ML14338A059

301 Draft SRO Written Exam

ML14338A059
Person / Time
Site:	Vogtle
Issue date:	12/03/2014
From:	NRC/RGN-II
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Download: ML14338A059 (242)

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Question Documentation Anatomy Prefix List:

A.0 - Question A.1 - Distractor Analysis A.2 - Misc Comments/Feedback/History A.3 - NUREG ES4015 A.4 - Original Question (if question was MODIFIED from a BANK question)

A.5 - Supplied Reference B.0 - Regulator Documents B.0 - Tech Specs B.1 - Tech Spec Bases B.2 - TRM B.3 - COLR B.4 - PTLR B.5 - ODCM B.6 - PLS B.7 - PTDB C.0 - Procedures C.1 - EOP C.2 - AOP C.3 - SOP C.4 - UOP C.5 - ARP C.6 - Admin C.9 - Other D.0 - Drawings D.1 - P&IDs D.2 - Oneline D.3 - Elementary D.4 - Logic D.5 Other E.0 - Misc Other E.1 - Photographs E.2 - Maps E.3 - Rad Surveys E.4 - Lesson Plan The bookmark sidebar will be very helpful in viewing this document as it will show the breakdown of each question more easily.

1. 001AA2.04 001/LOIT/SRO/C/A 4.2/4.3/001AA2.04/LO-TA-63013///

Initial conditions:

- Unit 1 is at 70% reactor power.

- Main Turbine load is stable.

- Rods are in automatic with CBD at 190 steps.

- Control rods start stepping out without demand.

Current conditions:

- Rod motion stops when the Rod Bank Selector Switch is placed in Manual.

- Reactor power indication has risen to 75%.

- CBD rods are at 211 steps, EXCEPT control rod H8, which is at 190 steps by DRPI indication.

With NO other actions taken, which one of the following completes the following statement?

When the inadvertent rod motion stops, reactor power indication will __(1)__,

and per the Bases of Tech Spec 3.1.4, "Rod Group Alignment Limits," CBD rod H8 is

__(2)__ at this time.

A. (1) stabilize and remain near 75%

(2) OPERABLE B. (1) stabilize and remain near 75%

(2) inoperable C. (1) trend down from the peak value observed to slightly above 70%

(2) OPERABLE D. (1) trend down from the peak value observed to slightly above 70%

(2) inoperable K/A 001 Continuous Rod Withdrawal AA2.04 Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

- Reactor power and its trend Thursday, February 20, 2014 8:28:29 AM 1

K/A MATCH ANALYSIS The question sets up a plausible scenario where rods withdrawal due to some internal failure inserts some amount of positive reactivity. Once operator action is taken rod motion stops and the candidate must address how the core will respond to the inserted reactivity and the impact of the stuck rod, therefore the two elements are in place that the KA requires. The OPERABILITY determination for the affect rod is SRO required knowledge.

EXPLANATION OF REQUIRED KNOWLEDGE As control rods are withdrawn, an increase in fission rate will occur. The resulting fuel centerline temperature increase from the positive reactivity addition will be offset by FTC and MTC feedback. As a result, reactor power will trend down toward the orginial power level with and increase RCS average temperature. Since RCS temperature is higher than the orginial temperature, SG pressure will also be slightly elevated. The higher SG pressure will result in slight increase in steam flow as compared to the original conditions. Therefore, reactor power will be slightly higher than the pre-transient value. This is a fundamental reactor theory concept.

Per TS 3.1.4 Bases, the OPERABILITY requirements (i.e. trippability) are seperate from the alignment requirements. Even if a rod is >12 steps misaligned, it is still OPERABLE.

Where rod(s) are not moving, the rod(s) must be considered untrippable unless there is verification that a rod control system failure is preventing rod motion. Since rod motion was demanded and did not occur and there is nothing given in the stem to explain the loss of motion, the candidate must declare rod H8 inoperable.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Part 1 is 'plausible' however incorrect in that the candidate may determine that reactor power will rise to a new higher equilibrium value and stabilize. Thus, the candidate failed to take into count the reactivity feedback mechanisms of MTC and FTC on core behavior. In addition, the answer would be correct under some plant condition like during startup steam dumps in steam pressure mode.

Part 2 is 'plausible' but is also incorrect in that the candidate must consider the affected rod inoperable until proven trippable as stated in Tech Spec 3.1.4 Rod Group Alignment Limits bases; However, where rod(s) are not moving, the rod(s) must be considered untrippable unless there is verification that a rod control system failure is preventing rod motion. If the rod control system is demanding motion properly and no motion occurs, the rod is considered untrippable (i.e., inoperable). This operability call is contrary to the standard philosophy that a component is considered OPERABLE until determined otherwise.

B. Incorrect. Plausible. Part 1 is 'plausible' however incorrect. See Part 1 of choice A above.

Thursday, February 20, 2014 8:28:29 AM 2

Part 2 is correct requiring the candidate to recall Tech Spec 3.1.4 Rod Group Alignment Limits bases which ties OPERABILITY to trippability of the rods. The rod OPERABILITY (i.e., trippability) requirement is satisfied provided that the rod will fully insert in the required rod drop time assumed in the safety analyses. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability. However, where rod(s) are not moving, the rod(s) must be considered untrippable unless there is verification that a rod control system failure is preventing rod motion. If the rod control system is demanding motion properly and no motion occurs, the rod is considered untrippable (i.e., inoperable).

C. Incorrect. Plausible. Part 1 is correct which is addressing core response and correctly predicts that the positive reactivity will be offset by FTC and MTC feedback. Over time the reactor power will lower to near original values, only slightly higher due to TAVG being raised which results in an increase in Steam Generator pressure. This would result in a small increase in steam flow.

Part 2 is 'plausible' but is also incorrect. See Part 2 of choice A above.

D. Correct. Part 1 is correct. See Part 1 of choice C above.

Part 2 is correct. See Part 2 of choice B above.

SRO JUSTIFICATION (10CFR43(b))

(2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No the question is not addressing any TS action times.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, the question is not addressing above-the-line TS information. The required knowledge is TS Bases.

-Can question be answered solely by knowing the TS Safety Limits? No, the question is not related to TS Safety Limits.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)

• Knowledge of TS bases that is required to analyze TS required actions and terminology. Yes. Specific knowledge of TS Bases is required to make the OPERABILITY call required and this call is contrary to the standard philosophy for OPERABILITY calls.

Thursday, February 20, 2014 8:28:29 AM 3

Level: SRO Tier # / Group # T1 / G2 K/A# 001AA2.04 Importance Rating: 4.2 / 4.3 Technical

Reference:

Tech Spec 3.1.4 Bases Rev 1-8/03 References provided: None Learning Objective: LO-LP-36990-07 State the Technical Specification bases for the restrictions on control rod insertion limit, and alignment (SRO only).

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: Whether any Tech Spec LCO's of section 3.1 are exceeded.The required actions for all section 3.1. LCO's.

LO-TA-63013 Implement Technical Specification LCO using 10008-C (SRO Only)

Question origin: MODIFIED - Vogtle HL18 Question #005AG2.1.07 001 Cognitive Level: C/A 10 CFR Part 55 Content: 41.1 / 43.2 Comments:

You have completed the test!

Thursday, February 20, 2014 8:28:29 AM 4

1. 005AG2.1.07 001/1/2/STUCK CRDM/C/A - 4.4/4.7/BANK-HL-17/HL-18 NRC/SRO/AML Initial conditions: Original Question

- Time = 0900.

- Unit 1 is at 60% power following a refueling outage.

- The OATC is withdrawing rods when one DRPI is seen not moving with its group.

- The OATC immediately stops withdrawing rods, and all rod motion stops.

- CBD, Group 2, Rod H-8 DRPI indicates 198 steps.

- CBD, Group 2, step counters indicate 209 steps.

Current conditions:

- Time = 0945.

- No rod motion has occurred since 0900.

- I&C has verified no faults on the DRPI system.

- I&C has verified that the rod lift coil for Control Rod H-8 is failed.

Which one of the following completes the below statements?

Based on the initial conditions, at 0900 Control Rod H-8 was __________ in accordance with the Bases of Tech Spec 3.1.4, Rod Group Alignment Limits.

Based on the current conditions, at 0945 Control Rod H-8 was __________ in accordance with the Bases of Tech Spec 3.1.4, Rod Group Alignment Limits.

Rod H-8 status Rod H-8 Status at 0900 at 0945 A. OPERABLE inoperable B. inoperable inoperable C. OPERABLE OPERABLE D. inoperable OPERABLE Monday, January 20, 2014 9:38:55 AM 1

Rod Group Alignment Limits B 3.1.4 BASES (continued)

LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment.

The rod OPERABILITY (i.e., trippability) requirement is satisfied provided that the rod will fully insert in the required rod drop time assumed in the safety analyses. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability. However, where rod(s) are not moving, the rod(s) must be considered untrippable unless there is verification that a rod control system failure is preventing rod motion. If the rod control system is demanding motion properly and no motion occurs, the rod is considered untrippable (i.e.,

inoperable).

The requirement to maintain the rod alignment to within plus or minus 12 steps of their group step counter demand position is conservative. The safety analysis assumes a total misalignment from fully withdrawn to fully inserted. When required, movable incore detectors may be used to determine rod position and verify the rod alignment requirement of this LCO is met.

Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which a self-sustaining chain reaction (Keff t 1) occurs, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are fully inserted and the reactor is shut down, with no self-sustaining chain reaction. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.

(continued)

Vogtle Units 1 and 2 B 3.1.4-5 Rev. 1-8/03

1. 001G2.1.19 001/LOIT/SRO/C/A 3.9/3.8/001G2.1.19/LO-TA-05002///G2.1.37 Initial conditions:

- Unit 1 reactor power lowered due to an inadvertent turbine runback.

- ALB10-D04 ROD BANK LO-LO LIMIT alarm was received.

Current conditions:

- Main Turbine load has been stabilized.

- RCS Tavg is 2°F below Tref.

- The OATC has requested a 2-step rod withdrawal for temperature control.

Which one of the following completes the following statement?

Based on the current conditions and using the Plant Computer data provided, ALB10-D04 __(1)__ valid, and based on the current conditions and NMP-OS-001, "Reactivity Management Program,"

guidance, the Shift Supervisor is __(2)__ to authorize a 2-step control rod withdrawal for Tavg control.

REFERENCE PROVIDED

__(1)__ __(2)__

A. is allowed B. is NOT allowed C. is NOT allowed D. is NOT NOT allowed K/A 001 Control Rod Drive G2.1.19 Ability to use plant computers to evaluate system or component status.

K/A MATCH ANALYSIS The question test the candidate ability to evaluate and validate a plant computer (IPC) generated alarm for the current plant conditions associated with the Rod Control system. Then the SRO is required to determine if control rod withdrawal is both prudent and allowed under the stated conditions in the stem per the NMP-OS-001 Reactivity Management Program.

Thursday, February 20, 2014 8:45:05 AM 1

EXPLANATION OF REQUIRED KNOWLEDGE The IPC calculates the rod insertion limit using real time data. When the alarm is generated the candidate would be expected to verify the validity of the alarm. This evaluation would be accomplished using IPC and Tech Spec data with backup from QMCB indications. In addition, the Shift Supervisor is required the make a decision to withdrawal control rods based on NMP-OS-001 Reactivity Management Program guidance.

The control room team shall not immediately dilute or withdraw control rods in an attempt to restore RCS Tavg/Tref deviations caused by a secondary plant transient.

Attempts to immediately restore RCS Tavg/Tref deviations caused by a secondary plant transient can be aggravated by withdrawing control rods or reducing boron concentration with reactor power rising. Per NMP-OS-001, once turbine load has been stabilized and RCS Tavg has been restored to within 3°F of Tref, positive reactivity can be added by withdrawing control rods.

ANSWER / DISTRACTOR ANALYSIS A. Correct. The first part is correct. Per the IPC printout, RCS DeltaT Power is approximately 83% and CBD position is 116 steps.

The IPC uses Auctioneered High DeltaT Power for the RIL calculation. Per the COLR, the RIL for 83% power is 122 steps on CBD. Therefore, ALB10-D04 is a valid alarm.

The second part is correct. Per NMP-OS-001 step 6.1.2.4, once turbine load has been stabilized and RCS Tavg has been restored to within 3° of Tref, positive reactivity can be added by withdrawing control rods. Both of these conditions have been met. Therefore, control rod withdrawal is not restricted.

B. Incorrect. Plausible. The first part is correct. See the first part of Choice A above.

The second part is incorrect. Per NMP-OS-001 step 6.1.2.4, rod withdrawals are allowed if Tavg/Tref devation is within 3°F.

However, NMP-OS-001 step 6.1.2.4 states it is non-conservative to withdraw control rods in response to a transient and anomalies are to be mitigated utilizing the secondary plant. It is reasonable for a candidate not familiar with the specific guidance of NMP-OS-001 to believe that the Tavg/Tref of 2°F is still considered a "transient" and not allow the use of control rods. Therefore, this distractor is plausible.

C. Incorrect. Plausible. The first part is incorrect. Per the IPC printout, RCS DeltaT Power is approximately 83% and CBD position is 116 steps.

Per the COLR, the RIL for 83% power is 122 steps on CBD.

Therefore, ALB10-D04 is a valid alarm. However, if the candidate does not understand how RIL is calculated and uses NIS Power instead of DeltaT Power, an RIL of 105 steps would be determined and the candidate would conclude the alarm was Thursday, February 20, 2014 8:45:05 AM 2

not valid. Therefore, this distractor is plausible.

The second part is correct. See the second part of choice B above.

D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.

The second part is correct. See the second part of choice B above.

SRO JUSTIFICATION (10CFR43(b))

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, specific knowledge of Reactivity management during a transient per NMP-OS-001 is required.

-Can the question be answered solely by knowing immediate operator actions? No, there are not associated IOA's.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, the only associated AOP (18013-C) does not directly address the RIL alarm or the associated temperature requirement.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, the conditions are specific to NMP-OS-001.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. Yes, the question requires specific knowledge of NMP-OS-001, which is an administrative procedure on Reactivity Management and more specifically the guidance associated with management of a transient condition and the duties of the SRO (Reactivity Management SRO) when reactivity manipulations are being performed.

(6) Procedures and limitations involved in initial core loading, alternations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

Thursday, February 20, 2014 8:45:05 AM 3

The question is about "Procedures and limitations involved in control rod movement as it relates to internal/external effects on core reactivity".

Although not listed as an example for this category in the "Clarification Guidance for SRO-only Questions", this question does include procedural administrative requirements and controls associated with external effects on core reactivity.

Level: SRO Tier # / Group # T2 / G2 K/A# 001G2.1.19 Importance Rating: 3.9 / 3.8 Technical

Reference:

NMP-OS-001 Rev 17.0, page 10 Unit 1 Cycle 18 COR, Figure 3 ARP 17010-1 Rev 50, page 3 and 41 References provided: Plant Computer RIL screenshot and COLR Figure 3 Learning Objective: LO-LP-39205-07 State the reasons for maintaining rods above the RIL.

LO-LP-60301-08 Describe how placing the delta T defeat switch to a failed channel will affect the response of the rod insertion limit computer.

LO-PP-27101-21 State the alarms associated with the rod insertion limits; include set points and source of the set points.

LO-TA-05002 Obtain Data From the Integrated Plant Computer using 13505-1/2 Question origin: BANK - Hatch 2011 NRC Question # G2.1.37 Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments: Early submittal 401-9 response:

Question appears to match the KA. Question appears to be at the SRO level. Question appears to be okay as-is.

- JAT 12/19/13 (SAT)

You have completed the test!

Thursday, February 20, 2014 8:45:05 AM 4

CORE OPERATING LIMITS REPORT, VEGP UNIT 1 CYCLE 18 JULY 2012 FIGURE 3 ROD BANK INSERTION LIMITS VERSUS % OF RATED THERMAL POWER (Fully Withdrawn*)

220 (28.0%, 225) (78.0%, 225) 200 ROD BANK POSITION (Steps Withdrawn)

BANK B 180 160 (0%, 161) (100%, 161) 140 BANK C 120 100 80 BANK D 60 40 (0%, 46) 20 (30.2%, 0) 0 0 10 20 30 40 50 60 70 80 90 100 POWER (% of Rated Thermal Power)

Fully withdrawn shall be the condition where control rods are at a position within the interval t225 and d 231 steps withdrawn.

NOTE: The Rod Bank Insertion Limits are based on the control bank withdrawal sequence A, B, C, D and a control bank tip-to-tip distance of 115 steps.

Page 9 of 11

CORE OPERATING LIMITS REPORT, VEGP UNIT 1 CYCLE 18 JULY 2012 FIGURE 3 ROD BANK INSERTION LIMITS VERSUS % OF RATED THERMAL POWER (Fully Withdrawn*)

220 (28.0%, 225) (78.0%, 225) 200 ROD BANK POSITION (Steps Withdrawn)

BANK B 180 160 (0%, 161) (100%, 161)

IPC Calculated RIL at 122 steps on 140 CBD BANK C 120 100 Current Position Position if NI below RIL Power used to 80 calculate RIL BANK D 60 40 (0%, 46) 20 (30.2%, 0) 0 0 10 20 30 40 50 60 70 80 90 100 POWER (% of Rated Thermal Power)

Fully withdrawn shall be the condition where control rods are at a position within the interval t225 and d 231 steps withdrawn.

NOTE: The Rod Bank Insertion Limits are based on the control bank withdrawal sequence A, B, C, D and a control bank tip-to-tip distance of 115 steps.

Page 9 of 11

Approved By Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 17010-1 50 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL Page Number 08/16/2011 1C1 ON MCB 3 of 66 ALB 10 (1) (2) (3) (4) (5) (6)

A SR/IR NIS SOURCE AND POWER RANGE HI REACTOR BYPASS REACTOR BYPASS ROD CONTROL SIG PROCESSOR INTMD RANGE NEUTRON FLX HI BRKR BYA BRKR BYA NON URGENT TROUBLE TRIP BYPASS SETPOINT ALERT IN-OPERATE CLOSE FAILURE B SOURCE RNG HI POWER RANGE REACTOR BYPASS REACTOR BYPASS ROD CONTROL SHUTDOWN FLUX HI NEUTRON FLX BRKR BYB BRKR BYB URGENT FAILURE ALARM BLOCKED LOW SETPOINT IN-OPERATE CLOSE C SOURCE RANGE POWER RANGE OVERPOWER T ROD BANK RPI NIS CHANNEL HI FLUX LEVEL CHANNEL ROD BLOCK AND LO LIMIT NON URGENT ON TEST AT SHUTDOWN DEVIATION RUNBACK ALERT ALARM D INTMD RANGE PWR RANGE UP OVERPOWER ROD BANK RPI ROD DEV HI FLUX DET HI FLX DEV ROD STOP LO-LO LIMIT URGENT ALARM LEVEL ROD STOP E SR/IR REMOTE PWR RANGE LWR OVERTEMP T ROD AT BOTTOM RADIAL TILT SIG PROCESSOR DET HI FLX DEV ROD BLOCK AND DPU-B TROUBLE RUNBACK ALERT F SR/IR POWER RANGE ROD DRIVE M-G TWO OR MORE DELTA FLUX AMPLIFIER HI NEUTRON FLX SET TROUBLE RODS AT BOTTOM DEVIATION TROUBLE RATE ALERT Printed September 26, 2013 at 11:48

Approved By Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 17010-1 50 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL Page Number 08/16/2011 1C1 ON MCB 41 of 66 IPC and Rod Control function WINDOW D04 ORIGIN SETPOINT ROD BANK IPC Calculated Rod Insertion LO-LO LIMIT Limit UD0366 1.0 PROBABLE CAUSE RCS Boron concentration too low for present reactor power level due to:

1. Plant transient.

2. Xenon transient.

3. IPC failure 2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS

1. Check indications and determine if actual control bank rod position is below the Lo-Lo insertion limit by referring to the COLR and Technical Specification LCO 3.1.6.

2. IF actual control bank position is below the Lo-Lo Insertion Limit, perform the following:

a. Within 1 hour:

Verify shutdown margin is within the limits specified in the COLR per 14005-1 Shutdown Margin Calculation; Refer To TR 13.1.1 for applicability.

OR Initiate and maintain Emergency Boration per 13009-1, CVCS Reactor Makeup Control System, until the Control Banks Lo-Lo Limit Annunciator clears.

b. Restore the affected control bank(s) above the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Printed September 26, 2013 at 11:48

Southern Nuclear Operating Company Nuclear NMP-OS-001 Management Reactivity Management Program Version 17.0 Procedure Page 10 of 39 6.1.2.3 Administrative Controls The following administrative requirements ensure that nuclear safety is maintained during activities that affect reactivity:

Planned reactivity manipulations are peer checked.

Calculations that involve reactivity control are independently verified prior to use.

Reactor operators use redundant instrumentation when monitoring the effects of reactivity manipulations.

Reactor engineering is actively engaged in activities that change reactivity significantly, including any special tests with the potential to affect reactivity.

The reactor operator performing rod movement activities is free from distractions and will have no other duties while performing reactivity manipulations.

See Attachment 2 for summarized expectations for reactivity manipulations.

6.1.2.4 Conduct of Reactivity Changes Operators anticipate the effects of reactivity manipulations and monitor core parameters carefully until parameters stabilize. Any unanticipated reactivity change is immediately brought to the attention of management and the resolution of the change is pursued to its conclusion.

Adding positive reactivity is never an appropriate way to address unstable plant conditions. It is non-conservative to withdraw control rods in response to primary plant anomalies caused by unplanned secondary plant transients. For the PWRs, once turbine load has been stabilized and RCS Tavg has been restored to within 3 degrees of Tref, positive reactivity can be added by withdrawing control rods.

Whenever the status of reactor criticality becomes unknown, the reactor is shutdown.

Under normal conditions, positive reactivity changes will not be performed by more than one means at a time.

During approach to criticality, two positive reactivity additions will not be performed simultaneously.

The operator at the controls shall suspend turnover if a reactivity manipulation is required during turnover.

Rod Position IPC calculation of RIL based on Rod Position and Delta T Power Delta T Power NI Power as a used to calculate distractor RIL

1. 003G2.4.3 001/LOIT AND LOCT/SRO/M/F 3.7/3.9/003G2.4.3/LO-TA-37015///

Initial conditions:

- Unit 1 experienced a small break LOCA.

- 19000-C, "Reactor Trip or Safety Injection," is in progress.

- Loop 3 RCS Tcold instrumentation is not available.

Current condition:

- RCS pressure is 1350 psig and slowly lowering.

Which one of the following completes the following statement?

Stopping the RCPs is required to minimize the risk of __(1)__,

and per the Bases of Tech Spec 3.3.3, "Post Accident Monitoring (PAM) Instrumentation,"

for diverse indication of RCS Tcold temperature, the operator is directed to use __(2)__

instrumentation.

__(1)__ __(2)__

A. core uncovery RCS Thot B. core uncovery steam generator pressure C. RCP damage RCS Thot D. RCP damage steam generator pressure K/A 003 Reactor Coolant Pump G2.3.4 Ability to identify post-accident instrumentation.

K/A MATCH ANALYSIS The question tests the candidate's ability to identify the post-accident instrument that would be utilized as a diverse indication of RCS Tcold temperature with RCPs secured.

EXPLANATION OF REQUIRED KNOWLEDGE Per WOG Background for RCP Trip, the reason for purposely tripping the RCPs during a small break LOCA is to prevent excessive depletion of RCS water inventory through a saml break in the RCS which might lead to severe core uncovery if the RCPs were tripped for some reason later in the accident. The RCPs should be tripped before the RCS inventroy is depleted to the point where tripping of the pumps would cause the Thursday, February 20, 2014 1:26:28 PM 1

break to immediately uncover. The WOG gives options in parameter that may be utilzied as RCP trip criteria. Vogtle has chosen the option based solely on RCS pressure. Therefore, RCPs are manually tripped if RCS pressure is <1375 psig provided either CCPs or SIPs are injecting into the core.

Per TS 3.3.3 FU 2,3 Bases, steam line pressure provides diverse indication for the RCS cold leg temperature. With either forced or natural circulation flow through the steam generators, SGs will be at saturation pressure for the RCS Cold Legs.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Part 1 of the answer is correct and states the reason provided in the Westinghouse background documents for tripping the Reactor Coolant Pumps for small break LOCAs as the potential for core damage due to core uncover and exceeding peek centerline temperatures criteria.

Part 2 is not correct but is 'plausible' because T-hot instrumentation is addressed in the same bases as diverse indication for the CETCs as opposed to Tcold instrumentation.

The candidates would see there is a relationship since both are measuring temperature as opposed to the correct instrument which is using pressure.

B. Correct. Part 1 is correct. See Part 1 of choice A above.

Part 2 is correct. Per Tech Spec 3.3.3 'Post Accident Monitoring Instrumentation,' steam line pressure provides diverse indication for the RCS cold leg temperature.

C. Incorrect. Plausible. Part 1 is not correct for the small break LOCA but would be true for large break LOCAs since the Reactor Coolant Pumps would be stopped under this condition due to loss of support conditions for continued operation and subsequent pump damage.

Part 2 is not correct. See Part 2 of choice A above.

D. Incorrect. Plausible. Part 1 is not correct. See Part 1 of choice C above.

Part 2 of the answer is correct. See Part 2 of choice B above.

SRO JUSTIFICATION (10CFR43(b))

(2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the question is not addressing Tech Spec action times.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, the question is not addressing Tech Spec above-the-line information.

Thursday, February 20, 2014 1:26:29 PM 2

-Can question be answered solely by knowing the TS Safety Limits? No, the question is not related to Tech Spec Safety Limits.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)

• Knowledge of TS bases that is required to analyze TS required actions and terminology. Yes, the answer to the question is only found in Tech Spec bases.

Level: SRO Tier # / Group # T2 / G1 K/A# 003G2.4.3 Importance Rating: 3.7 / 3.9 Technical

Reference:

Westinghouse Background - RCP Trip Rev 2, 4/30/2005 Tech Spec 3.3.3 Bases page B 3.3.3-6, Rev 0 Reference Provided: None Learning Objective: LO-TA-37006 Conduct a Natural Circulation Cooldown per 19002-C LO-TA-37015 Perform the Initial Recovery Actions for a small Loss of Reactor or Secondary Coolant per 19010-C LO-TA-63013 Implement Technical Specification LCO using 10008-C (SRO Only)

LO-LP-39207-04 Describe the bases for any given Tech Spec in section 3.3.

LO-LP-39208-01 For any given item in section 3.4 of Tech Specs, be able to: State the LCO. State any one hour or less required actions.

Question origin: NEW Cognitive Level: M/F 10 CFR Part 55 Content: 41.10 / 43.2 Comments:

You have completed the test!

Thursday, February 20, 2014 1:26:29 PM 3

PAM Instrumentation B 3.3.3 BASES LCO 2,3. Reactor Coolant System (RCS) Hot and Cold Leg Temperatures (Wide Range) (continued)

RCS hot and cold leg temperatures are used to determine RCS subcooling margin. RCS subcooling margin will allow termination of safety injection (SI), if still in progress, or reinitiation of SI if it has been stopped. RCS subcooling margin is also used for unit stabilization and cooldown control.

In addition, RCS cold leg temperature is used in conjunction with RCS hot leg temperature to verify the unit conditions necessary to establish natural circulation in the RCS.

Reactor outlet temperature inputs to the Reactor Protection System are provided by two fast response resistance elements and associated transmitters in each loop. The channels provide indication over a range of 50qF to 700qF.

The core exit thermocouples provide diverse indication for the RCS hot leg temperature.

Steam line pressure provides diverse indication for the RCS cold leg temperature.

4. Steam Generator Water Level (Wide Range)

Wide range SG water level (Loops 501, 502, 503, & 504) is a Type A variable used to determine if an adequate heat sink is being maintained through the SGs for decay heat removal, primarily for the response to a loss of secondary heat sink event when the level is below the narrow range. The wide range SG level indication may also be used in conjunction with auxiliary feedwater flow for SI termination. In addition, the wide range level is cold calibrated and provides a complete range for monitoring SG level during a cooldown. Auxiliary feedwater flow provides the diverse indication for wide range SG water level.

(continued)

Vogtle Units 1 and 2 B 3.3.3-6 Revision No. 0

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1. 004G2.2.44 001/LOIT AND LOCT/SRO/C/A 4.2/4.4/004G2.2.44/LO-TA-16009///

Given the following procedure titles:

- 19000-C, "Reactor Trip or Safety Injection"

- 13003-1, "Reactor Coolant Pump Operation"

- 18005-C, "Partial Loss of Flow" Initial condition:

- Unit 1 is at 16% reactor power with a startup in progress.

Current conditions:

- ALB08-A04 RCP 1 NO. 2 SEAL LKOF HI FLOW is received.

- ALB08-A05 RCP 1 CONTROLLED LKG HI/LO FLOW is received.

- 1FI-160A, #1 SEAL LEAK-OFF for RCP #1 is indicating 6.0 gpm.

Which one of the following completes the following statement?

RCP #1, seal number __(1)__ has failed, and per 13003-1, the Shift Supervisor will direct the crew to __(2)__.

__(1)__ __(2)__

A. one initiate 18005-C B. one trip the reactor and initiate 19000-C C. two initiate 18005-C D. two trip the reactor and initiate 19000-C K/A 004 Chemical and Volume Control G2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

K/A MATCH ANALYSIS The question tests the candidates ability to interpret the control room annunciators associated with RCP seal #1 and #2 leak-offs. Seal injection and leakoff are part of the CVCS system per the K/A catalog. The candidate is then required to utilize these annunciators to diagnose the issue and select the appropriate procedure path to Tuesday, February 25, 2014 8:43:22 AM 1

address the degraded conditions.

EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17008-1 window A05, either low or high seal flow can actuate the alarm. High seal flow is an indication of Seal #1 failure, and low seal flow is an indication of Seal #2 failure. The candidate is given a seal flow of 6.0 gpm to allow differentiation between these two. Window A04 lists a probable cause of Seal #2 failure only. Knowledge of RCP seal package construction is required to understand why this annunciator would be consistent with a Seal #1 failure. As Seal #1 fails, the seal surfaces open and backpressure lowers within the package. Flow down the shaft lowers and increases both to the leak-off path and to Seal #2 .

Per 13003-1, Seal #1 leak-off >5.5 gpm requires stopping the RCP immediately per step 4.2.1.4. With reactor power greater than 15% RTP, the operator is directed to trip the reactor and initiate 19000-C. When the IOA's are complete, the operator is direct to perform steps 4.2.1.4.d thru h to stop the RCP, close the associated spray valve, and isolate seal #1 leak-off.

The question stem specifies 1FI-160A, #1 SEAL LEAK-OFF for RCP #1 is indicating 6.0 gpm. This is top of scale for the indicator. Simulations show that a seal failure equivalent to approximately 9 gpm leak leak-off is required for annunciators ALB08-A04 RCP 1 NO. 2 SEAL LKOF HI FLOW and ALB08-A05 RCP 1 CONTROLLED LKG HI/LO FLOW to be in alarm. At a seal leak-off of >4.8 gpm and <9 gpm, only annunciator ALB08-A05 RCP 1 CONTROLLED LKG HI/LO FLOW is in alarm.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. The combination of annunciator alarms and elevated seal #1 leak-off are symptoms of a seal #1 failure.

The second part is incorrect. With seal #1 leak-off > 5.5 gpm, an immediate shutdown of the RCP is required. Per 13003-1 step 4.2.1.4, the reactor must be tripped and 19000-C initiated.

However, if a candidate does not realize RTP is >15%, then step 4.2.1.4 would direct initiation of AOP 18005-C.

B. Correct. The first part is correct. See the first part of choice A above.

The second part is correct. With seal #1 leak-off > 5.5 gpm, an immediate shutdown of the RCP is required. Per 13003-1 step 4.2.1.4, the reactor must be tripped and 19000-C initiated.

C. Incorrect. Plausible. The first part is incorrect. The combination of annunciator alarms and elevated seal #1 leak-off are symptoms of a seal #1 failure. However, if the candidate does not recognize 1FI-160A indicating 6.0 gpm as being abnormally high, then annunciator ALB08-A04 RCP 1 NO. 2 SEAL LKOF HI FLOW would be a symptom of a seal #2 failure.

The second part is incorrect. With seal #1 leak-off > 5.5 gpm, Tuesday, February 25, 2014 8:43:22 AM 2

an immediate shutdown of the RCP is required. Per 13003-1 step 4.2.1.4, the reactor must be tripped and 19000-C initiated.

However, if a candidate has diagnosed a seal #2 failure, 13003-C still supports an immediate stop of the RCP given other conditions. Since the question does not give an option to leave the RCP running, the candidate must assume a valid reason to stop the RCP has been met. As such, if the candidate then does not realize RTP is >15%, then step 4.2.1.4 would direct initiation of AOP 18005-C.

D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.

The second part is correct. See the second part of choice B above.

SRO JUSTIFICATION (10CFR43(b))

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, the first part is system knowledge however, the second part is discriminating to the SRO level as it requires an operational decision.

-Can the question be answered solely by knowing immediate operator actions?

No, the requrired actions are specific direction associated with ARPs and SOPs and a shutdown decision must be made.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, entry conditions to 18005-C and 19000-C both address plant conditions in a generic nature. Specific procedure knowledge is required to differentiate the required procedure flow path.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, the question requires specific knowledge of a specific SOP step.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Yes, the candidate must use specific knowledge of SOP 13003-C decision flow charts from memory as well as specific knowledge of step 4.2.1.4 in order to direct the specific actions to be taken to shutdown the reactor and address the RCP seal failure.

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures Tuesday, February 25, 2014 8:43:22 AM 3

Level: SRO Tier # / Group # T2 / G1 K/A# 004G2.2.44 Importance Rating: 4.2 / 4.4 Technical

Reference:

ARP 17008-1, Rev 18.0, pages 10-12 SOP 13003-1, Rev 47.1, pages 13-15 & 38-40 AOP 18005-C, Rev 11.1, pages 1 & 3 V-LO-PP-16401, Rev 5.4, pages 12-14 References provided: None Learning Objective: LO-PP-16401-02 Describe the function of RCP seals 1, 2, and 3 including DP across each seal and expected flow rate.

LO-PP-16401-03 Describe the control room indications for a failure of a RCP seal.

LO-TA-60015 Respond to a Partial Loss of Flow per 18005-C LO-TA-16009 Respond to abnormal RCP seal per 13003-1/2 Question origin: MODIFIED - HL14 NRC Question #015/017G2.4.4 Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments: - JAT 12/19/13 (U/E)

Early submittal 401-9 response:

Amandas comments incorporated to include removing the VCT pressure and trend in the stem and moved the question to procedure knowledge as opposed to Tech Spec bases.

- JAT 2/4/14 (U/E)

Response following revision from early submittal:

The first part of the question is improved. However, I am having difficulty seeing the TS completion time connection to the second part of the question, and the way it's written, it appears as though the question can be answered solely by knowing what specific direction is contained within the ARP (which is likely not SRO-only, because it does not involve selection of procedures or sections of a procedure, nor does it require knowledge of >1h TS.).

- JCC 2/5/14 Question replaced with modified question from HL14 NRC associated with RCP seal abnormality and procedure You have completed the test!

Tuesday, February 25, 2014 8:43:22 AM 4

1. 015/017G2.4.4 002/1/1/RCP MALF - EOP ENTRY/C/A - 4.3/MODIFIED/SRO/HL-14 NRC/TNT / RLM Unit 2 is at 12% power when the following annunciators are received.

- ALB08B05 "RCP # 3 CONTROLLED LKG HI / LO FLOW"

- ALB08B04 "RCP # 3 NO. 2 SEAL LKOF HI FLOW" The OATC reports the following indications: Original Question

- RCP # 3 seal leakoff flow Hi Range meter is 6.0 gpm.

- RCP # 3 seal injection flow is 9.9 gpm.

- RCP # 3 Seal Water Inlet temperature is 223°F and stable.

Which one of the the following is the correct procedurally directed action(s) for the SS to take?

A. Per 12004-C, "Power Operations (Mode 1)", commence a unit shutdown to be in Mode 3 in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

B. Trip the reactor and enter 19000-C, "E-0 Reactor Trip or Safety Injection", per 13003-2, "RCP Operation", stop RCP # 3 and close seal leakoff valve HV-8141C.

C. Per 13003-2, stop RCP # 3, close seal leakoff valve HV-8141C, enter 18005-C, "Partial Loss of RCS Flow", commence unit shutdown per 12004-C.

D. Per 12004-C, maintain reactor power at 25%, monitor the RCP per 13003-2 section 4.2.1 "Pump Operation With A Seal Abnormality", contact Duty Engineering.

Wednesday, February 05, 2014 12:44:26 PM 1

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 18005-C 11.1 Effective Date Page Number PARTIAL LOSS OF FLOW 08/15/2012 1 of 5 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure addresses the loss of forced RCS flow during power operation below the P-8 (48%) setpoint.

SYMPTOMS ALB11-E04 RCP TRIP ALB12-A01(B01, C01, D01) RCP LOOP 1 (2, 3, 4) LOW FLOW ALERT ALB08-A01(B01, C01, D01) RCP 1 (2, 3, 4) MTR OVERLOAD ALB11-E06 UNDERVOLTAGE RCP BUS ALERT ALB11-F06 UNDERFREQUENCY RCP BUS ALERT UNIT 1 ALB33-A01(A02) 13.8KV SWGR 1NAA(1NAB) TROUBLE UNIT 2 ALB33-A01(A02) 13.8KV SWGR 2NAA(2NAB) TROUBLE MAJOR ACTIONS Stabilize plant conditions.

Shutdown to Mode 3.

Restart RCP.

Select appropriate UOP.

Printed February 5, 2014 at 10:22

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 18005-C 11.1 Effective Date Page Number PARTIAL LOSS OF FLOW 08/15/2012 3 of 5 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1

1. Check Reactor power - LESS THAN 1. Perform the following:

OR EQUAL TO 15%.

a. Trip the Reactor. 1.a 1.b

b. Go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

2

2. Stop any power changes in progress. 2.

3

3. Initiate the Continuous Actions Page. 3.

4

• 4. Check affected loop SG NR Level - *4. Control feed flow to maintain TRENDING TO 65%. affected loop SG NR level between 60% and 70%.

5

5. Check Tavg - TRENDING TO 5. Adjust control rods to restore Tavg.

PROGRAM.

6

6. Verify PRZR level - TRENDING TO 6.

PROGRAM.

7

7. Verify PRZR pressure - TRENDING 7.

TO 2235 PSIG.

8

8. Check RCP 1 and RCP 4 - 8. Close the affected loop spray valve:

RUNNING.

Loop 1: PIC-0455C Loop 4: PIC-0455B 9

9. Initiate shutdown to Mode 3 by 9.

initiating 12004-C, POWER OPERATION (MODE 1). (TS 3.4.4) 10

10. Determine and correct the cause of 10.

the pump trip.

11

11. Check shutdown to Mode 3 - 11. Return to Step 9.

COMPLETE.

S Printed February 5, 2014 at 10:22

Approved By Procedure Version M.G. Brill Vogtle Electric Generating Plant 13003-1 47.1 Effective Date Page Number 06/12/2013 REACTOR COOLANT PUMP OPERATION 13 of 42 INITIALS 4.2 SYSTEM OPERATION 4.2.1 Pump Operation With A Seal Abnormality 4.2.1.1 IF the Plant Computer is available, trend the computer data points listed in Table 2. ________

4.2.1.2 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 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. ________

b. IF NO further seal degradation exists after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, consult with the Shift Supervisor (SS) for less frequent monitoring. ________

4.2.1.3 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°F RCP shaft vibration 20 mils RCP Frame vibration 5 mils

1. 1 seal Differential Pressure <200 psid

1. 1 seal leakoff flow (sum of #1 seal < minimum on Figure 2 with pump leakoff as indicated on the MCB and #2 bearing / seal inlet temperature seal leakoff read locally in containment) increasing Total loss of ACCW for a duration of 10 minutes Printed February 5, 2014 at 10:24

Approved By Procedure Version M.G. Brill Vogtle Electric Generating Plant 13003-1 47.1 Effective Date Page Number 06/12/2013 REACTOR COOLANT PUMP OPERATION 14 of 42 INITIALS

d. WHEN directed by Figure 1, stop the affected RCP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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. ________

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:

(1) Trip the Reactor and initiate 19000-C, "E-0 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-1E Control Switch in STOP and then placing the RCP 1E Control Switch in STOP:

RCP Non-1E Control Switch 1E Control Switch Loop 1 1-HS-0495B 1-HS-0495A ________

Loop 2 1-HS-0496B 1-HS-0496A ________

Loop 3 1-HS-0497B 1-HS-0497A ________

Loop 4 1-HS-0498B 1-HS-0498A ________

Printed February 5, 2014 at 10:24

Approved By Procedure Version M.G. Brill Vogtle Electric Generating Plant 13003-1 47.1 Effective Date Page Number 06/12/2013 REACTOR COOLANT PUMP OPERATION 15 of 42 INITIALS 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. ________

Printed February 5, 2014 at 10:24

Approved By Procedure Version M.G. Brill Vogtle Electric Generating Plant 13003-1 47.1 Effective Date Page Number 06/12/2013 REACTOR COOLANT PUMP OPERATION 38 of 42 FIGURE 1 - RCP SEAL ABNORMALITIES DECISION TREE 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 D-04 Printed February 5, 2014 at 10:24

Approved By Procedure Version M.G. Brill Vogtle Electric Generating Plant 13003-1 47.1 Effective Date Page Number 06/12/2013 REACTOR COOLANT PUMP OPERATION 39 of 42 FIGURE 2 5.6 gpm is in the non-operate range

1. If the No. 1 seal leak rates are outside the normal (1.0-5.0 gpm) but within the operating 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 pump and seal parameters and contact engineering for further instructions.

IF the No.1 seal leak off is between 0.6 and 0.8 gpm within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> determine the leak off with the No.1 plus No. 2 seals. IF the total leakoff is less than 0.8 gpm perform an orderly shutdown of the pump (see Note 4) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

At 0.6 gpm on the No. 1 seal immediately shutdown the pump.

(The 0.8 gpm and 0.6 gpm value includes the #2 seal leak off value from containment as well.)

Printed February 5, 2014 at 10:24

Approved By Procedure Version M.G. Brill Vogtle Electric Generating Plant 13003-1 47.1 Effective Date Page Number 06/12/2013 REACTOR COOLANT PUMP OPERATION 40 of 42

2. 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).

3. No.1 Seal Differential Press = RCS WR Press - VCT Press.

4. Per Westinghouse Technical Bulletin ESBU-TB-93-01-R1, total #1 seal leakoff is the sum of #1 seal leakoff and #2 seal leakoff. #1 seal leakoff is read directly at the MCB and #2 seal leakoff can be obtained from instrumentation in Containment.

Printed February 5, 2014 at 10:24

Approved By Procedure Number Rev J.B. Stanely Vogtle Electric Generating Plant 17008-1 18 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON Page Number 07/08/11 PANEL 1A2 ON MCB 3 of 55 (1) (2) (3) (4) (5) (6)

A RCP 1 RCP 1 RCP 1 RCP 1 RCP 1 MTR OVERLOAD STANDPIPE STANDPIPE NO. 2 SEAL LKOF CONTROLLED LKG LO LEVEL HI LEVEL HI FLOW HI/LO FLOW B RCP 2 RCP 2 RCP 2 RCP 2 RCP 2 MTR OVERLOAD STANDPIPE STANDPIPE NO. 2 SEAL LKOF CONTROLLED LKG LO LEVEL HI LEVEL HI FLOW HI/LO FLOW C RCP 3 RCP 3 RCP 3 STANDPIPE RCP 3 RCP 3 MTR OVERLOAD STANDPIPE HI LEVEL NO. 2 SEAL LKOF CONTROLLED LKG LO LEVEL HI FLOW HI/LO FLOW D RCP 4 RCP 4 RCP 4 RCP 4 RCP 4 RCP MTR OVERLOAD STANDPIPE STANDPIPE NO. 2 SEAL LKOF CONTROLLED LKG NO. 1 SEAL LO LEVEL HI LEVEL HI FLOW HI/LO FLOW LO P E RCP 1 RCP 2 RCP RCP VIBRATION VIBRATION VIBRATION SEAL WATER INJ ALERT ALERT HIGH FILTER HI P F RCP 3 RCP 4 RCP VIB RCP VIBRATION VIBRATION MON PNL SEAL WATER INJ ALERT ALERT TROUBLE LO FLOW Printed February 5, 2014 at 10:18

Approved By Procedure Number Rev J.B. Stanely Vogtle Electric Generating Plant 17008-1 18 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON Page Number 07/08/11 PANEL 1A2 ON MCB 10 of 55 WINDOW A04 ORIGIN SETPOINT RCP 1 1-FIS-0194 0.9 gpm NO. 2 SEAL LKOF HI FLOW 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

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 on RCP operation with seal malfunctions.

4.0 SUBSEQUENT OPERATOR ACTIONS NONE 5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB114, 1X6AB09-119, PLS Printed February 5, 2014 at 10:18

Approved By Procedure Number Rev J.B. Stanely Vogtle Electric Generating Plant 17008-1 18 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON Page Number 07/08/11 PANEL 1A2 ON MCB 11 of 55 WINDOW A05 ORIGIN SETPOINT RCP 1 1-FT-0161 4.8 gpm CONTROLLED LKG 1-FT-0157 0.8 gpm HI/LO 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 Printed February 5, 2014 at 10:18

Approved By Procedure Number Rev J.B. Stanely Vogtle Electric Generating Plant 17008-1 18 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON Page Number 07/08/11 PANEL 1A2 ON MCB 12 of 55 WINDOW A05 (Continued) 3.0 INITIAL OPERATOR ACTIONS NOTE RCP 1 No. 1 seal water leakoff high range flow may be monitored using computer point F0161.

1. Observe seal injection flow and seal leakoff 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.

4.0 SUBSEQUENT OPERATOR ACTIONS NONE 5.0 COMPENSATORY OPERATOR ACTIONS

1. Verify proper seal leakoff using 1-FI-0156A and 1-FI-0160A once per shift, and refer to 13003-1, "Reactor Coolant Pump Operation" if leakoff is outside the limits.

2. Log corrective actions to repair the disabled annunciator or reasons for no action on 10018-C, "Annunciator Control", Figure 2.

3. Log compensatory actions on 10018-C, "Annunciator Control", Figure 5.

END OF SUB-PROCEDURE

REFERENCES:

1X4DB114, 1X6AB09-119, PLS Printed February 5, 2014 at 10:18

Objective 4 Demonstrate using the seal package model.

The seal package consist of three seals

1) Number 1 seal (film riding) a) The primary seal b) Seal is accomplished with a hydrostatic film between the shaft runner and seal ring.

c) No mechanical contact between seal ring and shaft runner (must keep P >200 psid)

2) Number 2 seal (face rubbing) a) Provides back up for #1 seal b) Consist of carbon graphite (face rubbing seal) c) Graphite makes contact with runner which rotates with shaft d) If #1 seal fails , #2 seal converts to a film riding seal if #1 seal leak off valve is closed and seal is exposed to full RCS pressure. #2 seal designed to allow plant shutdown and should last approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e) Placing #2 seal in service with the RCP shaft still rotating will tend to score the shaft at the #2 seal area. This can require extensive repairs before placing the RCP back in service. Vogtle chooses to remove RCP from service and allow its shaft to come to a standstill before closing the #1 seal leak off valve to avoid this problem.

3) Number 3 seal (face rubbing) a) Prevents the leakage of liquid and gases from the RCS into containment.

b) Consist of carbon graphite seal which makes contact with runner (face rubbing) c) The runner is around the shaft and rotates with it.

d) The seal is actually two graphite sealing surfaces called dams.

V-LO-PP-16401 12

Objective 1d

2) RCP Motor Auxiliaries A) Motor Cooler

1) Containment Air is drawn into the motor by fan blades on motors rotor

2) It is then exhausted through the motor cooler

3) ACCW is the cooling medium used in the cooler (cools the outgoing air)

4) This arrangement limits containment air temperature rise and in turn limits motor temperature.

3) Flywheel A) Addressed in tech spec administrative section B) Stores rotational energy of the pump and motor while running then releases energy by maintaining pump motion to slowly reduce core flow following loss of power for core protection.

4) RCP motor space heater A) Each RCP motor has a electric resistance heater.

B) Used to prevent moisture accumulation in windings when motor is shutdown.

C) Not needed when motor is in operation because of heat generation from motor windings.

D) Heaters are automatically energized when either of the RCP motor breakers are opened.

E) Heaters are supplied from 480 V MCCs RCP motor burned up at Vogtle that was attributed to moisture from space heater breaker being open when pump was shutdown. The space heater did not energize therefore moisture accumulated while the pump was shutdown during the outage. Upon restart the motor windings shorted out. Motor rebuild was required.

V-LO-PP-16401 13

Objective 2 RCP Seal Injection a) Provided from CVCS b) 8 gpm per pump c) 5 gpm is directed through to lower pump radial bearing and into the RCS loop.

d) The remaining 3 gpm supplies #1 and #2 seals e) #3 seal injection is from small tanks called Standpipes. (Gravity Fed) f) Flow path

1) 8 gpm from CVCS enters RCP at 2250 psig

2) 5 gpm passes through the lower pump bearing lubricating and cooling it.

3) Seal injection at 2250 psig prevents RCS water from escaping the loop.

4) 3 gpm is directed through #1 seal

5) A pressure drop at 2220 psid across the #1 seal occurs.

6) Approximately 3 gph (0.05 gpm) leak off from #1 seal is used as seal injection to #2 seal.

7) The remainder of #1 seal leak off is directed back to the VCT via seal water return.

8) 3 gph passes through #2 seal and the leak off is directed to the RCDT (~5-6 psig)

9) 800 cc/hr seal injection for #3 seal is provided by standpipe (~10 psig)

10) The standpipes are located at a higher elevation than the RCP and gravity feeds #3 seal; standpipes Auto fill from RMWST.

11) #3 seal injection is injected between the two dams and sealing surfaces.

12) #3 seal injection pressure is slightly higher than #2 seal injection leak off.

13) This prevents RCS liquids or gases from escaping to the containment environment.

14) #3 seal has two leak off paths a) The outer dam leak off (400 cc/hr) combines with #2 seal leak off and is routed to RCDT b) The inner dam leak off (400 cc/hr) is directed to the containment sump.

SMART - Solid Knowledge. If the #1 seal leakoff was isolated, the #2 seal would become a film riding seal due to increased pressure across the #2 seal facing.

V-LO-PP-16401 14

1. 008A2.02 001/LOCT AND LOIT/SRO/C/A 3.2/3.5/008A2.02/LO-TA-60026//HL-18 NRC/

Initial conditions:

- Unit 1 is at 100% reactor power.

- CCW pump #5 is tagged out for maintenance.

LSLL-1854, CCW Surge Tank level switch for CCW pump #3, has failed.

- ALB02-A05 CCW TRAIN A SURGE TK LO-LO LVL is received.

Current conditions:

- ALB02-A06 CCW TRAIN A LO HDR PRESS is received.

- ALB02-B06 CCW TRAIN A LO FLOW is received.

Which one of the following completes the following statement?

Demin Water Makeup Valve to the CCW Train 'A' Surge Tank __(1)__ automatically open, and per Tech Spec 3.7.7, "Component Cooling Water (CCW) System," Train 'A' CCW is declared __(2)__.

__(1)__ __(2)__

A. will OPERABLE B. will inoperable C. will NOT OPERABLE D. will NOT inoperable 008 Component Cooling Water System (CCWS)

A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

- High / Low surge tank level K/A MATCH ANALYSIS The candidate is presented with a plausible scenario where a CCW Surge Tank level Thursday, February 20, 2014 3:54:06 PM 1

transmitter fails low and is required to determine the impact of the failure on the system.

With one CCW pump tagged out and the inability to start the pump with the failed level transmitter, a decision of operability must be determined which is SRO required knowledge.

EXPLANATION OF REQUIRED KNOWLEDGE Each train of CCW is comprised of 3 pumps. Two pumps are required to met LCO 3.7.7. The pumps automatically start on SI, LOSP, Lo header pressure, and trip of a running pump. Only the SI and LOSP starts are required per SR 3.7.7.2.

When level switch 1-LSLL-1854 fails low, CCW pump #3 trips. With CCW pump #5 already tagged out, the system header pressure lowers and annunciators ALB02-A06 &

B06 alarm. Since automatic Demin Water Makeup Valve LV-1850 is controlled by level switches 1-LSL-1850 & 1-LSH-1850, its operation is not affected. (Reference P&ID 1X4DB136) No actual low level condition exists, therefore LV-1850 will not open.

Since 1-LSLL-1854 failing low resulted in a pump trip, SR 3.7.7.2 to verify each CCW pump starts automatically on an actual or simulated actuation signal can not be satsified. Per TS SR 3.0.1, failure to meet a Survelliance, whether such failure is experienced during the performance of the Survelliance or between performances of the Survelliance, shall be failure to meet the LCO. Therefore, CCW Pump #3 must be declared inoperable. Since CCW Pump #5 was already inoperable, TS 3.7.7 LCO is not met and RAS 'A' must be entered.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Since automatic makeup valve LV-1850 is controlled by level switches 1-LSL-1850 &

1-LSH-1850, its operation is not affected. No actual low level condition exists, therefore LV-1850 will not open. However, if the candidate believes makeup is controlled off the same level transmitter or misses that the level transmitter failed and believes there is an actual lo level, then makeup valve LV-1850 would be expected to open.

The second part is incorrect. SR 3.7.7.2 can not be met and CCW Pmp #3 and #5 are both inoperable. LCO 3.7.7 cannot be met. However, a candidate not familiar with the requirements of SR 3.7.72 & SR 3.0.1 or the pump start logic could conclude that the non-safety related surge tank low level trip would be bypassed on an SI or LOSP or that auto makeup would restore a low level condition and CCW Pmp #3 would remain OPERABLE.

B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.

The second part is correct. SR 3.7.7.2 can not be met and CCW Pmp #3 and #5 are both inoperable. LCO 3.7.7 cannot be met.

Thursday, February 20, 2014 3:54:06 PM 2

C. Incorrect. Plausible. The first part is correct. Since automatic makeup valve LV-1850 is controlled by level switches 1-LSL-1850 &

1-LSH-1850, its operation is not affected. No actual low level condition exists, therefore LV-1850 will not open.

The second part is incorrect. See the second part of choice A above.

D. Correct. The first part is correct. See the first part of choice C above.

The second part is correct. See the second part of choice B above.

SRO JUSTIFICATION (10CFR43(b))

(2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, generic LCO SR3.0.1 knowledge is required. No 1 hr or less actions exist.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, the required knowledge is found in generic LCO SR3.0.1 and SR 3.7.7.2, which is below the line.

-Can question be answered solely by knowing the TS Safety Limits? No, this question does not involve a TS Safety Limit.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes, the question requires an OPERABILITY call be made by utilizing generic LCO SR3.0.1 as applied to SR 3.7.7.2.

• Knowledge of TS bases that is required to analyze TS required actions and terminology Thursday, February 20, 2014 3:54:07 PM 3

Level: SRO Tier # / Group # T2 / G1 K/A# 008A2.02 Importance Rating: 3.2 / 3.5 Technical

Reference:

ARP 17002-1, Rev 24.1, pages 3, 14, & 26 P&ID 1X4DB136, Rev 33.0 TS 3.7.7, Amendment No. 96, pages 3.7.7-1 & 2 TS 3.0.1, Amendment No. 125, page 3.0-4 References provided: None Learning Objective: LO-PP-10101-04 From memory, describe the expected system response and operator corrective actions for each of the following:

d. Surge Tank Low Level LO-LP-39211-04 Describe the bases for any given Tech Spec in section 3.7.

LO-TA-60026 Respond to a Loss of CCW per 18020-C LO-TA-10006 Draw the CCW System Question origin: BANK - Reuse - HL18 Question # 008A2.02 Cognitive Level: C/A 10 CFR Part 55 Content: 43.2 Comments:

You have completed the test!

Thursday, February 20, 2014 3:54:07 PM 4

CCW System 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Component Cooling Water (CCW) System LCO 3.7.7 Two CCW trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CCW train A.1 -------------NOTE-------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops - MODE 4,"

for residual heat removal loops made inoperable by CCW.

Restore CCW train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Vogtle Units 1 and 2 3.7.7-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

CCW System 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 ------------------------------NOTE----------------------------

Isolation of CCW flow to individual components does not render the CCW System inoperable.

Verify each CCW manual, power operated, and In accordance with automatic valve in the flow path servicing safety the Surveillance related equipment, that is not locked, sealed, or Frequency Control otherwise secured in position, is in the correct Program position.

SR 3.7.7.2 Verify each CCW pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program Vogtle Units 1 and 2 3.7.7-2 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.

Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per . . ."

basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

(continued)

Vogtle Units 1 and 2 3.0-4 Amendment No. 125 (Unit 1)

Amendment No. 103 (Unit 2)

Approved By Procedure Version P.H. Burwinkel Vogtle Electric Generating Plant 17002-1 24.1 Effective Date ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 02 ON PANEL Page Number 07/27/2012 1A1 ON MCB 3 of 42 (1) (2) (3) (4) (5) (6)

NSCW TRAIN A NSCW TRAIN A NSCW TRAIN A NSCW TRAIN A CCW TRAIN A CCW TRAIN A A F-1 HI VIB F-2 HI VIB F-3 HI VIB F-4 HI VIB SURGE TK LO HDR PRESS LO-LO LVL NSCW TRAIN A NSCW TRAIN A CCW TRAIN A CCW TRAIN A B LO HDR PRESS TRANSF PMP SURGE TK LO FLOW LO DISCH PRESS HI/LO LVL NSCW TRAIN A NSCW TRAIN A NSCW TRAIN A CCW TRAIN A CCW TRAIN A C CLG TWR BASIN DG CLR RHR PMP & MTR SURGE TK RHR HX HI/LO LVL LO FLOW CLR LO FLOW MAKE UP LVL HI FLOW NSCW TRAIN A NSCW INTERTIE CCW TRAIN A D CNMT CLR 1 & 2 TRN A TO TRN B RHR HX LO FLOW HI FLOW LO FLOW NSCW TRAIN A NSCW TRAIN A RMWST CCW TRAIN A PRIMARY E CNMT CLR 5 & 6 NORM/BYP VLV VAC DEGASIFIER RHR PMP SEAL EQUIPMENT LO FLOW MISPOSITIONED PNL ALARM LO FLOW HI TEMP NSCW TRN A RX RX MAKE UP RX MAKE UP F CVTY CLG COIL STOR TK STOR TK LOW FLOW LO-LO LVL HI/LO LVL Printed February 20, 2014 at 15:04

Approved By Procedure Version P.H. Burwinkel Vogtle Electric Generating Plant 17002-1 24.1 Effective Date ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 02 ON PANEL Page Number 07/27/2012 1A1 ON MCB 14 of 42 WINDOW A05 ORIGIN SETPOINT CCW TRAIN A 1-LSLL-1852 4.75 in. below CL SURGE TK 1-LSLL-1854 (equal to 42%) LO-LO LVL 1-LSLL-1856 1.0 PROBABLE CAUSE

1. Failure of automatic make-up from Demineralized Water System.

2. Failure of manual make-up from Reactor Makeup Water System.

3. Leak in Component Cooling Water System.

2.0 AUTOMATIC ACTIONS LO-LO level will trip Component Cooling Water Pumps.

3.0 INITIAL OPERATOR ACTIONS Go To 18020-1, "Loss Of Component Cooling Water.

4.0 SUBSEQUENT OPERATOR ACTIONS NONE 5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB136, 1X3D-BD-L01A, 1X3D-BD-L01C, 1X3D-BD-L01E, 1X5DN091-1, -2, -3, 1X5DT0022, CX5DT101-96 Printed February 20, 2014 at 15:04

Approved By Procedure Version P.H. Burwinkel Vogtle Electric Generating Plant 17002-1 24.1 Effective Date ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 02 ON PANEL Page Number 07/27/2012 1A1 ON MCB 26 of 42 WINDOW C05 ORIGIN SETPOINT CCW TRAIN A 1-LSL-1850 1.25 in. above CL SURGE TK (equal to 52%) MAKE UP LVL 1.0 PROBABLE CAUSE Component Cooling Water (CCW) System leakage.

2.0 AUTOMATIC ACTIONS Makeup Valve 1-LV-1850 opens.

3.0 INITIAL OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Monitor level using 1-LIT-1846 or computer point L2671.

2. IF 1-LV-1850 fails to open:

a. Open the valve using 1-HS-1850 on QMCB,

b. Continue to monitor level,

c. Open 1-LV-1848 using 1-HS-1848 if level continues to fall.

3. IF equipment failure is indicated, initiate maintenance as required.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB136, 1X3D-BD-L01G, 1X5DT0022, CX5DT101-95 Printed February 20, 2014 at 15:04

Pump Trip Level Switches

1. 008AG2.4.41 001/LOCT AND LOIT/SRO/C/A 2.9/4.6/008AG2.4.41/LO-TA-40002///

Initial condition:

- Unit 1 is at 100% reactor power.

Current conditions:

- ALB12-F01 PRZR SAFETY RELIEF DISCH HI TEMP is received.

- PRT level, temperature, and pressure are all increasing.

- CVCS charging flow is 110 gpm.

- All RCP seal injection and #1 seal return flows are within normal operating range.

- CVCS letdown is isolated.

- Pressurizer level is 55% and stable.

Which one of the following completes the following statement?

Per Tech Spec 3.4.13, "RCS Operational Leakage," the RCS leakage is classified as

__(1)__,

and per NMP-EP-110, "Emergency Classification Determination and Initial Action," the Shift Manager is required to declare as a minimum a(n) __(2)__.

REFERENCE PROVIDED

__(1)__ __(2)__

A. identified NOUE B. identified Alert C. unidentified NOUE D. unidentified Alert K/A 008 Pressurizer Vapor Space Accident G2.4.41 Knowledge of the emergency action level thresholds and classifications.

K/A MATCH ANALYSIS The question sets up plausible scenario where Main Control annunciators are received and various plant parameters and indications are provided. Based on the information provided the SRO candidate is required to assess the type of leakage indicated and determine if any emergency action level thresholds were exceeded. The type of Friday, February 21, 2014 10:13:47 AM 1

leakage reflects the first part of the KA requirement for a vapor space accident, the second action to quantify the leakage amount and determine any Emergency Plan requirements meets the second part of the KA and brings the question to the SRO knowledge level.

EXPLANATION OF REQUIRED KNOWLEDGE The given plant conditions are indicitive of a Pressuirzer Code Safety valve relieving into the PRT. Per TS 3.4.13 Bases, identified leak is defined as being from a specifically known and located source, ie it must be collected and quantifiable, and not pressure boundary. Based on the given conditions, the leakrate is approximately 98 gpm, [charging-(letdown + seal leakoff)]. Total seal leakoff is normally about 12 gpm.

Since the leakage is across a vavle seat, it is not pressure boundary. Therefore, the leakage described fully meets the definition of IDENTIFIED.

Per NMP-EP-110-GL03 Figure 1, a potential loss of the RCS barrier exists if RCS leak rate is non-isolatabe and >120 gpm. Since leakrate is calculated to be 98 gpm, this threshold has not been exceeded. Per NMP-EP-110-GL02 Figure 2, NOUE threshold SU5 is exceeded if identified leakage is > than 25 gpm. Therefore, this threshold has been exceeded.

ANSWER / DISTRACTOR ANALYSIS A. Correct. Part 1 is correct it required the candidate to determine the type of RCS leakage based on information provided and knowledge of the Tech Spec definition of the various leakage categories.

From the information the candidate should determine this is identified leakage. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank.

Part 2 required the candidate to take the information gathered in Part 1 and apply it to the various Emergency Action Level classification thresholds. From that a leak rate of 98 gpm is determined and limits for NOUE recognized as exceeded.

B. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice A above.

Part 2 is incorrect but plausible because the candidate may determine that 110 gpm charging added to 32 gpm seal injection (which is alreay included in the charging flow indicator reading) going into RCS is 142 GPM leak. This minus the RCP seal return flow of 3.2 each gpm would be above the ALERT threshold of 120 gpm for the potential loss of the RCS barrier. (Reference P&IDs 1X4DB114 and 1X4DB116-1)

C. Incorrect. Plausible. Part 1 is incorrect but 'plausible' in that the candidate must first identify the type of RCS leakage. As the PRT continues to recieve effluent from the Przr Code Safety, it will eventually Friday, February 21, 2014 10:13:47 AM 2

rupture. If the candidate assumes that once this occur the leakage is no longer being "collected" in the PRT and not think about the leakage now being colleted in the containment sump, then it would be reasonable to interpret the indications as UNIDENTIFIED RCS leakage.

Part 2 is correct. See Part 2 of choice A above.

D. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice C above.

Part 2 is incorrect. See Part 2 of choice B above.

SRO JUSTIFICATION (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, the answer requires specific knowledge of emergency classification thresholds.

-Can the question be answered solely by knowing immediate operator actions?

No, IOAs are not addressed by this question.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, the question does not address AOP or EOP entry conditions.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, the answer requires specific knowledge of emergency classification thresholds.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. Yes, the answer requires specific knowledge of emergency classification thresholds and determination of the specific classification based on current plant conditions. This is an SRO ONLY job link associated with an SRO ONLY objective. [LO-LP-40101-13 Given an emergency scenario, and the procedure, classify the emergency (SRO only).]

Friday, February 21, 2014 10:13:47 AM 3

Level: SRO Tier # / Group # T1 / G1 K/A# 008G2.4.41 Importance Rating: 2.9 / 4.6 Technical

Reference:

NMP-EP-110-GL03, Rev 3.0, page 58 NMP-EP-110-GL03, Figure 1, Rev 3.0, page 121 NMP-EP-110-GL03, Figure 2, Rev 3.0, page 122 P&ID 1X4DB114, Rev 50.0 P&ID 1X4DB116-1, Rev 41.0 TS 3.4.13 Bases, Rev 2-9/06, page B.3.4.13-2 TS 3.4.13 Bases, Rev 1-9/03, page B.3.4.13-3 References provided: NMP-EP-110-GL03, Figure 1 NMP-EP-110-GL03, Figure 2 NMP-EP-110-GL03, Figure 3 Learning Objective: LO-TA-40002 Emergency Classification and Implementing Instructions using NMP-EP-110 (SRO Only)

LO-TA-60014 Respond to Reactor Coolant System Leakage per 18004-C LO-LP-40101-13 Given an emergency scenario, and the procedure, classify the emergency (SRO only).

LO-LP-39202-02 Demonstrate a working knowledge of the application of all Technical Specification definitions.

LO-LP-39202-01 Define the following terms, as per Plant Vogtle Tech Specs:

i. identified leakage

q. unidentified leakage LO-LP-60304-04 Given the symptoms of RCS leakage into an area or system, correctly identify the leakage area or system.

LO-LP-60304-10 Given conditions and/or indications of leaks identified in Attachment "A" of AOP 18004-C, determine the probable location of the leakage per 18004-C.

LO-LP-60304-12 Discuss how approximate RCS leak rate is determined.

Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments:

You have completed the test!

Friday, February 21, 2014 10:13:47 AM 4

RCS Operational LEAKAGE B 3.4.13 BASES (continued)

APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses SAFETY ANALYSES do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analyses for an event resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is one gallon per minute or increases to one gallon per minute as a result of accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of an off-normal condition. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary (continued)

Vogtle Units 1 and 2 B 3.4.13-2 Rev. 2-9/06

RCS Operational LEAKAGE B 3.4.13 BASES LCO c. Identified LEAKAGE (continued)

LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day. The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage,"

measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

(continued)

Vogtle Units 1 and 2 B 3.4.13-3 Rev. 1-9/06

NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 Figure 1 VOGTLE ELECTRIC GENERATING PLANT Figure 1 - Fission Product Barrier Evaluation NMP-EP-110- GL03 Rev 3.0 General Emergency Site Area Emergency Alert Unusual Event FG1 FS1 FA1 FU1 Loss of ANY Two Barriers AND Loss or Loss or Potential Loss of ANY Two ANY Loss or ANY Potential Loss of ANY Loss or ANY Potential Loss of Potential Loss of Third Barrier Barriers EITHER Fuel Clad OR RCS Containment Fuel Clad Barrier Loss Potential Loss

1. Critical Safety Function Status 1. Critical Safety Function Status Core-Cooling RED Core Cooling-ORANGE OR Heat Sink-RED

2. Primary Coolant Activity Level 2. Primary Coolant Activity Level Indications of RCS Coolant Activity greater than 300 Ci/gm Dose Not Applicable Equivalent I-131

3. Core Exit Thermocouple Readings 3. Core Exit Thermocouple Readings Core Exit TCs greater than 1200F Core Exit TCs greater than 711F

4. Reactor Vessel Water Level 4. Reactor Vessel Water Level Not Applicable RVLS LEVEL less than 63%

5. Containment Radiation Monitoring 5. Containment Radiation Monitoring Containment Radiation Monitor RE-005 OR 006 greater than 6E+6 Not Applicable mR/hr

6. Other Indications 6. Other Indications Not applicable Not applicable

7. Emergency Director Judgment 7. Emergency Director Judgment Judgment by the ED that the Fuel Clad Barrier is lost. Consider conditions not Judgment by the ED that the Fuel Clad Barrier is potentially lost. Consider addressed and inability to determine the status of the Fuel Clad Barrier conditions not addressed and inability to determine the status of the Fuel Clad Barrier.

RCS Barrier Loss Potential Loss

1. Critical Safety Function Status 1. Critical Safety Function Status Not Applicable RCS Integrity-RED OR Heat Sink-RED

2. RCS Leak Rate 2. RCS Leak Rate RCS subcooling less than 24F {less than 38 F Adverse} due to an Non-isolable RCS leak (including SG tube Leakage) greater than 120 RCS leak greater than Charging / RHR capacity gpm

3. SG Tube Rupture 3. SG Tube Rupture SGTR resulting in an SI actuation Not Applicable

4. Containment Radiation Monitoring 4. Containment Radiation Monitoring CTMT Rad Monitor RE-005 OR 006 greater than 2.0E+4 mR/hr Not Applicable

5. Other Indications 5. Other Indications Not applicable Unexplained level rise in ANY of the following:

Distractor Containment sump Reactor Coolant Drain Tank (RCDT)

Waste Holdup Tank (WHT)

6. Emergency Director Judgment 6. Emergency Director Judgment Judgment by the ED that the RCS Barrier is lost. Consider conditions not Judgment by the ED that the RCS Barrier is potentially lost. Consider conditions addressed and inability to determine the status of the RCS Barrier not addressed and inability to determine the status of the RCS Barrier.

Containment Barrier Loss Potential Loss

1. Critical Safety Function Status 1. Critical Safety Function Status Not Applicable Containment-RED

2. Containment Pressure 2. Containment Pressure Rapid unexplained CTMT pressure lowering following initial pressure CTMT pressure greater than 52 psig rise OR OR CTMT hydrogen concentration greater than 6%

Intersystem LOCA indicated by CTMT pressure or sump level response OR not consistent with a loss of primary or secondary coolant CTMT pressure greater than 21.5 psig AND Less than the following minimum operable equipment:

Four CTMT fan coolers AND One train of CTMT spray

3. Core Exit Thermocouple Reading 3. Core Exit Thermocouple Reading Not applicable CORE COOLING CSF - RED OR - ORANGE for greater than 15min AND RVLS LEVEL less than 63%

4. SG Secondary Side Release with Primary to Secondary Leakage 4. SG Secondary Side Release with P-to-S Leakage RUPTURED S/G is also FAULTED outside of containment Not applicable OR Primary-to-Secondary leakrate greater than 10 gpm with nonisolable steam release from affected S/G to the environment

5. CNMT Isolation Valves Status After CNMT Isolation 5. CNMT Isolation Valves Status After CNMT Isolation CTMT isolation valve(s) OR damper(s) are NOT closed resulting in a Not Applicable direct pathway to the environment after containment isolation is required

6. Significant Radioactive Inventory in Containment 6. Significant Radioactive Inventory in Containment Not Applicable CTMT Rad monitor RE-005 OR 006 greater than 2.4E+8 mR/hr

7. Other Indications 7. Other Indications Pathway to the environment exists based on VALID Not applicable RE-2562C Alarm AND RE-12444C OR RE-12442C Alarms

8. Emergency Director Judgment 8. Emergency Director Judgment Judgment by the ED that the CTMT Barrier is lost. Consider conditions Judgment by the ED that the CTMT Barrier is potentially lost. Consider not addressed and inability to determine the status of the CTMT Barrier conditions not addressed and inability to determine the status of the CTMT Barrier 121

NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 FIGURE 2 Distractor Classification 122

NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 SU5 Initiating Condition RCS Leakage.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Threshold Values: (1 OR 2)

1. Unidentified OR pressure boundary leakage greater than 10 gpm.

2. Identified leakage greater than 25 gpm.

Basis:

This IC is included as a NOUE because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was selected as it is observable with normal control room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances). The Threshold Value for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.

58

Total charging flow indicator upstream of seal injection flow indicator To normal charging nozzle To seal injection

Seal Injection flow meter downstream of FI-0121

1. 022G2.1.19 001/LOCT AND LOIT/SRO/M/F 3.9/3.8/022G2.1.19/LO-TA-63013///

Initial conditions:

- Unit 1 is at 100% reactor power.

- Containment Cooler #1 high speed fan breaker trips.

Currect conditions:

- ALB01-E06 CNMT HI TEMP is received.

- IPC data for containment temperature is collected.

Which one of the following completes the following statement?

Based on the IPC data provided, the Tech Spec Surveillance for containment temperature (Tech Spec SR 3.6.5.1) __(1)__ within Tech Spec limits, and per the applicable Annunciator Response Procedure and System Operating Procedure, the crew is required, as a minimum, to __(2)__.

REFERENCE PROVIDED A. (1) is (2) start one additional Containment Cooler B. (1) is (2) stop Containment Cooler #2, then start an additional pair of Containment Coolers C. (1) is NOT (2) start one additional Containment Cooler D. (1) is NOT (2) stop Containment Cooler #2, then start an additional pair of Containment Coolers K/A 022 Containment Cooling AA2.01 Ability to use plant computers to evaluate system or component status.

K/A MATCH ANALYSIS Friday, February 21, 2014 11:25:14 AM 1

The question sets up plausible scenario where the candidate must first determine which instruments are required to be referenced when complying with the Containment temperature monitoring verification surveillance of Technical Specifications. A decision is required based on the operating limit in comparison with an IPC screen shot. The second part matches the KA for the Shift Supervisor to evaluate Containment Cooling requirements used to control within environmental limits and the first part makes this SRO required knowledge when the surveillance requirement is tested.

EXPLANATION OF REQUIRED KNOWLEDGE Per TS SR 3.6.5.1, containment average temperature is required to be verified <120F on a periodic bases. The IPC screen for Control Room TS Rounds lists values for CMNT Levels 2, C, & B and an average temperature. Individual readings above 120F are acceptable as long as the average temperature is <120F.

Per OSP 14000-1, page 15, if the IPC points for containment temperature are not available, annunciator ALB01-E06 can be verified extinquished as an alternate. This annunciator utilzes temperature elements from all three levels and has a setpoint of

<120F. If the annunciator is in alarm and the IPC is unavailable, then local temperature readings are required.

Per ARP 17001-1 for ALB01-E06, if any of the three containment temperature indicators rise above 120F, the operator is instructed to start an additional pair of Containment Coolers. Containment Coolers must be started in specific pairs (1&2, 3&4, 5&6, and 7&8) due to the backdrafter damper for the individual fans being de-energized open on a common plentum (reference P&ID 1X4DB212 and SOP 13120-1 section 4.1). Starting the fans one at a time will result in the non-running fan of the pair to spinning backwards. When the second fan is subsequently started, the supply breaker will trip open from high in-rush current produced by the increased electrical slip angle.

Since fan #1 tripped, fan #2 should be stopped prior to starting an additional pair. Fan

1. 2 is mostly recirculating short-cycled flow in the current configuration and is producing little cooling to containment. Simply starting a second fan on another pair would only create the same condition on a second pair and would also do little to lower containment temperature.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Part 1 is correct and sets up a circumstance were the candidate must determine which instrumentation in particular is required to be monitored the ensure compliance with SR 3.6.5.1 for Containment temperature. The SR requires the average temperature of three levels in Containment be noted to ensure Containment temperature remains within the accident analyses.

Part 2 is incorrect but 'plausible' because the candidate may determine that the only required action is to start an additional Containment Cooler to replace the tripped Containment cooler not recalling that the system operating procedure would require Friday, February 21, 2014 11:25:14 AM 2

the fans be operated in pairs.

B. Correct. Part 1 is correct. See Part 1 of choice A above.

Part 2 is correct per the ARP 17001-1, and the SOP 13120-1, an additional pair of Containment Coolers would be started.

C. Incorrect. Plausible. Part 1 is incorrect but 'plausible' in that the candidate may determine that Technical Specification if any of the three levels listed on the IPC printout exceeded 120F. This assumption would be consistent with the use of ALB01-E06 as an alternate as described in OSP 14000-1.

Part 2 is incorrect. See Part 2 of choice A above.

D. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice C above.

Part 2 is correct. See part 2 of choice B above.

SRO JUSTIFICATION (2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the question does not address 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tech Spec actions.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, the question is not related to above-the-line information in Tech Spec.

-Can question be answered solely by knowing the TS Safety Limits? No, the question is not related the Tech Spec Safety Limits.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1). Yes, the question requires specific knowledge related to Tech Spec surveillance requirements and instruments used to satisfy the requirement.

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4). No, the question is not related to these aspects of Tech Spec.

• Knowledge of TS bases that is required to analyze TS required actions and terminology. No, the question is not related to Tech Spec bases.

Friday, February 21, 2014 11:25:14 AM 3

Level: SRO Tier # / Group # T2 / G1 K/A# 022G2.1.19 Importance Rating: 3.9 / 3.8 Technical

Reference:

TS 3.6.5, Amendment No. 158, page 3.6.5-1 TS 3.6.5 Bases Rev 14.0, page 3.6.5-4 SOP 13120-1, Rev 24.0, pages 3 & 5-8 ARP 17001-1, Rev 31.1, page 3 & 47 OSP 14000-1, Rev 88.2, page 15 P&ID 1X4DB212, Rev 12.0 References provided: IPC Screen shot of CNMT Temps Learning Objective: LO-TA-63013 Implement Technical Specification LCO using 10008-C (SRO Only)

LO-LP-39209-01 For any given item in section 3.5 of Tech Specs, be able to: State the LCO. State any one hour or less required actions LO-LP-39209-03 Describe the bases for any given Tech Spec in section 3.5.

LO-PP-29101-15 State the starting interlocks associated with the Containment Cooling fans.

Include set points and coincidence where applicable.

Question origin: MODIFIED - LORQ Question # V-LO-PP-29101-09 001 Cognitive Level: M/F 10 CFR Part 55 Content: 41.9 / 43.2 Comments:

You have completed the test!

Friday, February 21, 2014 11:25:14 AM 4

1. V-LO-PP-29101-09 001/SRO/0022A2.04/LO-TA-29003/C/A/3/DIABLO CANYON 05/SOP 13120-1/2/TECH SPEC 3.6.5 Due to the high summer air temperature and fouling of the Containment Air Coolers are causing the following conditions to occur on Unit 1:

- ALB01-E06 CNMT HI TEMP is in alarm Original Question

- Containment Level 2 temperature - 122°F

- Containment Level C temperature - 114°F

- Containment Level B temperature - 120°F

- Containment Coolers 1,2,5, and 6 are running in high speed

- Containment pressure is currently 1.0 psig and increasing slowly Based on the current plant conditions, which of the following actions should the Shift Supervisor the operator to perform?

A. Direct a start of the Train B Containment Coolers in Hi speed, and consider venting Containment.

B. Direct a start of Train B Containment Coolers in Hi speed, and restore pressure within limits within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in MODE 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

C. Direct a start of the Train B Containment Coolers in Hi speed, reduce Containment temperature to less than 120°F within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in MODE 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and consider venting Containment.

D. Direct a start of the Train B Containment Coolers in Hi speed, restore pressure within limits within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in MODE 3 in the next 6 hrs and consider venting Containment.

Monday, January 20, 2014 11:59:42 AM 1

Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be 120°F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air A.1 Restore containment 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> temperature not within average air temperature limit. to within limit.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is In accordance with within limit. the Surveillance Frequency Control Program Vogtle Units 1 and 2 3.6.5-1 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

Containment Air Temperature B 3.6.5 BASES (continued)

SURVEILLANCE SR 3.6.5.1 REQUIREMENTS Location Tag Number

a. Level 2 TE-2563

b. Level B TE-2613

c. Level C TE-2612 NOTE: A local sample may be taken at a corresponding location in lieu of using one of the instruments designated above.

Verifying that containment average air temperature is within the LCO limit ensures that containment operation remains within the limit assumed for the containment analyses. In order to determine the containment average air temperature, an arithmetic average is calculated using measurements taken at locations within the containment selected to provide a representative sample of the overall containment atmosphere. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. FSAR, Section 6.2.

2. 10 CFR 50.49.

Approved By Procedure Number Rev M. D. Askew Vogtle Electric Generating Plant 13120-1 24 Date Approved Page Number 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM 3 of 50 1.0 PURPOSE This procedure provides instructions for operation of the Containment Building Cooling Systems, which consist of these subsystems: Containment Heat Removal System (CHRS), Control Rod Drive Mechanism (CRDM) Cooling Units, and the Containment Building (CTB) Cavity Cooling System. Instructions are provided in the following subsections:

4.1 CTB Cooling System Startup To Standby 4.2 Containment Heat Removal System Startup 4.3 CRDM Cooling Units Startup 4.4 Shifting CRDM Cooling Units 4.5 CTB Cavity Cooling System Startup 4.6 Shifting CTB Cavity Cooling Units 4.7 Shifting CTB Reactor Support Cooling Fans 4.8 Shifting CTB Cooling Unit Fans 4.9 Shifting Auxiliary Coolers 4.10 Post LOCA Purge Cavity Fan Operation 4.11 Containment Heat Removal System Shutdown 4.12 CRDM Cooling Units Shutdown 4.13 CTB Cavity Cooling System Shutdown Printed February 21, 2014 at 10:34

Approved By Procedure Number Rev M. D. Askew Vogtle Electric Generating Plant 13120-1 24 Date Approved Page Number 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM 5 of 50 INITIALS 4.0 INSTRUCTIONS 4.1 CTB COOLING SYSTEM STARTUP TO STANDBY 4.1.1 Perform Table 1 to align the system handswitches for startup. ________

4.1.2 If required, perform the system startup alignment per 11120-1, "Containment Building Cooling System Alignment". ________

4.1.3 Close the links for the K2 Relay and close the breakers for the CNMT COOLING UNITS (IV REQUIRED), document on Checklist 1:

1ABE-26 ________

1ABE-27 ________

1ABC-07 ________

1ABC-08 ________

1BBE-26 ________

1BBE-27 ________

1BBC-07 ________

1BBC-08 ________

Printed February 21, 2014 at 10:34

Approved By Procedure Number Rev M. D. Askew Vogtle Electric Generating Plant 13120-1 24 Date Approved Page Number 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM 6 of 50 INITIALS NOTE Panel LR01 (1-1816-U3-019) is located in Control Building, Corridor R-149, outside the Radiochemistry Lab.

4.1.4 Open the CTB Cooling Units Outlet Dampers at Panel LR01.

Damper Position may also be verified by 1-ZLB indications on 1-QHVC. (IV REQUIRED), IF available, the IPC may also be used to verify damper position, document on Checklist 1.

DAMPER DESCRIPTION HANDSWITCH 1-QHVC 1-HV-2582A CTB CLG UNIT 1 1-HS-2582G 1ZLB-43 ________

1-HV-2582B CTB CLG UNIT 2 1-HS-2582H 1ZLB-43 ________

1-HV-2583A CTB CLG UNIT 3 1-HS-2583G 1ZLB-44 ________

1-HV-2583B CTB CLG UNIT 4 1-HS-2583H 1ZLB-44 ________

1-HV-2584A CTB CLG UNIT 5 1-HS-2584G 1ZLB-43 ________

1-HV-2584B CTB CLG UNIT 6 1-HS-2584H 1ZLB-43 ________

1-HV-2585A CTB CLG UNIT 7 1-HS-2585G 1ZLB-44 ________

1-HV-2585B CTB CLG UNIT 8 1-HS-2585H 1ZLB-44 ________

Printed February 21, 2014 at 10:34

Approved By Procedure Number Rev M. D. Askew Vogtle Electric Generating Plant 13120-1 24 Date Approved Page Number 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM 7 of 50 INITIALS 4.1.5 Open the links for the K2 Relay for the following breakers.

(IV REQUIRED), document on Checklist 1:

1ABE-26 ________

1ABE-27 ________

1ABC-07 ________

1ABC-08 ________

1BBE-26 ________

1BBE-27 ________

1BBC-07 ________

1BBC-08 ________

Printed February 21, 2014 at 10:34

Approved By Procedure Number Rev M. D. Askew Vogtle Electric Generating Plant 13120-1 24 Date Approved Page Number 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM 8 of 50 INITIALS NOTES CTB Cooling Units Outlet Dampers are OPEN and DE-ENERGIZED with their breakers LOCKED OPEN to preclude inadvertent closure of dampers.

Damper indication on 1ZLB-43 and 1ZLB-44 will have no indication after performing the following step.

Indication will still be available for damper indication on the IPC. Computer points for damper indication are; ZD9430, ZD9432, ZD9434, ZD9436, ZD9438, ZD9440, ZD9442 and ZD9444.

4.1.6 Open and lock the following breakers for the CTB Cooling Units Outlet Dampers. (IV REQUIRED), document on Checklist 1:

DESCRIPTION BREAKER CTB CLG UNIT A7-001 1-HV-2582A, 1ABE-26 ________

CTB CLG UNIT A7-002 1-HV-2582B, 1ABE-27 ________

CTB CLG UNIT A7-003 1-HV-2583A, 1BBE-26 ________

CTB CLG UNIT A7-004 1-HV-2583B, 1BBE-27 ________

CTB CLG UNIT A7-005 1-HV-2584A, 1ABC-07 ________

CTB CLG UNIT A7-006 1-HV-2584B, 1ABC-08 ________

CTB CLG UNIT A7-007 1-HV-2585A, 1BBC-07 ________

CTB CLG UNIT A7-008 1-HV-2585B, 1BBC-08 ________

Printed February 21, 2014 at 10:34

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17001-1 31.1 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 01 ON PANEL Page Number 08/16/2010 1A1 ON MCB 3 of 48 (1) (2) (3) (4) (5) (6)

CIRC WTR CIRC WTR TPCW TPCW SERVICE AIR AIR CMPSR A P-1 MOTOR P-2 MOTOR PUMP 1 PUMP 2 SWING CMPSR MSTR SEP OVERLOAD OVERLOAD TRIPPED TRIPPED MISALIGNED DISCH HI TEMP CIRC WTR CIRC WTR TPCW PMP SERVICE AIR INSTR AIR B P-1 DISCH VLV P-2 DISCH VLV DISCH HDR CMPSR EQUIP TROUBLE TROUBLE LO PRESS TROUBLE LO PRESS CIRC WTR P-1 CIRC WTR P-2 CLG TOWER UNIT 1 SERVICE AIR C LO PIT LVL LO PIT LVL BASIN HI LVL SERV AIR HDR HDR LO PRESS TIED TO UNIT 2 CIRC WTR P-1 CIRC WTR P-2 TPCCW TPCCW INSTR AIR D SCREEN WTR SCREEN WTR PUMP 1 PUMP 2 CNMT SPLY HI DIFF LVL HI DIFF LVL TRIPPED TRIPPED LINE BREAK CONDR CIRC TPCCW TPCCW CNMT E WTR SURGE TK DISCH HDR HI TEMP ISO VLV HI/LO LVL LO PRESS CLOSED CNMT F HI MSTR Printed October 2, 2013 at 9:13

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17001-1 31.1 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 01 ON PANEL Page Number 08/16/2010 1A1 ON MCB 47 of 48 WINDOW E06 ORIGIN SETPOINT CNMT 1-TSH-2563 120F HI TEMP 1-TSH-2612 1-TSH-2613 1.0 PROBABLE CAUSE Insufficient number of Containment Building Cooling Units operating.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Start an additional pair of Containment Cooling Units or a Containment Auxiliary Cooling Unit per 13120-1, "Containment Building Cooling Systems".

2. Verify Nuclear Service Cooling Water flow to coolers, and IF necessary, dispatch an operator to inspect the Containment Heat Removal System.

3. Refer to Technical Specification LCO 3.6.5 and 3.6.6.

4. IF equipment failure is indicated, initiate maintenance as required.

5.0 COMPENSATORYOPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB212, CX5DT101-66, CX5DT101-71 Printed October 2, 2013 at 9:13

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 14000-1 88.2 Effective Date Page Number 09/25/2013 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS 15 of 36 Sheet 9 of 10 DATA SHEET 1 MODE 1 & 2 MODE _______________

DATE _______________

LCO TECH SPEC I N D I C A T I O N LIMIT(S)

METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE LCO/PROC CREFS ACTUATION SR 3.3.7.1 CR INTAKE 1RE-12116 OPERABLE FCN 3 RADIATION CHANNEL CHECK 3.3.7 CHANNEL CHECK MONITORS 1RE-12117 REQUIRED 2 (INIT)

FHB ACTUATION TRS 13.3.6.1 FHB EFFL ARE-2532A 13.3.6 OPERABLE RADIOGAS

• CHANNEL CHECK FHB ISO ARE-2532B REQUIRED 1 (INIT)

FHB ACTUATION TRS 13.3.6.1 FHB EFFL ARE-2533A

• 13.3.6 OPERABLE RADIOGAS CHANNEL CHECK FHB ISO ARE-2533B REQUIRED 1 (INIT)

• INDICATING NORMALLY. ALL STATUS AND ALARM LIGHTS EXTINGUISHED.

DG1A FUEL OIL INVENTORY SR 3.8.3.1 DG 1A LEVEL 1-LI-9024 82% 3.8.3 VERIFY FUEL OIL STORAGE (%)

TANK LEVEL DG1B FUEL OIL INVENTORY SR 3.8.3.1 DG 1B LEVEL 1-LI-9025 82% 3.8.3 VERIFY FUEL OIL STORAGE (%)

TANK LEVEL TWO INDEPENDENT SR 3.7.10.1 NOTE: TEMPERATURE INDICATION IS OBTAINED FROM HAND-HELD TEST EQUIPMENT.

CONTROL ROOM EMERGENCY SR 3.7.11.1 RECORD INSTRUMENT INFORMATION BELOW.

FILTRATION SYSTEMS INSTRUMENT ID NO. N/A SHALL BE OPERABLE VERIFY CONTROL ROOM CAL DUE DATE TEMP CONTROL ROOM M&TE <85F 3.7.10 TEMPERATURE 3.7.11 (F)

THE RWST SHALL BE SR 3.5.4.1 RWST >51F

• 3.5.4 OPERABLE TRS 13.1.7.1 TEMPERATURE 1TIS-10980 13.1.7 VERIFY TEMPERATURE (F) <109F *

• WITH INDICATED RWST TEMPERATURE OUTSIDE THE LIMITS, THEN VERIFY RWST TEMPERATURE IS WITHIN TECHNICAL SPECIFICATION LIMITS BY PLACING THE RWST ON RECIRC USING SLUDGE MIXING PUMP WITH HEATER OFF AND OBSERVING 1-TI-10982 TO BE WITHIN 44F AND 116F.

THE ULTIMATE HEAT COMPUTER POINT SINK SHALL BE OPERABLE T2601* <90F 3.7.9 VERIFY WATER -OR-TEMPERATURE AND LEVEL SR 3.7.9.2 TEMPERATURE 1TJI-1692 (F) POINT 2*

COMPUTER POINT T2602*

-OR-1TJI-1692 POINT 17*

• IF COMPUTER POINT AND RECORDER POINT ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT. RECORD INSTRUMENT INFORMATION BELOW.

INSTRUMENT ID NO. N/A CAL DUE DATE 1LI-1606 >73%

SR 3.7.9.1 LEVEL

(%) 1LI-1607 CONTAINMENT AIR SR 3.6.5.1 COMPUTER POINT TEMPERATURE SHALL NOT T2501 EXCEED 120F TEMPERATURE COMPUTER POINT VERIFY AVERAGE AIR (F) T2502 NA TEMPERATURE COMPUTER POINT T2503 COMPUTER POINT 3.6.5 UT2501 (AVG) <120F

• IF COMPUTER POINT IS NOT AVAILABLE ALB-01 (E06)

VERIFY CNMT HI TEMP ALARM NOT IN ALARM ALB-01 (E06) IS NOT IN ALARM.

• IF COMPUTER POINT AND ALB-01 (E06) ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT FOR 1TE-2612 FOR POINT T2502 AND 1TE-2613. FOR POINT T2503 RECORD INSTRUMENT INFORMATION BELOW. USE MCB INDICATOR 1TI-2563 FOR POINT T2501 AND AVERAGE THE THREE.

INSTRUMENT ID NO.

<120F CAL DUE DATE COMPLETED BY: DAY: TIME: NIGHT: TIME:

SS REVIEW: DAY: TIME: NIGHT: TIME:

Printed February 21, 2014 at 11:05

Short-cylced flow with Fan #1 tripped Containment Coolers must be started in pairs due to MOV de-energized open resulting in reverse rotation and breaker tripping on start.

1. 025AG2.4.30 001/LOCT AND LOIT/SRO/C/A 2.7/4.1/025AG2.4.30/LO-TA-40002///

At time 1055:

- Unit 1 is in Mode 5.

- Containment integrity is NOT established.

- Pressurizer level is 27%.

- All Pressurizer Safety Valves are removed.

At time 1100:

- An LOSP occurs.

- DG1A trips on overspeed.

- RHR pump 'B' will NOT start following the load sequence.

- RCS temperature is 200°F and increasing.

Which one of the following completes the following statement?

At time 1115, the Shift Manager must declare a(n) __(1)__ per NMP-EP-110, "Emergency Classification Determination and Initial Action,"

and no later than time 1130, as a minimum, the __(2)__ must be notified of the declaration per NMP-EP-111, "Emergency Notifications."

REFERENCE PROVIDED

__(1)__ __(2)__

A. NOUE NRC, state, and local authorities B. NOUE state and local authorities C. Alert NRC, state, and local authorities D. Alert state and local authorities K/A 025 Loss of RHR System G2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

K/A MATCH ANALYSIS The question sets up a plausible scenario which includes the first required element of Friday, February 21, 2014 12:40:33 PM 1

the KA loss of RHR cooling, then has the candidate evaluate plant conditions and make determination of reporting requirements which meets the second part of the KA and brings the knowledge to the SRO required level.

EXPLANATION OF REQUIRED KNOWLEDGE Based on the conditions in the stem, 2 EPIPs classification thresholds have been exceed. First, an LOSP has existed with only the 'B' DG energizing 1BA03. As such NOUE CU3 has been exceeded. Additonally, due to both RHR pumps not running, RCS temperature has risen above 200F while in Mode 4. Containment integrity is not established. Since all pressurizer safeties are removed, RCS integrity is also not established. As shuch, ALERT CA4 has also been exceeded. Therefore, the Emergency Director is required to declare an ALERT emergency on or before 11:15 per NMP-EP-110.

Per NMP-EP-111 steps 5.1.1 ad 5.1.3, state and local agencies must be notified within 15 minutes of the declaration, which would be 11:30. Additionally, the NRC shall be notified immediately follwoing the state and local agencies and within an hour of the declaration.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Part 1 is incorrect but plausible because the candidate may determine the loss of power event is driving the initial classification not recognizing the loss of RHR Cooling and CU3 would be the correct.

Part 2 is incorrect but plausible because the candidate may determine that the NRC must be notified in this condition within 15 minutes. Per NMP-EP-111, Notification of the NRC shall be completed immediately following notifications to the state and local agencies and within an hour of the declaration of an emergency. Therefore, the NRC is not REQUIRED to be notified by 11:30, but instead is allowed additonal time if needed up to an hour, but is to be done as soon as possible following the state and locals.

B. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice A above.

Part 2 is correct in that within 15 minutes of the classification of the NOUE the state and local authorities must be notified per NMP-EP-111, "Emergency Notifications". Notifications of applicable State and Local Agencies shall be accomplished as soon as practicable and within 15 minutes of the declaration of an emergency, an upgrade to a higher emergency classification level, or the approval of protective actions recommendations.

C. Incorrect. Plausible. Part 1 is correct and has the candidate evaluate plant conditions and determine the event meets the ALERT threshold for CA4 due to being in Mode 5 with RCS temperature >200F without containment and RCS integrity established.

Friday, February 21, 2014 12:40:33 PM 2

without containment and RCS integrity established.

Part 2 is incorrect. See Part 2 of choice A above.

D. Correct. Part 1 is correct. See Part 1 of choice C above.

Part 2 is correct. See Part 2 of choice B above.

SRO JUSTIFICATION (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, the answer requires specific knowledge of emergency classification thresholds.

-Can the question be answered solely by knowing immediate operator actions?

No, IOAs are not addressed by this question.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, the question does not address AOP or EOP entry conditions.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, the answer requires specific knowledge of emergency classification thresholds.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. Yes, the answer requires specific knowledge of emergency classification thresholds and determination of the specific classification based on current plant conditions. This is an SRO ONLY job link associated with an SRO ONLY objective. [LO-LP-40101-13 Given an emergency scenario, and the procedure, classify the emergency (SRO only).]

Friday, February 21, 2014 12:40:33 PM 3

Level: SRO Tier # / Group # T1 / G1 K/A# 025G2.4.30 Importance Rating: 2.7 / 4.1 Technical

Reference:

NMP-EP-110-GL03, Rev 3.0, page 58 NMP-EP-110-GL03, Figure 3, Rev 3.0, page 123 NMP-EP-111, Rev 8.0, page 7 References provided: NMP-EP-110-GL03, Figure 1, Rev 3.0, page 121 NMP-EP-110-GL03, Figure 2, Rev 3.0, page 122 NMP-EP-110-GL03, Figure 3, Rev 3.0, page 123 Learning Objective: LO-TA-40002 Emergency Classification and Implementing Instructions using NMP-EP-110 (SRO Only)

LO-TA-40003 Emergency Notifications using NMP-EP-111 LO-LP-40101-13 Given an emergency scenario, and the procedure, classify the emergency (SRO only).

LO-LP-40101-16 List the state and federal authorities that are notified in an emergency.

Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments:

You have completed the test!

Friday, February 21, 2014 12:40:33 PM 4

NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 FIGURE 3 Correct Answer Distractor 123

Southern Nuclear Operating Company Emergency NMP-EP-111 Implementing Emergency Notifications Version 8.0 Procedure Page 7 of 12 4.3.3 The ENS functions DO NOT normally transfer to the EOF. The EOF has an ENS communicator position that coordinates with the TSC ENS communicator to maintain a continuous open communication line with the NRC. The role of the EOF ENS Communicator is to assist in communications with the NRCOC relevant to activities performed from the EOF (i.e., offsite interface, public information, PAR development, Dose assessment activities, etc.).

5.0 PROCEDURE 5.1 Precautions and Limitations 5.1.1 Notifications of applicable State and Local Agencies shall be accomplished as soon as practicable and within 15 minutes of the declaration of an emergency, an upgrade to a higher emergency classification level, or the approval of protective actions recommendations.

5.1.2 Electronic notification using the electronic Emergency Notification Form (ENF) in WebEOC is the preferred method of notification of State and Local Agencies. Should WebEOC be unavailable the back-up notification method is completion of a hard copy ENF and reading the form to the applicable State and Local Agencies via the ENN. Whether using the electronic or back-up ENF the emergency notification to an agency is considered complete when the agency verbally confirms receipt of the message via the ENN. To expedite availability of WebEOC in an emergency, the crew members responsible for completing the ENF and making electronic notifications should login to WebEOC as soon as possible and remain logged-in.

5.1.3 Notification of the NRC shall be completed immediately following notifications to the state and county agencies and within an hour of the declaration of an emergency. Follow-up notifications of the NRC shall be made promptly after any further degradation in the plant conditions, any change from one emergency class to another, or for the termination of an emergency. NRC notifications are typically performed utilizing the Federal Telephone System (FTS). The Emergency Notification System (ENS) line is normally utilized. An open line is maintained for the duration of the event at the request of the NRC communicator receiving the initial notification.

5.1.4 For security based emergencies, notifications to the NRC should be performed within 15 minutes of discovery of an imminent threat or attack against the plant to ensure proper mobilization of federal resources.

CAUTION An initial notification of an upgrade in emergency Classification should take precedence over a follow-up message of a lower ranking emergency. (i.e., an initial site area emergency notification takes precedence over an alert follow-up notification.)

5.1.5 If the plant condition degrades and a higher emergency classification is declared before the notifications are confirmed for the lesser emergency declaration, then a notification reflecting the higher emergency classification should be made. This notification should be made within 15 minutes of the lesser emergency declaration. This should be performed IF the notification can be made within 15 minutes of the lesser (first) classification.

1. 026AA2.06 001/LOIT AND LOCT/SRO/C/A 2.8/3.1/026AA2.06/LO-TA-60005///

At time 1000:

- Unit 1 is at 100% reactor power.

At time 1005:

- ALB04-A03 ACCW RCP 1 CLR LOW FLOW is received.

- ALB04-A04 ACCW RCP 1 CLR OUTLET HI TEMP is received.

- 18022-C, "Loss of Auxiliary Component Cooling Water," is entered.

At time 1007:

- RCP #1 seal water inlet temperature is 220°F and rising at 1°F per minute.

- RCP #1 motor stator winding temperature is 307°F and rising at 1°F per minute.

Which one of the following completes the following statement?

To prevent damage to the RCP, the RCP must be stopped no later than time __(1)__,

and after the reactor is tripped, the Shift Supervisor directs the OATC to stop the affected RCP per __(2)__ direction.

A. (1) 1012 (2) 18022-C, "Loss of Auxiliary Component Cooling Water" B. (1) 1012 (2) 19000-C, "Reactor Trip or Safety Injection" C. (1) 1016 (2) 18022-C, "Loss of Auxiliary Component Cooling Water" D. (1) 1016 (2) 19000-C, "Reactor Trip or Safety Injection" K/A 026 Loss of Component Cooling Water AA2.06 Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water:

- The length of time after the loss of CCW flow to a component Friday, February 21, 2014 1:21:51 PM 1

before that component may be damaged.

K/A MATCH ANALYSIS To answer this question, the applicant must know that loss of ACCW will lead to entry into 19000-C, requiring a manual trip (even though it does not approach or exceed any automatic setpoints), and which procedure provides the action. Since the Shift Supervisor is required to recall the specific strategy from the 18022-C procedure, the choices involve SRO only knowledge of the procedure flow path of the AOP (past the entry conditions and there are no immediate actions in 18022-C), and knowledge of the need to enter 19000-C. In addition, the question requires the candidate to recall not only how long the RCPs can operate without ACCW cooling but to evaluate other parameters that may require immediate stopping of the pumps.

EXPLANATION OF REQUIRED KNOWLEDGE At 1005, annunciators ALB04-A03 and A04 indicate a complete loss of ACCW flow to RCP #1 only. AOP 18022-C is entered. Step 6 checks to see if the RCP should be stopped. At 1016, RCP #1 would have sustained a total loss of ACCW for >10 minutes and must be stopped. The stem also states that seal water inlet temperature is rising at a rate of 1F/min. At 1018, the RCP must be stopped due to seal water inlet temperature. Additionally, motor stator winding temperature is also rising at a rate f 1F/min. At 1012, 311F would have been exceeded and the RCP must be stopped.

Therefore, the most limiting operationg condition would be the stator winding temperature and the the RCP must be stopped no later than 1012. The RCP operationg parameters in 18022-C are the same as those listed in SOP 13003-1 Limitation 2.2.10.

Since the reactor is greater than 15% RTP, the reactor must be tripped prior to tripping the RCP. Tripping the reactor is a direct entry condition into EOP 19000-C. On step 11 of 19000-C, a check of ACCW status is made and the RNO directs stopping the RCP.

However, 18022 step 6 is specifically design to trip the reactor and stop the RCP prior to performing EOP 19000-C. This is done to ensure the RCP is not damaged in the time required to perform the initial steps of 19000-C before direction to stop the RCP is encountered.

ANSWER / DISTRACTOR ANALYSIS A. Correct. Part 1 is correct. Based of Westinghouse vendor requirements the RCPs must be stopped under normal conditions when ACCW cooling is lost to the motors to prevent potential damage. This is supported by SOP 13003-1, which establishes the ACCW operating time limit of 10 minutes and other trip parameters. In the case presented to the candidate the most limiting parameter is the stator limit of 311F which will require the pump to be stopped within 5 minutes of 10:07 (i.e. 10:12).

Part 2 is correct because 18022-C directs manual reactor trip, verify trip, the stop RCPs then perform 19000-C, therefore the correct direction to stop the RCP is 18022-C as opposed to 19000-C.

Friday, February 21, 2014 1:21:51 PM 2

B. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice A above.

Part 2 is incorrect but plausible because the candidate may determine that the direction that is provided in 19000-C step 11 would be used to stop the RCPs as opposed to 18022-C step 1.

The 10 minute time allowance could lead the candidate to determine there is no real pressure to speed up the action to stop the RCPs, prior to the initial steps in 19000-C.

C. Incorrect. Plausible. Part 1 is incorrect however is plausible since the candidate may determine that none of the operating limits provided in the stem is above limits or will reach their limit within 10 minutes, therefore the 10 minutes becomes the most limiting factor.

Part 2 is correct. See Part 2 of choice A above.

D. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice C above.

Part 2 is correct. See Part 2 of choice B above.

SRO JUSTIFICATION (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, the system parameter needed to determine if the RCP should be stopped is system level knowledge. However, the direction to stop the RCP requires specific knowledge and priortization of steps in both the AOP and EOP.

-Can the question be answered solely by knowing immediate operator actions?

No, there are no IOAs associated with the actions addressed in this question.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, stopping an RCP has nothing to do with the entry conditions for either AOP 18022-C or EOP 19000-C. 19000-C does have entry conditions associated with trip of the reactor, which would stem from stopping an RCP at 100% RTP. However, this would lead a candidate down the incorrect path.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, specific knowledge of individual procedure steps is required.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency Friday, February 21, 2014 1:21:51 PM 3

contingency procedures Yes, knowledge of specific diagnotics in step 6 of AOP 18022-C is required in contrast to step 11 of EOP 19000-C.

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures Level: SRO Tier # / Group # T1 / G1 K/A# 026AA2.06 Importance Rating: 2.8 / 3.1 Technical

Reference:

EOP 19000-C Rev 37.1, page 21 AOP 18022-C Rev 15.2, page 6 SOP 13003-1, Rev 47.1, page 7 References provided: None Learning Objective: LO-TA-60005 Respond to a Loss of ACCW per 18022-C LO-LP-60318-05 Describe the operator actions required during a loss of ACCW with the plant in operation and the RCP temperature time limits are exceeded.

LO-PP-16401-07 List the RCP components that are cooled by the ACCW system.

Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments:

You have completed the test!

Friday, February 21, 2014 1:21:51 PM 4

Approved By Procedure Version M.G. Brill Vogtle Electric Generating Plant 19000-C 37.1 Effective Date Page Number E-0 REACTOR TRIP OR SAFETY INJECTION 7-5-13 21 of 35 OATC INITIAL ACTIONS Sheet 5 of 4 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 11

11. Check ACCW Pumps - AT LEAST 11. Try to start one ACCW Pump.

ONE RUNNING.

IF an ACCW Pump can NOT be started within 10 minutes of loss of ACCW, 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:

ACCW SPLY HDR ORC ISO VLV HV-1979 ACCW SPLY HDR IRC ISO VLV HV-1978 ACCW RTN HDR IRC ISO VLV HV-1974 ACCW RTN HDR ORC ISO VLV HV-1975 12

12. Adjust Seal Injection flow to all RCPs 12.

8 TO 13 GPM.

13

13. Dispatch Operator to ensure one train 13. IF one train of SFP COOLING can of SPENT FUEL POOL COOLING in NOT be restored to service, service per 13719, SPENT FUEL THEN initiate 18030-C, LOSS OF POOL COOLING AND SPENT FUEL POOL LEVEL OR PURIFICATION SYSTEM. COOLING.

END OF SUB-PROCEDURE TEXT Printed October 24, 2013 at 09:00 4

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 18022-C 15.2 Effective Date Page Number LOSS OF AUXILIARY COMPONENT COOLING 05/07/2013 WATER 6 of 13 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 6

• 6. Check if RCPs should be stopped: 6.

6.a

a. Check the following RCP a. Perform the following:

parameters (using plant 6.a.1) computer): 1) IF any parameter limit is exceeded, THEN perform Step 6.b.

Motor bearing (upper or 6.a.2) lower radial or thrust) - 2) Go to Step 7.

GREATER THAN 195°F.

Motor stator winding -

GREATER THAN 311°F.

Seal water inlet - GREATER THAN 230°F.

Loss of ACCW - GREATER THAN 10 MINUTES.

6.b

b. Perform the following: b.

6.b.1)

1) Trip the reactor. 1) 6.b.2)

2) WHEN Reactor is verified 2) tripped, THEN stop affected RCP(s).

6.b.3)

3) Initiate 19000-C, E-0 3)

REACTOR TRIP OR SAFETY INJECTION.

S Printed January 20, 2014 at 12:46

Approved By Procedure Version M.G. Brill Vogtle Electric Generating Plant 13003-1 47.1 Effective Date Page Number 06/12/2013 REACTOR COOLANT PUMP OPERATION 7 of 42 INITIALS 2.2.8 The following starting duty cycle for the RCP should be observed: ________

Only one RCP shall be started at any one time.

Two successive starts are permitted, provided the motor is permitted to coast to a stop between starts.

A third start may be made when the winding and core have cooled by running for a period of 20 minutes, or by standing idle for a period of 45 minutes.

2.2.9 During RCS filling and venting, RCS pressure must be greater than 325 psig prior to starting an RCP to verify adequate seal D/P is maintained throughout RCS fill and vent. If necessary, the RCP should be stopped prior to seal D/P dropping less than 200 psid. If the seal D/P goes below 200 psid during pump operation or coast down, the RCP should be evaluated before restarting the RCP. ________

2.2.10 An RCP shall be stopped IF any of the following conditions exist: ________

Motor bearing temperature exceeds 195°F.

Motor stator winding temperature exceeds 311°F.

Seal water inlet temperature exceeds 230°F Total loss of ACCW for a duration of 10 minutes.

RCP shaft vibration of 20 mils or greater.

RCP frame vibration of 5 mils or greater.

Differential pressure across the number 1 seal of less than 200 psid.

2.2.11 If a loss of RCP seal cooling (Seal Injection and/or ACCW to Thermal barrier) occurs, resulting in RCP shutdown due to exceeding operating limits, then the unit should be cooled down to Mode 5 to facilitate recovery. Upon reaching Mode 5, ACCW to the Thermal barrier should be restored. Seal injection should then be returned to service.

This sequence should prevent seal damage, RCP shaft bowing, ACCW System damage, etc. due to excessive thermal stresses. ________

Printed October 2, 2013 at 13:30

1. 028AG2.4.8 001/LOIT AND LOCT/SRO/C/A 3.8/4.5/028AG2.4.8/LO-TA-37021///

Initial condition:

- Unit 1 reactor trip and SI occurred.

Current conditions:

- 19011-C, "SI Termination," Step 21, is in progress to evaluate if a bubble exists in the pressurizer.

- Controlling pressurizer level channel, 1LT-459, fails low.

- Actual pressurizer level is 92% and slowly rising.

Which one of the following completes the following statement?

Per 10020-C, "EOP and AOP Rules of Usage," the Shift Supervisor __(1)__ direct the use of 18001-C, "Systems Instrumentation Malfunction," guidance to restore pressurizer heaters to service while performing 19011-C actions, and in response to actual pressurizer level, the Shift Supervisor __(2)__ required to transition to 19261-C, "Response to High Pressurizer Level."

__(1)__ __(2)__

A. may is NOT B. may is C. may NOT is NOT D. may NOT is K/A 028 Pressurizer Level Malfunction G2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

K/A MATCH ANALYSIS The question is requires the candidate to determine if AOP 18001-C can be used in conjunction with EOP 19011-C to mitigate complications arising from a failed pressurizer level instrument. The question is elevated to the SRO level by requiring the candidate to make decision on whether to direct implementation of procedures during EOP implementation.

Friday, February 21, 2014 3:33:47 PM 1

EXPLANATION OF REQUIRED KNOWLEDGE Per EOP and AOP rules of usage 10020-C step 3.5.9, other procedures such as AOPs and ARPs may be performed in parallel with EOPs as long as their actions do not conflict with the EOP steps. EOP actions take priority. In both steps 21 and 26 of EOP 19011-C, pressurizer heaters are required to be energized to saturate the pressurizer.

With pressurizer level at 92% and rising, this becomes a crucial action, however the step cannot be accomplished due to heaters being tripped off as a result of LT-459 failing low. The use of 18001-C to select away from the failed channel and restore pressurizer heaters is a prudent and necessary action.

With pressurizer level at 92% and rising, a yellow path on the CSFST for INVENTORY should exist. Per 19200-C step 1 RNO, if a yellow path exist, then initiate FRP based on plant conditions with Shift Supervisor approval. Per step 11 RNO, transition to FRP 19261-C should only be made if solid plant condtions exist, which are not currently present. Therefore, 19261-C is not required to be entered and should not be entered at this time. Instead, the priority should be placed on restoring pressurizer heaters and continuing efforts to lower pressurizer level via charging/letdown mismatch.

ANSWER / DISTRACTOR ANALYSIS A. Correct. Part 1 is correct. Per 10020-C 'EOP and AOP Rules of Usage', Other procedures such as AOPs or ARPs may be performed in parallel with EOPs as long as their actions do not conflict with the EOP steps. EOP actions take priority.

Additionally, performance of 18001-C to restore pressurizer heaters is prudent and necessary.

Part 2 is correct. 19200-C 'Critical Safety Function Status Trees', states IF a Yellow condition exists, THEN initiate FRP after evaluating plant conditions with Shift Supervisor's approval. 19011-C "SI Termination' states IF solid plant conditions are present, THEN refer to 19261 C, FR I.1 Response to High Pressurizer Level. Solid plant conditions are not present. Entry into 19261-C would actually impede the success path in 19011-C.

B. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice A above.

Part 2 is incorrect but 'plausible' because the candidates know that EOP actions always take priority. The candidate may not realize 19261-C does not contain any steps that would be different from those already in progress in 19011-C and believe transition to 19261-C is necessary to prevent going solid in the pressurizer.

C. Incorrect. Plausible. Part 1 is incorrect but plausable because 10020-C 'EOP and AOP Rules of Usage', Other procedures such as AOPs or ARPs may be performed in parallel with EOPs as long as their actions do not conflict with the EOP steps. EOP actions take priority.

Friday, February 21, 2014 3:33:47 PM 2

The candidate may not realize that the sucess path for current plant conditions resides in restoring pressuirzer heaters and deems use of 18001-C as unnecessarily slowing the progress of the EOP 19011-C.

Part 2 is correct. See Part 2 of choice A above.

D. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice C above.

Part 2 is incorrect. See Part 2 of choice B above.

SRO JUSTIFICATION (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, the question revolves around transition decisions and does not involve system knowledge.

-Can the question be answered solely by knowing immediate operator actions?

No, the question does not address any IOAs.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, specific knowledge of procedure steps contained within the EOP and AOPs is needed in addition to entry conditions.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, specific knowledge of procedure steps contained within the EOP and AOPs is needed in addition to overall knowledge of the associated procedures.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures Yes, the question requires specific knowledge of the RNO of step 21 as well as the specifics of both 18001-C and 19261-C to determine if transition to these procedures would mitigate the challenges currently observed.

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures Friday, February 21, 2014 3:33:47 PM 3

Level: SRO Tier # / Group # T1 / G2 K/A# 028G2.4.8 Importance Rating: 3.8 / 4.5 Technical

Reference:

EOP 10020-C, Rev 9.0, page 10 EOP 19011-C, Rev 29.2, page 12 & 18 EOP 19200-C, Rev 24.2, page 2, 3, & 10 References provided: None Learning Objective: LO-TA-37021 Respond to High Pressurizer Level per 19261-C LO-TA-60030 Respond to a Failure of Pressurizer Level Instrumentation per 18001-C LO-TA-37005 Terminate Safety Injection per 19000-C or 19011-C LO-TA-05003 Respond to CSFST Trouble alarm and evaluate CSFSTs using SPDS and the PSMS per 17006-1/2, 13521-1/2, 13505-1/2, and 19200-C LO-LP-60301-12 Given that the pressurizer level control selector switch is in the NORMAL position (459/460), describe how and why the plant will respond to the following instrument failures. Consider each separately and include effects on pressurizer level control, alarms, RPS, and ESF actuations.

b. 459 fails low LO-LP-37002-09 Using EOP 19200, as a guide, briefly describe how the steps are accomplished.

Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments:

You have completed the test!

Friday, February 21, 2014 3:33:47 PM 4

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 19011-C 29.2 Effective Date Page Number ES-1.1 SI TERMINATION 05/01/2013 12 of 32 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 20

20. Align CCP suction to VCT: 20.

20.a

a. Open VCT OUTLET ISOLATION a.

valves:

LV-0112B LV-0112C 20.b

b. Close RWST TO CCP A&B b.

SUCTION valves:

LV-0112D LV-0112E 21

• 21. Control PRZR pressure: 21.

21.a

a. Check Stub Busses - a. Energize Stub Busses by ENERGIZED: performing the following as necessary:

NB01 NB01 NB10 NB10 1) Open breaker 1) Open breaker NB01-01 NB10-01

2) Close breaker 2) Close breaker AA02-22 BA03-18

3) Close breaker 3) Close breaker NB01-01 NB10-01 21.b

b. Check for a bubble in the PRZR b. Control charging and to enhance RCS pressure letdown flows to avoid control. sudden pressure changes.

Energize PRZR Heaters.

IF solid plant conditions are present, THEN refer to 19261-C, FR-I.1 RESPONSE TO HIGH PRESSURIZER LEVEL.

Go to Step 22.

Step 21 continued on next page Printed September 27, 2013 at 12:07

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 19011-C 29.2 Effective Date Page Number ES-1.1 SI TERMINATION 05/01/2013 18 of 32 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 26

26. Check RCP status: 26.

26.a

a. RCPs - ALL STOPPED. a. Go to Step 27.

26.b

b. Check RVLIS full range indication b. Perform the following:

- GREATER THAN 94%.

Raise PRZR level With LT-459 failed greater than 90% [90%

low, heaters cannot ADVERSE].

be energized.

18001-C would be Raise RCS Subcooling used to select away based on core exit TCs from failed channel greater than 60°F [74°F and energize ADVERSE].

heaters.

Use PRZR Heaters, as necessary to saturate the Pressurizer water.

26.c

c. Start an RCP using c. IF an RCP can NOT be ATTACHMENT A. (RCP 4 or started, RCP 1 preferred) THEN verify natural circulation using ATTACHMENT B.

IF natural circulation NOT established, THEN raise rate of dumping steam using Steam Dumps.

After natural circulation is verified, maintain rate of dumping steam.

S Printed September 27, 2013 at 12:09

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 19200-C 24.2 Effective Date Page Number F-0 CRITICAL SAFETY FUNCTION STATUS TREES 7/25/12 2 of 11 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES If SPDS display of the Plant Computer is not operable or questionable, manual monitoring of CSFSTs should be performed by a licensed operator.

CSFSTs should be monitored continuously if a RED or ORANGE condition is present or each 10 to 15 minutes if the highest priority CSFST is no higher than YELLOW.

CSFSTs should be checked in order listed.

Priority of operator action is given by the following:

Red (Solid) Path - Extreme challenge, in Tree Order per Step 1.

Orange (Dashed) Path - Severe challenge, in the Tree Order per Step 1.

Yellow (Dotted) Path - Not satisfied, in Tree Order per Step 1.

Green (Outlined) Path - Satisfied.

If using the Plant Computer (if available) to monitor CSFSTs:

The mode indication of the Plant Computer CSFSTs should be indicating zero.

RCP breakers should be opened for RCPs NOT running in order to provide proper RVLIS indication.

If SPDS is operable, CSFSTs may be checked by scanning the display console for alarm conditions.

Color status of CSFSTs will also be indicated by letter R for red, O for orange, Y for yellow, G for green, and M for magenta.

CSFSTs will indicate active (alarming) paths as solid lines and non-active paths as empty or hollow lines.

1

1. Check CSFSTs- SATISFIED: 1. IF a Red condition exists, THEN immediately go to FRP.

1.a

a. Subcriticality (F-0.1) IF an Orange condition exists, THEN go to FRP after completion of 1.b

b. Core Cooling (F-0.2) present pass thru CSFSTs.

Step 1 continued on next page Printed February 21, 2014 at 14:30

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 19200-C 24.2 Effective Date Page Number F-0 CRITICAL SAFETY FUNCTION STATUS TREES 7/25/12 3 of 11 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1.c

c. Heat Sink (F-0.3) IF a Yellow condition exists, THEN initiate FRP after evaluating plant conditions with Shift 1.d

d. Integrity (F-0.4)

Supervisor's approval.

1.e

e. Containment (F-0.5) 1.f

f. Inventory (F-0.6) 2

2. Report change of status of any 2.

CSFST to the Shift Supervisor, if necessary (i.e., change in status not understood).

3

3. Check EOP usage - NO LONGER 3. Return to Step 1.

REQUIRED.

4

4. Monitoring of CSFSTs is no longer required.

END OF PROCEDURE TEXT Printed February 21, 2014 at 14:30

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 19200-C 24.2 Effective Date Page Number F-0 CRITICAL SAFETY FUNCTION STATUS TREES 7/25/12 10 of 11 Sheet 1 of 1 F-0.6 INVENTORY RVLIS INDICATES UPPER HEAD GREATER THAN 98 % FULL AT LEAST ONE RCP RUNNING RVLIS DYNAMIC RANGE:

> 95 % - 4 RCP

> 68 % - 3 RCP

> 46 % - 2 RCP

> 35 % - 1 RCP PRESSURIZER LEVEL LESS THAN 92%

PRESSURIZER LEVEL RVLIS INDICATES UPPER HEAD GREATER THAN 98 % FULL AT LEAST ONE RCP RUNNING RVLIS DYNAMIC RANGE:

> 95 % - 4 RCP

> 68 % - 3 RCP

> 46 % - 2 RCP

> 35 % - 1 RCP Printed February 21, 2014 at 14:30

Approved By Procedure Number Rev C.S. WALDRUP Vogtle Electric Generating Plant 10020-C 9 Date Approved Page Number 01/26/2011 EOP AND AOP RULES OF USAGE 10 of 27 3.5.8 ES-0.0, REDIAGNOSIS, may be entered any time based on operator judgement and may be entered as follows:

3.5.8.1 ES-0.0 may be entered when there is doubt in being in correct EOP.

3.5.8.2 Safety injection is in service or is required.

3.5.8.3 E-O, REACTOR TRIP OR SAFETY INJECTION, has been executed and a transition has been made to another EOP which is bounded by the ORGs only (not FRGs).

3.5.9 Other procedures such as AOPs or ARPs may be performed in parallel with EOPs as long as their actions do not conflict with the EOP steps. EOP actions take priority.

3.6 STEP PLACE-KEEPING 3.6.1 When exiting an EOP step it is necessary to track what procedure and step was exited such that when directed to "return to procedure step in affect", the correct procedure step may be re-entered. A red ribbon page marker has been provided in the Simulator and Main Control Room EOP sets for assistance in tracking such transitions.

3.6.2 Step by step place-keeping is a valuable human performance tool. It shall be performed in accordance with plant standard and management expectations.

3.7 NOTES AND CAUTIONS All NOTES and CAUTIONS shall be reviewed by the Shift Supervisor. Those NOTES or CAUTIONS that are pertinent to the evolution in progress shall be read aloud to the operating crew.

3.8 MODES OF APPLICABILITY 19000-C E-0 1,2,3 Assumes RHR system not in service and SI operable 19001-C ES-01 1,2 Assumes trip from power 19002-C ES-0.2 1,2.3 Assumes No-load conditions 19003-C ES-0.3 1,2,3 Assumes No-load conditions 19004-C ES-0.4 1,2,3 Assumes No-load conditions Printed September 27, 2013 at 12:05

1. 032AA2.03 001/LOIT AND LOCT/SRO/C/A 2.8/3.1/032AA2.03/LO-LP-39213-04///

Given the following conditions:

- Unit 1 is in Mode 6.

- Source Range N31/32 each indicate ~10 cpm.

- Source Range N31 is powered from 1AY1A.

- Source Range N32 is powered from 1NLP39.

Which one of the following completes the following statement?

Tech Spec LCO 3.9.3, "Nuclear Instrumentation," __(1)__ met, If the Source Range N32 detector high voltage power supply breaker were to trip on overcurrent, the OATC would observe the QMCB N32 meter indicating __(2)__.

__(1)__ __(2)__

A. is as-is B. is bottom of scale C. is NOT as-is D. is NOT bottom of scale K/A 032 Loss of Source Range NI AA2.03 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation:

- Expected values of source range indication when high voltage is automatically removed K/A MATCH ANALYSIS The question tests the candidate's ability to determine the indication of the Source Range NIs following the trip of the high voltage power supply breaker in Mode 6 with 10CPM indicated.

(Note: At Vogtle, Source Range NIs high voltage power supplies are no longer automatically removed - GAMAMETRICS now installed.)

EXPLANATION OF REQUIRED KNOWLEDGE Wednesday, February 26, 2014 8:20:39 AM 1

The candidate is required to know TS 3.9.3 Bases to determine the Operability of the Source Range NIs. In Mode 6, one SR NI can be powered from a Non-1E power supply provided the other SR NI is powered from it's normal 1E power supply.

Additionally, the candidate is required to determine the SR NI indication following a loss of the high voltage power supply. In this situation, the NI detector would be down powered and the indication would drop to bottom of scale.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. When any of the safety-related busses supplying power to one of the detectors (NI-0031 or NI-0032) associated with the source range neutron flux monitors are taken out of service, the corresponding source range neutron flux monitor may be considered OPERABLE when its detector is powered from a temporary nonsafety-related power source, provided the detector for the opposite source range neutron flux monitor is powered from its normal source.

The second part is incorrect. Loss of the high voltage power supply will constitute a complete loss of power to the SR NI detectors, resulting in indication failing to bottom of scale.

However, the NI's have 3 different power supplies that support various functions. If a loss of control power only occurs, all bistables trip, however NI indication is unchanged and remains "as-is". Therefore, this distractor is plausible.

B. Correct. The first part is correct. See the first part of choice A above.

The second part is correct. The loss of the High Voltage Power Supply will result in a complete loss of power to the SR NIs and indication will read bottom of scale.

C. Incorrect. Plausible. The first part is incorrect. Per TS 3.9.3 Bases, one SR NI can be powered from a Non-1E power supply provided the other SR NI is powered from it's normal 1E power supply in Mode 6. It is abnormal for a Safety-Related, Tech Spec required component to be powered from a non-1E power supply and be considered OPERABLE. It is reasonable for a candidate without knowledge of this specific exception to determine the LCO not met. Therefore, this distractor is plausible.

The second part is incorrect. See the second part of choice A above.

D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.

The second part is correct. See the second part of choice B above.

SRO JUSTIFICATION (10CFR43(b))

Wednesday, February 26, 2014 8:20:39 AM 2

(2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the knowledge required is not included in any TS or TRM action.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, the knowledge required does not exist above the line in any TS or TRM.

-Can question be answered solely by knowing the TS Safety Limits? No, SR NIs are not discussed in the TS Safety Limits.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)

• Knowledge of TS bases that is required to analyze TS required actions and terminology. Yes, specific knowledge of the Mode 6 power supply alignments listed in TS Bases 3.9.3 are required to determine if the LCO is met.

Wednesday, February 26, 2014 8:20:39 AM 3

Level: SRO Tier # / Group # T1 / G2 K/A# 032AA2.03 Importance Rating: 2.8 / 3.1 Technical

Reference:

17010-1 Rev 50 , page 60 Tech Spec Bases 3.9.3 Rev 3-4/09, page B3.9.3-1&2 References provided: None Learning Objective: LO-LP-60302-05 Describe how and why a reactor startup would be affected by a source range instrument failure when the reactor is at the following power levels: above and below P-6.

LO-PP-17201-01 Discuss the operation of the Source &

Intermediate Range Detectors to include:

a. Type of detector

b. Gamma compensation

c. When they are used

g. Power supplies (also including the effects on loss of instrument or control power)

LO-LP-39213-04 Describe the bases for any given Tech Spec in section 3.9.

Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 41.7 / 43.2 Comments:

You have completed the test!

Wednesday, February 26, 2014 8:20:39 AM 4

Nuclear Instrumentation B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Nuclear Instrumentation BASES BACKGROUND The source range neutron flux monitors are used during refueling operations to monitor the core reactivity condition. The installed source range neutron flux monitors (NI-0031 and NI-0032) are part of the Nuclear Instrumentation System (NIS). These detectors are located external to the reactor vessel and detect neutrons leaking from the core. Temporary neutron flux detectors which provide equivalent indication may be utilized in place of installed instrumentation.

The installed source range neutron flux monitors are fission chamber detectors. The detectors monitor the neutron flux in counts per second. The instrument range covers seven decades of neutron flux (1E-1 cps to 1E +6 cps) with a 2% instrument accuracy. The detectors also provide continuous visual indication in the control room.

The NIS is designed in accordance with the criteria presented in Reference 1.

APPLICABLE Two OPERABLE source range neutron flux monitors are required SAFETY ANALYSES to provide a signal to alert the operator to unexpected changes in core reactivity such as an improperly loaded fuel assembly. The need for a safety analysis for an uncontrolled boron dilution accident is minimized by isolating all unborated water sources except as provided for by LCO 3.9.2, "Unborated Water Source Isolation Valves."

The source range neutron flux monitors satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO This LCO requires that two source range neutron flux monitors be OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity. To be OPERABLE each monitor must provide visual indication.

When any of the safety-related busses supplying power to one of the detectors (NI-0031 or NI-0032) associated with the source range neutron flux monitors are taken out of service, the corresponding source range neutron flux monitor may be considered OPERABLE when its detector is powered from a temporary nonsafety-related (continued)

Vogtle Units 1 and 2 B 3.9.3-1 Rev. 3-4/09

Nuclear Instrumentation B 3.9.3 BASES LCO source of power, provided the detector for the opposite source range (continued) neutron flux monitor is powered from its normal source.

APPLICABILITY In MODE 6, the source range neutron flux monitors must be OPERABLE to determine changes in core reactivity. There are no other direct means available to check core reactivity levels. In MODES 2, 3, 4, and 5, the operability requirements for the installed source range detectors and circuitry are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation."

ACTIONS A.1 and A.2 With only one source range neutron flux monitor OPERABLE, redundancy has been lost. Since these instruments are the only direct means of monitoring core reactivity conditions, CORE ALTERATIONS and positive reactivity additions must be suspended immediately. Performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position or normal cooldown of the coolant volume for the purpose of system temperature control.

B.1 Condition B is modified by a Note to clarify the requirement that entry into or continued operation in accordance with Condition A is required for any entry into Condition B. The Note reinforces conventions of LCO applicability as stated in LCO 3.0.2 and as reflected in examples in 1.3, Completion Times.

With no source range neutron flux monitor OPERABLE, action to restore a monitor to OPERABLE status shall be initiated immediately.

Once initiated, actions shall be continued until a source range neutron flux monitor is restored to OPERABLE status.

B.2 With no source range neutron flux monitor OPERABLE, there are no direct means of detecting changes in core reactivity. However, since CORE ALTERATIONS and positive reactivity additions are not to be (continued)

Vogtle Units 1 and 2 B 3.9.3-2 Rev. 1-4/09

Approved By Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 17010-1 50 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL Page Number 08/16/2011 1C1 ON MCB 3 of 66 ALB 10 (1) (2) (3) (4) (5) (6)

A SR/IR NIS SOURCE AND POWER RANGE HI REACTOR BYPASS REACTOR BYPASS ROD CONTROL SIG PROCESSOR INTMD RANGE NEUTRON FLX HI BRKR BYA BRKR BYA NON URGENT TROUBLE TRIP BYPASS SETPOINT ALERT IN-OPERATE CLOSE FAILURE B SOURCE RNG HI POWER RANGE REACTOR BYPASS REACTOR BYPASS ROD CONTROL SHUTDOWN FLUX HI NEUTRON FLX BRKR BYB BRKR BYB URGENT FAILURE ALARM BLOCKED LOW SETPOINT IN-OPERATE CLOSE C SOURCE RANGE POWER RANGE OVERPOWER T ROD BANK RPI NIS CHANNEL HI FLUX LEVEL CHANNEL ROD BLOCK AND LO LIMIT NON URGENT ON TEST AT SHUTDOWN DEVIATION RUNBACK ALERT ALARM D INTMD RANGE PWR RANGE UP OVERPOWER ROD BANK RPI ROD DEV HI FLUX DET HI FLX DEV ROD STOP LO-LO LIMIT URGENT ALARM LEVEL ROD STOP E SR/IR REMOTE PWR RANGE LWR OVERTEMP T ROD AT BOTTOM RADIAL TILT SIG PROCESSOR DET HI FLX DEV ROD BLOCK AND DPU-B TROUBLE RUNBACK ALERT F SR/IR POWER RANGE ROD DRIVE M-G TWO OR MORE DELTA FLUX AMPLIFIER HI NEUTRON FLX SET TROUBLE RODS AT BOTTOM DEVIATION TROUBLE RATE ALERT Printed September 27, 2013 at 13:52

Approved By Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 17010-1 50 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL Page Number 08/16/2011 1C1 ON MCB 60 of 66 WINDOW F01 ORIGIN SETPOINT SR/IR NC-35M Not Applicable AMPLIFIER NC-36M TROUBLE 1.0 PROBABLE CAUSE

1. High Voltage Power Supply greater than 875V or less than 660V.

2. Loss or degraded +15v Power Supply in WR amplifier (or isolator for N36)

3. Loss or degraded -15v Power Supply in WR amplifier (or isolator for N36)

4. Degraded +5V Power Supply(s) in Isolator Assembly (N36 only) 2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS Go to 18002-C, "Nuclear Instrumentation System Malfunction".

4.0 SUBSEQUENT OPERATOR ACTIONS NONE 5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCE:

1X6AS01-154 Printed September 27, 2013 at 13:52

1. 033A2.03 001/LOIT AND LOCT/SRO/C/A 3.1/3.5/033A2.03/LO-TA-25010///

Procedure 13719-1, "Spent Fuel Pool Cooling and Purification," sections as follows:

- Section 4.2.2, "SFP Makeup from the RWST through the SFP Purification Loop"

- Section 4.2.4, "SFP Makeup from the RMWST" Initial conditions:

- Unit 1 is defueled.

- Transfer canal is drained for transfer cart inspection.

- Spent fuel shuffle is in progress in the FHB.

Current conditions:

- ALB05-E02 SPENT FUEL PIT LO LEVEL is received.

- Personnel in the FHB report SFP level is slowly lowering.

- 18030-C, "Loss of Spent Fuel Pool Level or Cooling," is entered.

Which one of the following completes the following statement?

To mitigate the consequences of the event, the Shift Supervisor is required to direct makeup to the SFP using 13719-1, Section __(1)__,

and per the Bases of Tech Spec 3.7.15, "Fuel Storage Pool Water Level," maintaining the required minimum water level in the SFP __(2)__ ensure adequate iodine decontamination factors are met for a fuel handling accident.

__(1)__ __(2)__

A. 4.2.2 does B. 4.2.2 does NOT C. 4.2.4 does D. 4.2.4 does NOT K/A 033 Spent Fuel Pool Cooling A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal spent fuel pool water level or loss of water level Friday, February 21, 2014 3:41:24 PM 1

K/A MATCH ANALYSIS The question tests the candidate's ability to predict the impact of low spent fuel pool level by having to recall the bases for TS 3.7.15 level. The candidate is also required to mitigate the consequence of the event through the selection of a makeup source by selecting the appropriate procedure section to perform.

EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17005-1, ALB05-E02 alarms at a Spent Fuel Pool Level of 217'-0". Reports from the field have verified and low and decreasing pool level. AOP 18030-C is entered to mitigate the event. Per step 6 of this procedure, makeup to the SFP per SOP 13719-1 is directed. The specific procedure section is not specified. The candidate must recognize that makeup due to leakage will be from the RWST and not the RMWST. This ensures SFP boron concentration will be maintained. This requirement is stipulated in SOP 13719-1 Precaution and Limitation 2.1.7 and in a CAUTION at the beginning of sections 4.2.3 and 4.2.4. These state that non-borated makeup is usually only allowed for normal evaporative level losses. Borated water sources are the preferred maekup source for abnormal or unexplained level losses. Objective LO-PP-25102-11 is utilized during LOIT to re-enforce the use of borated water sources only during leakage.

With SFP level less than the Tech Spec limit of 217'-0", less than 23 feet of water exist over the spent fuel stored in the racks. Per TS 3.7.15 Bases, the minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The bases also discusses TS 3.7.15 water level as providing shielding to minimize general area dose and provide shielding during spent fuel movement.

ANSWER / DISTRACTOR ANALYSIS A. Correct. The first part is correct. Per SOP 13719-1 Spent Fuel Pool Cooling and Purification System, makeup due to leakage is from a borated source.

The second part is correct. Per Technical Specification 3.7.15 Fuel Storage Pool Water Level bases, the minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.

B. Incorrect. Plausible. The first part is correct. See the first part of choice A above.

The second part is incorrect. Per Technical Specification 3.7.15 Fuel Storage Pool Water Level bases, the minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.

However, the bases also discusses the shielding function of the water, which is a more commonly known benefit. A candidate without adequate knowledge of the TS Bases, may conclude that iodine decontamination is not a factor at all for water level, Friday, February 21, 2014 3:41:24 PM 2

or may assume the normal minimum water level of 218'-0" required. Therefore, this distractor is plausible.

C. Incorrect. Plausible. The first part is incorrect. Per SOP 13719-1 Spent Fuel Pool Cooling and Purification System, makeup due to leakage is from a borated source. However, normally SFP makeup is made from either Demin Water or the RMWST since the level change is due to evaporative loss and the boron is left in solution. A candidate with inadequate knowledge of the purpose behind using the different sources would find it reasonable to makeup using a normal source. Therefore, this distractor is plausible.

The second part is correct. See the second part of choice A above.

D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.

The second part is incorrect. See the second part of choice B above.

SRO JUSTIFICATION (10CFR43(b))

(2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the question requires specific knowledge of the bases for TS 3.7.15. The immediate action of the associated RAS does not address the purpose of the level or how to restore it.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? The information above the line deals with the required level to be maintained only. It does not address the reason for the level.

-Can question be answered solely by knowing the TS Safety Limits? No, Spent Fuel Pool level is not a Safety Limit.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)

• Knowledge of TS bases that is required to analyze TS required actions and terminology. Yes, the reason SFP level is maintained >23 ft above the fuel is only specified in the Bases for TS 3.7.15.

Friday, February 21, 2014 3:41:24 PM 3

Level: SRO Tier # / Group # T2 / G2 K/A# 033A2.03 Importance Rating: 3.1 / 3.5 Technical

Reference:

SOP 13719-1 Rev 55.2, pages 6, 19, & 20 ARP 17005-1 Rev 34.2, pages 43-45 Tech Spec 3.7.15 Amendment No. 158, page 3.7.15-1 Tech Spec Bases 3.7.15 Rev 1-10/01, page B 3.7.5.15-1 References provided: None Learning Objective: LO-PP-25102-12 Describe the minimum allowable water level over spent fuel and the basis for this level.

LO-PP-25102-11 Describe when the different sources of makeup to the spent fuel pool would be used. For evaporation, For leakage LO-TA-25010 Makeup to the SFP per 13719-1/2, 13903-C, and 18030-C Attachment C Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 41.10 / 43.2 Comments: Early submittal 401-9 response:

-Need to make sure Section 4.2.4 cannot be argued as a correct answer. 13719 and EOP caution say borated water "should" rather than "shall" be used.

-Having Keff requirements as a basis for the SFP water level does not seem plausible. However, I think this can be solved by reframing the second question to state "per Bases of Tech Spec 3.7.15", maintain the required minimum water level in the SFP does/does not ensure adequate idodine decontamination factors are met.

- JAT 12/19/13 (Editorial)

The new question incorporates the above suggestion, and the first concern with the original question has a learning objective to reinforce the use of borated water sources.

Need to make sure this is definitive enough as a technical source.

- JAT 2/4/2014 You have completed the test!

Friday, February 21, 2014 3:41:24 PM 4

Fuel Storage Pool Water Level 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Fuel Storage Pool Water Level LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool water A.1 -------------NOTE--------------

level not within limit. LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies in the fuel storage pool.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft In accordance with above the top of the irradiated fuel assemblies the Surveillance seated in the storage racks. Frequency Control Program Vogtle Units 1 and 2 3.7.15-1 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

Fuel Storage Pool Water Level B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the fuel storage pool design is given in the FSAR, Subsection 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Subsection 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR, Subsection 15.7.4 (Ref. 3).

APPLICABLE The minimum water level in the fuel storage pool meets SAFETY ANALYSES the assumptions of the fuel handling accident described in Regulatory Guide 1.25 (Ref. 4). The resultant 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose per person at the exclusion area boundary is a small fraction of the 10 CFR 100 (Ref. 5) limits.

According to Reference 4, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The fuel storage pool water level satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

(continued)

Vogtle Units 1 and 2 B 3.7.15-1 Rev. 1-10/01

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 13719-1 55.2 Date Approved Page Number 03/27/2012 SPENT FUEL POOL COOLING AND PURIFICATION SYSTEM 6 of 82 INITIALS 2.0 PRECAUTIONS AND LIMITATIONS 2.1 PRECAUTIONS 2.1.1 The SFPCPS should be operated as necessary to maintain the SFP temperature below the high temperature alarm setpoint of 130°F. ________

2.1.2 The differential pressure across the SFP Skimmer Filter should not exceed 20 psid. ________

2.1.3 The differential pressure across the SFP Purification Loop cartridge filter should not exceed 70 psid. ________

2.1.4 The purification flow through the SFP Demineralizer System should not exceed 120 gpm. ________

2.1.5 Thoroughly fill and vent all applicable SFPCPS components prior to returning them to service after maintenance. This minimizes system performance degradation due to gas entrainment. ________

2.1.6 Any time that water is being removed from the RWST for makeup to the SFP or when the Refueling Water Purification Pump is taking suction from the RWST, the RWST level shall be maintained above the applicable Technical Specification low limit. ________

2.1.7 Non-borated makeup is usually only allowed for normal evaporative level losses. Borated water sources are the preferred makeup source for abnormal or unexplained level losses. ________

2.1.8 IF in Modes 1-4 opening of 1-1204-U4-158 must NOT be performed, since this would result in RWST declared inoperable ________

Printed October 2, 2013 at 14:49

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 13719-1 55.2 Date Approved Page Number 03/27/2012 SPENT FUEL POOL COOLING AND PURIFICATION SYSTEM 19 of 82 INITIALS CAUTION SFP boron concentration should be checked following makeup to assure a minimum boron concentration of 2000 ppm.

4.2.3 SFP Makeup from the Demineralized Water System (SNC16999)

(SNC12987)

CAUTION Non-borated makeup is only allowed for normal evaporative level losses.

Abnormal or unexplained level losses should be compensated for by using only borated sources.

4.2.3.1 Open SFP DEMIN WTR SPLY ISO, 1-1213-U4-055. (RA53) ________

CAUTIONS When gravity filling from the RWST, the Spent Fuel Pool level must be monitored continuously to prevent overflowing the SFP.

Spent fuel Pool lighting receptacles are at the 2189 level.

Spent fuel Pool HI Level Alarm setpoint is at 219.

4.2.3.2 Monitor Spent Fuel Pool level (see Figure 1). ________

4.2.3.3 WHEN the required water level is reached, close and lock SFP DEMIN WTR SPLY ISO, 1-1213-U4-055, (RA53);

(IV REQUIRED) ________

4.2.3.4 Request Chemistry sample the Spent Fuel Pool to determine the boron concentration. ________

Printed October 2, 2013 at 14:49

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 13719-1 55.2 Date Approved Page Number 03/27/2012 SPENT FUEL POOL COOLING AND PURIFICATION SYSTEM 20 of 82 INITIALS CAUTION SFP boron concentration should be checked following makeup to assure a minimum boron concentration of 2000 ppm.

4.2.4 SFP Makeup from RMWST (SNC16999) (SNC12987)

CAUTION Non-borated makeup is usually only allowed for normal evaporative level losses. Borated water sources are the preferred makeup source for abnormal or unexplained level losses.

4.2.4.1 Open SFP CLG RMWST ISOLATION VALVE, 1-1213-U4-054.

(RA53) ________

CAUTIONS When gravity filling from the RWST, the Spent Fuel Pool level must be monitored continuously to prevent overflowing the SFP.

Spent fuel Pool lighting receptacles are at the 2189 level.

Spent fuel Pool HI Level Alarm setpoint is at 219.

4.2.4.2 Monitor Spent Fuel Pool level (see Figure 1). ________

4.2.4.3 WHEN required water level is reached, close SFP CLG RMWST ISOLATION VALVE, 1-1213-U4-054, (RA53); (IV REQUIRED) ________

4.2.4.4 Request Chemistry sample the Spent Fuel Pool to determine the boron concentration. ________

Printed October 2, 2013 at 14:49

Approved By Procedure Version C. H. Williams, Jr. Vogtle Electric Generating Plant 17005-1 34.2 Effective Date ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON Page Number 6/21/13 PANEL 1A2 ON MCB 43 of 67 WINDOW E02 ORIGIN SETPOINT SPENT FUEL PIT 1-LSHL-625 217 feet elevation LO LEVEL 1.0 PROBABLE CAUSE

1. Insufficient inventory during filling or refueling operation.

2. Normal evaporation.

3. System leak.

4. Loss of air to the Fuel Transfer Canal and/or Cask Loading Pit Gate Seals.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE Printed December 4, 2013 at 13:27

Approved By Procedure Version C. H. Williams, Jr. Vogtle Electric Generating Plant 17005-1 34.2 Effective Date ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON Page Number 6/21/13 PANEL 1A2 ON MCB 44 of 67 WINDOW E02 (Continued) 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Dispatch an operator to determine actual level locally.

(see Figure 1 in this procedure).

2. Notify the Security Alarm Station (CAS) to dispatch a security patrol to check for any indications of sabotage.

3. Refer to 13719-1, "Spent Fuel Pool Cooling And Purification" and return the Spent Fuel Pit to normal level (218.5 feet).

4. IF level cannot be maintained greater than 217 feet with fuel movement in containment in progress or 216.5 feet with the Spent Fuel Pool Gate Valve closed, THEN suspend movement of irradiated fuel assemblies in the Spent Fuel Pool and all crane operations over the Spent Fuel Pool. Initiate 18030-C, "Loss Of Spent Fuel Pool Level Or Cooling" and 18006-C Fuel Handling Event.

5. Check service air to gate seals and refer to 13710-1, "Service Air System" to restore service air if lost.

6. Refer to Technical Specification LCO 3.7.15.

5.0 COMPENSATORY OPERATOR ACTIONS NOTE If the East and West pools are connected through the cask loading pit, Unit 1 annunciator ALB05E02 will detect a low level condition for both pools.

Verify Spent Fuel Pool Level every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per 11883-1, Radwaste Rounds Sheets.

END OF SUB-PROCEDURE

REFERENCES:

1X4DB130, PLS, 1X5DT0037, Technical Specifications LCO 3.7.15 Commitments SNC11369, 1986308950; SNC4521, 1984301472; SNC16061, 1996332947 Printed December 4, 2013 at 13:27

Approved By Procedure Version C. H. Williams, Jr. Vogtle Electric Generating Plant 17005-1 34.2 Effective Date ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON Page Number 6/21/13 PANEL 1A2 ON MCB 45 of 67 Figure 1 - Spent Fuel Pool Local Water Level Indication

1) LEVEL NUMBERS ARE PLANT ELEVATIONS IN FEET

2) POOL VOLUME APPROX.

453,000 GALS AT "N" ELEV.

3) 1 FOOT OF POOL (ONLY)

ELEVATION EQUALS APPROX.

11,408 GALS HIGH ALARM 219' NORMAL LEVEL 218' 6" 218' 217' 6" LOW ALARM 217' Printed December 4, 2013 at 13:27

1. 055G2.2.42 001/LOIT/SRO/M/F 3.9/4.6/055G2.2.42/LO-TA-60020A///

At time 1000:

- Unit 1 is at 100% reactor power.

- 18009-C, "Steam Generator Tube Leak," is in progress.

- SG sample results indicate high activity on SG #1.

At time 1020:

- 1RE-0724, Steam Line Rad Monitor, indicates 105 gpd.

- 1RE-0810, SJAE Exhaust Rad Monitor, indicates 120 gpd.

- 1RE-0724 ROC is 55 gpd/hour.

- 1RE-0810 ROC is 60 gpd/hour.

Which one of the following completes the following statement?

Per Tech Spec 3.4.13, "RCS Operational Leakage," the primary to secondary leakage

__(1)__ exceed the limit, and per 18009-C, the Shift Supervisor is required to initiate __(2)__ to lower reactor power.

__(1)__ __(2)__

A. does 18013-C, "Rapid Power Reduction" B. does 12004-C, "Power Operation (Mode 1)"

C. does NOT 18013-C, "Rapid Power Reduction" D. does NOT 12004-C, "Power Operation (Mode 1)"

K/A 055 Condenser Air Removal G2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

K/A MATCH ANALYSIS The question requires the candidate to utilize radiation monitors associated with Condenser Air Removal (1RE-0810) to recognize entry level for Tech Specs on RCS leakage. The candidate is then required to select which procedure will be utilzied based on the current primary to secondary leakage for plant shutdown.

EXPLANATION OF REQUIRED KNOWLEDGE Friday, February 21, 2014 3:43:48 PM 1

Per TS 3.4.13, 150 gpd primary to secondary LEAKAGE through any one steam generator exceeds allowable RCS operational LEAKAGE. In this condidtion, a shutdown to Mode 3 in 6 hrs and Mode 5 in 36 hrs is required. This shutdown can be accomplished using either the guidance of 12004-C or 18013-C. The decision is based on the characteristics of the leak and directed out of 18009-C.

Per 18009-C, if the tube leak is <5gpm and changing at a rate of < 30 gpd/hr, then a shutdown per 12004-C is sufficiently aggressive. If the tube leak is >5gpm or <5 gpm but changing at a rate >30 gpd/hr, then a more agressive shutdown utilizing 18013-C is necessary to ensure the plant is shutdown before the leak propogates into a rupture.

The rate of change is determined using 1RE-0724, N-16 Rad Monitor and/or 1RE-0810, SJAE Exhaust Rad Monitor, whose rate of change indications become valid after 20 mintues.

This decission tree is a complex set encompassing steps 5 thru 10 of 18009-C and is often mis-navigated. A high level understanding of steps are required to ensure internal self-checking of this operational decision.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is incorrect. Per TS 3.4.13, 150 gpd LEAKAGE in any steam generator exceeds the TS limit. Since both rad montiors have leakage below this value, the limit has not be exceeded. However, the threshold between a tube leak and a tube rupture is 120 gpm. Therefore, a candidate without sufficient knowledge of the TS limits could transpose in their minds the 120 and 150 values and conclude that the TS limit has been exceeded. Therefore, this distractor is plausible.

The second part is correct. Per 18009-C, with the leak rate <5 gpm and the rate of change > 30gpd/hr, a power reduction using 18013-C would be required.

B. Incorrect. Plausible. The first part is incorrect. See the first part of choice A above.

The second part is incorrect. Per 18009-C, with the leak rate

<5 gpm and the rate of change > 30gpd/hr, a power reduction using 18013-C would be required. However, a candidate without specific knowledge of the procedure decision tree values could conclude that the leak rate (120 gpd = 0.0833 gpm) is not sufficiently large to justify such an aggressive shutdown as 18013-C. Therefore, this distractor is plausible.

C. Correct. The first part is correct. Per TS 3.4.13, 150 gpd LEAKAGE in any steam generator exceeds the TS limit. Since both rad montiors have leakage below this value, the limit has not be exceeded.

The second part is correct. See the second part of choice A above.

Friday, February 21, 2014 3:43:48 PM 2

D. Incorrect. Plausible. The first part is correct. The first part is correct. See the first part of choice C above.

The second part is incorrect. See the second part of choice B above.

SRO JUSTIFICATION (10CFR43(b))

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, the procedure direction decision is based out of NEI guidance on leak characteristics. It is not a logic decision, it is purely based on imperical data from the industry.

-Can the question be answered solely by knowing immediate operator actions?

No, the procedure knowledge required is not an IOA.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, the procedure knowledge required is neither associated with entry conditions. It is specific to plant conditions the procedure utilizes to make operational decisions.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, detailed and not overall knowledge of steps and sequencing is required to answer the question.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Yes, the question requires the candidate to have a high level of understanding of the operational goals associated with a decision tree encompassed by 5 steps in AOP 18009-C which then determine the mitigating strategy that will be utilized.

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures Friday, February 21, 2014 3:43:48 PM 3

Level: SRO Tier # / Group # T2 / G2 K/A# 055G2.2.42 Importance Rating: 3.9 / 4.6 Technical

Reference:

TS 3.4.13, Rev Amendment No. 144, page 3.4.13-1 AOP 18009-C, Rev 29.2, page 6 & 7 References provided: None Learning Objective: LO-TA-60020A Respond to a Steam Generator Tube Leak per 18009-C LO-TA-16010 RCS Leakage Calculation (Inventory Balance) using 14905-1/2 LO-LP-39208-05 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.

b.The required actions for all section 3.4 LCOs.

LO-LP-39208-04 Describe the bases for any given Tech Spec in section 3.4.

LO-LP-37311-02 Describe the response of the following parameters to a Steam Generator Tube Rupture: (include in the discussion the response at power, during a reactor startup, and after a reactor trip/safety injection)

k. Steam jet air ejector and steam packing exhauster radiation monitor LO-LP-60309-10 Discuss how changes in the following affect radiation monitor response to a steam generator tube leak/rupture:

a. RCS activity

b. Power level

c. Process flow rate (i.e., SG blowdown)

d. Rupture size Question origin: MODIFIED - HL17 NRC Question # 37AA2.10 Cognitive Level: M/F 10 CFR Part 55 Content: 41.5 / 43.5 Comments:

You have completed the test!

Friday, February 21, 2014 3:43:48 PM 4

1. 037AA2.10 001/1/2/SGTL TECH SPEC/F 3.2/4.1/NEW/HL-17 NRC/SRO/EMT/GCW Unit 2 is experiencing a Steam Generator Tube Leak on SG 4. The crew is performing 18009-C, "Steam Generator Tube Leak".

Current conditions:

- Reactor power is 100% and stable.

- 2RE-0724 N-16 Rad Monitor indicates 155 gpd.

- 2RE-0810 SJAE Exhaust Rad Monitor indicates 160 gpd.

- RCS specific activity is 1.31 X 10-3 micro Curies per gram DOSE EQUIVALENT I-131.

Based on these conditions, which one of the following correctly completes the following statement?

Per Tech Spec 3.4.13, "RCS Operational Leakage", the primary to secondary leakage is ___(1)___ the limit and per the Tech Spec Bases 3.4.17, "Steam Generator Tube Integrity," the limit ensures that under the stress of a LOCA or MSLB a single crack leaking this amount will not

___(2)___ .

A. (1) within (2) exceed the limits for secondary coolant activity B. (1) within (2) propagate to a SGTR C. (1) exceeding (2) exceed the limits for secondary coolant activity D. (1) exceeding (2) propagate to a SGTR Monday, January 20, 2014 2:25:25 PM 1

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE; and

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

Vogtle Units 1 and 2 3.4.13-1 Amendment No. 144 (Unit 1)

Amendment No. 124 (Unit 2)

Approved By Procedure Version J Thomas Vogtle Electric Generating Plant 18009-C 29.2 Effective Date Page Number STEAM GENERATOR TUBE LEAK 08/16/2012 6 of 34 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE If available, both RE-0810 and RE-0724 should be used to determine the leakage rate of change in Step 7 RNO; however, if only one of the two radiation monitors is Functional, then the reading from the Functional monitor should be used to determine leakage rate of change.

6

6. Check Radiation monitors available: 6. Go to Step 8 RE-810 OR RE-724 7.

7

7. Check leakage rate of change: 7.

7.a

a. Greater than or equal to 30 a. Perform the following:

GPD/HR based on a 20 minute 7.a.1) trend: 1) After a 20 minute trend has elapsed, determine the leakage rate of change.

IPC Points:

IF leakage rate of RE-0810: UR6810(GPD) change is greater than UR6811(ROC) or equal to 30 gpd/hr, THEN go to Step 8.

RE-0724: UR6724(GPD)

-OR-UR6725(ROC)

IF leakage rate of change is less than 30 gpd/hr, THEN go to Step 9.

Distractor flow path S

Printed January 20, 2014 at 14:32

Approved By Procedure Version J Thomas Vogtle Electric Generating Plant 18009-C 29.2 Effective Date Page Number STEAM GENERATOR TUBE LEAK 08/16/2012 7 of 34 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 8

8. Check leakage rate - LESS THAN 8. Perform the following:

75 GPD. 8.a

a. Initiate 18013-C, RAPID POWER REDUCTION.

8.b

b. Be less than 50% power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

8.c

c. Be in Mode 3 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

8.d

d. Go to Step 12.

9

9. Check leakage rate - LESS THAN 9. Perform the following:

150 GPD. 9.a

a. Initiate 12004-C, POWER OPERATION (MODE 1).

9.b

b. Be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

9.c

c. Go to Step 12.

10

10. Check leakage rate - LESS THAN 10. IF leakrate has remained greater 75 GPD. than or equal to 75 gpd for one hour, THEN perform the following:

10.a

a. Initiate 12004-C, POWER OPERATION (MODE 1).

10.b

b. Be in Mode 3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

10.c

c. Go to Step 12.

S Printed January 20, 2014 at 14:32

1. 058AA2.03 001/LOIT AND LOCT/SRO/C/A 3.5/3.9/058AA2.03/LO-TA-60040A//HL15/

Procedure titles are as follows:

- 18034-1, "Loss of Class 1E 125 VDC Power"

- 19000-C, "Reactor Trip or Safety Injection" Initial condition:

- Unit 1 is at 100% reactor power.

Current conditions:

- All Train 'A' MSIV red and green handswitch lights extinguish.

- RTB 'A' red and green lights extinguish.

- RCP #1 1E breaker red and green handswitch lights extinguish.

- Channel I TSLB bistable lights illuminate.

Which one of the following completes the following statement?

The Shift Supervisor __(1)__ required to enter 18034-1, and the Shift Supervisor __(2)__ required to enter 19000-C.

__(1)__ __(2)__

A. is is B. is is NOT C. is NOT is D. is NOT is NOT K/A 058 Loss of DC Power AA2.03 Ability to determine and interpret the following as they apply to the Loss of DC Power:

- DC loads lost; impact on ability to operate and monitor plant systems.

K/A MATCH ANALYSIS The question tests the candidates ability to relate multiple indications and the immediate impact to plant operations. They must interpret these various indications Friday, February 21, 2014 3:47:35 PM 1

and make a decision on which procedures would address the problem.

EXPLANATION OF REQUIRED KNOWLEDGE The indications given are the symptoms of a loss of power to 125VDC bus 1AD1. All 'A' train 1E switchgear breakers will loose control power. As such, the breaker will remain in its current state without electrical protection and all handswitch indication lights will be de-energized. The MSIVs and MFIVs will fail CLOSED as their solenoids are de-energized resulting in a Reactor Trip. All Channel I TSLBs lights will illuminate due to the loss of 1AY1A, which is normally feed from an inverter supplied by 1AD1.

Entry conditions for 18034-1 are met. Step 1 of 18034-1 directs a reactor trip and INITIATION of 19000-C. Per Admin procedure 10020-C step 3.3, "Initiate" means the referenced procedure will be used as a supplement to, and it will be performed concurrently with the one in effect. Therefore, 18034-C and 19000-C are expected to be worked in conjunction due to the complications resulting from a loss of DC.

18034-1 will address all issues required by the loss of the supported 120V Vital AC bus.

ANSWER / DISTRACTOR ANALYSIS A. Correct. The first part is correct. Entry conditions for 18034-1 are met.

The second part is correct, Step 1 of 18034-1 directs a reactor trip and INITIATION of 19000-C.

B. Incorrect. Plausible. The first part is correct. Entry conditions for 18034-1 are met.

The second part is incorrect. Per step 1 RNO of 18034-1, a reactor trip should have occured and is required. However, step 1 is not an IOA. A candidate without specific knowledge of the procedure and who did not realize the MISIVs could conclude the reactor did not trip and entry into 19000-C is not required. Therefore, this distractor is plausible.

C. Incorrect. Plausible. The first part is incorrect. The loss of indication on the associated handswitches and TSLBs indicate a loss of 1AD1.

However, a candidate with insufficient knowledge of power supplies could conclude that the handswitches are feed from 120V Vital AC and not the 125VDC supply. This is a common misconception. As such, it would be reasonable for the candidate to conclude a loss of 1AY1A to have occurred and entry into 18034-1 is not required. Therefore, this distractor is plausible.

The second part is correct. Step 1 of 18034-1 directs a reactor trip and INITIATION of 19000-C. However, a candidate that believes 18032-1 was entered instead of 18034-1, would still find it reasonable that the reactor would trip based on other trips. Therefore, this distractor is plausible.

D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.

Friday, February 21, 2014 3:47:35 PM 2

The second part is incorrect. Step 1 of 18034-1 directs a reactor trip and INITIATION of 19000-C. However, a candidate that believes 18032-1 was entered instead of 18034-1 and has knowledge of 18032-1, would not expect the reactor to trip.

SRO JUSTIFICATION (10CFR43(b))

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, selection of the appropriate AOP/EOP combination is NOT associated with system knowledge.

-Can the question be answered solely by knowing immediate operator actions?

No, the direction to initiate 19000-C concurrent with 18034-C is the RNO of step 1 of 18034-C and is not an IOA.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, the entry condition for each AOP/EOP is part of the question. However, the specific knowledge of 18034-C step 1 RNO is required. Normally, AOP use is discontinued upon entry into 19000-C. Most AOPs that direct tripping the reactor say "go to 19000-C", indicating that a transition to the EOP network is made. There are only a select few AOPs that required concurrent implementation of the AOP with the EOP. In most situation, the AOP is only performed at SS discretion and is typically not needed. 19000-C is designed to be successful with only one train of electrical power.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, specific knowledge of 18034-C step 1 RNO is required.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Yes, the candidate is required to select the appropriate procedure/combination of procedures to mitigate the plant conditions.

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures Level: SRO Tier # / Group # T1 / G1 K/A# 058AA2.03 Importance Rating: 3.5 / 3.9 Friday, February 21, 2014 3:47:36 PM 3

Technical

Reference:

AOP 18034-1 Rev 13.1, page 4 Admin Proc 10020-C Rev 9.0, page 8 References provided: None Learning Objective: LO-LP-60329-01 Given that a loss of power has occurred to any of the following 125VDC vital buses and given the appropriate plant procedures, describe the operator actions required and why these actions are taken.

a. 1AD1 LO-LP-60329-04 Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable).

LO-TA-60040A Respond to a Loss of Class 1E 125 VDC Power per 18034-1/2.

Question origin: BANK - HL15 NRC Question # 058AA2.03 Cognitive Level: C/A 10 CFR Part 55 Content: 41.7 / 41.10 / 43.5 Comments: Question appears to match the KA. Not sure if the procedure question is at the SRO-only level. The second question is systems knowledge and not at the SRO-only level.

Choices A-D have 4 different answers; the applicant does not need to know the answer to the second question to answer the question correctly.

The explanation states that an automatic reactor trip has occurred, however, the justification for the correct answer is that the RNO step for "Verify Reactor Trip," states to initiate 19000-C. Does this mean that, although it occurred automatically, Reactor Trip cannot be verified?

One possible fix could be dropping the second question (and anything in the stem that was solely used to answer the second question) and separate the first question to ask:

"the Shift Supervisor <is/is not> required to enter 18034-1 and the Shift Supervisor <is/is not> required to enter 19000-C."

(I still need to think about whether asking it this way is at the SRO-only level. I'm leaning towards it IS at the SRO-only level.)

- JAT 12/19/2013 (Editorial)

New question incorporates the above comments. SRO-only Friday, February 21, 2014 3:47:36 PM 4

appears to be met because the question requires knowledge of procedure rules-of-usage.

- JAT 2/4/14 You have completed the test!

Friday, February 21, 2014 3:47:36 PM 5

Approved By Procedure Number Rev J. Thomas Vogtle Electric Generating Plant 18034-1 13.1 Date Approved Page Number 3/16/12 LOSS OF CLASS 1E 125V DC POWER 1 of 84 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure provides operator actions to be followed in the event that power is lost to one of the 125V DC Vital Busses (1AD1, 1BD1, 1CD1, or 1DD1).

Specific instructional steps will be found in the following sections:

A. LOSS OF 125V DC BUS 1AD1 B. LOSS OF 125V DC BUS 1BD1 Handswitch lights C. LOSS OF 125V DC BUS 1CD1 for RCP#1 and RTB 'A' D. LOSS OF 125V DC BUS 1DD1 extinguished SYMPTOMS Channel 1 TSLB illuminated SECTION A. LOSS OF 125V DC BUS 1AD1 125V DC Vital Bus 1AD1 voltage low.

Loss of power to 1AY1A and 1AY2A 120V AC Vital Instrument Panels.

Loss of indicating lights on 1AA02, 1AB04, 1AB05, and 1AB15 Switchgear Controls.

Train A Main Steamline Isolation.

Train A Main Feedwater Isolation. No handswitch lights on MSIV 'A' SECTION B. LOSS OF 125V DC BUS 1BD1 125V DC Vital Bus 1BD1 voltage low.

Loss of power to 1BY1B and 1BY2B 120V AC Vital Instrument Panels.

Loss of indicating lights on 1BA03, 1BB06, 1BB07, and 1BB16 Switchgear Controls.

Train B Main Steamline Isolation.

Train B Main Feedwater Isolation.

Printed January 21, 2014 at 12:44

Approved By Procedure Number Rev J. Thomas Vogtle Electric Generating Plant 18034-1 13.1 Date Approved Page Number LOSS OF CLASS 1E 125V DC POWER 3/16/12 4 of 84 A. LOSS OF 125V DC BUS 1AD1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES This procedure should be performed concurrent with 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

RCP 1 undervoltage and underfrequency trips will NOT actuate.

See ATTACHMENT A for equipment responses, breaker and valve control loss, valve failures from loss of instrument air, and annunciator failures.

A1

__A1. Verify reactor trip. A1. Perform the following:

A1.a

__a. Trip the reactor.

A1.b

__b. Initiate 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

A2

__A2. Initiate the Continuous Actions Page. A2.

A3

__A3. Dispatch an operator to 1AA02 A3.

SWGR Room (CB-A48).

NOTE IF DG1A is NOT running, it can NOT be started.

A4

__A4. Check DG1A - RUNNING. __A4. Go to Step A7.

S Printed November 25, 2013 at 12:20

Approved By Procedure Number Rev C.S. WALDRUP Vogtle Electric Generating Plant 10020-C 9 Date Approved Page Number 01/26/2011 EOP AND AOP RULES OF USAGE 8 of 27 3.2 GO TO STEPS To maintain consistency in referencing or branching to another procedure:

3.2.1 "Go to" is used when it is desired to branch to another procedure or to a preceding step in the procedure.

Example: IF the reactor trips, THEN go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

Branching implies the procedure in use shall be exited and a new procedure entered.

3.2.2 "return to" is used when it is desired to branch to a previous step in the procedure.

3.3 BY INITIATING STEPS When "by initiating" is used, the referenced procedure will be used as a supplement to, and it will be performed concurrently with the one in effect.

3.4 IMMEDIATE OPERATOR ACTIONS STEPS 3.4.1 These are actions that, for EOPs are to be committed to memory for immediate performance upon initiation of the procedure. These actions, which typically involve verification of automatic actions, are listed starting on top of the next page after the symptoms section with "IMMEDIATE OPERATOR ACTIONS" typed above Step 1.

3.4.2 Immediate Operator Action Steps shall be performed by memory by the operator.

The Unit Shift Supervisor will state the high level steps as written in the procedure. Upon restatement the operator will repeat the step including all substeps to ensure completeness.

3.4.3 All EOP immediate actions must be completed prior to taking any early action or non-EOP action.

Printed December 4, 2013 at 14:49

1. 062AG2.2.12 001/LOIT/SRO/C/A 3.7/2.7/062AG2.2.12/LO-LP-39204-04///

Initial condition:

- Unit 1 is at 100% reactor power.

Current conditions:

- The Shift Supervisor discovers the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Channel Check for Train 'A' NSCW basin level, 1LI-1606, was missed.

- The last performance of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Channel Check was 0030 on 5-11-14.

- Time of discovery of the missed surveillance was 1700 on 5-12-14.

- A risk evaluation will NOT be performed.

Which one of the following completes the following statement?

To prevent declaring the LCO NOT met, the surveillance is required to be performed satisfactorily no later than ________.

A. 0030 on 5-12-14 B. 0630 on 5-12-14 C. 1700 on 5-13-14 D. 0100 on 5-14-14 K/A 062 Loss of NSCW G2.2.12 Knowledge of surveillance procedures.

K/A MATCH ANALYSIS The question tests the candidates knowledge of generic TS survelliance SR 3.0.3 as specifically applied to a missed NSCW survelliance. If the missed survelliance results in an inoperable declaration, a loss of one train of NSCW would occur since the survelliance affects the Ultimate Heat Sink LCO 3.7.8.

EXPLANATION OF REQUIRED KNOWLEDGE A Surveillance has been identified on an NSCW System as being missed and the candidate must use Tech Spec SR 3.0.3 to determine when the missed surveillance is to be performed by. If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow Monday, February 24, 2014 9:43:58 AM 1

performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Incorrect but plausible because the candidate may determine the surveillance must be perform within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the time missed as opposed to time of discovery.

B. Incorrect. Plausible. Incorrect but plausible because the candidate may determine the surveillance must be perform from time missed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> plus the 125% grace period described in SR 3.0.2. The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

C. Correct. The answer is correct the missed surveillance must be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of point of discovery. (e.g. If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.)

D. Incorrect. Plausible. Incorrect but plausible because the candidate may determine the surveillance must be perform from point of discovery within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> plus the 125% grace period described in SR 3.0.2.

The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

SRO JUSTIFICATION (2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the question is not related to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action time requirements.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, the question is not related to above-the-line information.

Monday, February 24, 2014 9:43:58 AM 2

-Can question be answered solely by knowing the TS Safety Limits? No, the question is not related to Safety Limits.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes, the required knowledge is application of SR 3.0.3 in Tech Spec.

• Knowledge of TS bases that is required to analyze TS required actions and terminology Level: SRO Tier # / Group # T1 / G1 K/A# 062G2.2.12 Importance Rating: 3.7 / 4.1 Technical

Reference:

OSP 14000-1, Rev 88.1, page 15 TS SR 3.0.3, Amendment No. 125, page 3.0-4 Surv Frequency Control Program, Rev 3, page 15 References provided: None Learning Objective: LO-LP-39204-04 State the allowable time intervals for extension of surveillances. State the result of failure to perform surveillances within this period.

LO-LP-39204-06 In regard to surveillances, determine when time delay may be applied and the maximum time allowed to perform the surveillance.

Question origin: BANK Cognitive Level: C/A 10 CFR Part 55 Content: 41.10 / 43.2 Comments:

You have completed the test!

Monday, February 24, 2014 9:43:58 AM 3

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.

Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per . . ."

basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

(continued)

Vogtle Units 1 and 2 3.0-4 Amendment No. 125 (Unit 1)

Amendment No. 103 (Unit 2)

VEGP - Surveillance Frequency Control Program Surveillance LDCR Frequency Notes Requirement No.

SR 3.7.7.2 18 months on a STAGGERED 2012-015 TEST BASIS SR 3.7.8.1 31 days N/A SR 3.7.8.2 18 months For the following N/A components only:

1HV1668A 2HV1668A 1HV1668B 2HV1668B 1HV1669A 2HV1669A 1HV1669B 2HV1669B 18 months on a STAGGERED For the following 2012-015 TEST BASIS components only:

1HV1806 2HV1806 1HV1808 2HV1808 1HV1822 2HV1822 1HV1830 2HV1830 1HV2134 2HV2134 1HV2138 2HV2138 1HV1807 2HV1807 1HV1809 2HV1809 1HV1823 2HV1823 1HV1831 2HV1831 1HV2135 2HV2135 1HV2138 2HV2138 SR 3.7.8.3 18 months on a STAGGERED 2012-015 TEST BASIS SR 3.7.9.1 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> N/A SR 3.7.9.2 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> N/A SR 3.7.9.3 31 days N/A SR 3.7.9.5 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> N/A Page 15 of 19 Revision 3

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 14000-1 88.1 Effective Date Page Number 06/21/2013 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS 15 of 36 Sheet 9 of 10 DATA SHEET 1 MODE 1 & 2 MODE _______________

DATE _______________

LCO TECH SPEC I N D I C A T I O N LIMIT(S)

METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE LCO/PROC CREFS ACTUATION SR 3.3.7.1 CR INTAKE 1RE-12116 OPERABLE FCN 3 RADIATION CHANNEL CHECK 3.3.7 CHANNEL CHECK MONITORS 1RE-12117 REQUIRED 2 (INIT)

FHB ACTUATION TRS 13.3.6.1 FHB EFFL ARE-2532A 13.3.6 OPERABLE RADIOGAS

• CHANNEL CHECK FHB ISO ARE-2532B REQUIRED 1 (INIT)

FHB ACTUATION TRS 13.3.6.1 FHB EFFL ARE-2533A

• 13.3.6 OPERABLE RADIOGAS CHANNEL CHECK FHB ISO ARE-2533B REQUIRED 1 (INIT)

• INDICATING NORMALLY. ALL STATUS AND ALARM LIGHTS EXTINGUISHED.

DG1A FUEL OIL INVENTORY SR 3.8.3.1 DG 1A LEVEL 1-LI-9024 82% 3.8.3 VERIFY FUEL OIL STORAGE (%)

TANK LEVEL DG1B FUEL OIL INVENTORY SR 3.8.3.1 DG 1B LEVEL 1-LI-9025 82% 3.8.3 VERIFY FUEL OIL STORAGE (%)

TANK LEVEL TWO INDEPENDENT SR 3.7.10.1 NOTE: TEMPERATURE INDICATION IS OBTAINED FROM HAND-HELD TEST EQUIPMENT.

CONTROL ROOM EMERGENCY SR 3.7.11.1 RECORD INSTRUMENT INFORMATION BELOW.

FILTRATION SYSTEMS INSTRUMENT ID NO. N/A SHALL BE OPERABLE VERIFY CONTROL ROOM CAL DUE DATE TEMP CONTROL ROOM M&TE <85F 3.7.10 TEMPERATURE 3.7.11 (F)

THE RWST SHALL BE SR 3.5.4.1 RWST >51F

• 3.5.4 OPERABLE TRS 13.1.7.1 TEMPERATURE 1TIS-10980 13.1.7 VERIFY TEMPERATURE (F) <109F *

• WITH INDICATED RWST TEMPERATURE OUTSIDE THE LIMITS, THEN VERIFY RWST TEMPERATURE IS WITHIN TECHNICAL SPECIFICATION LIMITS BY PLACING THE RWST ON RECIRC USING SLUDGE MIXING PUMP WITH HEATER OFF AND OBSERVING 1-TI-10982 TO BE WITHIN 44F AND 116F.

THE ULTIMATE HEAT COMPUTER POINT SINK SHALL BE OPERABLE T2601* <90F 3.7.9 VERIFY WATER -OR-TEMPERATURE AND LEVEL SR 3.7.9.2 TEMPERATURE 1TJI-1692 (F) POINT 2*

COMPUTER POINT T2602*

-OR-1TJI-1692 POINT 17*

• IF COMPUTER POINT AND RECORDER POINT ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT. RECORD INSTRUMENT INFORMATION BELOW.

INSTRUMENT ID NO. N/A CAL DUE DATE 1LI-1606 >73%

SR 3.7.9.1 LEVEL

(%) 1LI-1607 CONTAINMENT AIR SR 3.6.5.1 COMPUTER POINT TEMPERATURE SHALL NOT T2501 EXCEED 120F TEMPERATURE COMPUTER POINT VERIFY AVERAGE AIR (F) T2502 NA TEMPERATURE COMPUTER POINT T2503 COMPUTER POINT 3.6.5 UT2501 (AVG) <120F

• IF COMPUTER POINT IS NOT AVAILABLE ALB-01 (E06)

VERIFY CNMT HI TEMP ALARM NOT IN ALARM ALB-01 (E06) IS NOT IN ALARM.

• IF COMPUTER POINT AND ALB-01 (E06) ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT FOR 1TE-2612 FOR POINT T2502 AND 1TE-2613. FOR POINT T2503 RECORD INSTRUMENT INFORMATION BELOW. USE MCB INDICATOR 1TI-2563 FOR POINT T2501 AND AVERAGE THE THREE.

INSTRUMENT ID NO.

<120F CAL DUE DATE COMPLETED BY: DAY: TIME: NIGHT: TIME:

SS REVIEW: DAY: TIME: NIGHT: TIME:

Printed January 3, 2014 at 14:02

1. 076A2.02 001/LOIT AND LOCT/SRO/C/A 2.7/3.1/076A2.02/LO-TA-60003/63013///

Initial conditions:

- Unit 1 is at 100% reactor power.

- SI Pump 'A' is tagged out.

Current conditions:

- Train B NSCW supply header pressure is 75 psig and lowering.

- Train 'B' NSCW supply header flow is 25,000 gpm.

- Train 'B' NSCW return header flow is 10,000 gpm.

- 18021-C, "Loss of Nuclear Service Cooling Water System," is entered.

Which one of the following completes the following statement?

Per 18021-C, the crew is required to __(1)__ the standby Train B NSCW pump, and after completing the actions of 18021-C, the Shift Supervisor will determine per 10008-C, "Recording Limiting Conditions for Operation," that a LOSF __(2)__ exist.

__(1)__ __(2)__

A. start does B. start does NOT C. place in PTL does D. place in PTL does NOT K/A 076 Service Water A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

- Service water header pressure K/A MATCH ANALYSIS The question addresses a problem identified on an NSCW System to include low Monday, February 24, 2014 2:02:36 PM 1

header pressure and the candidate must determine the correct procedure action based on the indications provided. In addition, the candidates must determine the impact on plant operations using the information provided in the stem as related to LOSF evaluation. This concept brings the question to the SRO knowledge level.

EXPLANATION OF REQUIRED KNOWLEDGE Per 18021-C symptoms, a drop in NSCW header pressure accompanied by a large difference between supply and return header flows indicates a large (catastrophic) leak.

Per steps 1 and 2, all pumps in the affected train will be placed in PTL.

Since NSCW 'B' is a support system for all 'B' train ECCS pumps, a LOSF function exists with SIPs. With SIP 'A' tagged out in the stem and a subsequent loss of SIP 'B' due to loss of a support system, no medium head injection is available. Since TS 3.5.2 does not have a condition for two trains of ECCS inoperable, TS 3.0.3 must be entered.

(Reference 10008-C for LOSF evaluation guidance.)

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Part 1 is incorrect however plausible since the candidate may believe the stem indications are due a pump problem such as a broken coupling, as opposed to a leak since the symptoms would be the same with exception of the supply and return flow deviation. Per 18021-C step 6, the correct action for this condition would be to start the standby pump.

Part 2 is correct. The candidate should determine that with both Train A SI Pump and Train B NSCW System inoperable a LOSF exists per 10008-C 'Recording Limiting Conditions for Operation'. However, if the candidate misses the catastrophic leak in Part 1, it would still be plausible for them to incorrectly determine that the standby NSCW pump is inoperable due to the failure to start on low header pressure and determine a LOSF exists for the wrong reason.

B. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice A above.

Part 2 is incorrect. The candidate should determine that with both Train A SI Pump and Train B NSCW System inoperable a LOSF exists per 10008-C 'Recording Limiting Conditions for Operation'. However, if the candidate misses the catastrophic leak in Part 1 and recognizes that the failure of the start on low header pressure in not a required function, then determining that a LOSF does not exist would be correct for these erroneous conditions..

C. Correct. Part 1 is correct. The candidate should determine that from the stem information provided that a large leak has occurred in the NSCW piping and the correct actions per 18021-C Loss of NSCW, is to place the affected pumps in PTL.

Monday, February 24, 2014 2:02:36 PM 2

Part 2 is correct. The candidate should determine that with both Train A SI Pump and Train B NSCW System inoperable a LOSF exists per 10008-C 'Recording Limiting Conditions for Operation'.

D. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice C above.

Part 2 is incorrect however plausible since the candidate may not make the operability connection between the two systems or assume that NSCW could be run in single pump operations to supply cooling and therefore believe a LOSF condition is not present.

SRO JUSTIFICATION (2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the question only addresses TS action of 72 hrs in association with implementation of TS 3.0.3.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, all necessary knowledge is below the line.

-Can question be answered solely by knowing the TS Safety Limits? No, the question does not address Safety Limits.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes, the candidate is required to determine the applicability of LCO 3.0.3 as applied to a LOSF.

• Knowledge of TS bases that is required to analyze TS required actions and terminology (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, system knowledge will not address the operability/safety function determination.

-Can the question be answered solely by knowing immediate operator actions?

No IOA's are addressed by the question.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, knowledge of an administrative process is required associated with Tech Specs.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, sequence or overall stategy of 18021-C will not answer either part of the question.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with Monday, February 24, 2014 2:02:36 PM 3

which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures Yes, specific knowledge of Admin procedure 10008-C is required to perform a LOSF evaluation to determine which LCO(s) are not met based on the inoperability of support systems and the impact to pre-existing inoperabilities.

Level: SRO Tier # / Group # T2 / G1 K/A# 076A2.02 Importance Rating: 2.7 / 3.1 Technical

Reference:

ADMIN 10008-C, Rev 30.0, pages 1-6, 15-19, & 28 AOP 18021-C, Rev 19.0, pages 1-3 TS 3.5.2, Amendment No. 136, page 3.5.2-1 References provided: None Learning Objective: LO-LP-63508-04 Define the following terms per 10008-C.

a. LCO

d. Loss of Safety Function

e. Supported System

f. Support System LO-PP-06101-04 Describe the indications of the following:

c. Supply header leak

d. Return header leak

e. NSCW pump trip

f. NSCW piping leak in a pump room LO-TA-63013 Implement Technical Specification LCO using 10008-C (SRO Only)

LO-TA-60003 Respond to a Loss of NSCW per 18021-C Question origin: MODIFIED - HL15 Question # 062AA2.02 Cognitive Level: C/A 10 CFR Part 55 Content: 41.4 / 41.10 / 43.2 / 43.5 Comments:

You have completed the test!

Monday, February 24, 2014 2:02:36 PM 4

1. 062AA2.02 001/1/1/LOSS OF NSCW/C/A - 3.6/NEW/HL15/SRO/DS/TNT Given the following conditions at 38% power:

- ACCW pump 2 is in service

- CCW pumps 2 & 4 are in service

- NSCW Pump 5 is danger tagged Train A NSCW indications: Train B NSCW indications:

- Supply header pressure 45 psig - Supply header pressure 58 psig

- Supply header flow 8,000 gpm - Supply header flow 25,000 gpm

- Return header flow 8,000 gpm - Return header flow 10,000 gpm Which of the following choices contains the correct procedural entry and actions?

A. Enter AOP 18021-C, Loss of NSCW, due to loss of both NSCW Trains.

Place all NSCW pumps in PTL, trip the reactor and initiate EOP 19000-C. Trip the RCPs and isolate CVCS letdown.

B. Enter AOP 18021-C, Loss of NSCW, due to leakage on Train A NSCW.

Place all Train A NSCW pumps in PTL, trip the reactor and initiate EOP 19000-C.

Trip the RCPs and isolate CVCS letdown if cooling not restored in 10 minutes.

C. Enter AOP 18021-C, Loss of NSCW, due to leakage on Train B NSCW.

Place all Train B NSCW pumps in PTL. Shift to Train A CCW pumps. Start ACCW pump #1 and remain in 18021-C.

D. Enter AOP 18021-C, Loss of NSCW, due to loss of both NSCW Trains.

Place NSCW Train B in single pump operation and all Train A NSCW pumps in PTL. Trip RCPs if seal temperatures exceed 230 F and remain in 18021-C.

Tuesday, January 21, 2014 1:35:43 PM 1

ECCS - Operating 3.5.2 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.

---

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

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 REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> No condition exists for 2 trains inop -

TS 3.0.3 must be entered.

Vogtle Units 1 and 2 3.5.2-1 Amendment No. 136 (Unit 1)

Amendment No. 115 (Unit 2)

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 18021-C 19 Effective Date Page Number LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM 11/09/2012 1 of 15 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure addresses the loss or degraded operation of one or more trains of Nuclear Service Cooling Water.

SYMPTOMS Trip of operating NSCW pumps and failure of standby pump to start.

Dropping NSCW Supply Header pressure.

Large difference between Supply Header flow and Return Header flow, indicating a large leak.

NSCW Tower Basin temperature rising above 90°F.

High temperature or low flow alarms on any components or systems cooled by NSCW.

MAJOR ACTIONS Determine condition causing loss or degraded operation of NSCW.

Transfer loads to unaffected train.

Correct or repair condition causing loss or degraded operation of NSCW.

Printed October 3, 2013 at 12:28

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 18021-C 19 Effective Date Page Number LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM 11/09/2012 2 of 15 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1

1. Check if catastrophic leakage from 1. Go to Step 6.

NSCW system - EXISTS.

2

2. Place affected train NSCW pump 2.

handswitches in PULL-TO-LOCK.

3

3. Depress both Emergency Stop 3.

pushbuttons for the affected DG.

4

4. Verify proper operation of 4. IF neither NSCW train can be placed UNAFFECTED NSCW train: in normal, two pump operation, THEN perform the following:

4.a Two pumps running. a. Trip the reactor.

4.b Supply header pressure greater b. Initiate 19000-C, E-0 REACTOR than 70 psig: TRIP OR SAFETY INJECTION.

Train A: PI-1636 4.c Train B: PI-1637 c. Trip all reactor coolant pumps.

4.d Supply header temperature d. Isolate letdown.

computer indication less than 90°F:

Train A: T2601 4.e Train B: T2602 e. Place one train of NSCW in single pump operation by initiating 13150, NUCLEAR Supply header flow SERVICE COOLING WATER approximately 17,000 gpm: SYSTEM.

Train A: FI-1640B 4.f Train B: FI-1641B f. Verify train-related CCP or NCP running and seal injection flow established using 13006, CHEMICAL AND VOLUME CONTROL SYSTEM.

Step 4 continued on next page Printed October 3, 2013 at 12:28

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 18021-C 19 Effective Date Page Number LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM 11/09/2012 1 of 15 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure addresses the loss or degraded operation of one or more trains of Nuclear Service Cooling Water.

SYMPTOMS Trip of operating NSCW pumps and failure of standby pump to start.

Dropping NSCW Supply Header pressure.

Large difference between Supply Header flow and Return Header flow, indicating a large leak.

NSCW Tower Basin temperature rising above 90°F.

High temperature or low flow alarms on any components or systems cooled by NSCW.

MAJOR ACTIONS Determine condition causing loss or degraded operation of NSCW.

Transfer loads to unaffected train.

Correct or repair condition causing loss or degraded operation of NSCW.

Printed October 3, 2013 at 13:06

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 18021-C 19 Effective Date Page Number LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM 11/09/2012 2 of 15 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1

1. Check if catastrophic leakage from 1. Go to Step 6.

NSCW system - EXISTS.

2

2. Place affected train NSCW pump 2.

handswitches in PULL-TO-LOCK.

3

3. Depress both Emergency Stop 3.

pushbuttons for the affected DG.

4

4. Verify proper operation of 4. IF neither NSCW train can be placed UNAFFECTED NSCW train: in normal, two pump operation, THEN perform the following:

4.a Two pumps running. a. Trip the reactor.

4.b Supply header pressure greater b. Initiate 19000-C, E-0 REACTOR than 70 psig: TRIP OR SAFETY INJECTION.

Train A: PI-1636 4.c Train B: PI-1637 c. Trip all reactor coolant pumps.

4.d Supply header temperature d. Isolate letdown.

computer indication less than 90°F:

Train A: T2601 4.e Train B: T2602 e. Place one train of NSCW in single pump operation by initiating 13150, NUCLEAR Supply header flow SERVICE COOLING WATER approximately 17,000 gpm: SYSTEM.

Train A: FI-1640B 4.f Train B: FI-1641B f. Verify train-related CCP or NCP running and seal injection flow established using 13006, CHEMICAL AND VOLUME CONTROL SYSTEM.

Step 4 continued on next page Printed October 3, 2013 at 13:06

Approved By Procedure Version J. B. Stanley Vogtle Electric Generating Plant 18021-C 19 Effective Date Page Number LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM 11/09/2012 3 of 15 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 4.g

g. Check RCP No. 1 seal temperatures less than 220°F.

4.h

h. IF RCP No. 1 seal temperatures greater than 220°F, THEN do NOT attempt to restart RCPs prior to a status evaluation.

5

5. Go to Step 13. 5.

6

6. Verify two or more NSCW pumps on 6. Perform the following:

the affected train are operating properly by checking the following parameters exist:

6.a Supply header pressure greater a. Place affected train NSCW than 70 psig. pump handswitches in PULL-TO-LOCK.

Train A: PI-1636 6.b Train B: PI-1637 b. Depress both Emergency Stop pushbuttons for the affected DG.

6.c Supply header flow c. Investigate cause for trip of approximately 17,000 gpm. running pump(s).

Train A: FI-1640B Train B: FI-1641B Step 6 continued on next page Printed October 3, 2013 at 13:06

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 1 of 32 RECORDING LIMITING CONDITIONS FOR OPERATION PROCEDURE LEVEL OF USE CLASSIFICATION PER NMP-AP-003 CATEGORY SECTIONS Continuous: NONE

Reference:

NONE Information: ALL Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 2 of 32 TABLE OF CONTENTS PAGE 1.0 PURPOSE 3 2.0 PRECAUTIONS AND LIMITATIONS 3 3.0 DEFINITIONS 4 3.1 EXTENT OF CONDITION REVIEW 4 3.2 LIMITING CONDITION FOR OPERATION (LCO) 4 3.3 TECHNICAL REQUIREMENT (TR) 4 3.4 INFORMATION ONLY LIMITING CONDITION FOR OPERATION/TECHNICAL REQUIREMENT (Info LCO/TR) 5 3.5 LOSS OF SAFETY FUNCTION (LOSF) 5 3.6 SUPPORT SYSTEM 5 3.7 SUPPORTED SYSTEM 6 4.0 PROCEDURE 7 4.1 LCO/TR STATUS SHEET PREPARATION 7 4.1.1 Initiation Of LCO/TR Status Sheet For An LCO/TR 8 4.1.2 Initiation Of LCO/TR Status Sheet For An Info LCO/TR 11 4.1.3 Restoration Of LCO/TR s And Info LCO/TR s 12 4.2 CONVERSION OF LCO/TR(S) TO INFO LCO/TR(S) 14 4.3 CONVERSION OF INFO LCO/TR(s) TO LCO/TR(s) 14 4.4 LCO/TR STATUS BINDER 14 4.4.1 Part I. LCO/TR Status Log 14 4.4.2 Part II. Active LCO Status Sheets 14 4.4.3 Part III. Completed LCO/TR Status Sheets 15 4.5 LOSS OF SAFETY FUNCTION (LOSF) EVALUATION 15 5.0 RECORDS 20

6.0 REFERENCES

20 Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 3 of 32 1.0 PURPOSE 1.1 This procedure prescribes the method to record the failure to meet the Limiting Conditions for Operation (LCO), or Technical Requirement, the associated ACTION requirements, any change in status effecting the ACTION, and the return to compliance with LCO/TR.

1.2 This procedure also includes instructions for implementing Technical Specification 5.5.15, the Safety Function Determination Program. As required by LCO 3.0.6, this program ensures that proper actions are taken such that multiple inoperable Structures, Systems, or Components (SSC) do not result in an undetected LOSS OF SAFETY FUNCTION.

1.3 This procedure also ensures that the allowed out of service time of SUPPORTED SYSTEMS is not inappropriately extended as a result of multiple inoperable SUPPORT SYSTEMS.

2.0 PRECAUTIONS AND LIMITATIONS 2.1 Technical Specification LCO 3.0.2 states that the required Actions of an LCO MUST be performed when the requirements of the LCO are NOT met. LCO 3.0.6 provides an exception to LCO 3.0.2 for SUPPORTED SYSTEMS by NOT requiring the Required Actions for the SUPPORTED SYSTEMS to be performed WHEN the failure to meet an LCO is SOLELY due to the inoperability of a SUPPORT SYSTEM. In this situation, LCO 3.0.6 requires ONLY the Required Actions of the SUPPORT SYSTEM to be performed. Since cascading is NOT required in this case, a possibility exists that unrelated concurrent failures of more than one SUPPORT SYSTEM could result in the complete loss of both trains of a SUPPORTED SYSTEM. THEREFORE, upon a failure to meet two or more LCOs during the same time period, an evaluation SHALL be conducted to determine if a LOSS OF SAFETY FUNCTION (LOSF) exists. This LOSF Evaluation satisfies the criteria of Technical Specification Administrative Control 5.5.15, Safety Function Determination Program.

2.2 If the failure of a SSC not addressed by Technical Specification results in the inoperability of a required SUPPORT and/or SUPPORTED SYSTEM, then the LCO(s) for the required SUPPORT and/or SUPPORTED SYSTEM would be entered. Example: If 1NB10 is de-energized, the operability of D/G '1B' or the Pressurizer Heaters may be impacted.

2.3 A single component inoperability can result in multiple inoperabilities within a single train and affect multiple Technical Specification LCOs. LCO 3.0.6 limits the amount of "cascading" of actions that is required when an inoperable SSC renders a SUPPORT SYSTEM inoperable.

Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 4 of 32 2.4 A single component inoperability CAN also impact operability on redundant trains. Example: IF 1HV-8716A is closed, both trains of ECCS may be impacted.

2.5 A Loss of Safety Function evaluation must be performed for each inoperability of a SCC impacting a required SUPPORT or SUPPORTED SYSTEM(s).

2.6 The LOSF Evaluation MUST be reinitiated whenever an additional required structure, system, or component (SSC) is declared inoperable. This includes LCOs with Required Actions that specify declaring additional components inoperable.

2.7 Alternating between LCO Conditions, in order to allow indefinite continued operation while not meeting the LCO, is not allowed.

3.0 DEFINITIONS 3.1 EXTENT OF CONDITION REVIEW A review to determine the scope of SSCs (in other trains, units, or subcomponents) that are affected by a condition adverse to quality.

3.2 LIMITING CONDITION FOR OPERATION (LCO)

A condition specified in the plant Technical Specifications (TS) or Technical Requirements Manual (TRM) which limits unit operations. An LCO may be entered by an equipment malfunction or a change in a unit parameter. If an LCO is not met, the associated ACTION requirements shall be met.

3.3 TECHNICAL REQUIREMENT (TR)

A condition specified in the plant Technical Requirements Manual (TRM) which limits unit operations. A TR may be entered by an equipment malfunction OR a change in a unit parameter. If a TR is not met, the associated ACTION requirements SHALL be met. TR and Technical Requirement Surveillances (TRS) associated with each TR are implemented in the same way as Technical Specifications. However, TRs and TRSs are treated as plant procedures and are not part of the Technical Specification. Therefore exceptions apply (Reference TRM Section 11.5).

Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 5 of 32 3.4 INFORMATION ONLY LIMITING CONDITION FOR OPERATION/TECHNICAL REQUIREMENT (Info LCO/TR)

A method of tracking an equipment malfunction or change in plant parameter which would restrict unit operation in another mode OR prevent a mode change in which it would become applicable, or may become an LCO/TR for the present mode should other Technical Specification related equipment or redundant safety related equipment become inoperable.

Information Only LCOs should NOT be prepared for conditions that are not applicable in the present operating mode unless used for tracking for entry into a mode in which a transition is to be directly made.

In addition, as an administrative tool to help track compliance with the ODCM, Information LCOs WILL be used when the requirements of the Offsite Dose Calculation Manual (ODCM) Sections 2.5 and 3.5 are not met.

Log entry (Electronic Log) of Information Only LCOs should NOT be made.

3.5 LOSS OF SAFETY FUNCTION (LOSF)

A LOSF exists WHEN; assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed.

3.6 SUPPORT SYSTEM 3.6.1 A SSC which is needed by another Technical Specification LCO required SSC to perform a safety function.

Example: The Component Cooling Water System (SUPPORT SYSTEM) is required by the Residual Heat Removal System (SUPPORTED SYSTEM) to fulfill its safety function. A SUPPORT SYSTEM may also be a SUPPORTED SYSTEM. Example: The Component Cooling Water System requires the Nuclear Service Cooling Water System to fulfill its safety function.

In the question the Train 'B' NSCW supports the Train 'B" SI Pump.

Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 6 of 32 3.6.2 For the purpose of implementing LCO 3.0.6, a SSC which monitors or maintains a process parameter or operating limit is not a SUPPORT SYSTEM; however, specific functions of Technical Specification instrumentation required to fulfill a credited safety function, may be considered a SUPPORT SYSTEM.

Examples:

The Digital Rod Position Indicators (DRPI) are used to monitor control rod insertion limits, however; inoperability of DRPI does not result in the control rods not being within insertion limits. Control rod insertion limits are monitored separately and actions taken as appropriate when insertion limits are not met or Surveillance Requirements not performed when required.

Likewise, parameter limits that could affect other parameter limits if exceeded are also NOT considered SUPPORT SYSTEMS for the purposes of implementing LCO 3.0.6. Example: Exceeding control rod insertion limits could affect hot channel factors Auto Actuation Logic and Actuation Relays, although identified as Instrumentation, MAY be considered a SUPPORT SYSTEM.

3.7 SUPPORTED SYSTEM A SSC, required by the Technical Specifications, which requires a SUPPORT SYSTEM to ensure its safety function can be performed. For the purposes of implementing LCO 3.0.6, process parameters, operating limits, or individual instrument channels are NOT SUPPORTED SYSTEMS; however, specific functions of Technical Specification instrumentation required to fulfill a credited safety function, MAY be considered a SUPPORTED SYSTEM.

Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 15 of 32 4.4.3 Part III. Completed LCO/TR Status Sheets Part III contains copies of LCO/TR Status Sheets for LCO/TRs that have been restored. Sheets are filed in order of their LCO/TR number. Copies of completed sheets SHOULD be retained in the binder for at least 30 days after they have been closed out.

4.5 LOSS OF SAFETY FUNCTION (LOSF) EVALUATION 4.5.1 Review Precautions and Limitations PRIOR to performing next step.

4.5.2 Identify the applicable Technical Specification conditions and required actions for the inoperable SSCs PRIOR to entering the LCO, IF possible.

NOTE A flow chart of the LOSF Evaluation process is shown in Figure 5.

4.5.3 Generate a list of impacted SUPPORT/ SUPPORTED Systems.

4.5.3.1 Considering the Conditions identified in step 4.5.2 as well any LCO Condition(s) previously in effect, determine if required SUPPORT or SUPPORTED SYSTEM(s) are rendered inoperable on redundant safety-related trains.

Train A Train B System i System i System ii System ii Inoperable system System iii System iii System iv System iv For the above example, IF Train A System iii is inoperable THEN Train B Systems i, ii and iii (support systems) and System iv (supported system) MUST be verified operable.

IF all Conditions in effect are limited to a single train, THEN no LOSF exists. All applicable Conditions SHOULD be entered, the provisions of LCO 3.0.6 may be applied, and NO additional evaluation is required.

Below is an example of a list when Unit 2 SSPS is rendered inoperable while performing U2 RTB testing.

Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 16 of 32 4.5.4 Procedure 10005-C SHALL be used to manually ILLUMINATE SSMP for the systems/components identified in steps 4.5.2 and 4.5.3.

Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 17 of 32 4.5.5 Using flowchart (Figure 5) determine if a Loss of Safety Function will exist IF the component/system is rendered inoperable. A method of place keeping should be used ensuring correct flow path is used. A (SRO) SHALL conduct an independent peer check of flowchart.

4.5.6 Determine IF concurrent inoperable SUPPORT or SUPPORTED systems on required redundant train, results in the loss of a credited safety function.

4.5.6.1 Equipment supported by an inoperable Offsite Source OR Diesel Generator should NOT be considered inoperable for the purpose of this evaluation, UNLESS required by LCO 3.8.1 Required Action A.2 or B.3. IF LCO 3.8.1 Condition A OR Condition B is in effect AND implementation of Required Action A.2 or B.3 subsequently results in the inoperability of a required supported system, THEN a LOSF Evaluation MUST be re-performed.

Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 18 of 32 4.5.6.2 The TS related systems that SHOULD be evaluated when determining if a potential loss of safety function exists are:

Reactor Trip System Automatic Trip Logic Reactor Trip and Bypass Breakers ESFAS Automatic Actuation Logic & Actuation Relays:

Safety Injection Containment Spray Containment Isolation Steamline Isolation Turbine Trip and Feedwater Isolation Auxiliary Feedwater Containment Sump Semi-automatic Switchover LOSP Instrumentation - Loss of either Undervoltage or Degraded Voltage Functions CVI Automatic Actuation Logic & Actuation Relays CREFS Automatic Actuation Logic & Actuation Relays High Flux at Shutdown Alarm (HFASA)

Decay Heat Removal (including refueling operations)

Pressurizer PORVs and associated Block Valves Cold Overpressure Protection System ECCS (See Step 4.5.6.8)

Containment Penetrations Containment Spray and Cooling Systems Main Steam Isolation Valves MFIVs, MFRVs, and associated Bypass Valves Atmospheric Relief Valves Auxiliary Feedwater System Component Cooling Water System Nuclear Service Cooling Water System Ultimate Heat Sink Control Room Emergency Filtration Systems Piping Penetration Area Filtration and Exhaust System ESF Room Cooler and Safety-Related Chiller System AC Sources (including Safety Systems Sequencer)

Diesel Fuel Oil, Lube Oil, Starting Air, and Ventilation DC Sources Inverters Electrical Distribution Systems Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 19 of 32 4.5.6.3 A credited safety function is a function required to mitigate the consequences of a design basis event as described in the FSAR (reference FSAR Chapters 6 and 15), including all assumptions of the initiating event such as loss of offsite power.

NOTE FSAR assumptions such as loss of offsite power are considered as part of the initiating event and should not be considered as an additional concurrent failure in the following step.

4.5.6.4 A LOSF exists WHEN; assuming that with no additional concurrent failure during a design basis event, a required safety function assumed in the accident analysis CANNOT be performed.

4.5.6.5 If a LOSF is determined to exist, the appropriate Conditions and Required Actions of the LCO in which the LOSF exists SHALL be entered. IF no Condition within the LCO addresses the LOSF, THEN LCO 3.0.3 shall be entered. Results of the LOSF Evaluation SHOULD be entered in the Unit Control Log and/or by initiation of an LCO tracking sheet documenting the LCO in which the LOSF exists.

4.5.6.6 IF a LOSF does not exist, THEN the Required Actions for the LCO SUPPORT SYSTEM(s) address the condition AND the required actions of the SUPPORTED SYSTEM(s) do NOT have to be performed as permitted by LCO 3.0.6.

4.5.6.7 Ensure that the Completion Time of any SUPPORTED SYSTEM has not been inappropriately extended as shown in Figure 6. Completion Time Extensions are considered inappropriate if the SUPPORTED SYSTEM remains inoperable for longer than the allowed out of service time of the SUPPORT SYSTEM which caused the initial inoperability.

4.5.6.8 Technical Specification 3.5.2 Condition A, allows an ECCS train to be inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS Train remains available. Analyses have been performed for many of the potential flowpaths available that can be used to credit this allowance (reference 5.3). For cases where it is unclear if the flow equivalent of a single ECCS train remains operable, system engineering should be contacted for guidance.

Printed October 3, 2013 at 12:42

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 10008-C 30 Effective Date Page Number 02/08/2013 RECORDING LIMITING CONDITIONS FOR OPERATION 28 of 32 LOSF EVALUATION FLOWCHART Figure 5 Printed October 3, 2013 at 12:42

1. 076AG2.4.47 001/LOIT AND LOCT/SRO/C/A 4.2/4.2/076G2.4.47/LO-TA-63013///

Initial condition:

- Unit 1 is at 100% reactor power.

Current conditions:

- 18013-C, "Rapid Power Reduction," is entered due to a secondary transient.

- The following reactor power trends are recorded:

TIME POWER 1100 100%

1115 97%

1130 93%

1145 89%

1200 86%

1215 81%

1230 79%

Which one of the following completes the following statement?

Chemistry sampling of the RCS __(1)__ required per Tech Spec 3.4.16, "RCS Specific Activity," Surveillance Requirements, and per the Bases of Tech Spec 3.4.16, "RCS Specific Activity," the required action to reduce RCS Tavg below 500°F if the gross specific activity is exceeded is to prevent opening of the __(2)__.

__(1)__ __(2)__

A. is Atmospheric Relief Valves B. is Main Steam Safety Valves C. is NOT Atmospheric Relief Valves D. is NOT Main Steam Safety Valves K/A 076 High Reactor Coolant Activity G2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

K/A MATCH ANALYSIS Monday, February 24, 2014 4:10:27 PM 1

The question sets up a plausible scenario which includes all the required KA elements.

First the SRO candidate must recognize in a timely manner the requirements for RCS sampling following the power reduction trend provided in the stem. Timely RCS activity sampling following a power reduction of more than 15% in one hour is required verify no fuel damage, which leads to high coolant activity. Then the candidate must determine the Tech Spec bases for lowering the energy level in the RCS if limits are exceeded.

EXPLANATION OF REQUIRED KNOWLEDGE TS SR 3.4.16.2 requires verification of DOES EQUIVALENT I-131 specific activity less than or equal to 1.0 uCi/gm between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a power change of greater than or equal to 15% RTP within a 1 hr period. By definition, the 1 hr period is a rolling hour. As stated in the stem, the power change between 1100 and 1200 is 14%RTP.

The power change between 1115 and 1215 is 18%. Therefore, the 15% RTP in an hour has been exceeded.

Per TS 3.4.13, if I-131 is in excess of limits or LCO completion time of Cond A or B cannot be met, then the plant is required to be placed in Mode 3 with RCS Tavg <500F.

Per TS 3.4.13 Bases, with RCS Tavg <500F, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valve lift setpoint.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Part 1 is correct the candidate should determine that from the stem information provided that a power reduction of more than 15% in one hour occured between 1115 and 1215 and require chemistry sampling of the RCS to verify no fuel damage.

Part 2 is incorrect however plausible since the candidate may determine that the bases for the RCS temperature limit is to prevent the ARVs from lifting since they would normally open an a lower setpoint.

B. Correct. Part 1 is correct. See Part 1 of choice A above.

Part 2 is correct. Per Tech Spec 3.4.16 Bases, the purpose of lowering RCS temperature below 500°F is to prevent radioactive releases due to main stream safety valves lifting.

C. Incorrect. Plausible. Part 1 is incorrect however plausible since the candidate may determine based on the time line given that RCS sampling is not required. The rolling hour starting at times 1100 and 1130 are <15% change.

Part 2 is incorrect. See Part 2 of choice A above.

D. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice C above.

Part 2 is correct. See Part 2 of choice B above.

Monday, February 24, 2014 4:10:27 PM 2

SRO JUSTIFICATION (2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the question requires Bases knowledge.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, the applicability statement for TS 3.4.16 does state Mode 3

>500F Tavg. However, the Bases of this applicability and the Required Actions for both Conditon B & C are only listed in the Bases document.

-Can question be answered solely by knowing the TS Safety Limits? No, Safety Limits are not addressed by this question.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)

• Knowledge of TS bases that is required to analyze TS required actions and terminology Yes, the question requires the candidate to specifically know the Bases for TS 3.4.13.

Level: SRO Tier # / Group # T1 / G2 K/A# 076G2.4.47 Importance Rating: 4.2 / 4.2 Technical

Reference:

TS 3.4.16, Ammendment No. 158, page 4.3.16-2 TS Bases 3.4.16, Rev 1-10/01, page B 3.4.16-3 References provided: None Learning Objective: LO-TA-63013 Implement Technical Specification LCO using 10008-C (SRO Only)

LO-PP-16001-04 State the LCO, applicability, bases, and the 1 hr or less actions for each of the following: 3.4.16 RCS Specific Activity Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 41.5 / 43.2 Comments:

You have completed the test!

Monday, February 24, 2014 4:10:27 PM 3

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Tavg < 500°F.

Time of Condition A not met.

OR DOSE EQUIVALENT I-131 in the unacceptable region of Figure 3.4.16-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific In accordance with activity 100/ Ci/gm. the Surveillance Frequency Control Program SR 3.4.16.2 ----------------------------NOTE-----------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-131 In accordance with specific activity 1.0 Ci/gm. the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period (continued)

Vogtle Units 1 and 2 3.4.16-2 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO The specific iodine activity is limited to 1.0 Ci/gm DOSE EQUIVALENT I-131, and the gross specific activity in the reactor coolant is limited to the number of Ci/gm equal to 100 divided by (average disintegration energy of the sum of the average beta and gamma energies of the coolant nuclides). The limit on DOSE EQUIVALENT I-131 ensures the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose to an individual at the exclusion area boundary during the Design Basis Accident (DBA) will be a small fraction of the allowed thyroid dose. The limit on gross specific activity ensures the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> whole body dose to an individual at the exclusion area boundary during the DBA will be a small fraction of the allowed whole body dose.

The SGTR accident analysis (Ref. 2) shows that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> site boundary dose levels are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of an SGTR, lead to exclusion area boundary doses that exceed the 10 CFR 100 dose guideline limits.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS average temperature t 500qF, operation within the LCO limits for DOSE EQUIVALENT I-131 and gross specific activity are necessary to contain the potential consequences of an SGTR to within the acceptable site boundary dose values.

For operation in MODE 3 with RCS average temperature < 500qF, and in MODES 4 and 5, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves.

(continued)

1. G2.1.34 001/LOIT AND LOCT/SRO/M/F 2.7/3.5/G2.1.34/LO-TA-60024///

Given the following:

- Unit 1 is at 100% reactor power.

Which one of the following completes the following statement?

Per Tech Spec 3.7.16, "Secondary Specific Activity," the specific activity of the secondary coolant shall be < __(1)__ Ci/gm Dose Equivalent I-131, and operating within this limit ensures that the off-site dose will be limited to within a small fraction of the 10 CFR 100 dose guideline values in the event of a __(2)__.

__(1)__ __(2)__

A. 0.10 steam line break B. 0.10 loss of all AC power C. 1.0 steam line break D. 1.0 loss of all AC power Tuesday, February 25, 2014 9:35:23 AM 1

K/A G2.1.34 Knowledge of primary and secondary plant chemistry limits.

K/A MATCH ANALYSIS The question tests the candidate's knowledge of primary and secondary plant chemistry limits by requiring the student to select Tech Spec 3.7.16, "Secondary Specific Activity,"

specific activity limit for Dose Equivalent I-131 and the associated limiting design bases accident. This value is pitted against the limit for Tech Spec 3.4.16, "RCS Specific Activity".

EXPLANATION OF REQUIRED KNOWLEDGE Per Tech Spec 3.7.16, "Secondary Specific Activity" bases, the accident analysis of the main steam line break (MSLB), assumes the initial secondary coolant specific activity to have a radioactive isotope concentration of 0.10 µCi/gm DOSE EQUIVALENT I-131.

This assumption is used in the analysis for determining the radiological consequences of the postulated accident. The accident analysis, based on this and other assumptions, shows that the radiological consequences of an MSLB do not exceed a small fraction of the unit EAB limits for whole body and thyroid dose rates.

Per Tech Spec 3.4.16, "RCS Specific Activity" bases, the limit of 1.0 µCi/gm DOSE EQUIVALENT I-131 on specific activity of ensures that the doses are held to a small fraction of the 10 CRF 100 limits during a steam generator tube rupture accident. The maximum dose to the whole body and the thyroid considers exposure to an individual at the exclusion boundary for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ANSWER / DISTRACTOR ANALYSIS A. Correct. Part 1 is correct. Per Tech Spec 3.7.16, Secondary Specific Activity, the specific activity of the secondary coolant shall be <

or = 0.10 µCi/gm DOSE EQUIVALENT I-131.

Part 2 is correct. Per Tech Spec 3.7.16, Secondary Specific Activity bases, the accident analysis of the main steam line break (MSLB), assumes the initial secondary coolant specific activity to have a radioactive isotope concentration of 0.10

µCi/gm DOSE EQUIVALENT I-131. This assumption is used in the analysis for determining the radiological consequences of the postulated accident. The accident analysis, based on this and other assumptions, shows that the radiological consequences of an MSLB do not exceed a small fraction of the unit EAB limits for whole body and thyroid dose rates.

B. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice A above.

Part 2 is incorrect however plausible since the candidate may Tuesday, February 25, 2014 9:36:10 AM 1

recall the LOSP as the most limiting accident not taking into account the remaining steam generators are available for core decay heat dissipation by venting steam to the atmosphere through the MSSVs and steam generator atmospheric dump valves (ARVs). The Auxiliary Feedwater System supplies the necessary makeup to the steam generators. Venting continues until the reactor coolant temperature and pressure have decreased sufficiently for the Residual Heat Removal System to complete the cooldown.

C. Incorrect. Plausible. Part 1 is incorrect however plausible since its reasonable to assume the candidate may confuse primary limit for specific activity with the secondary limit since the numbers are only distinguish by the movement of the decimal point.

Part 2 is correct. See Part 2 of choice A above.

D. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice C above.

Part 2 is correct. See Part 2 of choice B above.

SRO JUSTIFICATION (10CFR43(b))

(2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the question is not addressing Tech Spec action times.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, the question is not addressing Tech Spec above-the-line information.

-Can question be answered solely by knowing the TS Safety Limits? No, the question is not related to Tech Spec Safety Limits.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)

• Knowledge of TS bases that is required to analyze TS required actions and terminology. Yes, the answer to the question is only found in Tech Spec bases.

Tuesday, February 25, 2014 9:36:10 AM 2

Level: SRO Tier # / Group # T3 K/A# G2.1.34 Importance Rating: 2.7 / 3.5 Technical

Reference:

TS 3.4.16, Ammendment No. 137, page 3.4.16-1 TS 3.7.16, Ammendment No. 158, page 3.7.16-1 TS Bases 3.4.16, Rev 0, page B 3.4.16-1 TS Bases 3.7.16, Rev 1-10/01, page B 3.7.16-2 References provided: None Learning Objective: LO-TA-63013 Implement Technical Specification LCO using 10008-C (SRO Only)

LO-PP-16001-04 State the LCO, applicability, bases, and the 1 hr or less actions for each of the following: 3.4.16 RCS Specific Activity.

LO-LP-39211-01 For any given item in section 3.7 of Tech Specs, be able to:

a. State the LCO.

b. State any one hour or less required actions.

LO-TA-60024 Respond to Abnormal Plant Chemistry per 18014-C or 18015-C Question origin: NEW Cognitive Level: M/F 10 CFR Part 55 Content: 41.5 / 41.10 / 43.2 Comments:

You have completed the test!

Tuesday, February 25, 2014 9:36:10 AM 3

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity 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

---

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

LCO 3.0.4c is applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT A.1 Verify DOSE Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> I-131 > 1.0 Ci/gm. EQUIVALENT I-131 within the acceptable region of Figure 3.4.16-1.

AND A.2 Restore DOSE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EQUIVALENT I-131 to within limit.

B. Gross specific activity of B.1 Perform SR 3.4.16.2. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the reactor coolant not within limit. AND B.2 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Tavg < 500°F.

(continued)

Vogtle Units 1 and 2 3.4.16-1 Amendment No. 137 (Unit 1)

Amendment No. 116 (Unit 2)

Secondary Specific Activity 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Secondary Specific Activity LCO 3.7.16 The specific activity of the secondary coolant shall be 0.10 Ci/gm DOSE EQUIVALENT I-131.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> within limit.

AND A.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify the specific activity of the secondary In accordance with coolant is 0.10 Ci/gm DOSE EQUIVALENT the Surveillance I-131. Frequency Control Program

RCS Specific Activity B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND The maximum dose to the whole body and the thyroid that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during an accident is specified in 10 CFR 100 (Ref. 1). The limits on specific activity ensure that the doses are held to a small fraction of the 10 CFR 100 limits during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident.

The LCO limits specific activity for both DOSE EQUIVALENT I-131 and gross specific activity. The allowable levels are intended to limit the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the exclusion area boundary to a small fraction of the 10 CFR 100 dose guideline limits. The limits in the LCO are standardized, based on parametric evaluations of offsite radioactivity dose consequences for typical site locations.

The parametric evaluations showed the potential offsite dose levels for a SGTR accident were an appropriately small fraction of the 10 CFR 100 dose guideline limits. Each evaluation assumes a broad range of site applicable atmospheric dispersion factors in a parametric evaluation.

APPLICABLE The limits on the specific activity of the reactor coolant SAFETY ANALYSES ensures that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed a small fraction of the 10 CFR 100 dose guideline limits following a SGTR accident. The SGTR safety analysis (Ref. 2) assumes that the reactor has been operating at the maximum allowable Technical Specification limit for primary coolant activity and primary to secondary leakage for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and in the secondary coolant.

(continued)

Vogtle Units 1 and 2 B 3.4.16-1 Revision No. 0

Secondary Specific Activity B 3.7.16 BASES (continued)

APPLICABLE The accident analysis of the main steam line break (MSLB),

SAFETY ANALYSES as discussed in the FSAR, Chapter 15 (Ref. 2) assumes the initial secondary coolant specific activity to have a radioactive isotope concentration of 0.10 Ci/gm DOSE EQUIVALENT I-131. This assumption is used in the analysis for determining the radiological consequences of the postulated accident. The accident analysis, based on this and other assumptions, shows that the radiological consequences of an MSLB do not exceed a small fraction of the unit EAB limits (Ref. 1) for whole body and thyroid dose rates.

With the loss of offsite power, the remaining steam generators are available for core decay heat dissipation by venting steam to the atmosphere through the MSSVs and steam generator atmospheric dump valves (ARVs). The Auxiliary Feedwater System supplies the necessary makeup to the steam generators.

Venting continues until the reactor coolant temperature and pressure have decreased sufficiently for the Residual Heat Removal System to complete the cooldown.

In the evaluation of the radiological consequences of this accident, the activity released from the steam generator connected to the failed steam line is assumed to be released directly to the environment. The unaffected steam generator is assumed to discharge steam and any entrained activity through the MSSVs and ARVs during the event. Since no credit is taken in the analysis for activity plateout or retention, the resultant radiological consequences represent a conservative estimate of the potential integrated dose due to the postulated steam line failure.

Secondary specific activity limits satisfy Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO As indicated in the Applicable Safety Analyses, the specific activity of the secondary coolant is required to be d 0.10 PCi/gm DOSE EQUIVALENT I-131 to limit the radiological consequences of a Design Basis Accident (DBA) to a small fraction of the required limit (Ref. 1).

Monitoring the specific activity of the secondary coolant ensures that when secondary specific activity limits are exceeded, appropriate actions are taken in a timely manner (continued)

Vogtle Units 1 and 2 B 3.7.16-2 Rev. 1-10/01

1. G2.1.37 001/LOIT AND LOCT/SRO/M/F 4.3/4.6/G2.1.37/LO-LP-63500-09///

Given the following:

- Unit 1 is at 3% reactor power and raising power following an outage.

- Unit 2 is at 100% reactor power.

- The Shift Manager and the following assigned personnel are in the control room:

Unit 1 Unit 2 Shift Supervisor Shift Supervisor Reactivity Management SRO 1 NPO*

2 NPOs*

(*) NPO - Individual with a Reactor Operator License.

Which one of the following completes the following statement?

Per TRM 15.1, "Unit Staffing," the total number of NPOs assigned __(1)__ meet the minimum required for the given conditions, and per NMP-OS-001, "Reactivity Management Program," any changes to the Unit 1 reactivity plan must be approved by the __(2)__.

__(1)__ __(2)__

A. does Reactivity Management SRO B. does Shift Supervisor C. does NOT Reactivity Management SRO D. does NOT Shift Supervisor K/A Conduct of Operations 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.

K/A MATCH ANALYSIS The question tests the candidate's knwoledge of administrative procedural requirements assoicated with the required approvals for changing reactivity plans during low power ascention following a reactor startup.

Tuesday, February 25, 2014 9:43:40 AM 1

EXPLANATION OF REQUIRED KNOWLEDGE The candidate is required to evaluate the minimum shift staffing for the given Modes per TRM 15.1 for the RO positions. Per Table 15.1.2-1, (3) ROs are required with both Units in Modes 1-4. (2) RO's must be assigned to the OATC position on each of the Units.

Per NMP-OS-001 step 6.1.2.1, the Shift Manager, who has ultimate responsibility for controlling the reactor core, approves formal Reactivity Management Plans as described in paragraph 6.6. The Shift Supervisor, or a designated Senior Reactor Operator, directly supervises reactivity changes. A written plan is developed by reactor engineering and approved for significant reactivity changes. The Shift Manager, Shift Supervisor, OATC, Shift Technical Advisor, and Reactor Engineer concur on Reactivity Management Plans and changes thereto, prior to implementation.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. Per TRM 15.1 Table 15.1.2-1, (3) ROs are required with both Units in Modes 1-4. (2) RO's must be assigned to the OATC position on each of the Units. This condition is met.

The second part is incorrect. Per NMP-OS-001 step 6.1.2.1, the Shift Manager, Shift Supervisor, OATC, Shift Technical Advisor, and Reactor Engineer concur on Reactivity Management Plans and changes thereto, prior to implementation. However, the Reactivity Management SRO has oversight of all reactivity changes. As such, it is reasonable for a candidate without specific knowledge of the procedural requirements to assume the Reactivity Management SRO would also have this authority. Therefore, this distractor is plausible.

B. Correct. The first part is correct. See the first part of choice A above.

The second part is correct. Per NMP-OS-001 step 6.1.2.1, the Shift Manager, Shift Supervisor, OATC, Shift Technical Advisor, and Reactor Engineer concur on Reactivity Management Plans and changes thereto, prior to implementation.

C. Incorrect. Plausible. The first part is incorrect. Per TRM 15.1 Table 15.1.2-1, (3)

ROs are required with both Units in Modes 1-4. (2) RO's must be assigned to the OATC position on each of the Units. This condition is met. However, administrative procedure 00012-C Shift Manning limits are not currently met. Per 00012-C, the ENN communication position is normally filled by each of the UOs which results in (4) NPOs being required. This is the normal shift alignment. It is reasonable for a candidate who does not have adequate knowledge of TRM 15.1 requirements to determine insufficient NPOs exist. Therefore, this distractor is plausible.

Tuesday, February 25, 2014 9:43:40 AM 2

The second part is incorrect. See the second part of choice A above.

D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.

The second part is correct. See the second part of choice B above.

SRO JUSTIFICATION (10CFR43(b))

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, the question is not system related in any way.

-Can the question be answered solely by knowing immediate operator actions?

No, the question does not involve any operator actions.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, the question does not involve an AOP or EOP.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, the question requires the knowledge of a specifc detail. Overall understanding of approvals actually directs the candidate to an incorrect answer.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. Yes, the question requires specific knowledge of the hierachy of an administrative procedure for approving Reactivity Plan changes once developed.

(6) Procedures and limitations involved in initial core loading, alternations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

Yes, the question requires the candidate to have specific knowledge of the approval process and responsibilities of control room personnel that approve Reactivity Plans and oversee Reactivity Manipulations.

Tuesday, February 25, 2014 9:43:40 AM 3

Level: SRO Tier # / Group # T3 K/A# G2.1.37 Importance Rating: 4.3 / 4.6 Technical

Reference:

NMP-OS-001 Rev 17.0, page 9 00012-C Rev 17.2, pages 5&6 TRM 15.1, Table 15.1.2-1 Rev 0 12/26/96, page 15.1-2 References provided: None Learning Objective: LO-LP-36110-01 Per Technical Specifications, state the shift manning requirements. (SRO)

LO-LP-63503-02 State the requirements of the OATC with regards to shift manning when fuel is in either reactor.

LO-LP-63503-05 State the requirements of shift manning for the following conditions: (SRO only)

f. minimum shift crew LO-LP-63510-04 Name the two site groups that have the most day-to-day effect on reactivity management.

LO-LP-63500-09 State reactivity manipulation expectations including: monitoring, briefing, pre-plans, peer checks, transient operation, pull-and-wait, load change concurrence, instrument response, changing temperature and power, and Reactivity Management SRO.

Question origin: BANK - Hatch 2011 NRC # G2.1.37 Cognitive Level: M/F 10 CFR Part 55 Content: 43.5 / 43.6 Comments:

You have completed the test!

Tuesday, February 25, 2014 9:43:40 AM 4

Unit Staff TR 15.1 Table 15.1.2-1 MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH A COMMON CONTROL ROOM Position Number of Individuals Required to Fill Position One Unit in MODE 1, 2, 3, or4 and Both Units in One Unit in Both Units in MODE 5 or 6 or MODE 5 or 6 or MODE 1, 2, 3, or 4 DEFUELED DE FUELED ss 1 1 1 SRO 1 None( 1l 1 3(2)

RO 3(2)

NLO STA 1(3) None (1) At least one licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities must be present during CORE ALTERATIONS on either unit.

(2) At least one of the required individuals must be assigned to the designated position for each unit.

(3) See TS 5.2.2.g.

ss- Shift Superintendent with a Senior Operator License.

SRO- Individual with a Senior Operator License.

RO- Individual with an Operator License.

NLO- Non-licensed operator.

STA- Shift Technical Advisor.

Vogtle Units 1 and 2 15.1 - 2 Rev. 0 Technical Requirement 12/26/96

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 00012-C 17.2 Date Approved Page Number 03/17/2009 SHIFT MANNING REQUIREMENTS 5 of 6 DATA SHEET 1 Sheet 1 of 2 Minimum Shift Manning (Either Unit in Mode 1-4)

Date: Shift (Day/Night):

POSITION UNIT #1 COMMON UNIT #2 Shift Manager V-OPS-SS, V-ERO-CR01, and Also assigned as Emergency Director V-ERO-CR10 SS V-OPS-USS, V-ERO-CR02, AND Also assigned as ENS Communicator Also assigned as ENS Communicator V-ERO-CR10 OATC 1 2 V-OPS-RO/BOP UO 3 4 V-OPS-RO/BOP and V-ERO- Also assigned as ENN Communicator Also assigned as ENN Communicator CR04 SO V-OPS-SO SO/NPO SO/NPO STA (May be assigned other duties) (SM, or SSS or SS not assigned to FB V-OPS-STA or ENN Communicator)

Fire Team Captain V-FP-FIRE BRIGADE LEADER SSS, or SS C&T FB Member 1.

V-FP-FIRE BRIGADE SO(Also fulfills Common SO FSAR req)

FB Member 2.

V-FP-FIRE BRIGADE SO FB Member 3.

V-FP-FIRE BRIGADE SO FB Member 4.

V-FP-FIRE BRIGADE SO Security V-ERO-SEC or V-ERO-SEC02 Per Security Procedure 90101-C SAT Operator 5.

V-OPS-SO-OAO Assigned per procedure 13419-C SO/NPO/SRO Wilson Operator 6.

V-OPS-WILSON BLKSTRT Assigned per procedure 13419-C SO/NPO Emergency Plan POSITION UNIT #1 COMMON UNIT #2 Emergency Shift Manager Director V-OPS-SS ENN UO Unaffected Unit UO Unaffected Unit Communicator V-ERO-CR04 or V-ERO-CR10 ENS SS Unaffected Unit SS Unaffected Unit Communicator V-OPS-USS or V-OPS-STA Printed November 22, 2013 at 16:32

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 00012-C 17.2 Date Approved Page Number 03/17/2009 SHIFT MANNING REQUIREMENTS 6 of 6 DATA SHEET 1 Sheet 2 of 2 Emergency Plan (cont)

POSITION UNIT #1 COMMON UNIT #2 Dose Assessment V-ERO-TSC09 OR V-ERO-TSC10 HP Foreman Field Monitoring Team (FMT) 1.

V-ERO-CR08 OR V-ERO-OSC16 HP Tech/Chem Tech/SO/I&C Tech 2.

HP Tech/Chem Tech/SO/I&C Tech FMT Communicator V-ERO-TSC18 Chem Foreman/Chem Tech/HP Tech/

Maint. Shift ATL Chemistry Sampler V-ERO-OSC09 Chemistry Tech Mechanical Repair and Corrective Action V-ERO-OSC07 Mechanic Electrical Repair and Corrective Action V-ERO-OSC06 Electrician I&C Repair and Corrective Action V-ERO-OSC08 I & C Technician In Plant Monitors 1.

V-ERO-OSC17 HP Technician 2.

HP Technician Search & Rescue/First Aid 1.

(May be assigned other Duties) V-ERO-OSC15 HP Technician 2.

HP Technician Minimum Dual Unit Safe Shutdown POSITION UNIT #1 COMMON UNIT #2 Emergency Director SM ENN UO Unit #1 or #2 ENS UO Unit #2 or #1 Shutdown Panel B SS SS Shutdown Panel A OATC OATC Shutdown Panel C SO SO Fire Brigade Same as Normal Operations

1. Personnel may NOT be assigned to more than one position unless specifically noted next to the position label.

2. If both units are in Modes 5, 6, or defueled, minimum shift manning for operations may be reduced per Operations Manager (not Emergency Plan or Fire Brigade staffing).

COMMENTS:

Approved by: Date: Time:

Shift Manager Printed November 22, 2013 at 16:32

Southern Nuclear Operating Company Nuclear NMP-OS-001 Management Reactivity Management Program Version 17.0 Procedure Page 9 of 39 6.1.2 Expectations 6.1.2.1 Reactivity Management Controls NOTE: Prior to implementing any activity that has the potential to add positive reactivity; the plant shall be ramped down, as necessary, to ensure that the reactor does not exceed 100.0% of the unit's Rated Thermal Power limit.

Reactivity management involves a systematic process of controlling evolutions with the potential to affect reactivity:

Planned reactivity changes are conducted in a controlled and conservative manner Unexpected changes in reactivity are minimized Conservative actions are taken in response to unexpected reactivity changes Reactivity control systems, including reactivity monitoring instrumentation, are available and reliable Modifications, analyses, and predictions are correct and effectively implemented The Shift Manager, who has ultimate responsibility for controlling the reactor core, approves formal Reactivity Management Plans as described in paragraph 6.6. The Shift Supervisor, or a designated Senior Reactor Operator, directly supervises reactivity changes.

A strong relationship exists between reactor engineering and operations. Reactor engineering is actively engaged in the planning of significant reactivity changes such as reactor startup, reactor shutdown, planned power changes, and special tests with the potential to affect reactivity. A written plan is developed by reactor engineering and approved for significant reactivity changes. The Shift Manager, Shift Supervisor, OATC, Shift Technical Advisor, and Reactor Engineer concur on Reactivity Management Plans and changes thereto, prior to implementation.

6.1.2.2 Control Room Operations Only operators with an active license (NPO or SRO) manipulate the controls of the reactor. An individual in a NRC-approved license training program may manipulate reactor controls when under the direction and in the presence of a licensed operator.

The OATC obtains concurrence from the Shift Supervisor prior to allowing or performing planned reactivity manipulations. Directions that affect reactivity go through the Shift Supervisor.

A briefing is conducted prior to the start of a planned reactivity change. The reactor operator performing rod movement, the OATC and Shift Supervisor monitor reactivity manipulations and verify that the end state of the reactivity manipulation is as expected.

1. G2.2.20 001/LOCT AND LOIT/SRO/M/F 2.6/3.8/G2.2.20/LO-LP-63350-07//HL18 NRC/

Initial condition:

- ALB34-D01 125 VDC SWGR 1AD1 TROUBLE is received due to a bus ground.

Current conditions:

- Per NMP-AD-002, "'Problem Solving and Troubleshooting Guidelines," a troubleshooting plan has been written.

- As part of the plan, operations personnel will open various breakers and maintenance personnel will open links to measure resistance inside the 1AD1 panel.

Which one of the following completes the following statement?

This type of Troubleshooting Monitoring is called __(1)__,

and the tracking of the equipment out-of-service time while troubleshooting 1AD1 is the responsibility of the __(2)__ Department.

A. (1) Intrusive (2) Maintenance B. (1) Intrusive (2) Operations C. (1) Non-Intrusive (2) Maintenance D. (1) Non-Intrusive (2) Operations G2.2.20 Equipment Control Knowledge of the process for managing troubleshooting activities.

K/A MATCH ANALYSIS:

The candidate is presented with a plausible scenario where Troubleshooting is to be performed in the 1AD1 Panel by both maintenance personnel and operations personnel. The activity involves opening links, breakers and measuring resistance to Tuesday, February 25, 2014 10:15:47 AM 1

resolve a ground problem. The candidate must determine if this is intrusive or non-intrusive trouble shooting and also has to determine the person responsible for maintaining the system status of the panel during Trouble Shooting activities.

EXPLANATION OF REQUIRED KNOWLEDGE Per NMP-AD-002 definition 3.5, non-intrusive monitoring is defined as the act of monitoring a component or system by not affecting normal oepration of the conponent or system.

Per NMP-AD-002 definition 3.6, intrusive monitoring is defined as the act of temporarily altering the system to allow monitoring a component or system. This applies to electrical or mechanical testing methods.

Per NMP-AD-002 responsibility 4.2, the Operation Department is responsible for maintaining approved system status during troubleshooting activities (ie Out of Service).

ANSWER / DISTRACTOR ANALYSIS:

A. Incorrect. Plausible. Part 1 is correct. Opening breakers by Operations and opening links by Maintenance is an intrusive troubleshooting per NMP-AD-002.

Part 2 is incorrect. Operations Department is responsible for tracking systems status, however the answer is 'plusable' because opening links, measuring resistance, etc. would all be maintenance activities and some amount of individual component status control is required. These components are typically tracked using a "lifted lead" sheet inside the MWO.

Even though maintenance tracks status of some components, Operations still retains overall responsibility for ensuring equipment status and configuration control.

B. Correct. Part 1 is correct. See Part 1 of choice A above.

Part 2 is correct. Per NMP-AD-002 responsibility 4.2, the Operation Department is responsible for maintaining approved system status during troubleshooting activities.

C. Incorrect. Plausible. Part 1 is incorrect however 'plusable' because the word intrusive is very subjective and would require procedure knowledge to make this distinction. The candidate may determine that allowing intrusive troubleshooting on 1AD1 with the unit on line would imply too much risk to operations and thus eliminate this possibility.

Part 2 is incorrect. See Part 2 of choice A above.

D. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice C above.

Part 2 is correct. See Part 2 of choice B above.

Tuesday, February 25, 2014 10:15:47 AM 2

SRO JUSTIFICATION (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, system knowledge will not answer any part of this question.

-Can the question be answered solely by knowing immediate operator actions?

No, IOAs are not addressed in any way in this question.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, this question does not pertain to an EOP or AOP.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, the question pertains to specific guidance in NMP-AD-002, which cannot be answered by broad knowledge of the procedure.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures Yes, the question requires specific knowledge of an administrative procedure that specifies the hiearchy by which plant configuration is statused and configuration control is maintained during normal plant operation.

Tuesday, February 25, 2014 10:15:47 AM 3

Level: SRO Tier # / Group # T3 K/A# G2.2.20 Importance Rating: 2.6 / 3.8 Technical

Reference:

NMP-AD-002, Rev 10.0, pages 4 & 5 References provided: None Learning Objective: LO-LP-63350-07 Define the following terms:

f. Trouble shooting LO-LP-63354-03 Describe the Shift Manager's responsibility concerning maintenance activities.

Question origin: BANK - HL18 NRC # G.2.2.20 Cognitive Level: M/F 10 CFR Part 55 Content: 41.10 / 43.5 Comments:

You have completed the test!

Tuesday, February 25, 2014 10:15:47 AM 4

Southern Nuclear Operating Company Nuclear Problem Solving and Troubleshooting NMP-AD-002 Management Guidelines Version 10.0 Procedure Page 4 of 14 1.0 Purpose The purpose of this procedure is to provide a process for performance of troubleshooting when required for plant problem resolution. These problems may include equipment failures, abnormal operating conditions, negative performance trends or recurring events.

2.0 Applicability 2.1 This procedure is applicable to troubleshooting activities at any of the SNC sites.

2.2 Entry into the formal troubleshooting process is not intended for simple problems where the cause appears straightforward or known. In these cases investigation will be controlled by the Work Order process.

2.3 Formal troubleshooting activities shall be performed in accordance with this procedure unless waived by plant management or management within the affected department. If waived, the justification shall be documented in the appropriate location (Condition Report, Work Order, etc).

2.4 This procedure does not apply to special tests.

2.5 All troubleshooting shall use high impedance M&TE and/or isolation transformers on the signal and AC power source unless low impedance is specifically called for in equipment procedure.

3.0 Definitions 3.1 Troubleshooting - A systematic approach to data collection, failure analysis, or a measurement plan that results in high confidence that the complete cause of system/equipment degradation has been determined. There may be potential personnel safety risk.

3.2 High Risk Troubleshooting - Potential impacts are assessed as high risk when evaluated per NMP-DP-001, Operational Risk Awareness.

3.3 Medium Risk Troubleshooting - Potential impacts are assessed as medium risk when evaluated per NMP-DP-001, Operational Risk Awareness.

3.4 Low Risk Troubleshooting - Potential impacts are assessed as low risk when evaluated per NMP-DP-001, Operational Risk Awareness.

3.5 Non-Intrusive Monitoring - The act of monitoring a component or system by not affecting normal operation of the component or system. Examples would be using Voltage Test Jacks, monitoring voltages across relay contacts, power supplies, etc.

3.6 Intrusive Monitoring - The act of temporarily altering the system to allow monitoring a component or system. This applies to electrical or mechanical testing methods.

3.7 Stop-Decision Points - Administrative and Physical Hold Points within the Troubleshooting Plan to limit and control activities.

Southern Nuclear Operating Company Nuclear Problem Solving and Troubleshooting NMP-AD-002 Management Guidelines Version 10.0 Procedure Page 5 of 14 4.0 Responsibilities 4.1 Troubleshooting Leader- Individual assigned to develop the troubleshooting plan, coordinate work and team discussions, act as a single point of contact and/or obtain changes to the plan, as assigned by the responsible department manager. Perform Just In Time Risk Assessment as described in NMP-DP-001, Operational Risk Awareness.

4.2 Operations Department The Operations Department, under the direction of the Operations Manager, is responsible for ensuring the troubleshooting activities are supported by:

Approve the Troubleshooting plan where risk has been assessed as Medium or High Providing personnel in support of the Troubleshooting Team Maintain approved system status during troubleshooting activities (i.e. Out Of Service) 4.3 Work Planning The group developing the Troubleshooting plan will determine the level of risk associated with the Troubleshooting plan by using Procedure NMP-DP-001, Operational Risk Awareness.

The plan should consider elimination of worst case, long lead time components early in the process, as potential causes.

Troubleshooting plan steps that alter the configuration of the plant will be implemented and controlled by a planned work order or use of referenced instructions from an approved procedure. This requirement may be waived by the operations shift manager in which case configuration changes will be controlled using detailed instructions in the troubleshooting plan and will be approved by operations before implementation.

Troubleshooting plan steps that do not affect plant configuration control, for example system walkdown, data collection and trending, field observation, and other similar fact finding steps may be implemented by the troubleshooting plan steps.

4.4 Maintenance Department The Maintenance Department, under the direction of the Maintenance Manager, is responsible for:

Maintenance Manager or his designee will approve High or medium risk Troubleshooting activities where personal safety or economic safety are assessed as Medium or High risk Providing personnel in support of the Troubleshooting Team Ensuring that Troubleshooting Plan is performed and documented in accordance with approved site procedures and safe work practices Determining the need for additional support for troubleshooting activities

1. G2.2.25 001/LOIT AND LOCT/SRO/M/F 3.2/4.2/G2.2.25/LO-LP-39209-02//HL-17 AUDIT/

Given the following:

- Unit 1 is at 100% reactor power.

- The following RWST parameters are recorded:

Temperature is 47°F.

Level is 93%.

Which one of the following completes the following statement?

Tech Spec action is required for RWST __(1)__,

and the Tech Spec Basis for this parameter limit is to ensure __(2)__.

A. (1) Level (2) sufficient borated water to support the ECCS during the injection phase of a design basis main steam line break B. (1) Level (2) sufficient borated water to support the ECCS during the injection phase of a design basis loss of coolant accident C. (1) Temperature (2) that the amount of cooling provided from the RWST during the heatup phase of a main steam line break is consistent with safety analysis assumptions D. (1) Temperature (2) that the amount of cooling provided from the RWST during the heatup phase of a main feed line break is consistent with safety analysis assumptions K/A 2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

K/A MATCH ANALYSIS The question asks the candidate straight forward and RO level question of which parameter will place the unit in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown LCO, RWST level or temperature.

Once determining RWST level is the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown, the SRO portion requires the candidate to know the correct bases for the RWST level.

Tuesday, February 25, 2014 11:49:11 AM 1

EXPLANATION OF REQUIRED KNOWLEDGE Per TS SR 3.5.4.1, the RWST borated water temperature must be maintained greater than or equal to 44°F and less than or equal to 116°F. Per TS 3.5.4 Bases, the maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a feedline break is consistent with the safety analysis assumptions. The minimum temperature is an assumption in both the MSLB and inadvertent ECCS actuation. The inadvertent ECCS actuation is typically non-limiting.

Per TS SR 3.5.4.1, the RWST borated water must be maintained greater than or equal to 686,000 gallons. Per Tech Spec rounds OSP 14000-1, 686,000 gallons corresponds to an indicated level of 94%. Per TS 3.5.4 Bases, the RWST volume is an explicit assumption for LOCA events. The desired volume limit is set by LOCA and containment analyses. The volume is not an explicit assumption for non-LOCA events since the required volume is a small fraction of the available volume. The deliverable volume is different from the total volume contained since, due to the design of the tank, more water can be maintained than can be delivered.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. 93% level will place the unit in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown.

The second part is incorrect. RWST level ensures sufficient injection volume for a DBA LOCA, not a DBA MSLB. However, Safety Injection is required to mitigate a main steam line break and requires suction from the RWST, but is not a limiting factor since the required volume would be a small fraction of that available.

B. Correct. The first part is correct. See the first part of choice A above.

The second part is correct. Per TS 3.5.4 Bases, the RWST volume is an explicit assumption for LOCA events C. Incorrect.Plausible. The first part is incorrect. Per TS SR 3.5.4.1, the RWST borated water temperature must be maintained greater than or equal to 44°F and less than or equal to 116°F.

The second part is incorrect. Per TS 3.5.4 Bases, the maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a feedline break is consistent with the safety analysis assumptions, not a main steamline break.

D. Incorrect.Plausible. The first part is incorrect. See the first part of choice C above.

The second part is the correct for the first part. Per TS 3.5.4 Bases, the maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a Tuesday, February 25, 2014 11:49:11 AM 2

feedline break is consistent with the safety analysis assumptions, not a main steamline break.

SRO-ONLY JUSTIFICATION (2) Facility operating limitations in the technical specifications and their bases.

-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the question requires knowledge of the TS Bases for the limit.

-Can question be answered solely by knowing the LCO/TRM information listed above-the-line? No, the question is not related to above-the-line Tech Spec.

The information is found in a survelliance and TS Bases.

-Can question be answered solely by knowing the TS Safety Limits? No, the question is not related to any Tech Spec Safety Limit.

-Does the question involve one or more of the following for TS,TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)

• Knowledge of TS bases that is required to analyze TS required actions and terminology. Yes, the question asks the bases for a Tech Spec survelliance.

Tuesday, February 25, 2014 11:49:11 AM 3

Level: SRO Tier # / Group # T3 K/A# G2.2.25 Importance Rating: 3.2 / 4.2 Technical

Reference:

TS 3.5.4, Ammendment No. 96, page 3.5.4-1 & 2 TS Bases 3.5.4, Rev 0, pages B3.5.4-3 & 4 OSP 14000-1, Rev 88.1, pages 8 & 15 References provided: None Learning Objective: LO-LP-39209-01 For any given item in section 3.5 of Tech Specs, be able to:

a. State the LCO.

b. State any one hour or less required actions.

LO-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. Whether any Tech Spec LCOs of section 3.5 are exceeded.

b. The required actions for all section 3.5 LCOs.

LO-LP-39209-03 Describe the bases for any given Tech Spec in section 3.5.

Question origin: BANK - LOIT Question # 006G2.2.39 001 Cognitive Level: M/F 10 CFR Part 55 Content: 41.8 / 43.2 Comments:

You have completed the test!

Tuesday, February 25, 2014 11:49:11 AM 4

RWST 3.5.4 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RWST boron A.1 Restore RWST to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> concentration not within OPERABLE status.

limits.

OR RWST borated water temperature not within limits.

B. One or more sludge B.1 Restore the valve(s) to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mixing pump isolation OPERABLE status.

valves inoperable.

C. Required Action and C.1 Isolate the sludge mixing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion system.

Time of Condition B not met.

D. RWST inoperable for D.1 Restore RWST to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reasons other than OPERABLE status.

Condition A or B.

(continued)

Vogtle Units 1 and 2 3.5.4-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

RWST 3.5.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or D AND not met.

E.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 --------------------------NOTE-------------------------------

Only required to be performed when ambient air temperature is < 40°F.

Verify RWST borated water temperature is In accordance with 44°F and 116°F. the Surveillance Frequency Control Program SR 3.5.4.2 Verify RWST borated water volume is 686,000 In accordance with gallons. the Surveillance Frequency Control Level specified by Program

% in TS Rounds.

SR 3.5.4.3 Verify RWST boron concentration is 2400 ppm In accordance with and 2600 ppm. the Surveillance Frequency Control Program SR 3.5.4.4 Verify each sludge mixing pump isolation valve In accordance with automatically closes on an actual or simulated the Surveillance RWST Low-Level signal. Frequency Control Program Vogtle Units 1 and 2 3.5.4-2 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

RWST B 3.5.4 BASES BACKGROUND reduction of SDM or excessive boric acid precipitation in the core (continued) following the LOCA, as well as excessive stress corrosion of mechanical components and systems inside the containment.

APPLICABLE During accident conditions, the RWST provides a source of SAFETY ANALYSES borated water to the ECCS and Containment Spray System pumps.

As such, it provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for reactor shutdown (Ref. 1). The design basis transients and applicable safety analyses concerning each of these systems are discussed in the Applicable Safety Analyses section of B 3.5.2, "ECCS Operating"; B 3.5.3, "ECCS Shutdown"; and B 3.6.6, "Containment Spray and Cooling Systems." These analyses are used to assess changes to the RWST in order to evaluate their effects in relation to the acceptance limits in the analyses.

The RWST must also meet volume, boron concentration, and temperature requirements for non-LOCA events. The volume is not an explicit assumption in non-LOCA events since the required volume is a small fraction of the available volume. The deliverable volume limit is set by the LOCA and containment analyses. For the RWST, the deliverable volume is different from the total volume contained since, due to the design of the tank, more water can be contained than can be delivered. The minimum boron concentration is an explicit assumption in the main steam line break (MSLB) analysis to ensure the required shutdown capability. The maximum boron concentration is an explicit assumption in the inadvertent ECCS actuation analysis, although it is typically a nonlimiting event and the results are very insensitive to boron concentrations. The maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a feedline break is consistent with safety analysis assumptions; the minimum is an assumption in both the MSLB and inadvertent ECCS actuation analyses, although the inadvertent ECCS actuation event is typically nonlimiting.

The MSLB analysis has considered a delay associated with the interlock between the VCT and RWST isolation valves, and the (continued)

Vogtle Units 1 and 2 B 3.5.4-3 Revision No. 0

RWST B 3.5.4 BASES APPLICABLE results show that the departure from nucleate boiling design SAFETY ANALYSES basis is met. The delay has been established as 27 seconds, (continued) with offsite power available, or 39 seconds without offsite power (includes 12 seconds for the Emergency Diesel Generator). This response time includes an electronics delay, a stroke time for the RWST valves, and a stroke time for the VCT valves.

For a large break LOCA analysis, the minimum water volume limit of 499,091 gallons and the lower boron concentration limit of 2400 ppm are used to compute the post LOCA sump boron concentration necessary to assure subcriticality. The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core.

The upper limit on boron concentration of 2600 ppm is used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA. The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation in the core following the accident.

In the ECCS analysis, the containment spray temperature is assumed to be equal to the RWST lower temperature limit of 44qF. If the lower temperature limit is violated, the containment spray further reduces containment pressure, which decreases the rate at which steam can be vented out the break and increases peak clad temperature. (The reduction in containment pressure correspondingly reduces the density of the vented steam. This reduces the flow of steam out of the core, which translates into a decrease in the ECCS flooding rate. This decrease in the flooding rate causes the increase in peak clad temperature.) The upper temperature limit of 116qF is used in the small break LOCA analysis and containment OPERABILITY analysis.

Exceeding this temperature will result in a higher peak clad temperature, because there is less heat transfer from the core to the injected water for the small break LOCA and higher containment pressures due to reduced containment spray cooling capacity. For the containment response following an MSLB, the lower limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment.

The RWST satisfies Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

(continued)

Vogtle Units 1 and 2 B 3.5.4-4 Rev. 1-10/01

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 14000-1 88.1 Effective Date Page Number 06/21/2013 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS 8 of 36 Sheet 2 of 10 DATA SHEET 1 MODE 1 & 2 MODE _______________

DATE _______________

LCO TECH SPEC I N D I C A T I O N LIMIT(S)

METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE LCO/PROC EACH ACCUMULATOR SHALL 1HS-8808A BE OPERABLE SR 3.5.1.1 MLB001 2.3 VERIFY DISCHARGE VALVE VALVE POSITION 1HS-8808B OPEN 3.5.1 POSITION (INIT) MLB002 2.3 1HS-8808C MLB001 2.4 1HS-8808D MLB002 2.4 TWO ECCS FLOW TRAINS 1HS-8806 OPEN AND POWER SHALL BE OPERABLE SR 3.5.2.1 REMOVED VERIFY VALVES POSITIONED 1HS-8835 OPEN AND POWER AND POWER REMOVED REMOVED BY ASSOCIATED 1HS-8813 OPEN AND POWER LOCKOUT SWITCH LIGHT REMOVED EXTINGUISHED AND VALVE STATUS 1HS-8802A CLOSED AND 3.5.2 SWITCH IN LOCKOUT (INIT) POWER REMOVED POSITION 1HS-8802B CLOSED AND POWER REMOVED 1HS-8840 CLOSED AND POWER REMOVED 1HS-8809A OPEN AND POWER REMOVED 1HS-8809B OPEN AND POWER REMOVED ESFAS INSTRUMENTATION SR 3.3.2.1 1LI-0991A CHANNEL CHECK SHALL BE OPERABLE FCN 7B CHANNEL CHECK RWST LEVEL 1LI-0993A REQUIRED 4 3.3.2(K)

(%)

ACCIDENT MONITORING SR 3.3.3.1 1LI-0990A REQUIRED 2 3.3.3 INSTRUMENTATION SHALL FCN 9 (B,G,H,J)

BE OPERABLE 1LI-0992A CHANNEL CHECK SR 3.5.4.2 >94% 3.5.4 COMPLETED BY: DAY: TIME: NIGHT: TIME:

SS REVIEW: DAY: TIME: NIGHT: TIME:

Printed January 21, 2014 at 15:43

Approved By Procedure Version J.B. Stanley Vogtle Electric Generating Plant 14000-1 88.1 Effective Date Page Number 06/21/2013 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS 15 of 36 Sheet 9 of 10 DATA SHEET 1 MODE 1 & 2 MODE _______________

DATE _______________

LCO TECH SPEC I N D I C A T I O N LIMIT(S)

METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE LCO/PROC CREFS ACTUATION SR 3.3.7.1 CR INTAKE 1RE-12116 OPERABLE FCN 3 RADIATION CHANNEL CHECK 3.3.7 CHANNEL CHECK MONITORS 1RE-12117 REQUIRED 2 (INIT)

FHB ACTUATION TRS 13.3.6.1 FHB EFFL ARE-2532A 13.3.6 OPERABLE RADIOGAS

• CHANNEL CHECK FHB ISO ARE-2532B REQUIRED 1 (INIT)

FHB ACTUATION TRS 13.3.6.1 FHB EFFL ARE-2533A

• 13.3.6 OPERABLE RADIOGAS CHANNEL CHECK FHB ISO ARE-2533B REQUIRED 1 (INIT)

• INDICATING NORMALLY. ALL STATUS AND ALARM LIGHTS EXTINGUISHED.

DG1A FUEL OIL INVENTORY SR 3.8.3.1 DG 1A LEVEL 1-LI-9024 82% 3.8.3 VERIFY FUEL OIL STORAGE (%)

TANK LEVEL DG1B FUEL OIL INVENTORY SR 3.8.3.1 DG 1B LEVEL 1-LI-9025 82% 3.8.3 VERIFY FUEL OIL STORAGE (%)

TANK LEVEL TWO INDEPENDENT SR 3.7.10.1 NOTE: TEMPERATURE INDICATION IS OBTAINED FROM HAND-HELD TEST EQUIPMENT.

CONTROL ROOM EMERGENCY SR 3.7.11.1 RECORD INSTRUMENT INFORMATION BELOW.

FILTRATION SYSTEMS INSTRUMENT ID NO. N/A SHALL BE OPERABLE VERIFY CONTROL ROOM CAL DUE DATE TEMP CONTROL ROOM M&TE <85F 3.7.10 TEMPERATURE 3.7.11 (F)

THE RWST SHALL BE SR 3.5.4.1 RWST >51F

• 3.5.4 OPERABLE TRS 13.1.7.1 TEMPERATURE 1TIS-10980 13.1.7 VERIFY TEMPERATURE (F) <109F *

• WITH INDICATED RWST TEMPERATURE OUTSIDE THE LIMITS, THEN VERIFY RWST TEMPERATURE IS WITHIN TECHNICAL SPECIFICATION LIMITS BY PLACING THE RWST ON RECIRC USING SLUDGE MIXING PUMP WITH HEATER OFF AND OBSERVING 1-TI-10982 TO BE WITHIN 44F AND 116F.

THE ULTIMATE HEAT COMPUTER POINT SINK SHALL BE OPERABLE T2601* <90F 3.7.9 VERIFY WATER -OR-TEMPERATURE AND LEVEL SR 3.7.9.2 TEMPERATURE 1TJI-1692 (F) POINT 2*

COMPUTER POINT T2602*

-OR-1TJI-1692 POINT 17*

• IF COMPUTER POINT AND RECORDER POINT ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT. RECORD INSTRUMENT INFORMATION BELOW.

INSTRUMENT ID NO. N/A CAL DUE DATE 1LI-1606 >73%

SR 3.7.9.1 LEVEL

(%) 1LI-1607 CONTAINMENT AIR SR 3.6.5.1 COMPUTER POINT TEMPERATURE SHALL NOT T2501 EXCEED 120F TEMPERATURE COMPUTER POINT VERIFY AVERAGE AIR (F) T2502 NA TEMPERATURE COMPUTER POINT T2503 COMPUTER POINT 3.6.5 UT2501 (AVG) <120F

• IF COMPUTER POINT IS NOT AVAILABLE ALB-01 (E06)

VERIFY CNMT HI TEMP ALARM NOT IN ALARM ALB-01 (E06) IS NOT IN ALARM.

• IF COMPUTER POINT AND ALB-01 (E06) ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT FOR 1TE-2612 FOR POINT T2502 AND 1TE-2613. FOR POINT T2503 RECORD INSTRUMENT INFORMATION BELOW. USE MCB INDICATOR 1TI-2563 FOR POINT T2501 AND AVERAGE THE THREE.

INSTRUMENT ID NO.

<120F CAL DUE DATE COMPLETED BY: DAY: TIME: NIGHT: TIME:

SS REVIEW: DAY: TIME: NIGHT: TIME:

Printed January 21, 2014 at 15:43

1. G2.3.4 001/LOCT AND LOIT/SRO/M/F 3.2/3.7/G2.3.4/LO-LP-40101-08///

Initial condition:

- General Emergency has been declared.

Current conditions:

- A first responder is briefed to rescue an injured worker.

- Health Physics estimates the first responder will receive 11 rem TEDE dose while performing the rescue.

Which one of the following completes the following statement?

Per 91301-C, "Emergency Exposure Guidelines," the dose received by the first responder during the rescue __(1)__ be added to the responder's occupational non-emergency exposure, and the LOWEST level of approval required to authorize the first responder's rescue exposure is the __(2)__.

__(1)__ __(2)__

A. will Health Physics Supervisor B. will Emergency Director C. will NOT Health Physics Supervisor D. will NOT Emergency Director K/A 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

K/A MATCH ANALYSIS The question sets up a plausible scenario which includes all the required KA elements.

First the SRO candidate must determine if the exposure in the General Emergency would be added to normal exposure already accumulated and the authorization level required for the given dose.

EXPLANATION OF REQUIRED KNOWLEDGE Per Admin procedure 91301-C, "Emergency Exposure Guidelines" NOTE on TABLE 1, dose to workers performing emergency services may be treated as a once-in-a-lifetime Tuesday, February 25, 2014 1:44:48 PM 1

exposure and should not be added to occupational exposure accumulated under non-emergency conditions. Per Responsibilities 2.1, the Emergency Director (ED) has the sole authority to allow radiation exposures in excess of 10CFR20 limits. Per Responsibilities 2.2.4, the HP Supervisor or designedd can authorize individuals to recieve radiation exposures in excess of VEGP Admin Guidelines, but not in excess of 10CFR20 limits.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Part 1 is incorrect however plausible since its reasonable to assume the candidate may determine that all radiation exposure for the year would be additive to ensure accurate accounting for health reasons.

Part 2 is incorrect however plausible since the candidate may determine, based on plant conditions, that with the exposure less than 25 Rem, that Emergency Director involvement would not be required and therefore the Health Physics Supervisor could authorizes this.

B. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice A above.

Part 2 is correct. Per 91301-C EMERGENCY EXPOSURE GUIDELINES, the Emergency Director (ED) has the sole authority to allow radiation exposures in excess of 10CFR20 limits.

C. Incorrect. Plausible. Part 1 is correct. Per 91301-C EMERGENCY EXPOSURE GUIDELINES, dose to workers performing emergency services may be treated as an once-in-a-lifetime exposure and should not be added to occupational exposure accumulated under non-emergency conditions.

Part 2 is incorrect. See Part 2 of choice A above.

D. Correct. Part 1 is correct. See Part 1 of choice C above..

Part 2 is correct. See part 2 of choice B abvoe.

SRO JUSTIFICATION (10CFR43(b))

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, the knowledge required pertains to administrative guidance and ED non deligable duties.

-Can the question be answered solely by knowing immediate operator actions?

No, information found in IOAs is not involved in the question.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, EOPs and AOPs Tuesday, February 25, 2014 1:44:48 PM 2

are not involved with this questioon.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, detailed knowledge of process and responsibilities within an admin procedure are required. Overall knowledge will not answer the question.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. Yes, specific knowledge of the duties of the ED and HP Supervisor as well as how emergency exposures are recorded for dose purposes is required.

Level: SRO Tier # / Group # T3 K/A# G2.3.4 Importance Rating: 3.2 / 3.7 Technical

Reference:

ADMIN 91301-C, Rev 12.1, pages 3 & 9 References provided: None Learning Objective: LO-LP-40101-35 State what group of people should be first considered for emergency exposure, and what group should not be allowed to receive an emergency exposure (91301-C). (SRO only)

LO-LP-40101-08 State from memory ED duties that cannot be delegated (SRO only).

Question origin: MODIFIED - Turkey Point 2011 NRC Question # G.2.3.4 Cognitive Level: M/F 10 CFR Part 55 Content: 43.5 Comments:

You have completed the test!

Tuesday, February 25, 2014 1:44:48 PM 3

Approved By Procedure No. Version S. C. Swanson Vogtle Electric Generating Plant 91301-C 12.1 Effective Date Page Number 03/12/2013 EMERGENCY EXPOSURE GUIDELINES 3 of 16 INFORMATION USE 1.0 PURPOSE The purpose of this procedure is to provide instructions and controls for radiation exposures in excess of the Vogtle Electric Generating Plant (VEGP)

Administrative Guidelines, or in excess of the 10CFR20 occupational limits during emergency conditions.

2.0 RESPONSIBILITIES 2.1 The Emergency Director (ED) has the sole authority to allow radiation exposures in excess of 10CFR20 limits in accordance with the provisions of this procedure.

2.2 The Health Physics (HP) Supervisor, or designee, shall have the following responsibilities:

2.2.1 Preparing Permits for Emergency Radiation Exposure (PERE). (1985304698) 2.2.2 Maintaining records of emergency exposures for each individual.

2.2.3 Providing recommendations to the ED on exposure control measures including issuance of Dosimetry, use of protective equipment and issuance of thyroid blocking agents such as potassium iodide (KI).

2.2.4 Authorizing individuals to receive radiation exposures in excess of the VEGP Administrative Guidelines, but which do not exceed the 10CFR20 limits.

3.0 PREREQUISITES An emergency situation exists which results in a need to initiate corrective actions, protective actions, sampling activities, or lifesaving measures which might result in exposures greater than 10CFR20 limits.

4.0 PRECAUTIONS 4.1 Personnel authorized to receive exposures in excess of 10CFR20 limits shall meet the following criteria: (1985305022) 4.1.1 Personnel shall be familiar with the risks of exposure to the higher radiation levels which are likely during emergency conditions as outlined in Table 2 and Table 3.

Printed October 4, 2013 at 12:52

Approved By Procedure No. Version S. C. Swanson Vogtle Electric Generating Plant 91301-C 12.1 Effective Date Page Number 03/12/2013 EMERGENCY EXPOSURE GUIDELINES 9 of 16 TABLE 1 EMERGENCY EXPOSURE GUIDELINES (1985305256) (1985305827)

NOTES Dose limits listed in this table apply to doses incurred over the duration of the emergency.

Dose to workers performing emergency services may be treated as an once-in-a-lifetime exposure and should not be added to occupational exposure accumulated under non-emergency conditions.

Workers performing services during emergencies shall limit dose to the lens of the eyes to three times the listed value and doses to any other organ (including skin and body extremities) to ten times the listed value.

Dose Limit (REM)

Total Effective Activity Condition Dose Equivalent 5 All 10 Protecting Valuable Property Lower Dose not practicable 25 Lifesaving or protection of Lower Dose not practicable large population

>25 Lifesaving or protection of Only on a voluntary basis to large population persons fully aware of the risks involved Printed November 19, 2013 at 15:22

1. G2.3.7 001/LOIT AND LOCT/SRO/M/F 3.5/3.6/G2.3.7/LO-LP-63920-03///

Given the following:

- A Systems Operator (SO) will make multiple entries into AB-A-33 to place the CVCS cation demineralizer in service.

- The SO will use RWP 14-0108 (red RWP).

- The SO will exceed the Annual Administrative 4000 mrem per year TEDE limit during the task.

Which one of the following completes the following statement?

The SO __(1)__ required to receive an ALARA briefing prior to each AB-A-33 entry, and per NMP-HP-001, "Radiation Protection Standard Practices," the __(2)__ is the LOWEST level of approval required to exceed the Administrative dose limit.

__(1)__ __(2)__

A. is HP Manager B. is Plant Manager C. is NOT HP Manager D. is NOT Plant Manager Tuesday, February 25, 2014 2:59:06 PM 1

K/A G2.3.7 Radiation Control Ability to comply with radiation work permit requirements during normal or abnormal conditions:

K/A MATCH ANALYSIS:

The candidate is presented with a scenario where an Auxiliary Building Operator is required to enter several rooms with high dose rates in the area. The Operator is also on the verge of exceeding his Annual TEDE limits of 4000 mrem. The candidate has to determine the minimum level of authority that may approve exceeding the annual TEDE limits.

EXPLANATION OF REQUIRED KNOWLEDGE Per NMP-HP206 step 5.3.31.2 states that RED RWPs are "Single Use" type briefings.

Step 5.3.11.1.4 defines Single Use as "individuals must be authorized and the authorization is good for one entry only."

Per NMP-HP-001 step 6.2.3, there are 3 different administrative levels that require 3 different levels of approval:

1. 2000 mrem in a year requires HP Supervisor, Physicist, or Manager approval.

2. 4000 mrem in a year requres AGM or Plant General Manager approval.

3. 4500 mrem in a year requires Project Vice President approval.

Dose in excess of 5mrem requires special NRC approval for normal operation, ED approval during emergencies. These different admin limits do not have a noun description/name and therefore are generally referred to by the associated dose limit.

DISTRACTOR ANALYSIS:

A. Incorrect. Plausible. Part 1 is correct. RWP 14-0108 is a red RWP and therefore requires a Single Use briefing on each entry per NMP-HP-206.

Part 2 is incorrect. Per NMP-HP-001, exceeding the 4000mrem per year admin dose limit will required Plant General Manager approval. However, the HP Supervisor can approval all dose up to the 4000mrem limit. Therefore, this distractor is plausible.

B. Correct. Part 1 is correct.RED RWP and requires a briefing for each entry. The minimum authority level to approve the TEDE extension is the Plant Manager from the choices presented.

Part 2 is correct. Per NMP-HP-001, exceeding the 4000mrem per year admin dose limit will required Plant General Manager approval.

Tuesday, February 25, 2014 2:59:48 PM 1

C. Incorrect. Plausible. Part 1 is incorrect. RWP 14-0108 is a red RWP and therefore requires a Single Use briefing on each entry per NMP-HP-206.

However, both yellow and green RWPs allow re-entry into the room with a single HP brief. Therefore, this distractor is plausible.

Part 2 is incorrect. See Part 2 of choice A above.

D. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice C above.

Part 2 is correct. See Part 2 of choice B above.

SRO JUSTIFICATION (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Yes, specific knowledge of administrative procedures associated with radiological safety and rad exposure authorization levels during normal plant conditions is tested.

Level: SRO Tier # / Group # T3 K/A# G2.3.7 Importance Rating: 3.5 / 3.6 Technical

Reference:

NMP-HP-001, Rev 5.2, page 14 & 15 NMP-HP-206, Rev 3.0, pages 8 &12 V-LO-LP-63930, page 12 References provided: None Learning Objective: LO-LP-63930-06 State the entry requirements applicable to each of the following:

b. Radiation Control Area (RCA)

c. Radiation Area

d. High Radiation Area

e. Locked High Radiation Area LO-LP-63920-03 State the plant administrative limits/guidelines for radiation dose.

LO-LP-63920-04 State the actions to be taken if administrative dose limits are being approached.

Question origin: MODIFIED - HL18 NRC - G2.3.7 Cognitive Level: M/F 10 CFR Part 55 Content: 41.12 / 43.4 You have completed the test!

Tuesday, February 25, 2014 2:59:48 PM 2

1. G2.3.7 003/LOCT AND LOIT/SRO/M/F 3.5/3.6/G2.3.7///HL18 NRC/079G2.1.27 (Original Question from HL18 NRC)

Given the following:

- A Fuel Handling Coordinator (FHC) is entering the Spent Fuel Pool area.

- The FHC is reviewing his RWP prior to beginning work and notices an ALARA briefing is required.

- The dose rate is 900 mrem/hour due to damaged fuel assemblies.

- The FHC will also exceed 2000 mrem Annual TEDE limits while in the area.

Which one of the following completes the following statement?

Based on the area dose rate, the FHC will be required to receive an ALARA briefing prior to __(1)__ entry, and per NMP-HP-001, "Radiation Protection Standard Practices", the ___(2)___ is the MINIMUM authority level required to exceed the Annual TEDE limit.

__(1)__ __(2)__

A. each HP Manager B. each Plant General Manager C. ONLY the first HP Manager D. ONLY the first Plant General Manager Friday, November 15, 2013 3:12:19 PM 1

Southern Nuclear Operating Company NMP-HP-001 Nuclear Radiation Protection Standard Practices Version 5.2 Management Page 14 of 16 Procedure 6.1.55 20.2202 Notification of Incidents Paragraph (a)(2) and (b)(2), for reporting the release of radioactive material inside or outside of a restricted area is caveated as not applicable to, "locations where personnel are not normally stationed during routine operations, such as hot cells or process enclosures." For purposes of this section, consistent with the answer to question 56 of the first NRC 10CFR20 Q & A document, locations where personnel are not normally stationed will be interpreted as areas, rooms and enclosures which are not normally occupied nor periodically patrolled during normal plant operations and maintenance.

6.1.56 20.2203 Reports of Exposures, Radiation Levels and Concentrations of Radioactive Material Exceeding the Limits No fleet practices identified.

6.1.57 20.2204 Reports of Planned Special Exposures No fleet practices identified.

6.1.58 20.2206 Reports of Individual Monitoring The intent of Regulatory Guide 8.7, "Instructions for Recording and Reporting Occupational Radiation Exposure Data," will be met in complying with this paragraph with NRC Form 5.

6.1.59 20.2301 Applications for Exemptions No fleet practices identified.

6.1.60 20.2302 Additional Requirements No fleet practices identified.

6.1.61 20.2401 Violations No fleet practices identified.

6.2 Other Consensus Positions 6.2.1 Whole Body Count Performance Frequency Monitored workers will be given an entrance and exit whole body count (WBC) or whole body scan (WBS). The exit WBC from another SNC site can be used in lieu of an entrance WBC if the SNC site was the last site the worker entered an RCA and/or monitored. Upon request from a worker, WBCs will be provided to the worker on a voluntary and reasonable basis.

6.2.2 Dose Limits for Workers Who Provide Outage Support at a SNC Plant Other Than Their Home Plant.

Workers should be limited to 500 mrem per visit, unless express consent is given by the home plants management to exceed that limit.

6.2.3 Administrative Annual TEDE Dose Limits and the Approval Authority Necessary to Exceed Limits 6.2.3.1 2000 mrem in a year requires HP Support Supervisor, Plant Health Physicist, or HP Manager approval.

Southern Nuclear Operating Company NMP-HP-001 Nuclear Radiation Protection Standard Practices Version 5.2 Management Page 15 of 16 Procedure 6.2.3.2 4000 mrem in a year requires AGM or Plant General Manager approval.

6.2.3.3 4500 mrem in a year requires Project Vice President approval.

6.2.4 Discrepant Dosimeter Investigation Criteria An assessment of workers dose should be initiated for discrepant dosimetry results when the following criteria are met: the primary or secondary dosimeter dose exceeds 100 mrem; and the secondary dosimeter reading differs by more than 25% from the primary dosimeter.

6.2.5 Dose Monitoring Threshold All individuals entering an RCA will be monitored for radiation exposure. A single dosimeter suffices for Visitors or Radiation Workers whose annual dose from sources external to the body is not expected to exceed 100 mrem at a particular station. An Optically Stimulated Luminescent Dosimeter (OSLD) and a self-reading dosimeter, such as an Electronic Dosimeter (ED), will be provided to all other individuals.

6.2.6 Training Requirements for Visitors or Temporary Radiation Workers Who Enter RCAs All SNC plants will administer training to all who must enter the RCAs in the following manner:

6.2.6.1 If the individual is expected to receive < 100 mrem in a year, the individual will be escorted by a GET qualified worker and will be provided instructions.

6.2.6.2 If the individual is expected to receive >100 and <500 mrem in a year, the individual will receive a visitor handout (containing all of the instruction elements required by 10CFR19.12) and will acknowledge receipt of the instructions or handout by signing a form. The individual will have a GET-trained radiological escort. If special circumstances dictate that entries into contaminated areas are required, the individual will receive dress-out training (if he has no history of such training at our plants). Similarly, if the individual requires entry to high radiation areas, special training or instructions may be required.

6.2.6.3 If the individual is expected to receive >=500 mrem in a year, the individual is required to complete GET which includes testing. If the individual has previous GET training at a nuclear facility within the past two years or as allowed by the Training Department, then exemption GET and the exemption GET test can be administered.

6.2.7 Air Flow for Fume Hoods Containing Radioactive Materials/Fluids Sample station fume hoods containing radioactive materials/fluids will meet a minimum air flow acceptance criteria of 100 LFPM, and will be labeled to ensure that 100 LFPM is maintained or exceeded.

6.2.8 Respirator Training and Fit Test Frequencies Sites will apply a program whereby classroom or Computer-based training with an examination will be conducted annually. For individuals required to utilize respirator

Southern Nuclear Operating Company Nuclear NMP-HP-206 Issuance, Use and Control of Management Version 3.0 Radiation Work Permits Procedure Page 8 of 32 5.3.11 Enter the End Date in the same manner as step 5.3.10. When the End Date is entered, press the Tab key to advance to the Authorization Type field.

5.3.11.1 The type of worker authorization is based upon the type of RWP being written.

Worker authorizations fall into 4 categories:

5.3.11.1.1 All - Anyone can use this RWP.

5.3.11.1.2 Work Group - Work Group must be authorized.

5.3.11.1.3 Individual - Individuals must be authorized.

5.3.11.1.4 Single Use - Individuals must be authorized and the authorization is good for one entry only.

5.3.12 From the pull-down menu, select All or Work Group for a Green RWP. Select Individual for a Yellow RWP. Select Single Use for a Red RWP.

5.3.13 Press the Tab key to advance to the RWP Type field.

5.3.14 With the cursor in the Type field, select the appropriate RWP type (General or Specific) from the pull-down table. Press the Tab key to advance to the Principle Work Document field.

5.3.15 With the cursor in the Principle Work Document field, type the activity or MWO number, if applicable. If the RWP is not written for a specific activity or MWO, this field may be either left blank or type N/A in the field.

5.3.16 Press the Tab key and advance past the ALARA Review Number field. Tab to the HP Job Coverage field.

5.3.17 In the HP Coverage field, enter None, Intermittent, or Continuous as appropriate for the RWP. Table 1 defines the type of HP coverage used.

5.3.18 Press the Tab key to advance to the Job Description field.

5.3.19 Type a short general description of the work to be performed in the Job Description field. Press Tab to advance to the Location field.

NOTE For an RWP, Location is the area where the majority of the work should be performed.

5.3.20 Type the work location code or select a location from the pull-down table in the Location field.

5.3.21 When the work location is selected, press Tab to advance to the Area field.

5.3.22 The Area field is normally left blank or N/A is typed in this field. Press Tab to advance to the Comments field. Enter comments as needed or leave blank.

Southern Nuclear Operating Company Nuclear NMP-HP-206 Issuance, Use and Control of Management Version 3.0 Radiation Work Permits Procedure Page 12 of 32 5.3.29.2 When Detail is selected, a Dosimetry Types dialog box will appear. Check the box next to the type of dosimetry required for the RWP. Click on the OK button.

5.3.29.3 Click on the Apply button at the bottom of the screen.

5.3.30 Select the Worker Instructions tab in the Maintain RWP screen.

5.3.30.1 With the cursor in the blank Worker Instructions field, click on the right mouse button and select Detail from the pop-up table.

5.3.30.2 When Detail is selected, a Worker Instructions dialogue box will appear. Place a check mark next to the applicable work instructions, and select OK.

5.3.30.3 If working from a model, Rem_occ may be used to remove any instructions that may be no longer necessary. To remove an instruction, place the cursor in the line of the instructions and click on the Rem_occ button.

5.3.30.4 Worker Instructions may be added as free form text.

5.3.30.5 Additional Worker Instructions may be added by using the Add_occ button.

5.3.30.6 When all Worker Instructions are entered, click on Apply at the bottom of the screen.

5.3.31 Select the Briefing tab in the Maintain RWP screen.

5.3.31.1 With the cursor in the Briefing field, right click on the mouse and select Detail from the pop-up table.

5.3.31.2 When Detail is selected, a dialogue box will open. Check the appropriate type of briefing for the RWP and select OK.

For Red RWPs, use Single Use briefing type.

For Yellow RWPs, the briefing type is conditional based on the activity.

For Green RWPs, no briefing is required.

5.3.31.3 Select Yes or No in the Required field if applicable.

5.3.31.4 After entering the briefing type, click on the Apply button at the bottom of the screen.

5.3.32 Select the Supervisors tab in the Maintain RWP screen.

5.3.32.1 Type the name of the job supervisor in the Job Supervisor field. Press the Tab key.

5.3.32.2 In the Department field, select the appropriate department from the pull-down table. Press the Tab key.

5.3.32.3 In the Phone/Ext field, type the phone number of the job supervisor. Press the Tab key.

V-LO-LP-63930 III. LESSON OUTLINE NOTES Risk-Based RWP Format and Requirements Color Code Radiological Types of Type of Briefing Required General Radiological Significance RWPs Conditions Dose Rate: < 100 All General mrem/hr RWPs Contamination Levels: < 200,000 No ALARA briefing required.

dpm/100 And 2 Green Low Specific RWPs Airborne Levels: < 0.3 DAC with low

• Workers should always refer to the radiological most recent survey information for risk the area(s) being worked in.

Initial ALARA briefing required Specific RWPs prior to first entry. Dose Rate: < 1000 that are tied to mrem/hr unique Work Additional ALARA briefing Groups - required when specified rad Contamination Levels: < 500,000 conditions are exceeded. dpm/100 cm2 Specific RWPs A pre-job ALARA briefing will be Airborne Levels: < 0.3 DAC Yellow Moderate covering work required if:

• Workers should always refer to the in areas with Radiological conditions that are intermediate addressed in the Worker most recent survey information for levels of Instructions section may be the area(s) being worked in.

radiological exceeded, or risk. If the RWP default settings for the accumulated dose or dose rate alarms may be exceeded, or Breach of a contaminated system ALARA Briefing required prior to each Dose Rate: > 1000 entry.

Radiological conditions on the mrem/hr RWP will be based on actual, Contamination Levels: > 500,000 Specific RWPs projected or historical survey covering work in information. dpm/100 cm2 High areas with high Red levels of Latest rad conditions and specific instructions will be covered in the Airborne Levels: > 0.3 DAC radiological risk. pre-job ALARA briefing

• Workers should always refer to the most recent survey information for the area(s) being worked in.

5. HP reviews active RWPs on a routine basis

6. Normally, HP will survey the work area prior to issuing an RWP a) In high radiation areas, survey performance may not be consistent with ALARA 12 of 22

1. G2.4.46 001/LOIT AND LOCT/SRO/C/A 4.2/4.2/G2.4.46/LO-TA-40002///056AG2.4.45 At time 1000:

- Unit 1 is in Mode 6.

At time 1005 the following alarms illuminate:

- ALB32-D02 RESV AUX XFMR 1NXRA HI SIDE PHOC LOR TRIP

- ALB32-E02 RESV AUX XFMR 1NXRB HI SIDE PHOC LOR TRIP

- ALB35-A10 DG1A TRIP OVERSPEED

- ALB35-F10 DG1A EMERGENCY START

- ALB36-A01 4160V SWGR 1AA02 TROUBLE

- ALB37-A01 4160V SWGR 1BA03 TROUBLE alarms, then subsequently clears.

- ALB38-F10 DG1B EMERGENCY START Current time is 1025:

Based on the current time, which one of the following is the correct Emergency Classification required to be declared?

REFERENCE PROVIDED A. Alert Emergency (CA3)

B. Alert Emergency (SA5)

C. Notification of Unusual Event (SU1)

D. Notification of Unusual Event (CU3)

G2.4.46 Emergency Procedures / Plan Ability to verify that the alarms are consistent with the plant conditions.

K/A MATCH ANALYSIS:

The candidate must analyze various alarms and indications associated with the electrical distribution system to determine plant status and the correct emergency classification. The event initiation time will also affect the classification.

EXPLANATION OF REQUIRED KNOWLEDGE The following annunciators are symptomatic of both train RATs being de-energized.

- ALB32-D02 RESV AUX XFMR 1NXRA HI SIDE PHOC LOR TRIP Tuesday, February 25, 2014 4:23:03 PM 1

- ALB32-E02 RESV AUX XFMR 1NXRB HI SIDE PHOC LOR TRIP The following annunciators are symptomatic of 1AA02 being de-energized.

- ALB35-A10 DG1A TRIP OVERSPEED

- ALB35-F10 DG1A EMERGENCY START

- ALB36-A01 4160V SWGR 1AA02 TROUBLE The following annunciators are symptomatic of 1BA03 de-energizing and then being re-energized by the 1B DG.

- ALB37-A01 4160V SWGR 1BA03 TROUBLE alarms, then subsequently clears.

- ALB38-F10 DG1B EMERGENCY START The question stem states that the plant is in Mode 6 and 20 minutes has elapsed since the loss of power event occured. Per NMP-EP-110 Figure 3, an NOUE should be classified based on Loss of All Offsite Power to Essential Buses for GREATER THAN 15 minutes and one EDG is supplying the 4160VAC bus.

An upgrade to an ALERT would occur if at any time the 1B DG fails to keep 1BA03 energized.

On NMP-EP-110 Figure 2, there are two similar classifications. The ALERT threshold is the same as the NOUE threshold in Mode 6. The difference arise out of a requirement to maintain 1 train in Mode 6 and 2 trains in Modes 1-4. A similar NOUE also exist for a Loss of All Offsite Power to Essential Buses for GREATER THAN 15 minutes and one EDG is supplying each of the 4160VAC buses.

ANSWER / DISTRACTOR ANALYSIS:

A. Incorrect. Plausible. CU3 is the correct classification. However, if the candidate does not recognize that 1BA03 is energized by the 1B DG and believes both 4160V buses are de-energized, then CA3 would be the correct classificaton.

B. Incorrect. Plausible. CU3 is the correct classification. However, if the candidate incorrectly utilizes NMP-EP-110 Figure 2 instead of Figure 3, then SA5 would match the conditions of the stem for the, but for the incorrect mode.

C. Incorrect. Plausible. CU3 is the correct classification. However, if the candidate incorrect diagnoses 1AA02 and believes it is energized by the 1A DG and also incorrect utilizes NMP-EP-110 Figure 2 instead of Figure 3, then SU1 would be the correct threshold.

D. Correct. CU3 is the correct classification. Loss of All Offsite Power to Essential Buses for GREATER THAN 15 minutes and one EDG is supplying the 4160VAC bus in Mode 6.

ANSWER / DISTRACTOR ANALYSIS Tuesday, February 25, 2014 4:23:03 PM 2

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, the answer requires specific knowledge of emergency classification thresholds.

-Can the question be answered solely by knowing immediate operator actions?

No, IOAs are not addressed by this question.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, the question does not address AOP or EOP entry conditions.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, the answer requires specific knowledge of emergency classification thresholds.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. Yes, the answer requires specific knowledge of emergency classification thresholds and determination of the specific classification based on current plant conditions. This is an SRO ONLY job link associated with an SRO ONLY objective. [LO-LP-40101-13 Given an emergency scenario, and the procedure, classify the emergency (SRO only).]

Tuesday, February 25, 2014 4:23:03 PM 3

Level: SRO Tier # / Group # T3 K/A# G2.4.46 Importance Rating: 4.2 / 4.2 Technical

Reference:

NMP-EP-110-GL03, Figure 2, Rev 3.0, page 122 NMP-EP-110-GL03, Figure 3, Rev 3.0, page 123 References provided: NMP-EP-110-GL03, Figure 1, Rev 3.0, page 121 NMP-EP-110-GL03, Figure 2, Rev 3.0, page 122 NMP-EP-110-GL03, Figure 3, Rev 3.0, page 123 Learning Objective: LO-TA-40002 Emergency Classification and Implementing Instructions using NMP-EP-110 (SRO Only)

LO-LP-40101-13 Given an emergency scenario, and the procedure, classify the emergency (SRO only).

LO-PP-11101-56 Predict the possible consequences of paralleling and loading the EDG when a loss of offsite power is anticipated.

LO-PP-11101-30 Describe in general terms the actions that occur on a normal or emergency start of a diesel engine up to and including the final condition of the diesel and the differences between a normal and emergency start.

LO-LP-60323-05 Given the entire AOP, describe:

a. Purpose of selected steps

b. How and why the step is being performed

c. Expected response of the plant/parameter(s) for the step LO-TA-60009A Respond to a Loss of Class 1E Electrical Systems per 18031-1/2 LO-TA-37018 Respond to a Loss of All AC Power per 19100-C LO-TA-11021 Respond to Diesel Generator Alarms Using Procedure 17035-1/2 or 17038-1/2 Question origin: MODIFIED - HL18 NRC Question # 056AG2.4.45 Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments:

You have completed the test!

Tuesday, February 25, 2014 4:23:03 PM 4

1. 056AG2.4.45 001/1/1/LOSP- EP/C/A-4.1/4.3/NEW/HL-18 NRC/SRO/

At 10:00:

- Unit 1 is in Mode 4.

At 10:05 the following alarms illuminate:

- ALB32-D02, RESV AUX XFMR 1NXRA HI SIDE PHOC LOR TRIP

- ALB32-E02, RESV AUX XFMR 1NXRB HI SIDE PHOC LOR TRIP

- ALB35-A10, DG1A TRIP OVERSPEED

- ALB35-F10, DG1A EMERGENCY START

- ALB36-A01, 4160V SWGR 1AA02 TROUBLE

- ALB37-A01, 4160V SWGR 1BA03 TROUBLE alarms, then subsequently clears.

- ALB38-F10, DG1B EMERGENCY START Current time is 10:25:

Based on the current time, which one of the following is the correct Emergency Classification required to be declared?

REFERENCE PROVIDED A. Alert Emergency (CA3)

B. Alert Emergency (SA5)

C. Notification of Unusual Event (SU1)

D. Notification of Unusual Event (CU3) 056AG2.4.45 Loss of Offsite Power Ability to prioritize and interpret the significance of each annunciator or alarm:

(CFR: 41.10 / 43.5 / 45.3 / 45.12)

K/A MATCH ANALYSIS:

The candidate is given various alarms and indications associated with the electrical distribution system. The candidate has to analyze the alarms to determine the plant status and determine the correct emergency classification, there is a time given when the event occurred that will also play into the classification.

The question is SRO only due to the Vogtle specific objective for Classification of an Emergency is an SRO only objective.

Tuesday, January 21, 2014 4:27:30 PM 1

ANSWER / DISTRACTOR ANALYSIS:

A. Incorrect. CA3 is a Cold Matrix classification, the plant is in Mode 4, not Mode 5 or 6. The mode was NOT stated in the question but just an RCS temperature given to increase the plausibility the candidate may select the wrong classification matrix. If the candidate selects the wrong matrix with the given alarms, it is plausible he could misinterpret the event and classify wrong. With the multiple alarms and indications, this can easily occur.

B. Correct. SA5 is the correct classification using the Hot Matrix, this is still a difficult determination with the multiple annunciator windows illuminated. The plant is only one failure away from a total plant blackout in this condition but the candidate has to determine this and correlate the event has been ongoing for > 15 minutes.

C. Incorrect. SU1 is a Hot Matrix classification. The plant is only one failure away from a total plant blackout in this condition but the candidate has to determine this and correlate the event has been ongoing for > 15 minutes. This choice is very plausible as the only difference between this and SA5 is that both diesels have to be carrying the buses to classify as SU1 versus 1 DG as in the correct choice.

This is a difficult determination with the mulitiple annuciators illuminated.

D. Incorrect. CA3 is a Cold Matrix classification, the plant is in Mode 4, not Mode 5 or 6. The mode was NOT stated in the question but just an RCS temperature given to increase the plausiblity the candidate may select the wrong classification matrix. IF, the plant were in Mode 5 and the Cold Matrix required to be used, this choice would then be correct.

REFERENCES:

The following references will be provided to the candidates during the exam.

NMP-EP-110, GL03, Figure 3, Cold Initiating Condition Emergency Action Level Matrix

- Modes 5, 6, and Defueled Only NMP-EP-110, GL03, Figure 2, Hot Initiating Condition Emergency Action Level Matrix -

Modes 1, 2, 3, and 4 Only NMP-EP-110, GL03, Figure 1, Fission Product Barrier Evaluation VEGP learning objectives:

LO-LP-40101-13 Given an emergency scenario, and the procedure, classify the emergency (SRO only).

This question is SRO only because the Emergency Plan is linked to a learning objective that is specifically labeled in the lesson plan as SRO Only.

You have completed the test!

Tuesday, January 21, 2014 4:27:30 PM 2

NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 FIGURE 2 Distractor Distractor 122

NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 4.0 FIGURE 3 Distractor Answer 124

1. WE11EA2.02 001/LOIT/SRO/M/F 3.4/3.9/WE11EA2.02/LO-TA-37020//HL15 NRC/

Initial conditions:

- Unit 1 experienced a LOCA.

- 19111-C, "Loss of Emergency Coolant Recirculation," was entered.

- RCS cooldown to cold shutdown has been initiated.

- RWST level is 8% and slowly lowering.

Current condition:

- Critical Safety Function Status Tree (CSFST) is ORANGE on Integrity.

Which one of the following completes the following statement?

The crew is required to __(1)__,

and then the Shift Supervisor __(2)__ required to transition to 19241-C, "Response to Imminent Pressurized Thermal Shock Condition."

A. (1) stop all pumps taking suction from the RWST (2) is B. (1) reduce ECCS flow from the RWST to one running train (2) is C. (1) stop all pumps taking suction from the RWST (2) is NOT D. (1) reduce ECCS flow from the RWST to one running train (2) is NOT K/A W/E11 Loss of Emergency Coolant Recirc. / 4 EA2.02 -Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation):

- Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments K/A MATCH ANALYSIS Thursday, February 27, 2014 9:13:31 AM 1

The question requires the candidate to make two decisions based on plant conditions while in 19111-C, "Loss of Emergency Coolant Recirculation" - stopping all ECCS pumps and transitioning out of 19111-C. Both decisions challenge the candidate's ability to adhere to the rules for using EOPs in compliance with WOG and facility requirements. Actions taken in accordance with these EOPs are part of the bases for granting the facility's license.

EXPLANATION OF REQUIRED KNOWLEDGE Per 19111-C steps 6 and 33 if RWST level lowers to <8%, all ECCS pumps taking suction from the RWST are to be placed in Pull-to-Lock (PTL).

Transition to any ORANGE or RED CSFST will be made when conditions are met. Per the rules of EOP usage, CSFSTs are initiated when either step 22 of 19000-C is reached, or a transition out of 19000-C is made. CSFSTs remain in effect during the entire EOP network unless otherwise directed. 19111-C does NOT contain any exceptional guidance on CSFST implementation.

ANSWER / DISTRACTOR ANALYSIS A. Correct. The first part is correct. Per continuous actions step 6 and steps 33 and 34, if RWST lowers to <8%, all ECCS pumps taking suction from the RWST are placed in PTL.

The second part is correct. CSFST monitoring was initiated on transition out of 19000-C. There is no guidance in 19111-C that prohibits actions based on CSFSTs. Therefore, a transition to 19241-C will be made as soon as ORANGE path conditions are verified.

B. Incorrect. Plausible. The first part is incorrect. Per continuous actions step 6 and steps 33 and 34, if RWST lowers to <8%, all ECCS pumps taking suction from the RWST are to be placed in PTL.

However, step 15 does reduce ECCS flow to only one train.

This is a mitigation strategy that prolongs RWST inventory. A candidate who does not possess the knowledge of the overall mitigating strategy of 19111-C could find it unreasonable to stop all ECCs pumps with a LOCA in progress and then transition to another procedure that does not address loss of injection flow.

The second part is correct. See the second part of choice A above.

C. Incorrect. Plausible. The first part is correct. See the first part of choice A above.

The second part is incorrect. There is no guidance in 19111-C that prohibits actions based on CSFSTs. Therefore, a transition to 19241-C would be made as soon as conditions for the ORANGE path are verified to exist. However, step 1 of EOP 19113-C, "Recirculation Sump Blockage" directs the operator to "initiate monitoring CSFSTs for information only. Function Thursday, February 27, 2014 9:13:31 AM 2

Restoration Procedures (FRP) should NOT be implemented." A candidate may confuse the two EOPs and believe transition out of 19111-C on CSFSTs is not allowed.

D. Incorrect. Plausible. The first part is incorrect. See the first part of choice B above.

The second part is incorrect. See the second part of choice C above.

SRO JUSTIFICATION (10CFR43(b))

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

-Can the question be answered solely by knowing systems knowledge, i.e.,

how the system works, flowpath, logic, component location? No, CSFSTs are symptom based, not system based.

-Can the question be answered solely by knowing immediate operator actions? No, stopping the ECCS pump and transitioning to FRPs is not governed by IOA's.

-Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No, the decisions made in the question require specific step knowledge.

-Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No, the decision to transition is based on specific parameters and direction.

-Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures. Yes, the question requires the SRO to make a transition decision to an FRP after stopping all ECCS pumps. The decision requires application of EOP rules of usage, knowledge of specific limitations on FRP implementation, and a high level knowledge of ECA and FRP mitigations strategy and interrelationships.

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures Thursday, February 27, 2014 9:13:31 AM 3

Level: SRO Tier # / Group # T1 / G1 K/A# WE11EA2.02 Importance Rating: 3.4 / 3.9 Technical

Reference:

19111-C REV 33.2, pages 6, 11, & 21 References provided: None Learning Objective: LO-LP-37114-12 State the intent of EOP 19111, Loss of Emergency Coolant Recirculation.

LO-PP-37117-04 Describe the differences between the actions for 19113-C and 19111-C and the reason for the differences.

LO-TA-37020 Respond to a Loss of Emergency Coolant Recirculation Capability per 19111-C Question origin: BANK - HL15 Question # WE11EG2.4.2 Cognitive Level: M/F 10 CFR Part 55 Content: 43.5 / 45.13 Comments: Question appears to match the KA. Transitioning to the FRG's on an Orange path is not SRO-only knowledge.

Knowledge of what to do when RWST reaches 8% may be SRO-only knowledge. Since the 8% RWST does not govern a procedure transition, please make sure this is required knowledge of the operators (i.e., not minutia).

- JAT 12/19/2013 (SAT)

You have completed the test!

Thursday, February 27, 2014 9:13:31 AM 4

Approved By Procedure Version C. S. Waldrup Vogtle Electric Generating Plant 19111-C 33.2 Effective Date Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013 RECIRCULATION 6 of 49 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTIONS If offsite power is lost after SI reset, action is required to restart the following ESF equipment if plant conditions require their operation:

RHR Pumps SI Pumps Post-LOCA Cavity Purge Unit Containment Coolers in low speed (Started in high speed on a UV signal).

ESF Chilled Water Pumps (IF CRI is reset).

4

4. Reset SI if necessary. 4. IF SI will NOT reset, THEN initiate ATTACHMENT E.

5

5. Check Containment Cooling Units - 5. Start Cooling Units in low speed.

RUNNING IN LOW SPEED.

6

• 6. Check RWST level - GREATER *6. Go to Step 33.

THAN 8%.

7

7. Determine Containment Spray 7.

requirements:

7.a

a. Check CS Pump suction - FROM a. IF CS Pump suction from RWST: Sump, THEN go to Step 9.

HV-9017A - CNMT SPRAY PMP-A RWST SUCT ISO VLV - OPEN HV-9017B - CNMT SPRAY PMP-B RWST SUCT ISO VLV - OPEN Step 7 continued on next page Printed November 25, 2013 at 09:05 27

Approved By Procedure Version C. S. Waldrup Vogtle Electric Generating Plant 19111-C 33.2 Effective Date Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013 RECIRCULATION 11 of 49 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 14

14. Check if ECCS is in service: 14. Go to Step 24.

CCPs - ANY RUNNING.

-OR-BIT NOT ISOLATED.

-OR-RHR Pumps - ANY RUNNING IN INJECTION MODE.

15

15. Establish one train of ECCS flow: 15.

15.a

a. CCP - ONLY ONE RUNNING. a. Start or stop a CCP to establish only one Pump running.

15.b

b. SI Pump - ONLY ONE b. Start or stop an SI Pump to RUNNING. establish only one Pump running.

15.c

c. RCS pressure - LESS THAN c. Stop RHR Pumps.

300 PSIG.

Go to Step 16.

15.d

d. RHR Pump - ONLY ONE d. Start or stop an RHR Pump RUNNING. to establish only one Pump running.

S Printed November 25, 2013 at 09:05 27

Approved By Procedure Version C. S. Waldrup Vogtle Electric Generating Plant 19111-C 33.2 Effective Date Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT 05/01/2013 RECIRCULATION 21 of 49 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 30

• 30. Check if RCPs must be stopped: 30.

30.a

a. Check the following: a. IF neither condition satisfied, THEN go to Step 31.

Seal number 1 differential pressure - LESS THAN 200 PSID.

-OR-Seal number 1 leakoff flow -

LESS THAN 0.2 GPM.

30.b

b. Stop affected RCPs. b.

30.c

c. Close Spray Valve for idle RCP: c.

RCP 1: PIC-0455C RCP 4: PIC-0455B 31

31. Check RCS WR Hot Leg 31. Go to Step 45.

temperature - GREATER THAN 200°F.

32

32. Check RWST level - LESS THAN 32. Return to Step 2.

8%.

33

33. Stop Pumps taking suction from 33.

RWST and place switches in PULL-TO-LOCK positions:

RHR Pumps SI Pumps CCPs CS Pumps S

Printed November 25, 2013 at 09:05 27