a. PRZR PORVs and associated block valves -ANY BLEED PATH OPEN b. Close all but ONE PRZR PORV: D 1) Block one train of COPS. D 2) Place associated PRZR PORV in AUTO. D 3) Verify proper operation of PORV. 29 PAGE NO. 34 of 45 RESPONSE NOT OBTAINED D a. Go to Step 76. Db. Close associated PORV block valve. D IF block valve can NOT be closed, THEN go to 19010-C, E-1 LOSS OF REACTOR OR SECONDARY COOLANT. D 76. Check Instrument Air -AVAILABLE r D 76. Establish Safety Grade Charging by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM. ( ( PROCEDURE NO. VEGP 19231-C REVISION NO. PAGE NO. 29 35 of 45 COZ,A 2.,rOt ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 77. Establish charging flow: a. Open CCP normal miniflow isolation valves: O. HV-SlllA -CCP-A MINI FLOW O. HV-SlllB -CCP-B MINI FLOW O. HV-Sll0 -CCP A&B COMMON MINI FLOW b. Close CCP alternate miniflow valves: O. HV-S50SA -CCP-A RV TO RWST ISOLATION O. HV-S50SB -CCP-B RV TO RWST ISOLATION

c. Verify white Pressure Control Mode light -NOT LIT: O. HV-S50SA -CCP-A RV TO RWST ISOLATION O. HV-S50SB -CCP-B RV TO RWST ISOLATION

d. Perform the following:

o 1) Block remaining train of COPS. o 2} Place associated PRZR PORV in AUTO. o 3} Verify proper operation of PORV. De. Set SEAL FLOW CONTROL HC-1S2 to maximum seal flow {HV-01S2 CLOSED}. f. Close BIT DISCH ISOLATION valves: g. O. HV-SS01A O. HV-SS01B Open CHARGING TO RCS ISOLATION valves: O. HV-Sl05 O. HV-Sl06 o *7S. Maintain Seal Injection flow to all Reps -8 TO 13 GPM Od. Close associated PORV Block Valve. D IF block valve can NOT be closed, ---THEN go to 19010-C, E-l LOSS OF REACTOR OR SECONDARY COOLANT. ( ( ( PROCEOURE NO. VEGP 19231-C REVISION NO. ACTION/EXPECTED RESPONSE D *79. Check RCS Bot Leg temperatures -STABLE OR LOWERING. PAGE NO. 29 36 of 45 RESPONSE NOT OBTAINED D *79. Control feed flow and dump steam as necessary to establish stable RCS Hot Leg temperatures.

80. Check if RHR pumps should be stopped: D a. RHR Pumps -ANY RUNNING WITH SUCTION ALIGNED TO RWST b. RCS pressure:

D 1) Greater than 300 psig. D 2) Stable or rising. D c. Stop RHR pumps. D *81. Control charging flow to maintain PRZR level at 25%. D 82. Go to 19011-C, ES-1.1 SI TERMINATION, Step 14. D a. Go to step 81. Db. Go to 19010-C, E-1 LOSS OF REACTOR OR SECONDARY COOLANT. END OF PROCEDURE TEXT ( 3. 007G2.4.6001 The unit trips from 100% power when a lightning strike causes a fault on 13.8 Kv bus 1NAB. -The other 13.8Kv bus, 1 NAA fails to fast bus transfer but is capable of being re-energized. -4160 non-1 E buses 1 NA01, 1 NA04, and 1 NA05 are still energized. -The crew is performing the actions of 19001-C, "Reactor Trip Response. -Control bank D (CDB) rod M12 indicates 18 steps on DRPI. -CST # 1 level is 52% and slowly lowering. -CST # 2 level is 74% and stable. -Letdown isolates post Rx. trip, attempts to re-establish are unsuccessful because LV-460 will not re-open. Which ONE of the following would be the CORRECT actions to take? A'! Re-energize 1 NAA per 13420-1, "13.8Kv Electrical Distribution System" and start RCP # 4 per 19001-C Attachement "A", "Starting a Reactor Coolant Pump". B. Perform an Emergency Boration of the RCS in accordance with SOP-13009-1/2, CVCS Reactor Makeup Control System and perform a Shutdown Margin. C. Perform the actions of AOP-18007-C, section A for Loss of Letdown and place Safety Grade Letdown in service per 13006, "Chemical Volume Control System". D. Immediately swap AFW pump suctions to CST # 2 per 13610-1, "Auxiliary Feedwater System" and transition to 19002-C, "Natural Circulation Cooldown". Page: S of48 6/6/2007 ( 007 Reactor Trip -Stabilization -Recovery G2.4.6 Knowledge of Symptom based EOP mitigation strategies KIA MATCH ANALYSIS Question gives a plausible scenario post reactor trip with several malfunctions, the candidate must choose the correct action / procedure to respond to the symptoms. Question meets 1 OCFR55.43(b) item # 5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions. ANSWER I DISTRACTOR ANALYSIS A. Correct. 13.8 Kv bus should be re-energized and RCP # 4 started. B. Incorrect. Plausible the candidate may consider this action but the emergency boration is only required for 2 or more stuck rods. Shutdown margin is required. C. Incorrect. Plausible the candidate would consider performing the actions for the Loss of Letdown AOP in parallel, however, Excess Letdown would be placed in service versus Safety Grade Letdown which is used on a loss of instrument air. Safety Grade Letdown is mentioned in notes and cautions of 19001-C. D. Incorrect. AFW suctions should only be swapped when one CST level lowers to less than 15%, there would be no reason to go to 19002 for Natural Circ cooldown when it is possible to start an RCP and CST levels are not a problem. REFERENCES 19001-C, "Reactor Trip Response". 18007-C, "cvcs Malfunction section A for Loss of Letdown." VEGP learning objectives: LO-LP-37011-03, State the bases for the "Reactor Trip Recovery" procedure. LO-LP-37011-04, State and describe the major action categories of 19001, "Reactor Trip Recovery". Page: 6 of 48 6/6/2007 ( C--" . Number Text LO-LP-37011-02 State how the following control systems are employed to automatically stabilize the plant after a reactor trip: a. steam dumps b. feedwater

c. pressurizer level and pressure d. auxiliary feedwater LO-LP-37011-03 State the bases for the "Reactor Trip Recovery" procedure.

LO-LP-37011-04 State and describe the major action categories of 19001, "Reactor Trip Recovery." LO-LP-37011-05 State the bases for the "Reactor Trip or Safety Injection" procedure. LO-LP-37011-06 State from memory the immediate action steps form 19000. Include substeps and RNO actions. LO-LP-37011-07 State the definition for the term "stable" as used in the EOPs. LO-LP-37011-0B State why the control of AFW is so important following a reactor trip. LO-LP-37011-09 Of multiple reactor trip alarms showing on the annunciator panel, state how the operator would be able to recognize which was the first one to be received. LO-LP-37011-10 List the most common reasons for a reactor trip at a PWR. Give three root causes for each type of trip (if applicable). LO-LP-37011-11 State how the operator participates in the industry-wide reactor trip reduction program. State the operator's responsibility in reducing the number of trips that may occur at Plant Vogtle. LO-LP-37011-13 List the parameters used to verify NC flow in accordance with Attachment 0 of EOP 19001. LO-LP-37011-14 Given a scenario requiring the use of the foldout page, state the actions that the operator would be required to take. LO-LP-37011-15 Given a NOTE or CAUTION statement from the EOP, state the bases for that NOTE or CAUTION statement. LO-LP-37011-16 State the importance of verifying the turbine tripped after a reactor trip. (commitment) LO-LP-37012-01 State the immediate and long-term effect on the Primary System of a loss of forced coolant flow. LO-LP-37012-02 State the natural circulation cooldown transient performed at the St. Lucie Plant in 1980 and the pressurizer level anomaly which occurred. (commitment) LO-LP-37012-03 State the operational guidelines used to establish a controlled cooldown using natural circulation. LO-LP-37012-04 State the bases for "Natural Circulation Cool down" procedure. LO-LP-37012-05 Using 19002, 19003, and 19004 as guides, summarize the actions of these emergency procedures which guide operator response in a natural circulation condition. Friday, June 01, 2007 Page 28 of 165 ( ( , I PROCEDURE NO. VEGP 18007-C REVISION NO. PAGE NO. 19 7 of 17 A. TOTAL LOSS OF LETDOWN FLOW ACTION/EXPECTED RESPONSE A6. Identify and correct cause for loss of letdown: Da. Check for letdown path valve failures or mispositions.

b. Check instrumentation:

0-PI-131A 0-TI-130 Dc. Check PIC-131. Dd. Check for other causes. DA7. Check normal letdown -AVAILABLE RESPONSE NOT OBTAINED A7. Perform the following: ?"" Da. Establish Excess Letdown by initiating 13008, CHEMICAL AND VOLUME CONTROL SYSTEM EXCESS LETDOWN. (/C If \ ( IVc d I\,f\.f c i \ £",Jr-c e 5 S Lt-J if (1/ 0 b. Restore normal letdown by initiating 13006, CHEMICAL Go to Step A9. DA8. AND VOLUME CONTROL SYSTEM. DA9. Initiate the Continuous Actions Page. D*A10. Verify PRZR level -TRENDING TO PROGRAM *A10. IF PRZR level can NOT be maintained, THEN isolate charging by performing the following: Da. Adjust FV-121 while closing HV-182 to maintain seal injection 8 to 13 gpm. Db. Close HV-8106. Dc. Notify Engineering of charging nozzle thermal cycle. Kefe,{l-fNCt-00 7 e-z, ( it.t.p I ( C REVISION N

O. PROCEDURE

110. VEGP 19001-C 29 ACTION/EXPECTED RESPONSE *10. Check PRZR level control: D a. Instrument Air -AVAILABLE Db. PRZR Level -GREATER THAN 17% Dc. Charging and -IN SERVICE letdown Dd. Maintain PRZR level at 25%. PAGE NO. 10 of 20 RESPONSE NOT OBTAINED a. Perform the following:

b. Dc. D 1) Establish Safety Grade Charging by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM. Establish Safety Grade Letdown: a) Open RX HEAD VENT TO LETDOWN ISOLATION VLVs: B: D-HV-8095A HV-8096A HV-8095B HV-8096B b) Open REACTOR HEAD VENT TO PRT flow control valves as necessary: D -HV-0442A D -HV-0442B D 3) Go to Step 10d. Perform the following:

D 1) Verify letdown isolated. D 2) Verify PRZR Heaters de-energized. D 3) Control charging to restore PRZR level to greater than 17%. D 4) WBBN PRZR level greater than 17%, THEN place letdown in service and energize PRZR Beaters as necessary. D 5) Go to Step 10d. Place charging and letdown in service by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM. PROCEDURE NO. VEGP 19001-C REVISION NO. ACTION/EXPECTED RESPONSE 14. Transfer Steam Dumps to STEAM PRESSURE mode: D a. Check Condenser -AVAILABLE Db. Place PIC-507 in Manual. D c. Match demand on SG header pressure controller PIC-507 and SD demand meter UI-500. Dd. Transfer Steam Dumps to STM PRESS mode. e. Control Tavg: D Manual control -OR-D Auto control 15. Check RCP status: D a. RCPs -ALL STOPPED Db. Start an RCP using ATTACHMENT A. (RCP 4 or RCP 1 preferred) 29 PAGE NO. 14 of 20 RESPONSE NOT OBTAINED D a. Use SG ARVs. DGo to Step 15. D a. Go to Step 16. Db. IF an RCP can NOT be started, THEN verify natural circulation using ATTACHMENT B. D IF natural circulation NOT established, THEN raise rate of dumping steam using Steam Dumps. D IF Steam Dumps not available, THEN dump steam using SG ARVs. c PROCEDURE NO. VEGP 19001-C REVISION NO. ACTION/EXPECTED RESPONSE 07. Check all Rods -FULLY INSERTED 29 PAGE NO. 7 of 20 RESPONSE NOT OBTAINED IF two or more Rods NOT fully inserted, THEN EMERGENCY BORATE 154 ppm for each Rod not fully inserted by initiating 13009, evcs REACTOR MAKEUP CONTROL SYSTEM. o Verify adequate shutdown margin as required by Technical Specification SR 3.1.1.1. ( PROCEDURE NO. VEGP 19001-C REVISION NO. ACTION/EXPECTED RESPONSE *19. Maintain stable plant conditions: O. PRZR pressure -AT 2235 PSIG O. PRZR level -AT 25% O. SG NR levels -BETWEEN 10% AND 65%

• RCS temperature:

o With RCP(s) running -RCS AVERAGE TEMPERATURE AT 557°F -OR-29 o Without RCP(s) running -RCS WR COLD LEG TEMPERATURES AT 557°F 20. Check if natural circulation cooldown is required: o a. Db. Dc. Any RCP -RUNNING Go to 12006-C, RCS COOLDOWN TO COLD SHUTDOWN. At GREATER THAN 66% .1.. / c::, J d. Perform one of the following: o Maintain hot standby conditions by returning to Step 19. -OR-o Go to 19002-C, ES-0.2 NATURAL CIRCULATION COOLDOWN based on RCP(s) restart status. PAGE NO. 16 of 20 RESPONSE NOT OBTAINED o a. Dc. Go to Step 20c. Go to 19002-C, ES-0.2 NATURAL CIRCULATION COOLDOWN. (r ( . Dip {etc.. 7 t'b It b vt + {IN'C () lL/2...e c. + I END OF PROCEDURE TEXT c ( PROCEDURE NO. VEGP 19001-C REVISION NO. PAGE NO. 29 20 of 20 1. FOLDOUT PAGE SI ACTUATION CRITERIA Actuate SI and go to Procedure 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION, if EITHER condition listed below occurs:

• RCS subcooling

-LESS THAN 24°F.

• PRZR level -CANNOT BE MAINTAINED GREATER THAN 9%. 3. AFW SUPPLY SWITCHOVER CRITERION Switch to alternate CST by initiating 13610, AUXILIARY FEEDWATER SYSTEM when CST level lowers to less than 15%.

( l 4. 00SAA2.15 001 Given the following plant conditions with both units at 100% power: -Unit 1 has one Block valve closed and de-energized to isolate a PORV that is partially stuck open. -Unit 2 has both Block valves closed and still energized to isolate both PORVs which have excessive seat leakage. Which ONE of the units would be required to shutdown due to INOPERABLE PORV I Block valve status per Tech Specs and what is the CORRECT bases? A. Unit 1 -both PORVs are required to be capable of automatically cycling to limit RCS pressure following the blowdown of a faulted Steam Generator. B. Unit 2 -both PORVs are required to be capable of automatically cycling to mitigate events such as a Steam Generator Tube Rupture or Loss of Heat Sink. Unit 1 -both PORVs are required to be capable of being manually cycled to mitigate events such as a Steam Generator Tube Rupture or Loss of Heat Sink. D. Unit 2 -both PORVs are required to be capable of being manually cycled to limit RCS pressure following the blowdown of a faulted Steam Generator. Page: 7 of 48 6/6/2007 ( 008 Pressurizer Vapor Space Accident AA2.15 Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident. ESF control board, valve controls, and indicators KIA MATCH ANALYSIS Question gives a plausible scenario where both units have PORVs closed to due either excessive seat leakage or stuck open PORVs. The candidate must determine which unit is required to shutdown per Tech Specs and the Tech Spec bases. Question meets 10CFR55.43(b) criteria for item # 2 -Facility operating limits in Tech Specs and their bases. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible the candidate may think both PORVs have to automatically cycle and this is not a correct bases. B. Incorrect. Plausible the candidate may think both PORVs have to automatically cycle and this is the correct bases. C. Correct. Both PORVs must be capable of being manually cycled and this is a correct bases. D. Incorrect. Plausible the candidate may know both capable of manually cycling but this is an incorrect bases. REFERENCES Technical Specification 3.4.11 and bases for PORVs and Block Valves VEGP learning objectives: LO-LP-39208-02, 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 3.4 are exceeded. Page: 8of48 6/6/2007 ( ( Number Text LO-LP-39206-08 Define nuclear enthalpy rise hot channel factor. LO-LP-39206-09 Define quadrant power tilt ratio. State the required action for exceeding the limit at various power levels. LO-LP-39207-01 For any given item in section 3.3 of Tech Specs, be able to: a.O State the LeO. b.O State anyone hour or less required actions. LO-LP-39207 -02 Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a.D Whether any Tech Spec LeOs of section 3.3 are exceeded. b.O The required actions for all section 3.3 LeOs. LO-LP-39207-03 For any given item in section 13.3 of the Technical Requirements Manual, be able to: a.D State the Technical Requirement (TR) for operation. b.D State anyone hour or less required actions. LO-LP-39207-04 Describe the bases for any given Tech Spec in section 3.3. LO-LP-39207-05 Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: a.D Whether any TR of section 13.3 has been exceeded. b.O The required actions for all section 13.3 TRs. LO-LP-39207-06 State the values for the Limiting Safety System Settings. LO-LP-39208-01 For any given item in section 3.4 of Tech Specs, be able to: a.O State the LeO. b.O State anyone hour or less required actions. LO-LP-39208-02 Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a.D Whether any Tech Spec LeOs of section 3.4 are exceeded. b.O The required actions for all section 3.4 LeOs. LO-LP-39208-03 For any given item in section 13.4 of the TRM, be able to: a.O State the TR for operation. b.O State anyone hour or less required actions. LO-LP-39208-04 Describe the bases for any given Tech Spec in section 3.4. LO-LP-39208-05 State why the mode 5 ReS loops and coolant circulation specifications for loops filled vs. loops not filled vary with regard to the number of RHR trains required and steam generator availability. LO-LP-39208-06 State the reason for limiting the ReS specific activity. Friday, June 01, 2007 Page 42 of 165 -l U IIIJ-r /) r JI 0 n !?cl\.l 00 % A..Jl Pressurizer PORVs f"\ E',e'fl...-{" lire -, ( , I rr l,r I ) 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) LCO 3.4.11 Each PORV and associated block valve shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS -------------------------------NOTE--------------------------------- Separate Condition entry is allowed for each PORV. -------------------------------------------- CONDITION REQUIRED ACTION COMPLETION TIME A. One or more PORVs A.1 Close and maintain power 1 hour () v l 1-inoperable and capable to associated block valve. O(1t of being manually 5f-,'* ( cycled. o eN.l pne PORV inoperable B.1 Close associated block 1 hour f /'and not capable of being valve. U('vfJt-l 4. manually cycled. AND A-B.2 Remove power from 1 hour associated block valve. AND (d-.. f V hctse) B.3 Restore PORV to 72 hours OPERABLE status. COI\.I\ wrJ CU/\,cL1'c.rJ 8 IIIpt" D --:; eCT (continued) hCt 'JC) J Vogtle Units 1 and 2 3.4.11-1 Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2) ( ACTIONS (continued) CONDITION C. One block valve C.1 inoperable. AND C.2 D. Required Action and D.1 associated Completion Time of Condition A, B, AND or C not met. D.2 E. Two PORVs inoperable E.1 and not capable of being manually cycled. AND E.2 AND E.3 AND E.4 F. More than one block F.1 valve inoperable. AND c. Vogtle Units 1 and 2 Lt 00 f1;r Pressurizer PORVs 3.4.11 REQUIRED ACTION COMPLETION TIME Place associated PORV 1 hour in manual control. Restore block valve to 72 hours OPERABLE status. Be in MODE 3. 6 hours Be in MODE 4. 12 hours Close associated block 1 hour valves. Remove power from 1 hour associated block valves. Be in MODE 3. 6 hours Be in MODE 4. 12 hours Place associated PORVs 1 hour in manual control. 3.4.11-2 (continued) Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ( ( ACTIONS CONDITION REQUIRED ACTION F. (continued) F.2 Restore one block valve to OPERABLE status. AND F.3 Restore remaining block valve to OPERABLE status. G. Required Action and G.1 Be in MODE 3. associated Completion Time of Condition F not AND met. G.2 Be in MODE 4. SURVEILLANCE REQUIREMENTS SR 3.4.11.1 SR 3.4.11.2 SURVEILLANCE

---

NOTE--------------------------------

Not required to be performed with block valve closed in accordance with the Required Action of Conditions A, B, or E. Perform a complete cycle of each block valve. Perform a complete cycle of each PORV. Pressurizer PORVs 3.4.11 COMPLETION TIME 2 hours 72 hours 6 hours 12 hours FREQUENCY 92 days 18 months Vogtle Units 1 and 2 3.4.11-3 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) c ( c ... Pressurizer PORVs B 3.4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) BASES BACKGROUND Vogtle Units 1 and 2 The pressurizer is equipped with two types of devices for pressure relief pressurizer safety valves and PORVs. The PORVs are related DC solenoid operated valves that are controlled to open at a specific set pressure when the pressurizer pressure increases and close when the pressurizer pressure decreases. The PORVs may also be manually operated from the control room. Block valves, which are normally open, are located between the pressurizer and the PORVs. The block valves are used to isolate the PORVs in case of excessive leakage or a stuck open PORV. Block valve closure is accomplished manually using controls in the control room. A stuck open PORV is, in effect, a small break loss of coolant accident (LOCA). As such, block valve closure terminates the RCS depressurization and coolant inventory loss. The PORVs and their associated block valves may be used by plant operators to depressurize the RCS to recover from certain transients if normal pressurizer spray is not available. Additionally, the series arrangement of the PORVs and their block valves permit performance of surveillances on the block valves during power operation. The PORVs may also be used for feed and bleed core cooling in the case of multiple equipment failure events that are not within the design basis, such as a total loss of feedwater. The power supplies to the PORVs, their block valves, and their controls are Class 1 E. Two PORVs and their associated block valves are powered from two separate safety trains (Ref. 1). The plant has two PORVs, each having a relief capacity of 210,000 Ib/hr at 2385 psig. The functional design of the PORVs is based on maintaining pressure below the Pressurizer Pressure-High reactor trip setpoint up to and including the design step-load decreases with steam dump. In addition, the PORVs minimize challenges to the pressurizer (continued) B 3.4.11-1 Revision No. 0 c c BASES BACKGROUND (continued) APPLICABLE SAFETY ANALYSES f ,,-,1' 1 fQtVA-+D ?i,.,J \ct(V Vogtle Units 1 and 2 Pressurizer PORVs B 3.4.11 safety valves and also may be used for cold overpressure protection. See LCO 3.4.12, "Cold Overpressure Protection System (COPS)." Plant operators may employ the PORVs to depressurize the RCS in response to certain plant transients if normal pressurizer spray is not available. For the Steam Generator Tube Rupture (SGTR) event, the safety analysis assumes that manual operator actions are required to mitigate the event. A loss of offsite power is assumed to accompany the event, and thus, normal pressurizer spray is unavailable to reduce RCS pressure. The PORVs or auxiliary pressurizer spray may be used for RCS depressurization, which is one of the steps performed to equalize the primary and secondary pressures in order to terminate the primary to secondary break flow and the radioactive releases from the affected steam generator. In addition, in the event of an inadvertent safety injection actuation at power, the potential for pressurizer filling and subsequent water relief via the pressurizer safeties (PSVs) is evaluated (FSAR section 15.5.1). Operator action to make one PORV available is credited in the analysis to mitigate this event. If the PORV is available for automatic actuation, the event consequences would be mitigated directly by preventing water relief through the PSVs. However, automatic actuation is not required to mitigate this event. The analysis includes an acceptable delay for the operator to open a block valve and to manually control the PORV if necessary. The PORVs also provide the safety-related means for reactor coolant system depressurization to achieve safety-grade cold shutdown and to mitigate the effects of a loss of heat sink or an SGTR. They are modeled in safety analyses for events that result in increasing RCS pressure for which departure from nucleate boiling ratio (DNBR) criteria, pressurizer filling, or reactor coolant saturation are critical (Ref. 2). By assuming PORV actuation, the primary pressure remains below the high pressurizer pressure trip setpoint, thus the DNBR calculation is more conservative. As such, automatic actuation is not f-lee required to mitigate these events, and PORV automatic operation is, ..J therefore, not an assumed safety function. Events that assume this ovtt-fI-.."..,J; condition include a turbine trip, loss of normal feedwater, and feedwater line break (Ref. 2). Pressurizer PORVs satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii). (continued) B 3.4.11-2 Rev. 3-10/01 c BASES LCO APPLICABILITY Vogtle Units 1 and 2 Pressurizer PORVs B 3.4.11 The LCO requires the PORVs and their associated block valves to be OPERABLE for manual operation to mitigate the effects associated with an SGTR, or loss of heat sink, and to achieve safety grade cold shutdown. The PORVs are considered OPERABLE in either the manual or automatic mode. The PORVs (PV-455A and PV-456A) are powered from 125 V MCCs 1I2AD1M and 1/2BD1M, respectively. If either or both of these MCCs become inoperable, the affected PORV(s) are to be considered inoperable. By maintaining two PORVs and their associated block valves OPERABLE, the single failure criterion is satisfied. An OPERABLE PORV is required to be capable of manually opening and closing, and not experiencing excessive seat leakage. Excessive seat leakage, although not associated with a specific criteria, exists when conditions dictate closure of the block valve to limit leakage. An OPERABLE block valve may be either open and energized, or closed and energized with the capability to be opened, since the required safety function is accomplished by manual operation. Although typically open to allow PORV operation, the block valves may be OPERABLE when closed to isolate the flow path of an inoperable PORV that is capable of being manually cycled (e.g., as in the case of excessive PORV leakage). Similarly, isolation of an OPERABLE PORV does not render that PORV or block valve inoperable provided the relief function remains available with manual action. Satisfying the LCO helps minimize challenges to fission product barriers. The PORVs are required to be OPERABLE in MODES 1, 2, and 3 for manual actuation to mitigate a steam generator tube rupture event, an inadvertent safety injection, and to achieve safety grade cold shutdown. In addition, the block valves are required to be OPERABLE to limit the potential for a small break LOCA through the flow path. The most likely cause for a PORV small break LOCA is a result of a pressure increase transient that causes the PORV to open. Imbalances in the energy output of the core and heat removal by the secondary system can cause the RCS pressure to increase to the PORV opening setpoint. The most rapid increases will occur at the higher operating power and pressure conditions of MODES 1 and 2. Pressure increases are less prominent in MODE 3 because the core input energy is reduced, but the RCS pressure is high. Therefore, the LCO is applicable in MODES 1, 2, and 3. The LCO is not applicable in MODES 4, 5, and 6 with the reactor vessel head in place when both pressure and core energy are decreased and the pressure surges become much less significant. LCO 3.4.12 addresses the PORV (continued) B 3.4.11-3 Rev. 1-2/00 ( BASES APPLICABILITY (continued) ACTIONS Vogtle Units 1 and 2 Pressurizer PORVs B 3.4.11 requirements in MODES 4, 5, and 6 with the reactor vessel head in place. A Note has been added to clarify that all pressurizer PORVs are treated as separate entities, each with separate Completion Times (Le., the Completion Time is on a component basis). PORVs may be inoperable and capable of being manually cycled (e.g., excessive seat leakage, instrumentation problems, or other causes that do not create a possibility for a small break LOCA). In this condition, either the PORVs must be restored or the flow path isolated within 1 hour. The associated block valve is required to be closed, but power must be maintained to the associated block valve, since removal of power would render the block valve inoperable. The PORVs may be considered ((\..il c, l-OPERABLE in either the manual or automatic mode. This permits e 6't.. operation of the plant until the next refueling outage (MODE 6) so that Pr ¢-'D maintenance can be performed on the PORVs to eliminate the problem condition. Quick access to the PORV for pressure control can be made when power remains on the closed block valve. The Completion Time of 1 hour is based on plant operating experience that has shown that minor problems can be corrected or closure accomplished in this time period. B.1, B.2. and B.3 If one PORV is inoperable and not capable of being manually cycled, it must be either restored or isolated by closing the associated block valve and removing the power to the associated block valve. The Completion Times of 1 hour are reasonable, based on challenges to the PORVs during this time period, and provide the operator adequate time to correct the situation. If the inoperable valve cannot be restored to OPERABLE status, it must be isolated within the specified time. Because there is at least one PORV that remains OPERABLE, an additional 72 hours is provided to restore the inoperable PORV to ('c v a IV )LvefU (continued) B 3.4.11-4 Rev. 2 -6/05 BASES ACTIONS Vogtle Units 1 and 2 B.1. B.2. and B.3 (continued) Pressurizer PORVs B 3.4.11 OPERABLE status. If the PORV cannot be restored within this additional time, the plant must be brought to a MODE in which the LCO does not apply, as required by Condition D. C.1 and C.2 If one block valve is inoperable, then it is necessary to either restore the block valve to OPERABLE status within the Completion Time of 1 hour or place the associated PORV in manual control. The prime importance for the capability to close the block valve is to isolate a stuck open PORV. Therefore, if the block valve cannot be restored to OPERABLE status within 1 hour, the Required Action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck open PORV at a time that the block valve is inoperable. The Completion Time of 1 hour is reasonable, based on the small potential for challenges to the system during this time period, and provides the operator time to correct the situation. The time allowed to restore the block valve is based upon the Completion Time for restoring an inoperable PORV in Condition B since the PORV may not be capable of mitigating an event if the inoperable block valve is not fully open. If the block valve is restored within the Completion Time of 72 hours, the PORV may be restored to automatic operation. If it cannot be restored within this additional time, the plant must be brought to a MODE in which the LCO does not apply, as required by Condition D. D.1 and D.2 If the Required Action of Condition A, B, or C is not met, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, maintaining PORV OPERABILITY may be required. See LCO 3.4.12. (continued) B 3.4.11-5 Rev. 1-2/00 ( ( BASES ACTIONS (continued) Vogtle Units 1 and 2 Pressurizer PORVs B 3.4.11 E.1. E.2. E.3. and E.4 If more than one PORV is inoperable and not capable of being manually cycled, it is necessary to either restore at least one valve within the Completion Time of 1 hour or isolate the flow path by closing and removing the power to the associated block valves. The Completion Time of 1 hour is reasonable, based on the small potential for challenges to the system during this time and provides the operator time to correct the situation. If one PORV is restored and one PORV remains inoperable, then the plant will be in Condition B with the time clock started at the original declaration of having two PORVs inoperable. If no PORVs are restored within the Completion Time, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, maintaining PORV OPERABILITY may be required. See LCO 3.4.12. F.1. F.2. and F.3 If more than one block valve is inoperable, it is necessary to either restore the block valves within the Completion Time of 1 hour, or place the associated PORVs in manual control and restore at least one block valve within 2 hours and restore the remaining block valve within 72 hours. The Completion Times are reasonable, based on the small potential for challenges to the system during this time and provide the operator time to correct the situation. G.1 and G.2 If the Required Actions of Condition F are not met, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant (continued) B 3.4.11-6 Revision No. a ( c BASES ACTIONS SURVEILLANCE REQUIREMENTS REFERENCES Vogtle Units 1 and 2 Pressurizer PORVs B 3.4.11 G.1 and G.2 (continued) conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, maintaining PORV OPERABILITY may be required. See LCO 3.4.12. SR 3.4.11.1 Block valve cycling verifies that the valve(s) can be closed if needed. The basis for the Frequency of 92 days is the ASME Code, Section XI (Ref. 2). The Note modifies this SR by stating that it is not required to be performed with the block valve closed, in accordance with the Required Actions of Conditions A, B, or E. SR 3.4.11.2 SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. The Frequency of 18 months is based on a typical refueling cycle and industry accepted practice.

1. Regulatory Guide 1.32, February 1977. 2. ASME, Boiler and Pressure Vessel Code, Section XI. C(/6 Iht 2-( {5 B 3.4.11-7 Revision No. 0

( ( 5. 0151017G2.4.4 001 Unit 2 is at 25% power when the following annunciators are received. -ALB08 window B05 for RCP # 3 CONTROLLED LKG HI! LO FLOW -ALB08 window B04 for RCP # 3 NO.2 SEAL LKOF HI FLOW The RO reports the following indications: -RCP # 3 seal leakoff flow Hi Range meter is 6.0 gpm. -RCP # 3 seal injection flow is 7.9 gpm. -RCP # 3 Seal Water Inlet temperature is 223 degrees F and stable. Which ONE of the the following is the CORRECT procedurally directed action(s) for the SS to take? A. Per UOP-12004-C, "Power Operations (Mode 1)", commence a unit shutdown to be in Mode 3 in 8 hours. B:o' Trip the reactor and enter E-O, "Reactor Trip and Safety Injection", per SOP-13003, "RCP Operation", stop RCP # 3 and close sealleakoff valve HV-8141C. C. Per SOP-13003, stop RCP # 3, close sealleakoff valve HV-8141 C, enter AOP-18005, "Partial Loss of RCS Flow", commence unit shutdown per UOP-12004. D. Per UOP-12004, maintain reactor power at 25%, monitor the RCP per SOP-13003 section 4.2.1 "Pump Operation With A Seal Abnormality", contact Duty Engineering. Page: 9 of 48 6/6/2007 ( 015/017 RCP Malfunctions G2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. KIA MATCH ANALYSIS Question gives a plausible scenario with indications of an RCP # 3 seal failure. Candidate must choose the correct procedural actions to address the failure. Question meets 1 OCFRSS.43(b) criteria item # S -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency situations. ANSWER 1 DISTRACTOR ANALYSIS A. Incorrect. Plausible the candidate may not recognize the sealleakoff indications of >S.S gpm require an immediate RCP shutdown and a normal shutdown desired. B. Correct. per SOP-13003 and AOP-1800S if reactor power> 1S% and an RCP is to be tripped, correct action is to trip reactor and go to E-O. RCP should be stopped after IOAs of E-O and sealleakoff valve closed. C. Incorrect. Plausible the candidate may recognize the plant is < P-8 and recognize an auto reactor trip would not occur and think AOP-1800S-C Partial Loss of Flow entry is appropriate. SOP13003 and 1800S both state if reactor power> 1S% trip the reactor and enter E-O. D. Incorrect. Immediate trip criteria for RCP is exceeded due to excessive sealleakoff. Plausible the candidate may not recognize the immediate trip criteria and know that Duty Engineering would be contacted. REFERENCES AOP-1800S-C, "Partial Loss of RCS Flow". SOP-13003-1/2, "Reactor Coolant Pump Operation" Precautions and Limitations, and section 4.2.1 for Operation With a Seal Abnormality with the appropriate Decision Tree Flow Chart (figure). ARP-17008-1/2 windows COS and C04 for RCP # 3 Controlled Leakage Hi I Lo Flow and # 3 Seal Leakoff Hi Flow. Farley December 2004 NRC SRO Exam question # 4. c=.. VEGP learning objectives: LO-PP-16401-03, Describe the control room indications for a failure of an RCP seal. Page: 10 of 48 6/6/2007 ( ( NO OBI TX OBI LO-PP-16302-0 Compare and contrast hot and cold calibrated Pressurizer level indications under various PZR operating conditions using the PTDB. LO-PP-16302-0 State the LCO, applicability, bases, and the 1 hr or less actions for each of the following:

a. 3.3.1 Function 9 Pressurizer Water Level-High

b. 3.3.3 Function 6 Pressurizer Level c. 3.3.4 Function 8 Pressurizer Level LO-PP-16303-0 Describe the response of the pressurizer pressure control system to variations in pressurizer pressure I level. LO-PP-16303-0 Describe how the response of pressurizer pressure 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. stuck open PORV e. stuck open spray valve LO-PP-16303-0 State the reactor trips and SI actuation signals, including set points and coincidences associated with pressurizer pressure protection channels.

LO-PP-16303-0 Describe the permissives associated with pressurizer pressure including set point and coincidences and what it provides. LO-PP-16303-0 State the set point, coincidence, and protective actuations associated with the low pressurizer PORV interlock. LO-PP-16303-0 State the LCO, applicability, bases, and the 1 hr or less actions for each of the following:

a. 3.3.1 Function 6 OTDT (Rx Trip) b. 3.3.1 Function 8a. Pressurizer Pressure Low (Rx Trip) c. 3.3.1 Function 8b. Pressurizer Pressure High (Rx Trip) d. 3.3.2 Function 1d. Pressurizer Pressure Low (SI) e. 3.3.2 Function 8b. Pressurizer Pressure P-11 LO-PP-16401-0 Explain the function of the following RCP components:

a. Thermal Barrier b. Pump Seal Package c. Thrust bearing d. Motor Flywheel e. Anti Rotation Device 1. Oil Lift pump LO-PP-16401-0 Describe the function of RCP seals 1, 2, and 3 including DP across each seal and expected flow rate. LO-PP-16401-0 Describe the control room indications for a failure of a RCP seal. LO-PP-16401-0 State what the effect of closing the #1 seal leak off valve. Friday, June 01, 2007 Page 22 of 68 c ( Approved By C. H. Williams, Jr. Date Approved 1/1/2004 Vogtle Electric Plant A ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON PANEL 1A2 ON MCB Procedure Number Rev 17008-1 13.3 Page Number 24 of 46 WINDOW C04 ORIGIN SETPOINT 1-FIS-0192 0.9 gpm 1.0 PROBABLE CAUSE 1. Number 2 Seal failure. 2. Sudden reduction in RCDT level or pressure.

2.0 AUTOMATIC

ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS RCP3 NO.2 SEAL LKOF HI FLOW 1. Check RCDT pressure on 1-PISL-9699 (QPCP) 3 psig or greater. 2. Dispatch Operator to check RCDT pressure and level at PLPP: a. Pressure 2-3 psig, b. Level 20-75%. 3. IF RCDT pressure and level are normal, Go To 13003-1, "Reactor Coolant Pump Operation" for instructions covering RCP operation with seal malfunctions.

4.0 SUBSEQUENT

OPERATOR ACTIONSS NONE 5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE (

REFERENCES:

1X4DB114, 1X6AB09-119, PLS 015/ 00&-2.,.,'+, 4 Pnnted June 6, 2007 at 1.16 ( ( Approved By Procedure Number Rev C. H. Williams, Jr. Vogtle Electric Generating Plant .A 17008-1 13.3 Date Approved 1/1/2004 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON PANEL 1A2 ON MCB Page Number 25 of 46 ORIGIN 1-FT-0159 1-FT-0155 SETPOINT 4.8 gpm 0.8 gpm WINDOWC05 RCP3 CONTROLLED LKG HIILO FLOW 1.0 PROBABLE CAUSE 1. High Flow: a. Flashing in the Seal Leakoff Line due to loss of seal injection flow or high seal injection temperature, b. Failure of Number 1 Seal. 2. Low Flow: a. Low differential pressure across Number 1 Seal, b. High Volume Control Tank (VCT) pressure, c. Excess letdown in service, d. Failure of Number 2 Seal. 2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NOTE RCP 3 No. 1 seal water leakoff high range flow may be monitored using computer point F0159. 1. Observe seal injection flow and sealleakoff flow, as well as excess letdown temperature and pressure for indication of an actual seal anomaly. 2. IF a seal problem is indicated, Go To 13003-1, "Reactor Coolant Pump Operation" . 3. IF an instrument problem is indicated, initiate maintenance as required. Pnnted June 6, 2007 at 1: 16 ( ( ( PROCEDURE NO. VEGP 18005-C REVISION NO. ACTION/EXPECTED RESPONSE 10 PAGE NO. 3 of 4 RESPONSE NOT OBTAINED 1. Perform the following: D 1. Check Reactor power -LESS THAN OR EQUAL TO 15% < > D a . Trip the Reactor. Db. Go to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION. D 2. Stop any power changes in progress. D 3. Initiate the Continuous Actions Page. D

• 4. Check affected loop SG NR Level -TRENDING TO 65% D 5. Check Tavg -TRENDING TO PROGRAM D 6. Verify PRZR level -TRENDING TO PROGRAM D 7. Verify PRZR pressure -TRENDING TO 2235 PSIG D 8. Check RCP 1 and RCP 4 -RUNNING D 9. Initiate shutdown to Mode 3 by initiating 12004-C, POWER OPERATION

{MODE 1}. {TS 3.4.4} D 10. Determine and correct the cause of the pump trip. D 11. Check shutdown to Mode 3 -COMPLETE D

• 4. Control feed flow to maintain affected loop SG NR level between 60% and 70%. D 5. Adjust control rods to restore Tavg. 8. Close the affected loop spray valve: D Loop 1: PIC-0455C D Loop 4: PIC-0455B D 11. Return to Step 9.

Approved By S. A. Phillips Date Approved C 3-8-2007 4.2 4.2.1 4.2.1.1 4.2.1.2 4.2.1.3 ( VogtJe Electric Generating Plant A Procedure Number Rev 13003-1 35 REACTOR COOLANT PUMP OPERATION Page Number 11 of 32 I INITIALS SYSTEM OPERATION GiS 0/7 &-Lrl.{-, Pump Operation With A Seal Abnormality IF the Plant Computer is available, trend the computer data points listed in Table 2. IF the Plant Computer is NOT available, perform the following:

a. Monitor the QMCB indication listed in Table 2 at least hourly for the next 8 hours. b. IF NO further seal degradation exists after 8 hours, consult with the Shift Supervisor (55) for less frequent monitoring.

Monitor the No. 1 seal for further degradation using Figure 1 and RCP Trip Criteria as follows: a. Evaluate the monitored indications using Figure 1, "RCP Seal Abnormalities Tree". b. IF evaluation of the monitored indications using Figure 1 requires immediate pump shutdown, Go To Step 4.2.1.4. c. IF any of the following RCP Trip Criteria is exceeded, Go To Step 4.2.1.4 for immediate RCP shutdown. RCP TRIP CRITERIA Motor bearing temperature >195°F Motor stator-winding temperature >311°F Seal water inlet temperature >230 0 F ,'N RCP shaft vibration =20 mils U\ RCP Frame vibration =5 mils 3f-f fV\.. ________ ___ <_2O_O __ ps_i_d ____ Total loss of ACCW for a duration of 10 minutes J Pnnted June 6,2007 at 1:14 ( ( Approved By S. A. Phillips VOgtie Electric Generating Plant Date Approved 3-8-2007 REACTOR COOLANT PUMP OPERATION FIGURE 1 -RCP SEAL ABNORMALITIES DECISION TREE Yes Yes No r No N°'4-___ No No Yes No NOl1FY DUTY E NG TO CO N SULT PLANT Yes MANA G EM E NT FO R Note 1 MANAGEMENT AD VI SE S P U MP BE SHUTlXl lMl Yes A C 11 0 N S No SHUTDO lMI WITH I N e H OU R S PER 42.1 No No No NO Note 2 ..---Yes-----' Note 1: Abnormal Operating Range of Figure 2 Note 2: Non-operating Range of Figure 2 Note 3: ALB08 A-04, B-04, C-04 or 0-04 Printed June 6, 2007 at 1:15 Yes Note 2 No CHECK NO.2 SEAL LEAi<OFF FLOW Yes Yes FAI LU RE OF N O.3 OUTER DAM FAIL U RE OF N O.2 S EAL REPAIR A T NE XT OUT A G E Procedure Number Rev 13003-1 35 Page Number 29 of 32 No-------, Yes No No Yes Yes No No Approved By Procedure Number Rev S. A. Phillips 13003-1 35 Date Approved Page Number C 3-8-2007 REACTOR COOLANT PUMP OPERATION 30 of 32 ( ... o z NOTE 2 '\. FIGURE 2 NO. 1 SEAL NORMAL OPERA liNG RANGE 200 500 1,000 NORMAL OPERATING RANGE 1,500 2,000 I 2,500 2,250 No. 1 Seal Differential Pressure (PSI) NOTE 3 1 [ 1. If the No.1 seal leak rates are outside the normal (1.0-5.0 gpm) but within the {\ VI., -e5 limits <<0.8-5.5 gpm), continue pump operation. VERIFY that seal injection flow exceeds No. 1 seal leak rate for the affected RCP. Closely monitor II: +-0 pump and seal parameters and contact Engineering for further instructions.

2. Printed June 6.2007 at 1:15 Minimum startup requirements are 0.2 gpm at 200 PSID differential across the No.1 seal. For startups at differential pressures greater than 200 PSID, the minimum No. 1 seal leak rate requirements are defined in the NO. 1 SEAL NORMAL OPERATING RANGE (e.g., at 1000 psi differential pressure, do not start the RCP with less than 0.5 gpm). No.1 Seal Differential Press = RCS WR Press -VCT Press.

( l Approved By S. A. Phillips Date Approved 3-8-2007 4.2.1.4 Vogtle Electric Generating Plant A. REACTOR COOLANT PUMP OPERATION

d. WHEN directed by Figure 1, stop the affected RCP within 8 hours as follows: (1) Establish 9 gpm or greater seal injection flow to the affected pump. (2) Stop the affected RCP by continuing with step 4.2.1.4. WHEN directed by the SS, perform an RCP shutdown as follows: a. Start the RCP Oil Lift Pump for affected RCP, if available.

b. IF Reactor Power is greater than 15% Rated Thermal Power: ('; 0-.. (( (1) Trip the Reactor and initiate 19000-C, "E-O ,/ LJ Reactor Trip Or Safety Injection".

(2) WHEN the immediate operator actions of 19000-C are complete, Go To Step 4.2.1.4.d.

c. IF Reactor Power is less than 15% Rated Thermal Power, initiate 18005-C, "Partial Loss Of Flow". d. Stop the RCP by placing the RCP Non-1 E Control Switch in STOP and then placing the RCP 1 E Control Switch in STOP: RCP Non-1 E Control Switch 1 E Control Switch

• Loop 1 1-HS-04958 1-HS-0495A
• Loop 2 1-HS-04968 1-HS-0496A
• Loop 3 1-HS-04978 1-HS-0497A
• Loop 4 1-HS-04988 1-HS-0498A Pnnted June 6, 2007 at 1: 14 Procedure Number Rev 13003-1 35 Page Number 12 of 32 INITIALS Approved By S. A. Phillips Vogtle Electric Generating Plant A Date Approved C 3-8-2007 ( ( REACTOR COOLANT PUMP OPERATION CAUTION If RCP #1 or #4 is stopped, the associated Spray Valve is placed in manual and closed to prevent spray short cycling. e. IF RCP #1 OR #4 is stopped, verify its associated spray valve is placed in MANUAL AND closed.
• RCP 1: 1-PIC-0455C
• RCP 4: 1-PIC-0455B

f. WHEN the RCP comes to a complete stop (as indicated by reverse flow), close the RCP Seal Leakoff Isolation valve for the affected pump.

• RCP 1: 1-HV-8141A
• RCP 2: 1-HV-8141B
• RCP 3: 1-HV-8141C
• RCP 4: 1-HV-8141D

g. Secure the associated RCP Oil Lift Pump. h. IF RCP shutdown was due to loss of RCP seal cooling, review Limitation 2.2.11 for recovery action. Of::) Pnnted June 6,2007 at 1:14 Procedure Number Rev 13003-1 35 Page Number 13 of 32 INITIALS c l Page: 1 You are the Unit 2 SRO. Unit 2 is at 100% steady-state power. All systems are in automatic and functioning properly.

The following annunciators are received: -DC1, "RCP #1 SEAL LKOF FLOW LO" -DA5, "2A RCP #2 SEAL LKOF FLOW HI" The plant operator reports the following parameters: RCP #1 seal injection flow (gpm) #1 seal leakoff flow (gpm) #1 seal DIP (psid) o RCP lower seal water brg. temp ( F) 2A 7.4 stable 0.0 stable >400 stable 190 increasing 2B 7.3 stable 4.0 stable >400 stable 124 stable 2C 7.4 stable 4.0 stable >400 stable 123 stable Which ONE of the following is the most probable cause of these indications and the required actions? A. 2A RCP #1 seal failure, trip the reactor, and secure the RCP. B. 2A RCP #1 seal failure, perform a controlled shut down to Mode 3 in 6 hours. 2A RCP #2 seal failure, trip the reactor, and secure the RCP. D. 2A RCP #2 seal failure, monitor 2A RCP parameters for further degredation, contact Westinghouse for further guidance. Source: Modified from Farley Exam Bank Question #RCP SEAL-52522A03 This question satisfy the criteria in 10CFR55.43(b)(5). A -Incorrect, Flow into the #1 seal is satisfactory. DIP across the #1 seal is satisfactory. Correct action for a failed #1 seal. B -Incorrect, Flow into the #1 seal is satisfactory. DIP across the #1 seal is satisfactory. Action is the required TS action for one RCS loop becoming inoperable. C -Correct, Evidenced mostly be the #1 sealleakoff flow at 0.0 which shows that all the flow is going through the #2 seal indicating its failure. Annunciator DA5, 1A RCP #2 SEAL LKOF FLOW HI, confirms this failure, along with the slighty elevated RCP radial brg temp. RCP radial bearing not being stable is annunciator DC1 criteria for tripping the reactor and securing the RCP. D -Incorrect, #2 seal failure is indicated. Action is for stable nondegrading parameters. fqCG(ey I/ec-7ft If: O[ S j(j/7G-2-(i+c L\

6. 025AA2.03 001 Given the following conditions: ( -Unit 1 cooldown for refueling outage in progress.

-RCS temperature is 185 degrees F. -Charging and letdown flows are balanced. -RHR pump 1B is operating in shutdown cooling mode, RHR pump 1A is in standby. The following occurs: -Containment sump levels begin to rise with radiation alarms received on RE-002 and RE-003. -PRZR level is 90% and continuing to lower. The Unit SS should implement which ONE of the following: At! 18004-C, RCS Leakage, section C for RCS Leakage (Mode 5). B. 18004-C, RCS Leakage, section B for RCS Leakage (Mode 3 < 1000 psig and 4). C. 18019-C, Loss of RHR, section A for Loss of RHR capability in Mode 4 or Mode 5 with PRZR Level in the IR. D. 18019-C, Loss of RHR section B for Loss of RHR in Mode 5 or 6 Below PRZR IR or SG Nozzle Dams Installed. Page: 11 of 48 6/6/2007 c 025 Loss of Residual Heat Removal (RHR) System AA2.03 Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System. Increasing reactor building sump level. KIA MATCH ANALYSIS Question gives a plausible scenario where Containment sump levels, radiation, and PRZR level indicate an RCS leak in Containment. The candidate must determine the plant Mode from the given conditions and select the apppriate section of 18004 for RCS Leakage or 18019 for Loss of RHR to enter. Question meets 1 OCFR55.43(b) criteria for item # 5 -Assessment of facility conditions and selection of procedures during normal, abnormal, or emergency conditions. ANSWER I DISTRACTOR ANALYSIS A. Correct. 18004-C, section C should be used for this condition.This is a new section and a recent change to our procedures B. Incorrect. Plausible the candidate may not properly recognize the plant mode and think section B of 18004 is appropriate. C. Incorrect. Plausible the candidate may select this section of 18019-C since the description fits the plant conditions, section A is just for loss of capability, not leakage. D. Incorrect. Plausible the candidate may select this section of 18019-C since there is an RCS leak while on RHR, however PRZR level rules out this section. REFERENCES 18004-C, RCS Leakage 18019-C, Loss of Residual Heat Removal (RHR) VEGP learning objectives: LO-LP-60304-08, Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). (18004 objective) LO-LP-60314-04, Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). ( 18019 objective) Page: 11 of 48 6/6/2007 Number Text LO-LP-S0303-12 State the Immediate operator actions required for an uncontrolled continuous rod motion. Include RNO and substeps of the immediate action. LO-LP-S0303-14 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-S0303-15 Describe the effects of failing to reset the PIA converter (Bank Demand Position Display) following a misaligned rod recovery. LO-LP-S0303-1S Describe how the radial flux profile may be affected by a misaligned rod. LO-LP-60303-17 Describe why reactor power must be less than 65% or 10% below most limiting power distribution restriction prior to commencing a realignment of a control rod. LO-LP-60303-18 Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). LO-LP-S0303-19 Given the entire AOP, describe:

a. Purpose of selected steps b. how and why the step is being performed

c. Expected response of the plantlparameter(s) for the step LO-LP-60303-20 Describe the required operator actlon(s) if two or more. rods drop. LO-LP-60303-21 Given power level, QPTR and/or AFD conditions, describe ail Tech Spec required actions of 1 hour or less. LO-LP-60304-01 State the allowable limits for Reactor Coolant System leakage, as defined in Technical Specifications.

LO-LP-60304-02 Describe how leakage into containment from the RCS will affect the containment pressure, temperature, humidity , and activity. LO-LP-60304-03 State the Immediate operator action required for Reactor Coolant System leakage within RCS makeup capabilities. Include RNO and substeps of the immediate action. LO-LP-60304-04 Given the symptoms of RCS leakage into an area or system, correctly identify the leakage area or system. LO-LP-60304-05 State the setpoint at which the charging pump suction shifts to RWST on low VCT level. LO-LP-60304-06 Describe why the actions required for inability to maintain pressurizer level are different depending on whether you are in the procedure for *RCS leakage within RCS makeup capabilities* or in the procedure for *RCS leakage greater than RCS makeup capabilities while in modes 4 or 5.* LO-LP-60304-0B Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). LO-LP-60304-09 Given the entire AOP, describe:

a. Purpose of selected steps. b. How and why the step is being performed.

c. Expected response of the plantlparameter(s) for the step. LO-LP-60304-10 Given conditions and/or indications of leaks identified in Table 1 of AOP 1 B004-C, determine the probable location of the leakage per 1 B004-C. LO-LP-60304-11 Explain what is meant by 'RCS leak before break criteria.

Friday, June 01, 2007 Page 71 of 165 (/ c. Number Text LO-LP-60314-01 Given that the operator does not adjust turbine load, describe how and why the following parameters change subsequent to a loss of two strings of feedwater heaters. Indicate direction of change only. a. reactor power b. indicated turbine load (pimp) c. generator output (MW) d. Tavg e. Tc f. Th LO-LP-60314-02 Describe the operator actions required if during the performance of AOP 18016-C *Condensate and Feedwater Malfunction* a loss of SG level is imminent. . LO-LP-60314-03 State actions required on loss of both SGFP's with power> 20% and power < 20%. LO-LP-60314-04 Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). LO-LP-60314-0S Given the entire AOP, describe:

a. Purpose of selected steps b. How and why the step is being performed

c. Expected response of the plantlparameter(s) for the step LO-LP-6031S-01 Describe factors that can lead to a loss of RHA. LO-LP-6031S-02 State the possible consequences of a sustained loss of RHA. LO-LP-6031S-03 Given figures 1-S of AOP 18019-C , determine minimum ECCS flow , time to saturation, time to core uncovery, and heatup rate. LO-LP-6031S-04 Given the entire AOP, describe:

a. Purpose of selected steps b. How and why the step is being performed

c. Expected response of the plantlparameter(s) for the step LO-LP-6031S-0S Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable).

LO-LP-6031S-06 Given the CAUTIONs or NOTEs from AOP 18019-C, explain the reason for specific ones. LO-LP-60316-02 Describe how a CCW heat exchanger plugging or fouling problem may be indicated. Include the following:

a. higher than normal pressure drop b. lower than normal flow rate c. lower than normal temperature rise in the cooling water LO-LP-60316-03 Given conditions and/or indicat i ons , determine the required AOP to enter (including subsections , as applicable).

LO-LP-60316-04 Given the entire AOP, describe:

a. Purpose of selected steps b. How and why the step is performed

c. Expected response of the plantlparameter(s) for the step LO-LP-60316-0S Given the CAUTION(s) or NOTE(s) from AOP 18020-C, explain the reason for specific ones. LO-LP-60317

-01 Describe how the loss of NSCW System affects the operation of the diesel generators. Friday, June 01, 2007 Page 75 of 165 Approval Dlte C c Procedure 110. VogtJe Electric Generating Plant A. 18004-C NUCLEAR OPERATIONS Revision 110. 21.1 Unit COMMON Plge 110. 1 of 77 Abnormal Operating Procedures REACTOR COOLANT SYSTEM LEAKAGE PURPOSE PRB REVIEW REQUIRED Section A specifies the actions to be taken for Reactor Coolant System leakage during Modes 1, 2, and 3 with RCS pressure greater than 1000 psig. Section B specifies the actions to be taken for Reactor Coolant System leakage during Mode 3 with RCS pressure less than 1000 psig, and Mode 4. Section C specifies the actions to be taken for Reactor Coolant System leakage during Mode 5. SYMPTOMS

• Unexplained change in charging flow.
• A rise in VCT makeup frequency.
• Unexplained lowering of PRZR level and pressure.
• PRT temperature, pressure or level rising.
• CNMT moisture alarm or activity rising. No+-c'tv McJ
• CNMT sump level rising. 3 OA-If
• CNMT Air Cooler condensate flow rising alarm. ,') e,,-, I-e
• Transition from 18019-C, LOSS OF RESIDUAL HEAT REMOVAL. S '\

PROCEDURE NO. VEGP REVISION NO. 18004-C PAGE NO. 21.1 51 of 77 C RCS LEAKAGE (MODE 5) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED C1. Check plant conditions: Oa. In Mode 5. b. Both of the following: 0-RCS inventory -IN PRZR INDICATION RANGE -AND-0-SG nozzle dams -NOT INSTALLED OC2. Initiate 91001-C, EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS. C3. Check RCS drainage path isolated: Oa. Check HV-8716A RHR TRAIN A TO HOT LEG CROSSOVER ISO -CLOSED Db. Check HV-8716B RHR TRAIN B TO HOT LEG CROSSOVER ISO -CLOSED Dc. Check 1205-U6-027 or 1205-U4-226 RHR TEST RECIRC ISOL TO RWST -CLOSED a. Go to the appropriate section of this procedure: 0 SECTION A, RCS LEAKAGE (MODE 1, 2, AND 3 WITH RCS PRESSURE > 1000 PSIG) . -OR-O SECTION B, RCS LEAKAGE (MODE 3 <1000 PSIG AND 4) . Db. Go to 18019-C, LOSS OF RESIDUAL HEAT REMOVAL. Oa. Close HV-8716A. Db. Close HV-8716B. Dc. Dispatch operator to close valves. [Located in Aux Bldg D48 (UNIT 1) D22 (UNIT 2) .] c***** ( l I PROCEDURE NO. VEGP 18004-C REVISION NO. PAGE NO. 21.1 10 of 77 B RCS LEAKAGE (MODE 3 <1000 PSIG AND 4) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: RCS depressurization steps should be initiated as necessary to maintain RCS to PRZR liquid differential temperature less than 270°F and meet RCS leak before break criteria. B1. Check plant conditions: D D In Mode 3 with RCS pressure less than 1000 psig. -OR-In Mode 4. B1. Go to the appropriate section of this procedure: D D SECTION A, RCS LEAKAGE (MODE 1, 2, AND 3 WITH RCS PRESSURE > 1000 PSIG) . -OR-SECTION C, RCS LEAKAGE (MODE 5) . DB2. ;1t1 ",ttl e.5 () eflv Sf-e Initiate 91001-C, EMERGENCY (-CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS. DB3. Initiate the Continuous Actions Page. Approval Procedure 110. Vogtie Electric Generating Plant A. 18019-C NUCLEAR OPERATIONS Revis10n 110. Date 25 C Unit COMMON Page No. 1 of 60 Abnormal Operating Procedures LOSS OF RESIDUAL HEAT REMOVAL PURPOSE PRB REVIEW REOUIRED Section A of this procedure specifies actions to be taken for a Q of RHR capability while in:

• Mode 4 or,
• Mode 5 with RCS level within the PRZR indication range and with no SG nozzle dams installed.

A " ,1-, Section B of this procedure specifies actions to be taken for a !Ll c1ll loss of RHR capability or imminent loss of RHR due to RCS leakage " while in Mode 5 or 6 with RCS level below the PRZR indication range l( :1-' 5 SG nozzle dams installed. C of this procedure specifies actions to be taken for a loss of RHR capability while the Rx head is removed, Rx cavity flooded, and transfer canal open. Section D of this procedure specifies actions to be taken for a loss of RHR capability while the RCS is under vacuum pressure conditions (while performing 12009, ReS Vacuum Refill) . SYMPTOMS

• Unexplained change in RHR flow or discharge pressure.
• Any unexplained rise in RCS temperature while RHR is in operation.
• Any observed loss of RHR system capability while RHR is in operation.
• RHR motor amps fluctuating (C40mputer Points J9623 or J9624)
• From 18004-C on imminent loss of RHR cooling due to RCS leakage during reduced inventory operations or during operation with nozzle dams installed.
• SPDS alarm on:
• RHR pump current (core cooling CSFST -Computer Points UD4623 or UD4624)
• Both RHR loops not operating (core cooling or heat sink CSFST -Computer Point UM5626)
• RHR trouble (core cooling CSFST -Computer Point UD0626)

7. 026G2.2.25 001 The Containment Spray System is designed to reduce containment pressure during a c LOCA or a (1) . In addition, the Containment Spray System is designed to retain (2) in water solution.

A"! (1) Main Steamline Break IRC, (2) Iodine B. (1) Main Steamline Break IRC, (2) Cesium C. (1) Main Feedwater Line Rupture IRC, (2) Iodine D. (1) Main Feedwater Line Rupture IRC (2) Cesium ( Page: 13 of 48 6/6/2007 026 Containment Spray. G2.2.25 Knowledge of bases in Technical Specifications for limiting conditions for operations and safety limits. KIA MATCH ANALYSIS Question asks which DBA event will produce the highest peak containment temperature and the purpose of the Containment Spray Test Line. Question meets 10CFR55.43(b) criteria for item # 2 -Facility operating limits in Tech Specs and their bases. ANSWER I DISTRACTOR ANALYSIS A. Correct. DBA event is correct and iodine is the proper isotope to remove. B. Incorrect. Plausible the candidate may think a SLB is the proper event and Cesium is to be removed as it is one of the most common radioisotopes in radwaste. C. Incorrect. Plausible the candidate may think that a FLB produces a high pressure and iodine is the correct isotope. D. Incorrect. Plausible the candidate may think a FLB produces high pressure and Cesium is to be removed as it is one of the most common radioisotopes in radwaste. REFERENCES Technical Specifications 3.6.6 and bases for Containment Spray and Cooling Systems. LO-PP-15101-01-001, Vogtle LO Active Exam Bank. VEGP learning objectives: LO-PP-15101-08, State the LCO, applicability, and the bases of all Containment Spray related technical specifications. Page: 14 of 48 6/6/2007 NO OBI TX OBI LO-PP-14310-0 Describe how generator seal oil temperature and quality are maintained under the following conditions:

a. Normal operation with the Main Seal Oil Pump (MSOP) In service b. Operation with the Emergency Seal Oil Pump (ESOP) in service and the Main Seal Oil Pump (MSOP) shutdown LO-PP-14310-0 For the following conditions describe the system response, effect on generator operation, and any manual actions that must taken. a. Failure of the Main Seal Oil Pump (MSOP) b. Failure of both the Main Seal 011 Pump (MSOP) and the Emergency Seal Oil Pump (ESOP) c. Failure of Differential Pressure control valve PDV 6870 d. Failure of Seal Oil Float Trap Valve LCV 6871 LO-PP-15101-0 List the two main functions of the Containment Spray system and how each function is accomplished.

LO-PP-15101-O Describe what will actuate the Containment Spray System, including coincidence and set point. LO-PP-15101-0 Describe the Containment Spray Systems normal standby alignment. LO-PP-15101-0 List all components that receive a Containment Spray Actuation signal and their change in status. LO-PP-15101-0 Describe how and when the transition from normal flow path to the recirculation flow path is performed. LO-PP-15101-0 State the reason for a minimum required time the Containment Spray system is left on recirculation following a LOCA. LO-PP-15101-0 Describe how the Containment Spray pumps are provided minimum flow protection during test. lO-PP-15101-0 State the lCO, applicability, and the bases of all Containment Spray related technical specifications. lO-PP-16001-0 Indicate the values for the following RCS parameters:

a. Normal operating pressure b. No load Tavg. c. Full load Tavg. d. Reactor thermal power output e. RCP power output f. RCS total flow g. Flow from each RCP lO-PP-16001-0 Define the following terms related to RCS leakage: a. Pressure boundary leakage b. Unidentified leakage c. Identified leakage lO-PP-16001-0 State the safety limit, the applicability, bases, and the action as found in the Safety Limits section of Tech Specs for: a. Reactor Core b. Reactor Coolant System pressure Page 18 of 68 BASES BACKGROUND f} .n () Containment Spray and Cooling Systems f\e\t..-i-rtVCc... B3.6.6 Containment Cooling System (continued)

During normal operation, four fan units are operating. The fans are normally operated at high speed with NSCW supplied to the cooling coils. The Containment Cooling System, operating in conjunction with the Containment Ventilation and Air Conditioning systems, is designed to limit the ambient containment air temperature during normal unit operation to less than the limit specified in LCO 3.6.5, "Containment Air Temperature." This temperature limitation ensures that the containment temperature does not exceed the initial temperature conditions assumed for the DBAs. In post accident operation following an actuation signal, the Containment Cooling System fans are designed to start automatically in slow speed if not already running. If running in high (normal) speed, the fans automatically shift to slow speed. The fans are operated at the lower speed during accident conditions to prevent motor overload from the higher mass atmosphere. The temperature of the NSCW is an important factor in the heat removal capability of the fan units. APPLICABLE SAFETY ANALYSES The Containment Spray System and Containment Cooling System limit the temperature and pressure that could be experienced \.. following a DBA. The limiting DBAs considered are the loss of coolant (" accident (LOCA) and the steam line break (SLB). The LOCA and R r----7-;?* SLB are analyzed using computer codes designed to predict the V\; e.5 0\.1\.'-resultant containment pressure and temperature transients. No DBAs c are assumed to occur simultaneously or consecutively. The C-+-D postulated DBAs are analyzed with regard to containment ESF systems, assuming the loss of one ESF bus, which is the worst case 8 single active failure and results in one train of the Containment Spray fW 8 /'J Of., I-I-J System and Containment Cooling System being rendered inoperable. ev 0( .f 0 0 (1. ') If Ie. The analysis and evaluation show that under the worst case I (,( llO scenario, the highest peak containment pressure is 36.5 psig (experienced during a LOCA). The analysis shows that the peak Vogtle Units 1 and 2 containmenttemperature is 303.1 of (experienced during an SLB). Both results meet the intent of the design basis. (See the Bases for LCO 3.6.4A, "Containment Pressure," and (continued) B 3.6.6-3 Revision No. 0 c BASES APPLICABILITY (continued) ACTIONS Vogtle Units 1 and 2 Containment Spray and Cooling Systems B 3.6.6 In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the Containment Spray System and the Containment Cooling System are not required to be OPERABLE in MODES 5 and 6. With one containment spray train inoperable, the inoperable containment spray train must be restored to OPERABLE status within 72 hours. In this Condition, the remaining OPERABLE spray and It cooling trains are adequate to perform the iodine removal and n N5wl'(L/ containment cooling functions. The 72 hour Completion Time takes (r LL \. ( into account the redundant heat removal capability afforded by the 11 Containment Spray System, reasonable time for repairs, and low probability of a DBA occurring during this period. The 6 day portion of the Completion Time for Required Action A.1 is based upon engineering judgment. It takes into account the low probability of coincident entry into two Conditions in this Specification coupled with the low probability of an accident occurring during this time. Refer to Section 1.3, "Completion Times," for a more detailed discussion of the purpose of the "from discovery of failure to meet the LCO" portion of the Completion Time. I Ce SI 'lAfW\ f lc ,,,-sl'ble S/,,{ e.-!U. tn1 v7 -t f "t IfI\ vf' /5 Q -h(1e t'1IJ With one of the required containment cooling trains inoperable, the 1... n . inoperable required containment cooling train must be restored to / ctX Wt; s f Q.. OPERABLE status within 72 hours. The components in this degraded condition provide iodine removal capabilities and are capable of providing at least 100% of the heat removal needs. The 72 hour Completion Time was developed taking into account the redundant heat removal capabilities afforded by combinations of the Containment Spray System and Containment Cooling System, and the low probability of a DBA occurring during this period. (continued) B 3.6.6-6 Revision No. 0

8. 034G2.1.32 001 Given the following conditions:

c -Unit 2 refueling outage core reload in progress. -Fuel reconstitution is being performed in the Unit 1 New Fuel Elevator. -The FH Machine has just placed a spent fuel assembly in the elevator. -The reconsitution crew wishes to raise the spent fuel assembly in the elevator. Which ONE of the following is CORRECT regarding controls to minimize the possibility of personnel overexposure for this evolution? A. The New Fuel Elevator is interlocked to prevent raising the basket with the weight of a spent fuel asembly. The elevator must be hand cranked up using a handwheel. B. The New Fuel Elevator is interlocked to prevent raising the basket with the Fuel Handling Machine positioned over it. The elevator must be hand cranked up using a handwheel. The New Fuel Elevator is interlocked to prevent raising the basket with the weight of a spent fuel asembly. A bypass interlock key switch on the elevator pendant controller is required to raise the assembly. D. The New Fuel Elevator is interlocked to prevent raising the basket with the Fuel Handling Machine positioned over it. A bypass interlock key switch on the elevator pendant controller is required to raise the assembly. Page: 15 of 48 6/6/2007 034 Fuel Handling Equipment. G2.1.32 Ability to explain and apply all system limits and precautions. KIA MATCH ANALYSIS Question gives a plausible scenario where a spent fuel assembly needs to be raised in the New Fuel Elevator for fuel reconstitution during an outage. The candidate must determine the proper interlock required to raise the assembly. Question meets 1 OCFR55.43(b) criteria item # 7 -Fuel handling facilites and procedures. Question meets 1 OCFR55.43(b) criteria item # 4 -Radiation hazards that may arise during normal and abnormal situation, including maintenance activities and various contamination conditions. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. The interlock portion is correct. There is a manual handwheel for operation of the Fuel Transfer cart, not the elevator. B. Incorrect. Plausible the candidate may confuse the interlocks with some of the transfer system interlocks which require the FH Machine to be moved away to allow operation. There is a manual handwheel for operation of the Fuel Transfer cart, not the elevator. C. Correct. D. Incorrect. Plausible the candidate may confuse the interlocks with some of the transfer interlocks which require the FH Machine to be moved away to allow operation. Key bypass portion is correct. REFERENCES 93210-C, "New Fuel Elevator Operating Instructions" Precaution and Limitations. LO-OR-251 01-09-002 from Vogtle LO Active Exam Bank LO-PP-25101-01-001 from Vogtle LO Active Exam Bank LO-PP-25101 , Fuel Handling System and Refueling Power point slide 82,83, and 84 VEGP learning objectives: ( LO-PP-25101-06, Identify the interlocks and bypasses associated with the following:

d. New Fuel Elevator Page: 16 of 48 6/6/2007

( l NO OBI TX OBI LO-PP-25101-0 State the function of the following:

a. New fuel storage pit b. Spent fuel pool c. Transfer canal/transfer tube d. Refueling cavity e. Fuel Handling Machine Bridge Crane f. New fuel handling tool g. Spent fuel handling tool h. New fuel elevator . Refueling machine j. New RCCA Handling Tool k. RCCA Change Tool I. Thimble Plug Handling Tool LO-PP-25101-0 Describe the operations of the refueling machine controls to include: a. Bridge control b. Trolley control c. Mast control d. Auxiliary monorail hoist control e. Remote cameras and graphic display control LO-PP-25101-0 Describe the operation of the following tools: a. New fuel handling tool b. Spent fuel handling tool c. Control Rod Drive Shaft Unlatching Tool LO-PP-25101-0 Explain the function of the components of the fuel transfer system to include: a. Transfer carriage b. Upenders LO-PP-25101-0 Identify the interlocks and bypasses associated with the following:

a. Spent Fuel Cask crane b. Fuel Handling Machine Bridge Crane c. Refueling machine d. New fuel elevator e. Fuel transfer system LO-PP-25101-0 Describe how the pendant control system is used to manipulate bridge, trolleys, and hoists on the following:

a. Spent Fuel Cask Crane b. Fuel Handling Machine Bridge Crane LO-PP-25101-0 Describe the control functions the fuel transfer system consoles can perform: a. Fuel Handling Building console b. Containment (Rx side) console LO-PP-25101-0 Describe how a new fuel element moves through the fuel handling system from the new fuel receipt area to the reactor core. LO-PP-251 01-1 Explain which operations are covered by the Unloading, Inspection, and Storage of New Fuel Procedure (93010-C).

LO-PP-251 01-1 Describe the radiological precautions that must be followed prior to unloading new fuel. Page 38 of 68 ( Approved By Procedure Number Rev J. B. Beasley, Jr. Vogtle Electric Generating Plant A 93210-C 6 FUEL ELEVATOR OPERATING INSTRUCTIONS Date Approved 8/5/96 Page Number lof6 PRB REVIEW REQUIRED REFERENCE USE PROCEDURE 1.0 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 PURPOSE This procedure provides the operating instructions for the fuel elevators used at Vogtle. The East (Unit 1) elevator is now used exclusively as a fuel reconstitution basket. The West (Unit 2) elevator is used as the new fuel elevator, which is used to wet new fuel. PRECAUTIONS AND LIMITATIONS NOTE The use of the word "tool" in this procedure is understood to mean both the New and Spent Fuel Handling Tools unless otherwise specified. Care shall be exercised for personnel protection during all fuel movement activities. The load monitor or spring scale shall be monitored during all fuel assembly insertion and withdrawal operations to detect possible binding. Prior to lifting a fuel assembly, ensure the latching fingers are in the latched position with the handle in the down position and the locking pin in the operating handle. If the tool cannot be unlatched from a fuel assembly, the fuel handling operation shall be stopped and the Fuel Handling Coordinator notified. The new fuel elevator key bypass switch is for use only in abnormal situations, or during reconstitution, to allow a fuel J1 assembly to be raised from the bottom of the Spent Fuel Pool. l/ Never attempt to raise a fuel assembly in the elevator the permission of both the Unit Shift Supervisor and the Fuel 1\"", (( a", I Handling Coordinator. 1 Complete reliance on limit switches and indicating lights to protect the fuel assemblies and core components during handling is not recommended. Visual observation during the handling of the assemblies could preclude possible damage. Be aware of abnormal equipment conditions. r3 J-0 June 6, 2007 at 1:51 Key Lock Light Push Button Push Button Push I Pull Button OVERLOAD llMITaVPASS HOIST o DOWN Elevator controls, pull for on, push in to turn off. Up and down pushbuttons, Overload limit bypass key switch to allow raising an assembly. It is administratively controlled. 82 ( "-I OVE RLOAD I lIMJT S"tPASS 0

• 0 D Mashing the up or down button will allow the operator to raise or lower an empty basket. I HOIST I 0 Basket will not raise with the weight of an assembly due to an overload trip designed to I DOWN I prevent raising an irradiated assembly from out D ofthe water. Down to lower elevator, up to raise elevator.

Have to bypass with key to raise if assembly in elevator. 83 An overload trip will prevent raising an assembly in the new fuel elevator. However, you can bypass this trip with a key that must be checked out, and administrative levels of supervision must approve use of this key. Light will illuminate when In bypass position. THERE ARE EXTREME CONSEQUENCES FOR PULLING AN IRRADIATED ASSEMBLY OUT OF THE WATER III! Once again to raise the elevator with an assembly in it you have to check out an administratively controlled key. This would bypass the overload trip and allow raising the elevator with the weight of an assembly in it. However, pulling an assembly out of the water shielding would have fatal consequences for the operator and anyone else in the vicinity. We NEVER raise the elevator with assemblies in it other than the dummy for testing the overload trip OR for fuel re-contstitution. During this evolution which is performed by Westinghouse personnel who are essentially rebuilding or replacing broken parts on assemblies, we allow the key to be checked out and the bypass used to raise a spent assembly near the surface where they can work on it with long handled tools. However, we also put in a physical hard stop so that if the elevator raise button were to stick while in the bypass position it wouldn't raise the assembly out of the water with fatal consequences for the operating personnel. 84

1. LO-PP-2SI01-01 001 Page: 1 Which of the following design interlocks reduces the possibility of personnel overexposure from spent fuel while operating with spent fuel in the New Fuel Elevator? The New Fuel Elevator is interlocked to prevent inadvertantly raising the basket with the weight of a spent fuel assembly in it. B. The Fuel Handling Machine is interlocked such that it cannot be positioned over the New Fuel Elevator while a spent fuel assembly is in the Elevator.

C. The New Fuel Elevator is interlocked to prevent raising the basket with with the Fuel Handling Machine positioned directly over it. D. A radiation monitor is located to sense increasing radiation levels in the New Fuel Elevator area and stop upward movement of the Elevator. 6/6/2007

1. LO-OR-25 101-09 002 Page: 1 Which of the following correctly states how the possibility of personnel overexposure from spent fuel is minimized while operating with spent fuel in the New Fuel Elevator.

A'I Administrative controls and guidelines (Le. Fuel Handling Procedure restrictions) and the New Fuel Elevator is interlocked to prevent raising the basket with the weight of a spent fuel assembly in it. B. The Fuel Handling Machine is interlocked such that it cannot be positioned over the New Fuel Elevator with a spent fuel assembly in it. c. The New Fuel Elevator is interlocked to prevent raising the basket with the Fuel Handling Machine positioned directly over it. D. A radiation monitor is located to sense increasing radiation levels in the New Fuel Elevator area and stop upward movement. 6/6/2007

9. 039A2.03 001 The plant has experienced a SGTR with the following radiation monitors in alarm: c: -RE-12839, Condenser Air Ejector and Steam Packing Exhaust Effluent Monitor -RE-13121, Main Steam Line Radiation Monitor for SG # 2 read on the SRDC. A reactor trip and manual SI were actuated and E-O, Reactor Trip or Safety Injection was entered. -Both MDAFW pumps are UNAVAILABLE.

-The TDAFW pump is RUNNING. Prior to transitioning to 19030-C, "Steam Generator Tube Rupture", the BOP operator requests permission to isolate the steam supply to the TDAFW pump. Which ONE of the following actions should the Unit SS take regarding the requested early operator action ? A. Allow isolation of steam to the TDAFW pump only if S/G # 2 wide range level is> 29%. B. Allow the BOP to shut the Trip and Throttle Valve to the TDAFW pump to isolate the steam supply. C. Do Not allow isolation of steam to the TDAFW pump because the ruptured S/G is not positively identified. Do Not allow isolation of steam to the TDAFW pump, this action will be performed after transition to 19030-C. 039 Main and Reheat Steam. A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Indications and alarms for main steam and area radiation. KIA MATCH ANALYSIS Question gives a plausible scenario where a SGTR has taken place on SG # 2. The BOP asks for permission to isolated the steam supply to the TDAFW pump. The candidate must determine the correct operator action. This question meets 1 OCFR55.43(b) criteria for item # 4 -Radiation hazards that may arise during normal, abnormal, and situations, including maintenance activites and various contamination conditions. The SS has to make a decision on whether to Page: 17 of 48 6/6/2007 c perform early isolatation of the radiation release to limit contamination. This question meets 1 OCFR55.43(b) criteria for item # 5 -Assessment of facility conditions and selections of procedures during normal, abnormal, and emergency conditions. Per Vogtle Rules of EOP Usage procedure 10020-C, isolation would not be allowed until a transition to 19030-C has occurred. This implies a selection or transition of procedures must take place before the isolation is allowed. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible the candidate may confuse this with the isolation of AFW I feed flow to the S/G which should not occur until S/G level is > 10% NR. B. Incorrect. Plausible the candidate may choose this as it is method to shut down the TDAFW pump,however, it is an RNO action if the individual steam supply is not able to be closed. Early actions are not allowed per 10020-C, Vogtle's EOP Rules of Usage Procedure. C. Incorrect. Plausible the candidate may not feel there is enough information to identify the ruptured S/G, but with the given indications and the BOP requesting permission to isolate implies he has identified the SG that is ruptured. However, per our EOP Rules of Usage this action would not be allowed early and is the reason it cannot be perfomred. D. Correct. S/G # 2 should NOT be isolated until procedurally directed per 10020-C, "EOP Rules of Usage". REFERENCES 19030-C, "Steam Generator Tube Rupture" steps # 7 and # 8 in particular. 10020-C, "EOP Rules of Usage" steps 3.1.2 and 3.1.3 for early operator actions. V. C. Summer October 2006 NRC SRO Retake Exam question # 40. VEGP learning objectives: Not applicable. Page: 18 of 48 6/6/2007 ( Approved By D. R. Vineyard Date Approved 6/30/06 2.1.2 3.0 NOTE 3.1 3.1.1 3.1.2 Vogtle Electric Generating Plant Procedure Number Rev 10020-C 5 EOP AND AOP RULES OF USAGE Page Number AOP Numbering Example Numbering: 18 033 I I I I I C +-------AOPs are common, or similar for Unit and Unit 2 +---------------- Sequential Number +--------------------- Indicates AOP INSTRUCTIONS FOR PERFORMING EOPs AND AOPs 30f27 Sections 3.1.0 through 3.6.0 apply to both EOPs and AOPs unless specific EOP instructions are given. Section 3.7 applies to EOPs only. GUIDELINES FOR FOLLOWING STEPS Operators shall respond to abnormal and emergency conditions in a methodical manner, assessing the event and utilizing the diagnostics with DISCERNABLE PAUSE to ensure that "undue haste" does not result in misdiagnosis, misoperation, or undesired plant conditions. THIS STEP SHALL BE IMPLEMENTED AS THE "EXCEPTION" NOT "THE RULE" TO NORMAL PROCEDURAL RESPONSE TO ABNORMAL PLANT CONDITIONS AND SHOULD ONLY BE IMPLEMENTED WHEN PROCEDURAL DIRECTION REGARDING RAPID RESPONSE TO EITHER OF THE CONDITIONS BELOW DOES NOT READILY EXIST. Consistent with training and knowledge, operators are expected to take actions that stabilize the plant and mitigate consequences of events AFTER PERFORMING AOP OR EOP IMMEDIATE OPERATOR ACTIONS when the following conditions exist: The setpoint for actuation is imminent OR exceeded and automatic actuation does not occur. OR System or equipment failures require operator intervention for reactor or personnel safety. OPERATORS SHALL NOTIFY THE SS WHEN ACTIONS ARE REQUIRED TO BE TAKEN PER THIS GUIDANCE. June 6, 2007 at 2:10 ( Approved By Procedure Number Rev D. R. Vineyard Vogtle Electnic Generating Plant 10020-C 5 Date Approved 6/30/06 3.1.3 3.1.4 3.1. 5 3.1. 6 Page Number 40f27 EOP AND AOP RULES OF USAGE The following actions have been analyzed and may be taken prior to procedural direction following the completion of immediate operator actions. THESE ACTIONS SHALL ONLY BE IMPLEMENTED AFTER REACTOR TRIP HAS BEEN VERIFIED AND SHIFT SUPERVISOR PERMISSION HAS BEEN OBTAINED. This applied method for verifying the Reactor is tripped and obtaining permission prior to taking these actions will increase crew awareness before actions are taken and remove "undue haste" from the evolution. These actions are limited to vital responses deemed necessary to protect plant personnel as well as the public. This specific list of actions is as follows:

• Isolate Main Steam Lines on a Secondary fault.
• Throttle Aux Feed Water for current plant conditions.
• Isolate Aux Feed Water to a faulted SG.
• Isolate Aux Feed Water on a ruptured SG after required level (adverse or non-adverse containment considered) is achieved.
• Control heatup following blowdown of faulted SG. EOPs (as well as AOPs) are entered based on the Entry Conditions or Symptoms at the beginning of the procedure.

Operators are expected to be knowledgeable of these without referral. Initial entry into the EOPs will be to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION, unless both emergency AC buses are de-energized. The operator may enter 19100-C, ECA-O.O LOSS OF ALL AC POWER, directly based on symptoms. After verifying symptoms or entry conditions upon entering a procedure, the operator should go to Step 1 ACTION/EXPECTED RESPONSE (AER) column unless directed otherwise by the procedure just exited. 0e C I{ lA. 5{ (\J. e A ,'f:'e) Pr1nted June 6, 2007 at 2:06 PROCEDURE NO. VEGP 19030-C REVISION NO. ACTION/EXPECTED RESPONSE o 7 . Check at least one MDAFW Pump -RUNNING AND CAPABLE OF FEEDING SG(s) NEEDED FOR RCS COOLDOWN 35 PAGE NO. 6 of 46 RESPONSE NOT OBTAINED 7. Perform the following: o a. Maintain at least one steam supply to the TDAFW Pump. b. Close ONLY ONE TDAFW Pump Steam Supply Valve as necessary to isolate ruptured SG: OHV-3009, (SG 1) LP-1 MS SPLY TO AUX FW TD PMP-1 -OR-OHV-3019, (SG 2) LP-2 MS SPLY TO AUX FW TD PMP-1 o c. Go to Step 9. 8. Close affected TDAFW Pump Steam Supply Valve(s): o 8. IF at least one MDAFW Pump running, OHV-3009 (SG 1) LP-1 MS SPLY TO AUX FW TD PMP-1 OHV-3019 (SG 2) LP-2 MS SPLY TO AUX FW TD PMP-1 o 9. Verify SG Blowdown Isolation Valves -CLOSED WITH HANDSWITCHES IN CLOSE POSITION THEN trip the TDAFW Pump. o IF TDAFW Pump NOT tripped, THEN locally isolate TDAFW Pump steam supply from ruptured SG(s) . ( 1.039 A2.03 001 The plant has experienced a SGTR with the following annunciators in alarm:

• RM-A9 -CNDSR EXHAUST GAS ATMOS
• RM-G19C -STMLN HI RNG GAMMA A reactor trip and manual SI were actuated and EOP-1.0, Reactor Trip/Safety Injection Actuation, was entered.
• BOTH MDEFW pumps are UNAVAILABLE.
• The TDEFW Pump is RUNNING. Prior to transitioning to EOP-4.0, Steam Generator Tube Rupture, the BOP operator requests permission to isolate steam to the TDEFP. Which ONE (1) of the following actions should the Control Room Supervisor take? A. Do Not allow isolation of steam to the TDEFP because the ruptured S/G is not positively identified.

B. Do Not allow isolation of steam to the TDEFP from either SG because the TDEFP is required to maintain secondary heat sink. C. Allow isolation of steam from 'B' and 'C' S/Gs to the TDEFP only if S/G narrow range level is > 30%. Allow isolation of steam from 'C' S/G only if the supply from 'B' S/G to the TDEFP is open. Student must identify "C" SG as ruptured SG and apply OAP-103.4 EOP deviation (6.14.b.2) which allows isolation of steam from a ruptured SG to the TDEFP to minimize off-site releases if the steam supply from the non-ruptured SG is open. A is incorrect because although the condenser exhaust does not identify the ruptured SG, the MSL monitor identifies C as the correct SG. B is incorrect because an exception is granted per the OAP. C is incorrect; misapplication of ruptured Sg isolation. NR level >30% to throttle EFW flow, not steam supply from EFW. Vc Page: 1 6/6/2007

10. 040G2.2.22 001 Which ONE of the following describes the MOST restrictive condition assumed to ensure that the minimum shutdown reactivity of accident analysis is met during a ( guillotine break of a main steam line inside containment?

A. At the beginning of core life, with Tavg at full load operating temperature. B. At the beginning of core life, with Tavg at no load operating temperature. C. At the end of core life, with Tavg at full load operating temperature. At the end of core life, with Tavg at no load operating temperature. 040 Steam Line Break -Excessive Heat Transfer G2.2.22 Knowledge of limiting conditions for operations and safety limits. KIA MATCH ANALYSIS Question asks the basis for the minimum shutdown reactivity accident analysis for Shutdown Margin during a main steam line break IRC. Candidate must choose the correct bases. Question meets 10CFR55.43(b) criteria item # 2 -Facility operating limits in Tech Specs and their bases. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. EOl, no load Tavg is bases. B. Incorrect. EOl, no load Tavg is bases. C. Incorrect. EOl, no load Tavg is bases. D. Correct. Per bases of Tech Spec REFERENCES Technical Specifications and Bases for 3.1.1, "Shutdown Margin". Farley December 2003 NRC SRO Exam question # 8. VEGP learning objectives: lO-lP-39205-02, 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 3.1 are exceeded. Page: 19 of 48 6/6/2007 , Number Text LO-LP-39204-08 State what actions are required if a component is rendered inoperable soley due to the inoperability of a support system. LO-LP-39205-01 For any given item in section 3.1 of Tech Specs, be able to: a. State the limiting condition for operation (LCO), and b.O State anyone hour or less required actions. LO-LP-39205-02 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 LCO's of section 3.1 are exceeded.

b. The required actions for all section 3.1. LCO's LO-LP-39205-03 For any given item in section 13.1 of the TRM, be able to: a. State the Technical Requirement (TR) for operation.

b. State anyone hour or less required actions. LO-LP-39205-04 Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any TR of section 13.1 has been exceeded.

b. The required actions for all section 13.1 TRs. LO-LP-39205-05 State the reason for having a shutdown margin in all modes. LO-LP-39205-06 State the reason for limitations on MTC. LO-LP-39205-07 State the reasons for maintaining rods above the RIL. LO-LP-39205-08 Describe the bases for any given Tech Spec in section 3.1. LO-LP-39206-01 For any item in section 3.2 of Tech Specs, be able to: a.O State the LCO. b.O State anyone hour or less required actions. LO-LP-39206-02 Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a.O Whether any Tech Spec LCOs of section 3.2 are exceeded.

b.O The required actions for all section 3.2 LCOs. LO-LP-39206-03 State the relationship between Relaxed Axial Offset Control and AFD. LO-LP-39206-04 Describe the bases for any given Tech Spec in section 3.2. LO-LP-39206-05 Define AFD target band. State when it is used. LO-LP-39206-06 State the action required for being outside the band at various power levels. LO-LP-39206-07 Define heat flux hot channel factor. State how the core height correction factor curve is used. Friday, June 01, 2007 Page 41 of 165 ( BASES (continued) APPLICABLE SAFETY ANALYSES Vogtle Units 1 and 2 The minimum required SDM is assumed as an initial condition in safety analyses. The safety analysis (Ref. 2) establishes a SDM that ensures specified acceptable fuel design limits are not exceeded for normal operation and AOOs, with the assumption of the highest worth rod stuck out on scram. The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are not exceeded. This is done by ensuring that: a. b. c. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events; The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature limits for AOOs, and < 200 cal/gm average fuel pellet enthalpy at the hot spot for the rod ejection accident); and The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. The most limiting accident for the MODES 1 and 2 SDM requirements is a guillotine break of a main steam line inside containment initiated at the end of core life with RCS average temperature at no-load operating temperature, as described in the accident analYSis (Ref. 2). The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. As RCS temperature decreases, the severity of an MSLB decreases until the MODE 5 value is reached. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs, is a guillotine break of a main steam line inside containment initiated at the end of core life at no-load operating temperature. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus (continued) B 3.1.1-2 Revision No. 0

1. 040AKl.05 001 Which ONE of the following describes the most restrictive conditions assumed to ensure that the minimum shutdown reactivity of accident analysis is met during a C guillotine break of a main steam line inside containment?

A. At the begining of core life, with T avg at full load operating temperature. B. At the end of core life, with T avg at full load operating temperature. C. At the begining of core life, with T avg at no load operating temperature. At the end of core life, with T avg at no load operating temperature. Source: Slightly modified from a Farley Bank Question #052302E01 005. A. Incorrect, the conditions listed in Basis of TIS are EOl, No load Tavg. B. Incorrect, the conditions listed in Basis of TIS are EOl, No load Tavg. C. Incorrect, the conditions listed in Basis of TIS are EOl, No load Tavg. D. Correct. These are the conditions listed in the basis of TIS. 5AO Page: 1 6/6/2007

11. 056AA2.24 001 Given the following conditions:

-The crew is in 18031-C due to an L05P on 2BA03 with DG2B tying to the bus. -Various CCW Train Train B low flow and pressure alarms annunciate, then clear. -The crew notes 3 CCW train B pump red lights illuminated on the QMCB. Which ONE of the following is CORRECT regarding CCW Train B and the actions the 55 should take? A. CCW pump locked rotor has occurred. Monitor pump amps on the QEAB to determine which pump to stop. Enter LCO 3.7.7 for CCW. B. CCW pump locked rotor has occurred. Monitor pump amps on the QEAB to determine which pump to stop. Enter INFO LCO 3.7.7 for CCW. C. CCW pump shaft shear has occurred. Monitor pump amps locally at 2BA03 to determine which pump to stop. Enter LCO 3.7.7 for CCW. D!' CCW pump shaft shear has occurred. Monitor pump amps locally at 2BA03 to determine which pump to stop. Enter INFO LCO 3.7.7 for CCW. Page: 20 of 48 6/6/2007 ( ( 056 Loss of Off-site Power AA2.24 Ability to determine and interpret the following as they apply to the Loss of Offsite Power. CCW pump ammeter, flowmeter and run indicator KIA MATCH ANALYSIS Question gives a plausible scenario during an LOSP on a class 1 E electrical bus which starts the DG and the bus is re-energized. Three CCW pumps will be running due to low pressure. The candidate must determine which pump to stop and if an LCO entry is required. Question meets 10CFR55.43(b) criteria for item # 2 -Facility operating limits in Tech Specs and their bases. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Locked rotor would result in a breaker trip. Pump amps not available on the QEAB. Plausible the candidate may think amps available on QEAB or an LCO entry is required. B. Incorrect. Locked rotor would result in a breaker trip. Pump amps not available on the QEAB. Plausible the candidate may think amps available on QEAB and know an INFO LCO entry is required. C. Incorrect. Plausible the candidate may know pump amps monitored at swgr and think an LCO entry is required. D. Correct. REFERENCES Tech Spec 3.7.7 for CCW and the Bases. AOP-18031-C, "Loss of Class 1 E Electrical Systems" section B for Loss of Power With DG Tying to Bus. VEGP learning objectives: LO-LP-39211-02, 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 3.7 are exceeded. LO-LO-39211-04, Describe the bases for any given Tech Spec in section 3.7 Page: 21 of 48 6/6/2007 Number Text LO-LP-39211-02 Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, eqiupment availability, and operational mode: a.O Whether any Tech Spec LeOs of section 3.7 are exceeded. b.O The required actions for ali section 3.7 LeOs. LO-LP-39211-03 For any given item in section 13.7 of the TAM, be able to: a.O State the TA for operation. b.D State anyone hour or less required actions. LO-LP-39211-04 Describe the bases for any given Tech Spec in section 3.7. LO-LP-39211-05 List the five relief setpoints of the steam generator safety valves. LO-LP-39211-06 Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: a.D Whether any TR of section 13.7 has been exceeded. b.O The required actions for all section 13.7 TRs. LO-LP-39212-01 For any given item in section 3.8 of Tech Specs, be able to: a. State the LCO. b. State anyone hour or less required actions. 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 LeOs. LO-LP-39212-03 For any item in section 13.8 of the TAM, be able to; a. State the TA for operation.

b. State any 1 hour or less required actions. LO-LP-39212-04 Describe the bases for any given Tech Spec in section 3.8. LO-LP-39212-05 Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any TR of section 13.8 has been exceeded.

b. The required actions for all section 13.8 TAs. LO-LP-39213-01 For any given item in section 3.9 of Tech Specs, be able to: a. State the LeO. b. State anyone 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 LeOs of section 3.9 are exceeded.

b. The required actions for all section 3.9 LeOs. Friday, June 01, 2007 Page 44 of 165

( ( PROCEDURE NO. VEGP 18031-C REVISION NO. PAGE NO. 22 13 of 26 B LOSS OF POWER WITH DG TYING TO BUS ACTION/EXPECTED RESPONSE B1. Perform the following for the affected bus: Oa. Verify bus frequency -AT 60 HZ Db. Verify bus voltage -AT 4160V AC o B2. Check charging pumps -ONLY ONE OPERATING OB3. Check CCW pumps on affected train -TWO RUNNING RESPONSE NOT OBTAINED B2. Perform the following: o a. Start or stop pumps as necessary to establish one running charging pump. b. IF charging pump can NOT be started, THEN perform the following:

01) Isolate normal letdown. 02) Initiate 18007-C, CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION.

B3. Perform the following: Start or stop pumps as necessary to establish two pumps running in the affected train. IF two pumps can NOT be started, THEN initiate 18020-C, LOSS OF COMPONENT COOLING WATER. ( BASES CCW System B 3.7.7 LCO (continued) A CCW train is considered OPERABLE when: a. fj--tV SWt' !\..l Two pumps and associated surge tank are OPERABLE; and The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety related function are OPERABLE. {Zv. [e {( DLl The isolation of CCW from other components or systems not (}V\.1 required for safety may render those components or systems inoperable but does not necessarily make the CCW System ilL C inoperable. Consideration should be given to the size of the load IT f) V{ [0 (, (.-I( 0, {( isolated and the impact it will have on the rest of the CCW system 1\ \...; vt ., C./ before determining OPERABILITY. St',f'Ce, Wc,,\ f) 3h cy -{-(LtJ (J h()",J '3 1 APPLICABILITY ACTIONS Vogtle Units 1 and 2 In MODES 1, 2, 3, and 4, the CCW System is a normally operating system, which must be prepared to perform its post accident safety functions, primarily RCS heat removal, which is achieved by cooling the RHR heat exchanger. In Modes 5 or 6, there are no TS OPERABILITY requirements for the CCW System. However, the functional requirements of the CCW System are determined by the systems it supports. Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -MODE 4," be entered if an inoperable CCW train results in an inoperable RHR loop. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this Condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function. The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period. oSC Ibt-t< 2.-t ( continued) B 3.7.7-3 Rev. 1-8/05 \ IVc-t-,t."./ LccJ ( 12. 063A2.02001 If the 125V DC class 1 E battery room fans were lost while the batteries were charging, the primary concern and actions you should take would be: A. Battery room temperatures exceeding Tech Spec limits, perform 13405, "125V DC Electrical Distribution System" to stop the battery charge and close and lock the room doors to prevent personnel entry. BlI" Explosive Hydrogen gas could accumulate in the battery rooms, prop doors open per 0031 O-C, "Standards for Use of Doors" and establish portable ventillation per 13302, "Control Building ESF Ventillation Systems". C. Smoke purge function may be lost in the event of a battery room fire, place the battery rooms in the smoke purge mode per 13302, "Control Building ESF Ventillation Systems". D. Toxic fume buildup could affect affect battery room habitability I personnel safety, prop doors open per 00310-C, "Standards for Use of Doors" and establish portable ventillation per 13302, "Control Building ESF Ventillation Systems". 063 DC Electrical Distribution. A2.02 Ability to (a) predict the impacts of the following malfuncitons or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations. Loss of ventilation during battery charging. KIA MATCH ANALYSIS Question gives a plausible scenario where the 1 E battery room fans are lost while the batteries are charging. The candidate must determine the primary concern and the correct actions to mitigate. Question meets 1 OCFR5S.43(b) criteria item # 5 -assessment of facility conditions and selection of procedures during normal, abnormal, and emergency situations. Question is also SRO only by KA Catalog # for SRO giving an importance factor rating of 3.1 while the RO importance factor rating is only 2.3. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausbile the candidate may recall Tech Spec limits on room temperature and consider stopping charging of the batteries. The room doors should be propped open, not shut to prevent explosive H2 concentration. B. Correct. Explosive Hydrogen gas is the main concern and 13302 gives direction to establish portable ventillation and to prop open room doors per 0031 O-C. Page: 22 of 49 6/6/2007 c ( ( C. Incorrect. Battery room ventillation fans do not perform a smoke purge function but plausible the candidate may consider the use of Smoke Purge Mode to remove hydrogen or fumes, smoke purge would not be used until after a fire is out to prevent feeding 02 to the fire. D. Incorrect. Plausible the candidate may consider toxic fumes to be a personnel hazard and actions same as that for "B" above but not the primary concern. REFERENCES SOP-13302-1/2, CB ESF Ventillation Systems, Precautions 2.1.1 and 2.1.2 13404-1/2, "125V DC Electrical Distribution Systems" Precaution 2.1.3. 00310-C, "Standard For Use of Doors". Vogtle October 2005 SRO Audit Exam question # 21 modified to meet 10CFR55.43(b). VEGP learning objectives: Not applicable. Page: 23 of 49 6/6/2007 ( c. Approved By draft Date Approved draft 1.0 2.0 2.1 2.1.1 2.1.2 2.1.3 2.1.4 2.2 2.2.1 2.2.2 2.2.3 2.2.4 2.2.5 2.2.6 2.2.7 Vogtle Electric Generating Plant A 125V DC A TRAIN 1 E ELECTRICAL DISTRIBUTION SYSTEM PURPOSE Procedure Number Rev 13405A-1 1.0 Page Number 30f40 This procedure provides instructions for energizing, operating and de-energizing the Unit 1 125V DC 1E Electrical Distribution System. PRECAUTIONS AND LIMITATIONS Precautions Batteries produce hydrogen. Smoking, using open flames, or operating space heaters is prohibited in the vicinity of the Batteries. Battery Room Ventilation Systems should be in operation to limit the buildup of hydrogen in the Battery Rooms. (I 0 q If Battery Room Ventilation System is not available, doors to Battery Room should be propped open per 00310-C, "Standard for Use of Doors." {("'\h 5 In MODE 5 or 6 when a battery must be removed from service for long periods such as for testing, it is preferable to transfer the 120V vital busses to their regulated AC power supply, provided the inverter is not required per Technical Specification 3.8.8. C. J.. Transferring the vital busses to the regulated source will reduce the potential for power . \I losses due to 125V DC switchgear instabilities. (Tech Spec 3.8.7 and 3.8.8) Limitations The 125V DC 1 E Electrical Busses shall be OPERABLE in MODEs, 1, 2, 3 and 4 per Technical Specification LCO 3.8.4 and LCO 3.8.9. The 125V DC 1 E Electrical Busses shall be OPERABLE in MODEs 5 and 6 per Technical Specification LCO 3.8.5 and LCO 3.8.10. The DC Input Breaker to the 25kVA Inverters can not be closed until the Inverter Internal Capacitor Bank has been charged. The Battery Charger 480V AC input voltage shall be 480V AC +/-10% (432V-528V). If the electrical switchgear must be energized by the battery chargers alone (without battery breaker closed in), only one charger should be energized to supply the bus. A 72-hour equalizing charge should be performed every six months on all 1 E batteries. Removal of 125V DC equipment from service may necessitate performance of 14235-1, "On Site Power Distribution Operability Verification," to document that sufficient equipment (batteries and chargers) remain OPERABLE to satisfy Technical SpeCifications LCO 3.8.10. D Pnnted June 6, 2007 at 2:36 Approved By S. E. Prewitt Date Approved c 1-16-2006 2.0 2.1 2.1.1 2.1.2 2.1.3 2.1.4 Procedure Number Rev 13302-1 12.1 Vogtle Electric Generating Plant A Page Number 30f23 CONTROL BUILDING ESF VENTILATION SYSTEMS PRECAUTIONS AND LIMITATIONS PRECAUTIONS The Control Building Safety Features Battery Room Exhaust Fans should be running any time the batteries are energized to prevent the accumulation of Hydrogen in the Battery Rooms. If power is lost to Safety Features Battery Room Exhaust Fans and cannot be ." restored in a timely manner, the Maintenance Department should be directed to t' .J provide portable ventilation within 48 hours to prevent hY Q rogen buil 9 up. -t-p/\v(J o(li'N\ ,%-t-(Oc-k 0.105 tl:>{ f e. u Each ventilation system sho'uld De operated as necessary to maintain the room A temperature of the areas being served below the temperature alarm setpoints. Unless Emergency conditions exist, the Chemistry Foreman should be notified prior to initiating Smoke Purge Mode in the Control Building. This is necessary to determine the presence of any airborne activity. Based on the sample results, a Batch Release Permit may be required. If an emergency exists, the Chemistry Foreman shall be notified as soon as possible.

13. 067G2.1.33 001 Given the following:

c -The Unit is at 100% Rated Thermal Power (RTP). -SIP "A" inoperable, Tech Spec LCO 3.5.2 for ECCS -Operating in Effect -A fire occurs at DG1 B -At 0400 LCO 3.8.1, A. C. Sources Operating is entered due to DG1 B inoperable. Which ONE of the following is REQUIRED in accordance with Tech Specs? A. Immediately take actions of LCO 3.0.3 (Motherhood) B. No other LCO entry is required, LCO 3.8.1 addresses this condition. By 0800 take actions of LCO 3.5.2 for both trains of ECCS inoperable. D. Immediately take actions of LCO 3.5.2 for both trains of ECCS inoperable. ( Page: 24 of 49 6/6/2007 ( 067 Plant Fire on Site G2.1.33 Ability to recognize indications for system operating parameters which are entry level conditions for Technical Specifications. KIA MATCH ANALYSIS Question gives a plausible scenario where a fire at a DG has resulted in inoperable Question meets 10CFR55.43(b) criteria for item # 2 -Facility operating limits in Tech Specs and their bases. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible candidate may think motherhood is immediately required to be entered. B. Incorrect. Plausible candidate may think no other LCO required. SIP "A" has to be declared inoperable at 0800. C. Correct. At 0800 SIP "B" would have to be declared inoperable. D. Incorrect. Plausible candidate may think actions required for LCO 3.5.2 but there is 4 hours to try to restore per 3.8.1. REFERENCES Technical Specification 3.8.1 for A. C. Sources -Operating and the bases. Technical Specification 3.5.2 for ECCS -Operating VEGP learning objectives: LO-LP-39209-02, 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 3.5 are exceeded. 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. Page: 25 of 49 6/6/2007 c Number Text LO-LP-39208-07 Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: a.O Whether any TR of section 13.4 has been exceeded. b.O The required actions for all section 13.4 TRs. LO-LP-39208-08 List the DNB parameters and the reason for these limits. LO-LP-39208-09 State the reason for the limits on AFD and DNB parameters. LO-LP-39209-01 For any given item in section 3.5 of Tech Specs, be able to: a.O State the LeO. b.O State anyone hour or less required actions. LO-LP-39209-02 Given a set of the Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a.O Whether any Tech Spec LeOs of section 3.5 are exceeded. b.O The required actions for all section 3.5 LeOs. LO-LP-39209-03 Describe the bases for any given Tech Spec in section 3.5. LO-LP-39209-04 For any given item in section 13.5 of the TRM, be able to: o a.O State the TR for operation b.O State anyone hour or less actions. LO-LP-39209-05 Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: o a.O Whether any TR of section 13.5 has been exceeded. o b.O The required actions for all section 13.5 TRs. LO-LP-3921 0-01 For any given item in section 3.6 of Tech Specs, be able to: a.O State the LeO. b.O State anyone hour or less required actions. LO-LP-3921 0-02 Given a set of Tech Specs and the bases, determine for a sepcific set of plant conditions, equipment aVailability, and operational mode: a.D Whether any Tech Spec LeOs of section 3.6 are exceeded. b.O The required actions for all section 3.6 LeOs. LO-LP-3921 0-03 Describe the bases for any given Tech Spec in section 3.6. LO-LP-39211-01 For any given item in section 3.7 of Tech Specs, be able to: a.O State the LeO. b.O State anyone hour or less required actions. Friday, June 01, 2007 Page 43 of 165 Number Text LO-LP-39211-02 Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, eqiupment availability, and operational mode: a.O Whether any Tech Spec LeOs of section 3.7 are exceeded. b.O The required actions for all section 3.7 LeOs. LO-LP-39211-03 For any given item in section 13.7 of the TRM, be able to: a.O State the TR for operation. b.O State anyone hour or less required actions. LO-LP-39211-04 Describe the bases for any given Tech Spec in section 3.7. LO-LP-39211-05 List the five relief setpoints of the steam generator safety valves. LO-LP-39211-06 Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: a.D Whether any TR of section 13.7 has been exceeded. b.O The required actions for all section 13.7 TRs. LO-LP-39212-01 For any given item in section 3.8 of Tech Specs, be able to: a. State the LCO. b. State anyone hour or less required actions. 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 LeOs of section 3.8 are exceeded.

b. The required actions for all section 3.8 LeOs. LO-LP-39212-03 For any item in section 13.8 of the TRM, be able to; a. State the TR for operation.

b. State any 1 hour or less required actions. LO-LP-39212-04 Describe the bases for any given Tech Spec in section 3.8. LO-LP-39212-05 Given the TRM , determine for a specific set of plant conditions, eqUipment availability, and operational mode: a. Whether any TR of section 13.8 has been exceeded.

b. The required actions for all section 13.8 TRs. LO-LP-39213-01 For any given item in section 3.9 of Tech Specs, be able to: a. State the LeO. b. State anyone 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 LeOs of section 3.9 are exceeded.

b. The required actions for all section 3.9 LeOs. Friday, June 01, 2007 Page 44 of 165

( L AC Sources -Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources -Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1 E AC Electrical Power Distribution System; and b. Two diesel generators (DGs) capable of supplying the onsite Class 1 E power distribution subsystem(s).

Automatic load sequencers for Train A and Train B ESF buses shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS --------------------------------------------------------------NOTE---------------------------

---

LCO 3.0.4b is not applicable to DGs. CONDITION A. One required offsite circuit inoperable. Vogtle Units 1 and 2 A.1 REQUIRED ACTION Perform SR 3.8.1.1 for required OPERABLE offsite circuit. 3.8.1-1 COMPLETION TIME 1 hour Once per 8 hours thereafter (continued) Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2) ACTIONS CONDITION A. (continued) Vogtle Units 1 and 2 A.2 AND A.3 REQUIRED ACTION Declare required feature( s) with no offsite power available inoperable when its redundant required feature(s) is inoperable. Restore required offsite circuit to OPERABLE status. AC Sources -Operating

3.8.1 COMPLETION

TIME 24 hours from discovery of no offsite power to one train concurrent with inoperability of redundant required feature(s) 72 hours 14 days from discovery of failure to meet LCO (continued) 3.8.1-2 Amendment No.1 00 (Unit 1) Amendment No. 78 Unit 2) ( ACTIONS (continued) CONDITION REQUIRED ACTION B. One DG inoperable. B.1 Perform SR 3.8.1.1 for the required offsite circuit(s). AND B.2 Verify SAT available. AND fti\l5wetL B.3 Declare required feature(s) supported by the inoperable DG inoperable when its required redundant feature(s) is inoperable. vt {es 0'-1 t-AND B.4.1 Determine OPERABLE DG fJ *-D is not inoperable due to common cause failure. OR f-B.4.2 Perform SR 3.8.1.2 for OPERABLE DG. f{ AND AC Sources -Operating

3.8.1 COMPLETION

TIME 1 hour AND Once per 8 hours thereafter 1 hour AND Once per 12 hours thereafter 4 hours from discovery of Condition B concurrent with inoperability of redundant required feature(s) 24 hours 24 hours (continued) 7 6-2,,( (\ 33 Vogtle Units 1 and 2 3.8.1-3 Amendment No.1 00 (Unit 1) Amendment No. 78 Unit 2) c 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS -Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ECCS -Operating 3.5.2 --------------------------------------N OTE---------------------------------------- In MODE 3, either residual heat removal pump to cold legs injection flow path may be isolated by closing the isolation valve to perform pressure isolation valve testing per SR 3.4.14.1. ACTIONS CONDITION A. One or more trains inoperable. AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available. B. Required Action and associated Completion Time not met. Vogtle Units 1 and 2 REQUIRED ACTION A.1 Restore train(s) to OPERABLE status. '{vile) ovt c-(-B.1 Be in MODE 3. AND B.2 Be in MODE 4. 3.5.2-1 COMPLETION TIME 72 hours A-+D(rv v+ 6 hours 12 hours Amendment No. 136 (Unit 1) Amendment No. 115 (Unit 2)

14. 078G2.1.32 001 The following annunciators I indications are present in the Unit 1 Control Room: -ALB01 window B05 for "SERVICE AIR CMPSR TROUBLE".

-ALB01 window C06 for "SERVICE AIR HDR LO PRESS". -Instrument air pressure is slowly decreasing as reasd on QMCB meter PI-9361. -All available compressors are running and the Turbine Building Operator (TBO) reports all compressors are loading and unloading properly. -The TBO reports and air dryer is malfunctioning. Which ONE of the following is CORRECT actions for the Unit SS to take I direct? A. Direct the TBO to iimplement SOP-1371 0-1, "Instrument Air System", if an instrument air dryer malfunction, bypass the instrument air dryer by opening the air dryer bypass, then slowly close the air dryer outlet, then close the air dryer inlet. B. Enter AOP-18028-C, "Loss of Instrument Air", if a service air dryer malfunction, isolate control air to the service air dryer by manually closing the petcock valve located on the air regulator at the dryer inlet. C. Direct the TBO to implement SOP-13711-1, "Service Air System", if a service air dryer malfunction, bypass the service air dryer by bleeding off control air to the dryer by depressing the Sullicon controller pushbutton at the dryer inlet. Enter AOP-18028-C, "Loss of Instrument Air", if an instrument air dryer malfunction, place the instrument air dryer in the two chamber full flow mode by depressing the pushbutton at the front end of the dryer. Page: 26 of 49 6/6/2007 078 Instrument Air System. G2.1.32 Ability to explain and apply all system limits and precautions. KIA MATCH ANALYSIS Question gives a plausible scenario with either an instrument or service air dryer malfunction. The candidate must pick the appropriate procedure and actions to mitigate the event. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible, isolate. Action is a correct action for a service air dryer. Instrument air dryers do not have a bypass. 13710-1/2 is the correct procedure for normal instrument air evolutions. B. Incorrect. Plausible, this is the old method for placing a service air dryer in 2 flow full chamber mode but no longer directed per procedure. Note in procedure says NOT to isolate control air to a dryer, an incorrect action. AOP-1B02B-C, is the proper procedure to use. C. Incorrect. Procedure directs placing in 2 chamber full flow mode by depressing button at front of dryer, not by bleeding off control air. This is method for causing a rotary air compressor to fully load. 13711-1/2 is the correct procedure for normal service air evolutions. D. Correct. Placing air dryer in two chamber full flow mode by depressing the pushbutton at the front of the dryer is the proper response. AOP-1B02B-C is the proper procedure to use. REFERENCES NOTE: This is a re-use question from Vogtle May 2006 NRC SRO Exam question # 15 with KA # (07BA2.01). This is the only question re-used from the last 2 previous SRO exams given at Plant Vogtle. This question will also fit KA # 07BG2.1.32 as it tests system limits and precautions for instrument air dryers. AOP-1B02B-C, "Loss of Instrument Air" SOP-13710-1/2, "Instrument Air System". SOP-13711-1/2, "Service Air System". VEGP learning objectives: LO-LP-60321-06 Describe the operator actions required during normal full power operation when instrument air header pressure falls to < BO psig or < 70 psig. Page: 27 of 49 6/6/2007 c , , Number Text LO-LP-S0321-02 State the fail position of the following valves on loss of instrument air: a. extraction steam non-return valves b. feedwater heater high level dump valves c. HV-182 (charging flow control valve) d. containment instrument air header isolation valves e. MSIVs 1. SGFP mini-flow valves g. FRV h. FRV bypass . RHR heat exchanger outlet valve j. RHR heat exchanger bypass valve k. CVCS letdown isolation valve I. Containment isolation valves m. CVI valves n. HV-128 (RHR to Letdown valve) o. FV-121 (Charging Flow Control valve) LO-LP-S0321-03 Describe why a loss of instrument air precludes plant operation. LO-LP-S0321-04 Describe why a plant cooldown from mode 3 to mode 4 should be delayed if a loss of instrument air occurs. LO-LP-S0321-05 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 S. LO-LP-S0321-0S 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-LP-S0321-08 Describe the effects on RCS pressure due to a loss of instrument air while solid on RHR. LO-LP-S0321-09 Describe the effect on Fuel Transfer System gate seals on a Loss Of Service Air and what operator action(s} is required. LO-LP-S0321-10 Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). LO-LP-S0321-11 Given the entire AOP, describe:

a. Purpose of selected steps b. How and why the step is being performed

c. Expected response of the plantlparameter(s}

for the step LO-LP-S0322-02 Describe why you are cautioned in AOP 18030-C not to align the Spent Fuel Pool System to provide bleed to the RWST if spent fuel pool level is less than 217 ft elevation. LO-LP-S0322-03 Describe the source of MlU to SFP to makeup for leakage and why this source is used. LO-LP-S0322-04 Given conditions and/or indications,determine the required AOP to enter (including subsections, as applicable). LO-LP-S0322-05 Given the entire AOP, describe:

a. Purpose of selected steps b. How and why the step is being performed

c. Expected response of the plantlparameter(s}

for the step LO-LP-S0323-01 State the immediate operator action required on loss of 1 E Electrical Systems, AOP-18031. Include RNO and substeps of the immediate action. LO-LP-S0323-02 Describe why the affected train diesel generator must be tripped following a loss of one train of 1 E Electrical Systems. Friday, June 01, 2007 Page 77 of 165 o7C6 G-{ 57." c 1. 078A2.01 002 The following annunciators I indications are present in the Unit 1 Control Room:

• ALB01-B05, "SERVICE AIR CMPSR TROUBLE"
• ALB01-C06, "SERVICE AIR HDR LO PRESS"
• Instrument air pressure is slowly decreasing as read on QMCB PI-9361.
• All available compressors are running and the Turbine Building Operator (TBO) reports all compressors are loading and unloading properly.
• The Turbine Building Operator (TBO) reports an air dryer is malfunctioning.

Which ONE of the following are CORRECT actions for the Unit SS to take I direct? A. Direct the TBO to implement SOP-13710-1, "Instrument Air System", if an instrument air dryer malfunction, bypass the instrument air dryer by opening the air dryer bypass, then slowly close the air dryer outlet, then close the air dryer inlet. B. Enter AOP-18028-C, "Loss of Instrument Air, if a service air dryer malfunction, isolate control air to the service air dryer by manually closing the petcock valve located on the air regulator at the dryer inlet. C. Direct the TBO to implement SOP-13711-1, "Service Air System", if a service air dryer malfunction, bypass the service air dryer by bleeding off control air to the dryer by depressing the Sullicon controller pushbutton at the dryer inlet. Enter AOP-18028-C, "Loss of Instrument Air", if an instrument air dryer malfunction, place the instrument air dryer in the two chamber full flow mode by depressing the pushbutton at the front of the dryer. Page: 1 of 2 6/6/2007 ( ( 078 Instrument Air System. A2.01. Ability to (a) predict the impacts of the following malfunctions or operations on the lAS; and based (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: KIA MATCH ANALYSIS Question gives a plausible scenario with either an instrument or service air dryer malfunction. Candidate must pick appropriate procedure and actions to mitigate event. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible, isolate. Action is a correct action for a service air dryer. Instrument air dryer does not have a is the correct procedure for normal instruement air evolutions. B. Incorrect. Plausible, this is the old method for placing a service air dryer in 2 flow full chamber mode but no longer directed per procedure. Note in procedure says to NOT isolate control air to a dryer, an incorrect action. AOP-18028 is the proper procedure to use. C. Incorrect. Procedure directs placing in 2 chamber full flow mode by depressing button at front of dryer, not by bleeding off control air. This is method for causing ---a rotary air compressor to fully load. 13711-1/2 is the correct procedure for normal service air evolutions. D. Correct. Placing air dryer in two chamber full flow mode by depressing the pushbutton at the front of the dryer is the proper response. AOP 18028-1/2 is the proper procedure to use. REFERENCES AOP-18028-C, Loss of Instrument Air ARP-17001-C, windows B05 and C06 SOP-13710-1/2, "Instrument Air System" VC J ;Vl"Ly VOp tirzc 5R--O SOP-13711-1/2, "Service Air System" VEGP learning objectives: LO-LP-60321-06 Describe the operator actions required during normal full power operation when instrument air header pressure falls to < 80 psig and or < 70 psig. cLC-t1 l 1-Page: 2 of2 -Fuv-0 7 u v V / ,'.J 6/6/2007 ( ( PROCEDURE NO. VEGP 18028-C REVISION NO. ACTION/EXPECTED RESPONSE INITIAL ACTIONS 1. Check proper operation of all available air compressors on affected unit: Da. All air compressors -RUNNING Db. All air compressors -PROPERLY LOADING AND UNLOADING 24 PAGE NO. 3 of 26 RESPONSE NOT OBTAINED Da. b. Start all available air compressors on affected unit. Dispatch operator to fully load any air compressor not loading properly:

01) 2) Reciprocating

-close service air compressor filter inlet valves: UNIT 1 (TB-A-T11)

• 1-2401-U4-627
• A-2401-U4-629 UNIT 2 (TB-A-T10)
• 2 -2401-U4-627 Rotary Da) Isolate instrument air to controllers PY-19315A and/or PY-19314A using local air isolation valves. Verify air pressure is bled off by depressing water drain pushbutton bill. f-on Sullicon at a1r 1nlet dampers. } ,

( PROCEDURE NO. VEGP 18028-C REVISION NO. ACTION/EXPECTED RESPONSE PAGE NO. 24 4 of 26 RESPONSE NOT OBTAINED CAUTION: Do NOT isolate control air to instrument or service air dryers as this will cause dryers to blow down continuously.

02. Verify proper operation of Instrument Air dryers. 02. IF an Instrument Air Dryer is malfunctioning, If THEN place in two chamber N 5W-e IV full flow mode by p,;!shing in local pushbutton

03. Verify proper operation of Service Air dryers. /\ tfe-c----e fife u o7C6

04. Initiate the Continuous Actions Page. Dryer A HS-0746 Dryer B HS-0747 3. IF a Service Air Dryer is malfunctioning, o THEN perform one of the following:

Bypass Service Air dryer: Oa. Open 2401-U4-551 Service Air Dryer Bypass . Db. Slowly close 2401-U4-554 Service Air Dryer Outlet . Dc. Close 2401-U4 -548 Service Air Dryer Inlet. -OR-Place service air dryer in two chamber full flow mode by depressing local pushbutton switch HS-0745. ( Approved By Procedure Number Rev C. H. Williams, Ir Vogtle Electric Generating Plant A 13710-1 34 Date Approved 12-31-2005

4.4.9 CAUTIONS

NOTES SERVICE AIR SYSTEM Placing The Service Air Dryer In The Two Chamber Full Flow Mode Of Operation Page Number 520f66

• Control Air to a Service Air Dryer shall NOT be isolated, as this will cause the dryer to blow down continuously.
• Time spent in Two Chamber Full Flow Mode should be minimized as this disables the dryer's moisture removal capability.
• Due to the configuration and operational mode of the inlet-switching valve, the vessel in regeneration mode will slowly pressurize and is an expected result in placing the Service Air Dryer in Two Chamber Full Flow Mode.
• Placement of the Service Air Dryers in full flow will remove the power to the MEC controller.

When this power is lost, the common trouble alarm (low air pressure/dryer trouble) on the compressor control panel will annunciate. There is no second hit on this alarm and it will stay in until the dryer is placed back in service. ,11/50 is preferable to bypass the Service Air Dryer, removing the Air Dryer from service and operating the Service Air (-System with the dryer bypassed. ()V\. '"'\ 4.4.9.1 c r \.( 4.4.9.2 To place Service Air Dryer in the Two Chamber Full Flow Mode of operation, perform the following:

a. b. c. d. Obtain SS permission to place Service Air Dryer in Two Chamber Full Flow Mode of Operation, De-energize controller by pushing in 1-HS-0745 (located next to the Service Air Dryer skid), Request the Control Room to enter the time and reason why the Service Air Dryer was re-aligned in the Unit Control Log, If operation in this mode is required for more than two hours, notify the System Engineer.

To remove Service Air Dryer from Two Chamber Full Flow Mode and restore to normal operation, perform the following:

a. b. Verify corrective action for the air dryer malfunction or abnormal indications, as applicable, has been completed, Obtain permission from the SS to restore the service air dryer to normal operation, Printed June 6, 2007 at 3:03 [ ] [ ] [ ] [ ] [ ] [ ]

( ( Approved By Procedure Number Rev C. H. Williams, Jr Vogtle Electric Generating Plant A 13711-1 13.1 Date Approved 1-16-2006 Page Number 11 of 19 INSTRUMENT AIR SYSTEM INITIALS 4.2.2 Placing The Instrument Air Dryer(s) In the Two Chamber Full Flow Mode Of Operation 4.2.2.1 CAUTIONS

• Permission to place the Instrument Air Dryers in the Two Chamber Full Flow Mode of operation must be obtained from the Shift Supervisor (SS) or his designee.
• Do not isolate control air to an Instrument Air Dryer, as this will cause the dryer to blow down continuously.
• Since Two Chamber Full Flow Mode disables the dryer's moisture removal capability, time spent in this mode should be minimized.

Notify the system engineer if operation in this mode is required for more than two hours. If only one air dryer is affected, it is preferable to remove the affected dryer from service and operate with the unaffected air dryer.

• If an air dryer is placed in Two Chamber Full Flow Mode due to a malfunction or abnormal indications, verify the condition is corrected prior to restoring the dryer to normal operation.

To place Instrument Air Dryer A in the Two Chamber Full Flow Mode of operation, perform the following:

a. Obtain permission from the SS to place Instrument Air Dryer "A" in the Two Chamber Full Flow Mode of operation, b. De-energize Instrument Air Dryer "A" controller by pushing in 1-HS-0746 located next to the Instrument Air Dryer "A" skid. c. Direct the Control Room enter the time and reason why the Instrument Air Dryer "A" was re-aligned in the Unit Control Log. Pnnted June 6, 2007 at 3:02

15. 079A2.01 001 Given the following conditions:

-Unit 1 at 100% RTP and Unit 2 at 55% RTP. -All Unit 1 Air Compressors are available. -The "Swing' Air compressor has been aligned to Unit 2 and the air headers crosstied. -No Unit 2 air compressors are available. -Unit 2 air pressure is 73 psig and slowly lowering, service air is isolated. -Unit 1 air pressure is 78 psig and slowly lowering for both instrument and service air. -The Unit 1 SERVICE AIR LO PRESS annunciator is illuminated. -AOP-18028-C, "Loss of Instrument Air" section A for Mode 1 is in effect. Which ONE of the following actions is CORRECT actions that should be performed? Art Per AOP-18028-C, isolate I separate the unit air headers and continue with actions of 18028-C, section A for Mode 1. B. Per AOP-18028-C", verify Unit 1 service air isolates, maintain the unit air headers crosstied and continue with actions of 18028-C, section A for Mode 1. C. Per AOP-18028-C, isolate I separate the unit air headers, trip the Unit 2 Main Turbine and implement 18011-C, Turbine Trip Below P-9, continue with 18028-C. D. Per AOP-18028-C, verify Unit 1 service air isolates, maintain the air headers crosstied, trip Unit 2 enter and E-O, perform AOP-18028-C section B for Mode 3. Page: 28 of 49 6/6/2007 c c 079 Station Air. A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the SAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Cross connection with lAS. KIA MATCH ANALYSIS Question gives a plausible scenario where one unit has lost all available air compressors, the candidate has to determine if crosstie of headers is allowed and at what point to isolate the headers and trip the affected unit. Question meets 1 OCFR55.43(b) criteria item # 5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency situations. ANSWER I DISTRACTOR ANALYSIS A. Correct. Air headers should be separated, Unit 2 trip criteria of 70 psig not met yet. B. Incorrect. Plausible the candidate may think verification of Unit 1 service air isolation should be verified to stabilize pressure. AOP directs isolation of air headers if Unit 1 pressure drops below 80 psig. C. Incorrect. Plausible the candidate may feel a Unit 2 Turbine trip is imminent and entry into 18012 appropriate. Air headers isolated part is correct. D. Incorrect. Plausible the candidate may feel reactor trip appropriate but Unit trip criteria of 70 psig not reached yet. Also, air headers should have been separated at 80 psig. REFERENCES 18028-C, "Loss of Instrument Air" section A for Mode 1 operations. VEGP learning objectives: LO-LP-60321-06 Describe the operator actions required during normal full power operation when instrument air header pressure falls to < 80 psig or < 70 psig. Page: 29 of 49 6/6/2007 ( Number Text LO-LP-60321-02 State the fail position of the following valves on loss of instrument air: a. extraction steam non-retum valves b. feedwater heater high level dump valves c. HV-182 (charging flow control valve) d. containment instrument air header isolation valves e. MSIVs f. SGFP mini-flow valves g. FRV h. FRV bypass . RHR heat exchanger outlet valve j. RHR heat exchanger bypass valve k. CVCS letdown isolation valve I. Containment isolation valves m. CVI valves n. HV-128 (RHR to Letdown valve) o. FV-121 (Charging Flow Control valve) LO-LP-60321-03 Describe why a loss of instrument air precludes plant operation. LO-LP-60321-04 Describe why a plant cool down from mode 3 to mode 4 should be delayed if a loss of instrument air occurs. 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-60321-06 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-LP-60321-08 Describe the effects on RCS pressure due to a loss of instrument air while solid on RHR. LO-LP-60321-09 Describe the effect on Fuel Transfer System gate seals on a Loss Of Service Air and what operator action(s) is required. LO-LP-60321-10 Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). 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 plantlparameter(s) for the step LO-LP-60322-02 Describe why you are cautioned in AOP 18030-C not to align the Spent Fuel Pool System to provide bleed to the RWST if spent fuel pool level is less than 217 ft elevation.

LO-LP-60322-03 Describe the source of M/U to SFP to makeup for leakage and why this source is used. LO-LP-60322-04 Given conditions and/or indications,determine the required AOP to enter (including subsections, as applicable). LO-LP-60322-05 Given the entire AOP, describe:

a. Purpose of selected steps b. How and why the step is being performed

c. Expected response of the plantlparameter(s) for the step LO-LP-60323-01 State the immediate operator action required on loss of 1 E Electrical Systems, AOP-18031.

Include RNO and substeps of the immediate action. LO-LP-60323-02 Describe why the affected train diesel generator must be tripped following a loss of one train of 1 E Electrical Systems. Friday, June 01, 2007 Page 77 of 165 071 A-2.,0 I C I PROCEDURE 110. VEGP 18028-C REVISION NO. ACTION/EXPECTED RESPONSE 016. Check UNAFFECTED unit Instrument Air pressure -GREATER THAN 80 PSIG ffN5wefL, CTifU 24 ;7 rz VI. (.e) <f B '( -+ ({D li 017. Identify source of leakage and isolate if possible. 018. Check affected Unit Mode -MODES 1 OR 2 PAGE NO. 9 of 26 RESPONSE NOT OBTAINED 16. Restore/isolate UNAFFECTED unit Instrument Air as follows: a. Perform one of the following: 0 IF Unit 1 is selected for the swing compressor, THEN close 2-2401-U4-510. -OR-O IF Unit 2 is selected for the swing compressor, THEN close 1-2401-U4-510. Db. Verify swing compressor is running (TB-A-TC11) . 18. Perform one of the following: D I F in Mode 3, THEN go to ATTACHMENT ATTACHMENT A" LOSS OF INSTRUMENT AIR IN MODE 3. -OR-o IF in Modes 4, 5, or 6, THEN go to ATTACHMENT ATTACHMENT B" LOSS OF INSTRUMENT AIR IN MODES 4, 5, OR 6. ( PROCEDURE NO. VEGP 18028-C REVISION NO. ACTION/EXPECTED RESPONSE 0* 19. Check Instrument Air header pressure -REMAINS GREATER THAN 70 PSIG 7 0 \ {# 020. Check header pressure -STABLE OR RISING Id e fIIHI\./ I eta (( (}r1Vc/ 7 f--? f-od 021. Check Instrument Air header pressure on PI-9361 -GREATER THAN 100 PSIG 022. Check PV-9375 Service Air System Trip Valve -OPEN PAGE NO. 24 10 of 26 RESPONSE NOT OBTAINED

• 19. Perform the following:

o a . Trip the reactor. Db. Initiate 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION. Dc. Go to ATTACHMENT ATTACHMENT A" LOSS OF INSTRUMENT AIR IN MODE 3. 20. IF leakage source can NOT be isolated, THEN restore/isolate UNAFFECTED unit Instrument Air as follows: a. Perform one of the following: 0 IF Unit 1 is selected for the swing compressor, THEN close 2-2401-U4-510. -OR-O IF Unit 2 is selected for the swing compressor, THEN close 1-2401-U4-510. Db. Verify swing compressor is running (TB-A-TC11) . 021. Go to Step 25. 022. IF PV-9375 is NOT required to isolate leak, THEN reset and open PV-9375 by initiating 13710, SERVICE AIR SYSTEM.

c. ( 16. }03A2.ol OO} The following conditions exist on Unit 2 while at 100% power. -The Shift Manager receives word that the containment air lock has failed the leakage rate surveillance test due to excessive air lock leakage. -The Containment overall leak rate is now being exceeded.

Which ONE of the following is the CORRECT required Tech Spec action(s) ? A. Apply Tech Spec 3.6.1, "Containment", restore Containment to operable status within 7 days or be in Mode 3 in the following 6 hours and in Mode 5 in 36 hours. B. Apply Tech Spec 3.6.2, "Containment Air Locks", within 1 hour close and lock at least 1 Air Lock door. Restore leakage within limits within 7 days or apply Tech Spec 3.6.1 "Containment" actions. C,-. Apply Tech Spec 3.6.1, "Containment", restore Containment to operable status within 1 hour or be in Mode 3 in the following 6 hours and in Mode 5 in 36 hours. D. Apply Tech Spec 3.6.2, "Containment Air Locks", immediately close and lock both Air Lock doors. Restore leakage within limits within 7 days or apply Tech Spec 3.6.1 "Containment" actions. 103 Containment. A2.01 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. Integrated Leak Rate Test KIA MATCH ANALYSIS Question gives a plausible scenario where Containment Air Lock doors cause Containment overall leak rate to exceed leakage limits. Note 3 of Tech Spec 3.6.2 states if Air Lock Leakage causes Containment Leakage to exceed limits, apply Tech Spec 3.6.1, "Containment". Candidate must determine the correct Tech Spec actions. Question meets 10CFR55.43(b) criteria item # 2 -Facility Operating limits in Tech Specs and their bases. Question is also SRO due to KA # for SRO has an importance factor of 2.6 where the KA # for RO only has an importance factor of 2.0. ANSWER I DISTRACTOR ANALYSIS Page: 30 of 49 6/6/2007 c A. Incorrect. Plausible since Condition A actions of 3.6.2 Air Locks, says to perform actions of 3.6.1 for Containment. Plausible could be a 7 day action. B. Incorrect. Plausible since it is Condition B actions of 3.6.2 Air Locks, but is for interlock mechanisms and implies the plant can run in this condition for 7 days before applying LCO 3.6.1. C. Correct. 3.6.1 for Containment must be applied and 1 hour to bring within limits or shutdown to Mode 3 in 6 hours and be in Mode 5 in 36 hours. D. Incorrect. Plausible since it is close to Condition C actions of 3.6.2 Air Locks but implies Tech Spec 3.6.1 does not have to be applied for 7 days. REFERENCES Tech Spec 3.6.1, "Containment" and bases. Tech Spec 3.6.2 "Containment Air Locks" and bases. VEGP learning objectives: LO-LP-3921 0-02, 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 3.6 are exceeded.

b. The required actions for all sections 3.6 LCOs. Page: 31 of 49 6/6/2007

( c Number Text LO-LP-39208-07 Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: a.D Whether any TR of section 13.4 has been exceeded. b.O The required actions for all section 13.4 TRs. LO-LP-39208-08 List the DNB parameters and the reason for these limits. LO-LP-39208-09 State the reason for the limits on AFD and DNB parameters. LO-LP-39209-01 For any given item in section 3.5 of Tech Specs, be able to: a.O State the LeO. b.O State anyone hour or less required actions. LO-LP-39209-02 Given a set of the Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a.O Whether any Tech Spec LeOs of section 3.5 are exceeded. b.O The required actions for all section 3.5 LeOs. LO-LP-39209-03 Describe the bases for any given Tech Spec in section 3.5. LO-LP-39209-04 For any given item in section 13.5 of the TRM, be able to: o a.o State the TR for operation b.o State anyone hour or less actions. LO-LP-39209-05 Given the TRM, determine for a specific set of plant conditions, equipment availability, and operational mode: o a.o Whether any TR of section 13.5 has been exceeded. o b.o The required actions for all section 13.5 TRs. LO-LP-3921 0-01 For any given item in section 3.6 of Tech Specs, be able to: a.O State the LeO. b.Q State anyone hour or less required actions. LO-LP-3921 0-02 Given a set of Tech Specs and the bases, determine for a sepcific set of plant conditions, equipment availability, and operational mode: a.D Whether any Tech Spec LeOs of section 3.6 are exceeded. b.O The required actions for all section 3.6 LeOs. LO-LP-3921 0-03 Describe the bases for any given Tech Spec in section 3.6. LO-LP-39211-01 For any given item in section 3.7 of Tech Specs, be able to: a.O State the LeO. b.O State anyone hour or less required actions. Friday, June 01, 2007 Page 43 of 165 c 3.6 CONTAINMENT SYSTEMS 3.6.1 Containment LCO 3.6.1 Containment shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. Containment inoperable. A.1 Restore containment to OPERABLE status. B. Required Action and B.1 Be in MODE 3. associated Completion Time not met. AND B.2 Be in MODE 5. Vogtle Units 1 and 2 3.6.1-1 Containment

3.6.1 COMPLETION

TIME 1 hour iffV' r,. '-6 hours 36 hours Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) SURVEILLANCE REQUIREMENTS SR 3.6.1.1 SR 3.6.1.2 SURVEILLANCE Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program. Verify containment structural integrity in accordance with the Containment Tendon Surveillance Program. Containment

3.6.1 FREQUENCY

In accordance with the Containment Leakage Rate Testing Program In accordance with the Containment Tendon Surveillance Program Vogtle Units 1 and 2 3.6.1-2 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) c 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 Containment Air Locks 3.6.2 ------------------------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 A. One or more containment air locks with one containment air lock door inoperable.

Vogtle Units 1 and 2 REQUIRED ACTION ---------------NOTES--------------------

1. Required Actions A.1, A.2, and A.3 are not applicable if both 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.

COMPLETION TIME ff--[ BtD ( continued) I l 3.6.2-1 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ( ACTIONS CONDITION A. (continued) Vogtle Units 1 and 2 A.1 AND A.2 AND REQUIRED ACTION Verify the OPERABLE door is closed in the affected air lock. Lock the OPERABLE door closed in the affected air lock. Containment Air Locks 3.6.2 COMPLETION TIME 1 hour 24 hours A.3 ---------NOTE------------ Air lock doors in high radiation areas may be verified locked closed by administrative means. Verify the OPERABLE door is locked closed in the affected air lock. Once per 31 days (continued) 3.6.2-2 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ACTIONS (continued) CONDITION B. One or more containment air locks with containment air lock interlock mechanism inoperable. Vogtle Units 1 and 2 REQUIRED ACTION ---------NOTES--------

1. Required Actions B.1, B.2, and B.3 are not applicable if both doors in the same air lock are inoperable and Condition C is entered. 2. Entry and exit of containment are permissible under the control of a dedicated individual.

Containment Air Locks 3.6.2 COMPLETION TIME B.1 Verify an OPERABLE 1 hour p {C1 " 7/\ b (t (((]ll AND B.2 AND door is closed in the affected air lock. Lock an OPERABLE door 24 hours closed in the affected air lock. O;t....-(f 0 l ( B.3 --------NOTE--------- Air lock doors in high radiation areas may be verified locked closed by administrative means. Verify an OPERABLE door is locked closed in the affected air lock. Once per 31 days (continued) 3.6.2-3 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) c ACTIONS (continued) CONDITION C. One or more C.1 containment air locks inoperable for reasons other than Condition A or B. AND C.2 AND C.3 D. Required Action and D.1 associated Completion Time not met. AND D.2 Vogtle Units 1 and 2 REQUIRED ACTION Containment Air Locks 3.6.2 COMPLETION TIME Initiate action to evaluate Immediately overall containment leakage rate per LCO 3.6.1. Verify a door is closed in the affected air lock. Restore air lock to OPERABLE status. Be in MODE 3. Be in MODE 5. 3.6.2-4 1 1 hour .MVike) 24 hours I?JJ-o Id"£*)1\ Vv v 6 hours 36 hours Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) c ( Containment Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SR 3.6.2.1 SR 3.6.2.2 SURVEILLANCE

---

N()TES----------------------------

1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. 2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1. Perform required air lock leakage rate testing in accordance with the Containment Leakage Rage Testing Program. FREQUENCY In accordance with the Containment Leakage Rate Testing Program Verify only one door in the air lock can be opened 18 months at a time. Vogtle Units 1 and 2 3.6.2-5 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) c c. 17. G2.1.5 001 Which ONE of the following would be CORRECT regarding the stated condition and action regarding shift crew minimum staffing requirements?

A. An SO has called in just prior to shift turnover and will be about an hour late due to a flat tire. This would put staffing 1 below the minimum. It will not be necessary to hold anyone over since he will arrive at the plant within the next hour. B. An SO has unexpectedly failed an FFD test and is not allowed back on site. This puts staffing 1 below the minimum. Shift crew minimum could be reduced by 1 person for a time period not to exceed 1 shift. 2 licensed control room operators have become extremely sick due to something they ate and have to leave work. This puts staffing 1 below the minimum. Shift crew minimum could be reduced by 1 person for a time period not to exceed 2 hours. D. An RO has called in sick just prior to shift turnover. Another RO has been called in and can arrive within 2 hours. This puts staffing 1 below minimum until he arrives. It would not be necessary to hold over the day shift RO. Page: 32 of 49 6/6/2007 G2.1.S Ability to locate and use procedures and directives related to shift staffing ( and activities. l KIA MATCH ANALYSIS Questions gives several plausible scenarios the candidate may encounter as an SRO regarding minimum shift staffing and the candidate has to determine the correct action. Question meets 10CFR55.43(b) criteria item # 2 -Facility operating limits in Tech Specs and their bases. There is no bases for the Shift Mannin Admin requirements. Question also is SRO due to KA importance factor for this topic is only 2.3 for the RO level and 3.4 for the SRO level. Therefore, it is an SRO only topic. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. You can only go below minimum staffing for an exepected absence of duty. In this instance, there would be time to hold shift personnel over to maintain the minimum staffing. Plausible the candidate could think it would be OK to allow the personnel to leave as long as the off duty person arrives within 2 hours or not think of minimun requirements pertaining to non-licensed personnel. B. Incorrect. Miminum for unexpected on duty person is limited to 2 hours. Would need a replacement within 2 hours. Plausible the candidate could think there is one shift to replace the persons since a lot of other admin requirements are within one shift. C. Correct. Could go below minimum due to unexpected absense but would be required to replace the personnel within 2 hours. D. Incorrect. There is a 2 hour time limit for unexpected absence of on duty personnel. Would have to hold someone over. Plausible the candidate would think it is OK to allow the person to leave since a replacement should arrive within 2 hours. REFERENCES 10003-C, "Manning the Shift", REQUIREMENT

1. 3.8 G2.1.5 from Vogtle October 2005 SRO Audit Exam VEGP learning objectives:

Not applicable. Page: 33 of 49 6/6/2007 ( ( Approved By Procedw-e Number Rev C. H. Williams, Jr. Vogtle Electric Generating Plant A 10003-C 24 Date Approved 111112004 Page Number 40f6 MANNING THE SHIFT 3.8 The shift crew minimum requirements may be reduced by one pers Ojl R JE 7 tJII\ for a period of time not to exceed 2 hours. This is to i Olll the unexpected absence of on-duty shift personnel. ff Immediate action shall be taken to restore manning to the minimum n ( c: requirements. This provision does not permit any shift to be unmanned upon shift change due to tardiness or absence of on-coming personnel. 1) ( 1-IC ( '" 1. I r 1" f,:, 0 \ 0 \. I 3.9 During any absence of the USS from the Control Room while either unit is in Mode 1, 2, 3 or 4, an individual with a valid SRO 3.10 3.11 3.12 4.0 4.1 4.2 4.2.1 4.2.2 license shall be designated to assume the Unit's Control Room operating responsibilities. During any absence of the USS from the Control Room while either unit is in Mode 5 or 6, an individual with a valid SRO or RO license shall be designated to assume the Unit's Control Room operating responsibilities. The balance of Plant (BOP) operator will normally remain in the Control Room when not needed elsewhere in the plant. The SS shall designate a qualified person to perform the Shift Technical Advisor (STA) function at the beginning of each shift. The Shift Technical Advisor (STA) provides engineering expertise during operational emergencies to assess plant status and assist in implementing EOPs. The STA may be a dual role position if the Shift Superintendent (SS), USS, or SSS holds a bachelors degree in engineering or a related science. This dual-roled position will not normally be the Emergency Plan Communicator or the Fire Brigade Captain. If an STA is assigned on shift, he or she will report to the SS. The SS should use the guidance provided in Procedure 00012-C, "Shift Manning Requirements", to ensure that the shift is adequately staffed to meet applicable regulatory requirements. REFERENCES Vogtle Technical Specifications Section 5.2.2 and TRM 15.1 PROCEDURBS 10000-C, "Conduct Of Operations" 00715-C, "Licensed Operator Requalification Program" END OF PROCEDURE TEXT Printed June 6, 2007 at 3:49 r 18. G2.1.34 001 Given the following conditions: -Unit 1 has just tripped from 100% power and is at no-load temperature and pressure. -Chemistry has sampled the primary and secondary plants. Which ONE of the following CORRECTLY states the Tech Spec limits for RCS Dose Equivalent 1-131 and Secondary Specific Activity? Res Dose Equivalent 1-131 A. 0.15 micro curies per gram B. 1.5 micro curies per gram C. 0.10 micro curies per gram 1.0 micro curies per gram Page: 34 of 49 Secondary Specific Activity 1.5 micro curies per gram 0.15 micro curies per gram 1.0 micro curies per gram 0.10 micro curies per gram 6/6/2007 c l G2.1.34 Ability to maintain primary and secondary plant chemistry within allowable limits. KIA MATCH ANALYSIS Question gives a plausible scenario . Question is SRO and meets 1 OCFR55.43(b) criteria item # 2 -Facility operating limits in Tech Specs and their bases. Question is also SRO due to the KA Catalog importance factor for this KA # is only a 2.3 for RO and is 3.4 for SRO. Therefore, it is an SRO level question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible the candidate may confuse the limit with the values for Chlorides or Fluorides from the TRM and invert the numbers. B. Incorrect. Plausible the candidate may confuse the limit with the values for Chlorides or Fluorides from the TRM. C. Incorrect. Plausible the candidate may invert the correct values for primary and secondary 1-131. D. Correct. REFERENCES TRM 13.4.1 Reactor Coolant System -Chemistry. HL-AW-39000-00-016 Vogtle SRO Audit question from HL-13. VEGP learning objectives: LO-LP-64101-03, State the LCO applicability and any action statement required within one hour for the following Technical Specifications:

a. T.R.M. 13.4.1 Chemistry

b. T.S. 3.4.16 Specific Activity Page: 35 of 49 6/6/2007 c , Number Text LO-LP-63930-05 State the posting requirements for each of the following areas: a. D Restricted Area b. D Radiation Control Area (RCA) c. D Radiation Area d. D High Radiation Area e. D Locked High Radiation Area f. D Hot Spot g. D Contaminated Area h. D Airbome Radioactivity Area i. D Radioactive Materials Area j. D Hot Particle Control Area k. D Hot Particle Buffer Area I.D Very High Radiation Area LO-LP-63930-06 State the entry requirements applicable to each of the following:

a. 0 Restricted Area b. D Radiation Control Area (RCA) c. D Radiation Area d. D High Radiation Area e. D Locked High Radiation Area f. D Hot Spot g. D Contaminated Area h. D Airbome Radioactivity Area i. D Radioactivity Materials Area j. D Hot Particle Control Area k. D Hot Particle Buffer Area I.D Very High Radiation Area LO-LP-63930-0B Given a requirement for entry into an area containing radiation or contamination, Btate the RWP requirement for that area. LO-LP-63930-09 Describe the routing required for RWP issuance.

LO-LP-63930-10 State the conditions under which a general and/or specific RWP may be terminated. LO-LP-63930-11 State the frisking requirements necessary when leaving the RCA. LO-LP-641 01-01 Describe major sample flowpaths from RCS and pressurizer through the NSSL. LO-LP-641 01-02 With regards to the Primary Sampling System, state the Primary Sampling sample points. LO-LP-64101-03 State the LCO applicability and any action statement required within one hour for the following Technical Specifications:

a. T.R.M 13.4.1 chemistry

b. T.S. 3.4.16 specific activity LO-LP-641 02-01 List eight sample points into the TPSS for which continuous monitoring is provided.

LO-LP-661 00-01 State methods used to deliver clear and concise messages. LO-LP-661 00-02 State the fundamentals of group communication and how these fundamentals are applied in the control room setting. Friday, June 01, 2007 Page 102 of 165 ( 15. lll..-A W-39000-00 016 c iven the following conditions:

• Unit 1 has just tripped from 100% power and is at no-load temperature and pressure.
• Chemistry has sampled the primary and secondary plants. Which ONE of the following correctly states the Tech Spec limits for RCS Dose Equivalent 1-131 and Secondary Specific Activity?

RCS Dose Equivalent 1-131 A. .15 micro curies per gram B. 1.5 micro curies per gram C. .10 micro curies per gram 1.0 micro curies per gram Secondary Specific Activity 1.5 micro curies per gram J_ P { 'h I L \' q ... '7 ( lL Til Q&--v1iA 5 15' . -, _ . micro cunes per gram ' V\. 'H 1.0 micro curies per gram :h le.L 'd."'f 1 .10 micro curies per gram 7 J7 '" (v ...... "r .( c c 13.4 Reactor Coolant System (RCS) TR 13.4.1 Chemistry RCS Chemistry TR 13.4.1 TR 13.4.1 RCS chemistry shall be maintained within the limits specified in Table 13.4.1-1. APPLICABILITY: At all times, except for dissolved oxygen when T avg 250 o f. ACTIONS CONDITION A. One or more chemistry parameters> steady-state limit and transient limit in MODES 1, 2, 3, or 4. B. One or more chemistry parameters> transient limit in MODES 1, 2, 3, or 4. OR Required Action and associated Completion Time of Condition A not met. Vogtle Units 1 and 2 Technical Requirement REQUIRED ACTION A.1 Restore parameter to within steady-state limit. B.1 Be in Mode 3. AND B.2 Be in Mode 5. 13.4 -1 COMPLETION TIME 24 hours 6 hours 36 hours ( continued) Rev.O 12/26/96 ( ACTIONS (continued) CONDITION REQUIRED ACTION C. --------------NOTE------------ C.1 Initiate action to reduce the All Required Actions must pressurizer pressure to :s; be completed whenever 500 psig. this Condition is entered. ---------------------------------- AND Chloride or fluoride C.2 Perform an engineering concentration > steady-evaluation to determine the state limit for> 24 hours effects of the out-of-limit in any condition other than condition on the structural MODES 1, 2, 3, or 4. integrity of the RCS. OR Chloride or fluoride concentration> transient AND limit in any condition other than MODES 1, 2, 3, or 4. C.3 Determine that the RCS remains acceptable for continued operation. TECHNICAL REQUIREMENT SURVEILLANCES SURVEILLANCE TRS 13.4.1.1 ---------------------------NOTE------------------------------ Not required to be performed for dissolved oxygen when Tavg:S; 250 of. Verify RCS chemistry within limits specified on Table 3.4.1-1. Vogtle Units 1 and 2 Technical Requirement 13.4 -2 RCS Chemistry TR 13.4.1 COMPLETION TIME Immediately Prior to increasing pressurizer pressure > 500 psig. Prior to entering MODE 4. Prior to increasing pressurizer pressure > 500 psig. OR Prior to entering MODE 4. FREQUENCY 72 hours Rev.O 12/26/96 c c .. Table 13.4.1-1 RCS Chemistry Limits PARAMETER STEADY-STATE LIMIT Dissolved Oxygen (a) 0.10 ppm Chloride 0.15 ppm Fluoride 0.15 ppm (a) Limits not applicable when T a 250 of. Vogtle Units 1 and 2 Technical Requirement 13.4 -3 RCS Chemistry TR 13.4.1 TRANSIENT LIMIT 1.00 ppm 1.50 ppm /" Rev.O 12/26/96 ( 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity RCS Specific Activity 3.4.16 LCO 3.4.16 The specific activity of the reactor coolant shall be within limits. APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 500°F. ACTIONS LCO 3.0.4c is applicable. CONDITION A. DOSE EQUIVALENT A.1 1-131 > 1.0 IlCi/gm. 41\15vv( IL/ AND A.2 (( D l( B. Gross specific activity of B.1 the reactor coolant not within limit. AND B.2 Vogtle Units 1 and 2 NOTE---, ------------ REQUIRED ACTION COMPLETION TIME Verify DOSE Once per 4 hours EQUIVALENT 1-131 within the acceptable region of Figure 3.4.16-1. Restore DOSE EQUIVALENT 1-131 to within limit. Perform SR 3.4.16.2. Be in MODE 3 with T avg < 500°F. 3.4.16-1 48 hours 4 hours 6 hours (continued) Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2) c ACTIONS (continued) CONDITION C. Required Action and C.1 associated Completion Time of Condition A not met. OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1. SURVEILLANCE REQUIREMENTS REQUIRED ACTION Be in MODE 3 with T avg < 500°F. SURVEILLANCE SR 3.4.16.1 SR 3.4.16.2 Verify reactor coolant gross specific activity S 1 OOIE NOTE Only required to be performed in MODE 1. RCS Specific Activity 3.4.16 COMPLETION TIME 6 hours FREQUENCY 7 days Verify reactor coolant DOSE EQUIVALENT 1-131 14 days specific activity S 1.0 JlCi/gm. Vogtle Units 1 and 2 3.4.16-2 Between 2 and 6 hours after a THERMAL POWER change of 15% RTP within a 1 hour period (continued) Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) c* SR 3.4.16.3 ----------NOTE---------- Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours. ------_._-------------- RCS Specific Activity 3.4.16 FREQUENCY Determine E from a sample taken in MODE 1 184 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours. Vogtle Units 1 and 2 3.4.16-3 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) c l 3.7 PLANT SYSTEMS 3.7.16 Secondary Specific Activity Secondary Specific Activity 3.7.16 LCO 3.7.16 The specific activity of the secondary coolant shall be 0.10 IlCi/gm DOSE EQUIVALENT 1-131. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION A. Specific activity not within limit. A.1 AND A.2 REQUIRED ACTION Be in MODE 3. Be in MODE 5. SURVEILLANCE REQUIREMENTS SR 3.7.16.1 SURVEILLANCE Verify the specific activity of the secondary coolant is 0.10 IlCi/gm DOSE EQUIVALENT 1-131. '{J COMPLETION TIME 6 hours 36 hours FREQUENCY 31 days P (qlA '7/ l e. tv (',.,; J fr-v i--fc-A-(( C l( Vogtle Units 1 and 2 3.7.16-1 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) A e-feC'-u c e.. 6--2--, / ( 19. G2.2.7001 A new system engineer has requested that CCP 1A be started with the discharge valve throttled to determine current and flow rate data under these conditions. Which ONE of the following describes the process for evaluating this test and who can perform this evaluation? The 10CFR50.59 process is required to determine if ......... . A. an NRC review of the results is required after conducting the test and the 50.59 evaluation can be performed by any currently licensed SRO. NRC approval is required prior to conducting the test and the 50.59 evaluation can only be performed by a qualified reviewer. C. PRB approval is required prior to conducting the test and the 50.59 evaluation can be performed by any currently licensed SRO. D. a PRB review of the results is required after conducting the test and the 50.59 evaluation can only be performed by a Qualified Reviewer. Page: 36 of 49 6/6/2007 ( G2.2.7 Knowledge of the process for conducting tests or experiments not described in the safety analysis report. KIA MATCH ANALYSIS Question gives a plausible scenario where engineering wants to perform a test on a safety related MOV. The candidate must determine what is required to allow performance of the test. Questions meets 10CFR55.43(b) criteria for item # 3 -Facility license procedures required to obtain authority for design and operating changes in the facility. Question also is SRO only due to the KA catalog # for RO only rates a 2.0 importance factor while the rating is 3.2 for SRO. Therefore, question is an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible candidate may think an NRC review of the test results is required and any qualified SRO may perform. B. Correct. Determines if NRC approval is required and a Qualified Reviewer to review. C. Incorrect. Plausible the candidate may think the PRB is highest level of approval but 10CFR50.59 is to determine if NRC approval required. Also, plausible the candidate may think the SRO level is all that is required. D. Incorrect. Plausible the candidate may think a PRB review of the test results is required and must be performed by a qualified reviewer. REFERENCES NMP-AD-010, "10CFR50.59 Screening and Evaluations" Oconee June 2004 NRC SRO Exam question # 21 VEGP learning objectives: LO-LP-63400-01 Explain the purpose of plant design control (SRO only) Page: 37 of 49 6/6/2007 l Number Text LO-LP-63350-04 Describe the requirements for emergency maintenance and when emergency maintenance can be performed. LO-LP-63350-05 State who, by title, must authorize work to begin for normal plant maintenance of plant systems, such as an idle condensate pump. LO-LP-63350-06 Describe the action taken by the SS if the need for an emergency Work Order (WO) arises. LO-LP-63350-07 Define the following terms: a. Tool pouch maintenance

b. Marker c. Emergency maintenance

d. Functional test e. Operations controlled equipment

f. Trouble shooting LO-LP-63350-08 Discuss the types of information that the SS should consider prior to authorizing work, including the documentation of work orders in LCO status sheets. LO-LP-63350-09 Describe how to determine an assigned functional test is adequate on work orders following maintenance where a functional test is not pre-assigned, and who is responsible for this determination.

LO-LP-63354-01 State the purpose of 10CFR50.65, Maintenance Rule. LO-LP-63354-02 Define the following terms relating to the Maintenance Rule: a.O Category (a)(1) b.O Category (a)(2) c.O Critical Safety Function d.O Maintenance Rule Functions e.O Maintenance Rule Scoping Manual f.O Out of Service Time g.O System Unavailable h.O System Reliability LO-LP-63354-03 Describe the Shift Superintendent's responsibility concerning maintenance activities LO-LP-63354-04 Describe what actions can be taken to minimize the impact of maintenance on plant safety. LO-LP-63354-05 Describe the rules concerning emergent work and pre-release activities. LO-LP-63400-01 Explain the purpose of the plant design control (SRO ONLY). LO-LP-63400-03 Briefly describe the essential elements of plant design control. LO-LP-63400-04 Give a brief description of the following plant documents that support plant design control: a.DELETED b.DELETED

c. Design change package d. Design change request e. Field change request f. Minor design change g. Component design change h. As-built notice i. Equivalency determination

j. DELETED k. Request for engineering review I. Request for engineering assistance

m. DELETED n. DELETED Friday, June 01, 2007 Page 96 of 165 c Question 21 OCONEE NRC SRO EXAM 06*25*2004 1 POINT Engineering wants to conduct a temporary test (IT) procedure not described in the safety analysis.

Which ONE of the following describes the process for evaluating this test and who can perform this evaluation? The 10CFR50.59 process Is required to determine if ... A. NRC approval is required prior to conducting this test and can be performed by any currently licensed SRO. B. NRC approval is required prior to conducting this test and can only be performed by a Qualified Reviewer. C. PORC approval Is required prior to conducting this test and can be performed by any currently licensed SRO. D. PORC approval Is required prior to conducting this test and can only be performed by a Qualified Reviewer. c l Question 21 T3 -CFR: 43.3/45.13 OCONEE NRC SRO EXAM 06-25-2004 G2.2.7, Knowledge of the process for conducting tests or experiments not described In the saf analysis. (2.013.2) Answer: B A. Incorrect, first part is correct. Second part is Incorrect. In Operations only Qualified Reviewers can perfonn 1OCFR50.59 screenings and evaluations. B. Correct, the 1OCFR50.59 process Is required to determine If NRC approval Is required prior to conducting a teet and in Operations can only be performed by a Qualified Reviewer. C. Incorrect, first part is incorrect. The 10CFR50.59 process is NOT used to determine If PORe approval is required. The 10CFR50.59 process is required to determine if NRC approval is requIred prior to conducting a test. Second part Is Incorrect. In Operations only Qualified Reviewers can perform 10CFR50.59 screenings and evaluations. D. Incorrect, first part Is incorrect. The 1OCFRSO.59 process is NOT used to detennine if PORC approval is required. The 1OCFR50.59 process is required to determine If NRC approval is required prior to conducting a test. Second part is correct. Technical Reference(s): NSD-209 (1OCFR50.59 Process) Proposed references to be provided to applicants during examination: None Learning Objective: None Question Source: NEW Question History: Last NRC Exam ____ _ Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis Rewrote question to better match KA. Southern Nuclear Operating Company SOUTHERN A Nuclear Management 10 CFR 50.59 Screenings and Evaluations COMPANY Procedure Lurt:l,.$nwY .. Procedure Owner: W. F. Kitchens 1 Fleet Improvement Mgr. Approved By: Effective Dates: (Print: Name I Title I Site) Original Signed by W. F. Kitchens (Procedure Owner's Approval Signature) 01/18/07 Corporate 01/18/07 FNP 1 03/30107 HNP NMP-AD-010 Version 1.0 Page 1 of 19 01/04/2007 (Approval Date) 02/16/07 VEGP The individuals listed below are members of the Performance Improvement Peer Team and are responsible for the creation and maintenance of this procedure. SNC Corporate Office -W. F. Kitchens (Champion) Farley Nuclear Plant -W. R. Bayne Hatch Nuclear Plant -K. A. Underwood Vogtle Electric Generating Plant -W. G. Copeland SNC Corporate Office -T. M. Milton (Procedure Writer) PROCEDURE USAGE REQUIREMENTS SECTIONS Procedure must be open and readily available at the Continuous Use: work location. Follow procedure step by step unless otherwise directed by the procedure. Reference Use: Procedure or applicable section(s) available at the work location for ready reference by person performing steps. Information Use: Available on site for reference as needed. ALL Printed: 6/6/20074:14 AM ( Southern Nuclear Operating Company Nuclear NMP-AO-010 SOUIHERNA Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY urrt:/_Sn-Y .. rVllnU" Procedure Page 2 of 19 Version Number 1.0 Printed: 6/6/20074:14 AM Procedure Version Description Version Descri tion Version 1.0 represents a common approach to be used at all SNC locations for preparing 10 CFR 50.59 Screenings and Evaluations. It supercedes the following site procedures which were previously used to perform this task:

• Corporate:

TS-003, Version 3.0

• Farley: FNP-0-AP-88, Version 8.0
• Hatch: 1 OAC-MGR-01 0-0, Version 6.0
• Vo tie: 00056-C, Version 21.2 r ( Southern Nuclear Operating Company SOUTHERN..\.

Nuclear NMP-AD-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY Y",r w.rIir Procedure Page 3 of 19 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 Table of Contents Page Purpose ...................................................................................................................................... 4 Applicability ................................................................................................................................ 4 References ................................................................................................................................. 5 Definitions .................................................................................................................................. 5 Responsibilities .......................................................................................................................... 5 Procedure .................................................................................................................................. 7 Records .................................................................................................................................... 18 Commitments ............................................. .............................................................................. 18 ( Southern Nuclear Operating Company SOUTHERN.\. Nuclear NMP-AD-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY U"D'.Sn-y .. ,.1¥WU' Procedure Page 4 of 19 1.0 Purpose This procedure describes the process for compliance with the requirements of 10 CFR 50.59 using the guidelines contained in NEI 96-07, Revision 1. This document defines the responsibilities and establishes the controls and methods for the 10 CFR 50.59 process including development and review. Methods described herein are sufficiently detailed to enable individuals to organize and prepare 10 CFR 50.59 screens/evaluations needed to support various activities to be performed within Southern Nuclear Operating Company (SNC). 2.0 Applicability This procedure applies to all personnel within the SNC organization who prepare and review 10 CFR 50.59 screens/evaluations. Activities for which 10 CFR 50.59 screens/evaluations must be prepared are identified by completion of an Applicability Determination Checklist. Refer to NMP-AD-008 (Reference 3.5) for information on preparation of an Applicability Determination. If an Applicability Determination concludes that 10 CFR 50.59 is applicable to an activity, then a 10 CFR 50.59 screen is performed which further determines if a 10 CFR 50.59 evaluation is required. Accordingly, a 10 CFR 50.59 evaluation establishes the basis for determining if NRC approval must be obtained prior to implementation of the activity. NRC approval is obtained by licensee application for a license amendment. An explanation of the process for performing a 10 CFR 50.59 screen and a 10 CFR 50.59 evaluation is provided in this procedure. Additional guidance regarding digital design activities is provided in Reference 3.4. For to-digital changes that appear to be like-for-like replacements, an equivalency evaluation should be performed to determine if the replacement is a plant design change (subject to 10 CFR 50.59) versus a maintenance activity (subject to 10 CFR 50.65). Digital-to-digital changes may not necessarily be like-for-like because the system behaviors, response time, failure modes, etc. for the new system may be different from the old system. If the vendor, hardware, firmware, application software, and configuration data are identical, then the upgrade may be a like maintenance activity to be performed in accordance with 10 CFR 50.65, and 10 CFR 50.59 would not apply. Note: If it has been predetermined and documented in a procedure that an activity or group of activities do not require a 10 CFR 50.59 evaluation, neither an Applicability Determination nor a 10 CFR 50.59 screen need to be performed or documented. The procedure serves as the documented Applicability Determination/10 CFR 50.59 screen. Otherwise, an Applicability Determination Checklist should be completed for proposed activities. In general, this procedure applies to the implementation of certain activities that affect the following:

• Changes to structures, systems, and components outlined, summarized, or completely described in the Updated Final Safety Analysis Report (FSAR) including items incorporated by reference. (Refer to NEI 98-03, Revision 1, "Guidelines for Updating Final Safety Analysis Reports" for additional guidance on this topic.)
• Permanent and temporary design changes.
• Changes to plant procedures outlined, summarized or completely described in the Updated FSAR. Printed: 6/6/20074:14 AM

( Southern Nuclear Operating SOUTHERN A Nuclear NMP-AD-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY &"'DIOSw"" Ho.,WfwV Procedure Page 5 of 19 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9

• Test or experiments not described in the Updated FSAR.
• Revisions to NRC approved analysis methodology or assumptions as described in the Updated FSAR.
• Proposed compensatory actions to address degraded or non-conforming conditions.

References 10 CFR 50.59: Changes, Tests, and Experiments. NEI 96-07, Revision 1: Guidelines for 10 CFR 50.59 Implementation (NEI96-07, Rev. 1). Regulatory Issue Summary (RIS) 2005-20, Revision To Guidance Formerly Contained In NRC Generic Letter 91-18, Information To Licensees Regarding Two NRC Inspection Manual Sections On Resolution Of Degraded and Nonconforming Conditions and on Operability, dated September 26,2005. RIS 2002-22, Use of EPRI/NEI Joint Task Force Report, Guideline on Licensing Digital Upgrades: EPRI TR-102348, Revision 1, NEI 01-01: A Revision of EPRI TR-102348 to reflect changes to the 10 CFR 50.59 Rule. NMP-AD-008, Applicability Determinations. NMP-AD-009, Licensing Document Change Requests. Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments. Plateau Qualification, S-CFR-EVAL (10 CFR 50.59/72.48 Training). Plant Review Board, NMP-GM-009. 3.10 NMP-AD-01 0-F01, 10 CFR 50.59 Screening/Evaluation Form. 4.0 Definitions Refer to NEI 96-07, Revision 1, Section 3 for the definition of terms used in conjunction with this procedure. Refer to Reference 3.4 for additional definition of terms applicable to digital design activities.

5.1 Manager

5.1.1 Assures

that 10 CFR 50.59 screens/evaluations prepared within the group receive adequate reviews. 5.1.2 Assures appropriate personnel within the group are qualified (Reference 3.8) to perform or review 10 CFR 50.59 screens/evaluations to support activities under their responsibility. Printed: 6/6/20074:14 AM ( ( Southern Nuclear Operating Company SOUTHERN A Nuclear NMP-AO-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY Procedure Page 6 of 19 EM'VNSnwY .. ,.VWU" 5.1.3 Recommends approval to the PRB for 10 CFR 50.59 screens that have a Yes answer in Section B of NMP-AO-01 0-F01 (Reference 3.10) for activities under their responsibility. 5.2 10 CFR 50.59 Preparer The preparer is a knowledgeable person trained in the requirements of 10 CFR 50.59 (Reference 3.8), with additional technical qualifications determined by the preparer's management sufficient to prepare or revise a 10 CFR 50.59 screen/evaluation to support an activity within the preparer's area of responsibility.

5.2.1 Performs

necessary research (includes obtaining data from various engineering disciplines or organizations, if necessary) to develop a 10 CFR 50.59 screen/evaluation that is technically correct. 5.2.2 Using the guidance contained in References 3.2 and 3.4, as applicable, prepares complete, consistent, clear, and accurate 10 CFR 50.59 screens/evaluations. 5.3 10 CFR 50.59 Reviewer The reviewer is a knowledgeable person trained in the requirements of 10 CFR 50.59 (Reference 3.8), with additional technical qualifications determined by the reviewer's management sufficient to prepare the 10 CFR 50.59 screens/evaluations being reviewed. The reviewer should not have participated in the preparation of the 10 CFR 50.59 screen/evaluation. The reviewer concurs that the 10 CFR 50.59 screen/evaluation is complete, consistent, clear, and accurate, and endorses the conclusion of the 10 CFR 50.59 screen/evaluation upon satisfactory disposition of the reviewer's comments. 5.4 10 CFR 50.59 Nuclear Regulatory Reviewer 5.5 The Nuclear Regulatory reviewer is a knowledgeable person trained in the requirements of 10 CFR 50.59 (Reference 3.8). The Nuclear Regulatory reviewer is also knowledgeable regarding other regulatory requirements that may be applicable to the activity. The Nuclear Regulatory reviewer concurs that the 10 CFR 50.59 screen/evaluation is complete, consistent, clear, and accurate, and endorses the conclusions of the 10 CFR 50.59 screen/evaluation upon satisfactory disposition of the reviewer's comments. A Nuclear Regulatory review signature is not required for a 10 CFR 50.59 screen/evaluation prepared at the plant site. 10 CFR 50.59 Nuclear Hazards Reviewer The Nuclear Hazards reviewer is a knowledgeable person trained in the requirements of 10 CFR 50.59 (Reference 3.8). The Nuclear Hazards reviewer is also knowledgeable regarding hazard requirements related to issues such as tornados, missiles, heavy loads, pipe break, floods, pressure/temperature analysis, hydrogen generation, radiation protection and shielding, and toxic chemicals. The Nuclear Hazards reviewer concurs that the 10 CFR 50.59 screen/evaluation has appropriately considered hazard issues. A Nuclear Hazards review signature is not required for a 10 CFR 50.59 screen/evaluation prepared at the plant site. Printed: 6/6/20074:14 AM ( Southern Nuclear Operating Company SOUTHERN..\. Nuclear NMP-AO-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY Procedure Page 7 of 19 W"D"'Snw,...,"ArU" 5.6 Personnel Requirements Personnel who prepare and review 1 0 CFR 50.59 screens/evaluations will be trained in accordance with Reference 3.8. Additionally, 10 CFR 50.59 screens/evaluations will be prepared and reviewed by persons working within a quality assurance program that conforms to 10 CFR 50, Appendix B. 5.7 Plant Review Board/Qualified Reviewer Reviews selected 10 CFR 50.59 screens/evaluations to make a determination as to whether a license amendment is involved.

5.8 Safety

Review Board Selected 10 CFR 50.59 screens/evaluations are reviewed for: 1) changes to procedures, structures, systems, or components and 2) tests or experiments completed under the provision of 10 CFR 50.59 to verify such actions do not involve a license amendment.

6.0 Procedure

This section establishes the basis for a common understanding and application of 10 CFR 50.59, and establishes minimum requirements to ensure consistency in compliance with 10 CFR 50.59. 6.1 10 CFR 50.59 Consideration

6.1.1 Quotation

from 10 CFR 50.59 "(c)(1) A licensee may make changes in the facility as described in the final safety analysis report (as updated), make changes in the procedures as described in the final safety analysis report (as updated), and conduct tests or experiments not described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to § 50.90 only if: Note: For the purposes of this procedure, Updated FSAR refers to the current FSAR as updated per 10 CFR 50.71(e), approved changes to the Updated FSAR which have not yet been submitted to the NRC by amendment, and documents incorporated into the Updated FSAR by reference. (i) A change to the technical specifications incorporated in the license is not required, and (ii) The change, test, or experiment does not meet any of the criteria in paragraph (cX2) of this section. (c)(2) A licensee shall obtain a license amendment pursuant to § 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would: Printed: 6/6/20074:14 AM ( SOUTHERN A. COMPANY bl'D,.SnwY .. Southern Nuclear Operating Company Nuclear NMP-AO-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 Procedure Page 8 of 19 (i) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the final safety analysis report (as updated); (ii) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the final safety analysis report (as updated); (iii) Result in more than a minimal increase in the consequences of an accident previously evaluated in the final safety analysis report (as updated); (iv) Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the final safety analysis report (as updated); (v) Create a possibility for an accident of a different type than any previously evaluated in the final safety analysis report (as updated); (vi) Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the final safety analysis report (as updated); (vii) Result in a design basis limit for a fission product barrier as described in the FSAR (as updated) being exceeded or altered; or (viii) Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses. (c)(3) In implementing this paragraph, the FSAR (as updated) is considered to include FSAR changes resulting from evaluations performed pursuant to this section and analyses performed pursuant to § 50.90 since submittal of the last update of the final safety analysis report pursuant to § 50.71 of this part. (c)(4) The provisions in this section do not apply to changes to the facility or procedures when the applicable regulations establish more specific criteria for accomplishing such changes. (d)( 1) The licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant to paragraph (c) of this section. These records must include a written evaluation which provides the bases for the determination that the change, test or experiment does not require a license amendment pursuant to paragraph (c)(2) of this section." (d)(2) The licensee shall submit, as specified in § 50.4, a report containing a brief description of any changes, tests, and experiments, including a summary of the evaluation of each. A report must be submitted at intervals not to exceed 24 months. (d)(3) The records of changes in the facility must be maintained until the termination of a license issued pursuant to this part or the termination of a license issued pursuant to 10 CFR Part 54, whichever is later. Records of changes in procedures and records of tests and experiments must be maintained for a period of 5 years. Printed: 6/6/20074:14 AM ( Southern Nuclear Operating Company SOUTHERN A Nuclear NMP-AO-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY Procedure Page 9 of 19 ur'V,.SnwY .. r.v 6.1.2 Explanation of 10 CFR 50.59 6.1.2.1 General 6.1.2.1.1 6.1.2.1.2 6.1.2.1.3 6.1.2.1.4 Printed: 6/6/20074:14 AM 10 CFR 50.59 is a legal requirement established to prevent plant changes from degrading the safety design basis of the plant.

• If the licensee can show (through a 10 CFR 50.59 screen/evaluation and a review of the Technical Specifications and Environmental Protection Plan) a proposed activity does not require a license amendment, the NRC, by authority of 10 CFR 50.59, grants the utility pre-approval to conduct the activity.

In such cases, the utility has the leeway to do what it determines is best for operation of the facility, provided the activity does not conflict with another license requirement.

• However, if a 10 CFR 50.59 screen/evaluation and/or review of the Technical Specifications or the Environmental Protection Plan concludes that an activity proposed by the utility is contrary to the basis for design and operation of the plant as described in the Updated FSAR, and/or requires a change to the Technical Specifications or the Environmental Protection Plan, prior NRC approval and an amendment to the operating license is required prior to conducting the activity. (Refer to subsection 6.4.2 for guidance on exceptions.)

The 10 CFR 50.59 screen/evaluation provides confidence that safety is maintained. A change involving a license amendment is not necessarily unsafe. In fact, a change would be proposed for implementation only if it were found to be safe. The fact that a change involves a license amendment only means prior NRC review and approval are necessary for implementation of the change. The range of changes covered in 10 CFR 50.59 is very broad. It includes changes to non safety-related, non safety-related that could impact safety, as well as safety-related systems and documents. If, during the process of preparing a 10 CFR 50.59 screen/evaluation for a proposed activity, it is determined the activity requires a license amendment, one of four actions will normally occur; (1) a more detailed evaluation incorporating all involved parties will be performed to determine whether the activity can be reclassified so it does not require a license amendment, (2) revision(s) to the activity will be attempted such that it no longer requires a license amendment, (3) the change will be withdrawn, or (4) a request for a license amendment will be prepared. ( Southern Nuclear Operating Company Nuclear NMP-AD-010 SOUTHERN A Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY &,'V.SnwY .. rWWU* Procedure Page 10 of 19 6.1.2.2 Simplified Statement of the Intent of 10 CFR 50.59 6.1.2.2.1 The maintenance of an operating license implies:

• Plant design basis is adequate to maintain previously NRC approved margins of safety.
• The structures, systems, and components described in the Updated FSAR have been "analyzed" to assure their proper function whenever required.

6.1.2.2.2 The intent of 10 CFR 50.59 is to prohibit changes which might:

• Defeat needed functions or exceed "analyzed" capabilities of structures, systems, or components.
• Adversely affect the frequency or likelihood that the plant can and will be operated without undue risk to public health and safety. 6.2 10 CFR 50.59 Screening Criteria

6.2.1 Structure

for Performing 10 CFR 50.59 Screenings 6.2.1.1 6.2.1.2 Printed: 6/6/20074:14 AM For the purposes of performing 10 CFR 50.59 screenings, the 10 CFR 50.59(c)(1) criteria identified in subsection 6.1.1 must be addressed. Using the guidance contained in NEI 96-07, Revision 1 (also refer to Reference 3.4 for additional guidance regarding digital design activities), these criteria have been rephrased into the following five questions:

1. Does the activity involve a modification, addition to, or removal of a structure, system, or component (SSC) such that a design function as described in the Updated FSAR is adversely affected?

2. Does the activity involve a change to procedures that adversely affects the performance or method of control of a design function as described in the Updated FSAR?

3. Does the activity involve an adverse change to a method of evaluation or use of an alternate method of evaluation from that described in the Updated FSAR that is used in establishing the design bases or in the safety analyses?

4. Does the activity involve a test or experiment not described in the Updated FSAR which is outside the reference bounds of the design bases as described in the Updated FSAR or is inconsistent with the analyses or descriptions described in the Updated FSAR? 5. Does the activity involve a change to the Technical Specifications and/or Environmental Protection Plan (See Sections 6.4 and 6.6.2.1.4.)?

The screening portion of the 10 CFR 50.59 screen/evaluation form shown in NMP-AD-010-F01 (Reference 3.10) is structured to list each of the screening ( Southern Nuclear Operating Company Nuclear NMP-AD-010 SOUTHERN A. Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY £"0'. Snw li.,.WWU* Procedure Page 11 of 19 6.3 questions with a YES and NO block before each question. A NOT APPLICABLE (N/A) block is also included before the 10 CFR 50.59 evaluation questions which are discussed in Section 6.3. An explanation of either a YES or NO response is required. The form also provides ample space to provide -the justification for each answer. For an activity not to require a 10 CFR 50.59 evaluation, the answer to all of the 10 CFR 50.59 screening questions must be NO. If 10 CFR 50.59 screening questions 1, 2, or 4 are answered YES, then 10 CFR 50.59 evaluation questions 1-7 must be answered; 10 CFR 50.59 evaluation question 8 is answered N/A. If only 10 CFR 50.59 screening question 3 is answered YES, then only 10 CFR 50.59 evaluation question 8 must be answered; evaluation questions 1-7 are answered N/A. If question 5 is answered YES, which indicates a license amendment request is required, and all aspects of the activity will be addressed in the license amendment request, then preparation of a 10 CFR 50.59 evaluation is not required. Refer to NEI 96-07, Revision 1, Section 4 for additional guidance on how to answer these questions (also refer to Reference 3.4 for additional guidance regarding digital design activities). 10 CFR 50.59 Evaluation Criteria 6.3.1 Structure for Performing 10 CFR 50.59 Evaluations 6.3.1.1 For the purposes of performing 10 CFR 50.59 evaluations, the 10 CFR 50.59( c)(2) criteria identified in subsection 6.1.1 must be addressed. Using the guidance contained in NEI 96-07, Revision 1 (also refer to Reference 3.4 for additional guidance regarding digital design activities), these criteria have been rephrased into the following eight questions. Note that a pre-screening question (7a) has been developed for criterion (vii): 1. Does the proposed activity result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the Updated FSAR? 2. Does the proposed activity result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the Updated FSAR? 3. Does the proposed activity result in more than a minimal increase in the consequences (Le., radiological) of an accident previously evaluated in the Updated FSAR? 4. Does the proposed activity result in more than a minimal increase in the consequences (Le., radiological) of a malfunction of an SSC important to safety previously evaluated in the Updated FSAR?

5. Does the proposed activity create the possibility for an accident of a different type than any previously evaluated in the Updated FSAR? 6. Does the proposed activity create the possibility for a malfunction of an sse important to safety with a different result than any previously evaluated in the Updated FSAR? Printed: 6/6/20074:14 AM

( l Southern Nuclear Operating Nuclear NMP-AD-010 SOUTHERN..\. Management 10 CFR 50.59 Screenings and E,(aluations Version 1.0 COMPANY Lf,'V,.s"..",y .... rllWU* Procedure Page 12 of 19 7. a. Does the proposed activity have any impact on the integrity of the fuel cladding, reactor coolant pressure boundary, or containment? If "YES", answer Question 7.b. b. Does the proposed activity result in a design basis limit for a fission product barrier as described in the Updated FSAR being exceeded or altered? Note: To answer this question, refer to the list of fission product barriers identified in Updated Farley FSAR Table 15.1-6, Hatch FSAR Table 15.1-3, and Vogtle FSAR Table 15.0.8-2, as applicable. 6.3.1.2 8. Does the proposed activity result in a departure from a method of evaluation described in the Updated FSAR used in establishing the design bases or in the safety analyses? The evaluation portion of the 10 CFR 50.59 screen/evaluation form shown in NMP-AD-010-F01 (Reference 3.10) is structured to list each of the evaluation questions with a YES, NO, and Not Applicable (N/A) block before each question. An explanation of the YES, NO, or N/A response is required. For an activity not to involve a license amendment, the answer to all of the foregoing questions must be either NO or N/A (with the exception of question 7.a). If the answer to one or more of the questions is YES (with the exception question 7.a), a license amendment must be obtained from the NRC. Refer to NEI 96-07, Revision 1, Section 4 for additional guidance on how to answer these questions (also refer to Reference 3.4 for additional guidance regarding digital design activities).

6.3.2 Level

of Detail 6.3.2.1 Printed: 6/6/20074:14 AM In describing the activity, the 10 CFR 50.59 screen/evaluation preparer should include:

• What systems and components are affected by the change, including their safety classification?
• What was the design function of the structure, system, or component?
• What parameters of the accident/transient analysis are affected by the change?
• What design basis accidents or operational transients were reviewed for impact?
• What failure modes of the change were reviewed?
• What systems and analyses are indirectly impacted (e.g., diesel generator loading, electromagnetic/radiofrequency interference, seismic II/I, etc.)? Documenting the effects considered and the references consulted will allow the independent reviewer to be able to note/address any effects that were not considered.

These effects must be addressed in responding to the eight c SOUTHERN A COMPANY htl"D" StTw r .. ,.1JWU' Southern Nuclear Operating Company Nuclear NMP-AD-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 Procedure Page 13 of 19 questions of the 10 CFR 50.59 evaluation. The 10 CFR 50.59 evaluation shall document how the evaluator reached the conclusion for each question by considering, as a minimum, the effects enumerated below. 6.3.2.2 In performing a 10 CFR 50.59 evaluation for a proposed activity, the preparer must answer the eight questions (refer to subsection 6.3.1.1) to determine whether a license amendment is required. Although the answers are either YES, NO, or N/A, there must be an accompanying explanation providing justification for the answer. These explanations must be complete in the sense that an independent reviewer could draw the same conclusion and arrive at the same answer. However, in the routine of conducting a 10 CFR 50.59 evaluation, evaluators may be inclined to short cut this aspect and provide explanations that are either too brief or deficient. For example, in support of a NO answer, the temptation is to reverse the word order of the question such that it becomes a simple statement of conclusion rather than provide words of explanation; e.g., "This activity does not increase the frequency of an accident previously evaluated in the Updated FSAR." Also, some evaluators provide an explanation in answer to one question and defer all explanations of other answers to that explanation. This approach will be used only when a single explanation clearly justifies all answers. Likewise, a single explanation will not be used in multiple cases unless it clearly justifies the answers to which it is being "matched." Sufficient documentation must be available to demonstrate to an independent reviewer that correct engineering judgment was applied or that the specifications based on safety requirements have been met. The importance of documentation is emphasized by the fact that often experience and engineering judgment are relied upon in making the license amendment determination. Since an important goal of the 10 CFR 50.59 screen/evaluation is completeness, the items considered by the evaluator must be clearly specified.

6.4 Technical

Specifications/Environmental Protection Plan Considerations 6.4.1 The provisions of 10 CFR 50.59 allow a licensee to engage in activities that do not require prior NRC approval provided such an activity satisfies the eight 50.59(c)(2) evaluation criteria discussed in subsection 6.3.1, and does not involve a change to the Technical Specifications or the Environmental Protection Plan. Note that 10 CFR 50.59 basically equates not meeting anyone of the eight 10 CFR 50.59(c)(2) evaluation criteria with a Technical Specification or Environmental Protection Plan change, in that for each, prior NRC approval in accordance with a license amendment is required. 6.4.2 Any activity that causes a change, however slight, to an existing Technical Specification or the Environmental Protection Plan requires prior NRC approval in accordance with a license amendment. The situation becomes less clear for an activity that does not cause a change to an existing Technical Specification or the Environmental Protection Plan but does have the potential to become an addition to the Technical SpeCifications or the Environmental Protection Plan. In such situations, the 10 CFR 50.59 evaluation should be completed, and the decision as to whether NRC approval is necessary should be based on whether or not a license amendment is required. Printed: 6/6/20074:14 AM ( ( SOUTHERN A COMPANY E.'t:/,.SnwY_rWWU" 6.4.3 Southern Nuclear Operating Company Nuclear NMP-AD-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 Procedure Page 14 of 19 A change to the Technical Specifications or the Environmental Protection Plan is typically associated with a physical or administrative required change to the plant. However, a change to the Technical Specifications or the Environmental Protection Plan can also occur outside those activities when simply revising a safety analysis. Therefore, when an activity only involves a change to an analytical basis, the associated Technical Specifications or the Environmental Protection Plan should be reviewed for potential impact. 6.4.4 A review of the activity will be performed/documented on the 10 CFR 50.59 screen/evaluation form shown in NMP-AD-010-F01 (Reference 3.10) to determine whether an existing Technical Specification or the Environmental Protection Plan will be impacted or whether an addition to the Technical Specifications or the Environmental Protection Plan is required.

6.5 Resolution

of Degraded and Non-conforming Conditions 6.5.1 NRC guidance regarding the corrective action associated with a degraded and conforming condition that may involve a 10 CFR 50.59 screen/evaluation is contained in RIS 2005-20 (Reference 3.3). This direction is consistent with the guidance for degraded and non-conforming conditions that is also contained in NEI 96-07, Revision 1. As discussed in NEI 96-07, Revision 1, there are three potential scenarios for addressing the applicability of 10 CFR 50.59 to a degraded and non-conforming condition (also refer to Reference 3.4 for additional guidance regarding digital design activities). They are as follows:

• If the licensee intends to restore the structure, system, or component back to its previous condition (as described in the Updated FSAR), then this corrective action should be performed in accordance with 10 CFR 50, Appendix B (Le., in a timely manner commensurate with safety). This activity is not subject to 10 CFR 50.59.
• If an interim compensatory action is taken to address the condition and involves a temporary procedure or facility change, 10 CFR 50.59 should be applied to the temporary change. The intent is to determine whether the temporary change/compensatory action itself (not the degraded condition) impacts other aspects of the facility described in the Updated FSAR. In considering whether a temporary change impacts other aspects of the facility, particular attention should be given to ancillary aspects of the temporary change that result from actions taken to directly compensate for the degraded condition.
• If the licensee corrective action is either to accept the condition "as-is" resulting in something different than described in the Updated FSAR or to change the facility or procedures to something different than described in the Updated FSAR, then 10 CFR 50.59 should be applied to the corrective action unless another regulation applies (e.g., 10 CFR 50.55a). In these cases the final resolution becomes the proposed change that would be subjected to 10 CFR 50.59. Printed: 6/6/20074:14 AM

( Southern Nuclear Operating Company Nuclear NMP-AD-010 SOUTHERN A. Management COMPANY 10 CFR 50.59 Screenings and Evaluations Version 1.0 Lt".",.SITw y .. ,wwu* Procedure Page 15 of 19 6.6 10 CFR 50.59 Screen/Evaluation Process 6.6.1 General Guidelines for Performing 10 CFR 50.59 Screens/Evaluations 6.6.1.1 6.6.1.2 6.6.1.3 6.6.1.4 6.6.1.5 Be as specific as possible. Use a computer generated form provided in NMP-AD-010-F01 (Reference 3.10). Identify and perform a thorough review of applicable sections of the Updated FSAR, the Technical Specifications, the Environmental Protection Plan, and other applicable licensing documents (e.g., the Offsite Dose Calculation Manual, Process Control Program, Core Operating Limits Report, PressureiTemperature Limits Report, Technical Requirements Manual, Technical Specification Bases, etc.). Utilize computer word search capability, if available, as an aid in accomplishing this task. Remember that a word search is only as good as the words that are chosen for the search. For example, to search the FSAR or TS for reactor coolant pump, you might have to enter "reactor coolant pump," "RCP," "pump," etc. If possible, state where the applicable system, component, etc., is addressed and why the specific activity is not covered. If the activity is not covered in these documents, so indicate and include a reference to the sections of these documents reviewed. From the nature of the change, determine whether it may affect any safety functions. For each potentially affected function, review the list of evaluated accidents in the Updated FSAR (e.g., Chapters 6 and 15) and determine whether the system in question (Le., the one covered/affected by the 10 CFR 50.59 review being performed) plays a part in initiating the accident or mitigating the severity of the accident. For each accident identified, review the following information in the Updated FSAR:

• Any analysis available for the accident in question and determine whether any of the variables are affected by the changes for which the 10 CFR 50.59 screen/evaluation is being prepared.
• The methods of accident mitigation to determine whether they are affected by the changes for which the 10 CFR 50.59 screen/evaluation is being prepared.
• The methods of evaluation which means the calculational framework as described in the Updated FSAR accident analyses used for evaluating behavior or response of the facility or an SSC. Ensure any conditional requirements contained within a 10 CFR 50.59 screen/evaluation are also contained within the applicable document.

For example, if the 10 CFR 50.59 screen/evaluation depends upon the fact the unit is in cold shutdown, an appropriate administrative control must be in effect to ensure the unit is in cold shutdown during performance of the activity. Printed: 6/6/20074:14 AM Southern Nuclear Operating_ Company Nuclear NMP-AD-010 SOUTHERN A Management COMPANY 10 CFR 50.59 Screenings and Evaluations Version 1.0 .. Procedure Page 16 of 19 6.6.1.6 6.6.1.7 6.6.1.8 Ensure the 1 0 CFR 50.59 screen/evaluation addresses any system interactions that could occur; be particularly careful of non-safety system actions that can impact safety-related systems. Ensure the 10 CFR 50.59 screen/evaluation is performed for the "scope" of the document or activity to which it applies by adhering to the following guidelines:

• The 10 CFR 50.59 screen/evaluation is to address the effect of a change.
• If a change involves a temporary plant alteration that supports maintenance or a temporary alteration that supports the installation and post-modification testing of an approved plant change, then those aspects of the change are considered maintenance activities to be assessed and managed in accordance with 10 CFR 50.65(a)(4) provided the temporary plant alteration will be restored to its as-designed condition prior to startup if shutdown or will be restored to its as-designed condition within 90 days during power operations (Modes 1 and 2). 10 CFR 50.59 would not apply to those aspects of the change. If the temporary plant alteration is not restored prior to startup, or within 90 days during power operation, a 10 CFR 50.59 screen/evaluation will be performed.

Significant changes to a 10 CFR 50.59 screen/evaluation will not normally be made once it has received plant approval; such changes will normally require a revision to the associated document. If it is necessary to make such changes to an approved 10 CFR 50.59 screen/evaluation, each change should receive the same level of review as before. Note: Pen and ink changes to 10 CFR 50.59 screenings/evaluations may be made for misspellings, typographical and grammatical errors, and obvious incorrect numerical transpositions such as step numbers. Pen and ink changes shall have concurrence of the 10 CFR50.59 Preparer and the 10 CFR 50.59 Reviewer. The person making the pen and ink change shall be 10 CFR 50.59 qualified, and shall initial and date the change with any needed explanations. Pen and ink changes shall not be used to correct technical deficiencies. Pen and ink changes shall not be made after site approval. 6.6.2 10 CFR 50.59 Screen/Evaluation Performance Note: The 10 CFR 50.59 screen/evaluation form shown in NMP-AD-010-F01 (Reference 3.10) will be completed in accordance with the instructions in this subsection. The Section Letter designation, where applicable in a step in this procedure, is a cross-reference to the Section Letter on the form. 6.6.2.1 The 10 CFR 50.59 screen/evaluation preparer will: 6.6.2.1.1 Printed: 6/6/20074:14 AM Indicate the following in the header of the form:

• Check the box for the applicable plant and the box for the applicable unit(s) (e.g., 1,2, or shared, but not both 1 and 2) to which the 10 CFR 50.59 screen/evaluation applies.

SOUTHERN A COMPANY Southern Nuclear Operating Company Nuclear NMP-AD-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 Procedure Page 17 of 19 6.6.2.1.2 6.6.2.1.3 6.6.2.1.4

• The type of activity/document to which the 1 0 CFR 50.59 screen/evaluation applies (e.g., Design Change Request, Request for Engineering Review, Licensing Document Change Request, Condition Report, etc.), that initiated the change.
• The activity/document number to which the 10 CFR 50.59 screen/evaluation applies such that the 10 CFR 50.59 screen/evaluation and the document describing the activity may be uniquely associated.
• The 10 CFR 50.59 screen/evaluation version number. This is a major version only beginning with 1.0. Each subsequent revision of a 10 CFR 50.59 screen/evaluation advances the version number to the next sequential major version, 2.0, 3.0, 4.0, etc.
• The title of the activity/document to which the 10 CFR 50.59 screen/evaluation applies or other means of identifying the document.
• The proposed version number of the activity/document to which the 10 CFR 50.59 screen/evaluation applies. In Section A, "Activity Summary," provide a description of the proposed change. This description should provide a general discussion of the proposed change and why it is necessary.

It should include any references necessary to support the responses to the 10 CFR 50.59 screening and 10 CFR 50.59 evaluation questions. Any limitations on how or when the activity or change is to be performed should be included. Answer the 10 CFR 50.59 screening questions contained in Section B, "10 CFR 50.59 Screening," and give the basis for the answers using the guidelines contained in NEI 96-07, Revision 1, Section 4 (also refer to Reference 3.4 for additional guidance regarding digital design activities). If the answer to any question in subsection 6.6.2.1.3 is YES, continue with subsection 6.6.2.1.5. NOTE: If any of the screening questions are answered "YES", PRB review of the 10 CFR 50.59 screening/evaluation is required PRIOR to implementation of the proposed activity. 6.6.2.1.5 Printed: 6/6/20074:14 AM If all of the answers to the 10 CFR 50.59 screening questions contained in Section B are NO, do not complete Sections C and D. Sections C and D should also be deleted from the 10 CFR 50.59 screen/evaluation form. Continue with subsection 6.6.2.1.8. If the entire change is to be submitted for pre-implementation approval to the NRC, then a license amendment determination is not required. Continue with subsection 6.6.2.1.8. Answer the 10 CFR 50.59 evaluation questions contained in Section C, "10 CFR 50.59 Evaluation," and give the basis for each answer using the ( ( Southern Nuclear Operating Company SOUTIIERNA Nuclear NMP-AD-010 Management 10 CFR 50.59 Screenings and Evaluations Version 1.0 COMPANY L",v,.SnwY .. ,IIlJrIJ'* Procedure Page 18 of 19 6.6.2.1.6 6.6.2.1.7 6.6.2.1.8 guidelines contained in NEI 96-07, Revision 1, Section 4 (also refer to Reference 3.4 for additional guidance regarding digital design activities). If the answer to any question answered in Section C is YES (with the exception of question 7.a), a license amendment must be obtained from the NRC before the change/activity may be implemented. Provide a 10 CFR 50.59 evaluation summary in Section D. It should include a brief description of the change and a concise summary of the responses to the evaluation questions provided in Section C. Also check the YES question in Section 0, indicating a copy of the completed 10 CFR 50.59 screen/evaluation will be forwarded to Nuclear licenSing. Sign and date the form in Section A. 6.6.2.2 The 10 CFR 50.59 screen/evaluation reviewer will review the completed form and indicate concurrence/endorsement by signing and dating the form in Section A. Note: For packages prepared by vendors and contractors that do not use NMP-AD-010-F01 to perform a 10 CFR 50.59 review, additional reviewer and approver slots can be pen and inked as needed on their form. 7.0 8.0 6.6.2.3 The 10 CFR 50.59 screen/evaluation Nuclear Hazards reviewer will review the completed form and indicate concurrence/endorsement by signing and dating the form in Section A. A Nuclear Hazards review signature is not required for a 10 CFR 50.59 screen/evaluation prepared at the plant site. 6.6.2.4 The 10 CFR 50.59 screen/evaluation Nuclear Regulatory reviewer will review the completed form and indicate concurrence/endorsement by signing and dating the form in Section A. A Nuclear Regulatory review signature is not required for a 10 CFR 50.59 screen/evaluation prepared at the plant site. 6.6.2.5 The preparer shall forward a copy of the completed 10 CFR 50.59 screen/ evaluation to Nuclear Licensing if Section 0 of the checklist is completed. 6.6.2.6 If required by NMP-GM-009 (Reference 3.9), the PRB will review the completed form and indicate concurrence by signing and dating the form, or providing the PRB meeting number which approved the document. 6.6.2.7 From this point on, the 10 CFR 50.59 screen/evaluation should remain with the document to which it applies and be maintained in plant records in accordance with 10 CFR 50.59(d)(3). Records Documents generated by this procedure are considered QA records and shall be maintained for the life of plants Farley, Hatch and Vogtle. Commitments Printed: 6/6/20074:14 AM ( ( Southern Nuclear Operating Company SOUTHERN A Nuclear Management 10 CFR 50.59 Screenings and Evaluations COMPANY Procedure wro"Snw YH'WWU-8.1 Farley -None 8.2 Hatch -None 8.3 Vogtle -C00010315, Section 1.0 Printed: 6/6/20074:14 AM C00010356, Section 1.0 C00019844, Section 1.0 NMP-AD-010 Version 1.0 Page 19 of 19 Southern Nuclear Operating Company SOUTHERN.A Nuclear 10 CFR 50.59 Management Page 1 of_ COMPANY Procedure Screen i ng/Evaluation EIN'D"s".., YtI.rw.ru" Plant: Farley 0 Hatch 0 Vogtle 0 Unit No. 10 20 Shared 0 Activity/Document No.: 10 CFR 50.59 Version No.: (Act./Doc. Initiating the Change) Activity/Document Version No.: Title: A. Activity Summary Preparer: ! Date: ____ _ Print Signature Reviewer: ! Date: -----Print Signature Nuclear Hazards Reviewer: ! Date: (If required) Print

---

Signature Nuclear Regulatory Reviewer:

! Date: (If required) Print

---

Signature Reviewer!

Approver: ! (As Needed) Print Signature Date: PRB Approval or Date: Meeting No.:

---

Print I Signature or PRB Meeting No. PRB Meeting No. (if applicable and not identified above ):, _____________

_

References:

1. (Plant) FSAR, (Version, date), Section(s)

2. (Plant) Technical Specifications, (Amendment, date), Section(s)

3. (Plant) Environmental Protection Plan, (Amendment, date) 4. Others as appropriate B. 10 CFR 50.59 Screening Does the activity to which this screening applies represent:

1. DYes D No A modification, addition to, or removal of a structure, system, or component (SSe) such that a design function as described in the Updated FSAR is adversely affected?

Basis for Answer: 2. DYes DNo Basis for Answer: A change to procedures that adversely affects the performance or method of control of a design function as described in the Updated FSAR? NMP-AD-01 0-F01, Version 1.0 NMP-AD-010 ( Southern Nuclear Operating Company SOUTHERN A Nuclear 10 CFR 50.59 COMPANY Management Screening/Evaluation Page 2 of_ E-ro" Snw Procedure Plant: Farley 0 Hatch 0 Vogtle 0 Unit No. 10 20 Shared 0 Activity/Document No.: 10 CFR 50.59 Version No.: (Act./Doc. Initiating the Change) Activity/Document Version No.: Title: 3. DVes DNo Basis for Answer: 4. DVes DNo Basis for Answer: 5. DVes DNo Basis for Answer: An adverse change to a method of evaluation or use of an alternate method of evaluation from that described in the Updated FSAR that is used in establishing design bases or in the safety analysis? A test or experiment not described in the Updated FSAR which is outside the reference bounds of the design basis as described in the Updated FSAR or is inconsistent with the analyses or descriptions described in the updated FSAR? A change to the Technical Specifications and/or Environmental Protection Plan incorporated in the operating license? IF the answer to all of the questions in section B is "NO", do not complete sections C and D. Sections C and D should also be deleted from the form. IF the answer to any of questions 1, 2, or 4 in section B is "VES", then only complete the answers to questions 1-7 in section C and complete the summary in Section D. IF only the answer to question 3 in section B is "VES", then only complete the answer to question 8 in section C and complete the summary in section D. IF question 5 is answered "VES", a license amendment is involved which requires NRC approval. Do not complete sections C and D if all aspects of the activity will be addressed in the license amendment request. C. 10 CFR 50.59 Evaluation

1. DVes DNo DN/A Basis for Answer: 2. DVes DNo DN/A Basis for Answer: 3. DVes ONo DN/A Basis for Answer: Does the proposed activity result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the Updated FSAR? Does the proposed activity result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, orcomponent (SSC) important to safety previously evaluated in the Updated FSAR? Does the proposed activity result in more than a minimal increase in the consequences of an accident previously evaluated in the Updated FSAR? NMP-AD-010-F01, Version 1.0 NMP-AD-010 Southern Nuclear Operating Company SOUTHERN A Nuclear 10 CFR 50.59 COMPANY Management Screening/Evaluation Page30f_ buru"Sww y.,.,,\IWlJ-Procedure Plant: Farley 0 Hatch 0 Vogtle 0 Unit No. 10 20 Shared 0 Activity/Document No.: 10 CFR 50.59 Version No.: (Act.lDoc.

Initiating the Change) Activity/Document Version No.: Title: 4. DVes DNo DN/A Basis for Answer: 5. DVes DNo DN/A Basis for Answer: 6. DVes DNo DN/A Basis for Answer: 7.aDves D No DN/A Basis for Answer: 7.bDves DNo DN/A Basis for Answer: 8. DVes DNo DN/A Basis for Answer: Does the proposed activity result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the Updated FSAR? Does the proposed activity create the possibility for an accident of a different type than any previously evaluated in the Updated FSAR? Does the proposed activity create the possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the Updated FSAR? Does the proposed activity have any impact on the integrity of the fuel cladding, reactor coolant pressure boundary, or containment? (Note: Answer Question 7b only if the answer to Question 7a is "VES.") Does the proposed activity result in a design basis limit for a fission product barrier as described in the Updated FSAR being exceeded or altered? Does the proposed activity result in a departure from a method of evaluation described in the Updated FSAR used in establishing the design bases or in the safety analyses? Provide a summary of the 10 CFR 50.59 evaluation in Section D. IF the answer to any of the questions in section C (excluding Question 7a) is "VES", a license amendment must be obtained from the NRC before the activity may be implemented. Do not complete section 0 if all aspects of the activity will be addressed in the license amendment request. D. 10 CFR 50.59 Evaluation Summary The 10 CFR 50.59 evaluation summary should include a brief description of the change and a concise summary of the responses to the evaluation questions provided in Section C. Summary: D Ves Check this box indicating a copy of the completed 10 CFR 50.59 screen/evaluation will be forwarded to Nuclear Licensing. NMP-AD-01 0-F01, Version 1.0 NMP-AD-010

20. G2.2.14 001 Which ONE of the following situations would require a piece of equipment to be added to the CAUTION Tag log? A. A normally locked open valve is tagged out in accordance with a clearance, the clearance is expected to be released on the following shift. B. Opening a normally closed valve in accordance with a troubleshooting work document that addresses promptly returning the valve to normal alignment.

C. Placing a safety related pump handswitch in PTL in accordance to a surveillance procedure and will be placed back to normal by the end of shift. A pump breaker is opened per SS direction due to a low oil level, the pump will be restored to normal on the next shift. Page: 38 of 49 6/6/2007 ( G2.2.14 Knowledge of the process for making configuration changes. KIA MATCH ANALYSIS Question gives a plausible scenario where the candidate has to determine which component will require an entry to be made into the CAUTION Tag Log. Question meets SRO only criteria by 10CFR55.43(b) item # 3 -Facility licensee procedures required to obtain authority for design and operating changes in the facility. Positions off normal condition require entry into logs such as OOPL and CAUTION Tag Log and this requires SRO authorization. Also, G2.2.14 in KA Catalog only has an importance factor of 2.1 for the RO level and an importance factor of 3.0 for the SRO level. Therefore, this is an SRO only level question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible, however configuration would be controlled by the clearance and would not require a CAUTION Tag Log entry. B. Incorrect. Plausible, not required since procedurally driven by the MWO with specific controls in place to return the equipment to normal configuration. C. Incorrect. Plausible, not required since the surveillance procedure drives this and the component will be placed back in normal by the end of shift. D. Correct. A component is added to the CAUTION TAG Log if it will not be restored prior to the end of shift. REFERENCES 10000-C, "Conduct of Operations" section 4.6 "Tracking Out of Position Components". Vogtle May 2006 SRO Retake Exam question # 18 Watts Bar July 2004 NRC SRO Exam question # 21. VEGP learning objectives: LO-LP-63500-03, Describe General work practices associated with Conduct of Operations.

i. Equipment return to service. Page: 39 of 49 6/6/2007 Number Text LO-LP-63404-01 State the purpose of the surveillance test program. 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. fallure of surveillance tests g.duties and responsibilities of surveillance test performer if a test falls or cannot be completed within the specified time LO-LP-63404-03 Describe the SM responsiblies as applicable to the surveillance test program. (SRO only) LO-LP-63404-04 With regards to implementation of the surveillance test program, describe the Shift Supervisor's responsibility for the following: (SRO only) a. prior authorization of performance tests b. actions for surveillance test failures c. operational mode change surveillance requirement LO-LP-63500-01 Briefly describe the reactor operator's overall responsibility.

Include a statement of the duties and responsibility concerning the following: a. direction of balance of plant operator (BOP) b.DELETED

c. plant operations with regards to Tech Specs and procedures

d. reportable notifications/unusual conditions

e. load dispatcher requests f. emergencies on site g. responsibility to shutdown the reactor LO-LP-63500-02 Briefly describe the BOP's overall responsibility.

Include a statement of the duties and responsibility conceming the following:

a. direction of plant equipment operators b.DELETED

c. plant operations with regards to Tech Specs and procedures

d. reportable notifications/unusual conditions

e. load dispatcher requests f. emergencies on site g. reactor operator relief LO-LP-63500-03 Describe general work practices for the following associated with the conduct of operations:

a. Abnormal indications

b. Instrument setpoints

c. Control room access d. Generator load changes e. Control room housekeeping

f. Manual operation of Motor Operated Valves g. Shift relief and evolution briefings

h. Reactor Trip Review I. Equipment return to service j. DELETED k. Procedure compliance I. Procedure implementation

m. System lineups and system status file n. Inaccessible component control o. Surveillance testing p. Operation of Dragon Needle Valves Friday,]uneOL2007 Page 97 of 165 Vogtle Electric Generating Plant A Procedure Number Rev 10000-C 70 Approved By C. S. Waldrup Page Number 55 of 70 Date Approved 2-15-2007 CONDUCT OF OPERATIONS

4.6 TRACKING

OUT OF POSITION COMPONENTS 4.6.1 When a plant component is manipulated to a position other than its normal alignment and its alignment is not controlled by existing administrative controls (e.g., procedure or clearance), then the component configuration will be tracked R V\ ("(5 as follows: ...&i O L a. If the component is intended to be restored to its normal position by the 01\ '( . end of the shift, it should be added to the Out of Position sublog (OOPL) of AutoLog by the SS/SSS. The text entry should contain the following information: (1 ) (2) Component tag number Component required position (3) Reason component removed from required position and resolution document, if applicable (e.g., MWO, CR, RER, etc.) (4) Name of person authorized to remove component from its required position LL b If the component is not intended to be restored to its normal position by IT WSW e/'LJ the end of the shift, it should be Caution Tagged and entered into the (f D \1 Caution Tag Log (figure 5), marked as an "Out of Position Component." c. If the component was added to the OOPL sublog per 4.6.1 a but will not be restored by the end of the shift, then it should be transferred to the Caution Tag Log per 4.6.1 b, and removed from the OOPL sublog with a notation stating that control was transferred to the Caution Tag Log. d. If the component was added to the OOPL per 4.6.1 a and is now ready to be returned to its normal position, then it will be removed from the OOPL sublog, with subsequent text entry specifying its return to normal position, by whom it was positioned, and IV performed if required. It is required that the OOPL sublog be cleared before the shift turns over. 4.6.2 Systems which are placed in an off normal configuration by a procedure and will remain in this configuration for greater than a shift should be Caution Tagged and entered into the Caution Tag Log, (figure 5) marked as an "Out of Position Component.". This also includes systems which are intended to be continuously in service, such as battery room fans, even if they are shutdown per the applicable SOP. This will ensure that their status is tracked and they are returned to service in a timely manner. Pnnted June 6. 2007 at 3:57 Vogtle Electric Generating Plant A Procedure Number Rev 10000-C 70 Approved By C. S. Waldrup Page Number 56 of 70 Date Approved 2-15-2007 CONDUCT OF OPERATIONS J 4.6.3 MWOs or surveillances that do not have specific procedure controls in place to "'/ maintain the configuration of a component should not be considered an 17 acceptable method for tracking the position of a component; they should be t _ k racked per this section. p 4.6.4 Tracking of Out of Position Components should not be used as a substitute for appropriate procedure changes or temporary modifications. The SS/SSS is responsible for ensuring that the listing of Out of Position Components is accurate and in their aggregate do not impact the operation of other systems, components, or equipment. 4.6.5 If the computer based programs are not available to track the Out of Position Components, then figure 5 should be used to document Out of Position Components until such time as availability is restored. At that time the Out of position Components should be transferred to the applicable tracking method. 4.6.6 The SS/SSS should ensure a Semi-Annual review of the Out of Position component in the Caution Tag Log is performed. The review should ensure that the components are still required to be out of position and the status of the resolution document. Pnnted June 6, 2007 at 3:57 ( 1. G2.2.14 001 Which ONE of the following situations would require a piece of equipment to be added to the Out of Position (OOPL) sublog of AutoLog ? A. Opening a normally closed valve in accordance with a troubleshooting work document that addresses returning the valve to normal alignment. B. Placing a safety related pump handswitch in PTL in accordance with a system operating procedure and will be placed back to normal by the end of shift. A pump breaker is opened in accordance with Shift Supervisor direction due to an inadvertent oil discharge and will be restored to normal position by the end of shift. D. A normally locked open valve will be repositioned in accordance with a clearance, release of the clearance is expected sometime on the following shift. G2.2.14 Knowledge of the process for making configuration changes. KIA MATCH ANALYSIS Question gives plausible scenarios for making an entry into the Out Of Position Sublog of AutoLog (OOPL). Candidate must determine which of the scenarios would require an entry to be made in the OOPL. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible, not required to be placed in OOPL since troubleshooting plan addresses returing component to normal status. B. Incorrect. Plausible, not required since procedurally driven and component will be placed back in normal by end of shift. C. Correct. Would require an OOPL entry since not procedurally driven. D. Incorrect. Plausible, however configuration would be controlled by the clearance and would not require an OOPL entry. REFERENCES 10000-C, "Conduct of Operations" section 4.6 "Tracking Out of Position Components" Watts Bar NRC SRO exam from July 2004 question # 21 VEGP learning objectives: LO-LP-63500-03 Describe general work practices associated with Conduct of Ops. 6/6/2007 G 2.2.14002 Which ONE of the following situations requires tracking via a Configuration Status Sheet? A'! A pump breaker is opened in accordance with Shift Manager direction due to inadvertent oil discharge. B. Repositioning a normally locked open valve in accordance with a system clearance. C. Opening a 125V DC control power breaker for a safety related pump in accordance with a system operating instruction. D. Opening a normally closed valve in accordance with a troubleshooting work document. The correct answer is A. A. Correct-unanticipated problems that require configuration changes must be tracked. The configuration Status Sheet would be included with the work order used for repairs. B. Incorrect -repositioning a valve in accordance with the system clearance procedure is allowable and the clearance process would provide for tracking component configuration. C. Incorrect -opening a breaker in accordance with the system operating procedure is allowable and the procedure would provide for tracking component configuration. D. Incorrect-component configuration changes are allowable during trouble shooting activities using a work document that contains general configuration control instruction.

REFERENCES:

3-0T-SPP1001 10CFR55.43.3/45.13 Knowledge of the process for making configuration changes RQ-NA SRO-96

Reference:

3-0T-SPPIOOI KlANumber: (12.214 K f A value: 3.0 Level: 2 Tier/Grp: Last Used: Source: NF.W SROOnly. YES W q t+J f;crrG 2, OcJ 4 SM ( 21. G2.3.6 001 Given the following: -The Shift Supervisor has declared A-RE-0014 Inoperable. -The radiation monitor will not come off the low end of scale. Which ONE of the following statements is CORRECT regarding approving a permit for the release of Waste Gas Decay Tank # 3? A. Approve. As long as the Shift Manager and the Chemistry Manager concurrently give permission to perform the release. B. Disapprove. The release CANNOT proceed because the discharge flow path cannot be aligned with A-RE-0014 failed offscale low. C. Disapprove. The release CANNOT proceed until A-RE-0014 has been returned to Operable status in accordance with ODCM requirements. Approve. As long as independent samples of tank contents are analyzed and the discharge valve alignment and release rate calculations are independently verified. Page: 40 of 49 6/6/2007 ( G2.3.6 Knowledge of the requirements for reviewing and approving release permits. KIA MATCH ANALYSIS Question gives a plausible scenario where A-RE-0014 in Inoperable and a release permit needs to be approved. The candidate must determine which requirement must be met to approve the release permit. Question meets 1 OCFR55.43(b) SRO criteria item # 4 -Radiation hazards that may arise during normal and abnormal situations, including maintenance actitivies and various contamination conditions. Question meets 10CFR55.43(b) SRO criteria item # 2 -Facility operating limits in Tech Specs and their bases. Question is also SRO only from KA Catalog since RO importance factor is only 2.1 and SRO importance factor is 3.1 for this topic. Therefore, question is SRO only. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible the candidate may think a higher level of authority from the two departments who normally approve releases is required. B. Incorrect. Plausible the candidate may think that the RE failure prevents opening RV-0014, this would be true on high failure, but not low. C. Incorrect. Plausible the candidate may think RE-0014 inoperable would prevent the release, ODCM actions allow. D. Correct. ODCM action # 45 requirements. REFERENCES ODCM Table 3-1 for Radioactive Gaseous Effluent Monitoring Instrumentation for ARE-0014, action # 45. Similar to Vogtle 2005 SRO Retake but for a different system and ODCM action. VEGP learning objectives: LO-PP-46101-10, Describe the major steps involved in releasing a gas decay tanks contents to the environment. LO-PP-46101-15, State the LCO, ODCM LCO, or TR, applicability, and anyone hour or less actions for the GWPS. Page: 41 of 49 6/6/2007 ( ( NO OBI TX OBI LO-PP-46101-0 List the parameters that will terminate oxygen I hydrogen flow to the recombiner. LO-PP-46101-0 Predict the consequences of RE-014 failing high or low on the Gaseous Radwaste System during a gas release. LO-PP-46101-0 Describe how the hydrogen recombiner prevents hydrogen and oxygen from reaching a flammable or explosive condition. . LO-PP-46101-0 Describe how VCT purge is established and maintained. LO-PP-46101-1 Describe the major steps involved in releasing a gas decay tanks contents to the environment. LO-PP-46101-1 State the events that require immediate termination of a gaseous release. LO-PP-46101-1 Describe why helium corrections must be performed on the waste gas analyzers. LO-PP-461 01-1 Describe how the waste gas system is used during a unit shutdown to degas the RCS. LO-PP-461 01-1 List the conditions that will cause PV-0115 to automatically close. LO-PP-461 01-1 State the LCO, ODCM LCO, or TR, applicability, and anyone hour or less actions for the GWPS LO-PP-47101-0 Describe how the FDT, WHUT, COT, LHST, and RHUT are normally processed. LO-PP-47101-0 State the normal and altemate discharge point for the Containment and Reactor Cavity sumps. LO-PP-47101-0 Briefly describe how a containment isolation signal affects the Containment Building Rad Drain System. LO-PP-47101-0 Describe the intended use of the CCW Drain Tank (CCWDT) and how the tank contents are handled. LO-PP-47101-0 Describe the preferred flow path and criteria to consider for draining of various plant systems to include: a. NSCW b. ACCW c. CCW d. Rad Systems e. Condensate

f. Fire Water LO-PP-47101-0 Describe the flow path from the Aux Building through the RPF for proceSSing of a Liquid waste stream. LO-PP-47101-0 State the Operability requirement for the HVAC and associated Radiation monitor for processing via the RPF. LO-PP-47101-0 Describe the major steps required for Operations to release a WMT. LO-PP-47101-0 State the conditions that require immediate termination of a Liquid waste release. LO-PP-47101-1 State the ODCM, TR, applicabilities, and anyone hour or less actions required for the Liquid Waste Processing System. Friday, June 01, 2007 Page 60 of 68 VEGPODCM ( Table 3-1. Radioactive Gaseous Effluent Monitoring Instrumentation OPERABILITY Requirements Minimum Instrument Channels OPERABLE Applicability ACTION 1. GASEOUS RADWASTE TREATMENT SYSTEM (Common) (M Noble Gas Activity Monitor, with Alarm and Automatic Termination of (\ Release (ARE-0014) 1 During releases a 45 b. Effluent System Flowrate Measuring Device (AFT-0014) 1 During releases a 46 2. Turbine Building Vent (Each Unit) a. Noble Gas Activity Monitor (RE-12839C) 1 During releases a 47 b. Iodine and Particulate Samplers (RE-12839A

& B) 1 During releases a 51 c. Flowrate Monitor (FT-12839 or FIS-12862)b 1 During releases a 46 d. Sampler Flowrate Monitor (FI-13211 ) 1 During releases a 46 3. Plant Vent (Each Unit) a. Noble Gas Activity Monitor (RE-12442C or RE-12444C) 1 At all times 47,48 b. Iodine Sampler/Monitor (RE-12442B or RE-12444B) 1 At all times 51 c. Particulate Sampler/Monitor (RE-12442A or RE-12444A) 1 At all times 51 d. Flowrate Monitor (FT-12442 or 12835) 1 At all times 46 e. Sampler Flowrate Monitor (FI-12442 or FI-12444) 1 At all times 46 4. Radwaste Processing Facility Vent (Common) a. Particulate Monitor (ARE-16980) 1 During releases a 51 a. "During releases" means "During radioactive releases via this pathway." b. During emergency filtration. VER23 ( VEGPODCM Table 3-1 (contd). Notation for Table 3-1 -ACTION Statements ACTION 45 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment for up to 14 days provided that prior to initiating the release: a. At least two independent samples of the tank's contents are analyzed, and b. At least two technically qualified members of the Facility Staff independently verify the discharge line valving, and verify the release rate calculations. Otherwise, suspend release of radioactive effluents via this pathway. ACTION 46 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flowrate is estimated at least once per 4 hours. ACTION 47 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours. ACTION 48 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend containment purging of radioactive effluents via this pathway. ACTION 49 -(Not Used) ACTION 50 -(Not Used) ACTION 51 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment. 3-4 VER23

22. G2.3.8 001 A release of Waste Gas Decay Tank # 1 is in progress when the following occurs during the tank release. -RE-0013, Waste Gas Process Monitor fails and is declared Inoperable.

-Auxiliary Building Continuous Exhaust Unit # 1 has just tripped on low flow. -The inlet Oxygen analyzer on the recombiner panel is declared Inoperable. -Waste Gas Decay Tank # 2 pressure is noted to be lowering by the ABO. Which ONE of the conditions listed above would REQUIRE release termination? A. RE-0013 monitor Inoperable. B. Auxiliary Building Continuous Exhaust trip. Waste Gas Decay Tank # 2 pressure lowering. D. Inlet Oxygen analyzer for recombiner inoperable. Page: 42 of 49 6/6/2007 ( G2.3.8 Knowledge of the process for performing a planned gaseous radioactive release. KIA MATCH ANALYSIS Question gives a plausible scenario where a Waste Gas Decay Tank release is in progress with several malfunctions occurring. The candidate must determine which of the malfunctions would require termination of the release. Question meets 1 OCFR55.43(b) criteria item # 4 -Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. Question is also SRO only since RO importance factor is 2.3 for this KA #, the SRO importance factor for this KA # is 3.2. Therefore, this is an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. RE-0013 not required per ODCM for Waste Gas Decay Tank release. Plausible the candidate may consider this since it is a gaseous radiation monitor but would only indicate radiation during processing, not releases. Not required to terminate the release. B. Incorrect. Plausible the candidate may consider an Aux. Building Exhaust unit trip as requiring to terminate the release as it could affect Aux. Building Pressure. C. Correct. Procedure specifically states to stop release if more than 1 tank lowering. D. Incorrect. Plausible but no procedure requirement to terminate due to this. Candidate may consider possible 02 or H2 concentrations a reason to stop the release. REFERENCES SOP-13202-1/2, "Gaseous Releases"step 4.2.13 and preceeding CAUTION. Vogtle 2002 NRC SRo Exam question # 85 VEGP learning objectives: LO-PP-461 01-11, State the events that require immediate termination of a gaseous release. Page: 43 of 49 6/6/2007 NO OBI TX OBI LO-PP-46101-0 List the parameters that will terminate oxygen / hydrogen flow to the recombiner. LO-PP-46101-0 Predict the consequences of RE-014 falling high or low on the Gaseous Radwaste System during a gas release. LO-PP-46101-0 Describe how the hydrogen recombiner prevents hydrogen and oxygen from reaching a flammable or explosive condition. . LO-PP-46101-0 Describe how VCT purge is established and maintained. LO-PP-461 01-1 Describe the major steps involved in releasing a gas decay tanks contents to the environment. LO-PP-461 01-1 State the events that require immediate termination of a gaseous release. LO-PP-461 01-1 Describe why helium corrections must be performed on the waste gas analyzers. LO-PP-46101-1 Describe how the waste gas system is used during a unit shutdown to degas the RCS. LO-PP-461 01-1 List the conditions that will cause PV-0115 to automatically close. LO-PP-461 01-1 State the LCO, ODCM LCO, or TR, applicability, and anyone hour or less actions for the GWPS LO-PP-47101-0 Describe how the FDT, WHUT, COT, LHST, and RHUT are normally processed. LO-PP-47101-0 State the normal and alternate discharge point for the Containment and Reactor Cavity sumps. LO-PP-47101-0 Briefly describe how a containment Isolation signal affects the Containment Building Rad Drain System. LO-PP-47101-0 Describe the intended use of the CCW Drain Tank (CCWDT) and how the tank contents are handled. LO-PP-47101-0 Describe the preferred flow path and criteria to consider for draining of various plant systems to Include: a. NSCW b. ACCW c. CCW d. Rad Systems e. Condensate

f. Fire Water LO-PP-47101-0 Describe the flow path from the Aux Building through the RPF for processing of a liquid waste stream. LO-PP-47101-0 State the Operability requirement for the HVAC and associated Radiation monitor for processing via the RPF. LO-PP-47101-0 Describe the major steps required for Operations to reiease a WMT. LO-PP-47101-0 State the conditions that require immediate termination of a Liquid waste release. LO-PP-471 01-1 State the ODCM, TR, applicabilities, and anyone hour or less actions required for the Liquid Waste Processing System. Friday, June 01, 2007 Page 60 of 68 R G-z., 3 ( 8

1. G2.3.8 002 Page: 1 A release of Waste Gas Decay Tank #1 is in progress. Which one of the following would require that the release be terminated?

A. RE-0013, Waste Gas process monitor fails low and is declared inoperable by the USS B:' Waste Gas Decay Tank #2 pressure is lowering in conjunction with Gas Decay Tank #1 C. The inlet Oxygen analyzer on the recombiner panel is declared inoperable D. Auxiliary Building Continuous Exhaust Unit #1 trips on low flow Ref: VG Procedure 13201-1 Waste Gas Processing System "CONTACT the USS and VERIFY that all LCOs for the Oxygen and Hydrogen Analyzers for the Recombiner to be placed in service have been exited. If not, DO NOT proceed until LCOs have been exited unless the USS approves operation under the action statement." 6/6/2007 Approved By Procedure Number Rev J. O. Williams Vogtle Electric Generating Plant A 13202-1 17 Date Approved 2-13-2007 4.2.13 4.2.14 4.2.15 GASEOUS RELEASES Page Number CAUTION All Unit 1 and Unit 2 GOT,s and SOT , s should be monitored to verify that only the GOT or SOT being released is decreasing in pressure. IF a GOT or SOT NOT being released decreases in pressure, immediately stop the release and notify the 55. 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 D 90f31 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, stop the release and notify the 55. IF AFT-0014 is operable, monitor A-RI-0014 and release flow rate while adjusting A-HIC-0014 to obtain the desired release rate. D Pnnted June 6, 2007 at 4:29 c ( ( 23. G2.4.46 001 Given the following conditions: -Unit 1 at 100% power. -A reactor trip and SI occur due to a large LOCA, on the reactor trip, an LOSP occurs to both RAT 1A and RAT 1B. -Both DGs re-energize 4160 1E busses 1AA02 and 1BA03. The following alarms are noted by the RO during the Initial Operator Actions of E-O. -ACCW LO HDR PRESSURE -ACCW RPC 1 (2, 3, 4) CLR LO FLOW -ACCW RCP 1 (2, 3, 4) CLR OUTLET HI TEMP -ACCW RX COOLANT DRN TK HX LO FLOW -ACCW EXCESS L TON HX LO FLOW -ACCW RTN HDR FROM RCP LO FLOW Which ONE of the following is CORRECT regarding the alarms and the actions the SS should direct the crew to perform? A'!I EXPECTED alarms, direct the RO to start an ACCW pump per direction of E-O Initial Operator Actions. B. UNEXPECTED alarms, direct the RO to trip the RCPs, direct an operator to perform actions of 18022-C, "Loss of ACCW" in parallel with E-O. C. EXPECTED alarms, direct the RO to trip the RCPs, direct an operator to perform actions of 18022-C in parallel with E-O, if ACCW is restored, re-start the RCPs. D. UNEXPECTED alarms, take no action at this time, the EOP network will address the loss of ACCW pumps in later steps of E-O or upon transition to another EOP. Page: 44 of 49 6/6/2007 l G2.4.46 Ability to verify that the alarms are consistent with the plant conditions. KIA MATCH ANALYSIS Question gives a plausible scenario during a simultaneous LOSP I SI where the ACCW pumps aren't running due to being locked out by the SI sequence. The candidate must determine the correct course of action I procedure to use to address the condition. Question meets 1 OCFR55.43(b) criteria item # 5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency situations. ANSWER I DISTRACTOR ANALYSIS A. Correct. ACCW pumps would not be running on a simultaneous LOSP I SI as SI sequence prevails and locks out the pumps. Initial Operator Actions of E-O would have the RO notify the SS of the condition and request permission to start the ACCW pumps, if they can't be started, trip the RCPs. B. Incorrect. Plausible if the candidate does not realize E-O initial actions address restoring ACCW or recognize why the pumps are stopped. An ACCW pump should be started and the RCPs not tripped. C. Incorrect. Plausible that candidate may realize why the ACCW pumps are stopped but not realize E-O Initial Operator Actions addresses. Stopping RCPs to protect them is a logical choice on loss of ACCW and AOP could be run in parallel by another operator. Restart of RCPs when available is also directed in many EOPs. D. Incorrect. Plausible the candidate may not recognize why the pumps are stopped or that E-O Immediate Operator Actions will address the condition. It is plausible that E-O would address the situation later, other EOPs such as the LOCA procedure which is the most likely transition in this situation would address the loss of ACCW pumps. REFERENCES 19000-C, "E-O Reactor Trip or Safety Injection", RO Initial Operator Actions step # 11. 18022-C, "Loss of ACCW", in particular symptoms and entry conditions for low flow annunciators which would occur on a loss of ACCW. LO-PP-041 01, ACCW Power Point slides 14 -19, 14 & 15 included here. VEGP learning obiectives: LO-PP-04101-04, From memory describe the expected system response and operator corrective actions for each of the following:

c. SI followed by LOSP Page: 45 of 49 6/6/2007 c , NO OBI LO-PP-021 01-1 Using AOP 18028-C describe the plant response to a loss of instrument air: a. with the unit at full power b. Unit in mode 3 c. Unit i n mode 4 , 5, or 6. LO-PP-021 01-1 Discuss the operation of loading and unloading the rotary and reCiprocating air compressors. LO-PP-021 01-1 Discuss the swing air compressor discharge flow path and required support systems (cooling water, power supplies).

LO-PP-03101-0 From memory state the following for the River Water Make Up System: a. System loads b. Normal configuration with both units at 100% power c. System Interfaces LO-PP-03101-0 Describe how plant cooling tower basin level is controlled. LO-PP-03101-O Describe the starting sequence of a River Water pump. LO-PP-03101-0 Describe the starting sequence of the traveling screen system for all modes of operation. LO-PP-03101-0 Describe the restrictions to water usage from the Savannah River. LO-PP-03101 -0 Describe the purpose of the emergency air compressor at the river intake structure. LO-PP-04101-0 From memory state the following for the ACCW system: a. Heat loads b. Heat loads cooled by either unit's ACCW cooling system c. Where heat is rejected to d. The impact on ACCW due to a loss of one train of NSCW LO-PP-04101-0 Describe how to shut down the ACCW system. LO-PP-04101-O Describe how ACCW surge tank level and RE-1950 are used to determine source of in-leakage and when the in-leakage is I solated. LO-PP-04101-0 From memory describe the expected system response and operator corrective actions for each of the following: a.SI b. LOSP c. SI followed by LOSP d. LOSP followed by SI e. LOW-LOW surge tank level 1. Pump shaft shear/locked rotor g. Thermal barrier heat exchanger leak h. Two pumps running i. CVCS letdown heat exchanger leak j. Stuck open relief valve k. TIC-130 failure I. Loss of air to TV-130 FriJIay, June 01, 2007 Page 2 of 68 PROCEDURE NO. VEGP REVISION NO. 19000-C PAGE NO. 32 20 of 34 Sheet 3 of 4 RO INITIAL ACTIONS 9. o 10. OIl. Check ECCS flows: o a. BIT flow Ob. RCS pressure -LESS THAN 1625 PSIG o c. SI Pump flow o d. RCS pressure -LESS THAN 300 PSIG o e. RHR Pump flow Check ECCS Valve alignment -PROPER INJECTION LINEUP INDICATED ON MLBs Check ACCW Pumps -AT LEAST ONE RUNNING 1\

fl\lswev (fA [( d-' ,f-er () (( ;;; rec-+-eJ sr

010. 011. aJJr-ef)e)

/N -f= 1(0 vve&' h( ld5P o a. Align Valves using ATTACHMENT B. o b. Go to Step 10. o c. Align Valves using ATTACHMENT C. o d. Go to Step 10. o e. Align Valves using ATTACHMENT D. Align valves using ATTACHMENTS B, C and D as necessary. Try to start one ACCW Pump. (((: l( I-D o IF an ACCW Pump can NOT be started within 10 minutes of rL loss of ACCW, I THEN stop all RCPs. IF an ACCW Pump can NOT be started within 30 minutes of loss of ACCW, THEN close ACCW Containment isolation valves: 0-ACCW SPLY HDR ORC ISO VLV HV-1979 0-ACCW SPLY HDR IRC ISO VLV HV-1978 0-ACCW RTN HDR IRC ISO VLV HV-1974 0-ACCW RTN HDR ORC ISO VLV HV-1975 Approval Procedure No. Vogtie Electric Generating Plant A. 18022-C NUCLEAR OPERATIONS Revision No. Date 14 ( Unit COMMON Page No. 1 of 10 Abnormal Operating Procedures LOSS OF AUXILIARY COMPONENT COOLING WATER PURPOSE PRB REVIEW REOUIRED This procedure provides operator actions for stabilizing the plant following a loss of Auxiliary Component Cooling Water. SYMPTOMS

• Observed gross system leakage.
• Low ACCW surge tank level that can NOT be restored.
• Both ACCW heat exchangers unavailable.
• High temperature on any heat exchanger serviced by ACCW.
• ALB04-A02 ACCW LO HDR PRESS
• ALB04-A03(B03, C03, D03) ACCW RCP 1(2, 3, 4) CLR LOW FLOW
• ALB04-A04(B04, C04, D04) ACCW RCP 1(2, 3, 4) CLR OUTLET HI TEMP
• ALB04-B02 ACCW RX COOLANT DRN TK HX LO FLOW
• ALB04-C02 ACCW EXCESS LTDN HX LO FLOW PlC{Vt 7 i hle
• ALB04-D02 ACCW RTN HDR FROM RCP LO FLOW
• ALB04-A01 ACCW SURGE TK HI/LO LVL rIL.-MAJOR ACTIONS
• Diagnose and attempt to correct cause of loss of ACCW.
• Monitor RCP parameters.
• Shutdown or isolate components cooled by ACCW. (

1\ j PUMP AUTO STARTS] 2. Loss of Offsite Power on the train related bus a. This start is blocked anytime a S a f e ty Injection signal is present ;:0 \1 )--l t I) \ AUTO START LOGIC SEQ BLOCK AUTO START OPENS ON SI OR LOSP ICLOSES ON Low Pressure AUTO START -SEQ START MOMENTARY I I LOSP I' CLOSES I ON LOP I "" .... OPENS ON SI " PUMP START f) It:) o (\J CfJ "" r c 24. WE03EA2.1 001 Given the following conditions: -The crew is performing the steps of 1901 O-C, "Loss of Reactor or Secondary Coolant" which states "check if RCS cooldown and depressurization is required". -RCS pressure is 450 psig and relatively stable. -RHR flow is reading 0 gpm on both trains. Which ONE of the following procedures would be the CORRECT procedure to perform with the given conditions? A. Transition to 19011-C, "51 Termination". B. Continue with subsequent steps of 1901 O-C. C. Transition to 19111-C, "Loss of Emergency Coolant Recirculation". Transistion to 19012-C, "Post LOCA Cooldown and Depressurization". Page: 46 of 49 6/6/2007 WE03 Facility conditions and selection of appropriate procedures during abnormal and emergency operations. EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations. KIA MATCH ANALYSIS Question gives a plausible scenario where the crew is performing1901 O-C. A transition point has been reached and the candidate must determine the proper procedure transition to make. Question meets 1 OCFR55.43(b) criteria item # 5 -Assessment of facility conditons and selection of procedures during normal, abnormal, and emergency conditions. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible the candidate may think 19011-C would be appropriate since RCS pressure is stable. B. Incorrect. Plausible the candidate may think continuing with 19010-C is appropriate. C. Incorrect. Plausible the candidate may think a transition to 19111-C, Loss of Emergency Coolant Recirculation is appropriate due to no RHR flow. D. Correct. Transition to 19012-C, Post LOCA Cooldown and Depressurization is required per the 1901 O-C step. REFERENCES 19011-C, "SI Termination. 19012-C, "Post LOCA Cooldown and Depressurization". VEGP learning objectives: LO-LP-37111-08, Using EOP 19010 as a guide, briefly describe how each step is accom plished. Page: 47 of 49 6/6/2007 c , ) l Number Text LO-LP-37071-0S State the intent of EOP 19241, Response to Imminent Pressurized Thermal Shock Condition. LO-LP-37071-06 Using EOP 19241 as a guide, briefly describe how each step is accomplished. LO-LP-37071-07 Given a NOTE or CAUTION statement from the EOP, state the bases for that NOTE or CAUTION statement. LO-LP-37101-0B Given a NOTE or CAUTION from 191 01-C, state the bases LO-LP-37111-01 State the physical bases for establishing equilibrium temperature and pressure in the RCS. LO-LP-37111-02 State the effect of various size breaks on the Primary System with respect to temperatures and pressures. LO-LP-37111-03 State the four characteristic stages of a large break LOCA. LO-LP-37111-04 State why a small break LOCA is a concem for propagation of cracks in the reactor vessel. LO-LP-37111-0S State how a LOCA is initially detected. State how the proper procedure is entered. LO-LP-37111-06 State the RCP trip criteria. Tell why it is especially important in the case of a small break LOCA. LO-LP-37111-07 State the bases for 'Loss of Reactor or Secondary Coolant' procedure. LO-LP-37111-0B Using EOP 19010 as a guide, briefly describe how each step is accomplished. LO-LP-37111-09 Given a NOTE or CAUTION statement from the EOP, state the bases for that NOTE or CAUTION statement. LO-LP-37111-10 Given a scenario requiring the use of the foldout page, state the actions that the operator would be required to take. LO-LP-37112-01 Using EOP 19012 as a guide, briefly describe how each step is accomplished. LO-LP-37112-02 State when post-LOCA cooldown and depressurization would be used. Page 34 of 165 c ( c PROCEDURE NO. VEGP 19010-C REVISION NO. ACTION/EXPECTED RESPONSE 21. Evaluate plant equipment:

22. o a. Secure unnecessary plant equipment.

b. Within 8 hours of SI actuation, isolate NSCW Corrosion Monitor Racks: o -Close 1202-U4-179 o -Close 1202-U4-180 (located in NSCTs on NSCW return header) o c. Repair or make available inoperable equipment which may be required.

d. Consult TSC for additional equipment to be started or actions to be taken to assist in recovery including: O-0-H2 Monitors CRDM Fans Within 5 days, initiate Containment inspection/cleanup if Containment Spray actuated and was terminated prior to recirculation.

Check if RCS cooldown and depressurization is required: RCS pressure -GREATER THAN 300 PSIG 30 PAGE NO. 14 of 21 RESPONSE NOT OBTAINED o a. { '5/vce tVb f(ow O (r (I ( {Ci<A. ,,)hle-C to { k1vt t-IF RHR Pump flow is? greater than 500 gpm THEN go to Step 23. Go to 19012-C, ES-1.2 POST-LOCA COOLDOWN AND IL DEPRESSURIZATION. \. fl lV5 Vefl.., 1(0 q ( 25. WE08G2.1.14 001 Given the following plant conditions: -A LOCA has just occurred at 0742 hours. -RHR flow is approximately 4500 gpm per train. -RCS lowest Cold leg temperature is 282 degrees F and stable. The Shift Manager (SM) makes an emergency classification at 0750 hours. Which ONE of the following is CORRECT regarding the emergency classification and required notifications? A. The SM was required to declare an NOUE and was expected to notify plant personnel by 0747 hours. The SM was required to declare an ALERT and was expected to notify plant personnel by 0755 hours. C. The SM was required to declare an NOUE and was expected to notify plant personnel by 0757 hours. D. The SM was required to declare an ALERT and was expected to notify plant personnel by 0805 hours. WE OS RCS Overcooling -PTS. G2.1.14 Knowledge of system status criteria which require the notification of plant personnel. KIA MATCH ANALYSIS Question gives a plausble scenarion where an RCS LOCA has occurred resulting in two differeent reasons the RCS Barrier is challenged (Potential Loss) and requires an emergency declaration. The candidate must determine if an NOUE is required or an ALERT emergency and the time frame for notifying plant personnel. Questions meets SRO only threshold by requirement of a classification in addition to the notification requirement. Per the KA Catalog Knowledge of Emergency action level thresholds and classification is an SRO importance factor of 4.1 with the RO importance factor level being only a 2.3. Therefore, this is an SRO only level question. Vogtle also has specific objectives for classification which are SRO only objectives. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible the candidate may not recognize the threshold for RCS Integrity Page: 48 of 49 6/6/2007 c barrier challenged yet due to Cold Leg Temperature> 265 degrees F. However, with RHR flow the break is greater than the capacity of one charging pump in the normal mode of operation. Also, plausible the candidate may think the time for plant personnel notification is 5 minutes from the time of the event. B. Correct. Alert emergency required and time is 5 minutes from declaration. C. Incorrect. Plausible the candidate may not recognize the threshold for RCS Integrity barrier challenged yet due to Cold Leg Temperature> 265 degrees F. However, with RHR flow the break is greater than the capacity of one charging pump in the normal mode of operation. Also, plausible the candidate may think the time for plant personnel notification is 15 minutes from the time of the event by confusing the time limit with that of notification of state and local authorities. D. Incorrect. Plausible the candidate may the threshold for RCS Integrity barrier challenged with RHR flow the break is greater than the capacity of one charging pump in the normal mode of operation. Also, plausible the candidate may think the time for plant personnel notification is 15 minutes from the time of the event by confusing the time limit with that of notification of state and local authorities. REFERENCES 91 001-C, Emergency Classification and Implementation Instructions. Vogtle May 2005 NRC SRO Exam question # 94 used as base for modification. VEGP learning objectives: LO-LP-401 01-02, State the four emergency classifications in order of severity from least to worst. Page: 49 of 49 6/6/2007 ( ( Number Text LO-LP-40101-01 Name the key individual responsible for the implementation of the EPIPs. LO-LP-40101-02 State the four emergency classifications in order of severity from least to worst. LO-LP-40101-04 Describe the basic function of the TSC, OSC, EOF, Recovery Organization, and HP and Chemistry groups during a declared emergency. LO-LP-40101-05 List the various radiological emergency teams (RETs), describe their function, and state who deploys them (91202-C). (SRO only) LO-LP-40101-06 State who fills the initial ED position when the primary is not on-site. LO-LP-40101-07 State who the primaries and altemates are for the ED position. (SRO only) LO-LP-40101-08 State from memory ED duties that cannot be delegated (SRO only). LO-LP-40101-09 State who has authority to approve modifications to EPIPs during an emergency situation (91002-C). LO-LP-401 01-1 0 List the three fission product barriers that are part of the criteria for clasSifying an emergency. LO-LP-40101-11 Describe how the status of fission product barrier integrity is obtained. LO-LP-40101-13 Given an emergency scenario, and the procedure, classify the emergency (SRO only). LO-LP-40101-15 State the individual responsible for making emergency notifications. LO-LP-40101-16 List the state and federal authorities that are notified in an emergency. LO-LP-40101-17 State the allotted time to contact the a. NRC b. State and local authorities LO-LP-40101-18 Describe when follow-up messages are required for state and local officials. LO-LP-40101-19 Describe when follow-up messages are required for the NRC. LO-LP-40101-20 Describe the communication system including power supplies, for notifing state and local officials. LO-LP-40101-21 Describe the communication system for notifying federal (NRC) officials. LO-LP-40101-22 List the backup communications for notifying state and local authorities. LO-LP-40101-23 List the backup communications for notifying federal officials. LO-LP-40101-24 State the circumstances requiring site dismissal with or without monitoring (SRO only). LO-LP-40101-25 State the individual responsible for conducting the site dismissal (91403-C). (SRO ONLY) LO-LP-40101-26 State the off-site reception center for VEGP personnel. LO-LP-40101-27 State the circumstances requiring assembly of VEGP non-essential personnel. (SRO ONLY) LO-LP-40101-28 State the group responsible for personnel accountability (91401-C). (SRO only) Friday, June 01, 2007 Page 47 of 165 ( c .---_. __ ..... __ .......... . n d1 I"\. -v-Alee., \(f) Vtif:::. Vogtle Nuclear Plant 2005-301 SRO Inital Exam 1 94. G2. 1. 14 Al_!!-R:.VISICIA __ __ ... _ While performing an emergency downpower, all annunciators in the Unit 1 Control Room are unexpectedly lost at 0900 hours and the SS makes an emergency classification at 0910 hours. Which ONE of the following describes the emergency classification and required notifications? A. The 5S was required to declare a NOUE and was expected to notify plant personnel by 0905. B. The S5 was required to declare a NOUE and was expected to notify plant personnel by 0915. C. The SS was required to declare an Alert and was expected to notify plant personnel: i DI' The SS was required to declare an Alert and was expected to notify plant personnel I by 0915 . .... _-.... _--_ ..... _ .... --_ ... , ---_ ..... _ ..... _ ..... ---""--KIA G2.1.14 Knowledge of system status criteria which require the notification of plant personnel. KIA MATCH ANAL VSIS ED is expected to notify plant personnel within 5 minutes of declaring an Alert or higher. The system status portion of the KIA is met by giving them a total loss of annunciators. ANSWER / DISTRACTOR ANAL VSIS A. Incorrect. Unplanned loss of annunciators places plant in automatic Alert. B. Incorrect. Unplanned loss of annunciators places plant in automatic Alert. C. Incorrect. Expectation is within 5 minutes of declaring Alert or higher. D. Correct. Alert declared at 0910 and expectation is within 5 minutes of declaring Alert. AU distractors are plausible based on memory nature of items. V 0) It ;Ill. REFERENCES <J /"I';! l Cij" 1. LO-LP-40101-39-C , EPIP Overview, Rev. 39, 05/03/2004. V/O....J MfZC 5!ZO 2. 91001-C, Emergency Classification and Implementing Instructions, Rev. 20.1, 09/1212000. Approved By I: J.D. Williams Vogtle Electric Generating Plant A Procedure Number Rev 91002-C 48 Date Approved Page Number ( ! 01123/2007 EMERGENCY NOTIFICATIONS 7 of 22 Sheet 1 of2 Reference Use ( CHECKLIST 1 PLANT PAGE ANNOUNCEMENT CHECKLIST (RADIOLOGICAL EMERGENCY) (1985304608 ) NOTES

• If the declared emergency involves an actual or credible imminent threat of attack on the plant by a hostile force, then go to Sheet 2 of 2 of Checklist

1.

• The completion of an initial plant page announcement that activated the ERO is expected to be completed within 5 minutes of the declaration of an Alert or higher. All subsequent upgrade announcements should be completed as soon as possible 1. Make an announcement with the plant page public address system: 0 2. 3. 4. 5. [select one] a. "ATTENTION ALL PERSONNEL

-THIS IS A (DRILL / ACTUAL EMERGENCY) [select one] (A NOTIFICATION OF UNUSUAL EVENT) (AN ALERT EMERGENCY) [select one] (A SITE AREA EMERGENCY) (A GENERAL EMERGENCY) HAS BEEN DECLARED FOR (Unit 11 Unit 21 THE SITE) [event description] __ _ NOTES

• For Alert declarations or higher, complete b. and c. as applicable.
• Parts b. and c. below are not required if assembly and accountability are complete.

b. "PERSONNEL WORKING (ONIIN) CALL CONTROL ROOM AT EXTENSION

& CONTINUE WORK." " c. "EMERGENCY RESPONSE PERSONNEL REPORT TO YOUR EMERGENCY RESPONSE FACILITY. NON EMERGENCY RESPONSE PERSONNEL, CONTRACTORS AND VISITORS EXIT THE PROTECTED AREA AND REPORT TO THE ADMINISTRATION BUILDING." [select one] THIS IS A (DRILL I ACTUAL EMERGENCY)" Sound the appropriate tone for 15 seconds: NOUE -None ALERT-Warble SITE AREA -Warble 0 GENERAL-Warble REPEAT above announcement(s). For an Alert or higher, repeat items 1 & 2, one (1) more time in about 10 minutes. (except for part 1.b. which should be repeated only for those affected personnel or locations who have not yet called the Control Room). Indicate the time the announcements are made. a. Initial Page Announcement Time: _____ _ b. Repeat Page Announcement Time: _____ _ o o o o Pnnted June 6, 2007 at 4:58 -. r ?, cp> ..... r----. k i n Approved By * . ... ---.. T.E. Tvnan Vogtle Electric Generating Plant A: ---_. I Procedure Number Rev i EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS

1. RCS Leak in progress AND RCS Subcooling is Less Than 24°F OR [38°F ADVERSE] 2. Indication that a SG is ruptured AND it has a NON-Isolable OR I LOSS Secondary Line Break Outside Containment

.. of RCS 3. Indication of a SGTR AND a Prolonged Release of Secondary OR I Barrier Coolant is occurring from the AFFECTED SG to the Environment

4. Containment Radiation Monitors RE-005 I 006 > 2.0E+4 mr/hr OR --------------------------------------------

HEAT SINK IV OR CSFST F RED RCS INTEGRITY I V OR I

• CSFST RED 1. NON-Isolable RCS leak (including SG tube leakage) GREATER OR THAN the Capacity of One Charging Pump in the normal charging mode--Potential Loss of RCS Barrier --------------------------------------------

JUDGMENT: Opinion of the ED that the RCS Barrier is Lost or Potentially Lost OR 91001-C 2S Page Number 8 of 11 I -OR 1-the inability to determine the status of the RCS Barrier ' FIGURE 2 -REACTOR COOLANT SYSTEM (RCS) INTEGRITY Printed June 6, 2007 at 4:59 ):::) 1') \) c (() --. --+= -", Approved By T.E. Date Approved 06/30/2006 G E N E R A L S I T E A R E A A L E R T N o U E ELECTRICAL POWER ModI;, LOM of AC,.,., to BOTH AN>>. AI<<) 8A03 MIll EITHER-ReUnlonof.t ..... bw HOT IUIy "'INn *... ot.". 01 bu, OR BARRER dMInrinecI

• -' Fila.,. Ptoducta.m.

EAL*.<F .... wat. -" ...... lftilofAC,.,.,on BOTH IW¥l AltO BIIOS 1Df ,. 'I mn. ModH1-4: l/npWIned ... aI.,... on All YhII DC tN.. (ADt. BOt. mt,MdDD1)b,.161rin. Mon.1-4: louof ...... onEnHEF ANJ2ORBA03 ... "'Im.MJ!J. ... h ..... Mdwp poMf"'" .wdabAe. LoesofAC power on BO," ANJ.2 AND BA03 b > ....... Lou ofQJ...ehs powIII' to BOTH AN12 AHD8A03_",I""'C_busle oonneded * ., ___ 0fI..Ms M/Il80lli ANn. ANOIIAIXS _ p-.d by DIMeI o.-.tDn ..,..W: on All WIll DC b.-.(ADt. B01 , 001 , ... DDt). >>15 "*'. (J-------., Number Rev ' Vogtle Electric Generating Plant ..:.. 91001-C 25 EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS of 11 RADIOACT1VITY Valki r..tIng on RE-12,," ht I,,. 3CIE+O,.cIIccfor>>16I'1'M.QlIIhiI value to be-=-ded for,. 15 ..... , (Not.') V.ad ** bound8ry __ .-....nt ofl ...... ... :. 1000 rrnm TEDE m."l5OOOrsnmCOE IhyroidQl:hIcI ....,., r-'ta1nclcD ,. 1000 nnmltv ..... gfIh1c1......,. ..... nIb. lhyRjd ... all5OOO_CDEb 1 hrall --V ... ,.,....onRE

• 1: ..........

o-3E+OpQteIor

.'6"*'.28 on RE-1213t ..... ,. &E+211 dIoc b .,61rin..Q8 .......... .. >>'6_ _'I vaw .... boundary do..-.-nt ot ........ IndIca ... "'OOrrnm TEOE m.,. 500 nnm me ."... QB Wd....,,.....incSa .. "'OO ........,.. hllaupedld to oontnu. ."",twQff hIcI...v..,..,... "'l¥okIdoM 011500 rrnm COEb1hrofinhaidon VaIId...slng on RE-12 ...... htll "2E-1 pdlc:cfor.,.m . .QI!on RE-12I3O .. , I." e:E+O,.c¥cc tor ,. 16 "*'. gs on RE.o18 .. , I. "'1E-1,.c11ccb""I5,,*,*C'*-'} ConfttrMd ....... .,."... tor gMeOUI "2ODX ClC)CM ..... b" 'I "*'-RfICIdon ..... hlclhwhn-'In Fwi HardIng<< ConIiIInnwIlBldv MID ........ <1 ........... r.v.IlnleluellncJ cnty ,...,. ... pooI. .... ....,., ctNI .. , ..... lnunr:lO'feltng ImIctiatIlCIbl(1Wb2) 0..".. n.dIa ...... vdd high al8rmon one Of rncnof .. foIooMng: RE-008 (2.1 mrlhr), RE-2S32J2533 AlB (1SE-4 flCl/oc). OR durlngModeO ... .....,., .... 1hr) Vddr.dtav.l.",l5l11'lhrlnh Cor*aI ....., ........ ( ** CI-. bcaI dwgtng.-.on) Valid tMdIncI 0fI RE*12444112442 ... le" 2£4 tdIcc-", eo .. Si!8on RE-t2I3D ..... "IE-l tdlccb > eo min. DBonRE'Ot ...... >1IE03

---

---

Cot6m.d .. ,...,.". mllquld ....... 1tIdIcatM

....... IIonsOf,... .. ,..."ZX ODa.Ilnfttfor",eo,,*,. AMidon .. vee. ..... nDmailn tim unooMoOed __ .. .,.. decrN. 1n"""lnoca¥lty,If*II .. lpooI , Of fuM hnsfeJ canal, BUT d n.dIatsd fueI ** -ean,....... co-.d will Valid .... r.d monitor rNdIngt 1ncrM .. lOOOX_ nonneI r.v.I .. (2) PLANT SYSTEMS .... .. ... ,.. .......... MJIJ. .... .. dldNOT NOT _-. .. cot*oI ..... t-4: SIgr6anI: ........ 1s In ,.......DI'I'ICItIlDf ... hn.-nt_noI avaIabIa MID ESF conRI board AND 1ndIca ... _NOT.vaIlabIe ..... ,-3: Aa.ItomIlIcl'Mdor.,. ..... _. exOMMd lJJDan autDnIdc,.....,.did NOT oc:cu tlllJallUOCMafu1l'11MUa1.,.OCICU'T.cI hm ... eot*oI rooM Unp6anMcIlouolrNlel.Of .. amunc:IatDn .. 6ndIcatDn In h conerofroomb,.......,..,.... IIIR EJTHEA

• eIpIIcaR( hnUstIl_ .. 1ndIcdons_

NOT -. Modes 4-6: Idomdc INC_'" .................. .4l!I2M MJlDmdc fUdot.., cId NOT occur IJIR. --U ___ .,.oacurq,d from"'t.ClnhltoOAI. MocIM 1-4: ............ 0I'1nctIc:Mon In"" LoN of AU of .. faDcMtng On-SI .. WephoM. GaI __ .. Sound LoN of ALL ofh foIbMng 0tJ.SI .. oormuallons"'-"": ENN and T ..... NATURAL PHENOMENON Tom.do pIInt .... ,..gs .....wn.dhurrlc:.lrMforoellllnd.oI 100ft1lhOffll"N .... on ..... by_ .............. (16mnM.--.). ..... rndDring 'f'ISm oonItmI ......-:: ....... ", 0.12 g. Re,oftoflomecto""'*"' .... ..-.-Huntc8ne ...... of 74 ..... fcncnt by'" NdoMt W ..... .,...(NWS-CoILwDIII !he nut feu hcua. ...... ... r.port ............. .....,.. HAZAROS CoMroI AMlmftKUdDn ... .... inIIIIItsdMI/J.ooNraIof ... ..... CNOK>T be .......... hm,.,.,. Ih*'-n pensts...., tl "*'. AInnt cr.lhcor6msd .ftectpCant ____________ _ Rsport Of cWsctIon oftoxic,lIiImIMbte , ga ... wHt*t. r.c.tr -..:u.1n concenhIon. t.t ....... ....:udon

01. room Of ... neecMd

___ _ ...... ....... 1JIJIJ..a.dad ....... .... ....... .......... ,.,..,."......, ,...... ...... Of..tscy I'IIMsd ........ ........ .... ------------------- Rsport: of .... 1IlrucUW ctan.gs (CMtMCI by ........ ........... abIIty .......,Udon DNfY of .. "", .,..,..AuceuNs: Contoimw>I ............... """""' ........ Fu.I HandlIng BulkIng D_.I o.n.nklf BuildIng Condenu .. 8kng11 Tri --NSCW CoolIngTOWIW RWSr c;;;. R;;.-.;;:';;' ....... Aftnft CIMh '*-" daINoe D...., N6etsd ..... 1IlNdurn Of sysIIImswllhlnhprotlc:tilcl .... -------------------otdMsdlonoflDldc."""""', ...... ........... ... t'*'d* __ Reportof ..... ea ... Flraln ........ oonIguousOf D.'III .......... .1Nn 11 min. of control room nod-Unenlclpstld upIosIon .... prolisdlsd .... ,..-ana In vi .... an.oe D perrnsnent strucUN Of SECURITY A HOSTilE FORCE t.. _ken-*Ol ofptMt.........., ......... 'plant per1JOfftII ........... operMIi D....,....,.. .. ty ........ s.c..tty ....... it .... WAlMEA PsAa.ct (l.,.,) ,)AfICIIIcdlJnhlfth ... exPoeM.1SaICt, aatInw '"'*" Of ohf HOSTD..E ACT10N l.occurrlngorh .. OCICU'T.cI.1hIn .. PROTECTED AREA , s.cunty e--.t In. PWIt PROTECTED NOV. NoeIcdon of ., AIrbatM Abc* *..* ) 1) A wIdatecI noIIcdon..,.., NRC of .,ebIhw ................. 30 -..., HoIkaIonofHOSTI.E ACTK)H ..... .. OWNER COHTROU.ED Nt£A (I ..... ) 1)A noIblonhmh ... sec:uIty .... 1hat .... .... , ..... ACTtoN I.CICCI..W'Ytng or hu oc:curred ...... t. OWNER CONTROllED AREA Cot6m.d 8ecwtty EWMI.tIk:h -..01.....,.of ......... O .... J 1. AcndIb6e ... ..-ctIc ""*"" --2.AnldUsdnolllcdofthmNRC -- PRDTECTED MEA (NeIIa.t) ....... Dtwm,.,-lOf AREA. N;tibiOn b; tU;. oISdat. ofpotental.., ..... 1on of III .. persomsI besed on en 0fJ..Us ...... BARRIERS LOM of THREE e.m.n LonofNfYTWOberrientlm Po .. nIaILonoft. ,"IRD Mmer YocIn 1-2; SubcrlIcaIIty CSFST .. RED &Q.ErrHER Core Cooing OR .... , SInk CSfST .. REO '--of TWO ........ lon ofDHE benW MR. PotsnIaI LON of. SECX>>m banIer P<*nCWlOMofBOTH F,,*CIed AJIlJ.RCSburieq Modes 1-4: MJ/1.Subc:riIcaIty CSfST I. RED Lo. ** Potanllalloss of F,,* a.d ...... LON Of PoWIIaILon ofRCS ..... Loa Of Lo. ofConWrwnent ...... ------------ .... _----RCSa....y.....,...inIII::etN "".......,.1-131 >1,.c¥gmtor " ......... orln_ofTechs,.c. .... ..... d'IIIly "'001'E JdfIm""o .. -. ------------------- RCSlJridenllecl .... "10gpm {l8 RCS P ..... boundary IN-. ", 10 gpm gs RCSIcf8nlhd ..... ,.26gpm SHUTDOWN SYSTEMS UodM W: lou ofreedDr __ ............. NlIt.MMIIty Ion of RHRCIOOIno .. * ...... "'JDP 180to-C MJIJ..,."OHE of .. v....I ...... ..... rad IfIOfI6Dn,REoOO2JD03v.1de.gh-..rrt (16 ndv) ........ hMdNlnDVed, OR CoreaH ........... " 711*F .. " ..... .... ,-4: .. offunclon ...... Of-ualn Hot ......... (fITliER Core CooIng OR .... t SIr*. CSF8T I. REO) MJIl no oe-.... t .... vaIabM. WodMi6oe: ooaIIno .6m.EITHEA RCS .....,..tw. I.,. mo-F OR RCS ..,...,. ... lalnctuling tn:OnhIed OTHER at. con4tor. ... ....,.In .. judgrtlsntoft. EMERGENCY DIRECTOR pc4enlalw can,..-..t.ty be....-,d D o:Md 1000 nnm TEDE Of 1000""", COE lh)lfOlddoMlevMouIIIdeh'" -..". gfAA.-l.""nI ........... po ...... ot.fGOndllons ... ",*"In .. EMERGENCY DltECTOR ...... _Jor ...... of .... ln:IoM r..ted ... protsc:Ion of .. pdIc. ot.c0n4 ........ wNc:fIln ... judgmsnC of .. EMERGENCY ORECTOR ....... plan!: ufIty .,...ms,be........, ...... ,tncne_ --oct.. oondIIIDM Gilt"" In OItE CT OR .,.....,.,.. ....... of..wy of ......... ......

• r..,nd operdng mode ..... Tem &p.c lCO gf TAM T .........

R..........-,..,nod III:IOn ---------------_ .. -----Mode. 1-4: Unoonnhd NOTE f: a.MIkaIonstlodd be be_ onOOCM orOlJ..-DoM '--.If'" monI_I'MdIng(.)I. _ kInget ..... h period NOT Of CAMrIOT be period , ..... dedMllon MUST be.,.. beNd on h..ld '*Ing. NOTE I: -..om.I.vee.. _ .. hlehMt reeding In .. &HI 24 hcMn prior 10" -vsncJ', ecb$1ng'" cuNftI .... k ...... NOTE3: Figure 4 -EMERGENCY CLASSIFICATION LEVEL DETERMINATION Printed June 6, 2007 at 4: 59}}