Text

Initial Exam 2015-301 Draft SRO Written Examination
	ML15119A478
Person / Time
Site:	Turkey Point
Issue date:	04/29/2015
From:	; Division of Reactor Safety II
To:	; Florida Power & Light Co
References
	Download: ML15119A478 (162)
	v • d • e

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 013 2.2.25 Importance Rating 4.2 Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

Proposed Question: SRO Question # 76 Which ONE of the following completes the following statement?

In accordance with WCAP-17070-P methodology for determining analytical limits in 0-ADM-536, Technical Specification Bases Control Program, Engineered Safety Features Actuation System Instrumentation Setpoints_____________

A. improve the overall reliability of RPS B. provide Steam Generator overfill protection C. ensure the plant design safety analysis is not exceeded D. guarantee RCS pressure is less than 120% design pressure Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible since this is the reason for the instrumentation setpoints in the Reactor Trip System (i.e., RPS), which assist the ESFAS in mitigating the consequences of accidents.

B. Incorrect. Plausible if examinee recognizes that feedwater isolation, which is an ESFAS function during a LOCA or steamline-break accident, serves (among other things) to prevent SG overfill; however, per 0-ADM-536, SG overfill protection is not itself part of the ESFAS.

C. Correct. Per 0-ADM-536, the instrumentation setpoints in the ESFAS ensure that the operation of plant systems is consistent with the assumptions used in the safety PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION analysis.

0. Incorrect. Plausible since this is a Technical Specification safety limit, which is not specifically addressed or mitigated by the ESFAS instrumentation setpoints.

Technical Reference(s): O-ADM-536, Section 3/4.3.2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: LP 6900163 Obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

11 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 20 of 192 PROCEDURE NO.:

O-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 4 of 176) 2 1 2 Reactor Coolant System Pressure The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The RCS piping, valves, and fittings are designed to ANSI B31 .1, which permits a maximum transient pressure of 120% of design pressure of 2485 psig.

The Safety Limit of 2735 psig is therefore more conservative than the ANSI B31 .1 design criteria and consistent with associated ASME Code requirements.

The entire RCS is hydro tested at 125% (3107 psig) of design pressure to demonstrate integrity prior to initial operation.

REVISION NO.: PROCEDURE TITLE: PAGE:

11 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 21 of 192 PROCEDURE NO.:

O-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 5 of 176) 2.2 Limiting Safety System Settings 2.2.1 Reactor Trip System Instrumentation Setpoints The Trip Setpoints or Nominal Trip Setpoints (NTS) specified in Table 2.2-1 are the nominal values at which the Reactor Trips are set for functional units 2a, 5, 6, 10, 11, and 12. ,The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the Nominal Trip Setpoint when the as measured setpoint is within the band allowed for calibration accuracy.

To accommodate instrument drift that may occur between operational tests and the accuracy to which setpoints can be measured and calibrated, allowances are provided for in the Nominal Trip Setpoint and Allowable Values in accordance with the setpoint methodology described in WCAPs 17070-P (for functional units 2a, 5, 6, 10, 11, and 12) and 12745. Surveillance criteria have been determined and are controlled in Plant procedures and in design documents. The surveillance criteria ensure that instruments which are NOT operating within the assumptions of the setpoint calculations are identified. An instrument channel is considered OPERABLE when the surveillance is within the Allowable Value and the channel is capable of being calibrated in accordance with Plant procedures to within the As-Left tolerance. If the As-Found setpoint is outside of the Plant procedure As-Found tolerance, the occurrence will be entered into the plant Corrective Action program and the channel will be evaluated to verify that it is functioning as required before returning the channel to service.

Sensor and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

The inability to demonstrate through measurement and/or analytical means, using the methods described in WCAPs 17070-P and 12745, that the Reactor Trip function would have occurred within the values specified in the design documentation provides a threshold value for REPORTABLE EVENTS.

REVISION NO.: PROCEDURE TITLE: PAGE:

11 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 66 of 192 PROCEDURE NO.:

O-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 50 of 176) 3/4.3 Instrumentation 3/4.3.1 3/4.3.2 Reactor Trip System andiineered Safety Features Actuation System Instrumentation The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) The associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint (2) The specified coincidence logic is maintained, (3) Sufficient redundancy is maintained to permit a channel to be out of-service for testing or maintenance (due to plant specific design, pulling fuses and using jumpers may be used to place channels in trip), and (4) Sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

Surveillances for the analog RPS/ESFAS Protection and Control rack instrumentation have been extended to quarterly in accordance with WCAP-1 0271, Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System, and supplements to that report as generically approved by the NRC and documented in their SERs (Letters to the Westinghouse Owners Group from the NRC dated February 21, 1985, February 22, 1989, and April 30, 1990).

Under some pressure and temperature conditions, certain surveillances for Safety Injection cannot be performed because of the system design.

Allowance to change modes is provided under these conditions as long as the surveillances are completed within specified time requirements.

REVISION NO.: PROCEDURE TITLE: PAGE:

11 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 70 of 192 PROCEDURE NO.:

0-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 54 of 176) 3/4.3.1 & 3/4.3.2 (Continued)

The NTS for Functional Units if, 4d, 5c, and 6b is the value at which the bistables are set and is the expected value to be achieved during calibratio The NTS value is the LSSS and ensures the safety analysis limits are met for the surveillance interval selected when a channel is adjusted based on stated channel uncertainties. Any bistable is considered to be properly adjusted when the As-Left NTS value is within the As-Left tolerance for CHANNEL CALIBRATION uncertainty allowance (i.e., +/-rack calibration). The NTS value is therefore considered a nominal value (i.e., expressed as a value without inequalities) for the purposes of the ANALOG/DIGITAL CHANNEL OPERATIONAL TEST (COT) and CHANNEL CALIBRATION. These functional Units have been modified by two notes as identified in Table 4.3-23. Note (a) requires evaluation of channel performance for the condition where the As-Found setting for the nominal trip setpoint is outside of the As-Found criterion. As stated above, these instances will be entered into the plant corrective action process and the channel will be evaluated to verify that it is functioning as required before returning the channel to service. Note (b) requires that the channel As-Left setting must be within the As-Left tolerance band. As noted above a channel is considered to be properly calibrated with the As-Left NTS value is within the As-Left tolerance band for CHANNEL CALIBRATION uncertainty allowance (i.e., +/- rack calibration accuracy).

WCAP-1 7070-P methodology for determining analytical limits only applies to those ESFAS functions that support the safety analysis. Certain ESFAS functions such as the loss of voltage UV signal to the 480V load centers only provide equipment protection and therefore their analytical limit is NOT required to meet the WCAP-1 7070-P methodology.

REVISION NO.: PROCEDURE TITLE: PAGE:

11 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 71 of 192 PROCEDURE NO.:

O-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 55 of 176) 3/4.3.1 & 3/4.3.2 (Continued)

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or NOT predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break 0!,

loss-of-coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) Feed water isolation, (4) Startup of the emergency diesel generators, (5) Containment spray pumps start and automatic valves position (6) Containment ventilation isolation, (7) Steam line isolation, (8) Turbine trip, (9) Auxiliary feedwater pumps start and automatic valves position, (10) Containment cooling fans start and automatic valves position, (11) Intake cooling water and component cooling water pumps start and automatic valves position, and (12) Control Room Isolation and Ventilation Systems start. This system also provides a feedwater system isolation to prevent SG overfill. Steam Generator overfill protection is NOT part of the Engineered Safety Features Actuation System (ESFAS) and is added to the Technical Specifications only in accordance with NRC Generic Letter 89-19.

The Steam Generator Pressure-Low and Steam Line Pressure- Low functions, items 1 .f and 4.d of Table 3.3-3, are anticipatory in nature and have a typical lead/lag ratio of 50/5. The 50/5-second lead/lag function is needed to assure acceptable results for the Hot Full Power and Hot Zero Power steam line break analyses.

SRO Question # 76 Clarification Guidance for SRO-only Questions Rev 1(03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TSITRM Action? RO question Can question be answered solely by knowing the Yes LCOITRM information listed above-the-line? RO question Can question be answered solely by knowing the Yes TS Safety Limits? RO 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)

SRO-only Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 039 2.4.45 Importance Rating 4.3 Emergency Procedures I Plan: Ability to prioritize and interpret the significance of each annunciator or alarm.

Proposed Question: SRO Question # 77 Given the following conditions:

Unit 4 is at 80% power.

The following annunciators alarm in a short period of time:

A616, SEAL WATER INJ FILTER HI tP C5/1, SG A STEAM> FEED D3/5, RDTA HI LEVEL D5/6, LEFM TROUBLE Which ONE of the following describes the first priority directed by the Unit Supervisor?

A. Place Standby Seal Water Injection Filter in service.

B. Manually adjust FCV-4-478, 4A FW Control Valve, to control S/G level.

C. Dispatch an operator to bypass and isolate the failed RDT control valve, CV-4-1505.

D. Bypass LEFM and select VENTURI on DCS calorimetric program.

Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible since this is an action in the ARP for A616, but it is not the initial action based on annunciator priority (in this case, priority 3 [second-lowest priority]).

B. Correct. This is the prompt action in the ARP for C5/1, which is a priority-2 annunciator (i.e., of the second-highest priority). At 80% power, the steam-feed mismatch is of a higher concern that the LEFM trouble (also a priority-2 concern); hence, it must be addressed first.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. Incorrect. Plausible since this is an action in the ARP for D315, but it is not the initial action based on annunciator priority (in this case, priority 4 [lowest priority]).

D. Incorrect. Plausible since this is an action in the ARP for D516 and it is (like C5/1) a priority-2 annunciator (i.e., of the second-highest priority). However, at 80% power the LEFM trouble is of less concern than maintaining SG level; hence, it is not the initial action.

4-ARP-097.CR.A, 4-ARP-Technical Reference(s): 097.CR.C, 4-ARP-097.CR.D, 0- (Attach if not previously provided)

ADM-219 Proposed References to be provided to applicants during examination: No Learning Objective: PTN 6900041, Obj. 4 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2012 Point Beach Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

3A 28 CONTROL ROOM RESPONSE - PANEL C PROCEDURE NO.: WINDOW:

4-ARP-097.CR.C TURKEY POINT UNIT 4 5/1 (Page 1 of 1)

CAUSES: 1. Steam Generator Level Control Malfunction

2. Instrument Failure C511

3. Feedwater or steam line break SGA STEAM > FEED DEVICE: SETPOINT: LOCATION:

FC-478A 1 out of 2 steam flow 0.5 x 106 lbs/hr greater than N/A

FC-478B feed flow PROMPT ACTIONS

1. IF malfinctioning SG level controls, THEN:

TAKE manual control of level.

RETURN SG levels to normal.

ALARM CONFIRMATION

1. CHECK FI-4-474, A STM GEN STM FLOW indicating 0.5 x 106 lbs/hr greater than FI-4-477, A STM GEN FEED FLOW on VPA.

2. CHECK FI-4-475, A STM GEN STM FLOW indicating 0.5 x 106 lbs/hr greater than F 1-4-476, A STM GEN FEED FLOW on VPA.

3. CHECK one out of two bistables FC478B1, FC478A2, SG A STM-FW FLO DEV status lights LIT on VPB.

4. CHECK recorder FR-4-478, A STEAM GENERATOR on console.

OPERATOR ACTIONS

1. IF condition is NOT due to faulty indication, THEN INVESTIGATE for feedwater or steam line breaks.

2. IF alarm is due to instrument failure, THEN REFER TO 4-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels.

REFERENCES:

1. FPL Control System Diagram 5610-T-D-17, 18B

2. Tech Spec Section 3/4.3.1

REVISION NO.: PROCEDURE TITLE: PAGE:

lB ANNUNCIATOR RESPONSE PROCEDURE USAGE 4 of 14 PROCEDURE NO.:

O-ADM-219 TURKEY POINT PLANT 1.0 PURPOSE Provide rules of usage for Annunciator Response Procedures (ARPs).

2.0 TERMS AND DEFINITIONS

1. Annunciator Response Procedures (ARP5) Plant procedures that specify the operator actions required to mitigate the consequences of transients that cause plant parameters to exceed alarm setpoints.

2. Local (Locally) An action performed by an operator outside the Control Room.

3. Immediate/Immediately The definition of immediate should be based on the specific situation. In all cases the action should be performed in a timely manner. In some cases this may require action in less than 15 minutes and in others, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months may be appropriate.

(This definition is not for Immediate Operator Actions).

4. ARP Prompt Actions Any actions specifically listed in the ARP as Prompt Actions. These actions may be performed from memory with SRO concurrence.

5. Manual (Manually) An action performed by the operator in the Control Room. This does NOT include automatic actions which take place without operator intervention.

6. Priority 1 (Red): Nuclear Safety These alarms require immediate response and reflect a potential challenge to plant safety and require protective systems to activate. These alarms include: SI and Reactor/Turbine/Gen Trips.

7. Priority 2 (Yellow): Power Production Availability These alarms require immediate response and reflect a challenge to plant equipment or systems that may affect continued plant availability or timely recovery. Immediate response to these alarms would be deferred only if action were required by a Priority 1 alarm. Failure to properly respond to a Priority 2 condition may lead to or contribute to a higher level condition.

8. Priority 3 (Blue): Investment Protection These alarms require prompt response and provide information that, if unattended, may result in a threat to higher level actions. Prompt action to this level of alarm may reduce the consequences of the problem by minimizing equipment damage or material waste.

REVISION NO.: PROCEDURE TITLE: PAGE:

lB ANNUNCIATOR RESPONSE PROCEDURE USAGE 5 Of 14 PROCEDURE NO.:

O-ADM-219 TURKEY POINT PLANT 2.0 TERMS AND DEFINITIONS (continued)

9. Priority 4 (White): Status/Information These alarms require non-priority response and reflect equipment status, transitions or conditions to be corrected, but do NOT threaten the unit availability.

Because they are NOT strictly Information Only items, they may warrant operator action. Priority 4 items are deferred in the face of higher priority items.

10. Transition A change from one place to another in the procedures, either from one step to another step or from one procedure to another procedure.

3.0 RESPONSIBILITIES

1. Shift Manager (SM) The SM shall provide technical guidance for event mitigation WHEN ARPs are in effect.

The SM is required to be fully aware/cognizant of all Annunciators at all times whether they have cleared or are locked in.

2. Unit Supervisor (US) The US should direct the detailed event mitigation strategy for the affected unit unless otherwise directed by the Shift Manager when ARPs are in effect.

The US is required to be fully aware/cognizant of all Annunciators at all times whether they have cleared or are locked in.

3. Field Supervisor The Field Supervisor should direct non-licensed operators, if necessary, to determine the cause of the alarm condition and the performance of corrective actions when ARPs are in effect.

4. Affected Unit Reactor Operator (RO) The affected unit RO is responsible for the following when ARPs are in effect:

A. Respond to alarms based on color code priority and plant conditions and is required to take immediate corrective actions.

B. Reading the ARP in effect and performing the event mitigation strategy for alarms received in the Control Room.

C. Transition to the appropriate procedures if required by the ARP.

REVISION NO.: PROCEDURE TITLE: PAGE:

lB ANNUNCIATOR RESPONSE PROCEDURE USAGE 7 of 14 PROCEDURE NO.:

O-ADM-219 TURKEY POINT PLANT 4.0 INSTRUCTIONS Procedure Entry A. Entry into the ARP begins by locating the panel and window with the alarm condition.

B. Control Room annunciator panels are designated by letters A through J and X. The applicable annunciator window is located numerically by column and row numbers. Therefore, an alarm on Panel A Column 1 Row 5 would be designated as A 1/5 in the ARP.

C. Once the alarm panel and window are identified, then (1) Acknowledge the alarm (2) Enter the appropriate ARP for response guidelines.

2. Step Performance A. Steps should be performed by referring to the written procedure directly.

B. The arnunciator panel procedures indicate appropriate operator action for Control Room panel annunciators. The actions listed are intended to be a guide for operators in responding to single annunciators and not intended to be a substitute for good judgment based on thorough understanding of plant conditions and equipment.

C. Many off normal plant conditions will result in several annunciators lighting almost simultaneously. In such a case, operators are expected to respond to the root cause of the problem and maintain the unit in a safe condition using applicable off normal and emergency procedures. This action may not necessarily correspond to that of the Annunciator Procedures.

D. The procedure should be performed in a step-by-step manner with each step being completed prior to the performance of the next step.

REVISION NO.: PROCEDURE TITLE: PAGE:

12 43 CONTROL ROOM RESPONSE - PANEL A PROCEDURE NO.:

WINDOW:

4-ARP-097.CR.A TURKEY POINT UNIT 4 6/6 (Page 1 of 2)

CAUSES: 1. Hi seal injection flow rate

2. Dirty in service seal water injection filter A616 SEAL WATER INJ FILTER HI AP DEVICE: SETPOINT: LOCATION:

PIC-4-157 20 psid N/A ALARM CONFIRMATION

1. CHECK the following locally:

DPI-4-1 54B for Seal Water Injection Filter 4A AP.

DPI-4-154A for Seal Water Injection Filter 4B AR OPERATOR ACTIONS CAUTION

RCPs are limited to a maximum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months operation without seal injection.

To prevent RCP seal damage, do NOT bypass seal injection filters.

1. IF seal water injection flow is above 13 gpm, THEN:

CAUTION Care must be exercised when throttling HCV-4-121 in the closed direction.

Throttling this valve completely closed can cause the Charging Pump discharge relief valve to lift resulting in a possible loss of charging if the relief valve fails to reseat.

A. ADJUST seal injection flow as necessary to establish 6 to 13 gpm per RCP using the applicable procedure(s):

4-NOP-041.O1A, 4A Reactor Coolant Pump Operations

4-NOP-041 .01 B, 4B Reactor Coolant Pump Operations

4-NOP-041.O1C, 4C Reactor Coolant Pump Operations B. REDUCE total charging pump speed.

2. IF fiIterP is high, THEN:

A. IF standby filter is available, THEN:

(1) PLACE standby filter in service using 4-OP-047, CVCS Charging and Letdown.

(2) ADJUST seal injection flow as necessary to establish 6 to 13 gpm per RCP using the applicable procedure(s):

4-NOP-041.O1A, 4A Reactor Coolant Pump Operations

4-NOP-041 .01 B, 4B Reactor Coolant Pump Operations

4-NOP-041.O1C, 4C Reactor Coolant Pump Operations

REVISION NO.: PROCEDURE TITLE: PAGE:

8A 20 CONTROL ROOM RESPONSE - PANEL D PROCEDURE NO.: WINDOW:

4-ARP-097.CR.D TURKEY POINT UNIT 4 (Page_1 of_1)

CAUSES: Malfunction of CV-4-1 515 D315 RDTA HI LEVEL DEVICE: SETPOINT: LOCATION:

LIT-4-1505A/B/C via 29.9 above bottom of tank N/A DCS Pt. L1505_V ALARM CONFIRMATION CHECK DCS indications.

OPERATOR ACTIONS

1. ENSURE alternate drain to condenser CV-4-1515 automatically OPENS.

2. DISPATCH operator to check Fl-4-5120 to confirm flow from normal drain to 6A FW HTR CV-4-1 505 and alternate drain to condenser CV-4-1515.

3. EVALUATE possible causes AND NOTIFY System Engineer.

4. IF high level alarm is a result of a secondary swing AND third Condensate Pump is available, THEN CONSIDER starting third condensate pump to minimize impact on SGFP suction pressure.

5 IF secondary swing is determined to be caused by a failed drain control val(CV)

THEN CONSIDER bypassing and isolating affected CV.

REFERENCES:

1. FPL Drwg 5614-M-3074

2. FPL Drwg 5610-E-26, Sh 39

3. FPL Drwg 5610-M-3

4. FPL Drwg 5614-M-3081, Sh 2

5. FPL Drwg 5614-J-FD-4RHDTA

6. EC 246909, MSR Replacement

REVISION NO.: PROCEDURE TITLE: PAGE:

8A 35 CONTROL ROOM RESPONSE - PANEL D PROCEDURE NO.:

WINDOW:

4-ARP-097.CR.D TURKEY POINT UNIT 4 5/6 (Page 1 of 1)

CAUSES: LEFM System status is Level 2 or 3 D516 LEFM TROUBLE DEVICE: SETPOINT: LOCATION:

DCS Various N/A NOTE Operation of Annunciators D 1/5, D 1/6, D 2/5, D 2/6, D 4/5, D 5/6, D 9/5, and D 9/6 is dependant upon a single DCS Field Bus module (FBM) 41 3Q25 in RPI Panel 4QR64.

All of these annunciators in alarm at the same time may indicate a failure of the FBM.

ALARM CONFIRMATION CHECK DCS monitor to confirm LEFM System status.

OPERATOR ACTIONS

1. IF LESS THAN OR EQUAL TO 25% Reactor Power, THEN PERFORM the following:

A. CONTACT Maintenance or System Engineer to determine source of trouble.

B. ENSURE corrective action is initiated for LEFM degraded status.

C. ENSURE VENTURI is selected as DATA SOURCE for Calorimetric Input.

D. WHEN LEFM corrective actions are complete and alarm is CLEAR, THEN:

(1) RESET CALORIMETRIC CORRECTION FACTOR.

(2) SELECT LEFM as DATA SOURCE.

2. IF GREATER THAN 25% Rector Power THEN GO TO 4-ONOP-074.1, Leading Edge FIow Monitor (LEFM) Trouble.

REFERENCES:

1. EC 242439

2. 4-ONOP-074.1, Leading Edge Flow Meter (LEFM) Trouble

[RO Question # 77 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1I II i RO question immediate operator actions? Yes j

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the questio n be answered solely by knowing the purpose, overall mitigative strategy of a procedur 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 063 A2.02 Importance Rating 3.1 Ability to (a) predict the impacts of the following malfunctions or operations on the dc electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of ventilation during battery charging Proposed Question: SRO Question # 78 Given the following conditions:

Unit 3 is at 100% power.

An equalizing charge is in progress on the 3B battery in accordance with 0-PME-003.16.

The Common HVAC Unit (E16D) and the North DC Equipment!Inverter Room HVAC Unit (E16E) in the DC Equipment and Inverter Rooms have failed.

DC Equipment and Inverter Rooms temperatures are at 100°F.

0-ONOP-025.3 is in progress.

Which ONE of the following completes the statements below?

The Vital Batteries 3A, 3B, 4A, and 4B and Spare Battery D52 are (1) . The Shift Manager (2) required to invoke 10CFR5O.54(x) and (y).

NOTE

0-PME-003.16, Individual Cell Equalizing Charge for Vital Batteries 3A, 3B, 4A, and 4B and Spare Battery D52

0-ONOP-025.3, DC Equipment and Inverter Rooms Supplemental Cooling A. (1)OPERABLE (2) is NOT B. (1) OPERABLE (2) is PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE NFORMATlON C. (1)NOT OPERABLE (2) is NOT D. (1) NOT OPERABLE (2)is Proposed Answer: A Explanation (Optional):

A. Correct. Per 0-ONOP-025.3, an ambient temperature of 115°F in a Battery Room is an acceptable operability limit for the batteries; hence, at 102°F, the batteries are considered operable. Technical Specifications (i.e., LCO 3.8.2) identify no limits based on battery/battery-room temperatures. As the batteries are operable, there is no requirement to invoke 1 OCFR5O.54(x) and (y).

B. Incorrect. The first part of the distractor is correct. Plausible if examinee believes that elevated battery temperature (i.e., 102°F) can lead to a declaration of battery inoperability, which would require IOCFR5O.54(x) and (y)to be invoked for continued plant operation.

C. Incorrect. Per 0-ONOP-025.3, an ambient temperature of 115°F in a Battery Room is an acceptable operability limit for the batteries; hence, at 102°F, the batteries are not considered inoperable due to excessive temperature. Technical Specifications (i.e.,

LCO 3.8.2) identify no limits based on battery/battery-room temperatures. The second part of the distractor is correct.

D. Incorrect. Per 0-ONOP-025.3, an ambient temperature of 115°F in a Battery Room is an acceptable operability limit for the batteries; hence, at 102°F, the batteries are not considered inoperable due to excessive temperature. Technical Specifications (i.e.,

LCO 3.8.2) identify no limits based on battery/battery-room temperatures. Plausible if examinee believes that elevated battery temperature (i.e., 102°F) can lead to a declaration of battery inoperability, which would require 10CFR5O.54(x) and (y)to be invoked for continued plant operation.

Technical Reference(s): 0-ONOP-025.3 Technical (Attach if not previously provided)

Specifications Proposed References to be provided to applicants during examination: N Learning Objective: PTN 6900139, Obj. 12 (As available)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

Procedure No Procedure TtIe: lage:

5 DC Equipment and Inverter Rooms Approval Date O-ONOP-025.3 Supplemental Cooling 4/22/11 5.0 SUBSEQUENT ACTIONS CAUTIONS

Battery room ambient temperature of I I F is an acceptable operability limit for the batteries.

Time intervals for actions are critical to limit room heatup and preclude exceeding fan operation temperature limitations.

Different scenarios included in attachments require RCA boundary doors, security doors, and fire doors associated with Halon Suppression System to be maintained open. Prior to implementing any supplemental cooling with any of these doors open, provisions are required to be made to establish personnel on a continuous basis to close the door. A security watch and Radiation Protection and Security control points also may be required.

5.1 IF units are in normal operation, THEN perform the following:

5.1.1 Obtain permission from the Shift Manager to initiate this procedure for plant continued operation with loss of HVAC to DC Equipment and Inverter Rooms.

NOTES I

Setting the thermostat for the North DC Equipment/lnverter Room HVAC (EI6E) at I less than 78°F will allow the HVAC unit to ice up during cool weather. I

The North (EI6E) DC Equipment/lnverter Room HVAC provides cooling to the North I Inverter Rooms, the Unit 3 MG Set Room, and the 38 Vital Battery Room. I I . The South (EI6F) DC Equipment/lnverter Room HVAC provides cooling to the South I I Inverter Rooms and the Unit 4 MG Set Room.

I

The common split HVAC unit (EI6D) provides cooling to the 3A, 4A, and 48 Vital I I Battery Rooms and supplements the cooling from the other units to the North and I South lnverter Rooms and both MG Set Rooms.

a a 5.1.2 Verify thermostat for the DC equipment and Inverter Rooms are set as follows:

HVAC Unit EI6D TIS-1071 75°F

HVAC Unit E16E TIS-6419 78°F

HVAC Unit EI6F TIS-6420 75°F

3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The following D.C. electrical sources shall be OPERABLE:*#

a. 125-volt D.C. Battery Bank 3A or spare battery bank D-52 and associated full capacity charger(s)

1) 3A1 powered by motor control center (MCC) 3C with EDG 3A OPERABLE, or

2) 3A2 powered by MCC 4D with EDG 4A and 4B OPERABLE, or

3) 3A1 powered by MCC 3C with EDG 3A OPERABLE and 3A2 powered by MCC 4D with EDG 4A and 4B OPERABLE,

b. 125-volt D.C. Battery Bank 3B or spare battery bank D-52 and associated full capacity charger(s)

1) 3B1 powered by MCC 3B with EDG 3B OPERABLE, or

2) 3B2 powered by MCC 4D with EDG 4A and 4B OPERABLE, or

3) 3B1 powered by MCC 3B with EDG 3B OPERABLE and 3B2 powered by MCC 4D with EDG 4A and 4B OPERABLE,

c. 125-volt D.C. Battery Bank 4A or spare battery bank D-52 and associated full capacity charger(s)

1) 4A1 powered by MCC 4C with EDG 4A OPERABLE, or

2) 4A2 powered by MCC 3D with EDG 3A and 3B OPERABLE, or

3) 4A1 powered by MCC 4C with EDG 4A OPERABLE and 4A2 powered by MCC 3D with EDG 3A and 3B OPERABLE,

d. 125-volt D.C. Battery Bank 4B or spare battery bank D-52 and associated full capacity charger(s)

1) 4B1 powered by MCC 4B with EDG 4B OPERABLE, or

2) 4B2 powered by MCC 3D with EDG 3A and 3B OPERABLE, or

3) 4B1 powered by MCC 4B with EDG 4B OPERABLE and 4B2 powered by MCC 3D with EDG 3A and 3B OPERABLE.

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

ACTION:

a. With one or more of the required battery chargers OPERABLE but not capable of being powered from its associated OPERABLE diesel generator(s), restore the capability within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This ACTION applies to both units simultaneously.

All battery chargers required to satisfy the LCO shall be powered from separate MCCs.

Inoperability of the required EDGs specified in the LCO requirements below does not constitute inoperability of the associated battery chargers or battery banks.

TURKEY POINT UNITS 3 & 4 3/4 8-13 AMENDMENT NOS. 138 AND 133

D.C. SOURCES LIMITING CONDITION FOR OPERATION ACTION: (Continued)

b. With one of the required battery banks inoperable, or with none of the full-capacity chargers associated with a battery bank OPERABLE, restore all battery banks to OPERABLE status and at least one charger associated with each battery bank to OPERABLE status within two hours* or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This ACTION applies to both units simultaneously.

SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and its associated full capacity charger(s) shall be demonstrated OPERABLE

a. At least once per 7 days by verifying that:

1) The parameters in Table 4.8-2 meet the Category A limits, and

2) The total battery terminal voltage is greater than or equal to 129 volts on float charge and the battery charger(s) output voltage is 129 volts, and

3) If two battery chargers are connected to the battery bank, verify each battery charger is supplying a minimum of 10 amperes, or demonstrate that the battery charger supplying less than 10 amperes will accept and supply the D.C. bus load independent of its associated battery charger.

b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 105 volts (108.6 volts for spare battery D-52), or battery overcharge with battery terminal voltage above 143 volts, by verifying that:

1) The parameters in Table 4.8-2 meet the Category B limits,

2) The average electrolyte temperature of every sixth cell is above 60°F, and

3) There is no visible corrosion at either terminals or connectors, or verify battery connection resistance is:

Battery Connection Limit (Micro-Ohms) 3B, 4A inter-cell I termination < 29 inter-cell (brace locations) < 30 transition cables < 125 or total battery connections 1958 Battery Connection Limit (Micro-Ohms) 3A, 4B, D-52 inter-cell / termination < 35 inter-cell (brace locations) < 40 transition cables < 125 or total battery connections < 2463

c. At least once per 18 months by verifying that:

1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,

Can be extended to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the oppsite unit is in MODE 5 or 6 and each of the remaining required battery chargers is capable of being powered from its associated diesel generator(s).

TURKEY POINT UNITS 3 & 4 3/4 8-14 AMENDMENT NOS. 252 AND 248

D.C. SOURCES SURVEILLANCE REQUIREMENTS (Continued)

2) The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material,

3) Each 400 amp battery charger (associated with Battery Banks 3A and 4B) will supply at least 400 amperes at 129 volts for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and each 300 amp battery charger (associated with Battery Banks 3B and 4A) will supply at least 300 amperes at 129 volts for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

4) Battery Connection resistance is:

Battery Connection Limit (Micro-Ohms) 3B, 4A inter-cell I termination < 29 inter-cell (brace locations) < 30 transition cables < 125 or total battery connections < 1958 Battery Connection Limit (Micro-Ohms) 3A, 4B, D-52 inter-cell I termination < 35 inter-cell (brace locations) < 40 transition cables < 125 or total battery connections < 2463

d. At least once per 18 months, during shutdown**, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test.

e. At least once per 12 months, during shutdown**, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% [75% for Batteries 4B and D52 (Spare) when used in place of Battery 4Bj of service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% [7% for Batteries 4B and D52 (Spare) when used in place of Battery 4B] of rated capacity from its average on previous performance tests, or is below 90% [93% for Batteries 4B and D52 (Spare) when used in place of Battery 4B} of the manufacturers rating.

f. At least once per 60 months, during shutdown**, by verifying that the battery capacity is at least 80% [87% for Batteries 4B and D52 (Spare) when used in place of Battery 4B] of the manufacturers rating when subjected to a performance discharge test. Once per 60-month interval this performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1 .d.

- Except that the spare battery bank D-52, and any other battery out of service when spare battery bank D-52 is in service may be tested with simulated loads during operation.

TURKEY POINT UNITS 3 & 4 3/48-15 AMENDMENT NOS. 252 AND 248

TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A)i) CATEGORY 2 B PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE(3)

DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL Electrolyte >Minimum level >Minimum level Above top of Level indication mark, indication mark, plates, and < 1/4 above and < 1/4 above and not maximum level maximum level overflowing indication mark indication mark Float Voltage 2.13 volts 2.13 volts(6) 2.07 volts Not more than 0.020 below the Specific average of all Gravity (4) i.2oo 1.195 connected cells Average of all Average of all Connected cells connected cells

> 1.205 1.195 TABLE NOTATIONS (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months all the Category B measurements are taken and found to be within their allowable values, and provided all Category A and B parameter(s) are restored to within limits within the next 6 days.

(2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that the Category B parameters are within their allowable values and provided the Category B parameter(s) are restored to within limits within 7 days.

(3) Any Category B parameter not within its allowable value indicates an inoperable battery.

(4) Corrected for electrolyte temperature and level.

(5) Or battery charging current is less than 2 amps when on charge.

(6) Corrected for average electrolyte temperature.

TURKEY POINT UNITS 3 & 4 3/48-16 AMENDMENT NOS. 138 AND 133

L Question #78 0

Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the questlo n be answered immediate ope rator actions?

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes I RO question that require direct entry to major EOPs?

Can the questio n be answered solely by knowi the purpose, overall mitigative strategy of a 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 onlJ

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

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group #

K/A# 076 A2.02 Importance Rating 3.1 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 Proposed Question: SRO Question # 79 Given the following conditions:

Unit 3 is operating at 100% power.

The 3A and 3B ICW pumps are running.

3C ICW pump is out of service for shaft replacement.

Annunciator 1-4/4, ICW HEADER A/B LO PRESS, is actuated.

ANPO reports 3B ICW Pump Motor is vibrating excessively.

BOP trips the 3B ICW pump.

ICW flow is at 19,300 GPM.

TPCW supply temperature is 102°F and rising at 1°F/hr.

CCW supply header temperatures are 102°F and rising at 1°F/hr.

Which ONE of the following completes the statement below?

With only one ICW pump in operation, the US directs a (1) , due to (2)

A. (1) load reduction lAW 3-GOP-i 03, Power Operation to Hot Standby (2) reaching the High CCW temperature limit B. (1) Reactor trip and enters 3-EOP-E-0, Reactor Trip or Safety Injection (2) operating one ICW pump with low pressure and high flow C. (1) Reactor trip and enters 3-EOP-E-0, Reactor Trip or Safety Injection.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION (2) reaching the High CCW temperature limit D. (1) load reduction lAW 3-GOP-i 03, Power Operation to Hot Standby (2) operating one lOW pump with low pressure and high flow Proposed Answer: D Explanation (Optional):

A. Incorrect. The first part of the distractor is correct. Plausible if examinee transposes the digits for the CCW supply header temperatures (102°F) and believes that the CCW HX high-temperature setpoint (120°F) has been attained.

B. Incorrect. Plausible if examinee believes that a single ICW Pump is insufficient to maintain adequate CCW System cooling, which places the RCPs in jeopardy and thereby warrants a reactor trip. The second part of the distractor is correct.

C. Incorrect. Plausible if examinee believes that a single lOW Pump is insufficient to maintain adequate CCW System cooling, which places the RCPs in jeopardy and thereby warrants a reactor trip. Plausible if examinee transposes the digits for the CCW supply header temperatures (102°F) and believes that the CCW HX high-temperature setpoint (120°F) has been attained.

D. Correct. Per 3-ONOP-019 and 3-ARP-097.CR.l, if a single ICW Pump is operating and total lOW flow 1 8,500 gpm (i.e., low-pressure/high-flow condition), unit load must be reduced using 3-GOP-i 03 (to limit heat input into the TPCW System).

3-ONOP-019 Rev 2 Technical Reference(s): 3-ARP-097.CR.H (Attach if not previously provided) 3-ARP-097.CR.I Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902277 Obj. 2, 7 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMA11ON Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN LOIT Exam bank Item #: 1.1.25.77.7.2 PTN L-154 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

11 27 CONTROL ROOM RESPONSE - PANEL I PROCEDURE NO.: WINDOW:

3-ARP-097.CR.I TURKEY POINT UNIT 3 4/4 (Page_1 of_1)

CAUSES: 1. Leak in ICW System

2. Trip of a running ICW pump 1414 ICw HEADERAIB LO PRESS DEVICE: SETPOINT: LOCATION:

PS-3-1619 (A HDR) 10 psig N/A

PS-3-1620 (B HDR)

ALARM CONFIRMATION

1. CHECK lOW header pressure indicators, P1-3-1619 or 3-1620 less than or equal to 10 psig on VPA.

2. IF operating a single ICW Pump, THEN CHECK total ICW flow is less than 18,500 gpm.

OPERATOR ACTIONS

1. START standby ICW pump using 3-NOP-Ol 9, Intake Cooling Water System.

2. Locally CHECK ICW piping and heat exchangers for leaks.

3. REFER TO 3-ONOP-019, Intake Cooling Water Malfunction.

4. IF operating a single ICW Pump AND total ICW flow is greater than 18,500 gpm, THEN immediately REDUCE total lOW flow by perl:orming the following:

A. THROTTLE TPCW Combined Outlet Valve, 3-50-401, while maintaining TPCW Hx outlet temperature less than 105°F.

B. THROTTLE 3-50-406, CCW HX lOW OUTLET SPOOL PIECE BYPASS and 3 407, CCW HX ICW OUTLET SPOOL PIECE ISOL while maintaining minimum ICW flows through CCW Hxs as determined by 3-NOP-019, Intake Cooling Water System.

5. IF unable to reduce total ICW flow through a single ICW Pump to less than 18,500 gpm, THEN REDUCE unit load using 3-GOP-i 03, Power Operation to Hot Standby, to limit heat input into TPCW AND THROTTLE TPCW Hx ICW flows using TPCW COMBINED OUTLET VALVE, 3-50-401, until total lOW flow is below 18,500 gpm.

6. IF a single ICW Pump has operated at flows greater than 18,500 gpm, THEN REFER TO 3-NOP-019, Intake Cooling Water System.

REFERENCES:

1. FPL Dwg 5613-M-3019, Sh 1

2. FPL EWD 561 0-E-27, Sh 25, Misc. Alarms

3. PTN-BFSM-98-01 6, Affects of Opening 3/4-50-402 While 3/4-50-401 is Fully Open

4. PC/M 02-018, ICW Header Low Alarm Setpoint Change

REVISION NO.: PROCEDURE TITLE: PAGE:

6 50 CONTROL ROOM RESPONSE - PANEL H PROCEDURE NO.: WINDOW:

3-ARP-097CR.H TURKEY POINT UNIT 3 8/5 (Page 1 of 1)

CAUSES: 1. High RCS to CCW flow in RHR HX

2. Dirty CCW HX tubes H815

3. Dirty ICW/CCW basket strainer

4. Low CCW System flow

5. Clogged CCW HX tube sheets

6. High RCS cooldown rate OUTLET HI TEMP DEVICE: SETPOINT: LOCATION:

TE-3-607A 1 2aF A header downstream of CCW heat exchangers

TE-3-607B B header downstream of CCW heat exchangers ALARM CONFIRMATION

1. CHECK the following:

Flow/temperature of RHR HX on VPB, if in operation

CCW header flows on VPB

ICW discharge pressure on VPA OPERATOR ACTIONS

1. DISPATCH operator to check AP across ICW/CCW basket strainer.

2. IF P greater than 1.0 psid, THEN BACKWASH ICW/CCW basket strainer.

3. CHECK all CCW HX in operation.

4. IF RHR in cooldown alignment, THEN REDUCE rate of RCS cooldown.

5. IF required, THEN START another CCW pump using 3-NOP-030, Component Cooling Water System.

6. IF required, THEN START another ICW pump using 3-NOP-019, Intake Cooling Water System.

7. REFER TO 3-ONOP-030, Component Cooling Water Malfunction.

8. REFER TO 3-ONOP-019, Intake Cooling Water Malfunction.

9. REFER TOTS 3.7.2, 3.7.3, 3.4.1.3, 3.4.1.4.1, and 3.4.1.4.2.

REFERENCES:

1. FPL Dwg 5613-M-3030

2. Tech Spec Sections 3.7.2, 3.7.3, 3.4.1.3, 3.4.1.4.1, and 3.4.1.4.2

STEP ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

4 Verify Intake Cooling Water Pumps - TWO Perform the following:

RUNNING a) Manually start any available Intake Cooling Water Pump to establish TWO RUNNING.

b) IF only one ICW Pump is operating AND total ICW flow is greater than 18,500 GPM, THEN immediately reduce total ICW flow by:

Throttling TPCW HX Outlet Combined ICW Iso Vlv 3-50-401 while maintaining TPCW Heat Exchanger outlet temperature less than 110 degrees.

Throttle 3-50-406, CCW HX Outlet Spool Piece Bypass Valve, and/or 3-50-407, CCW HX Outlet Spool Piece Iso Vlv, while maintaining minimum ICW flows through the CCW Heat Exchangers as determined by 3-NOP-019, INTAKE COOLING WATER SYSTEM.

c) IF unable to reduce total ICW flow through a single ICW Pump to less than 18,500 GPM, THEN reduce Unit Load using 3-GOP-i 03, POWER OPERATION TO HOT STANDBY, to limit heat input into the TPCW system and throttle ICW flow to the TPCW Heat Exchangers using TPCW HX Outlet Combined ICW Iso Vlv 3-50-401 until total ICW flow is less than 18,500 GPM.

d) IF a single ICW Pump has operated at flows greater than 18,500 GPM, THEN refer to 3-NOP-019, INTAKE COOLING WATER SYSTEM.

W97:/JBSIIn/mr/In

SRO Question # 79 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1I immediate operator actions? Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

9 Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 103 A2.03 Importance Rating 3.8 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation Proposed Question: SRO Question # 80 Given the following conditions:

Unit 3 trips on a large break loss of coolant accident.

MOV-3-381, RCP Seal Water Return and Excess Letdown Isolation fails to isolate.

MOV-3-6386, RCP Seal Return Isolation fails to isolate.

MOV-3-626, RCP Thermal Barrier CCW Outlet fails to isolate.

Dose rates at the above valves are very high.

Which ONE of the following choices properly identifies the minimum action required to isolate the Containment penetrations and whose approval is required to exceed 10 CFR 20 exposure limits?

A.

close MOV-3-381 and MOV-3-626.

RP Manager B.

close MOV-3-381 and MOV-3-626.

Emergency Coordinator C.

close MOV-3-6386 and MOV-3-381.

RP Manager D.

close MOV-3-6386 and MOV-3-381.

Emergency Coordinator Proposed Answer: B PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Explanation (Optional):

A. Incorrect. The first part of the distractor is correct. Plausible if examinee believes that the RP Manager has authority for approving radiation exposures; however, this individuals responsibilities are primarily administrative in nature and the Emergency Coordinator, per 0-EPIP-20101 and O-EPIP-201 11, authorizes radiation exposures in excess of regulatory limits.

B. Correct. The minimum action necessary to isolate Containment penetration P-43 is closure of MOV-3-626, whereas the minimum action necessary to isolate Containment penetration P-25 is closure of MOV-3-381 or MOV-3-6386 (which are in series). Per 0-EPIP-20101 and 0-EPIP-20111, the Emergency Coordinator authorizes radiation exposures in excess of regulatory limits.

C. Incorrect. The minimum action necessary to isolate Containment penetration P-25 is closure of MOV-3-381 or MOV-3-6386 (which are in series), but not both. Plausible if examinee believes that the RP Manager has authority for approving radiation exposures; however, this individuals responsibilities are primarily administrative in nature and the Emergency Coordinator, per 0-EPIP-20101 and 0-EPIP-20111, authorizes radiation exposures in excess of regulatory limits.

D. Incorrect. The minimum action necessary to isolate Containment penetration P-25 is closure of MOV-3-381 or MOV-3-6386 (which are in series), but not both. The second part of the distractor is correct.

Technical Reference(s): 0-EPIP-20101,O-EPIP-20111 561 3-M-3047, 561 3-M-3030 Technical Reference(s):

Proposed References to be provided to applicants during examination: N Learning Objective: LP 602353 Obj. 1, 3 (As available)

Question Source: Bank # 87106 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Comanche Peak Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

SRO Question # 80 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing I immediate operator actions? Yes I RO question 9

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

9 Can the question be answered solely by knowing the purpose, overall sequence of events, or J*festion overall mitigative strategy of a 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 L4onl

Knowledge of diagnostic steps and decision points in the FOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K/A# 002 2.2.22 Importance Rating 4.7 Equipment Control: Knowledge of limiting conditions for operations and safety limits.

Proposed Question: SRO Question # 81 Given the following conditions:

Unit 3 is in MODE 2.

A pinhole leak on the piping is discovered upstream of the first off RCS valve, 3-954B, RCS Hot Leg Sample Valve.

3-954B is closed with leakage continuing.

Which ONE of the following describes the correct operator response in accordance with Technical Specifications?

A. No Technical Specification actions are required with MODE 2 operation.

B. Reduce the leakage to less than 1 gpm in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months OR be in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

C. Place the Unit 3 in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND be in MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

D. Isolate the sample line with 2 valves within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months AND no further Technical Specification actions are required.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible if examinee believes that the TS leakage limitations for the RCS are only applicable during power operation (Mode 1).

PTN L-15-i. DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION B. Incorrect. Plausible if examinee believes that 3-954B is included in TS Table 3.4-1; the allowable leakage for such valves is 1 gpm or the plant must be in at least hot standby (Mode 3) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown (Mode 5) within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

C. Correct. Per TS 3.4.6.2, no pressure boundary leakage is allowed and, if it exists, the plant must be in at least hot standby (Mode 3) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown (Mode 5) within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

D. Incorrect. Plausible if examinee believes that 3-954B is isolable and included in TS Table 3.4-1; the leakage LCD action statement for such valves requires the closure of two associated isolation valves within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for continued operation.

Technical Specifications Technical Reference(s): .

(Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

PTN LOIT Exambank Item: 1.1.26.24.3.17 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATING 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary-to-secondary leakage through any one steam generator (SG),

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4-1 up to a maximum of 5 GPM at a Reactor Coolant System pressure of 2235 +/- 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 .

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, or with primary-to-secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding primary-to-secondary leakage, PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than allowed by 3.4.6.2.e above operation may continue provided:

1. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months verify that at least two valves in each high pressure line having a non functional valve are in, and remain in that mode corresponding to the isolated condition, i.e., manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supplies deenergized. Follow applicable ACTION statement for the affected system, and Test pressure less than 2235 psig are allowed. Minimum differential test pressure shall not be less than 150 psid. Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressure differential to the one-half power.

TURKEY POINT UNITS 3 & 4

- 3/4 4-14 AMENDMENT NOS. 260 AND 255

SRO Question # 81 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question Can question be answered solely by knowing the LCO/TRM information listed above-the-line?

Can question be answered solely by knowing 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) SRO-only

Knowledge of TS bases that is required to analyze TS i question required actions and terminology I I No Question might not be linked to I 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K/A# 015 2.1.23 Importance Rating 4.4 Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Proposed Question: SRO Question # 82 Given the following conditions:

Unit 3 is operating at 100% power.

The following annunciators are in alarm:

B213, POWER RANGE LOWER DET HI FLUX DEV!AUTO DEFEAT B316, OTAT B515, OTAT/OPzXT ROD STOP B6!4, POWER RANGE CHANNEL DEVIATION B7!1, NIS/RPI ROD DROP ROD STOP B9!2, AXIAL FLUX TILT

Rod Control is in AUTOMATIC.

No other alarms or automatic control actions occurred.

Which ONE of the following describes (1)the correct procedural guidance and (2) what action should be taken to mitigate the situation?

A. (1) 3-ONOP-028, Reactor Control System Malfunction (2) Direct a power reduction by boration to < 50% RTP since QPTR is greater than a Technical Specification limit.

B. (1) 3-ONOP-059.8, Power Range Instrumentation Malfunction (2) Bypass the affected channel at the NI Rack to allow the remaining channels to calculate AFD.

C. (1) 3-ONOP-059.8, Power Range Instrumentation Malfunction (2) Verify QPTR within limits using 3-OSP-059.10, Determination of Quadrant Power Tilt Ratio, every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. (1) 3-ON OP-028, Reactor Control System Malfunction (2) Place Rod Control in MANUAL until the channel is restored.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible if examinee believes that the given annunciators suggest a malfunction of a component(s) of the Reactor Control System and this malfunction has led to an excessive QPTR; under these conditions, per Technical Specifications, power must be reduced (via boration, since rod motion is hampered) to <50% of rated thermal power.

B. Incorrect. The first part of the distractor is correct. Plausible if examinee recognizes that many of the failed channels features are bypassed under these conditions, per 3-ONOP-059.8, although these actions are not performed for the purpose of allowing the remaining power-range channels to calculate the axial flux difference.

C. Correct. Based on the given annunciators, a lower power-range nuclear instrument has failed low (other than N44, since no rod motion occurred), rather than a malfunction of a component(s) of the Reactor Control System; hence, 3-ONOP-059.8 provides the appropriate mitigation strategy and requires that the QPTR be monitored (using 3-OSP-059.10, Determination of Quadrant Power Tilt Ratio) every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to satisfy TS 4.2.4.2.

D. Incorrect. Plausible if examinee believes that the given annunciators suggest a malfunction of a component(s) of the Reactor Control System and rods failed to insert in response to a failed-high power-range nuclear instrument (N44); under these conditions, per 3-ONOP-028, the Rod Motion Control Selector must be placed in the MAN position. However, indications suggest a failed-low power-range nuclear instrument (other than N44), since no rod motion occurred.

3-ONOP-059.8 3-ONOP-028 . .

Technical Reference(s): . (Attach if not previously provided)

Technical Specifications Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902206 Obj. 4, 5 (As available)

Question Source: Bank # X PTN L-1S-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Comanche Peak Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

Procedure No.: Procedure

Title:

Page 7

Power Range Approval Date:

3ONOP059.8 Nuclear Instrumentation Malfunction 11/14/07 5.0 WSEQUENT ACTIONS 5.1 Mode I - Power Operation 5.1.1 Malfunction of ONE channel:

1. Place the DROPPED ROD MODE switch for the failed channel in the BYPASS position.

2. Place the applicable ROD STOP BYPASS switch to the failed channel BYPASS position.

NOTE 1 I

If an Upper Section Deviation or Lower Section Deviation alarm occurs, or if ANNUNCIA TORS B-2/2 or B-2/3 annunciate, the actions of 3-OSP-059. 10, Determination of Quadrant Power Tilt Ratio, need to be performed if power is greater than 50 percent. I a a I

3. Transfer the UPPER SECTION comparator defeat switch to the failed channel.

4. Transfer the LOWER SECTION comparator defeat switch to the failed channel.

5. Transfer applicable POWER MISMATCH BYPASS switch to BYPASS the failed channel.

6. Transfer the COMPARATOR CHANNEL DEFEAT switch to the failed channel.

7. IF rod control is in manual due to failed Power Range N-44 channel AND automatic operation is desired, THEN transfer Rod Motion Control Selector switch to AUTO.

8. Perform the following within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of the failure determination:

a. Trip the Power Range bistables by removing the INSTRUMENT POWER fuses from drawer B of the failed channel.

a a a a a a a a NOTE I

The Reactor Protection System bistables associated with the failed power range channel I will be in the tripped condition when its bistable test switch is placed in the test (to the right)

position as indicated by the red trip LED on the Channel Status Light section of the EAGLE

21 Test Panel being ON.

b. IF Power Range I is failed, THEN place the following bistable test switches in Protection Channel I, Rack No. I in the TEST position:

BS-3-412B-l Overpower AT Trip BS-3-412B-2 Overpower AT Rod stop BS-3-412C-l Oveilemperature AT Trip BS-3-412C-2 Overtemperature AT Rod stop W97:TNM/ma/ev/mra

Procedure No Procedure

Title:

Page 8

Power Range Approval Date:

3-ONOP-059.8 Nuclear Instrumentation Malfunction 11/14/07 5.1.1.8 (Contd)

c. IF Power Range 2 is failed, THEN place the following bistable test switches in protection Channel II, Rack No. 11, in the TEST position:

BS-3-422B-1 Overpower AT Trip BS-3-422B-2 Overpower AT Rod stop BS-3-422C-1 Overtemperature AT Trip BS-3-422C-2 Overtemperature AT Rod stop

d. IF Power Range 3 is failed, THEN place the following bistable test switches in Protection Channel 111, Rack No. 14, in the TEST position:

BS-3-432B-1 Overpower AT Trip BS-3-432B-2 Overpower AT Rod stop BS-3-432C-1 Overtemperature AT Trip BS-3-432C-2 Overtemperature AT Rod stop

9. Notify 1&C.

I NOTE I

The first thermocouple QPTR calculation should be performed as soon as possible.

a a I

10. IF greater than 75 percent power, THEN monitor the Quadrant Power Tilt Ratio using 3-OSP-059.10, DETERMINATION OF QUADRANT POWER TILT RATIO. (Per Tech Spec 4.2.4.2)

OR Reduce power to less than 75 percent rated thermal power AND set Power Range Neutron Flux Trip less than 85 percent within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

11. IF a thermocouple QPTR calculation is performed AND acceptable results are NOT achieved, THEN contact Reactor Engineering.

12. IF maintenance is NOT to be performed immediately, THEN attach a clearance tag to each bistable switch tripped in Substep 5.1.1.8.b, 5.1.1.8.c, OR 5.1.1 .8.d in the tripped position.

W97:TNM/ma/ev/mrq

CAUTIONS

If the Rod Control System is inoperable due to Urgent Failure or other cause, the Shift Manager shall be notified immediately.

If a transient occurs and the Reactor cannot be stabilized by boration/dilution or changes in turbine load, the Reactor shall be tripped and a transition made to 3-EOP-E-O, REACTOR TRIP OR SAFETY INJECTION.

I_._

I NOTES I I I

Boration/dilution or changes in turb me load will effect shutdown margin and axial offset. If plant conditions permit, the Shift Manager shall be consulted for methods

used to achieve and maintain stable plant conditions. I

Failure of RCC(s) to move when demanded, (e.g., ROD CONTROL URGENT I FAILURE), constitutes inoperabiity of the associated RCC(s). The requirements of

T.S. 3.1.3.1 apply.

I a

4.0 IMMEDIATE ACTIONS 4.1 Immovable RCC 4.1 .1 IF the Rod Motion Control Selector is in Auto, THEN place in the MAN position.

4.1.2 DO NOT withdraw any control banks until the RCC(s) have been aligned.

4.2 Failure of an RCC Control Bank to Insert with Reactor Control in Automatic 4.2.1 Place the Rod Motion Control Selector switch to the MAN position.

4.3 Continuous Insertion of an RCC Control Bank 4.3.1 Place the Rod Motion Control Selector switch to the MAN position.

4.3.2 IF RCC control cannot be maintained manually, THEN trip the Reactor and Turbine and go to 3-EOP-E-0, REACTOR TRIP OR SAFETY INJECTION.

4.4 Continuous Withdrawal of an RCC Control Bank 4.4.1 Place the Rod Motion Control Selector switch to the MAN position.

4.4.2 IF RCC control cannot be maintained manually, THEN trip the Reactor and Turbine and go to 3-EOP-E-0, REACTOR TRIP OR SAFETY INJECTION.

4.5 Control Bank D Demanded Past ARO Position 4.5.1 None W97:/JWB/cIs/fm/ab

POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER*.

ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED THERMAL POWER.

See Special Test Exceptions Specification 3.10.2.

TURKEY POINT UNITS 3 & 4

- 3/42-11 AMENDMENT NOS. 260 AND 255

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1, within 30 minutes;

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED THERMAL POWER.

c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

TURKEY POINT UNITS 3 & 4

- 3/4 2-12 AMENDMENT NOS. 260 AND 255

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

3. Identify and correct the cause of the out-of -limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.

d. The provisions of Specifications 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a. Calculating the ratio at least once per 7 days when the Power Range Upper Detector High Flux Deviation and Power Range Lower Detector High Flux Deviation Alarms are OPERABLE, and

b. Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when either alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained either from two sets of four symmetric thimble locations or full-core flux map, or by incore thermocouple map is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.4.3 If the QUADRANT POWER TILT RATIO is not within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the POWER DISTRIBUTION LIMITS of 3.2.2 and 3.2.3 are within their limits, a Special Report in accordance with 6.9.2 shall be submitted within 30 days including an evaluation of the cause of the discrepancy.

TURKEY POINT UNITS 3 & 4

- 3/4 2-13 AMENDMENT NOS. 260 AND 255

SRO Question # 82 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1I immediate operator actions? Yes RO question I

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major FOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or iJ.Eestion overall mitigative strategy of a 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 LonlJ

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

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K/A # 068 A2.04 Importance Rating 3.3 Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of automatic isolation Proposed Question: SRO Question # 83 Given the following conditions:

Unit3isinMODE3.

During testing, CV-3-4668A, RCDT Pump Discharge to Holdup Tank CTMT Isolation valve is found stuck in the open position.

Repairs to CV-3-4668A will take approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

Which ONE of the following completes the statements below to meet the Technical Specification requirements?

Isolate CV-3-4668A by de-energizing SV-3-4668B to close CV-3-4668B within (1) hours or be in at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months to (2)

A. (1)4 (2) ensure valves are isolated within the specific isolation times of the LOCA analysis B. (1)6 (2) restrict the release of radioactive materials from Containment to within the assumptions of the COLR C. (1)4 (2) restrict the release of radioactive materials from Containment to within the assumptions of the COLR D. (1)6 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION (2) ensure valves are isolated within the specific isolation times of the LOCA analysis.

Proposed Answer: A Explanation (Optional):

A. Correct. TS 3.6.4 requires that the inoperable isolation valve (CV-3-4668A) be restored to operability within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or the associated penetration be isolated (by deactivating!

closing CV-3-4668B, which is downstream of CV-3-4668A) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, the plant must be in cold shutdown within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months (note that the plant is already in hot standby, hence the 6-hour action is not applicable). The basis for TS 3.6.4 states that the isolation time limit is consistent with the assumed isolation times for these valves in the LOCA analysis.

B. Incorrect. Plausible if examinee confuses the time requirement for the LCO action statement to be in hot standby (6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) with that for penetration isolation (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ).

Plausible because the primary intent of Containment isolation valves is to prevent the release of radioactive materials to the environment; however, the basis for the isolation time limit is a function of the assumptions in the LOCA analysis, not those in the COLR.

C. Incorrect. The first part of the distractor is correct. Plausible because the primary intent of Containment isolation valves is to prevent the release of radioactive materials to the environment; however, the basis for the isolation time limit is a function of the assumptions in the LOCA analysis, not those in the COLR.

D. Incorrect. Plausible if examinee confuses the time requirement for the LCO action statement to be in hot standby (6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) with that for penetration isolation (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ). The second part of the distractor is correct.

Technical Specifications, 0- .

Technical Reference(s): (Attach if not previously provided)

ADM-536 Proposed References to be provided to applicants during examination: No LP6902149Obj.8 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.:

11 PROCEDURE TITLE:

f PAGE:

PROCEDURE NO.:

TECHNICAL SPECIFICATION BASES CONTROL PROGRAM I 136 of 192 O-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 120 of 176) 3/4.6.3 Deleted 3/4.6.4 Containment Isolation Valves The OPERABILITY of the Containment Isolation Valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.

Containment isolation within the time limits specified in the In-Service Testing Program is consistent with the assumed isolation times of those valves with specific isolation times in the LOCA analysis.

Note that Tech Spec 3.6.4 applies only to automatic Containment Isolation Valves. Automatic Containment Isolation Valves are valves, which close automatically on a Containment Isolation Phase A signal, Containment Phase B, or a Containment Ventilation Isolation signal, and check valves.

3/4.7 Plant Systems 3/4.7.1 Turbine Cycle 3/4.7.1.1 Safety Valves The OPERABILITY of the main steam line Code Safety Valves ensures that the Secondary System pressure will be limited to within 110%

(1193.5 psig) of its design pressure of 1085 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e.,

NO steam bypass to the condenser).

The primary purpose of the Main Steam Safety Valves (MSSVs) is to provide overpressure protection for the secondary system. The MSSVs also provide protection against over pressurizing the Reactor Coolant Pressure Boundary (RCPB) by providing a heat sink for the removal of energy from the Reactor Coolant System (RCS) if the preferred heat sink, provided by the Condenser and Circulating Water System, is NOT available.

CONTAINMENT SYSTEMS 3/4.6.4 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.4 Each containment isolation valve shall be OPERABLE with isolation times less than or equal to required isolation times.

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

ACTION:

Vvith one or more isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or

b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic containment isolation valve secured in the isolation position, or

c. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange, or

d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.4.1 The isolation valves shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time.

CAUTION: The inoperable isolation valve(s) may be part of a system(s). Isolating the affected penetration(s) may affect the use of the system(s). Consider the technical specification requirements on the affected system(s) and act accordingly.

TURKEY POINT UNITS 3 & 4 3/46-16 AMENDMENT NOS. 260 AND 255

SRO Question # 83 Clarification Guidance for SRO-only Questions Rev 1 (03/1112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1 I immediate operator actions? Yes RO question I

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EQ Ps?

Can the question be answered solely by knowing the purpose, overall sequence of events, or question overall mitigative strategy of a 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

SRO Question # 83 Clarification Guidance for SRO-only Questions Rev 1(03111/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TSJTRM Action?

Can question be answered solely by knowing the LCOITRM information listed above-the-line? zLr;stio Can question be answered solely by knowing 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) I SRO-only

Knowledge of TS bases that is required to analyze TS required actions and terminology j question No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A # 007 EA2.02 Importance Rating 4.6 Ability to determine or interpret the following as they apply to a reactor trip: Proper actions to be taken if the automatic safety functions have not taken place Proposed Question: SRO Question # 84 Given the following conditions:

Unit 3 just completed a refueling outage and is now at 50% power.

Subsequently:

A steam break occurs just upstream 3-10-319, SG C AFW Pump Steam Supply Upstream Isolation.

The Reactor automatically trips from a Safety Injection Signal.

3-EOP-E-0, Rx Trip or Safety Injection is in progress.

The BOP performs Attachment 3, Prompt Action Verification, and finds Feedwater is NOT isolated.

Which ONE of the following is the reason why the Unit Supervisor directs the BOP to isolate Feedwater?

A. To prevent an excessive RCS cooldown.

B. To prevent Containment over-pressurization.

C. To prevent uncontrolled loss of Pressurizer level.

D. To prevent a restart accident from the positive reactivity added.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: A Explanation (Optional):

A. Correct. The basis document for 3-EOP-E-0 indicates that feedwater is isolated on a steam-line break to prevent uncontrolled filling of any SG and the associated excessive RCS cooldown, which could aggravate the transient.

B. Incorrect. Plausible if examinee believes that the steam-line break is inside Containment and feedwater isolation will serve to reduce the total amount of steam generated and released to Containment; however, 3-10-319 is located outside of Containment.

C. Incorrect. Plausible if examinee recognizes that the RCS cooldown caused by the steam-line break will also lead to contraction of the reactor coolant and a drop in Pressurizer level; however, a flow-limiting device in the SGs steam outlet nozzle (which is designed to choke steam flow during a postulated steam-Hne break and prevent rapid depressurization), in conjunction with automatic RCS-letdown isolation, should prevent an uncontrolled loss of Pressurizer level.

D. Incorrect. Plausible if examinee recognizes that the RCS cooldown caused by the steam-line break will also lead to a positive-reactivity addition; however, a postulated steam-line break is not expected to add sufficient reactivity to overcome the TS-required shutdown margin (i.e., 1.77% Ak/k) and facilitate a restart accident.

Technical Reference(s): BD-EOP-E-0, 0-ADM-536 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902321 Obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

Page 56 BD-EOP-E-O Reactor Trip or Safety Injection 8/15/14 BASIS DOCUMENT WOG Procedure Step 5 PTN Procedure Step ATT. 3 Step 2 Verify Feedwater Isolation BASIS:

The Main Eeedwater System is isolated on a EW Isolation signal to prevent uncontrolled filling of any steam generator and the associated excessive RCS cooldown, which could aggravate the transient, especially if it were a steamline break.

STEP DEVIATIONS FROM WOG GUIDELINES:

TYPE DESCRIPTION 2 Substep 2a was added to place main feedwater pumps in Stop. This will allow operation of AFW pumps later in the EOPs.

8 The words Flow control valves were changed to Feedwater control valves to conform with plant specific terminology.

8 The words Flow control bypass valves were changed to Feedwater bypass valves to conform with plant specific terminology.

8 The words FW isolation valves were changed to Feedwater isolation valves to conform with plant specific terminology.

9 Substep 2d was added as a result of the addition of Fast Acting Feedwater Bypass Isolation Valves for EPU. See ECs 242442(U3) and 246878(U4) for more information.

2 SIG Blowdown and S/G sample valves at PIN do not directly receive a Feedwater Isolation signal, and are verified during Phase A verification.

9 The RNO column was changed to provide specific operator tasks to be performed. The words Locally close valves was considered to be sufficient.

3 Turkey Point standby feedwater pumps do not trip on a safety injection signal. Flow from these pumps would continue if any feedwater bypass valve fails to close. Substep f was added to ensure all sources of non-emergency feedwater are isolated following safety injection actuation. Since these pumps are normally secured, the operator is directed to verifv the pumps off. If they are running and supplying this unit, the RNO column directs the operator to trip the standby feedwater pumps.

PLANT SPECIFIC SETPOINTS:

N/A W201 O:/fm/clstcls

REVISION NO.: PROCEDURE TITLE: PAGE:

11 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 46 of 192 PROCEDURE NO.:

O-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 30 of 176) 3/4.1 Reactivity Control Systems 3/4.1.1 Boration Control 3/4.1.1.1 &

3/4.1.1.2 Shutdown Margin A sufficient SHUTDOWN MARGIN ensures that: (1) The reactor can be made subcritical from all operating conditions, (2) The reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg.

With Tavg greater than 20a°F, the most restrictive condition occurs at EOL, with Tavg at NO load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. The COLR provides a curve (formerly TS Figure 3.1-1) showing the SHUTDOWN MARGIN equivalent to 1.77% Ak/k at the end-of-core-life with respect to an uncontrolled cooldown. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with UFSAR safety analysis assumptions.

The COLR figure provides a separate curve showing the minimum SHUTDOWN MARGIN equivalent to 1.77% Ak/k in MODE 4 without a Reactor Coolant Pump running. The reduced RCS mixing volume in this condition requires a higher minimum SHUTDOWN MARGIN is required to assure adequate operator response time is available to identify and terminate an inadvertent dilution event prior to loss of all SHUTDOWN MARGIN. With Tavg less than 200°F, a SHUTDOWN MARGIN equivalent to 1.77% Ak/k is required at all times to assure adequate operator response time is available to identify and terminate an inadvertent dilution event in MODE 5 prior to loss of all SHUTDOWN MARGIN. FormerTS Figure 3.1-1 titled Required Shutdown Margin versus Reactor Coolant Boron Concentration has been moved to the COLR per Amendments 247 and 243.

The boron rate requirement of 16 gpm of 3.0 wt% (5245 ppm) boron or equivalent ensures the capability to restore the SHUTDOWN MARGIN with one OPERABLE Charging Pump.

SRO Question # 84 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? j+restion Can the question be answered solely by knowing 1 immediate operator actions? I Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters question that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a 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 only

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

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1. DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A # 025 AA2.02 Importance Rating 3.8 Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Leakage of reactor coolant from RHR into closed cooling water system or into reactor building atmosphere Proposed Question: SRO Question # 85 Given the following conditions:

Unit 3 is being cooled down using 3B RHR loop.

Containment purge is in progress.

RCS temperature is 195° F.

RCS pressure is 310 psig.

Subsequently:

RCS pressure is lowering.

Annunciator H6/2, RHR HX HI/LO FLOW, alarms.

Annunciator G915, CNTMT SUMP HI LEVEL, alarms.

R-3-1 1, Containment Air Particulate Monitor, is in alarm.

R-14, Plant Vent Gas Monitor, is in alarm.

Which ONE of the following identifies (1) the procedure the Unit Supervisor will enter to mitigate the event and (2) why?

A. (1) 3-ONOP-041.7, SHUTDOWN LOCA (<1000 psig)

(2) To isolate the leak in containment.

B. (1) 3-ONOP-041.8, SHUTDOWN LOCA (2) To isolate the leak in the RHR HX.

C. (1) 3-ONOP-041.7, SHUTDOWN LOCA (<1000 psig)

(2) To isolate the leak in the RHR HX.

D. (1) 3-ONOP-041.8, SHUTDOWN LOCA (2) To isolate the leak in containment.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible if examinee does not recognize that the plant is in Mode 5 (i.e., RCS Tavg <200°F). The second part of the distractor is correct.

B. Incorrect. The first part of the distractor is correct. Plausible if examinee believes that the leak is in the RHR Heat Exchanger Room, since R-14 is in alarm (although this condition would not cause R-3-1 1 to alarm). However, the increasing containment sump level and R-3-11/R-14 alarms suggest that the leak is located inside containment.

C. Incorrect. Plausible if examinee does not recognize that the plant is in Mode 5 (i.e., RCS Tavg is <200°F) and believes that the leak is in the RHR Heat Exchanger Room, since R 14 is in alarm (although this condition would not cause R-3-1 1 to alarm). However, the increasing containment sump level and R-3-1 1/R-14 alarms suggest that the leak is located inside containment.

D. Correct. 3-ONOP-041 .8 is the correct procedure, as the plant is in Mode 5 (i.e., RCS Tavg 200°F). The increasing containment-sump level and R-3-11IR-14 alarms suggest that the leak is located inside containment.

3-ONOP-041 .7 Technical Reference(s): 3-ONOP-041 .8 (Attach if not previously provided)

TS Table 1.2 Proposed References to be provided to applicants during examination: NO PTN 6900265 and PTN 6900266 Learning Objective: (As available)

Obj. 2 and 7 Question Source: Bank # 100829 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2011 Point Beach Question Cognitive Level: Memory or Fundamental Knowledge PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

TABLE 1.2 OPERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, Keff THERMAL POWER* TEMPERATURE

1. POWER OPERATION 0.99 > 5% 350°F

2. STARTUP 0.99 5% 350°F

3. HOT STANDBY < 0.99 0 350°F

4. HOT SHUTDOWN 350°F > Tavg

< 0.99 0

> 200°F

5. COLD SHUTDOWN < 0.99 0 200°F

6. REFUELING** 0.95 0 140°F

ExcIuding decay heat.

- FueI in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

TURKEY POINT UNITS 3 & 4 1-8 AMENDMENT NOS. 137 AND 132

1.0 PURPOSE 1 .1 This procedure provides actions for protecting the reactor core in the event of a I OSS Of COOl ANT ACCIDENT (LOCA) that occurs during either Mode 3 less than 1000 psig, or in Mode 4, 1 .2 This procedure is applicable for Modes 3 after the accumulators are isolated, or in Mode 4.

2.0 SYMPTOMS OR ENTRY CONDITIONS 2.1 This procedure is entered from:

2.1.1 3-ONOP-041 .3, Excessive Reactor Coolant System Leakage, when leakage exceeds the capacity of the CVCS system.

2.1.2 3-EOP-ES-0.2, Natural Circulation Cooldown, was entered when the Unit 3 was in Mode 3 (less than 1000 psig) or Mode 4, AND SI is required.

2.1.3 3-EOP-ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (with RVLMS), if 3-EOP-ES-0.2, Natural Circulation Cooldown, was entered with Unit 3 in Mode 3 (less than 1000 psig) or Mode 4, AND SI is required.

2.1.4 3-EOP-ES-0.4, Natural Circulation Cooldown with Steam Void in Vessel (Without RVLMS), if 3-EOP-ES-0.2, Natural Circulation Cooldown, was entered with Unit 3 in Mode 3 (less than 1000 psig) or Mode 4, AND SI is required.

2.1.5 3-ONOP-047.I, Loss of Charging Flow in Modes I Through 4, Foldout page item

3 when PRZ level can NOT be maintained above 7% with the plant in Mode 3 (less than I 000#), or Mode 4.

\A!9fl1 flflI-1frI/fmfrk

REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN LOCA [MODE 5 OR 6] 4 of 43 PROCEDURE NO.:

3-ONOP-041.8 TURKEY POINT UNIT 3 1.0 PURPOSE This procedure provides actions necessary for maintaining core cooling and protecting the reactor core in the event that RHR cooling is lost during Mode 5 OR 6 EXCEPT when the Refueling Cavity is fIooded.

2.0 ENTRY CONDITIONS 2.1 Annunciators

H 6/2, RHR HX HI/LO FLOW

H 6/4, RHR PP A/B TRIP

I 7/6, RHR SUMP PUMP ROOM A HI LEVEL

I 8/6, RHR SUMP PUMP ROOM B HI LEVEL

I 3/6, RHR SUMP HX ROOM HI LEVEL

I 7/3, RX VESSEL DRAINDOWN LO-LO-LEVEL

A 7/1, PRT HI/LO LEVEL HI PRESS/TEMP 2.2 Indications

Neither RHR pump is operating when required for decay heat removal.

Low flow indicated on Fl-3-605.

Air Binding of the RHR pumps as evidenced by the following:

Motor current oscillations Erratic flow oscillations Excessive pump noise Pump Cavitation 2.3 Procedures

3-ONOP-041 .3, Excessive Reactor Coolant System Leakage

3-ONOP-050, Loss of RHR

SRO Question # 85 Clarification Guidance for SRO-only Questions Rev 1(03111/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

immediate operator actions?

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major FOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

1 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A# 026 2.1.19 Importance Rating 3.8 Conduct of Operations: Ability to use plant computers to evaluate system or component status.

Proposed Question: SRO Question # 86 Given the following conditions:

Unit 3 is at 95% power.

3A CCW Pump is OOS for maintenance.

3B CCW Pump tripped.

3C CCW Pump is running.

The Unit Supervisor reviews DCS - RCP Detailed Data Summary.

Which ONE of the following identifies the notification requirement, if any, for the given plant conditions after the crew has completed initial actions?

REFERENCE PROVIDED A. No notifications are required.

B. 15 minute to the State and Counties.

C. 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to the Nuclear Regulatory Commission Operations Center (NRCOC).

D. 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to the Nuclear Regulatory Commission Operations Center (NRCOC).

Proposed Answer: C Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION A. Incorrect. Plausible if examinee believes that no manual trip is warranted, hence no notification is necessary. Also plausible if examinee recognizes the need for a manual trip, but believes notifications are only warranted for automatic trips.

B. Incorrect. Plausible if examinee believes that conditions warrant entry into the sites Emergency Plan, which would require offsite notifications within 15 minutes (after event classification).

C. Correct. 0-ADM-1 15 and 0-ADM-560 state that an RPS actuation from criticality (manual or automatic) requires a notification within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 0-ADM-1 15 requires that a notification be made to the NRCOC.

D. Incorrect. Plausible if examinee recognizes that 0-ADM-1 15 also has an 8-hour notification requirement for a manual or automatic RPS actuation; however, 0-ADM-560 states that an 8-hour notification is required when the trip occurs from a non-critical condition. 0-ADM-1 15 requires that a notification be made to the NRCOC, for the conditions given.

O-ADM-115 0-ADM-560 3- .

Technical Reference(s): (Attach if not previously provided)

ONOP041.1 Proposed References to be provided to applicants during examination: No Learning Objective: PTN 6900205, Obj.4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

STEP ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

43 Check RCP Temperatures

a. Maintain all RCP motor bearing a. Perform the following:

temperatures LESS THAN 1 9°F

1) Manually trip the reactor perform 3-EOP-E-0, REACTOR TRIP OR SAFETY INJECTION, while continuing with this procedure.

2) WHEN reactor verified tripped, THEN stop the affected RCP(s).

b. Check RCP Motor Status b. Return to Step 10.

Any RCP stator temperature on DCS -

GREATER THAN 210°F OR INCREASING OR

Any RCP motor bearing temperature on DCS GREATER THAN 185°F OR INCREASING

c. Verify 3A and 3B 4KV bus voltages -

BETWEEN 3740 AND 4580 VOLTS

d. Consult with System Engineer and d. Perform the following:

Operations supervision to determine cause of high temperature(s) AND to determine if 1) Manually trip reactor AND perform RCP operation should continue 3-EOP-E-0, REACTOR TRIP OR SAFETY INJECTION, while continuing with this procedure.

2) WHEN the reactor verified tripped, THEN stop the affected RCP(s).

e. Return to Step 10 END OF TEXT AI)fl1 flIIkh-Ii,-I

A r E11 I -

MAIN MENU RCP DETAILED 3A MOTOR B MOTOR 3C MOTOR RELATED DATA

SUMMARY

DISPLAYS ALARM ALARM ALARM RCP A RCP B RCP C PRIMARY THERMALBARRIER p INH2O INH2O INH2O SECONDARY SEAL INJECTION FLOW GPM GPM CPM NUMBER ONE SEAL LEAKOFF GPM GPM FLOW OPM POWER THERMAL BARRiER COOLING WATER LO FLOW ALARM PUMP BEARING DEC F DEC F DEO F TEMPERATURES TURBINE NUMBER ONE SEAL LEAKOFF : DEC F DEC F I DEG F TEMPERATURE VCT TEMPERATURE DEC F DEC F DEC F ESF UPPER OIL RESERVOIR HLLO LEVEL ALARM LOWER OIL RESERVOIR HL/LO LEVEL ALARM SUPPORT SYS UPPERTHRUSTBEARINGTEMP iIiXi DEC F DEG F DEG F LOWER THRUST BEARING TEMP DEC F DEC F DEG F EMERG RESP UPPER GUIDE BEARING TEMP DEC F DEC F I DEC F LOWER GUIDE BEARING TEMP DEC F DEC F DEG F UTILITIES MOTOR BEARING COOLING WATER HI TEMP ALARM ;

MOTOR BEARING COOLING GPM GPM ALARMS WATER FLOW GPM STATOR WINDING TEMP I DEG F I DEC F : DEG F PRINT SEAL BYPASS LO FLOW ALARM [] !

I 3A PUMP B PUMP 3C PUMP PREVIOUS ALARM ALARM ALARM

SRO Question # 86 Clarification Guidance for SRO-only Questions Rev 1(03111/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 5543(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the questio n Yknow be answered in9ejRQestiofl immediate ope rator actions?

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A # 029 EA2.05 Importance Rating 3.4 Ability to determine or interpret the following as they apply to a ATWS: System component valve position indications Proposed Question: SRO Question # 87 Given the following conditions:

3-EOP-FR-S.1, Response to Nuclear Power Generation/ATWS, is in progress.

The Reactor fails to trip from the Control Room.

Annunciator B914, ROD CONTROL URGENT FAILURE, alarms

Local actions to trip the reactor are unsuccessful.

3B Boric Acid Transfer pump is running.

Prior to manipulation, MOV-3-350, Emergency Boration Valve, position red and green lights are found NOT lit.

CET temperature is 1250°F and rising.

Which ONE of the following identifies (1) the required action to initiate Emergency Boration and (2) whether the EC is required to issue PARS?

REFERENCE PROVIDED A. (1) Closing LCV-3-115C, VCT Outlet Isolation valve.

(2) PARS required.

B. (1) Opening 3-356, Manual Emergency Boration valve (2) PARS NOT required.

C. (1) Closing LCV-3-115C, VCT Outlet Isolation valve.

(2) PARS NOT required.

PTN L-15-]. DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. (1) Opening 3-356, Manual Emergency Boration valve.

(2) PARS required.

Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible if examinee doesnt recognize that the 3B Boric Acid Transfer pump is available and running; if it were not, the appropriate RNO for Step 4.d in 3-EOP-FR-S.1 would be to align the Charging Pump suction to the RWST by closing LCV-3-115C.

The second part of the distractor is correct.

B. Incorrect. The first part of the distractor is correct. Plausible if examinee does not recognize the red path on the Core Cooling CSF and believes that a Site Area Emergency (SS2) is in effect, which requires no PARs.

C. Incorrect. Plausible if examinee doesnt recognize that the 3B Boric Acid Transfer pump is available and running; if it were not, the appropriate RNO for Step 4.d in 3-EOP-FR-S.1 would be to align the Charging Pump suction to the RWST by closing LCV-3-1 15C.

Plausible if examinee does not recognize the red path on the Core Cooling CSF and believes that a Site Area Emergency (SS2) is in effect, which requires no PARs.

D. Correct. As MOV-3-350 cannot be opened, the associated RNO for Step 4.e in 3-EOP-FR-S.1 requires 3-356 to be opened. PARs are required in this case, since a General Emergency (SG2) is in effect (i.e., failed auto/manual/local trip, with a red path on the Core Cooling CSF [CET5 >1200°F]).

3-EOP-FR-S.1 3-EOP-F-0 Technical Reference(s): .

(Attach if not previously provided)

EPIP-20101 Attachment 1 Proposed References to be provided to applicants during examination: Yes Learning Objective: LP 6902346 Obj. 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Reference provided: Form 668 EAL Classifications, Hot Conditions, page 16.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

SRO Question # 87 Clarification Guidance for SRO-only Questions Rev 1(03/1112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

p Can the question be answered solely by knowing 1I immediate operator actions? Yes I RO question I 1 Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EQ Ps?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a 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 jOnlY

Knowledge of diagnostic steps and decision points in the EQPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN [-15-1 DRAFT NRC EXAM SECURE INFORMATI ON Examination Outline Cross-reference: Level RO SRO Tier#

Group #

K/A# 056 2.1.31 Importance Rating 4.3 Conduct of Operations: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

Proposed Question: SRO Question # 88 Given the following conditions:

Unit 4 is in MODE 3

The switchyard lost power 18 minutes earlier and remains de-energized.

4A EDG is running and its output breaker red light is lit.

4B EDG is running and its output breaker gree n light is lit.

4B RCP has no light indication.

The station blackout tie permissive blue light is off.

Which ONE of the following identifies the orde r to be given for the required initial operator action in accordance with 4-ONOP-004.3, Loss of 4B 4KV Bus, and the correct Emergency Classification?

REFERENCE PROVIDED A.

Close the 4B EDG Output Breaker.

Alert B.

Open the 4B RCP Breaker.

Alert C.

Open the 4B RCP Breaker.

Unusual Event PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-1S-1 DRAFT NRC EXAM SECURE INFORMATION D.

Close the 4B EDG Output Breaker.

Unusual Event Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible if examinee recognizes that after the 4B RCP breaker is opened, the 4B EDG output breaker may have to be manually closed (i.e., if it does not auto-close); however, the 4B RCP breaker needs to opened before the 4B EDG output breaker can be closed or the 4B EDG would be synchronized onto the bus, with the 4B RCP aligned. The second part of the distractor is correct.

B. Correct. With the SBO tie permissive blue light off (and no light indication for the 4B RCP, bus stripping is not assured; therefore, performance of ONOP-004.3 must be stopped and Attachment 1 (bus stripping) completed. If this were not done, the next action would be to synchronize the 4B EDG onto the bus, with the 4B RCP aligned, which would be a significant error. Manual or local closure of the EDG breaker is not protected by bus clearing. The 4A 4KV Bus is energized, since the 4A EDG is running and its output breaker is closed (red light indication); howev er, its subsequent loss would lead to a station blackout, therefore the correct classif ication is an Alert (SA5).

C. Incorrect. The first part of the distractor is correct. The correct classification is an Alert (SA5), rather than an Unusual Event, since a subsequent loss of the 4A 4KV Bus would lead to a station blackout.

D. Incorrect. Plausible if examinee recognizes that after the 4B RCP breaker is opened, the 4B EDG output breaker may have to be manually closed (i.e., if it does not auto-close); however, the 4B RCP breaker needs to opened before the 4B EDG output breaker can be closed or the 4B EDG would be synchronized onto the bus, with the 4B RCP aligned. The correct classification is an Alert (SA5),

rather than an Unusual Event, since a subsequent loss of the 4A 4KV Bus would lead to a station blackout.

4-ONOP-004.3, Step 7and Technical Reference(s): Attachment 1, Step 2; EPIP- (Attach if not previously provided) 20101, Attachment 1 Proposed References to be provided to applicants during examination: No Learning Objective:

(As available)

PTN L-15--1 DRAFT NRC EXAM SECURE IN FORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Level 2 because the operator must analyze the situation presented and conclude that the 4B RCP breaker is closed, based on the station blackout tie permis sive blue light being off and no light indication on the 4B RCP. Bus stripping is not verified and Attachment 1 is not complete.

Therefore performance of ONOP-004.3 must be stopped (reference ONOP-004.3 Step 7.B). If not, the next action is to synchronize the EDG onto the bus. It would be a significant error to attempt to close the 4B EDG breaker because manua l or local closure of the EDG breaker is not protected by bus clearing. The EDG would energize the bus with the RCP attached.

Then the SRO must recognize that with the 4A EDG runnin g and its output breaker closed (red light indication), the 4A 4KV Bus is energized. However, its subsequent loss would lead to a station blackout, therefore the correct classification is an Alert.

SRO level because the SRO is recalling what strategy or action is written into a plant procedure (4-ONOP-004.3 and 0-EPIP-201 01), including when the strategy or action is required. (Ref Guidance for SRO-only Questions Rev 0

- page 7 of 19)

Reference provided: Form 668 EAL Classifications, Hot Conditions, page 16.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE:

PAGE:

LOSS OF4B4KVBUS PROCEDURE NO.: 9of35 4-ONOP-004.3 TURKEY POINT UNIT 4 I STEP flACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 3.2 Subsequent Operator Actions (continued)

NOTE

When Unit 4 startup transformer is available, offsite power to 4B 4KV Bus should be restored using 4-ONOP-004.1, System Restoration Following Loss of Offsite Power.

If 4B Sequencer failure has occurred and SI has actuated, the 4B EDO output breaker may NOT CLOSE unless SI is RESET.

CAUTION Steady state loading on each Unit 4 EDG shall NOT exceed 2874 KW. Load transients up to 3162 KW are acceptable when starting additio nal equipment.

7. ENERGIZE 4B 4KV Bus from 4B Emergency Diesel Generator.

A. Manually START 4B Emergency Diesel PERFORM the following:

Generator from Control Room by any of the following methods: 1. IF 4A and 4B 4KV Buses are DEENERGIZED, THEN OBSERVE Emergency start Note prior to Section 3.2, Step 8 AND

GO TO Section 3.2, Step 8.

Rapid start

2. DIRECT operator to locally start Normal start 4B Emergency Diesel Generator per 4-ONOP-023.2, Emergency Diesel Generator Failure.

3. IF 4B Emergency Diesel Generator can NOT be started, THEN OBSERVE Note prior to Section 3.2, Step 8 AND GO TO Section 3.2, Step 8.

B. ENSURE 4B 4KV Bus stripping from WHEN bus stripping is COMPLETE, THEN Attachment 1, 4B 4KV Bus Stripping GO TO Section 3.2, Step 7.C.

COMPLETE.

REVISION NO.: PROCEDURE TITLE:

PAGE:

PROCEDURE NO.:

LOSS OF4B4KV BUS 29 of 35 4-ONOP-004.3 TURKEY POINT UNIT 4 ATTACHMENT I 4B 4KV Bus Stripping (Page 1 of 2)

1. IF 4B 4KV Bus is DEENERGIZED AND 4D 4KV Bus is aligned to 4B 4KV Bus, THEN:

A. CHECK Station Blackout Tie Permissive Blue light is ON.

B. ENSURE 3AD07, STATION BLACKOUT BREAKER, OPEN.

2. IF any of the following conditions exist:

4B 4KV Bus is DEENERGIZED and 4D 4KV Bus is NOT aligned to 4B 4KV Bus Station Blackout Tie Permissive Blue light is OFF THEN ENSURE the following breakers are OPEN:

4AB22, 4B 4KV Bus Tie to 4A or 40 4KV Bus

4AB05, Startup Transformer 4B 4KV Bus Supply

4AB02, Auxiliary Transformer4B Bus Supply

4AB1O, Heater Drain Pump 4B

4AB20, Condensate Pump 4B

4AB12, Safety Injection Pump 4B

4AB15, Residual Heat Removal Pump 4B

4AB13, Component Cooling Water Pump 4B

4ABO1, Reactor Coolant Pump 4B

4AB06, Reactor Coolant Pump 4C

4AB17, Intake Cooling Water Pump 4B

4AB1 1, Turbine Plant Cooling Water Pump 4B

4AB16, Circulating Water Pump 4B1

fsRc5QuestFo#88 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(

b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

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

1 Yes RD question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? question rCan the question be answered solely by knowing I the purpose, overall sequence of events, or question overall mitigative strategy of a 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 Lonly

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

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A# 062 2.4.20 Importance Rating 4.3 Emergency Procedures I Plan: Knowledge of operational implications of EOP warnings, cautions, and notes.

Proposed Question: SRO Question # 89 Given the following conditions:

Unit 3 is operating at 100% power.

Subsequently:

A LBLOCA occurs on Unit 3.

Crew enters 3-EOP-E-0, Reactor Trip or Safety Injection.

A loss of all AC power occurs on Unit 3.

Crew enters 3-EOP-ECA-0.0, Loss of All AC Power.

Unit Supervisor reviews the CAUTION on resetting the SI signal prior to attempting to restore a 4KV Bus.

Which ONE of the following identifies the basis for this CAUTION?

A. Ensures SI actuation when additional automatic SI setpoin ts are reached.

B. Defeats automatic loading of the 4KV bus on an additio nal automatic SI signal.

C. Permits automatic loading of the 4KV bus when power is restored.

D. Prevents a manual SI actuation when power requirements are limited.

Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible if examinee incorrectly believes that an SI signal reset is necessary to ensure subsequent SI signals provide proper actuation; however, resetting an SI PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-]. DRAFT NRC EXAM SECURE INFORMATION signal defeats automatic loading of a 4KV Bus, following loss and restoration of all AC power.

B. Correct. Should AC power be restored after an SI signal initiation, but prior to SI reset, safeguards equipment would automatically load on the bus (i.e.,

outside of operator and/or procedural control). Instead, resetting an SI signal defeats the automatic loading of a 4KV Bus, following a loss and restoration of all AC power.

C. Incorrect. Plausible if examinee incorrectly believes that resettin g an SI signal ensures its actuation upon restoration of power (rather than clearing it);

however, the intent of an SI signal reset is to defeat the automatic loading of a 4KV Bus, following a loss and restoration of all AC power.

D. Incorrect. Plausible if examinee incorrectly believes that an SI signal reset would defeat a subsequent manual SI actuation, thereby preventing the autom atic loading of a 4KV Bus when power requirements are a concern.

3-EOP-ECA-0.0 BD-EOP-ECA-Technical Reference(s):

(Attach if not previously provided) 0.0 561 0TL1 Sh. 11 Proposed References to be provided to applicants during examin ation: No Learning Objective: PTN 6900348, Obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE:

PAGE:

PROCEDURE NO.:

LOSS OF ALL AC POWER 9 Of 88 3-EOP-ECA-O.O TURKEY POINT UNIT 3 I STEP I ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I CAUTION

. If SI has been reset OR SI Actuation occurs on the other unit, Safe guards Equipment and at least one Computer Room Chiller needs to be restored to the required configuration.

. If an SI signal exists OR is actuated during this procedure, it must be reset to ensure restoration of a power source and to ensure contro lled loading of equipment on the 4KV bus.

NOTE

. Attachment 5 provides a reference for Emergency Diesel Genera tor loads.

. If a Sequencer Failure has occurred AND SI has actuated, the associated EDG Output Breaker may NOT close unless SI is reset.

. Unit Supervisor shall evaluate plant conditions and establish 4KV Bus Priority.

. An Available 4KV Bus is one that is NOT locked out, AND is or can be, successfully stripped.

. Reset of the EDG Lock-Out needs to occur immediately after pressing ALARM RESET pushbutton to prevent a subsequent Start Failure 15 seconds later.

5. Try To Restore Power To 3A OR 3B 4KV Bus:

a. Check 4KV BUS PRIORITY 3A a. Go to Step 5.o.

b. Check 3A Bus Lockout Relay b. Perform the following:

RESET

1) Reset Lockout Relay.

2) IF Lockout relay can NOT be reset, AND 4KV Bus Priority was 3B, THEN go to Step 5.y.

3) H Lockout relay can NOT be reset, THEN go to Step 5.o.

Page 16 BDEOP-ECA-O.O Loss of All AC Power 8/15/14 F WOG Procedure Step BASIS DOCUMENT 6 CAUTION 2 PTN Procedure Step 5 CAUTION 2 If an SI signal exists OR is actuated during this procedure, it must be reset to ensure restoration of a power source and to ensure controlled loading of equipment on the 4KV Bus.

BASIS:

Resetting the SI signal is consistent with the loss of all ac power philosophy of defeati ng automatic loading of the ac emergency buses. SI signal reset is also necessary to permit recovery as detailed in ECA-O.I for as wide a range of RCS conditions as possible, i.e.,

those conditions wherein RCS pressure or secondary system pressure are below their actuation setpoints but the recovery selection procedure criteria (RCS subcooling and pressurizer level) are above minimum values which permit recover per ECA-O.I. Since ECA-O.I attemp ts to recover nominal RCS conditions utilizing normal operating systems, automatic safegu equipment actuation must be prevented if RCS or secondary pressures are reduced ards below the SI actuation setpoints. SI signal reset is limited by the time delay on the reset logic. Should power he restored after SI signal initiation but prior to reset, the SI equipm ac ent will automatically load on the ac bus. This possibility is address in the recovery guideline selecti criteria. on STEP DEVIATIONS FROM WOG GUIDELINES:

TYPE DESCRIPTION 9 The WOG guidelines do not provide distinct definitions for the terms should and shall.

The word should was changed to must to denote a requirement.

8 The words it must be reset to ensure restoration of a power source and to ensure controlled were added to include a plant specific requirement associated with the bus sequencer. It was added as a result of information obtained during training on the plant reference simulator and classroom training on the sequencer.

9 WOG step 6, CAUTION 2 was divided into two cautions: one which reminds operators of the need to reset SI to allow manual loading, and one which reminds operators of the need for manual loading.

9 Caution was moved from Step 6 to Step 5 to address the potential need to reset SI in the event of certain failures to meet plant specific requirements needed to close the diesel breaker.

PLANT SPECIFIC SETPOINTS:

N/A W201 O:/DH/cls/cls/cls

SRO Question # 89 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? question Can the question be answered solely by knowing 1I I immediate operator actions? Yes RO question Ii Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

il,festio Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

il_festio 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 Lonl1 EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only I

Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier#

Group# 2 K/A# 028 2.4.21 Importance Rating 4.6 Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core coolin g and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Proposed Question: SRO Question # 90 Given the following conditions:

Unit 3 is at 100% power.

RVLMS is out of service.

Subsequently:

The unit trips.

Pressurizer Level Transmitter LT-3-460 fails low on the trip.

The crew transitions to 3-EOP-ES-0.1, Reactor Trip Respo nse.

Pressurizer level has rapidly risen to 93% and is oscillating.

CET temperatures are 608°F.

RCS pressure is 1510 psig.

Which ONE of the following identifies the applicable functio nal recovery procedure, if any?

REFERENCE PROVIDED A. 3-EOP-FR-l.1 B. 3-EOP-FR-I.2 C. 3-EOP-FR-l.3 D. Inventory critical safety functions are all satisfied.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible if examinee reads the triggers correctly, but does not recognize that head voids are indicated (e.g., believes that the high Pressurizer level is due to the failed level transmitter and resulting letdown isolation).

B. Incorrect. Plausible if examinee incorrectly reads the first Pressurizer-le vel trigger as 92% (instead of 92%) and the second as 14% (instead of 14).

C. Correct. With Pressurizer level >92%, all RCPs running, and an indication of head voiding, 3-EOP-FR-l.3 is applicable.

D. Incorrect. Plausible if examinee incorrectly reads the first Pressurizer-le vel trigger as 92% (instead of 92%), reads the second Pressurizer-level trigger correctly, selects RCPs on or off, and does not recognize that head voids are indicat ed (e.g., believes that the high Pressurizer level is due to the failed level transmitter and resulting letdown isolation).

Technical Reference(s): 3-EOP-F-0 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

Yes Learning Objective: PTN 6900922, Obj. 18 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-]. DRAFT NRC EXAM SECURE INFORMATION 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

SRO Question # 90 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1 immediate operator actions? I Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier#

Group# 2 K/A# 068 AA2.10 Importance Rating 4.4 Ability to determine and interpret the following as they apply to the Contro l Room Evacuation:

Source range count rate Proposed Question: SRO Question # 91 Given the following conditions:

Units 3 and 4 are at 100% power.

SM orders a Control Room evacuation due to smoke in the Control Room.

A controlled cooldown has been initiated from the Alternate Shutdown Panel.

Subsequently the following indications are observed during the cool down:

Corrected Steam Generator water levels are 22, 37, and 42%.

Corrected Pressurizer water level is 51%.

Source Range counts are 1000 cps and rising.

Steam Generator pressures are 900, 950, and 1005 psig.

Which ONE of the following describes (1) why the cool down should be stoppe d, and (2) what is the MINIMUM Emergency Classification for this event?

A. (1) A steam generator level is less than 27%.

(2) Alert B. (1) Pressurizer level is greater than 50%.

(2) Site Area Emergency.

C. (1) Source range counts are rising.

(2) Alert D. (1) The pressure difference between SJGs exceeds 100 psi.

(2) Site Area Emergency.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible because examinee may believe that a Steam Genera tor narrow-range level <27% represents a loss of heat sink (which would be correct

, if all Steam Generator narrow-range levels were less than this value and adverse contain ment conditions existed); however, if corrected Steam Generator levels on the ASP are maintained between 20% and 50%, the cooldown may proceed (per 0-ONO P-105, Attachment 22). The second part of the distractor is correct.

B. Incorrect. Plausible because examinee may believe that Pressurizer level is too high for the given (i.e., no-load) conditions, for which Pressurizer level would be 22%;

however, if corrected Pressurizer level on the ASP is maintained between 22% and 53%, the cooldown may proceed (per 0-ONOP-105, Attachment 22). The second part of the distractor is plausible if examinee misinterprets the question and choose s the MAXIMUM Emergency Classification for the event, which requires that plant control is not established within 15 minutes.

C. Correct. A Note in 0-ONOP-105, Attachment 22 states that the cooldown should be stopped if source-range counts show jy rise. The minimum classification for any Control Room evacuation is an Alert.

D. Incorrect. Plausible because examinee may believe that the mismatched Steam Generator pressures are indicative of a faulted Steam Generator, which could lead to an uncontrolled cooldown condition; however, if Steam Generator pressures are maintained within <120 psid, the cooldown may proceed (per 0-ONOP-105, Attachment 22). The second part of the distractor is plausible if examinee misinterprets the question and chooses the MAXIMUM Emergency Classification for the event, which requires that plant control is not established within 15 minutes.

Technical Reference(s): 0-ONOP-105, Attachment 22 (Attach if not previously provided)

EPIP-20101, Attachment 1 Proposed References to be provided to applicants during examination:

N Learning Objective: PTN 6900160, Obj. 5, PTN 3200003, Obj. 4 (As available)

Question Source: Bank # 92708 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-]. DRAFT NRC EXAM SECURE INFORMATION Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 Harris Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE:

PAGE:

11 CONTROL ROOM EVACUATION PROCEDURE NO.: 162 of 221 O-ONOP-105 TURKEY POINT PLANT ATTACHMENT 22 Unit 3 Cool Down from ASP (Page 2 of 23)

I STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED NOTE

CRDM Fans are NOT protected for Alternate Shutdown and may NOT be available due to fire damage.

Local control of CRDM Fans is NOT available. TSC staff shall determine how to operate CRDM Fans based on current plant conditions.

2. ENSURE all available CRDM Fans RUNNING.

NOTE To prevent uneven RCS temperature distributions, the pressure difference between SIGs shall be minimized. This ensures decay heat remov al is evenly distributed to each active coolant loop.

The cool down should be STOPPED if source range counts show py rise.

Makeup water sources for CST will be necessary if level lowers to less than 5 feet.

3. INITIATE RCS cool down to COLD SHUTDOWN:

A. MAINTAIN cool down rate in RCS Cold Legs less than 25°F/hr.

B. DUMP steam to atmosphere.

C. MAINTAIN pressure difference between S/Gs less than 120 psid.

D. Using curve on ASP, MAINTAIN Corrected S/G Level between 20% and 50%.

REVISION NO.: PROCEDURE TITLE:

PAGE:

CONTROL ROOM EVACUATION PROCEDURE NO.: 163 of 221 O-ONOP-105 TURKEY POINT PLANT ATTACHMENT 22 Unit 3 Cool Down from ASP (Page 3 of 23)

I STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I E. Using curve on ASP, MAINTAIN Corrected Pressurizer Level between 22% and 53%.

F. MAINTAIN RCS temperature and pressure to the of the 60°F/hr Cooldown Curve.

4. CHECK RCS Hot Leg Temperatures less RETURN TO Attachment 22, Step 3.

than 550°F.

5. DEPRESSURIZE RCS to 1950 psig as follows:

DE-ENERGIZE a Pressurizer Heaters.

ADJUST PRZ level as necessary.

6. CHECK Outside SNPO BLOCKED SI per WHEN SI is BLOCKED, THEN CONTINUE Attachment 18, Step 4 and with Attachment 22, Step 7.

Attachment 18, Step 6.

7. INITIATE RCS depressurization:

A. CHECK at least one CRDM Fan GO TO Attachment 22, Step 8.

RUNNING.

B. GO TO Attachment 22, Step 15.

SRO Question # 91 Clarification Guidance for SRO-only Questions Rev 1 (03/1112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1 immediate operator actions? I Yes RO question Can the questio n be answered solely by knowing entry conditions that require directentry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or eto overall mitigative strategy of a 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 LonlJ FOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 2 K!A# E07 2.4.6 Importance Rating 4.7 Emergency Procedures / Plan: Knowledge of EOP mitigation strategies.

Proposed Question: SRO Question # 92 Given the following conditions:

A Unit 4 Steam Generator Tube Rupture occurs.

Due to equipment failures, the crew is performing actions contained in 4-EOP-ECA-3.2, SGTR With Loss Of Reactor Coolant Saturated Recovery Desired.

All CSF Status Trees are GREEN with the exception of the following:

o Core Cooling YELLOW path for 4-EOP-FR-C.3, Response To Saturated Core Cooling o Inventory YELLOW path for 4-EOP-FR-l.2, Response To Low Pressurizer Level Which ONE of the following describes the required implementation of proced ures for this event, and the reason?

A. Transition from 4-EOP-ECA-3.2 to 4-EOP-FR-I.2 to restore the Inventory CSF to a GREEN condition B. Transition from 4-EOP-ECA-3.2 to 4-EOP-FR-C.3 to restore the Core Cooling CSF to a GREEN condition C. Remain in 4-EOP-ECA-3.2. The actions contained in 4-EOP-FR-C.3 and 4-EOP-FR-I.2 conflict with 4-EOP-ECA-3.2 actions D. Remain in 4-EOP-ECA-3.2. Implementation of YELLOW Path procedures is not allowed when using Emergency Contingency Action procedures.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible because the Inventory CSF is not satisfied and, lAW 4-EOP-F-O (Critical Safety Function Status Trees), entry into the YELLOW path functional restoration procedure is at the operators discretion; however, 4-EOP-FR-l.2 directs actions that will conflict with those in 4-EOP-ECA-3.2 and a Caution at the beginning of this procedure states that it is not to be performed, if 4-EOP-ECA-3.2 is in effect.

B. Incorrect. Plausible because the Core Cooling CSF is not satisfied and, lAW 4-EOP-F-o (Critical Safety Function Status Trees), entry into the YELLOW path functional restoration procedure is at the operators discretion; however, 4-EOP-FR-C.3 directs actions that will conflict with those in 4-EOP-ECA-3.2 and a Note at the beginning of this procedure states that it is not to be performed, if 4-EOP-ECA-3.2 is in effect.

C. Correct. 4-EOP-FR-C.3 and 4-EOP-FR-l.2 direct actions that will conflict with those in 4-EOP-ECA-3.2; a Note/Caution at the beginning of these procedures states that they are not to be performed, if 4-EOP-ECA-3.2 is in effect.

D. Incorrect. Plausible if examinee confuses 4-EOP-ECA-3.2 with 4-EOP-ECA-O.O, which states that CSF Status Trees are for information only and transitions to the Functional Restoration Procedures are not allowed. However, per guidance of 4-EOP-F-O, entry into the YELLOW path Functional Restoration Procedure is at the operators discretion and not prohibited in this case.

Technical Reference(s): 4-EOP-FR-C.3 4-EOP-FR-l.2, 4-EOP-F-O 4-EOP-ECA-O.o (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: LP 6900344, Obj. 4 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2011 Ginna PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Various WTSI Bank questions support this topic PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

2 RESPONSE TO SATURATED CORE COOLING 5 of 9 PROCEDURE NO.:

4-EOP-FR-C.3 TURKEY POINT UNIT 4 I STEP II ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED I 3.0 OPERATOR ACTIONS CAUTION If RWST level decreases to less than 155,000 gallons, AND SI system is in RWST Injection alignment, then the SI System shall be aligned for Cold Leg Recirculation using 4-EOP-ES-1 .3, TRANSFER TO COLD LEG RECIRCULATION.

NOTE

If 4-EOP-ECA-3.2, SGTR WITH LOSS OF REACTOR COOLANT-SATURATED RECOVERY DESIRED, is in effect, this procedure shall NOT be performed.

Foldout Page is required to be monitored throughout this procedure.

1. Check RHR System ALIGNED FOR Perform 4-ONOP-050, LOSS OF RHR, INJECTION OR RECIRCULATION concurrently to establish required alignment.

2. Verify SI Flow:

a. Check SI System

a. Verity one RHR pump running.

ALIGNED FOR RWST INJECTION Go to Step 2.d.

b. RCSpressure b. GotoStep2.d.

LESS THAN 275 PSIG [575 PSIG]

c. RHR Pump flow indicator c. Start RHR Pumps and align valves to CHECK FOR FLOW establish RHR flow.

d. High-Head SI Pump flow indicator d. Start pumps and align valves to CHECK FOR FLOW establish SI flow.

REVISION NO.: PROCEDURE TITLE:

PAGE:

LOSS OF ALL AC POWER 5 Of 89 PROCEDURE NO.:

4-EOP-ECA-O.O TURKEY POINT UNIT 4 I STEP fl ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED I 3.0 OPERATOR ACTIONS NOTE

Step 1 and Step 2 are IMMEDIATE ACTION steps.

CSF Status Trees are required to be monitored for Information Only.

FRPs shall NOT be implemented.

1. Verify Reactor Trip: Manually trip Reactor.

Rod Bottom Lights ON

Reactor Trip and Bypass Breakers OPEN

Rod Position Indicators AT ZERO

Neutron flux DECREASING

REVISION NO.: PROCEDURE TITLE: PAGE:

CRITICAL SAFETY FUNCTION STATUS TREES 17 of 19 PROCEDURE NO.:

4-EOP-F-O TURKEY POINT UNIT 4 ATTACHMENT I Rules of Usage for Critical Safety Function Status Trees (Page 1 of 2)

I. Critical Safety Function Status Trees shall be monitored in the following order of priority:

a. Subcriticality using ENCLOSURE 1

b. Core Cooling using ENCLOSURE 2

c. Heat Sink using ENCLOSURE 3

d. Integrity using ENCLOSURE 4

e. Containment using ENCLOSURE 5

f. Inventory:

1) H RVLMS in service, THEN using ENCLOSURE 6.

2) U RVLMS NOT in service, THEN using ENCLOSURE 7.

NOTE NOTES or CAUTIONS within EOPs which prohibit the use of Functional Restoration Procedures shall take precedence over the following rules.

2. IF an Extreme Challenge (RED PATH) is diagnosed, THEN the operator shall immediately stop procedure in effect and initiate functional restoration to restore the Critical Safety Function under Extreme Challenge.

3. IF a Severe Challenge (ORANGE PATH) is diagnosed, THEN the operator shall continue to check the status of all remaining Critical Safety Functions. IF NO extreme challenges exist, THEN the operator shall stop procedure in effect and initiate functional restoration to restore the highest priority Critical Safety Function under Severe Challenge.

4. N a NOT Satisfied Condition (YELLOW PATH) is diagnosed, THEN the operator shall continue to check the status of all remaining Critical Safety Functions. if NO Extreme or Severe Challenges exist, THEN it is the operators option to continue Optimal Recovery Procedures or to initiate functional restoration of the Critical Safety Function NOT satisfied.

5. IF a Satisfied Condition (GREEN PATH) is diagnosed, THEN NO challenge exists for the affected Critical Safety Function and the operator shall continue to check the status of all remaining Critical Safety Functions.

REVSON NO.: PROCEDURE TITLE:

PAGE:

2 RESPONSE TO LOW PRESSURIZER LEVEL PROCEDURE NO.: 5 of 13 4-EOP-FR-I.2 TURKEY POINT UNIT 4 ISTEP IIACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 3.0 OPERATOR ACTIONS CAUTION This procedure SHALL NOT be performed if gj of the following is in effect:

4-EOP-ECA-1 .1, LOSS OF EMERGENCY COOLANT RECIRCULATION

4-EOP-ECA-3.2, SGTR WITH LOSS OF REACTOR COOLANT -

SATU RATED RECOVERY DESIRED

4-EOP-ECA-3.3, SGTR WITHOUT PRESSURIZER PRESSURE CONTROL NOTE Foldout page is required be monitored throughout this procedure.

1. Check If SI Has Been Terminated

a. High-Head SI Pumps

a. Return to procedure and step in effect.

ALL STOPPED

2. Verify Letdown ISOLATED Manually isolate Letdown.

Normal Letdown

Excess Letdown

3. Verify Instrument Air To Containment:

a. CV-4-2803, Instrument Air Containment Isolation OPEN

b. P1-4-1444, Instrument Air Pressure b. Restore Instrument Air pressure using GREATER THAN 95 PSIG O-ONOP-013, LOSS OF INSTRUMENT AIR, while continuing with this procedure.

SRO Question #92 Clarification Guidance for SRO-only Questions Rev 1(03/1112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1 immediate operator actions? Yes RO question r

9 Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the questio n be answered solely by knowi the purpose, overall mitigative strategy of a 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 FOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-154 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 K/A# ElO EA2.2 Importance Rating 3.9 Ability to determine and interpret the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS) Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Proposed Question: SRO Question # 93 Given the following conditions:

Unit 3 is performing a natural circulation cooldown lAW 3-EOP-ES-0.2, Natural Circulation Cooldown.

RCS cold leg is 535°F.

RCS pressure is 1870 psig.

3A and 3B CRDM fans are running.

RCPs 3A, 3B, and 3C are tripped and cannot be restarted.

AFWflowis400gpm.

CST level is 8%.

PZR level is 65% and rising unexpectedly.

RVLMS is not available.

Which one of the following identifies (1) the procedure and (2) the maximum permitted cooldown rate to lower RCS Hot Leg Temperatures to 500° F?

NOTE

3-EOP-ES-0.2, Natural Circulation Cooldown.

3-EOP-ES-0.4, Natural Circulation Cooldown With Steam Void In Vessel (Without RVLMS).

A. (1) Continue with 3-EOP-ES-0.2.

(2) The maximum cooldown rate is <50°F/hr.

B. (1) Continue with 3-EOP-ES-0.2.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION (2) The maximum cooldown rate is <25°F/hr.

C. (1) Transition to 3-EOP-ES-0.4.

(2) The maximum cooldown rate is <50°F/hr.

D. (1) Transition to 3-EOP-ES-0.4.

(2) The maximum cooldown rate is <25°F/hr.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible because 3-EOP-ES-0.2 could be considered the preferred procedure, since a cooldown is already in progress (in this case, the maximum allowed cooldown rate would be 1 00°F/hr); however, an unexpected rise in Pressurizer level is indicative of void formation in the reactor head and 3-EOP-ES-0.2 directs transition to 3-EOP-ES-0.4 if, at any time, it is determined that the Natural Circulation Cooldown and depressurization must be performed at a rate that may form a steam void in the vessel (with RVLMS unavailable). The second part of the distractor is incorrect lAW 3-EOP-ES-0.2, but correct lAW 3-EOP-ES-0.4.

B. Incorrect. First part is plausible because 3-EOP-ES-0.2 could be considered the preferred procedure, since a cooldown is already in progress; however, an unexpected rise in Pressurizer level is indicative of void formation in the reactor head and 3-EOP-ES-0.2 directs transition to 3-EOP-ES-0.4 if, at any time, it is determined that the Natural Circulation Cooldown and depressurization must be performed at a rate that may form a steam void in the vessel (with RVLMS unavailable). Second part is plausible because a cooldown rate of <25°F/hr was permitted by the previous version of 3-EOP-ES-0.2.

C. Correct. 3-EOP-ES-0.2 directs transition to 3-EOP-ES-0.4 if, at any time, it is determined that the Natural Circulation Cooldown and depressurization must be performed at a rate that may form a steam void in the vessel (with RVLMS unavailable);

an unexpected rise in Pressurizer level is indicative of void formation in the reactor head. The maximum allowed cooldown rate is 50°F/hr in 3-EOP-ES-0.4, while RCS temperature is being reduced to 500°F.

D. Incorrect. The first part of the distractor is correct. Second part is plausible because a

cooldown rate of <25°F/hr was permitted by the previous version of 3-EOP-ES-0.2.

Technical Reference(s): 3-EOP-ES-0.2, 3-EOP-ES-0.4 (Attach if not previously provided)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed References to be provided to applicants during examination: No Learning Objective: PTN 6900324 Obj. 6; PTN 6900326, Obj. 4 (As available)

Question Source: Bank # 91583 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 Farley Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

2 NATURAL CIRCULATION COOLDOWN PROCEDURE NO.: WITH STEAM VOID IN VESSEL (WITHOUT RVLMS) of 36 3-EOP-ES-O.4 TURKEY POINT UNIT 3 I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED CAUTION If another procedure has reduced RCS pressure to less than 1937 psig, pressure is required to be held constant uritU directed to depressurize below the existing RCS pressure.

3. Decrease RCS Hot Leg Temperatures To 500°F:

a. Check RCS Hot Leg temperatures a. Go to Step 4.

GREATER THAN 500°F

b. Maintain cooldown rate in RCS Cold Legs LESS THAN 50°F/HR

c. Maintain RCS pressure BETWEEN 1900 PSIG AND 1937 PSIG

d. Maintain RCS temperature and pressure TO THE RIGHT OF THE TECHNICAL SPECIFICATION 100°F/HR COOLDOWN CURVE

e. Maintain stable PRZ level using Charging

f. Check RCS Hot Leg temperatures f. Return to Step 3.b.

LESS THAN 500°F

g. Stop RCS cooldown

REVISION NO.: PROCEDURE TITLE: PAGE:

2 NATURAL CIRCULATION COOLDOWN 21 of 69 PROCEDURE NO.:

3-EOP-ES-O.2 TURKEY POINT UNIT 3 STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I NOTE If at any time it is determined that the Natural Circulation 000ldown and depressurization must be performed at a rate that may form a steam void in the vessel, procedure 3-EOP-ES-0.3, NATURAL CIRCULATION 000LDOWN WITH STEAM VOID IN VESSEL (WITH RVLMS) or 3-EOP-ES-0.4, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL (WITHOUT RVLMS), shall be used.

15. Establish Required RCS Subcooling:

a. Check RCS pressure a. Maintain RCS pressure stable.

GREATER THAN 1600 PSIG Go to Step 15.c.

b. Maintain RCS pressure between 1600 psig and 1937 psi 9 untH subcooling in Step 15.c is established

c. Establish RCS Subcooling based on Core Exit TCs as follows:

With one OR more CRDM Cooler Fans running GREATER THAN 144°F OR With NO CRDM Cooler Fan running GREATER THAN 229°F

d. Check RCS Subcooling based on d. Return to Step 15.b.

CETs GREATER THAN THAT REQUIRED IN Step 15.c.

SRO Question # 93 Clarification Guidance for SRO-only Questions I

Rev 1 (03111/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? question Can the question be answered solely by knowing 1 immediate operator actions? Yes I RO question Ii Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a 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 only

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

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group K/A# Gi 2.1.29 Importance Rating 4.0 Conduct of Operations: Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

Proposed Question: SRO Question # 94 Given the following conditions:

Unit 3 is in MODE 3 following a refueling outage.

3-OSP-202.1 Safety Injection/Residual Heat Removal Flowpath Verification, Attachment 2

is in progress.

Prior to exceeding 380°F, the Shift Manager is verifying that the preceding steps of 3-GOP-503, Cold Shutdown to Hot Standby, are complete prior to signing the Shift Manager Verification Point.

Which ONE of the following identifies if subsequent 3-GOP-503 procedure steps, past the Shift Manager Verification Point, may be performed?

A. Steps may NOT be performed until 3-OSP-202.1, Attachment 2 is complete.

B. Steps may NOT be performed without the Operations Director concurrence.

C. Steps may be performed if it will not heat the RCS temperature above 380°F.

D. Steps may be performed if it does not affect the SI/RHR Flowpath verification.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible because the Shift Manager Verification Point in 3-GOP-503 (Section 5.31) requires verification that all prior Section 5.0 steps are complete:

however, this hold point does NOT prevent the performance of subsequent procedure steps that will not heat the RCS above 380°F. The examinee may assume that the OSP PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION attachment must be completed, with the verification point signed by the Shift Manager, prior to proceeding.

B. Incorrect. Plausible because the Shift Manager Verification Point in 3-GOP-503 (Section 5.31) requires verification that all prior Section 5.0 steps are complete; however, this hold point does NOT prevent the performance of subsequent procedure steps that will not heat the RCS above 380°F. The examinee may assume that the verification point must signed by the Shift Manager, prior to proceeding.

C. Correct. A note in the Shift Manager Verification Point in 3-GOP-503 (Section 5.31) states that this hold point does NOT prevent the performance of subsequent procedure steps, providing that they will not heat the RCS above 380°F.

D. Incorrect. Plausible because the Shift Manager Verification Point in 3-GOP-503 (Section 5.31) does NOT prevent the performance of subsequent procedure steps, providing that they will not heat the RCS above 380°F. The examinee may falsely assume that the verification point also requires that the SI/RHR flowpath verification (including Control Room switch alignments) is complete, to ensure that the requirements for TS 3.5.2 are met prior to exceeding 380°F; specifically, 0-ADM-536 (Technical Specification Bases Control program) states that the OPERABILITY of ECCS components and flowpaths are required in MODES 1, 2, and 3, to ensure sufficient emergency core cooling capability is available, in the event of a LOCA with any single active failure.

3-GOP-503, 0-ADM-536, Technical Reference(s): Technical Specifications (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: PTN 6900408, Obj.5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

Procedure No: Procedure TitIc Page 61 Approval Date 3-GOP-503 Cold Shutdown to Hot Standby 2/28/14 QA RECORD PAGE INITIALS CKD VERIF 5.30.3 Maximize Pressurizer heaters AND reduce Pressurizer spray flow as required to establish desired RCS pressurization rate.

5.31 Prior to increasing RCS temperature above 38(1°F, verifS that all prerequisites signed off in Subsection 3.3. have been 5.31.1 SHIFT MANAGER VERIFICATION POINT

1. Prior to exceeding 38Q F in the RCS, verify all Section 5.0 steps prior to this verification are complete, and the conditions of this verification point are satisfied. fCommitment Step 2.3.18 CAPRJ NOTE: This hold point does NOT prevent the performance of subsequent procedure steps which will NOT heat the RCS above 380 F.

Shift Manager Signature Print Date NOTE I Letdown Flow (Fl-3-150) will increase as RCS pressure increases.

I 5.32 Prior to reaching 120 gprn letdown flow, adjust Low Pressure Letdown Flow Controller, PC-3-145, to 275 psig.

5.33 WHEN RCS Pressure is between 700 psig and 1000 psig. THEN perform the following:

5.33.1 Unlock AND Close Breaker 30532. MOV-3-865A.

5.33.2 Unlock AND Close Breaker 30631, MOV-3-865B.

5.33.3 Unlock AND Close Breaker 30733, MOV-3-865C.

5.33.4 Open SI Accumulator Isol, MOV-3-865A.

tAiQ7iTMFA1I-frkh

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS Tavg GREATER THAN OR EQUAL TO 350°F LIMITING CONDITION FOR OPERATION 3.5.2 The following Emergency Core Cooling System (ECCS) equipment and flow paths shall be OPERABLE:

a. Four OPERABLE SafetX Injection (SI) pumps, each capable of being powered from its associated OPERABLE diesel generator#, with discharge aligned to the RCS cold legs,*

b. Two OPERABLE RHR heat exchangers,

c. Two OPERABLE RHR pumps with discharge aligned to the RCS cold legs,

d. An OPERABLE flow path capable of taking suction from the refueling water storage tank as defined in Specification 3.5.4, and

e. Two OPERABLE flow paths capable of taking suction from the containment sump.

APPLICABILITY: MODES 1, 2, and 3***

ACTION:

a. With any one of the required ECCS components or flow paths inoperable, except for inoperable Safety Injection Pump(s) or an inoperable RHR pump, restore the inoperable compo nent or flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. In the event the ECCS is actuated and injects water in the Reactor Coolant System

, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuati on cycles to date since January 1, 1990.

c. With one of the four required Safety Injection pumps inoperable and the opposi te unit in MODE 1, 2, or 3, restore the pump to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .***

Only three OPERABLE Safety Injection (SI) pumps (two associated with the unit and one from the opposite unit),

each capable of being powered from its associated OPERABLE diesel generator#,

with discharge aligned to the RCS cold leg are required if the opposite unit is in MODE 4, 5, or 6.

- The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the Safety Injection flow paths isolated pursuant to Specification 3.4.9.3 provided that the Safety Injection flow paths are restored to OPERABLE status prior to Tavg exceeding 380°F. Safety Injection flow paths may be isolated when Tavg is less than 380°F.

- - The provisions of Specifications 3.0.4 and 4.0.4 are not applicable.

lnoperability of the required EDGs does not constitute inoperability of the associated Safety Injection pumps.

TURKEY POINT UNITS 3 & 4 3/4 5-3 AMENDMENT NOS. 212 AND 206

REVISION NO.: PROCEDURE TITLE: PAGE:

11 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 121 of 192 PROCEDURE NO.:

O-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 105 of 176) 3/4.5.2 3/4.5.3 ECCS Subsystems The OPERABILITY of ECCS components and flowpaths required in Modes 1, 2 and 3 ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming any single active failure consideration. Two SI Pumps and one RHR Pump operating in conjunction with two Accumulators are capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all pipe break sizes up to and including the maximum hypothetical accident of a circumferential rupture of a reactor coolant loop. In addition, the RHR subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.

CAUTION Interim Compensatory Measure The provision for temporary restoration of power to locked ECCS valves listed in SR 4.5.2.a for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit temporary operation for surveillance and maintenance purposes has been determined to be non-conservative with respect to the safety analysis. Therefore, until appropriate changes to SR 4.5.2 via LAR 212 are approved and implemented, restoration of power to the valves listed in SR 4.5.2.a shall be limited to one hour in order to provide alternative valve position indication in the event that the continuous valve position indication (amber light) in the control room is unavailable. Ref AR 1811016 Motor Operated Valves (MOVs) 862A, 862B, 863A, 863B are required to take suction from the containment sump via the RHR System.

PC-600 supplies controlling signals to valves MOVs 862B and 863B, to prevent opening these valves if RHR Pump B discharge pressure is above 210 psig. P0-601 provides similar functions to valves MOVs 862A and 863A. Although all four valves are normally locked in position, with power removed, the capability to power up and stroke the valves must be maintained in order to satisfy the requirements for OPERABLE flow paths (capable of taking suction from the containment sump).

SRO Question #94 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

immediate operator actions?

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

9 Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page8ofl6

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 1 K/A# G1 2.1.9 Importance Rating 4.5 Conduct of Operations: Ability to direct personnel activities inside the control room.

Proposed Question: SRO Question # 95 Given the following conditions:

Unit3isinMODE2.

A Reactor startup is in progress.

The person providing reactivity oversight for the startup in the Control Room becomes ill and must be relieved.

Which ONE of the following identifies (1) who can be directed to relieve the person performing reactivity oversight in the Control Room and (2) who is the highest authority over the startup?

A. (1) Unit Supervisor (2) Shift Manager B. (1) WCC Supervisor SRO qualified (2) Shift Manager C. (1) Unit Supervisor (2) Reactivity Oversight SRO D. (1) WCC Supervisor SRO qualified (2) Reactivity Oversight SRO Proposed Answer: B PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Explanation (Optional):

A. Incorrect. The SRO-qualified individual that provides continuous oversight of reactiv ity manipulations may have no other assigned duties, as does the Unit Supervisor; the second part of the distractor is correct.

B. Correct. The individual that provides continuous oversight of reactivity manipulation s

must be SRO-qualified and have no other assigned duties (which exclude the Unit Supervisor); the Shift Manager provides oversight of the reactivity oversight SRO.

C. incorrect. The SRO-qualified individual that provides continuous oversight of reactiv ity manipulations may have jj other assigned duties, as does the Unit Supervisor; the Shift Manager provides oversight of the reactivity oversight SRO and, hence, is the higher authority.

D. Incorrect. The individual that provides continuous oversight of reactivity manipulation s

must be SRO-qualified and have no other assigned duties (which exclude the Unit Supervisor); the Shift Manager provides oversight of the reactivity oversight SRO and, hence, is the higher authority.

Technical Reference(s): O-ADM-200 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: PTN 6900043, Obj. 3 (As available)

Question Source: Bank # 99573 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Braidwood Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Corn ments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

21 OPERATIONS MANAGEMENT MANUAL 30 of 76 PROCEDURE NO.:

0-ADM-200 TURKEY POINT PLANT 4.4.3 Reactivity Control (continued)

2. (continued)

B. The following are requirements for controlling reactivity and providing oversight during low power operation, including reactor startup, reactor operation in the intermediate range or low in the power range, turbine roll, generator synchronization, operation with feedwater flow control valves in manual, and operation below 20% power:

A single Senior Reactor Operator, with NO other duties assigned, shall direct and provide continuous oversight for all reactivity manipulations. The reactivity oversight Senior Reactor Operator shall be responsible to the assigned Unit Supervisor during low power operations.

One or more Reactor Controls Operators, with NO other duties assigned, shall conduct all reactivity manipulations on the unit.

Reactivity manipulations, in this case, include RCS boration/dilution control, control rod motion, steam generator level control, turbine control, and main generator load control.

Licensed operators shall monitor reactor power on diverse indications (Analog PRNI, Analog IRNI, Plant Computer System PRNI, Plant Computer System IRNI).

The reactivity management Senior Reactor Operator shall set specific power bands on Power Range Nuclear Instruments and Intermediate Range Nuclear Instruments.

Reactor Controls Operator(s) performing reactivity manipulations shall receive a peer check from a licensed operator.

The Unit Supervisor and Shift Manager shall provide independent oversight of reactivity manipulations by reactivity oversight SRO and Reactor Controls Operators.

The Operations Manager and Assistant Operations Managers shall conduct periodic observations of reactivity management in the Control Room during low power operations.

SRO Question # 95 Clarification Guidance for SRO-only Questions Rev 1 (0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

immediate operator actions?

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters question that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or question overall mitigative strategy of a 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 FOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No j

I Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 2 K/A# G2 2.2.2 Importance Rating 4.1 Equipment Control: Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

Proposed Question: SRO Question # 96 Given the following conditions:

Unit 4 is in MODE 2, preparing to startup to MODE 1.

The 4C Charging pump is out of service for maintenance.

The Unit Supervisor is reviewing the MODE Change Report. The quarterly Charging Pump surveillance history is as follows:

o 4-OSP-047.1A, Charging Pump 4A Group A Pump Test, was completed 8 days ago.

o 4-OSP-047.1B, Charging Pump 4B Group A Pump Test, was completed 118 days ago.

o 4-OSP-047.1C, Charging Pump 4C Group A Pump Test, was completed 79 days ago.

Which ONE of the following identifies the maximum allowable time to perform the surveillance for the 4B Charging Pump, before declaring the pump inoperable?

A. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months D. One quarter Proposed Answer: D Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-1S-1 DRAFT NRC EXAM SECURE INFORMATION A. Incorrect. Plausible if examinee incorrectly believes that two pumps are not OPERABLE (since the 4B Charging Pump is beyond its normal surveillance interval and grace period) and, hence, TS 3.0.3 is applicable.

B. Incorrect. Plausible if examinee recognizes that the 4B Charging Pump is beyond its normal surveillance interval and TS 4.0.3 applies, but incorrectly identifies 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as the allowable surveillance delay period (rather than the longer period of one quarter).

C. Incorrect. Plausible if examinee believes that, since the 4B Charging Pump is beyond its normal surveillance interval and grace period, the pump is not OPERABLE. Thus, with only one OPERABLE Charging Pump (4A), a second pump will need to be returned to OPERABILITY within 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months .

D. Correct. The 4B Charging Pump is outside its quarterly surveillance schedule, which was completed 118 days ago. However, TS 4.0.3 allows for a delay of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified frequency, whichever is greater (i.e., one quarter). Therefore, this pump is OPERABLE. With the 4C Charging Pump OOS and one other Charging Pump (3A) still OPERABLE, entry into TS LCD 30.3 is NOT required.

Technical Reference(s): PTN Technical Specifications (Attach if not previously provided)

TS 3.0.3, 3.1.2.3, 4.0.2 4.0.3 Proposed References to be provided to applicants during examination:

Learning Objective: LP 6900234, Dbj.4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L1S-1 DRAFT NRC EXAM SECURE INFORMATION PTN L-1S-1 DRAFT NRC EXAM SECURE INFORMATION

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITIONS FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specification s is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met, except as provided in Specification 3.0.6.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals, except as provided in Specification 3.0.6. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initiated to place the unit, as applicable, in:

a. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

c. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements

, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.

This specification is not applicable in MODES 5 or 6.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Conditions for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATION AL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.

TURKEY POINT UNITS 3 & 4 3/4 0-1 AMENDMENT NOS. 235 AND 230

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.3 At least two charging pumps shall be OPERABLE.

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

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months or be in at least HOT STANDBY and borated to a boron concentration equivalent to at least the required SHUTDOWN MARGIN at COLD SHUTDOWN at 200°F within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUI REMENTS 4.1.2.3.1 The required charging pumps shall be demonstrated OPERABLE by testing pursuant to Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 and 4.

TURKEY POINT UNITS 3 & 4 3/4 1-10 AMENDMENT NOS. 260 AND 255

APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Limiting Condition for Operation. Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval. If an ACTION item 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.

4.0.3 If it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the Limiting Condition of Operation not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the surveillance is not performed within the delay period, the Limiting Condition of Operation must immediately be declared not met, and the applicable ACTION(s) must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition of Operation must immediately be declared not met, and the applicable ACTION(s) must be entered.

4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement(s) associated with a Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a.

Inservice testing of ASME Code Classi, 2, and 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50, Section 50.55a.

TURKEY POINT UNITS 3 & 4 3/4 0-3 AMENDMENT NOS. 225 AND 220

SRO Question # 96 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing I hour TSITRM Action?

Can question be answered solely by knowing the LCOITRM information listed above-the-line?

Can question 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)

SRO-only

Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 2 K/A # G2 2.2.43 Importance Rating 3.3 Equipment Control: Knowledge of the process used to track inoperable alarms.

Proposed Question: SRO Question # 97 Given the following conditions:

The plant is in MODE 3.

A problem with a CCW flow switch is causing an intermittent alarm for a CCW flow annunciator.

The alarm is determined to be a nuisance alarm.

Which ONE of the following describes the LOWEST level of authority required to (1) remove the annunciator from service and (2) the administrative process required?

A. (1) Shift Manager (2) A Work Request must be generated and the disabled alarm entered into the Annunciator Status Log and filed in the Equipment Out of Service Index.

B. (1) Shift Manager (2) If removal of the annunciator is controlled by a procedure, an 005 sticker is sufficient to document the disabled alarm; NO Equipment Out of Service Index entry is required.

C. (1) Operations Manager (2) A Work Request must be generated and the disabled alarm entered into the Annunciator Status Log and filed in the Equipment Out of Service Index.

D. (1) Operations Manager (2) If removal of the annunciator is controlled by a procedure, an DOS sticker is sufficient to document the disabled alarm; NO Equipment Out of Service Index entry is required.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: A Explanation (Optional):

A. Correct.

B. Incorrect. Plausible because an 005 sticker LY be used in this situation and examinee may confuse the requirement for procedural control to alleviate a work request with relaxation of the requirement to make an entry in the EOOS log; first part of distractor is correct.

C. Incorrect. Plausible because the Operations Manager may be involved in the process, but does not have responsibility for its approval; second part of distractor is correct D. Incorrect. See explanations for distractors B and C.

Technical Reference(s): OP-AA-100-1000,At115 .

0-OSP-200.5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: PTN 6900025, Obj. 2 (As available)

Question Source: Bank # 98338 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Wolf Creek Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

14 CONDUCT OF OPERATIONS 96 of 99 PROCEDURE NO.:

OP-AA-100-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 15 ANNUNCIATOR AND COMPUTER POINT CONTROL (Page 1 of 4) (Section 6.1.3, Voluntary Action Item 7) 1.0 PURPOSE Alarms provide early warning of process trends or equipment malfunction. Proper use of alarms is essential to resolve potential issues before they further degrade. A fundamental principle of safe operation is that indications must be believed.

Operators must be able to trust their indications and act appropriately when an alarming condition is received. A rigorous process is utilized to remove annunciators or computer points from service ensuring appropriate compensatory measures are established and proper reviews are performed.

2.0 STANDARD The bypassing, or removal from service, of annunciators and computer points will be reviewed and authorized by an SRO, typically the CRS or Shift Manager. A tracking mechanism will provide Operator awareness and form the basis for periodic reviews.

Compensatory actions and monitoring will be put in place when needed. These controls specifically include local annunciators and points outside the control room.

3.0 EXPECTATIONS

1. All alarms in the Control Room as well as in the field shall be treated as valid until proven otherwise.

2. If a procedure directs and controls the bypassing or removal from service of an annunciator or computer point, no requirement for review under 50.59 is necessary.

3. Alarms may be considered for removal from service under the following conditions:

A. Testing of components or equipment calibration B. The point has become unreliable C. A procedure directs and controls its removal from service

REVISION NO.: PROCEDURE TITLE:

PAGE:

14 CONDUCT OF OPERATIONS 97 Of 99 PROCEDURE NO.:

OP-M-1 00-1 000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 15 ANNUNCIATOR AND COMPUTER POINT CONTROL (Page2of4)

4. Authorization: Prior to authorizing the removal from service of any annunciator or computer point, the SRO shall ensure the following attributes are considered:

A. Station licensing basis is considered (Technical Specifications, Technical Requirement Manual, Offsite Dose Calculation Manual, UFSAR, Commitments, etc.)

B. Compensatory monitoring is considered and established where appropriate.

(1) The compensatory monitoring shall ensure the intended function of the original point is achieved to the maximum extent possible.

(2) The monitoring shall be formally documented via a plant process such as addition to logs, non-routine tracking sheet, or an Adverse Condition Monitoring Plan.

(3) The frequency of monitoring is commensurate with the importance of the function and the performance of the system.

(4) Impact of the compensatory monitoring on available resources is considered.

(5) If compensatory monitoring is established an AR is written to detem,ine if the condition creates a new Burden or Workaround.

C. The cause and effect of the issue is fully understood.

(1) Is the removal from service/bypassing due to a failed indicator?

(2) Is the point being deleted by procedure or due to a planned work order?

(3) Are there other indications of challenge within the system that must be better understood prior to authorizing the action?

Example: Generator Stator Winding temperature indication is viewed as unreliable and a request is received to delete the alarm. A further review of system performance indicates conductivity spikes and filter loading has been occurring. The point is not defeated and stator fouling is verified.

(4) Should the System Engineer be consulted for input?

REVISION NO.: PROCEDURE TITLE:

PAGE:

14 CONDUCT OF OPERATIONS PROCEDURE NO.: 98 of 99 OP-AA-1 00-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 15 ANNUNCIATOR AND COMPUTER POINT CONTROL (Page3of4)

(5) Does deletion of this point effect a calculation or mask multiple alarms?

(6) Are any automatic functions defeated?

(7) Are there any other points that have already been removed from service that are similar in nature to this one?

(8) Is there an aggregate impact from removing from service/bypassing this point along with others that have already been removed from service?

0. Does the annunciator or point support entry into or processing of Abnormal or Emergency Operating Procedures?

E. Notify the Shift Manager of the condition for consideration per OP-AA-1 00-1 000-10002, Daily Status Report Instruction.

5. Tracking and Awareness A. Each site shall maintain a tracking mechanism or log of defeated annunciators and computer points.

B. The following attributes are captured in the log at the time the point is removed from service:

(1) Point ID (2) Time/Date (3) Reason for removal (4) Work Order or AR tracking the issue (5) Compensatory Monitoring established

6. Flagging Where appropriate, a means of alerting Operators to the disabled annunciator should be considered. Examples include covers or overlays that highlight the defeated annunciator window or flagging placed on the window.

7. Periodic Review A periodic review process shall be in place to ensure configuration control, aggregate assessment, 1 OCFR5O .59/72.48 requirements are met and that changing conditions are considered.

REVISION NO.: PROCEDURE TITLE:

PAGE:

14 CONDUCT OF OPERATIONS 99 f 99 PROCEDURE NO.:

OP-AA-1 00-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 15 ANNUNCIATOR AND COMPUTER POINT CONTROL (Page 4 of 4)

A. A Monthly review shall be performed by Operations staff to ensure the following attributes are satisfied:

(1) The point is still required to be out of service (2) Compensatory monitoring, if applicable, remains adequate (3) Operations personnel are adequately aware of the item.

Consider utilizing flagging or adding to turnover documentation as necessary (4) No adverse trend within the effected system is being masked by the removed point (5) 10CFR5O.59/72.48 requirements are met. If the point was deleted via a procedure or is tracked on a Work Order this review is not required. An AR is written to track any item requiring review when the point has been out of service for greater than 30 days to prevent exceeding 90 days.

B. A Six Month assessment of the tracking mechanism is performed to ensure the following attributes are satisfied, in addition to the Monthly requirements:

(1) Operations SRO and Engineering representative(s) involved (2) Basis for removal/bypassing challenged (3) System trends for removed/bypassing points reviewed for possible indications that an adverse trend is being masked (4) Plant Operating Conditions and system performance still supports the removal/bypassing (5) Results of the review are provided to the Operations Manager and Plant Engineering Manager for review C. These reviews may be tracked by a mechanism appropriate to the site, typically a preventative maintenance tracking item or RTS is utilized.

D. Items are also reviewed under OP-AA-108 for potential Operations Workaround or Burden via the AR process

Procedure No Procedure Tdlc: age:

76 Miscellaneous Tests, Checks Approval Date O-OSP-200.5 and Operating Evolutions 2/21/14 ATTACHMENT 1 (Page 2 of 3)

DEFEATED/OUT-OF-SERVICE ANNUNCIATOR CHECKLIST QA RECORD PAGE 7 DEFEArED/OUT-OF-SERvICE ANNUNCIATOR Annunciator Window:

CHECKLIST (Page 2of3) Panel Col. Row 4N, Desibe actions necessary to restore this annunciator to normal operation including post stortion testing.

\

) I I r

/ I I /\ \

I ,\ \

I i__i___L N

5. Detatmg this :mmnciator has ee by:

Shift ti cinagel Si n tu,c znt (Date Ti/ne)

NOTE: 7 days is the maximum time allowed for ad inistratveIefeating an annunciator.

If the defeated annunciator or inputs to the annunciatfor L_}

will not be returned to service within 7 days, then noe I

the corrective action taken or the method used to traLik any temporary configuration in Block 8 by: /

(DateiTiine)

6. This annunciator has been defeated as described in Scctiun I of this form.

r Performed by: /

(Signature.) (Prinr) (Date Time) md Verified: /

Signature) (Prink) L -

(Date/Time)

Airnunciator Status Sheet Updated:

(Signature.) (Print) (Date/Time)

Sv stern Engmeer Notified:

(Signature) (Print,) (L)ate/Time)

N!A if block 2 was performed by System Engineer.

F259 2- of 3 1 Rev. 4 (O-OSP-200.5)

W201 O:JEClclslclsIcls

Procedure No. Procedure

Title:

lage 40 Miscellaneous Tests, Checks Approval Dale O-OSP-200.5 and Operating Evolutions 2/21/14 7.11 Annunciator Window Status Review Procedure Use requirement in Subsection 7.11 is: (Information Use)

INIT Date/Time Started: /

I N OTES

The following surveillance should be completed by the Unit Supervisor and reviewed by the Shift Manager.

The Annunciator Status Log (if required by Operations Department Instructions) and completed Attachment 1 forms should help in this review.

During outage periods, these requirements apply only to annunciators associated with systems required to be operable for the plant conditions. The Shift Manager or designee determines what annunciator will be addressed by this procedure under outage conditions.

j

Refer to O-ADM-051, Outage Risk Assessment and ControI and Technical Specification to determine systems required for shutdown/outage conditions.

Use the following list to support determination of alarm status (defeated or in service)

- Is a maintenance activity impacting alarm? I Are compensato,y actions in place?

I

- One input to alarm is NOT functional.

- Does the alarm have re-flash capability? I L I 7.11.1 Survey the Unit 3 and Unit 4 Control Room Annunciator Panels for any defeated/OOS Annunciator window(s) that should be operable for the plant conditions.

1. Survey the Unit 3 and Unit 4 1-IMI Annunciator Workstations for any disable points. Ensure all disabled points are being tracked using the Defeated/Out-Of-Service Annunciator Checklist.

7.11 .2 IF any windows are identified in Step 7.11 .1, THEN perform the following:

1. Merif the applicable blocks of Attachment 1 are completed for each defeated/ OOS annunciator AND that Attachment I is being retained in the EOOS Index.

2. Determine if there is a valid reason for the annunciator(s) to be defeated/OOS.

3. IF no valid reason identified, THEN have the annunciator(s) restored to its normal condition.

4. Document all remaining defeated/OOS annunciator windows that are NOT the result of a planned activity in the Remarks Section of the QA Record Page.

n iir i-i., 1.I.. 1,-I.-

Procedure No Procedure

Title:

Page:

41 Miscellaneous Tests, Checks Approval Date O-OSP-200.5 and Operating Evolutions 2/21/14 INIT 7.11 .2 (Contd)

5. Verify that each defeated/OOS annunciator that is NOT the result of a planned activity has been screened as an operator workaround using ODI-CO-040, Operator Workarounds and Operator Burdens.

a. Attach a copy of the Operator Workaround/Operator Burden Screening Checklist from ODI-CO-040 to each applicable Attachment I that is being retained in the EOOS Index.

6. IF an annunciator has been defeated/OOS for 60 days, THEN ensure a 10 CFR 50.59, Applicability Determination/Screening per Block 7 of Attachment 1, has been performed.

7. IF the 10 CFR 50.59, Applicability Determination/Screening shows that the defeated/OOS annunciator may adversely affect the safe operation of the unit(s), THEN notifj the Assistant Operations Manager for resolution.

NOTES I

The maximum time an annunciator may be administratively defeated is 7 days, unless I 7.11.2.8 is completed.

Substep

A Plant Work Order (PWO) that is tracking a defeated annunciator shall include a

statement that This Work Order shall not be canceled unless the affecte d

annunciator is returned to service and the original configuration established.

_ S S S a I

I

8. IF an annunciator can NOT be returned to service within 7 days, THEN complete Block 8 of Attachment 1.

9. IF an annunciator can NOT be returned to service within 7 days, THEN verify that Attachment 1, Section 8 has the appropriate tracking method for restoration.

10. WHEN the annunciator is returned to service, THEN complete Blocks 9 and 10 of Attachment 1.

7.11.3 Document the result of the determination on the QA Record Page.

7.11.4 Verify all log entries specified in Subsection 2.2 have been recorded.

W201 O:JEC/cls/cls/cls

SRO Question # 97 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1 I immediate operator actions? Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters J.Ftion that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

F 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page8ofl6

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 3 K/A# G3 2.3.4 Importance Rating 3.7 Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions.

Proposed Question: SRO Question # 98 Given the following conditions:

An Emergency Response Team is being briefed on a mission to rescue a worker who has suffered a broken leg.

Radiation Protection estimates each team member will receive a total dose (TEDE) of 9 rem and a thyroid dose of 6 rem while performing this rescue.

Which ONE of the following correctly describes (1) the estimated dose as compared to the TEDE limit of O-EPIP-201 11, Re-entry, and (2) the guidance for issuing Potassium Iodide (KI) to the team members?

REFERENCE PROVIDED A. (1) is within the TEDE limit; (2) KI should be issued to prevent iodine absorption B. (1) is within the TEDE limit; (2) KI should NOT be issued C. (1) exceeds the TEDE limit; (2) KI should NOT be issued D. (1) exceeds the TEDE limit; PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION (2) KI should be issued to prevent iodine absorption Proposed Answer: A Explanation (Optional):

A. Correct, because the estimated total dose (9 rem) does not exceed the TEDE rescue limit (10 rem) for a non-life-threatening situation; correct because KI should be distributed if the projected thyroid dose (CDE) is 5 rem.

B. Plausible if examinee does not recognize that a projected thyroid dose of 6 rem implies that the total committed organ dose (or CDE) will be 6 rem, which exceeds the CDE limit of 5 rem; the first part of the distractor is correct.

C. Plausible if examinee believes that the mission does not directly mitigate an event, minimize escalation, or minimize effluent releases, in which case the TEDE limit of 5 rem would be exceeded by the estimated total dose (9 rem); see explanation for distractor B.

D. See explanation for distractor C; the second part of the distractor is correct.

Technical Reference(s): 0-EPIP-20111 Step5.1.1.9and Enclosure 1, Page 1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: Yes Learning Objective: (As available)

Question Source: Bank # 98702 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 55.43 4 Comments:

Reference EPIP-20111, Enclosure 1(5 pages)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

SRO Question # 98 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

10 CFR 50.59 screening and evaluation processes.

Administrative processes for temporary modifications.

Administrative processes for disabling annunciators.

Administrative processes for the installation of temporary instrumentation.

Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

adiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

Process for gaseous/liquid release approvals, i.e., release permits.

Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 4 K/A# G4 2.4.16 Importance Rating 4.4 Emergency Procedures! Plan: Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, AOPs and SAMGs.

Proposed Question: SRO Question # 99 Given the following plant conditions:

The crew is implementing 3-EOP-ECA-0.0, Loss of All AC Power.

RCS pressure is 2340 psig and rising.

RCS Temperature (CETs) is 1050°F and rising slowly.

Loss of switchyard, U3 EDGs, and SBO crosstie.

Steam Generator Narrow Range levels are off-scale low.

AFW flow is not available.

Which one of the following identifies the appropriate procedure to restore Feedwater flow?

A. SACRG-1, Severe Accident Control Room Guideline Initial Response B. 3-EOP-FR-H.1, Response to Loss of Secondary Heat Sink C. 0-NOP-074.01, Standby Steam Generator Feedwater System D. 3-ONOP-075 Auxiliary Feedwater System Malfunction Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible if examinee thinks that CETs have reached the critical threshold, but a transition to SACRG-1 would jjjy occur if CETs are >1200°F and rising.

B. Incorrect. Plausible if examinee interprets all steam generator narrow range levels off scale low as a loss of secondary heat sink in all steam generators (which 3-EOP-FR-PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION H.1 provides a response for).

C. Incorrect. Plausible if examinee recognizes that Standby Steam Generator Feedwater Pumps LY be used to restore feedwater flow, but this is accomplished via 3-ONOP-075 (not 0-NOP-074.10).

D. Correct. This is the appropriate response, when proper AFW flow cannot be verified (i.e., when total AFW flow is not between 400 and 450 gpm) or narrow range levels cannot be maintained greater than 7% in at least one steam generator, per 3-EOP-ECA-0.0.

Technical Reference(s): 3-EOP-ECA-0.0 3-ONOP-075 .

3E0P-FR-H.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: PTN 6902348, Obj. 3 (As available)

Question Source: Bank # 87047 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Callaway Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

LOSS OF ALL AC POWER 18 of 88 PROCEDURE NO.:

3-EOP-ECA-O.O TURKEY POINT UNIT 3 I STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED CAUTION If CST level decreases to less than 12%, makeup water sources for the CST will be necessary to maintain secondary heat sink.

10. Check Intact SIG Levels:

a. Narrow Range Level a. Perform the following:

GREATER THAN 7%[27%]

1) Maintain maximum AFW flow until Narrow Range Level greater than 7%[27%] in at least one S/G.

2) H AFW flow can NOT be established, THEN try to restore AFW flow, or establish Standby Feed Flow, using 3-ONOP.-075, AUXILIARY FEEDWATER SYSTEM MALFUNCTION.

b. Control feed flow to maintain Narrow Range Level between 21%[27%j and 50%

c. Narrow Range Level c. Stop feed flow to any S/G with LESS THAN 50% Narrow Range Level greater than 50%.

REVISION NO.: PROCEDURE TITLE: PAGE:

LOSS OF ALL AC POWER 8 of 88 PROCEDURE NO.:

3-EOP-ECA-O.O TURKEY POINT UNIT 3 I STEP I [ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED I

4. Verify Proper AFW Flow:

a. Check AFW Pumps a. IF both units require AFW, AT LEAST TWO RUNNING THEN perform the following:

1) Establish 340 gpm flow to each unit.

2) Use a setpoint of 340 gpm to each unit for required AFW flow instead of the 400 (or 400 to 450) gpm specified in subsequent steps and procedures.

b. Verify total AFW flow b. Perform the following:

BETWEEN 400 AND 450 GPM

1) Verify AFW Pump running.

IL AFW Pump NOT running, THEN manually open steam supply valves.

2) Verify proper alignment of AFW valves.

IF alignment NOT proper, THEN manually align valves as necessary to establish proper alignment.

3) IF AFW can NOT be established, THEN restore AFW using 3-ONOP-075, AUXILIARY FEEDWATER SYSTEM MALFUNCTION, while continuing with Step 5.

REVISION NO.: PROCEDURE TITLE: PAGE:

LOSS OF ALL AC POWER 24 of 88 PROCEDURE NO.:

3-EOP-ECA-O.O TURKEY POINT UNIT 3 I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

19. Check Containment Pressure Perform the following:

HAS REMAINED LESS THAN 20 PSIG:

a. Verify Containment Isolation Phase B

PR-3-6306A actuated.

b. Verify Containment Isolation Phase B PR 3 6306B Valve white lights on VPB are aH bright.

c. U jjy Containment Isolation Phase B valve is NOT closed, THEN manually close valve.

IF valve(s) can NOT be manually closed, THEN manually or locally isolate the affected Containment Penetration -

d. Reset Containment Spray Signal.

20. Check Core Exit TCs IF Core Exit temperatures greater LESS THAN 1200°F than 1200°F AND increasing, THEN go to SACRG-1, SEVERE ACCIDENT CONTROL ROOM GUIDELINE INITIAL RESPONSE, Step 1.

REVISION NO.: PROCEDURE TITLE: PAGE:

6 RESPONSE TO LOSS OF SECONDARY HEAT SINK 4 of 60 PROCEDURE NO 3-EOP-FR-H.1 TURKEY POINT UNIT 3 1.0 PURPOSE This procedure provides actions to respond to Loss of Secondary Heat Sink in all Steam Generators.

2.0 SYMPTOMS AND ENTRY CONDITIONS This procedure is entered from:

1) E-0, REACTOR TRIP OR SAFETY INJECTION, Step 8, when minimum AFW flow is NOT verified AND Narrow Range Level in all S/Gs is less than 7%[27%].

2) ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS, Step 2 and Step 18, when adequate feed flow is NOT available to maintain RCS temperature control.

3) F-0.3, HEAT SINK CRITICAL SAFETY FUNCTION STATUS TREE, on a RED condition.

REVISION NO.: PROCEDURE TITLE: PAGE:

AUXILIARY FEEDWATER SYSTEM MALFUNCTION 13 of 27 PROCEDURE NO.:

3-ONOP-075 TURKEY POINT UNIT 3 STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED I 3.2 Subsequent Operator Actions (continued)

6. ESTABLISH Standby Feedwater flow to at least one S/G from A SSGFP.

A. CHECK 3C 4KV Bus energized. GO TO Section 3.2 Step 7.

B. ENSURE Feedwater Bypass Isolation, RESET.

C. ENSURE Feedwater Bypass Isolation PERFORM the following:

valves are OPEN.

1. ENSURE Feedwater Bypass Isolation

POV-3-477, 3A F/W BYPASS valve control switch is in either AUTO ISOLATION position.

POV-3-487, 3B FiW BYPASS 2. IF applicable Feedwater Bypass ISOLATION Isolation valve will NOT OPEN in AUTO, THEN locally OPEN valve.

POV-3-497, 3C F/W BYPASS ISO LATI ON D. CHECK PI-3-1616, FEEDWATER Locally THROTTLE OPEN, three turns HEADER REMOTE PRESSURE DWDS-3-012, ISOLATION VALVE STBY INDICATOR, greater than 500 psig. SG FEED PUMPS DISCH TO UNIT 3 MAIN FW HEADER.

E. START the A SSGFP. IF A SSGFP can NOT be started, THEN GO TO Section 3.2 Step 7.

F. Locally OPEN DWDS-3-012, ISOLATION VALVE STBY SG FEED PUMPS DISCH TO UNIT 3 MAIN FW HEADER.

REVISION NO.: PROCEDURE TITLE: PAGE:

AUXILIARY FEEDWATER SYSTEM MALFUNCTION 11 of 27 PROCEDURE NO.:

3-ONOP-075 TURKEY POINT UNIT 3 I STEP I I ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 3.2 Subsequent Operator Actions (continued)

NOTE Lost of AFW due to a fire in the foHowing Zones:

Zone 79 (outdoor area West of Unit 4 Containment)

Zone 83 (Instrument Air Equipment Area)

Zone 84 (Auxiliary Feedwater Pump Area)

Zone 117 (U3 and U4 Turbine Deck when Unit 4 is in an outage)

Will require B Standby SIG FW Pump to be started locally by placing the Master Control switch to RUN or MANUAL. Master Control switch is located on the Pump Control Panel.

5. ESTABLISH Standby Feedwater flow to at least one S/G from B SSGFP.

A. ENSURE Feedwater Bypass Isolation, RESET.

B. ENSURE Feedwater Bypass Isolation PERFORM the following:

valves are OPEN.

1. ENSURE Feedwater Bypass Isolation

POV-3-477, 3A FIW BYPASS valve control switch is in either AUTO ISOLATION position.

POV-3-487, 3B F/W BYPASS 2. IF applicable Feedwater Bypass ISOLATION Isolation valve will NOT OPEN in AUTO, THEN locally OPEN valve.

POV-3-497, 3C FAN BYPASS ISOLATION C. CHECK P1-3-1616, FEEDWATER Locally THROTTLE OPEN, three turns HEADER REMOTE PRESSURE DWDS-3-012, ISOLATION VALVE STBY INDICATOR, greater than 500 psig. SG FEED PUMPS DISCH TO UNIT 3 MAIN FW HEADER.

SRO Question # 99 Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing flowpath, logic, component location?

Can the question be answered solely by knowing 1 immediate operator actions? Yes RO question 9

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the questio n be answered solely by knowing the purpose, overquence of events, or Yes RO question overall mitigative strategy of a 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 4 K/A# G4 2.4.9 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of low power I shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Proposed Question: SRO Question # 100 The following plant conditions exist:

A SITE AREA EMERGENCY was declared 37 minutes ago.

The emergency response facilities are NOT activated.

The communicator has made the notification to the State and NRC.

Conditions are stabilized and the event no longer meets the emergency action level criteria.

Who is responsible for de-escalation of the classification?

A. Recovery Manager B. Emergency Coordinator C. Nuclear Division Duty Officer D. Emergency Control Officer Proposed Answer: A Explanation (Optional):

A. Correct. Once the plant classifies a SAE, jy the RM has the authority to de-escalate to a lower classification.

B. Incorrect. Plausible if examinee thinks that the EC has the authority to de-escalate a classification, but the EC may jjy de-escalate from an Alert or UE.

C. Incorrect. Plausible if examinee thinks that the NDDO has the authority to de-escalate a

classification, but the EC does not have the authority to de-escalate from a SAE.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. Incorrect. Plausible if examinee thinks that the EC or DCS have the authority to de escalate a classification, but the NDDO has no authority to de-escalate from jy classification.

Technical Reference(s): 0-LPIP-20101 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: PTN 3200003, Obj. 6 (As available)

Question Source: Bank # 98648 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Seabrook Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

SRO Question#1oO1 Clarification Guidance for SRO-orily Questions Rev 1(03111/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(

b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

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

1I Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? iJ+Edestion Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? 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 sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

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