ML15119A467

Initial Exam 2015-301 Draft Administrative Documents

ML15119A467
Person / Time
Site:	Turkey Point
Issue date:	04/29/2015
From:	Division of Reactor Safety II
To:	Florida Power & Light Co
References
Download: ML15119A467 (29)

Text

ES-401 PWR Examination Outline Form ES-401-2 I Facility: TURKEY POINT Date of Exam: JANUARY 2015 RO KIA Categorv Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

• Total

1. 1 3 3 3 3 3 3 18 3 3 6 Emergency&

Abnormal 2 1 1 2 NIA 2 2 NIA 1 9 2 2 4 Plant Evolutions Tier Totals 4 4 5 5 5 4 27 5 5 10 1 2 3 2 3 2 2 3 3 2 3 3 28 3 2 5 2.

Plant 2 1 1 1 0 1 1 1 1 1 1 1 10 0 I 2 3 Systems Tier Totals 3 4 3 3 3 3 4 4 3 4 4 38 4 4 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 3 3 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those KIAs having an importance rating (IR) of 2.5 or higher shall be selected .

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7.* The generic (G) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog , but the topics must be relevant to the applicable evolution or system. Refer to Section 0 .1.b of ES-401 for the applicable KIAs.

8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals(#) on Form ES-401 -3. Limit SRO selections to KIAs that are linked to 10 CFR 55.43.

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401 *2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 007EK1 .02 Reactor Trip - Stabilization - Recovery 3.4 3.8 ~ D D D D D D D D D D Shutdown margin

/1 009EK1 .01 SmallBreakLOCA/3 4.2 4.7 ~DD 0 0 0 0 0 0 0 0 Natural circulation and cooling, including reflux boiling 011 EG2.1 .20 Large Break LOCA I 3 4.6 4.6 D D D D D D D D D D ~ Ability to execute procedure steps.

01SAA1.05 RCP Malfunctions I 4 3.8 3.8 D D D D 0 0 ~ 0 D D D RCS flow 022AK3.01 Loss of Rx Coolant Makeup I 2 2.1 3.1 DD~ DD D 0 DD 0 0 Adjustment of RCP seal backpressure regulator valve to obtain normal flow 025AK2.03 Loss of AHR System 14 2.7 2.7 D ~ D D 0 D D D D D D Service water or closed cooling water pumps 027AA1.05 Pressurizer Pressure Control System 3.3 3.2 D O O D D D ~ D D D D Transfer of heaters to backup power supply Malfunction I 3 029EA1.03 ATWS/1 3.5 3.2 D 0 D D D 0 ~ D D D 0 Charging pump suction valves from VCT operating switch 038EK3.08 Steam Gen. Tube Rupture / 3 4.1 4.2 D D ~ D D 0 0 D D 0 D Criteria for securing RCP 055EG2.2.38 Station Blackout/ 6 3.6 4.s D D D D D D 0 D D 0 ~ Knowledge of conditions and limitations in the facility license.

057AA2.12 Loss of Vital AC Inst. Bus I 6 3.s 3.1 D D D D D D 0 ~ D 0 D PZR level controller, instrumentation and heater indications Page 1of2 03/18/2014 9:06 AM

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 058AK1 .01 Loss of DC Power I 6 2.8 3.1 ~ D D D D D D D D D D Battery charger equipment and instrumentation 062AG2.4.4 Loss of Nuclear Svc Water I 4 4.5 4.1 DD DD DD DD DD ~ Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

077 AK3.02 Generator Voltage and Electric Grid 3.6 3.9 D D ~ D D D D D D D D Actions contained in abnormal operating procedures for Disturbances I 6 voltage and grid disturbances WE04EK2.1 LOCA Outside Containment I 3 3.5 3.9 D ~ D D D D D D D D D Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features .

WE05EK2.2 Inadequate Heat Transfer - Loss of 3.9 4.2 D ~ D D D D D D D D D Facility's heat removal systems, including primary Secondary Heat Sink I 4 coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

WE11 EA2.1 Loss of Emergency Coolant Recirc. / 4 3.4 4.2 D D D D D D D ~ D D D Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

WE12EA2.2 Steam Line Rupture- Excessive Heat 3.4 3.9 D D D D D D D ~ D D D Adherence to appropriate procedures and operation Transfer/ 4 within the limitations in the facility's license and amendments.

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ES-401, REV 9 T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 001AA2.05 Continuous Rod Withdrawal I 1 4.4 4.6 0 0 0 0 0 0 0 ~ 0 0 0 Uncontrolled rod withdrawal from available indications 028AK3.03 Pressurizer Level Malfunction I 2 3.5 4.1 0 0 ~ 0 0 0 0 0 0 0 0 False indication of PZR level when PORV or spray valve is open and RCS saturated 051AA1 .04 Loss of Condenser Vacuum I 4 2.5 2.5 0 0 0 0 0 0 ~ 0 0 0 0 Rod position 060AA1.02 Accidental Gaseous Radwaste Rel. I 9 2.9 3.1 O O O O 0 0 ~ 0 0 0 0 Ventilation system 068AK2.07 Control Room Evac. I 8 3.3 3.4 O ~ O O O 0 0 0 D 0 0 ED/G 076AG2.4.46 High Reactor Coolant Activity I 9 4.2 4.2 0 0 0 0 0 0 0 0 0 0 ~ Ability to verify that the alarms are consistent with the plant conditions.

WE02EK3.4 SI Termination I 3 3.5 3.8 0 0 ~ 0 0 0 0 0 0 0 0 RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

WE08EK1 .1 RCS Overcooling - PTS I 4 3.5 3.8 ~ 0 0 0 0 0 0 0 0 0 0 Components, capacity, and function of emergency systems.

WE1 OEA2.1 Natural Circ. With Seam Void/ 4 3.2 3.9 0 0 0 0 0 0 0 ~ 0 0 0 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

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ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 003K4.07 Reactor Coolant Pump 3.2 3.4 D D D ~ D D D D D D D Minimizing RCS leakage (mechanical seals) 004A1.09 Chemical and Volume Control 3.6 3.8 D D D D D D ~ D D D D RCS pressure and temperature 004K5.43 Chemical and Volume Control 3.6 3.9 D D D D ~ D D D D D D Saturation, subcooling, superheat in steam/water 005K4.11 Residual Heat Removal 3.5 3.9 D D D ~ D D D D D D 0 Lineup for low head recirculation mode (external and internal) 0061<2.02 Emergency Core Cooling 2.s 2.9 D ~ O O D 0 0 D 0 0 0 Valve operators for accumulators 007A3.01 Pressurizer Relief/Quench Tank 2.7 2.9 0 0 0 0 0 0 0 0 ~ 0 0 Components which discharge to the PAT 008A1.02 Component Cooling Water 2.9 3.1 D D 0 0 D D ~ 0 D 0 0 CCW temperature 0101<2.02 Pressurizer Pressure Control 2.5 2.1 D ~ O O D 0 0 0 D 0 0 Controller for PZR spray valve 012A2.03 Reactor Protection 3.4 3.7 0 0 0 0 0 0 0 ~ 0 0 0 Incorrect channel bypassing 012K3.04 Reactor Protection 3.8 4.1 O D ~ O 0 0 0 D 0 0 0 ESFAS 013K6.01 Engineered Safety Features Actuation 2.7 3.1 O D 0 0 0 ~ 0 0 D D D Sensors and detectors Page 1of3 03/18/2014 9:06 AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 0221<2.01 Containment Cooling 3.o 3.1 D ~ D D D D D D D D D Containment cooling fans 022K3.01 Containment Cooling 2.0 3.2 D D ~ D D D D D D D D Containment equipment subject to damage by high or low temperature, humidity and pressure 026A4.01 Containment Spray 4.s 4.3 D D D D D D D D D ~ D CSS controls 039A3.02 Main and Reheat Steam 3.1 3.s D D D D D D D D ~ D D Isolation of the MASS 059A4.03 Main Feedwater 2.0 2.0 D D D D D D D D D ~ D Feedwater control during power increase and decrease 059G2.1.28 Main Feedwater 4.1 4.1 D D D D D D D D D D ~ Knowledge of the purpose and function of major system components and controls.

061 KS.01 Auxiliary/Emergency Feedwater 3.6 3.9 D D D D ~ D D D D D D Relationship between AFW flow and RCS heat transfer 061 K6.02 Auxiliary/Emergency Feedwater 2.6 2.1 D D D D D ~ D D D D D Pumps 062G2.2.39 AC Electrical Distribution 3.9 4.s D D D D D D D D D D ~ Knowledge of less than one hour technical specification action statements for systems.

063K1 .02 DC Electrical Distribution 2.1 3.2 ~ D D D D D D D D D D AC electrical system 063K4.04 DC Electrical Distribution 2.6 2.0 D D D ~ D D D D D D D Trips Page 2 of 3 03/18/2014 9:06 AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 064A1.03 Emergency Diesel Generator 3.2 3.3 0 0 0 0 0 0 ~ 0 0 0 0 Operating voltages, currents and temperatures 064K1.04 Emergency Diesel Generator 3.6 3.9 ~ 0 0 0 0 0 0 0 0 0 0 DC distribution system 073A2.01 Process Radiation Monitoring 2.5 2.9 0 0 0 0 0 0 0 ~ 0 0 0 Erratic or failed power supply 076G2.4.31 Service Water 4.2 4.1 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of annunciators alarms, indications or response procedures 078A4.01 Instrument Air 3.1 3.1 0 0 0 0 0 0 0 0 0 ~ 0 Pressure gauges 103A2.05 Containment 2.9 3.9 0 0 0 0 0 0 0 ~ 0 0 0 Emergency containment entry Page 3 of 3 03/18/2014 9:06 AM

ES-401, REV 9 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 001 K6.08 Control Rod Drive 2.9 3.2 D D D D D ~ D D D D D Purpose and position switch of alarm for high flux at shutdown 011A4.05 Pressurizer Level Control 3.2 2.9 D 0 0 0 0 0 0 D 0 ~ 0 Letdown flow controller 017K5.02 In-core Temperature Monitor 3.7 4.0 O D D D ~ 0 0 0 0 0 0 Saturation and subcooling of water 027K2.01 Containment Iodine Removal 3.1 3.4 O ~ D D 0 0 0 0 0 0 0 Fans 029A2.01 Containment Purge 2.9 3.6 O D D D 0 0 0 ~ 0 0 0 Maintenance or other activity taking place inside containment 041A3.05 Steam Dump/Turbine Bypass Control 2.9 2.9 0 O 0 0 0 0 0 0 ~ 0 0 Main steam pressure 045A1 .06 Main Turbine Generator 3.3 3.7 O O 0 0 0 0 ~ 0 0 0 0 Expected response of secondary plant parameters following T/G trip 055K1 .06 Condenser Air Removal 2.6 2.6 ~ 0 0 0 0 0 0 0 0 0 0 PAM system 068G2.4.11 Liquid Radwaste 4.0 4.2 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of abnormal condition procedures.

075K3.07 Circulating Water 3.4 3.5 D D ~ 0 0 0 0 0 0 0 0 ES FAS Page 1of1 03/18/2014 9:06 AM

ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO G2.1.3 Conduct of operations 3.7 3.9 D D D D D D D D D D ~ Knowledge of shift or short term relief turnover practices.

G2.1.42 Conduct of operations 2.s 3.4 D D D D D D D D D D ~ Knowledge of new and spent fuel movement procedures G2.2.17 Equipment Control 2.6 3.8 D D D D D D D D D D ~ Knowledge of the process for managing maintenance activities during power operations.

G2.2.18 Equipment Control 2.6 3.8 D D D D D D D D D D ~ Knowledge of the process for managing maintenance activities during shutdown operations.

G2.3.12 Radiation Control 3.2 3.7 DD DD DD DD DD ~ Knowledge of radiological safety principles pertaining to licensed operator duties G2.3.13 Radiation Control 3.4 3.8 D D D D D D D D D D ~ Knowledge of radiological safety procedures pertaining to licensed operator duties G2.3.14 Radiation Control 3.4 3.8 D D O O 0 0 0 D 0 0 ~ Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities G2.4.13 Emergency Procedures/Plans 4.o 4.6 O O O O 0 0 0 0 D 0 ~ Knowledge of crew roles and responsibilities during EOP usage.

G2.4.25 Emergency Procedures/Plans 3.3 3.7 D D D D D 0 D D D D ~ Knowledge of fire protection procedures.

G2.4.47 Emergency Procedures/Plans 4.2 4.2 D O O D 0 0 0 0 D 0 ~ Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

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ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 007EA2.02 Reactor Trip - Stabilization - Recovery 4.3 4.6 O D 0 0 0 0 0 ~ 0 0 0 Proper actions to be taken if the automatic safety func-I1 tions have not taken place 025AA2.02 Loss of AHR System I 4 3.4 3.8 D 0 0 0 0 0 0 ~ 0 0 0 Leakage of reactor coolant from AHR into closed cooling water system or into reactor building atmosphere 026AG2.1 .19 Loss of Component Cooling Water I 8 3.9 3.8 O D 0 0 0 0 0 0 0 0 ~ Ability to use plant computer to evaluate system or component status.

029EA2.05 ATWS I 1 3.4 3.4 D D 0 D D D D ~ D D 0 System component valve position indications 056AG2.1.31 Loss of Off-site Power I 6 4.6 4.3 D D D D 0 D D D D D ~ Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup.

062AG2.4.20 Loss of Nuclear Svc Water I 4 3.8 4.3 D D D D D D D D D D ~ Knowledge of operational implications of EOP warnings, cautions and notes.

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ES-401, REV 9 SRO T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 028AG2.4.21 Pressurizer Level Malfunction I 2 4.0 4.6 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of the parameters and logic used to assess the status of safety functions 068AA2.10 Control Room Evac. I 8 4.2 4.4 O O O O 0 D D ~ 0 D 0 Source range count rate we07EG2.4.6 Saturated Core Cooling Core Cooling 3.7 4.7 O O 0 O 0 0 0 0 0 0 ~ Knowledge symptom based EOP mitigation strategies.

14 WE1 OEA2.2 Natural Circ. With Seam Void/ 4 3.4 3.9 O O O O 0 0 D ~ 0 D 0 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

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ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 013G2.2.25 Engineered Safety Features Actuation 3.2 4.2 D D D D D D D D D D ~ Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

039G2.4.45 Main and Reheat Steam 4.1 4.3 D O O D D 0 0 D D 0 ~ Ability to prioritize and interpret the significance of each annunciator or alarm.

063A2.02 DC Electrical Distribution 2.3 3.1 O O D D D D D ~ D D D Loss of ventilation during battery charging 076A2.02 Service Water 2.1 3.1 D O O D D D D ~ D D D Service water header pressure 103A2 03 Containment 3.s 3.a D D D D D D D ~ D D D Phase A and B isolation Page 1of1 03/18/2014 9:06 AM

ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401 *2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 002G2.2.22 Reactor Coolant 4.o 4.7 D D D D D D D D DD ~ Knowledge of limiting conditions for operations and safety limits.

015G2.1 .30 Nuclear Instrumentation 4.4 4.o D D D D D D D D D D ~ Ability to locate and operate components, including local controls.

068A2.04 Liquid Radwaste 3.3 3.3 DD D D D D D ~ D D D Failure of automatic isolation Page 1of1 03/18/2014 9:06 AM

ES-401, REV 9 SRO T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO G2.1.29 Conduct of operations 4.1 4.0 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

G2.1.9 Conduct of operations 2.0 4.s O O O O 0 0 D 0 0 D ~ Ability to direct personnel activities inside the control room .

G2.2.2 Equipment Control 4.6 4.1 0 0 0 0 0 0 0 0 0 0 ~ Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

G2.2.43 Equipment Control 3.o 3.3 O O O O 0 0 D 0 0 0 ~ Knowledge of the process used to track inoperable alarms G2.3.4 Radiation Control 3.2 3.7 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of radiation exposure limits under normal and emergency conditions G2.4.16 Emergency Procedures/Plans 3.5 4.4 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines.

G2.4.9 Emergency Procedures/Plans 3.8 4.2 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of low power I shutdown implications in accident (e.g. LOCA or loss of AHR) mitigation strategies.

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ES-301 Administrative Topics Outline Form ES-301-1 Facility: Turkey Point Units 3 & 4 Date of Examination: 01/19/15 Examination Level: RO IZJ SROO Operating Tes.t No.: 2015-301 Administrative Topic (see Note) Type Code* Describe activity to be performed 2.1 .29 (4.1) - Knowledge of how to conduct system lineups, A.1 .a - Conduct of Operations N,R such as valves, breakers, switches, etc.

JPM: Blend To The RWST 2.1.23 (4.3) -Ability to perform specific system and integrated plant procedures during all modes of plant A.1.b - Conduct of Operations D,R operation.

JPM: Perform Reactor Coolant System Leak Rate Calculation

- Manual Method 2.2.40 (3.4) -Ability to apply Technical Specifications for a system.

A.2 - Equipment Control N, R JPM: Evaluate Containment Spray Pump Vibration data sheet and apply Technical Specifications, if applicable.

A.3 - Radiation Control N/A NOT SELECTED FOR RO EXAM 2.4.21 (4.0) - Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, A.4 - Emergency Procedures/Plan D,R reactor coolant system integrity, containment conditions, radioactivity release control, etc.

JPM: Perform 4-EOP-F-O, Evaluate Critical Safety Function Status Trees NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room , (S)imulator, or Class(R)oom (D)irect from bank (s 3 for ROs; s 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (;;:: 1)

(P)revious 2 exams (S 1; randomly selected)

Page 1of2

JPM

SUMMARY

STATEMENTS A.1.a Blend To The RWST This is a modified JPM. The US directs the RO to calculate the makeup to the RWST to clear the RWST Low Level Alarm by adding blended flow in accordance with O-OP-046, CVCS - Boron Concentration Control. The RO is directed to use Method 2 for the blended flow calculations. The operator is required to give the boric acid and water flowrates, total volumes, controller settings as required to makeup to the RWST. O-OP-046 and the Unit 3 Plant Curve Book (PCB) are provided for reference.

A. 1. b Perform Reactor Coolant System Leak Rate Calculation - Manual Method This is a bank JPM (02041036102). The operator is provided a list of plant conditions including power level, steady state, plant computer is OOS. 3-0SP-041.1, Reactor Coolant System Leak Rate Calculation, selected actions are complete with data provided for recording and to perform calculations. The RCS leak rate calculation uses the manual method using Attachment 3. The operator is required to review results and determine acceptance criteria is NOT met, and inform the Unit Supervisor to comply with TS LCO 3.4.6.2.b - Action b. The unit is required within four hours to reduce leakage to within limits to maintain continued plant operation. PTN Technical Specifications and 3-0SP-041.1, Reactor Coolant System Leak Rate Calculation, are provided for reference.

A.2 Evaluate Containment Spray Pump Vibration Data Sheet And Apply Technical Specifications This is a new JPM. The crew is performing 3-0SP-068.2, Containment Spray System lnservice Test. The operator provides Attachment 1, 3A Containment Spray Pump, for review. Vibration results are beyond acceptable specified range limits with other criteria is sat. The operator is required to identify if results are satisfactory and to list any required actions, if applicable. PTN Technical Specifications and 3-0SP-068.2 are provided for reference.

A.3 NOT SELECTED FOR RO EXAM A.4 Perform 4-EOP-F-O. Evaluate Critical Safety Function Status Trees This is a bank JPM (01200002501). During an emergency, the operator is provided with a list of plant parameters: containment, power level, core conditions, RCS data, and SIG data. The operator is directed to determine the highest priority procedure to restore the highest priority critical safety function. This determination is recorded on Attachment 2 of 4-EOP-F-O, Critical Safety Function Status Trees. This action results in an Orange determination on CSF Integrity and recommends transition to 4-EOP-FR-P.1, Response to Imminent Thermal Shock Condition.

Page 2 of 2

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Turkey Point Units 3 & 4 Date of Examination: 01/19/15 Examination Level: ROD SRO Lt! Operating Test No. : 2015-301 Administrative Topic (see Note) Type Code* Describe activity to be performed 2.1.34 (3.5) - Knowledge of primary and secondary plant N,R chemistry limits.

A.1 .a - Conduct of Operations JPM: Review Primary And Secondary Sample Results And Determine Applicable Actions.

2.1.23 (4.4) -Ability to perform specific system and integrated plant procedures during all modes of plant A.1.b - Conduct of Operations D,R operation.

JPM: Review Reactor Coolant System Leak Rate Calculation

- Manual Method 2.2.40 (4.7) -Ability to apply Technical Specifications for a system.

A.2 - Equipment Control N, R JPM: Evaluate Containment Spray Pump Vibration data sheet and apply Technical Specifications.

2.3.6 (3.8) -Ability to approve release permits.

A.3 - Radiation Control N,R JPM: Determine Requirements To Recommence A Radiological Release Once Suspended.

2.4.41 (4.6) - Knowledge of the emergency action level thresholds and classifications.

A.4 - Emergency Procedures/Plan D,R JPM: Classify the Event and Issue PARs NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; s 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (;? 1)47 (P)revious 2 exams (S 1; randomly selected)

Page 1 of2

JPM

SUMMARY

STATEMENTS A.1.a Review Primary And Secondary Sample Results And Determine Applicable Actions This is a new JPM. Chemistry delivers primary and secondary sample results. The current power increase is on hold at 55% for 3A Steam Generator Feedwater Pump repair. O-ADM-651, Nuclear Chemistry Parameters Manual, and PTN Technical Specifications are provided for the initial review of Chemistry parameters. Also, 3-0NOP-071.1 , Secondary Chemistry Deviation from Limits, and O-ONOP-041 .10, Primary Chemistry Deviation from Limits, are provided as needed. For the sample results reviewed, the operator identifies a secondary plant chemistry limit is exceeded and defines the minimum required plant response. With 3A SG sodium at 275 ppb from the sample taken 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago, the 3A SG sodium exceeds the Action 3 limit. Therefore, reduce power below 5% as quickly as safe plant operations permit.

A.1.b Review Reactor Coolant System Leak Rate Calculation - Manual Method This is a bank JPM (02041036102). The operator is provided a list of plant conditions including power level, steady state, plant computer is OOS. 3-0SP-041 .1, Reactor Coolant System Leak Rate Calculation, selected actions are complete with data taken. The RCS leak rate calculation is completed using the manual method using Attachment 3. The operator is required to review results and determine an error exists in the RCS leak rate, determine acceptance criteria is NOT met., and comply with TS LCO 3.4.6.2.b - Action b. The unit is required within four hours to reduce leakage to within limits to maintain continued plant operation. PTN Technical Specifications and 3-0SP-041 .1, Reactor Coolant System Leak Rate Calculation, are provided for reference.

A.2 Evaluate Containment Spray Pump Vibration Data Sheet And Apply Technical Specifications This is a new JPM. The crew is performing 3-0SP-068.2, Containment Spray System lnservice Test. The operator provides Attachment 1, 3A Containment Spray Pump, for review. Vibration results are beyond acceptable specified range limits with other criteria is sat. The operator is required to identify if results are satisfactory and to list any required actions, if applicable. PTN Technical Specifications and 3-0SP-068.2 are provided for reference.

A.3 Determine Requirements To Recommence A Radiological Release Once Suspended.

This is a new JPM. The operator is given initial conditions that R-18 , Waste Disposal System Liquid Effluent Monitor, has spiked high during a radioactive discharge of Waste Monitor Tank A. RCV-018, Liquid Waste Discharge Valve is closed. The R-18 is declared out-of-service by l&C and return to service time is expected to be two days from now. One Chemist verifies Waste Monitor Tank A effluent is higher than initial sample results. The operator determines if Waste Monitor Tank A discharge can be recommenced on the same Liquid Release Permit without R-18 available and lists any additional requirements, if applicable. O-NOP-061.11 C, Controlled Liquid Release From Waste Monitor Tank A and the ODCM are provided for reference.

A.4 Classify the Event and Issue PARs This is a bank JPM (02201052311). The operator will be provided with a list of plant conditions during an emergency. The operator will be directed to classify the event using O-EPIP-20101, Duties of the Emergency Coordinator, and issue protective action recommendations using O-EPIP-20134, Offsite Notifications and Protective Action Recommendations. The operator will be expected to classify the event as a Site Area Emergency, identify the PARs, and complete the Florida State Notification Form.

Page 2 of2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Turkey Point Units 3 & 4 Date of Examination: 01/19/15 Exam Level: RO ~ SRO-I D SRO-UO Operating Test No. : 2015-301 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System I JPM Title Type Code* Function A. 001 Control Rod Drive System [A2.11 - RO 4.4 SRO 4.7)

JPM: Recover A Dropped Rod A,M,S 1 B. 011 Pressurizer Level Control System [A4.01 RO 3.5 SRO 3.2)

JPM: Restore Pressurizer Level After A Loss Of Charging Flow N,S 2 C. 006 Emergency Core Cooling System [A 1.13 RO 3.5 SRO 3. 7]

JPM: Drain The 38 Accumulator For Boron And Level Adjustment A,M,EN,S 3 D. 005 Residual Heat Removal System [A2.03 RO 2.9 SRO 3.1)

JPM: Respond To A Loss Of RHR A, D, L,S 4P E. 045 Main Turbine Generator System [A4.02 RO 2.7 SRO 2.6*]

JPM: Synchronize Main Generator To Line (Manual Sync) D,S 4S F. 022 Containment Cooling System [A4.01 R03.6 / SR03.6]

JPM: Start 3A Normal Containment Cooler N,S 5 G. 015 Nuclear Instrumentation System [A4.01 RO 3.6* SRO 3.6*]

JPM: Respond To A Source Range Malfunction During Refueling D,L, S 7 H. 008 Component Cooling Water System [A2.01 RO 3.3 SRO 3.6]

JPM: Respond To 3A CCW Pump Trip During 3-0SP-030.5, Component A, N,*S 8 Cooling Water Pumps Low Header Pressure Start Test Page 1 of 4

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

I. E12 Uncontrolled Depressurization Of All Steam Generators

[EA2.2 RO 3.4 SRO 3.9)

N,E 4S JPM: Respond To A Steam Line Fault Of All SGs And Isolate 3A S/G J. 056 Loss Of Offsite Power [AA2.14 RO 4.4 SRO 4.6)

JPM: Start, Synchronize and Load 4A (48) Emergency Diesel Generator A,D, E 6 Locally K. 071 Waste Gas Disposal System [A4.26 RO 3.1 SRO 3.9)

JPM: Perform A Gaseous Radwaste Release D,R 9

@ All RO and SRO-I Control Room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the Control Room.

• Type Codes Criteria for RO I SRO-I / SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank :S9/:S8/:S4 (E)mergency or abnormal in-plant 2:1/2:1/2:1 (EN)gineered safety feature - I - I 2:1 (control room system)

(L)ow-Power I Shutdown 2:1/2:1/<::1 (N)ew or (M)odified from bank including 1(A) <::2/<::2/<::1 (P)revious 2 exams :S 3 / :S 3 / :S 2 (randomly selected)

(R)CA 2:1/<::1/<::1 (S)imulator Page 2 of 4

TURKEY POINT 2015 NRC EXAM JPM

SUMMARY

A. Recovery Of A Dropped Rod JPM A is a modified bank JPM. Unit 3 is in MODE 1 escalating power when a rod drops. Current power level is 48%. Shutdown Margin is verified adequate for the current plant conditions. l&C has determined and corrected a degraded condition causing the rod to drop. After the Shift Manager gives concurrence, the Unit Supervisor directs the RO to perform Attachment 1 of 3-0NOP-028.3, Dropped RCC. (Alternate Path) When the rod withdraws, a second rod drops requiring the RO to trip the Reactor. The RO continues with Reactor Trip verification.

B. Restore Pressurizer Level After A Loss Of Charging Flow JPM B is a new JPM. Unit 3 is in MODE 1 with Reactor Power at 23%. The Pressurizer level trend is at 30%

and lowering with a leak reported on 3A Charging Pump upstream 3A Chrg Pump Discharge, 3-286. During the automatic system response, LC-3-459G, Pressurizer Level Master Control fails low. The Unit Supervisor directs the RO to respond with 3-0NOP-047.1, Loss of Charging Flow in Modes 1Through4. The RO starts with isolating two Letdown Orifice Isolation Valves. Then, he closes MOV-3-6386 to isolate Excess UD and RCP Seal Return. He continues by stopping 3A and 3C Charging Pumps for leak isolation. The SNPO is dispatched to isolate 3-286. When the leak is isolated, the RO restores Charging in accordance with Attachment 1 and re-establishes Letdown with Attachment 2.

C. Drain The 38 Accumulator For Boron And Level Adjustment JPM C is a modified JPM (01047021301). Unit 3 is in MODE 1 with Reactor Power at 100%. The ECCS Accumulator has check valve leakage occurring. The 38 Accumulator is inoperable based on level being too high at about 6,890 gallons and boron too low at about 2,050 ppm. The Unit Supervisor directs the RO to perform 3-NOP-064, Safety Injection Accumulators , Section 5.2 to drain 38 Accumulator to 5,000 gallons for boron addition and a level adjustment. The RO verifies the operability of other accumulators (pressure and level) prior to draining 38 Accumulator. When CV-3-8528, 38 Accumulator Drain Valve, is opened for the 38 Accumulator, the CV-3-852A, 3A Accumulator Drain Valve, also leaks by. The Caution before the draining step states only one drain valve shall be opened at a time. (Alternate Path) With the 3A Accumulator level lowering, the RO closes CV-3-8528 which allows the leakage associated with CV-3-852A to stop. The 3A Accumulator low pressure annunciator alarms. The response directs the RO to re-pressurize 3A Accumulator.

D. Respond To A Loss Of RHR JPM D is a bank JPM (01050004301 ). Unit 3 is in MODE 4 with RCS temperature just below 350°F . RHR Shutdown Cooling is in service to maintain current conditions. The RO is directed to maintain current conditions. Subsequently, the operating RHR Pump amps fall and no flow is indicated. The RO enters 3-0NOP-050, Loss of RHR for response. The SNPO is dispatched to check the 3A RHR Pump. The report to the Control Room is the 3A RHR Pump's shaft is sheared. The 3B RHR Pump is started with 3-0NOP-050, Loss of RHR.

E. Synchronize Main Generator To Line (Manual Sync)

JPM Eis a bank JPM (01002002100). Unit 3 is in MODE 1 with Reactor Power at 6%. 3-GOP-301, Hot Standby to Power Operation, is in progress. The US directs the RO to manually synchronize the Main Generator to the East Bus. The JPM starts with using the new Turbine Control System (TCS). The RO demonstrates field flashing, transferring from manual to auto voltage control, speed control with using TCS.

Once setup conditions are finalized, one Main Generator Breaker is closed and initial load is picked up 10 MWe on the Turbine. If load is not picked up immediately above minimum, then a reverse power relay produces a Main Generator lockout.

F. Start 3B Normal Containment Cooling Fans JPM F is a new JPM. Unit 3 is in MODE; 1 at 100% power. Electrical Maintenance is performing Normal Containment Cooler Breaker maintenance PMs. 3A Normal Containment Cooler is ready for restart. The Unit Supervisor directs restarting 3A Normal Containment Cooler, and then securing 38 Normal Containment Cooler in accordance with 3-NOP-057, Containment Normal Ventilation And Cooling System for testing.

Page 3 of 4

G. Respond To A Source Range Malfunction During Refueling JPM G is a bank JPM (01059029301 ). Unit 3 is in MODE 6 with core alterations in progress. N32 isselected for audio counts and fails high. 3-0NOP-059.5, Source Range Nuclear Instrumentation, is entered. The Audio Count Rate Channel Selector is selected to the operable channel (N31). Verification of other operable channels occurs (Gamma Metrics). The RO notifies plant personnel of the erroneous Containment Evacuation Alarm. Additional switch manipulations are performed to remove the channel from service.

H. Respond To 3A CCW Pump Trip During 3-0SP-030.5. CCW Pumps Low Header Pressure Start Test JPM His a new JPM. Unit 3 is in MODE 1 with Reactor Power at 100%. The US directs the RO to test the 3C CCW Pump in accordance with 3-0SP-030.5, Component Cooling Water Pumps Low Header Pressure Start Test. This test ensures 3A CCW Pump is running. 3C CCW Pump is in Auto. The 3B CCW Pump is disabled by placing in Pull-To-Lock. The line in the field is vented off to simulate a low pressure signal. The alarm for the low pressure comes in, but 3C CCW Pump does not start. (Alternate Path) Concurrently, 3A CCW Pump seizes. The RO must suspend the test to focus on restoration of CCW flow. For success, the 3B CCW or 3C CCW Pump must be manually restarted to restore flow.

I. Respond To A Steam Line Fault Of All SGs And Isolate 3A SG JPM I is a new plant JPM. Unit 3 is in MODE 3 following a steamline line rupture outside Containment on CV-3-1606, 3A Steam Dump To Atmosphere. The crew enters 3-EOP-ECA-2.1, Uncontrolled Depressurization of All Stem Generators, for response. The Control Room loses the ability to close the MSIVs automatically or manually from the Control Room. SG pressure continues to lower. The Unit Supervisor directs the Unit 3 Turbine Operator (TO) to ensure a Secondary Pressure Boundary is established in *accordance starting with POV-3-2604, 3A MSIV. The Unit 3 TO continues with the RNO to perform Attachment 3 for SG 3A isolation.

Since POV-3-2604 is open, air is isolate and vented. Also with CV-3-1606 open, locally action is taken to close 3-10-001 for isolation.

J. Start. Synchronize. and Load 4A (4Bl Emergency Diesel Generator Locally JPM J is a bank JPM (04023021100). Unit 4 is in MODE 3 following a loss of offsite power. Power is restored to one of the 4KV Buses. The crew unsuccessfully attempts to restore power to one bus from an EOG from the Control Room. The Control Room directs the ANPO to check the operational status of 4A (4B) and restore power locally with the 4-0NOP-023.2, Emergency Diesel Generator Failure. The ANPO implements this procedure to take the Master Control Switch to Local, (Alternate Path) restore EOG Day Tank Level, reset the Start Failure Light, and reset the lockout relay. When the EOG does not start, further action is taken to locally start the affected Diesel Generator with the Emergency Stop/Start Switch. When the EOG starts, the EOG Output Breaker does not automatically close. After further adjustments, the ANPO requests permission from the RO to locally energize the associated 4KV Bus. These actions complete the synchronizing and energizing of the bus.

K. Perform A Gaseous Radwaste Release JPM K is a bank JPM (24061006100). Unit 3 and Unit 4 are in MODE 1. O-NOP-061.14B, Waste Gas Disposal System Controlled Release of Gas Decay Tank B, Attachment 1, Controlled Release from Gas Decay Tank B, initial steps are complete. The Auxiliary Building Ventilation System is in service as required.

The operator is directed by the Unit 3 RO to perform the Auxiliary Building actions for a Controlled Radioactive Gas Release from the "B" Gas Decay Tank using O-NOP-061 .14B, Waste Gas Disposal System Controlled Release of Gas Decay Tank B. After started, the release is terminated when the B Gas Decay Tank decreases to the limit by lowering RCV-014 hand loader pressure to 0 and checking RCV-014 valve stem position indicator indicates fully CLOSED.

Page 4 of4

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Turkey Point Units 3 & 4 Date of Examination: 01/19/15 Exam Level: RO D SRO-I 0 SRO-U0 Operating Test No.: 2015-301 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System I JPM Title Type Code* Function A. 001 Control Rod Drive System [A2.11 - RO 4.4 SRO 4. 7]

JPM: Recover A Dropped Rod A,M,S 1 B. 011 Pressurizer Level Control System [A4.01 RO 3.5 SRO 3.2]

JPM: Restore Pressurizer Level After A Loss Of Charging Flow N,S 2 C. 006 Emergency Core Cooling System [A1.13 RO 3.5 SRO 3.7)

JPM: Drain The 38 Accumulator For Boron And Level Adjustment A,M,EN,S 3 D. 005 Residual Heat Removal System [A2.03 RO 2.9 SRO 3.1)

JPM: Respond To A Loss Of RHR A,D,L,S 4P F. 022 Containment Cooling System [A4.01 R03.6 / SR03.6]

JPM: Start 3A Normal Containment Cooler N,S 5 G. 015 Nuclear Instrumentation System [A4.01 RO 3.6* SRO 3.6*)

JPM: Respond To A Source Range Malfunction During Refueling D, L, S 7 H. 008 Component Cooling Water System [A2.01 RO 3.3 SRO 3.6)

JPM: Respond To 3A CCW Pump Trip During 3-0SP-030.5, Component A,N,S 8 Cooling Water Pumps Low Header Pressure Start Test In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

I. E12 Uncontrolled Depressurization Of All Steam Generators

[EA2.2 RO 3.4 SRO 3.9)

N,E 4S JPM: Respond To A Steam Line Fault Of All SGs And Isolate 3A S/G Page 1 of4

J. 056 Loss Of Offsite Power [AA2.14 RO 4.4 SRO 4.6)

JPM: Start, Synchronize and Load 4A (48) Emergency Diesel Generator A,D, E 6 Locally K. 071 Waste Gas Disposal System (A4.26 RO 3.1 SRO 3.9)

JPM: Perform A Gaseous Radwaste Release D, R 9

@All RO and SRO-I Control Room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the Control Room.

• Type Codes Criteria for RO I SRO-I / SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank :59/:58/:54 (E)mergency or abnormal in-plant  :?!1/:?!1/:?!1 (EN)gineered safety feature - I - I :?!1 (control room system)

(L)ow-Power I Shutdown  :?!1/:?!1/:?!1 (N)ew or (M)odified from bank including 1(A)  :?!2/:?!2/:?!1 (P)revious 2 exams :s; 3 I :s; 3 I :s; 2 (randomly selected)

(R)CA  :?!1/:?!1/:?!1 (S)imulator Page 2 of 4

TURKEY POINT 2015 NRC EXAM JPM

SUMMARY

A. Recovery Of A Dropped Rod JPM A is a modified bank JPM. Unit 3 is in MODE 1 escalating power when a rod drops. Current power level is 48%. Shutdown Margin is verified adequate for the current plant conditions. l&C has determined and corrected a degraded condition causing the rod to drop. After the Shift Manager gives concurrence, the Unit Supervisor directs the RO to perform Attachment 1 of 3-0NOP-028.3, Dropped RCC. (Alternate Path) When the rod withdraws, a second rod drops requiring the RO to trip the Reactor. The RO continues with Reactor Trip verification.

B. Restore Pressurizer Level After A Loss Of Charging Flow JPM Bis a new JPM. Unit 3 is in MODE 1 with Reactor Power at 23%. The Pressurizer level trend is at 30%

and lowering with a leak reported on 3A Charging Pump upstream 3A Chrg Pump Discharge, 3-286. During the automatic system response, LC-3-459G, Pressurizer Level Master Control fails low. The Unit Supervisor directs the RO to respond with 3-0NOP-047.1, Loss of Charging Flow in Modes 1 Through 4. The RO starts with isolating two Letdown Orifice Isolation Valves. Then, he closes MOV-3-6386 to isolate Excess UD and RCP Seal Return. He continues by stopping 3A and 3C Charging Pumps for leak isolation. The SNPO is dispatched to isolate 3-286. When the leak is isolated, the RO restores Charging in accordance with Attachment 1 and re-establishes Letdown with Attachment 2.

C. Drain The 3B Accumulator For Boron And Level Adjustment JPM C is a modified JPM (01047021301). Unit 3 is in MODE 1 with Reactor Power at 100%. The ECCS Accumulator has check valve leakage occurring. The 3B Accumulator is inoperable based on level being too high at about 6,890 gallons and boron too low at about 2,050 ppm. The Unit Supervisor directs the RO to perform 3-NOP-064, Safety Injection Accumulators, Section 5.2 to drain 3B Accumulator to 5,000 gallons for boron addition and a level adjustment. The RO verifies the operability of other accumulators (pressure and level) prior to draining 3B Accumulator. When CV-3-852B, 3B Accumulator Drain Valve, is opened for the 3B Accumulator, the CV-3-852A, 3A Accumulator Drain Valve, also leaks by. The Caution before the draining step states only one drain valve shall be opened at a time. (Alternate Path) With the 3A Accumulator level lowering, the RO closes CV-3-852B which allows the leakage associated with CV-3-852A to stop. The 3A Accumulator low pressure annunciator alarms. The response directs the RO to re-pressurize 3A Accumulator.

D. Respond To A Loss Of RHR JPM D is a bank JPM (01050004301 ). Unit 3 is in MODE 4 with RCS temperature just below 350°F. RHR Shutdown Cooling is in service to maintain current conditions. The RO is directed to maintain current conditions. Subsequently, the operating RHR Pump amps fall and no flow is indicated. The RO enters 3-0NOP-050, Loss of RHR for response. The SNPO is dispatched to check the 3A RHR Pump. The report to the Control Room is the 3A RHR Pump's shaft is sheared. The 3B RHR Pump is started with 3-0NOP-050, Loss of RHR.

E. NOT SELECTED FOR SRO-I EXAM.

F. Start 3B Normal Containment Cooling Fans JPM F is a new JPM. Unit 3 is in MODE 1 at 100% power. Electrical Maintenance is performing Normal Containment Cooler Breaker maintenance PMs. 3A Normal Containment Cooler is ready for restart. The Unit Supervisor directs restarting 3A Normal Containment Cooler, and then securing 3B Normal Containment Cooler in accordance with 3-NOP-057, Containment Normal Ventilation And Cooling System for testing.

Page 3 of 4

G. Respond To A Source Range Malfunction During Refueling JPM G is a bank JPM (01059029301). Unit 3 is in MODE 6 with core alterations in progress. N32 isselected for audio counts and fails high. 3-0NOP-059.5, Source Range Nuclear Instrumentation, is entered. The Audio Count Rate Channel Selector is selected to the operable channel (N31 ). Verification of other operable channels occurs (Gamma Metrics). The RO notifies plant personnel of the erroneous Containment Evacuation Alarm. Additional switch manipulations are performed to remove the channel from service.

H. Respond To 3A CCW Pump Trip During 3-0SP-030.5, CCW Pumps Low Header Pressure Start Test JPM H is a new JPM. Unit 3 is in MODE 1 with Reactor Power at 100%. The US directs the RO to test the 3C CCW Pump in accordance with 3-0SP-030.5, Component Cooling Water Pumps Low Header Pressure Start Test. This test ensures 3A CCW Pump is running. 3C CCW Pump is in Auto. The 3B CCW Pump is disabled by placing in Pull-To-Lock. The line in the field is vented off to simulate a low pressure signal. The alarm for the low pressure comes in, but 3C CCW Pump does not start. (Alternate Path) Concurrently, 3A CCW Pump seizes. The RO must suspend the test to focus on restoration of CCW flow. For success, the 3B CCW or 3C CCW Pump must be manually restarted to restore flow.

I. Respond To A Steam Line Fault Of All SGs And Isolate 3A SG JPM I is a new plant JPM. Unit 3 is in MODE 3 following a steamline line rupture outside Containment on CV-3-1606, 3A Steam Dump To Atmosphere. The crew enters 3-EOP-ECA-2.1, Uncontrolled Oepressurization of All Stem Generators, for response. The Control Room loses the ability to close the MSIVs automatically or manually from the Control Room. SG pressure continues to lower. The Unit Supervisor directs the Unit 3 Turbine Operator (TO) to ensure a Secondary Pressure Boundary is established in accordance starting with POV-3-2604, 3A MSIV. The Unit 3 TO continues with the RNO to perform Attachment 3 for SG 3A isolation.

Since POV-3-2604 is open, air is isolate and vented. Also with CV-3-1606 open, locally action is taken to close 3-10-001 for isolation.

J. Start. Synchronize. and Load 4A (4B) Emergency Diesel Generator Locally JPM J is a bank JPM (04023021100). Unit 4 is in MODE 3 following a loss of offsite power. Power is restored to one of the 4KV Buses. The crew unsuccessfully attempts to restore power to one bus from an EDG from the Control Room. The Control Room directs the ANPO to check the operational status of 4A (4B) and restore power locally with the 4-0NOP-023.2, Emergency Diesel Generator Failure. The ANPO implements this procedure to take the Master Control Switch to Local, (Alternate Path) restore EOG Day Tank Level, reset the Start Failure Light, and reset the lockout relay. When the EOG does not start, further action is taken to locally start the affected Diesel Generator with the Emergency Stop/Start Switch. When the EDG starts, the EOG Output Breaker does not automatically close. After further adjustments, the ANPO requests permission from the RO to locally energize the associated 4KV Bus. These actions complete the synchronizing and energizing of the bus.

K. Perform A Gaseous Radwaste Release JPM K is a bank JPM (24061006100). Unit 3 and Unit 4 are in MODE 1. O-NOP-061.14B, Waste Gas Disposal System Controlled Release of Gas Decay Tank B, Attachment 1, Controlled Release from Gas Decay Tank B, initial steps are complete. The Auxiliary Building Ventilation System is in service as required .

The operator is directed by the Unit 3 RO to perform the Auxiliary Building actions for a Controlled Radioactive Gas Release from the "8" Gas Decay Tank using O-NOP-061 .14B, Waste Gas Disposal System Controlled Release of Gas Decay Tank 8. After started, the release is terminated when the B Gas Decay Tank decreases to the limit by lowering RCV-014 hand loader pressure to 0 and checking RCV-014 valve stem position indicator indicates fully CLOSED.

Page4 of 4

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Turkey Point Units 3 & 4 Date of Examination: 01/19/15 Exam Level: RO D SRO-I D SRO-U0 Operating Test No.: 2015-301 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System I JPM Title Type Code* Function C. 006 Emergency Core Cooling System [A1 .13 RO 3.5 SRO 3.7]

JPM: Drain The 38 Accumulator For Boron And Level Adjustment A,M,EN,S 3 D. 005 Residual Heat Removal System [A2.03 RO 2.9 SRO 3.1]

JPM: Respond To A Loss Of RHR A, D, L, S 4P G. 015 Nuclear Instrumentation System [A4.01 RO 3.6* SRO 3.6*]

JPM: Respond To A Source Range Malfunction During Refueling D, L, S 7 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

J. 056 Loss Of Offsite Power [AA2.14 RO 4.4 SRO 4.6]

JPM: Start, Synchronize and Load 4A (48) Emergency Diesel Generator A,D, E 6 Locally K. 071 Waste Gas Disposal System [A4.26 RO 3.1 SRO 3.9]

JPM: Perform A Gaseous Radwaste Release D,R 9 Page 1of3

@All RO and SRO-I Control Room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the Control Room.

• Type Codes Criteria for RO I SRO-I / SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank :S9/:S8/:S4 (E)mergency or abnormal in-plant ~1/~1/~1 (EN)gineered safety feature - I - I ~1 (control room system)

(L)ow-Power I Shutdown ~1/~1/~1 (N)ew or (M)odified from bank including 1(A) ~2/~2/~1 (P)revious 2 exams :s 3 I :s 3 / :s 2 (randomly selected)

(R)CA ~1/~1/~1 (S)imulator Page2 of 3

TURKEY POINT 2015 NRC EXAM JPM

SUMMARY

A. NOT SELECTED FOR SRO-U EXAM.

B. NOT SELECTED FOR SRO-U EXAM.

C. Drain The 3B Accumulator For Boron And Level Adjustment JPM C is a modified JPM (01047021301). Unit 3 is in MODE 1 with Reactor Power at 100%. The ECCS Accumulator has check valve leakage occurring. The 3B Accumulator is inoperable based on level being too high at about 6,890 gallons and boron too low at about 2,050 ppm. The Unit Supervisor directs the RO to perform 3-NOP-064, Safety Injection Accumulators, Section 5.2 to drain 3B Accumulator to 5,000 gallons for boron addition and a level adjustment. The RO verifies the operability of other accumulators (pressure and level) prior to draining 3B Accumulator. When CV-3-852B, 3B Accumulator Drain Valve, is opened for the 3B Accumulator, the CV-3-852A, 3A Accumulator Drain Valve, also leaks by. The Caution before the draining step states only one drain valve shall be opened at a time. (Alternate Path) With the 3A Accumulator level lowering, the RO closes CV-3-852B which allows the leakage associated with CV-3-852A to stop. The 3A Accumulator low pressure annunciator alarms. The response directs the RO to re-pressurize 3A Accumulator.

D. Respond To A Loss Of RHR JPM Dis a bank JPM (01050004301). Unit 3 is in MODE 4 with RCS temperature just below 350°F. RHR Shutdown Cooling is in service to maintain current conditions. The RO is directed to maintain current conditions. Subsequently, the operating RHR Pump amps fall and no flow is indicated. The RO enters 3-0NOP-050, Loss of RHR for response. The SNPO is dispatched to check the 3A RHR Pump. The report to the Control Room is the 3A RHR Pump's shaft is sheared. The 3B RHR Pump is started with 3-0NOP-050, Loss of RHR.

E. NOT SELECTED FOR SRO-U EXAM.

F. NOT SELECTED FOR SRO-U EXAM.

G. Respond To A Source Range Malfunction During Refueling JPM G is a bank JPM (01059029301 ). Unit 3 is in MODE 6 with core alterations in progress. N32 isselected for audio counts and fails high. 3-0NOP-059.5, Source Range Nuclear Instrumentation, is entered. The Audio Count Rate Channel Selector is selected to the operable channel (N31) . Verification of other operable channels occurs (Gamma Metrics). The RO notifies plant personnel of the erroneous Containment Evacuation Alarm. Additional switch manipulations are performed to remove the channel from service.

H. NOT SELECTED FOR SRO-U EXAM.

I. NOT SELECTED FOR SRO-U EXAM.

J. Start. Synchronize. and Load 4A (4B) Emergency Diesel Generator Locally JPM J is a bank JPM (04023021100). Unit 4 is in MODE 3 following a loss of offsite power. Power is restored to one of the 4KV Buses. The crew unsuccessfully attempts to restore power to one bus from an EOG from the Control Room. The Control Room directs the ANPO to check the operational status of 4A (4B) and restore power locally with the 4-0NOP-023.2, Emergency Diesel Generator Failure. The ANPO implements this procedure to take the Master Control Switch to Local, (Alternate Path) restore EOG Day Tank Level, reset the Start Failure Light, and reset the lockout relay. When the EOG does not start, further action is taken to locally start the affected Diesel Generator with the Emergency Stop/Start Switch. When the EOG starts, the EOG Output Breaker does not automatically close. After further adjustments, the ANPO requests permission from the RO to locally energize the associated 4KV Bus. These actions complete the synchronizing and energizing of the bus.

K. Perform A Gaseous Radwaste Release JPM K is a bank JPM (24061006100). Unit 3 and Unit 4 are in MODE 1. O-NOP-061.14B, Waste Gas Disposal System Controlled Release of Gas Decay Tank B, Attachment 1, Controlled Release from Gas Decay Tank B, initial steps are complete. The Auxiliary Building Ventilation System is in service as required.

The operator is directed by the Unit 3 RO to perform the Auxiliary Building actions for a Controlled Radioactive Gas Release from the "B" Gas Decay Tank using O-NOP-061 .14B, Waste Gas Disposal System Controlled Release of ~as Decay Tank B. After started, the release is terminated when the B Gas Decay Tank decreases to the limit by lowering RCV-014 hand loader pressure to 0 and checking RCV-014 valve stem position indicator indicates fully CLOSED.

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