ML17181A354

V.C. Summer 2015-301 Exam Draft Items for ADAMS 2A

ML17181A354
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Site:	Summer
Issue date:	06/30/2017
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ES-301 Administrative Topics Outline Form ES-301-1 NUREG-1021, Revision 9 Facility: VC SUMMER Date of Examination: 5/4/2015 Examination Level (circle one): RO/ SRO Operating Test Number: NRC-ILO-13-01 Administrative Topic (see Note) Type Code* Describe activity to be performed Conduct of Operations (A1-a) R,M RO/SRO Common Determine the volume required to be available in the holdup tanks to accommodate dilution of the RCS to the estimated critical boron concentration. K/A: 2.1.25 (RO: 3.9, SRO: 4.2) K/A: 2.1.37 (RO: 4.3, SRO: 4.6) Conduct of Operations (A1-b) R,N RO/SRO Common Calculate work hour limitations for a covered worker based on a current schedule and additional activities using SAP-0152. K/A: 2.1.5 (RO: 2.9, SRO: 3.9) Equipment Control (A2) R,N Given a loss of a DC power panel affecting annunciators use AOP-100.5 to determine applicable Surveillance requirements. K/A: 2.2.14 (RO: 3.9, SRO: 4.3) Radiation Control (A3) R,D,P RO/SRO Common Calculate the expected dose for two work options in a radiation area with airborne activity and prioritize them according to the VC Summer ALARA philosophy. K/A: 2.3.12 (RO: 3.2, SRO: 3.7) Emergency Plan (A4) Not selected for RO. 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 ( 3 for ROs; 4 for SROs & RO retakes) (N)ew or (M)odified from bank ( 1) (P)revious 2 exams ( 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 NUREG-1021, Revision 9 JPM SUMMARY STATEMENTS CONDUCT OF OPERATIONS (A1-a): This is modified from a JPM in the bank. The plant will be in Mode 3 pending Reactor start up. A current RCS boron concentration and the estimated critical boron concentration will be provided along with a copy of the Curve Book. The candidate will determine the volume required to be available in the holdup tanks to accommodate dilution of the RCS to the estimated critical boron concentration. This JPM will be modified from JPMs in the bank by changing the current and critical boron concentrations. (NJPA-021A) K/A 2.1.25 - Ability to interpret reference materials such as graphs, curves, tables, etc. (RO: 3.9, SRO: 4.2) K/A 2.1.37 - Knowledge of procedures, guidelines, or limitations associated with reactivity management. (RO: 4.3, SRO: 4.6) VCS Task: O-004-006-01-01: Perform boric acid concentration change calculations. CONDUCT OF OPERATIONS (A1-b): This is a new JPM. The candidate will be provided with a work hour history for three covered workers and a current work schedule for the individuals. Additional proposed work activities with planned durations will be provided. The additional activities will require inclusion in the work hour calculation. Candidate will calculate work hour limitations based on the current schedule and the added activities using SAP-0152, Fatigue Management and Work Hour Limits based on requirements of OAP-100.6, Control Room Conduct and Control of Shift Activities. (NJPA-1000) K/A: 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew compliment, overtime limitations, etc. (RO: 2.9, SRO: 3.9) VCS Task: O-341-038-03-02: Interpret and ensure compliance with plant administrative procedures during normal and off normal plant operations. EQUIPMENT CONTROL (A2): This is a new JPM. The candidate will be given a loss of a DC power panel and a copy of AOP-100.5, Loss of Main Control Board Annunciators. The candidate will determine applicable Surveillance requirements for the lost Annunciators using AOP-100.5. (NJPA-1006) K/A: 2.2.14 Knowledge of the process for controlling equipment configuration or status. (RO: 3.9, SRO: 4.3) VCS Task: O-000-170-05-01: Respond to loss of Main Control Board annunciators per AOP-100.5.

ES-301 Administrative Topics Outline Form ES-301-1 NUREG-1021, Revision 9 RADIATION CONTROL (A3): This is a bank JPM that was used on the 2011 NRC Exam. The candidate will compare two options to conduct work in a high radiation area with airborne activity due to a hydrogen explosion in the waste gas system. The candidate will calculate the expected dose for the two options and prioritize them according to the VC Summer ALARA philosophy. (NJPA-083A(R1)) K/A: 2.3.12 - Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirement, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (RO: 3.2,SRO: 3.7) VCS Task: O-000-061-05-01: Respond to area radiation monitoring system alarms. EMERGENCY PLAN (A4): Not selected for RO.

ES-301 Administrative Topics Outline Form ES-301-1 NUREG-1021, Revision 9 Facility: VC SUMMER Date of Examination: 5/4/2015 Examination Level (circle one): RO/ SRO Operating Test Number: NRC-ILO-13-01 Administrative Topic (see Note) Type Code* Describe activity to be performed Conduct of Operations (A1-a) R,M RO/SRO Common Determine the volume required to be available in the holdup tanks to accommodate dilution of the RCS to the estimated critical boron concentration. K/A: 2.1.25 (RO: 3.9, SRO: 4.2) K/A: 2.1.37 (RO: 4.3, SRO: 4.6) Conduct of Operations (A1-b) R,N RO/SRO Common Calculate work hour limitations for a covered worker based on a current schedule and additional activities using SAP-0152. K/A: 2.1.5 (RO: 2.9, SRO: 3.9) Equipment Control (A2) R,D Determine administrative actions required for transfer of 7.2Kv bus1DB to alternate feed using SAP-205. K/A: 2.2.14 (RO: 3.9, SRO: 4.3) Radiation Control (A3) R,D,P RO/SRO Common Calculate the expected dose for two work options in a radiation area with airborne activity and prioritize them according to the VC Summer ALARA philosophy. K/A: 2.3.12 (RO: 3.2, SRO: 3.7) Emergency Plan (A4) S,N Declare a Site Area Emergency in accordance with EPP-001 and complete the EPP-002 notification form. K/A: 2.4.41 (RO: 2.9 ,SRO: 4.6 ) K/A: 2.4.40 - (RO: 2.7 ,SRO: 4.5 ) 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 ( 3 for ROs; 4 for SROs & RO retakes) (N)ew or (M)odified from bank ( 1) (P)revious 2 exams ( 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 NUREG-1021, Revision 9 JPM SUMMARY STATEMENTS CONDUCT OF OPERATIONS (A1-a): This is modified from a JPM in the bank. The plant will be in Mode 3 pending Reactor start up. A current RCS boron concentration and the estimated critical boron concentration will be provided along with a copy of the Curve Book. The candidate will determine the volume required to be available in the holdup tanks to accommodate dilution of the RCS to the estimated critical boron concentration. This JPM will be modified from JPMs in the bank by changing the current and critical boron concentrations. (NJPA-021A) K/A 2.1.25 - Ability to interpret reference materials such as graphs, curves, tables, etc. (RO: 3.9, SRO: 4.2) K/A 2.1.37 - Knowledge of procedures, guidelines, or limitations associated with reactivity management. (RO: 4.3, SRO: 4.6) VCS Task: O-004-006-01-01: Perform boric acid concentration change calculations. CONDUCT OF OPERATIONS (A1-b): This is a new JPM. The candidate will be provided with a work hour history for three covered workers and a current work schedule for the individuals. Additional proposed work activities with planned durations will be provided. The additional activities will require inclusion in the work hour calculation. Candidate will calculate work hour limitations based on the current schedule and the added activities using SAP-0152, Fatigue Management and Work Hour Limits based on requirements of OAP-100.6, Control Room Conduct and Control of Shift Activities. (NJPA-1000) K/A: 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew compliment, overtime limitations, etc. (RO: 2.9, SRO: 3.9) VCS Task: O-341-038-03-02: Interpret and ensure compliance with plant administrative procedures during normal and off normal plant operations. EQUIPMENT CONTROL (A2): This JPM is direct from the bank. The candidate will determine administrative actions to place 7.2Kv bus 1DB on alternate feed. Candidate will complete SAP-205, Status Control and Removal and Restoration, Attachment 1, Removal and Restoration Checksheet for XSW1DB 16, BUS 1DB NORMAL INCOMING BKR to track all the requirements associated with transferring Bus 1DB to XTF-4/6. It will be critical to indicate that TS 3.8.1.1.a and 3.0.4 do apply. (NJPA-210A) K/A: 2.2.14 - Knowledge of the process for controlling equipment configuration or status. (RO: 3.9, SRO: 4.3) VCS Task: O-341-038-03-02: Interpret and ensure compliance with plant administrative procedures during normal and off normal plant operations.

ES-301 Administrative Topics Outline Form ES-301-1 NUREG-1021, Revision 9 RADIATION CONTROL (A3): This is a bank JPM that was used on the 2011 NRC Exam. The candidate will compare two options to conduct work in a high radiation area with airborne activity due to a hydrogen explosion in the waste gas system. The candidate will calculate the expected dose for the two options and prioritize them according to the VC Summer ALARA philosophy. (NJPA-083A(R1)) K/A: 2.3.12 - Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirement, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (RO: 3.2,SRO: 3.7) VCS Task: O-000-061-05-02: Respond to area radiation monitoring system alarms. EMERGENCY PLAN (A4): This is a new JPM. The candidate will declare a Site Area Emergency in accordance with EPP-001, Activation and Implementation of Emergency Plan. The candidate will declare the Site Area Emergency due to an inadequate core cooling event that results in a potential loss of the Fuel Clad Barrier and a loss of the Reactor Coolant System Barrier. The candidate will also be required to complete the EPP-002, Communication and Notification, Attachment I, Nuclear Power Plant Notification Form. This is a time critical JPM. (NJPA-1003) K/A: 2.4.41 - Knowledge of the emergency action level thresholds and classifications. (RO: 2.9, SRO: 4.6) K/A: 2.4.40 - Knowledge of the SRO's responsibilities in emergency plan implementation. (RO: 2.7, SRO: 4.5) VCS Task: O-344-019-03-02: Classify Events requiring Emergency Plan Implementation ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 Facility: VC Summer Date of Examination: 5/4/2015 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: NRC-ILO-13-01 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function a. Generic Abnormal Plant Evolution 001 [NJPSF-141A] Continuous Rod Withdrawal. (AOP-403.3, EOP-1.0, E-0) P,D,A,L,S 1 b. System 010 [NJPSF-007A] Steam Generator Tube Rupture (Depressurize RCS to < Ruptured S/G pressure). (EOP-4.0, E-3) M,A,L,S 3 c. Generic Abnormal Plant Evolution 025 [NJPS-065] Establish hot leg injection during loss of RHR at mid-loop conditions. (AOP-115.5) D,L,S 4P d. System 026 [NJPSF-019A] Manually initiate Reactor Building Spray. (EOP-1.0, E-0) M,A,L,S,EN 5 e. System 064 [NJPSF-025A] Start and load "A" Emergency Diesel Generator. (SOP-306) M,S,A 6 f. System 016 [NJPS-1000] Respond to Steam Generator Pressure Channel malfunction. (AOP-401.3) N,S 7 g. System 033 [NJPS-084] Restore Spent Fuel Pool level during refueling. (AOP-123.1) D,L,S 8 h. System 029 [NJPS-1001] Establish Reactor Building Purge Supply and Exhaust. (SOP-114) N,L,S 9 In-Plant Systems @ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. Generic Emergency Plant Evolution 055 [NJPP-402] Locally Dilute the Boric Acid Tanks (EOP-6.0, ECA-0.0) D,L,E,R 1 j. Generic Abnormal Plant Evolution 068 [NJPPF-049] Control Room evacuation (duties of BOP operator) (AOP-600.1) P,D,A,L,E 4S k. System 062 [NJPP-040] Transfer a Vital 120 volt Instrument Power Supply (SOP-310) D 6 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 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 / SRO-I / SRO-U RO (A)lternate path 4-6 / 4-6 / 2-3 5 (C)ontrol room (D)irect from bank 9 / 8 / 4 6 (E)mergency or abnormal in-plant 1 / 1 / 1 2 (EN)gineered safety feature NA / NA / 1(control room system) (L)ow-Power 1 / 1 / 1 8 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 5 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 2 (R)CA 1 / 1 / 1 1 (S)imulator 8 VC SUMMER 2015 NRC JPM SUMMARY a. This is a bank JPM that was previously used on the 2013 License exam. The candidate will respond to continuous rod withdrawal in accordance with AOP-403.3, Continuous Control Rod Motion, and EOP-1.0 (E-0), Reactor Trip/Safety Injection Actuation. The candidate will be told to withdraw rods to criticality (Shutdown Banks will be withdrawn, but Control Banks will be fully inserted). Candidate will withdraw Control Bank A to 10 steps and verify indications that all the Control Bank A rods came off the bottom. On pulling rods so that rods reach 103 steps, rods will continue to withdraw with no operator input. A failure of all automatic trips will make it critical that the candidate trip the reactor. The continuous rod motion is where this JPM becomes alternate path. The critical step will be to pull rods, then to trip the reactor when uncontrolled rod motion occurs prior to reaching the Estimated Critical Position of 100 steps on bank D. K/A 001AA1.05: Ability to operate and/or monitor the following as they apply to the Continuous Rod Withdrawal: Reactor Trip switches (RO: 4.3, SRO: 4.2) NUREG 1122 APE: Continuous rod withdrawal VCS Task: O-000-006-05-01: Respond to Continuous Rod Motion per AOP-403.3/SOP-403.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 b. This JPM is modified from a bank JPM. The candidate will take actions to depressurize the RCS to less than the pressure of the ruptured steam generator in accordance with EOP-4.0 (E-3), Steam Generator Tube Rupture. The candidate will attempt to open the only available pressurizer spray valve to depressurize the RCS but the valve will fail to open. This leads to the alternate path for this JPM. In order to accomplish the depressurization, the applicant will have to utilize a Pressurizer PORV. When criteria to stop the depressurization are met the chosen PORV will not shut. The candidate must then close the associate PORV block valve. This JPM is significantly modified from another JPM in the bank by failing the spray valve closed and by the failure of the selected PORV to close when demanded. The critical step is closing the PORV block valve to terminate the RCS depressurization. K/A 010000A203: Pressurizer Pressure Control System (PZR PCS). Ability to (a) predict the impacts of PORV failures on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PORV failures. (RO: 4.1, SRO: 4.2) NUREG 1122 System: Pressurizer Pressure Control System (PZR PCS) VCS Task: O-000-038-05-01: Respond to Steam Tube Rupture per EOP-4.0. c. This is a bank JPM. The initial conditions for this JPM have the plant in Mode 5 and RCS at Mid-loop conditions. Due to lowering hot leg level, the crew will have entered AOP-115.1, RHR Pump Vortexing and then AOP-115.5, Loss of RHR with the RCS not Intact (Modes 5 and 6). The candidate will be given the following parameters: RCS hot leg level off scale low, core exit TC temperatures >200°F and increasing and the Charging pump available. The candidate will be directed to implement AOP-115.5 Attachment 2, Establishing Hot Leg Injection as an alternative action from AOP-115.5 step 17. Candidate should manually align hot leg injection, start A Charging pump and raise hot leg level. The c K/A 000025K301: Knowledge of the reasons for the following responses as they apply to the Loss of Residual Heat Removal System: Shift to alternate flow path (RO: 3.1, SRO: 3.4) NUREG 1122 APE: Loss of Residual Heat Removal System VCS Task: O-000-083-05-01: Respond to Loss of Residual Heat Removal System While at Mid-loop Conditions per AOP-115.5/SOP-115.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 d. This JPM is modified from a bank JPM. The initial conditions for this JPM are a Reactor Trip from 100% power with Safety Injection actuated. Neither train of Reactor Building spray will have auto actuated and RB pressure will be greater than 12.0 psig. The Candidate will be directed to perform Step 8 of EOP-1.0 (E-0), Reactor Trip/Safety Injection Actuation. The Train A RB Spray manual actuation will fail to operate. If Train is manually actuated then RB Spray Pump must also be manually opened. attempted then the candidate must manually align flow paths (Spray and Phase B) and start RB Spray pumps. Candidate will then verify RB Spray flow and ensure that all RCPs are stopped. Manual alignment of the required equipment is where the JPM becomes alternate path. This JPM is significantly modified from another JPM in the bank by Critical steps will be to manually actuate at least one train of containment spray with >2500 gpm per EOP-1.0 and to secure RCPs to prevent damage to RCP motors due to loss of CCW as evident from Motor Bearing temperature exceeding 195°F or Lower Seal Water Bearing temperature exceeding 225°F or Seal Water Outlet temperature exceeding 235°F. K/A 026000A4.01 Ability to manually operate and/or monitor in the Control Room: CSS controls (RO: 4.5, SRO: 4.3) NUREG 1122 System Containment Spray System (CSS) VCS Task: O-026-005-01-01: Manually Initiate Reactor Building Spray per SOP-116/EOP1.0.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 e. This JPM is modified from a bank JPM. The initial conditions for this JPM have the plant operating at 100% power with normal AC power available to all buses. Relay testing will be in progress on 7.2Kv Switchgear bus 1DA and this has necessitated the removal of Bus 1DA from its normal and emergency feed. The "A" D/G is to be started and loaded onto bus 1DA. All pre-start check steps been completed. The candidate will be directed to start and load "A" D/G per SOP-306, Emergency Diesel Generator,Section IV.A, Steps 2.2.j. After starting the Diesel, Annunciator panel 636, 6-3, DG A ENG TEMP TRBL will light due to high lube oil temperature (167°F). Reports of high lube oil temperature (170°F rising, 176°F rising) will be provided to the candidate once the Diesel is started and after the alarm is lit. The Diesel will not trip automatically at 175°F Lube Oil Temperature. Candidate should trip the diesel prior to parallel due to failure of the automatic trip on high Lube Oil temperature. This JPM becomes alternate path upon reaching 175°F Lube Oil temperature with no automatic trip. This JPM is significantly modified from another JPM in the bank by adding the high Lube Oil temperature condition and auto trip failure prior to parallel. The critical step will be to manually trip the Diesel once the auto trip fails and prior to parallel. K/A 064000A401 Ability to manually operate and/or monitor in the control room: Local and remote operation of the ED/G (RO: 4.0, SRO: 4.3) K/A 064000A101 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including: ED/G lube oil temperature and pressure. (RO: 3.0, SRO: 3.1) NUREG 1122 System Emergency Diesel Generators VCS Task: O-064-003-01-01: Load the Diesel Generator. f. This is a new JPM. The initial conditions for this JPM have the plant operating at 100% powerSteam Generator pressure channel PT-485 will fail high. Since the pressure channel compensates the controlling Steam Flow channel this will cause the Steam Flow signal to fail high he FRV will then travel open in auto in response to the failed input. The candidate implements AOP-401.3, Steam Flow Feedwater Flow Protection Channel Failure and performs immediate actions. The operable Steam Flow and Feed Flow channels will be selected; Turbine load will be reduced 40 50 MWe, feed flow and Feedwater Pump speed will be adjusted as necessary. Candidate verifies SG level is within band and returns controls to auto. Candidate will identify the failed channel. The critical step will be to restore Feedwater ore reaching turbine trip criteria on high SG level. K/A 059000A211 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of feedwater control system. (RO: 3.0, SRO: 3.3) NUREG 1122 System Main Feedwater (MFW) System VCS Task: O-000-103-05-01: Respond to Excessive Feedwater Increase per AOP-401.3 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 g. This is a bank JPM. The initial conditions for this JPM have the plant in Mode 6 with core off load in progress. The 'A' RHR loop is in service providing core cooling. Due to decreasing level in the Spent fuel Pool, AOP-123.1, Decreasing Level in the Spent Fuel Pool or Refueling Cavity During Refueling has been entered. The leakage has been isolated in step 8. The candidate will be directed to respond to a decreasing level in the spent fuel pool in accordance with AOP-123.1. The critical steps are aligning RHR pump suction to the RWST and isolating RHR suction from the RCS. K/A 033000A203 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal spent fuel pool water level or loss of water level. (RO: 3.1, SRO: 3.5) NUREG 1122 System Spent Fuel Pool Cooling System (SFPCS) VCS Task: O-000-140-05-01: Respond to decreasing Water Level in the Spent Fuel Pool or Refueling Cavity per AOP-123.5/AOP-123.1. h. This is a new JPM. The initial conditions for this JPM have the plant in Mode 5 preparing for a refueling outage. The equipment hatch is open. The candidate will be directed to place Reactor Building purge in service using SOP-114, Reactor Building Ventilation System. After entering SOP-114 section III.C candidate will proceed with placing both RB Purge exhaust fans in service. Candidate should then start no more than one Purge supply fan to maintain negative pressure on the RB. The critical steps will be to start the exhaust fans and no more than one supply fan to maintain negative pressure on the RB and prevent an unmonitored release. K/A 029000A201 Ability to (a) predict the impacts of the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Maintenance or other activity taking place inside containment (RO: 2.9, SRO: 3.6) NUREG 1122 System Containment Purge System (CPS) VCS Task: O-088-505-01-04 Perform Line ups of the Reactor Building Ventilation Systems.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 i. This is a bank JPM. The initial conditions for this JPM have the plant shutdown and experiencing a Loss of all Offsite and Onsite AC power. Procedure EOP-6.0 (ECA-0.0) Loss of All ESF AC Power has been implemented. Annunciator panel 613, 4-2, "BAT A TEMP HI/LO" is standing (low setpoint 70°F) and local verification has indicated that using EOP-6.0, Attachment 6, Locally Diluting the Boric Acid Tanks. The candidate will simulate the following actions; connect the drain rig, open the drain isolation valve, drain the BAT to 50%, close the drain and remove the rig. The candidate then simulates flushing a nearby fire hose to a floor drain and simulates connecting the fill rig and use of the fire -95%. The critical step will be assuring that desired volume of fire water is added to the BAT and that the drain is isolated. K/A 000055K302 Knowledge of the reasons for the following responses as they apply to the Station Blackout: Actions Contained in EOP for loss of offsite and onsite power (RO: 4.3, SRO: 4.6) NUREG 1122 EPE Loss of Offsite and Onsite Power (Station Blackout) VCS Task: O-000-055-05-01 Respond to loss of offsite and on site ESF power per EOP-6.0/EOP-1.0 j. This is a bank JPM that was last used on the 2011 License exam. The initial conditions for this JPM have the reactor tripped due to the need to evacuate the control room. Evacuation is necessary due to a bomb threat and no equipment will have been tripped from the MCB. RCP will have been tripped. The candidate will be directed to take actions in accordance with AOP-600.1, Control Room Evacuation, Attachment 2 Duties of the BOP Operator. Candidate will simulate locally tripping the Main Feedwater pumps and MG set. Using a photograph of the inside of a 7.2 Kv breaker the candidate will simulate stopping two Condensate pumps by opening the breakers and three of four Feedwater Booster pumps by opening the breakers. The alternate path for the RO is that the A RCP is tripped and so B RCP will have to be left running. The critical step K/A 0000682130 Conduct of Operations: Ability to locate and operate components, including local controls (RO: 4.4, SRO: 4.0) NUREG 1122 APE Control Room Evacuation VCS Task: 0-000-068-05-01 Perform Control Room Evacuation per AOP-600.1.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 k. This is a bank JPM. The initial conditions for this JPM have the plant at 100% power. Vital AC Inverter, XIT-5901 is scheduled for preventive maintenance. The candidate is directed to remove XIT-5901 from service and place Vital AC distribution panel, APN-5901 on its alternate power source in accordance with SOP-310, Engineered Safety Features 120 VAC Instrumentation and Control Power System,Section IV.I. Initial conditions have been completed through step 1.4. Candidate simulate placing the Test Transfer switch to the ALT position and verifies ON Alternate light illuminates and the ON Inverter light goes out. Candidate simulates depressing the Inverter STOP push button and verifies the SYNCH MONITOR light is lit. Simulates placing the MAN Bypass switch to BYP TO ALT position. Simulates opening the Backup Source breaker and the Normal AC Source breakers. The critical step is to place the Test Transfer switch to the ALT position. K/A 062000A203 Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Consequences of improper sequencing when transferring to or from an inverter. (RO: 2.9, SRO: 3.4) NUREG 1122 System A.C. Electrical Distribution VCS Task: O-062-010-01-04 Remove Engineering Safety Features Vital Inverter from Service.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 Facility: VC Summer Date of Examination: 5/4/2015 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: NRC-ILO-13-01 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function a. Not selected for SRO-U b. System 010 [NJPSF-007A] Steam Generator Tube Rupture (Depressurize RCS to < Ruptured S/G pressure). (EOP-4.0, E-3) M,A,L,S 3 c. Not selected for SRO-U d. System 026 [NJPSF-019A] Manually initiate Reactor Building Spray. (EOP-1.0, E-0) M,A,L,S,EN 5 e. Not selected for SRO-U f. Not selected for SRO-U g. Not selected for SRO-U h. Not selected for SRO-U In-Plant Systems @ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. Generic Emergency Plant Evolution 055 [NJPP-402] Locally Dilute the Boric Acid Tanks (EOP-6.0, ECA-0.0) D,L,E,R 1 j. Generic Abnormal Plant Evolution 068 [NJPPF-049] Control Room evacuation (duties of BOP operator) (AOP-600.1) P,D,A,L,E 4S k. System 062 [NJPP-040] Transfer a Vital 120 volt Instrument Power Supply (SOP-310) D 6 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 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 / SRO-I / SRO-U SRO - U (A)lternate path 4-6 / 4-6 / 2-3 3 (C)ontrol room (D)irect from bank 9 / 8 / 4 3 (E)mergency or abnormal in-plant 1 / 1 / 1 2 (EN)gineered safety feature NA / NA / 1(control room system) 1 (L)ow-Power 1 / 1 / 1 4 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 2 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 1 (R)CA 1 / 1 / 1 1 (S)imulator 2 VC SUMMER 2015 NRC JPM SUMMARY a. Not selected for SRO-U b. This JPM is modified from a bank JPM. The candidate will take actions to depressurize the RCS to less than the pressure of the ruptured steam generator in accordance with EOP-4.0 (E-3), Steam Generator Tube Rupture. The candidate will attempt to open the only available pressurizer spray valve to depressurize the RCS but the valve will fail to open. This leads to the alternate path for this JPM. In order to accomplish the depressurization, the applicant will have to utilize a Pressurizer PORV. When criteria to stop the depressurization are met the chosen PORV will not shut. The candidate must then close the associate PORV block valve. This JPM is significantly modified from another JPM in the bank by failing the spray valve closed and by the failure of the selected PORV to close when demanded. The critical step is closing the PORV block valve to terminate the RCS depressurization. K/A 010000A203: Pressurizer Pressure Control System (PZR PCS). Ability to (a) predict the impacts of PORV failures on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PORV failures. (RO: 4.1, SRO: 4.2) NUREG 1122 System: Pressurizer Pressure Control System (PZR PCS) VCS Task: O-000-038-05-01: Respond to Steam Tube Rupture per EOP-4.0. c. Not selected for SRO-U ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 d. This JPM is modified from a bank JPM. The initial conditions for this JPM are a Reactor Trip from 100% power with Safety Injection actuated. Neither train of Reactor Building spray will have auto actuated and RB pressure will be greater than 12.0 psig. The Candidate will be directed to perform Step 8 of EOP-1.0 (E-0), Reactor Trip/Safety Injection Actuation. The Train A RB Spray manual actuation will fail to operate. If Train is manually actuated then RB Spray Pump must also be manually opened. attempted then the candidate must manually align flow paths (Spray and Phase B) and start RB Spray pumps. Candidate will then verify RB Spray flow and ensure that all RCPs are stopped. Manual alignment of the required equipment is where the JPM becomes alternate path. This JPM is significantly modified from another JPM in the bank by Critical steps will be to manually actuate at least one train of containment spray with >2500 gpm per EOP-1.0 and to secure RCPs to prevent damage to RCP motors due to loss of CCW as evident from Motor Bearing temperature exceeding 195°F or Lower Seal Water Bearing temperature exceeding 225°F or Seal Water Outlet temperature exceeding 235°F. K/A 026000A4.01 Ability to manually operate and/or monitor in the Control Room: CSS controls (RO: 4.5, SRO: 4.3) NUREG 1122 System Containment Spray System (CSS) VCS Task: O-026-005-01-01: Manually Initiate Reactor Building Spray per SOP-116/EOP1.0. e. Not selected for SRO-U f. Not selected for SRO-U g. Not selected for SRO-U h. Not selected for SRO-U ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 i. This is a bank JPM. The initial conditions for this JPM have the plant shutdown and experiencing a Loss of all Offsite and Onsite AC power. Procedure EOP-6.0 (ECA-0.0) Loss of All ESF AC Power has been implemented. Annunciator panel 613, 4-2, "BAT A TEMP HI/LO" is standing (low setpoint 70°F) and local verification has indicated that temperature is 68°F in 'A' BAT room. The candidate will be directed to dilutusing EOP-6.0, Attachment 6, Locally Diluting the Boric Acid Tanks. The candidate will simulate the following actions; connect the drain rig, open the drain isolation valve, drain the BAT to 50%, close the drain and remove the rig. The candidate then simulates flushing a nearby fire hose to a floor drain and simulates connecting the fill rig and use of -95%. The critical step will be assuring that desired volume of fire water is added to the BAT and that the drain is isolated. K/A 000055K302 Knowledge of the reasons for the following responses as they apply to the Station Blackout: Actions Contained in EOP for loss of offsite and onsite power (RO: 4.3, SRO: 4.6) NUREG 1122 EPE Loss of Offsite and Onsite Power (Station Blackout) VCS Task: O-000-055-05-01 Respond to loss of offsite and on site ESF power per EOP-6.0/EOP-1.0 j. This is a bank JPM that was last used on the 2011 License exam. The initial conditions for this JPM have the reactor tripped due to the need to evacuate the control room. Evacuation is necessary due to a bomb threat and no equipment will have been tripped will have been tripped. The candidate will be directed to take actions in accordance with AOP-600.1, Control Room Evacuation, Attachment 2 Duties of the BOP Operator. MG set. Using a photograph of the inside of a 7.2 Kv breaker the candidate will simulate stopping two Condensate pumps by opening the breakers and three of four Feedwater RCP will have to be left running. The critical step K/A 0000682130 Conduct of Operations: Ability to locate and operate components, including local controls (RO: 4.4, SRO: 4.0) NUREG 1122 APE Control Room Evacuation VCS Task: 0-000-068-05-01 Perform Control Room Evacuation per AOP-600.1.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9 k. This is a bank JPM. The initial conditions for this JPM have the plant at 100% power. Vital AC Inverter, XIT-5901 is scheduled for preventive maintenance. The candidate is directed to remove XIT-5901 from service and place Vital AC distribution panel, APN-5901 on its alternate power source in accordance with SOP-310, Engineered Safety Features 120 VAC Instrumentation and Control Power System,Section IV.I. Initial conditions have been completed through step 1.4. Candidate simulate placing the Test Transfer switch to the ALT position and verifies ON Alternate light illuminates and the ON Inverter light goes out. Candidate simulates depressing the Inverter STOP push button and verifies the SYNCH MONITOR light is lit. Simulates placing the MAN Bypass switch to BYP TO ALT position. Simulates opening the Backup Source breaker and the Normal AC Source breakers. The critical step is to place the Test Transfer switch to the ALT position. K/A 062000A203 Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Consequences of improper sequencing when transferring to or from an inverter. (RO: 2.9, SRO: 3.4) NUREG 1122 System A.C. Electrical Distribution VCS Task: O-062-010-01-04 Remove Engineering Safety Features Vital Inverter from Service.

The 2015 V.C. Summer written exam was drafted by the NRC utilizing various exam authors.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 - Scenario 1 NUREG -1021 R9 S1 Facility: VC SUMMER Scenario No: 1 Op Test No: NRC-ILO-13-01 Examiners: Operators: CRS: RO: BOP: Initial Conditions: The plant has completed a Mid-Cycle outage. The Reactor is Critical at 10-3 % power. Critical Data has been recorded. The National Weather Service has declared a Severe Weather Warning for Richland, Fairfield, and Kershaw counties for the next four (4) hours. The secondary has been warmed. B1 Train Work Week. Alternate Seal Injection is OOS. Turnover: Following turnover start RBCU 2B, then secure RBCU 1B per an Engineering Request to monitor the RBCU 2B. Following turnover raise Reactor Power to between 1% and 3%. Critical Tasks: Maintain SG levels using EFW without causing a Reactor trip. Align at least one CHG/SI flowpath prior to ORANGE path on Core Cooling. Isolate LOCA prior to exiting EOP-2.5. Event No. Malf No. Event Type* Event Description 1 NA N-BOP Start RBCU 2B, then secure RBCU 1B. 2 NA R-RO, N-CRS Raise power to between 1% and 3%. 3 EF002B EF002T C-BOP, CRS TS-CRS MD EFW B Pump Bearing Failure leading to trip of the pump. 4 EPS005C EPS006B C-BOP, CRS TS-CRS Loss of Emergency Auxiliary transformer (1DB). DG fails to AUTO start. 5 CRF004F8 CRF007 C-RO, CRS TS-CRS Partially Dropped Rod (F8) Rod slips to approximately 200 steps withdrawn.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 - Scenario 1 NUREG -1021 R9 S1 6 CRF004F8 CRF004D4 CRF007 Two Dropped Rods (F8 and D4) Trip the reactor 7 CVC015A I-RO, CRS Letdown pressure control valve PCV-145 fails CLOSED (AUTO ONLY). 8 RHR013E RHR013B RHR011 M-ALL LOCA Outside the Reactor Building. PCS005A Auto and Manual). Manually configure Pumps and Valves. CS004P CS006F CHG/SI pump trips (cannot be reset). CHG/SI pump fails to Auto-Start. Manually Start. pump breaker cannot be racked-up. * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 NRC 2015 - Scenario 1 NUREG -1021 R9 S1 The following notation is used in the ES-D- IOA designates Immediate Operator Action steps.

• designates Continuous Action steps. The crew will assume the watch having been pre-briefed on the Initial Conditions, the plan for this shift and any related operating procedures. The scenario involves a plant startup so GOP-3, Reactor Startup from Hot Standby to Startup (Mode 3 To Mode 2), is being implemented. Step 3.13, recording Critical Data, has been completed and the reactor is critical in Mode 2 at 10E-3% power. The secondary plant has been warmed with the turbine on turning gear. The simulator will be frozen until the crew assumes the watch. The crew will be briefed to have the BOP start the 2B RBCU and then stop the 1B RBCU following turnover due to a request from Engineering. The crew will then increase to between 1% and 3% in accordance with GOP-3 beginning at step 3.14. GTP-702 Attachment II G, Operational Mode Change Plant Startup - Entering Mode 1, has been completed. Sections of GOP-4A, Power Operation (Mode 1 - Ascending) have been completed to perform initial lineups and to warm the secondary plant. EVENT 1: Start RBCU 2B, then secure RBCU 1B. Three Reactor Building Cooling Units (RBCUs) will be running in fast speed at turnover. After turnover is complete, the BOP will start the 2B RBCU in fast speed then secure the 1B RBCU in accordance with SOP-114, Reactor Building Ventilation System. EVENT 2: Raise power to between 1% and 3%. The RO will increase Reactor Power by withdrawing control rods. The RO will recognize the negative reactivity feedback as the Point of Adding Heat is achieved and stabilize power between 1-3%. The BOP will adjust Emergency Feedwater flow to the Steam Generators as steam flow increases. The crew will transition to GOP-4A, Power Operation (Mode 1 - Ascending).

Appendix D Scenario Outline Form ES-D-1 NRC 2015 - Scenario 1 NUREG -1021 R9 S1 EVENT 3: TRIGGER 1 o PMP-EF002B XPP0021B MOTOR DRIVEN EFW PMP B BRG FAILURE RAMP = 5 seconds FINAL = 10 o PMP-EF002T XPP0021B MOTOR DRIVEN EFW PMP B TRIP ON COMMAND DELAY = 35 seconds On cue from the Examiner at approximately 2-3% power the B MDEFW bearing will fail and the EFW Pump will trip after a short delay if not stopped by the BOP. In accordance with XCP-623 1-5, MD EFP B MOTOR OVRLD and XCP-623 1-3, MD EFP B Trip, EFW flow must be reduced to below 400 gpm. The EFW flow requirement is approximately 180 gpm/percent so power is limited Aunless the crew decides to use the TDEFW Pump. The BOP will throttle EFW flow to the SGs using the A MDEFW in accordance with SOP-211, Emergency Feedwater System. The CRS will evaluate the failure and determine that the B MDEFW Pump is inoperable. The CRS will refer to Technical Specification 3.7.1.2, Emergency Feedwater System. EVENT 4: Loss of Emergency Auxiliary Transformer (1DB), DG fails to AUTO Start TRIGGER 2 o MAL-EPS006B DIESEL GENERATOR B FAILURE FAIL TO: No Auto Start o MAL-EPS005C LOSS OF ESF BUS 1DB (NORMAL FEED BREAKER) DELAY = 1 second On cue from the Examiner, power will be lost to the 1DB bus due to a breaker failure and the Auto-Start failure of the B onto the 1DB Bus. The CRS will refer to Technical Specification 3.8.1.1, AC Sources.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 - Scenario 1 NUREG -1021 R9 S1 EVENT 5: Partially Dropped Rod (F8) Rod slips to approximately 200 steps withdrawn. TRIGGER 3 o MAL-CRF004F8 DROPPED ROD F8 FINAL = STATIONARY DELETE = 1 second o LOA-CRF007 CONTROL ROD F8 STICKING POSITION FINAL VALUE = 200 DELETE = 1 second On cue from the Examiner, F8 will slip to the 200 steps withdrawn position. The RO will take immediate actions in accordance with AOP-403.6, Dropped Control Rod, by verifying that only one Control Rod has slipped and by placing the Rod Cntrl Bank Sel Switch in Manual. The CRS will refer to Technical Specifications: 3.1.1.1 Shutdown Margin, 3.1.3.1 Group Height, Insertion and Power Distribution Limits, 3.1.3.6 Control Rod Insertion Limits, and 3.2.4 Quadrant Power Tilt Ratio. EVENT 6: Two Dropped Rods (F8 and D4) Trip the Reactor TRIGGER 4 o LOA-CRF007 CONTROL ROD F8 STICKING POSITION (NOTE: This LOA is inserted allow F8 to fall when dropped) FINAL VALUE = 0 o MAL-CRF004F8 DROPPED ROD F8 FINAL = STATIONARY o MAL-CRF004D4 DROPPED ROD D4 FINAL =STATIONARY After Technical Specifications have been addressed for a single dropped rod the examiner will cue the booth operator to drop a second rod. The RO will trip the Reactor and implement EOP-1.0 (E-0) Reactor Trip/Safety Injection Actuation in accordance with AOP-403.6, Dropped Control Rod.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 - Scenario 1 NUREG -1021 R9 S1 EVENT 7: Letdown pressure control valve PCV-145 fails CLOSED (AUTO ONLY). TRIGGER 5 o MAL-CVC015A LETDOWN PRESSURE CONTROL VALVE PCV-145 FAILURE (AUTO ONLY) RAMP = 5 seconds FINAL = 35 On cue from the Examiner, PCV-145 pressure controller will drift LOW in auto causing letdown pressure to increase to the alarm setpoint. The RO will respond to annunciators, XCP-613 2-4, LP LTDN FLO/PRESS HI and take manual control of PCV-145. EVENT 8: LOCA Outside RB, CHG/SI CHG/SI Pump Fails to Auto-Start TRIGGER 6 o MAL-PCS005A SAFETY INJECTION FAILURE TRAIN A FAIL TO: Total Failure o MAL-RHR013E RHR DISCH CHECK VALVE 8973C LEAKAGE (0.05=800 GPM) SEVERITY = 0.05 o MAL-RHR013B RHR DISCH CHECK VALVE 8974B LEAKAGE (0.05=800 GPM) SEVERITY = 0.05 o FLX-RHR011 FLEX LEAK RLF VLV 8864B SEVERITY = 5000 AUTO-TRIGGER 7 LPPLSI ==1 SAFETY INJECTION ACTUATED o PMP-CS004T XPP0043A CHRG/SI PMP A TRIP ON COMMAND o PMP-CS006F XPP0043B CHRG/SI PMP B FAIL TO START TRIGGER 8 LOA-CVC041 CHARGING PUMP A SUPPLY BRKR POSITION TO: RACK OUT Appendix D Scenario Outline Form ES-D-1 NRC 2015 - Scenario 1 NUREG -1021 R9 S1 On cue from the Examiner, a LOCA will be inserted in the RHR suction line outside the Reactor Building. This leak is in the discharge line from the B RHR Pump to the RCS. The crew will implement AOP-101.1, Loss of Reactor Coolant Not Requiring SI, and determine that an SI is required. The crew will implement EOP-1.0 (E-0) and determine that the RCS leak is outside of containment and transition to EOP-2.5 (ECA-1.2). The crew will isolate the leak by closing 8888B, RHR LP A to Cold Legs, and transition to EOP-2.0 (E-1). tuate automatically or manually. Individual components will be started/positioned to their required SI condition. When Safety Injection actuation is attempted the running harging pump will trip Charging pump will fail to auto-start resulting in the loss of all High Head Safety injection. It is a critical task to start one High Head Safety Injection Pump. If the crew attempts to rack-will not rack-up. from High Head Safety Injection. Too much flow would mask the leak as RCS pressure is lowered. CRITICAL TASKS: It is a critical task to:

• align at least one CHG/SI flowpath by Charging Pump, prior to ORANGE path on Core Cooling.

• isolate the LOCA prior to exiting EOP-2.5 (ECA-1.2).

• maintain SG levels using EFW without causing a Reactor trip. TERMINATION: The scenario can be terminated after the crew has isolated the leak in EOP-2.5 (ECA-1.2), and transitions to EOP -2.0 then EOP-1.2 (ES-1.1) and terminates Safety Injection or at any time at the discretion of the Examiner.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 - Scenario 1 NUREG -1021 R9 S1 Scenario Attributes Events Total Malfunctions (5-8) 9

• MD EFW

• Loss of Emergency Auxiliary transformer (1DB)

• DG fails to auto-start

• Rod F8 slips to approximately 200 steps withdrawn

• 2 Dropped Rods (F8 and D4)

• Control Card Output for Letdown PCV-145 Drifts LOW

• Actuation Failure (Auto and Manual) * * -Start Malfunctions after EOP entry (1-2) 4 * * * -Start

• CHG/SI pump fails to Rack-Up Abnormal Events (2-4) 5 *

• Loss of Emergency Auxiliary transformer (1DB) with DG failing to auto-start

• Rod F8 slips to approximately 200 steps withdrawn

• 2 Dropped Rods (F8 and D4)

• Control Card Output for Letdown PCV-145 Drifts LOW Major Transient (1-2) 1

• LOCA Outside the Reactor Building. EOPs Entered (1-2) 3

• EOP-2.5 (ECA-1.2), LOCA Outside Containment

• EOP-2.0 (E-1), Loss Of Reactor Or Secondary Coolant

• EOP-1.2 (ES-1.1), Safety Injection Termination EOP Contingencies (0-2) 1

• EOP-2.5 (ECA-1.2), LOCA Outside Containment Critical Tasks (2-3) 3

• Maintain SG levels using EFW without causing a Reactor trip.

• Align at least one CHG/SI flowpath prior to ORANGE path on Core Cooling.

• Isolate LOCA prior to exiting EOP-2.5 (ECA-1.2).

Appendix D Scenario Outline Form ES-D-1 NRC 2015 - Scenario 1 NUREG -1021 R9 S1 SIMULATOR SCENARIO SETUP INITIAL CONDITIONS: IC Set 290 10-3% Power EOL Burnup = 20,000 MWD/MTU RCS Boron Concentration = 652 ppm FCV-113 Pot Setting = 2.80 Rod Position: Group D = 94 Tavg = 557.9 Xe = - 0.0 pcm Prior to the scenario, crew should pre-brief on conditions and expectations for the Shift (maintain power, repairs estimated to be complete well before LCO action time expires.) PRE-EXERCISE: Ensure simulator has been checked for hardware problems (DORT, burnt out light bulbs, switch malfunctions, chart recorders, etc.). VCS-TQP-0807 Attachment I-A, Unit 1 Booth Instructor Checklist, has been completed. Hang Red Tags for equipment out of service: o Hang Caution Tag on HCV-186 due to ASI being OOS  : o GOP-3, Reactor Startup From Hot Standby To Startup (Mode 3 To Mode 2) o GOP-4A, Power Operation (Mode 1 - Ascending) Conduct two-minute drill. PRE-LOAD: STANDARD SIMULATOR SETUP: PMP-LD003P, XPP0138 Leak Detection Sump Pmp Loss of Power VLV-FW028W, XVG01676-FW FW Hdr Recirc Isol Vlv Loss of Power VLV-FW029W, XVG01679-FW FW HTR Recirc Iso Vlv Loss of Power VLV-CS052W, XVT08141A-CS RCP A Seal Leakoff Vlv Loss of Power VLV-CS054W, XVT08141C-CS RCP C Seal Leakoff Vlv Loss of Power VLV-CS053W, XVT08141B-CS RCP B Seal Leakoff Vlv Loss of Power SCENARIO RELATED: ANN-TA030 , GEN AUX PNL TRBL FAIL TO: OFF ANN-CS044, ALT SEAL INJ PUMP TRBL FAIL TO: ON MAL-CVC027, ALT SEAL INJ D/G FAIL TO START MAL-CVC029, ALT SEAL INJ PUMP FAIL TO START Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 1 Page: 10 of 49 Event

Description:

Start RBCU 2B, then secure RBCU 1B. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOOTH OPERATOR: No triggers for this event. CRS Direct BOP to start RBCU 2B, then secure RBCU 1B in accordance with SOP-114, Reactor Building Ventilation System,Section III.A. NOTE 2.1 a. Due to eddy current brakes, RBCU control switches must be held in START position until the red breaker closed light is lit and starting current is indicated on appropriate meter. b. Normal and preferred lineup is three RBCUs running in NORM (fast speed). c. To increase stay times for teams entering containment, four RBCUs may be placed in service in NORM (fast speed). SOP-114 BOP 2.1 Place RBCUs in service by starting three or four RBCUs in SLOW or NORM as follows: SOP-114 BOP b. For XFN0064B-AH, REACTOR BLDG COOLING UNIT 1B EMERG FAN, start one of the following: 1) XFN 0064B-AH, 1B NORM. SOP-114 NOTE 2.1.e Contact PSE to evaluate, if RBCU fan motor amps exceed the values given. SOP-114 BOP e. Verify RBCU Fan motor amps return to normal operating range: 1) For fast speed operation, 275 amps to 300 amps. SOP-114 NOTE 2.1.f The RBCU TRAIN A (B) EMERG switch must be selected to an operable RBCU. SOP-114 BOP f. Verify the following switches are in the desired position: 2) XFN-64B/XFN 65B - RBCU TRAIN B EMERG. SOP-114 BOP 2.2 Shut down RBCUs by placing appropriate switch(es) in STOP: c. XFN 0064B-AH, 1B NORM. SOP-114 BOP Report that the 2B RBCU is running and the 1B RBCU has been secured. SOP-114 EVALUATOR NOTE: The next event is a power change which does not require a trigger.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 2 Page: 11 of 49 Event

Description:

Raise power to between 1% and 3%. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOOTH OPERATOR: No triggers for this event. EVALUATOR NOTE: The Unit is stable in Mode 3 at turnover with all surveillances completed for a Mode change to Mode 2. GOP-3 is complete to step 3.14. The RO will bring the Reactor to the POAH and transition to GOP-4A Power Operation (Mode 1 Ascending). GOP-4A steps have been signed off to indicate that lineups have been completed and the secondary warmed. NOTE 3.14 Ensure sufficient Emergency Feedwater Flow exists prior to raising power. GOP-3 RO 3.14 Increase Reactor Power to between 1% and 3%. GOP-3 RO 3.15 At the Point of Adding Heat, if NR-45, NIS RECORDER, had previously been selected to HI speed place the recorder in LO speed. GOP-3 CAUTION 3.16 a. Adjustment of Tavg with the Rod Control System must not be attempted with the ROD CNTRL BANK SEL Switch in any position other than MAN. b. Manual rod control is required to establish equilibrium conditions, since C-5 blocks automatic rod withdrawal. GOP-3 RO 3.16 Maintain Tavg between 555°F and 559°F. GOP-3 BOP Adjust EFW flow to the Steam Generators (SG) as power is increased to maintain Narrow Range SG levels between 60% and 65%. EVALUATOR NOTE: Attachment II.G was completed prior to turnover N/A 3.17 Complete Attachment II.G, Operational Mode Change Plant Startup - Entering Mode 1, of GTP-702. GOP-3 CRS 3.18 Proceed to GOP-4A, Power Operation (Mode 1 - Ascending). GOP-3 EVALUATOR NOTE: GOP-4A POWER OPERATION (MODE 1 - ASCENDING) has several line-up verifications. GOP-4A lineups and secondary plant warming have been completed. NOTE 3.1 through 3.11 Steps 3.1 through 3.11 raise Reactor Power from 1% to 25%. GOP-4A EVALUATOR NOTE: The next event may be initiated after GOP-4A is entered.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 3 Page: 12 of 49 Event

Description:

Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOOTH OPERATOR: When directed - Initiate Event 3 (TRIGGER 1). EVALUATOR NOTE: Event 3 should be initiated on entry into EOP-4A, Power Operation (Mode 1 - Ascending). Power should be between 2.0 - 3% when this event is triggered. Indications Available: MD EFP B Amps > 60 amps XCP-623 1-5, MD EFP B MOTOR OVRLD CRS Enters ARP-001-XCP-623, 1-5 XCP-623 1-5 CORRECTIVE ACTIONS: XCP-623 1-5 BOP 1. If possible, reduce demand to less than 400 gpm by throttling the flow control valves to the Steam Generators. XCP-623 1-5 BOP 2. Start Motor Driven Emergency Feedwater Pump A if necessary to maintain Steam Generator levels. XCP-623 1-5 EVALUATOR NOTE: The guidance in SOP-211, Emergency Feedwater System, is not relevant for this failure because both pumps are running initially. BOP 3. Refer to SOP-211. XCP-623 1-5 BOP 4. Determine if a single phasing event is in progress by diagnosis of any combination of the following symptoms: a. Vibration Alarms are received for other equipment. b. MCB Potential Lights are not lit. c. MCB Amber Overload lights are lit for running equipment or Motor Overload Alarms are received. d. MCB Undervoltage Alarms. e. Affected bus local 7.2 KV Bus ammeters XCP-623 1-5 EVALUATOR NOTE: The CRS could direct the RO to Stop the B MD EFW Pump, reduce The TD EFW Pump is not normally used for SG level control during heatup/cooldown. Indications Available: MD EFP B Amps > 60 amps XCP-623 1-5, MD EFP B MOTOR OVRLD XCP-623 1-3, MD EFP B TRIP Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 3 Page: 13 of 49 Event

Description:

Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 EVALUATOR NOTE: The crew will enter ARP-001-XCP-622, 1-EFW Pump will trip within one minute only the actions associated with XCP-623 1-3, MD EFP B TRIP are included. CRS Enters ARP-001-XCP-622, 1-5 XCP-623 1-5 CRS Enters ARP-001-XCP-623, 1-3 XCP-623 1-3 EVALUATOR NOTE: The CRS could direct the RO to reduce power to ensure MD EFW Astart the TD EFW Pump. The TD EFW Pump is not normally used for SG level control during heatup/cooldown. CORRECTIVE ACTIONS: XCP-623 1-3 BOP 1. Start Motor Driven Emergency Feedwater Pump A if necessary to maintain Steam Generator levels. XCP-623 1-3 RO 2. Reduce feedwater demand to less than 400 gpm. XCP-623 1-3 CRS 3. Refer to SOP-211. XCP-623 1-3 SUPPLEMENTAL ACTIONS: XCP-623 1-3 BOP 1. If Steam Generator levels cannot be maintained with one motor driven pump, start the Turbine Driven Emergency Feedwater Pump. 2. Place PUMP B control switch in NORMAL-AFTER-STOP to clear the alarm. XCP-623 1-3 CRS 3. Determine the cause of the trip and correct as soon as possible. 4. If the pump is inoperable, refer to Technical Specification 3.7.1.2. XCP-623 1-3 EVALUATOR NOTE: Emergency Feedwater requirements are approximately 180 gpm per percent power.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 3 Page: 14 of 49 Event

Description:

Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 CRITICAL TASK RO/BOP Maintain SG level without tripping the unit by reducing feedwater demand (Reactor Power) and/or controlling Emergency Feedwater flow. BOOTH OPERATOR: wait 3 minutes and report that the pump bearing are hot and the breaker is tripped with no flags. CRS Contacts Work Control and/or Maintenance for assistance. EVALUATOR NOTE: Technical Specification 3.0.4 is applicable so entry into Mode 1 is prohibited. CRS Enters Technical Specification 3.7.1.2, Action a: With one emergency feedwater pump inoperable, restore the required emergency feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. TECH SPEC EVALUATOR NOTE: The next event may be initiated after SG levels are under control and the Technical Specification determination is complete.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 4 Page: 15 of 49 Event

Description:

Loss of Emergency Auxiliary transformer (1DB). DG fails to AUTO start. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOOTH OPERATOR: When directed - Initiate Event 4 (TRIGGER 2). BOOTH OPERATOR: Silence the HVAC Alarms. INDICATIONS AVAILABLE: Multiple Alarms XCP-637 5-2 7KV ESF CHAN B BKR TRIP 1DB Volts = 0 1DB Feed Amps = 0 AUTOMATIC ACTIONS: XCP-637 5-2 1. If XSW1DB 07, TRANS 1DB1 & 1DB2, tripped, XSW1DB1 4B, MAIN INCOMING BREAKER, and XSW1DB2 4B, MAIN INCOMING BREAKER, will trip. 2. If XSW1DB 16, BUS 1DB NORMAL INCOMING BKR, tripped, Diesel Generator B will automatically start. 3. If XSW1EB 03, TRANSF 1EB1 FEEDER BREAKER, tripped, SW1EB1 4B, MAIN INCOMING BREAKER, will trip. XCP-637 5-2 EVALUATOR NOTE: B DB failed to start and load onto the 1DB bus. After determining that there are no lockout on the B DG or the 1DB Bus the operator should perform an emergency start of the B DG. BOP Perform an Emergency Start of the 1B Diesel Generator. CORRECTIVE ACTIONS: XCP-637 5-2 BOP 1. Using MCB indication, determine which breaker tripped. 2. Verify appropriate automatic actions. 3. Dispatch an operator to investigate the cause of the breaker trip. 4. Request Electrical Maintenance to troubleshoot and correct the cause of the breaker trip. XCP-637 5-2 SUPPLEMENTAL ACTIONS: XCP-637 5-2 1. When the cause has been corrected, reclose the breaker. XCP-637 5-2 CRS 2. Refer to Technical Specifications 3.8.1 and 3.8.3 for LCO requirements. XCP-637 5-2 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 4 Page: 16 of 49 Event

Description:

Loss of Emergency Auxiliary transformer (1DB). DG fails to AUTO start. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 CRS T.S. 3.8.1.1 AC SOURCES Actions c. With one offsite circuit and one EDG inoperable: 1. Demonstrate the OPERABILITY of the remaining offsite AC source by performing Surveillance Requirement 4.8.1.1.1 within one hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, and 2. *If the EDG became inoperable due to any cause other than preplanned preventative maintenance or testing: a) determine the OPERABLE EDG is not inoperable due to a common cause failure within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or b) demonstrate the OPERABILITY of the remaining EDG by performing Surveillance Requirement 4.8.1.1.2.a.3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and 3. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, verify that required systems, subsystems, trains, components and devices that depend on the remaining EDG as a source of emergency power are also OPERABLE and in MODE 1, 2, or 3, that the Turbine Driven Emergency Feed Pump is OPERABLE. If these conditions are not satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. 4. Restore one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, and 5. Restore the other AC power source (offsite circuit or diesel generator) to OPERABLE status in accordance with the provisions of Section 3.8.1.1 Action Statement a. or b.,as appropriate, with the time requirement of that Action Statement based on the time of initial loss of the remaining inoperable A.C. power source. T.S. 3.8 1.1 EVALUATOR NOTE: The Operators may have re-energized the 1DB bus using the 1B Diesel Generator.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 4 Page: 17 of 49 Event

Description:

Loss of Emergency Auxiliary transformer (1DB). DG fails to AUTO start. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 CRS T.S. 3.8.3.1 Actions a. With one of the required trains of AC Emergency busses not fully energized, re-energize the division within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. T.S. 3.8 3.1 EVALUATOR NOTE: The next event may be initiated after the 1B Diesel Generator is powering the 1DB 7.2 KV Bus. CRS Implement SOP-306, Emergency Diesel Generator, B. Operation Of Diesel Generator B After An Automatic Start And Load. BOP 2.1 Verify B TRN BLACKOUT SEQ COMPLETE Status Light is lit. SOP-306 RO 2.2 Ensure one Charging Pump is running. SOP-306 BOP 2.3 Ensure the following loads have started: a. RHR Pump B. b. One Train B Service Water Pump. c. One Train B HVAC Chilled Water Pump. d. One Train B CCW Pump. e. MD EFW Pump B. f. The Train B RBCU selected for emergency operation (slow speed). g. Train B FHB Exhaust Fan. h. Service Water Booster Pump B. i. The Train B HVAC Chiller associated with the running Train B HVAC Chilled Water Pump. SOP-306 BOP 2.4 Verify greater than or equal to 2000 gpm flow on FI-4496, SWBP B DISCH FLOW. SOP-306 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 4 Page: 18 of 49 Event

Description:

Loss of Emergency Auxiliary transformer (1DB). DG fails to AUTO start. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOP 2.5 Perform the following per SOP-220: a. Ensure Instrument Air is supplied by one of the following: 1) Either Station Instrument Air Compressor A or B. 2) Diesel Driven Air Compressor. SOP-306 BOP 2.6 Supply Reactor Building Instrument Air from Station Instrument Air with Reactor Building Instrument Air Compressors secured per SOP-121,Section IV. SOP-306 BOP 2.7 Maintain RB temperature as follows: a. Monitor RB temperature and pressure for indications of insufficient cooling. b. If required, supply Service Water to the Train A RBCUs per SOP-117. SOP-306 BOP 2.8 With Shift Supervisor concurrence perform the following: a. Secure Emergency Feedwater Pumps. b. Realign the Emergency Feedwater System for standby operation per SOP-211. SOP-306 NOTE 2.9 Spent Fuel Cooling Loop B is unavailable until NON-ESF LCKOUTS is reset. SOP-306 BOP 2.9 If required, startup Spent Fuel Cooling Loop A aligned to the Spent Fuel Pool per SOP-123. SOP-306 CAUTION 2.10 De-energizing the following Atmospheric Gaseous Module rate meters when the appropriate Interlock Switch is in NORMAL/OFF will result in the generation of a High Radiation signal and component realignment: a. RMA0001-RM, ATM GASEOUS IODINE - CONT ROOM SUPP AIR. b. RMA0002-RM, ATM GASEOUS IODINE - RB SAMPLE LINE. c. RMA0010-RM, WASTE GAS DISCHARGE RADIATION MONITOR. SOP-306 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 4 Page: 19 of 49 Event

Description:

Loss of Emergency Auxiliary transformer (1DB). DG fails to AUTO start. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOP 2.10 Perform either of the following for Train B radiation monitors: a. Restore Train B radiation monitors to normal operation per SOP-124. b. If Train B radiation monitors are unable to be restored to normal operation, contact Health Physics to perform compensatory actions per HPP-904 for loss of electrical power to Train B radiation monitors. SOP-306 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 5 Page: 20 of 49 Event

Description:

Partially Dropped Rod (F8) Rod slips to approximately 200 steps withdrawn. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOOTH OPERATOR: When directed - Initiate Event 5 (TRIGGER 3). Indication Available: XCP-620 2-5 CMPTR ROD DEV CORRECTIVE ACTIONS: XCP-621 3-1 RO 1 If a shutdown or control group rod fails to withdraw, implement AOP-403.5, Stuck or Misaligned Control Rod. 2 If a shutdown or control group rod has dropped and the Reactor did not trip, implement AOP-403.6, Dropped Control Rod. XCP-621 3-1 CRS Implement AOP-403.6, Dropped Control Rod. IOA RO 1 Verify only one Control Rod has dropped. AOP-403.6 IOA RO 2 Place ROD CNTRL BANK SEL Switch in MAN. AOP-403.6 RO 3 Stabilize the plant: a. Decrease Main Turbine load to maintain Tavg within 5°F of Tref . b. Verify PZR pressure is stable at OR trending to 2235 psig (2220 psig to 2250 psig). c. Verify PZR level is stable at OR trending to program level. AOP-403.6 RO 4 Check if Reactor power is LESS THAN 75%. AOP-403.6 CRS 5 Initiate GTP702, Attachments IV.A, IV.B, and IV.C. AOP-403.6 CRS 6 Notify the following plant personnel prior to moving Control Rods: Management Duty Supervisor. Rod Control System Engineer. Reactor Engineering AOP-403.6 BOOTH OPERATOR: As Reactor Engineering after receiving the report of plant conditions, recommend that Reactor Power be maintained at the current level until a recovery plan is developed.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 5 Page: 21 of 49 Event

Description:

Partially Dropped Rod (F8) Rod slips to approximately 200 steps withdrawn. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 CRS 7 Provide Reactor Engineering with the following information: Time rod dropped: ________. Dropped rod location: ________. Initial Reactor power level: ________. Current Reactor power level: ________. Current QPTR: ________. AOP-403.6 CRS 8 Determine and correct the cause of the failure. AOP-403.6 NOTE - Step 9 This Step must be completed before continuing with Step 10. AOP-403.6 CRS 9 Obtain the following information from Reactor Engineering: Power level at which recovery is to be performed: ________. Rate of Control Rod movement during recovery: ________. AOP-403.6 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 5 Page: 22 of 49 Event

Description:

Partially Dropped Rod (F8) Rod slips to approximately 200 steps withdrawn. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 CRS Enter Technical Specification 3.1.3.1.d.3 d. With one full length rod inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either: 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that: a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 - is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. c) A core power distribution measurement is obtained and F0(z) and Fj~5 are verified to be within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high neutron flux tip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. TECH SPEC EVALUATOR NOTE: This scenario does NOT include Control Rod recovery. The next event may be inserted after Technical Specifications have been addressed. The Technical Specifications may be addressed in post-exam questioning if it is desired to expedite the scenario.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 6 Page: 23 of 49 Event

Description:

Two Dropped Rods (F8 and D4) Trip the reactor Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOOTH OPERATOR: When directed - Initiate Event 6 (TRIGGER 4). EVALUATOR NOTE: The following steps occur after the 2nd rod drops. Indications Available: XCP-621, 3-2 RODS ON BOTTOM RO CORRECTIVE ACTIONS: 1 If two or more rods have dropped, manually trip the Reactor and implement EOP-1.0, Reactor Trip/Safety Injection Actuation. SUPPLEMENTAL ACTIONS: 1 Have I&C verify proper operation of the DRPI System and repair if necessary. XCP-621 3-2 CRS Direct EOP-1.0 (E-0) Reactor Trip/Safety Injection Actuation, entry. IOA Crew 1 Verify Reactor Trip: Trip the Reactor using either Reactor Trip Switch. Verify all Reactor Trip and Bypass Breakers are open. Verify all Rod Bottom Lights are lit. Verify Reactor Power level is decreasing. EOP-1.0 IOA BOP 2 Verify Turbine/Generator Trip: a. Verify all Turbine STM STOP VLVs are closed. b. Ensure Generator Trip (after 30 second delay): 1) Ensure the GEN BKR is open. 2) Ensure the GEN FIELD BKR is open. 3) Ensure the EXC FIELD CNTRL is tripped EOP-1.0 IOA BOP 3 Verify both ESF buses are energized. EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 6 Page: 24 of 49 Event

Description:

Two Dropped Rods (F8 and D4) Trip the reactor Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 IOA BOP 4 Check if SI is actuated: a. Check if either: (NO) SI ACT status light is bright on XCP-6107 1-1. OR Any red first-out SI annunciator is lit on XCP-626 top row. ALTERNATIVE ACTION a. GO TO Step 5. EOP-1.0 IOA CREW 5 Check if SI is required: a. Check if any of the following conditions exist: (NO) PZR pressure LESS THAN 1850 psig. OR RB pressure GREATER THAN 3.6 psig. OR Steamline pressure LESS THAN 675 psig. OR Steamline differential pressure GREATER THAN 97 psid. ALTERNATIVE ACTION a. GO TO EOP-1.1, REACTOR TRIP RECOVERY, Step 1. EOP-1.0 CRS Direct EOP-1.1 entry.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 6 Page: 25 of 49 Event

Description:

Two Dropped Rods (F8 and D4) Trip the reactor Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 REFERENCE PAGE FOR EOP-1.1 1 SI ACTUATION CRITERIA IF either of the following conditions occurs, THEN actuate SI and GO TO EOP-1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION, Step 1: PZR level can NOT be maintained GREATER THAN 8%. OR RCS subcooling on TI-499A(B), A(B) TEMP °F, is LESS THAN the value listed in the table below: RCS PRESSURE (psig) RCS SUBCOOLING (°F) 1576-3075 42.5 876-1575 45 576-875 47.5 476-575 50 375-475 52.5 EOP-1.1 CAUTION If SI actuation occurs during this procedure, EOP-1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION, should be performed to stabilize the plant. EOP-1.1 NOTE Main Turbine vibration should be monitored during coastdown. The EOP REFERENCE PAGE should be monitored throughout the use of this procedure. EOP-1.1 CREW 1 Announce plant conditions over the page system. EOP-1.1 EVALUATOR NOTE: Initiate Event 7 (TRIGGER 5) after EOP-1.1 has been entered. Remaining steps of EOP-1.1 will be run concurrently with Event 7 (next section) for PCV-145 failure.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 6 Page: 26 of 49 Event

Description:

Two Dropped Rods (F8 and D4) Trip the reactor Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOP 2 Check FW status: a. Check if RCS Tavg is LESS THAN 564°F. b. Verify FW Isolation: Ensure the FW Flow Control Valves, FCV-478(488)(498), are closed. Ensure the Main FW Isolation Valves, PVG-1611A(B)(C), are closed. Ensure the FW Flow Control Bypass Valves, FCV-3321 (3331)(3341), are closed. EOP-1.1 EVALUATOR NOTE Pump tripped during an earlier event. BOP c. Ensure EFW Pumps are running: 1) Ensure both MD EFW Pumps are running. (NO) 2) Verify the TD EFW Pump is running if necessary to maintain SG levels. d. Verify total EFW flow is GREATER THAN 450 gpm. e. Trip all Main FW Pumps. EOP-1.1

• RO 3 Check RCS temperature: With any RCP running, RCS Tavg is stable at OR trending to 557°F. OR With no RCP running, RCS Tcold is stable at OR trending to 557°F. EOP-1.1 BOP 4 IF EOP-1.0 was entered from AOP-112.2, THEN RETURN TO AOP-112.2, STEAM GENERATOR TUBE LEAK NOT REQUIRING SI, Step 7. (NO) ALTERNATIVE ACTION 4 GO TO Step 5. EOP-1.1 BOP 5 Verify all Control Rods are fully inserted. EOP-1.1 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 6 Page: 27 of 49 Event

Description:

Two Dropped Rods (F8 and D4) Trip the reactor Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOP 6 Check DA level control: a. Open LCV-3235, DEAER START UP DRAIN CNTRL, as necessary to maintain DA level LESS THAN 10.5 ft as indicated on LI-3135, DEAER STOR TK WR LVL FEET. b. Locally adjust ITV-3062A(B)(C), BD COOLER A(B)(C) CDSTE OUT TEMP, to 90% (XPN-0029, NUCLEAR BLOWDOWN PROCESSING PANEL, AB-436). EOP-1.1 RO 7 Check PZR level control: a. Verify PZR level is GREATER THAN 17%. b. Verify Charging and Letdown are in service. c. Verify PZR level is trending to 25%. EOP-1.1 RO 8 Check PZR pressure control: a. Verify PZR pressure is GREATER THAN 1850 psig. b. Verify PZR pressure is stable at OR trending to 2230 psig (2220 psig to 2250 psig). EOP-1.1

• BOP 9 Check SG levels: a. Verify Narrow Range level in all SGs is GREATER THAN 26%. b. Control EFW flow to maintain Narrow Range SG level between 40% and 60%. EOP-1.1

• BOP 10 Verify all AC buses are energized by offsite power: ESF AC buses. BOP AC buses. EOP-1.1 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 6 Page: 28 of 49 Event

Description:

Two Dropped Rods (F8 and D4) Trip the reactor Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOP 11 Transfer Condenser Steam Dumps to the Steam Pressure Mode: a. Verify PERMISV C-9 status light is bright on XCP-6114 1-3. b. WHEN RCS Tavg is LESS THAN P-12 (552"F), THEN place both STM DUMP INTERLOCK Switches to BYP INTLK. c. Perform the following: Verify the MS Isolation Valves, PVM-2801A(B)(C), are open. OR Open MS Isolation Bypass Valves: 1) Depress both MAIN STEAM ISOL VALVES RESET TRAIN A(B). 2) Open MS Isolation Bypass Valves, PVM-2869A(B)(C). d. Place the STM DUMP CNTRL Controller in MAN and closed. e. Ensure the STM DUMP CNTRL Controller is set to 8.4. f. Place the STM DUMP MODE SELECT Switch in STM PRESS. g. Place the STM DUMP CNTRL Controller in AUTO. EOP-1.1 NOTE - Step 12 Priority should be given to running RCP A to supply Normal PZR Spray. Since a time lag is expected after increasing steam flow before natural circulation parameters can be verified, this procedure should be continued concurrently with the establishment of natural circulation. EOP-1.1 RO 12 Verify RCP A is running. EOP-1.1 RO 13 Verify PERMISV C-9 status light is bright on XCP-6114 1-3. EOP-1.1 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 6 Page: 29 of 49 Event

Description:

Two Dropped Rods (F8 and D4) Trip the reactor Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 RO 14 Check the position of NR-45, NIS RECORDER: a. Verify Intermediate Range Power is LESS THAN P-6 (7.5x10-6%). b. Transfer NR-45, NIS RECORDER, to both Source Range channels. c. Initiate GTP-702, Attachment VI.KK. EOP-1.1 BOP 15 Shut down and stabilize the Secondary Plant. REFER TO AOP-214.1, TURBINE TRIP. EOP-1.1 RO 16 Maintain stable plant conditions: a. Maintain PZR pressure at 2230 psig (2220 psig to 2250 psig). b. Maintain PZR level at 25%. c. Maintain Narrow Range SG levels between 40% and 60%. d. Maintain RCS temperature: With any RCP running, Tavg at 557°F. OR With no RCP running, Tcold at 557°F. e. REFER TO GOP-5, REACTOR SHUTDOWN FROM STARTUP TO HOT STANDBY (MODE 2 TO MODE 3). EOP-1.1 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 7 Page: 30 of 49 Event

Description:

Letdown pressure control valve PCV-145 fails CLOSED (AUTO ONLY) Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOOTH OPERATOR: When directed - Initiate Event 7 (TRIGGER 5). Indications Available: XCP-613 2-4 LP LTDN FLO/PRESS HI CRS Direct implementation of ARP-001-XCP-613 CORRECTIVE ACTIONS: XCP-613 2-4 RO 1. Verify proper operation of PCV-145, LO PRESS LTDN. XCP-613 2-4 EVALUATOR NOTE: The Operator should identify the malfunction of PCV-145 automatic control and adjust letdown pressure with Manual control of PCV-145. RO 2. If necessary, place PCV-145, LO PRESS LTDN, in MAN and adjust as necessary to reduce flow or pressure. XCP-613 2-4 RO 3. Close Letdown orifice isolation valves as necessary to reduce flow or pressure. XCP-613 2-4 RO 4. Isolate Charging flow if Letdown is isolated. XCP-613 2-4 CRS Contacts Work Control/Maintenance for assistance. EVALUATOR NOTE: The next event may be initiated after letdown pressure control is established in manual.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 31 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOOTH OPERATOR: When directed - Initiate Event 8 (TRIGGER 6). EVALUATOR NOTE: The CRS may, at his/her discretion, implement AOP-101.1 LOSS OF REACTOR COOLANT NOT REQUIRING SI to attempt leak isolation in parallel with EOP-1.1. The leak rate will quickly exceed the capability of available makeup to maintain Pressurizer level. EOP-1.1, REACTOR TRIP RECOVERY reference page will also direct actuation of Safety Injection if AOP-101.1 is not entered. Indications Available: XCP-614 5-1 CHG LINE FLO HI/LO XCP-616 1-5 PZR LCS DEV HI/LO XCP-616 2-3 PZR PRESS HI/LO XCP-616 3-6 PZR PCS LO BU HTRS ON XCP-607 3-4 LD TRBL AB SMP/FLDRN LVL HI XCP-644 3-2 PLANT VENT PARTIC RM-A3 TRBL XCP-645 2-2 AB VENT DUCTS RM-A11 TRBL Decreasing Pressurizer level with increased Charging flow and normal Letdown flow. Decreasing Pressurizer pressure. Increased Heater output and Backup Heaters on. Increased VCT makeup frequency. CRS Diagnose an RCS Leak. CRS Implement AOP-101.1, Loss of Reactor Coolant Not Requiring SI. AOP-101.1 NOTE: If a Reactor Trip occurs AND SI is NOT required, this procedure should be continued after the actions of EOP-1.1, REACTOR TRIP RECOVERY, are completed. As valves are isolated, it may be necessary to monitor RCS pressure for a period of time to determine if the leak is isolated. AOP-101.1 EVALUATOR NOTE: PCV-145 has previously been placed in manual due to a failure of the controller in auto. Letdown should be controlled in manual rather the placing the failed controller back in service.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 32 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1

• RO 1. Verify PZR level is at or trending to program level. (NO) ALTERNATIVE ACTION 1. IF PZR level is decreasing, THEN perform the following: a) Open FCV-122, CHG FLOW, as necessary to maintain PZR level. b) IF PZR level continues to decrease, THEN reduce Letdown to one 45 gpm orifice: 1) Set PCV-145, LO PRESS LTDN, to 70%. 2) Ensure PVT-8149A, LTDN ORIFICE A ISOL, is open. 3) Close both PVT-8149B(C), LTDN ORIFICE B(C) ISOL. 4) Adjust PCV-145, LO PRESS LTDN, to maintain PI-145, LO PRESS LTDN PRESS PSIG, between 300 psig and 400 psig. 5) Place PCV-145, LO PRESS LTDN, in AUTO. AOP-101.1

• CRS, RO 2 Check if SI is required: a. Check if any of the following criteria are met: PZR level is decreasing with Charging maximized and Letdown minimized. (YES) OR PZR level is approaching 8%. OR PZR pressure is approaching 1870 psig. OR VCT level is approaching 5%. b. Perform the following: 1) Trip the Reactor. 2) GO TO EOP-1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION. WHEN EOP-1.0 Immediate Actions are complete, THEN actuate SI. AOP-101.1 CRS Implement EOP-1.0, Reactor Trip/Safety Injection Actuation.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 33 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 REFERENCE PAGE FOR EOP-1.0 1 RCP TRIP CRITERIA a. IF Phase B Containment Isolation has actuated (XCP-612 4-2), THEN trip all RCPs. b. IF both of the following conditions occur, THEN trip all RCPs: SI flow is indicated on FI-943, CHG LOOP B CLD/HOT LG FLOW GPM. AND RCS Wide Range pressure is LESS THAN 1418 psig. 2 REDUCING CONTROL ROOM EMERGENCY VENTILATION Reduce Control Room Emergency Ventilation to one train in operation within 30 minutes of actuation. REFER TO SOP-505, CONTROL BUILDING VENTILATION SYSTEM. 3 MONITOR SPENT FUEL COOLING Periodically check status of Spent Fuel Cooling by monitoring the following throughout event recovery: Spent Fuel Pool level. Spent Fuel Pool temperature. EOP-1.0 See step 11 note for RCP Trip Criteria IOA RO 1 Verify Reactor Trip: Trip the Reactor using either Reactor Trip Switch. Verify all Reactor Trip and Bypass Breakers are open. Verify all Rod Bottom Lights are lit. Verify Reactor Power level is decreasing. EOP-1.0 IOA BOP 2 Verify Turbine/Generator Trip: a. Verify all Turbine STM STOP VLVs are closed. b. Ensure Generator Trip (after 30 second delay): 1) Ensure the GEN BKR is open. 2) Ensure the GEN FIELD BKR is open. 3) Ensure the EXC FIELD CNTRL is tripped. EOP-1.0 IOA BOP 3 Verify both ESF buses are energized. EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 34 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 IOA RO 4 Check if SI is actuated: a. Check if either: SI ACT status light is bright on XCP-6107 1-1. OR Any red first-out SI annunciator is lit on XCP-626 top row. b. Actuate SI using either SI ACTUATION Switch. c. GO TO Step 6. EOP-1.0 EVALUATOR NOTE: The CRS may direct that SI be actuated during or after this step based on the instruction in AOP-101.1 to actuate SI following the EOP-1.0, Immediate Operator Actions. IOA RO 5 Check if SI is required: a. Check if any of the following conditions exist: PZR pressure LESS THAN 1850 psig. OR RB pressure GREATER THAN 3.6 psig. OR Steamline pressure LESS THAN 675 psig. OR Steamline differential pressure GREATER THAN 97 psid. EOP-1.0 RO b. Actuate SI using either SI ACTUATION Switch. EOP-1.0 EVALUATOR NOTE: Components to their SI condition in accordance with Attachment 3. Charging/SI pump will not Auto-Start on SI. Attachment 3, SI Equipment Verification, is included as an attachment to this guide on page 42.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 35 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOOTH OPERATOR: If contacted tripped. Use LOA-CVC041 to rack- -up. Acknowledge requests for support. BOP 6 Initiate ATTACHMENT 3, SI EQUIPMENT VERIFICATION. EOP-1.0 Crew 7 Announce plant conditions over the page system. EOP-1.0

• RO 8 Verify RB pressure has remained LESS THAN 12 psig on PR-951, RB PSIG (P-951), red pen. EOP-1.0

• RO 9 Check RCS temperature: With any RCP running, RCS Tavg is stable at OR trending to 557°F. OR With no RCP running, RCS Tcold is stable at OR trending to 557°F. EOP-1.0 RO 10 Check PZR PORVs and Spray Valves: a. PZR PORVs are closed. b. PZR Spray Valves are closed. c. Verify power is available to at least one PZR PORV Block Valve: MVG-8000A, RELIEF 445 A ISOL. MVG-8000B, RELIEF 444 B ISOL. MVG-8000C, RELIEF 445 B ISOL. d. Verify at least one PZR PORV Block Valve is open. EOP-1.0 NOTE - Step 11 Seal Injection flow should be maintained to all RCPs. EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 36 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 RO 11 Check if RCPs should be stopped: a. Check if either of the following criteria is met: Annunciator XCP-612 4-2 is lit (PHASE B ISOL). OR RCS pressure is LESS THAN 1418 psig AND SI flow is indicated on FI-943, CHG LOOP B CLD/HOT LG FLOW GPM. EOP-1.0 EVALUATOR NOTE: In accordance with OAP-Criteria for RCS pressure less than 1418 psig with SI flow do not apply in an event initiated from Mode 2. RO b. Stop all RCPs. EOP-1.0 RO 12 Verify no SG is FAULTED: No SG pressure is decreasing in an uncontrolled manner. No SG is completely depressurized. EOP-1.0 RO 13 Verify Secondary radiation levels indicate SG tubes are NOT RUPTURED: RM-G19A(B)(C), STMLN HI RNG GAMMA. RM-A9, CNDSR EXHAUST GAS ATMOS MONITOR. RM-L3, STEAM GENERATOR BLOWDOWN LIQUID MONITOR. RM-L10, SG BLOWDOWN CW DISCHARGE LIQUID MONITOR. EOP-1.0 RO 14 Check if the RCS is INTACT: a. RB radiation levels are normal on: RM-G7, CNTMT HI RNG GAMMA. RM-G18, CNTMT HI RNG GAMMA. b. RB Sump levels are normal. c. RB pressure is LESS THAN 1.5 psig. d. The following annunciators are NOT lit: XCP-606 2-2 (RBCU 1A/2A DRN FLO HI). XCP-607 2-2 (RBCU 1B/2B DRN FLO HI). EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 37 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 RO 15 Reset both SI RESET TRAIN A(B) Switches. EOP-1.0 RO 16 Reset Containment Isolation: RESET PHASE A - TRAIN A(B) CNTMT ISOL. RESET PHASE B - TRAIN A(B) CNTMT ISOL. EOP-1.0 RO 17 Place both ESF LOADING SEQ A(B) RESETS to: a. NON-ESF LCKOUTS. b. AUTO-START BLOCKS. EOP-1.0 RO 18 Establish Instrument Air to the RB: a. Start one Instrument Air Compressor and place the other in Standby. b. Open PVA-2659, INST AIR TO RB AIR SERV. c. Open PVT-2660, AIR SPLY TO RB. EOP-1.0 RO 19 Check if SI flow should be reduced: a. RCS subcooling on TI-499A(B), A(B) TEMP °F, is GREATER THAN 52.5°F. b. Secondary Heat Sink is adequate: Total EFW flow to the SGs is GREATER THAN 450 gpm. OR Narrow Range level is GREATER THAN 26% in at least one SG. c. RCS pressure is stable OR increasing. (NO) ALTERNATIVE ACTION c. GO TO Step 20. EOP-1.0 NOTE - Step 20 Procedures referenced in EOP-12.0, MONITORING OF CRITICAL SAFETY FUNCTIONS, may now be implemented. EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 38 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 CRS 20 Initiate monitoring of the Critical Safety Function Status Trees. REFER TO EOP-12.0, MONITORING OF CRITICAL SAFETY FUNCTIONS. EOP-1.0

• RO 21 Check SG levels: a. Verify Narrow Range level in all SGs is GREATER THAN 26%. b. Control EFW flow to maintain Narrow Range SG levels between 40% and 60%. EOP-1.0 BOOTH OPERATOR: Acknowledge request to sample SGs. RO 22 Check if Secondary activity is normal: a. Place SVX-9398A(B)(C), SG A(B)(C) SMPL ISOL, in AUTO. b. Notify Chemistry to sample all SG secondary sides for abnormal activity. EOP-1.0 CRS 23 Check for loss of Reactor Coolant outside Containment: a. Verify AB radiation levels are normal on: (NO) RM-A3, MAIN PLANT VENT EXH ATMOS MONITOR: PARTICULATE, IODINE, GAS. RM-A13, PLANT VENT HI RANGE. RM-A11, AB VENT GAS ATMOS MONITOR. Local area monitors. b. Verify annunciator XCP-631 6-1 is NOT lit (AB SMP LVL HI). c. Verify annunciators XCP-606 3-4 and XCP-607 3-4 are NOT lit (LD TRBL AB SMP/FLDRN LVL HI). ALTERNATIVE ACTION 23 Evaluate the cause of abnormal AB conditions. IF the cause is a loss of RCS inventory outside Containment, THEN GO TO EOP-2.5, LOCA OUTSIDE CONTAINMENT, Step 1. EOP-1.0 CRS Transition to EOP-2.5, LOCA OUTSIDE CONTAINMENT, Step 1.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 39 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 NOTE As valves are isolated, it may be necessary to monitor RCS pressure for a period of time to determine if the leak is isolated. Conditions for implementing Emergency Plan Procedures should be evaluated using EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN. Crew 1 Announce plant conditions over the page system. EOP-2.5 RO 2 Ensure the following are closed: a. RHR Pump Suction Valves from the RCS: 1) MVG-8701A and MVG-8702A, RCS LP A TO PUMP A (Status Lights XCP-6106 1-11(2-11)), for Train A. 2) MVG-8701B and MVG-8702B, RCS LP C TO PUMP B (Status Lights XCP-6106 1-12(2-12)), for Train B. b. Other paths out of Containment: 1) Normal Letdown Isolation:PVT-8149A(B)(C), LTDN ORIFICE A(B)(C) ISOL. PVT-8152, LTDN LINE ISOL. 2) RCP Seal Return Isolation: MVT-8100, SEAL WTR RTN ISOL. MVT-8112, SEAL WTR RTN ISOL. 3) PZR Sample Isolation: SVX-9356A, PZR STM SMPL ISOL. SVX-9356B, PZR LIQ SMPL ISOL. 4) RCS Loop B Sample Isolation: SVX-9364B, RCS LP B SMPL ISOL. SVX-9365B, RCS LP B SMPL ISOL. 5) RCS Loop C Sample Isolation: SVX-9364C, RCS LP C SMPL ISOL. SVX-9365C, RCS LP C SMPL ISOL. EOP-2.5 RO 3 Check if RCS pressure is continuing to decrease. EOP-2.5 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 40 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 RO 4 Try to identify and isolate the break: a. Close MVG-8888A, RHR LP A TO COLD LEGS. b. Check if RCS pressure is continuing to decrease. c. Open MVG-8888A, RHR LP A TO COLD LEGS. EOP-2.5 EVALUATOR NOTE: The following step isolates the leak from the RCS. CRITICAL TASK RO d. Close MVG-8888B, RHR LP B TO COLD LEGS. e. Check if RCS pressure is continuing to decrease. (NO) ALTERNATIVE ACTION e. GO TO EOP-2.0, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1. EOP-2.5 CRS Direct the implementation of EOP-2.0, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 41 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 REFERENCE PAGE FOR EOP-2.0 1 SI REINITIATION CRITERIA IF either of the following conditions occurs, THEN start Charging Pumps and operate valves as necessary: RCS subcooling on TI-499A(B), A(B) TEMP °F, is LESS THAN 52.5°F [67.5°F]. PZR level can NOT be maintained GREATER THAN 10% [28%]. 2 RCP TRIP CRITERIA IF either of the following criteria is met, THEN trip all RCPs: Annunciator XCP-612 4-2 is lit (PHASE B ISOL). RCS pressure is LESS THAN 1418 psig AND SI flow is indicated on FI-943, CHG LOOP B CLD/HOT LG FLOW GPM. 3 SECONDARY INTEGRITY TRANSITION CRITERIA IF any unisolated SG pressure is decreasing in an uncontrolled manner OR is completely depressurized, THEN GO TO EOP-3.0, FAULTED STEAM GENERATOR ISOLATION, Step 1. 4 TUBE RUPTURE TRANSITION CRITERIA IF any SG level increases in an uncontrolled manner OR if any SG has abnormal radiation, THEN start Charging Pumps and operate valves as necessary, and GO TO EOP-4.0, STEAM GENERATOR TUBE RUPTURE, Step 1. 5 COLD LEG RECIRCULATION TRANSITION CRITERION IF RWST level decreases to LESS THAN 18%, THEN GO TO EOP-2.2, TRANSFER TO COLD LEG RECIRCULATION, Step 1. 6 LOSS OF EMERGENCY COOLANT RECIRCULATION TRANSITION CRITERION IF Emergency Coolant Recirculation is established and subsequently lost, THEN GO TO EOP-2.4, LOSS OF EMERGENCY COOLANT RECIRCULATION, Step 1. 7 REDUCING CONTROL ROOM EMERGENCY VENTILATION Reduce Control Room Emergency Ventilation to one train in operation within 30 minutes of actuation. REFER TO SOP-505, CONTROL BUILDING VENTILATION SYSTEM. EOP-2.0 See note on next page for RCP Trip Criteria NOTE The EOP REFERENCE PAGE should be monitored throughout the use of this procedure. Seal Injection flow should be maintained to all RCPs. Conditions for implementing Emergency Plan Procedures should be evaluated using EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN. EOP-2.0 EVALUATOR NOTE: If flow has been High Head Safety Injection flow has been established with RCS pressure < 1418 psig and the RCPs are running they should be stopped in the following step.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 42 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 EVALUATOR NOTE: In accordance with OAP-Criteria for RCS pressure less than 1418 psig with SI flow do not apply in an event initiated from Mode 2. RO 1 Check if RCPs should be stopped: a. Check if either of the following criteria is met: Annunciator XCP-612 4-2 is lit (PHASE B ISOL). OR RCS pressure is LESS THAN 1418 psig AND SI flow is indicated on FI-943, CHG LOOP B CLD/HOT LG FLOW GPM. b. Stop all RCPs. EOP-2.0 RO 2 Verify no SG is FAULTED: No SG pressure is decreasing in an uncontrolled manner. No SG is completely depressurized. EOP-2.0

• RO 3 Check INTACT SG levels: a. Verify Narrow Range level in INTACT SGs is GREATER THAN 26%. b. Control EFW flow to maintain Narrow Range level in each INTACT SG between 40% and 60%. EOP-2.0 RO 4 Reset both SI RESET TRAIN A(B) Switches. EOP-2.0 RO 5 Reset Containment Isolation: RESET PHASE A - TRAIN A(B) CNTMT ISOL. RESET PHASE B - TRAIN A(B) CNTMT ISOL. EOP-2.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 43 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 RO 6 Check if Secondary radiation levels are normal: a. Check radiation levels normal on: RM-G19A(B)(C), STMLN HI RNG GAMMA. RM-A9, CNDSR EXHAUST GAS ATMOS MONITOR. RM-L3, STEAM GENERATOR BLOWDOWN LIQUID MONITOR. RM-L10, SG BLOWDOWN CW DISCHARGE LIQUID MONITOR. b. Place SVX-9398A(B)(C), SG A(B)(C) SMPL ISOL, in AUTO. c. Notify Chemistry to sample all SG secondary sides, and screen samples for abnormal activity using a frisker. EOP-2.0

• RO 7 Check PZR PORVs and Block Valves: a. Verify power is available to the PZR PORV Block Valves: 1) MVG-8000A, RELIEF 445 A ISOL. 2) MVG-8000B, RELIEF 444 B ISOL. 3) MVG-8000C, RELIEF 445 B ISOL. EOP-2.0 CAUTION - Step 7.b If any PZR PORV opens because of high PZR pressure, Step 7.b should be repeated after pressure decreases to LESS THAN 2330 psig, to ensure the PORV recloses. EOP-2.0 RO b. Verify all PZR PORVs are closed. c. Verify at least one PZR PORV Block Valve is open. EOP-2.0 RO 8 Place both ESF LOADING SEQ A(B) RESETS to: a. NON-ESF LCKOUTS. b. AUTO-START BLOCKS. EOP-2.0 RO 9 Establish Instrument Air to the RB: a. Start one Instrument Air Compressor and place the other in Standby. b. Open PVA-2659, INST AIR TO RB AIR SERV. c. Open PVT-2660, AIR SPLY TO RB. EOP-2.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 44 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1

• RO 10 Check if SI flow should be reduced: a. RCS subcooling on TI-499A(B), A(B) TEMP °F, is GREATER THAN 52.5°F. b. Secondary Heat Sink is adequate: Total EFW flow to INTACT SGs is GREATER THAN 450 gpm. OR Narrow Range level is GREATER THAN 26% in at least one INTACT SG. c. RCS pressure is stable OR increasing. EOP-2.0 NOTE - Step 10.d If PZR level is LESS THAN 10% [28%], the PZR should refill from SI flow after pressure is stabilized. EOP-2.0 RO d. PZR level is GREATER THAN 10%. EOP-2.0 RO e. GO TO EOP-1.2, SAFETY INJECTION TERMINATION, Step 1. EOP-2.0 REFERENCE PAGE FOR EOP-1.2 1 SI REINITIATION CRITERIA Following SI termination, IF either of the following conditions occurs, THEN start Charging Pumps and operate valves as necessary, and GO TO EOP-2.0, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1: RCS subcooling on TI-499A(B), A(B) TEMP °F, is LESS THAN 52.5°F [67.5°F]. OR PZR level can NOT be maintained GREATER THAN 10% [28%]. 2 SECONDARY INTEGRITY TRANSITION CRITERIA IF any unisolated SG pressure is decreasing in an uncontrolled manner OR is completely depressurized, THEN GO TO EOP-3.0, FAULTED STEAM GENERATOR ISOLATION, Step 1. 3 REDUCING CONTROL ROOM EMERGENCY VENTILATION Reduce Control Room Emergency Ventilation to one train in operation within 30 minutes of actuation. REFER TO SOP-505, CONTROL BUILDING VENTILATION SYSTEM. EOP-1.2 NOTE The EOP REFERENCE PAGE should be monitored throughout the use of this procedure. EOP-1.2 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 45 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 RO 1 Stop all but one Charging Pump and place in Standby. EOP-1.2 RO 2 Verify RCS pressure is stable OR increasing. EOP-1.2 RO 3 Establish Normal Charging: a. Close FCV-122, CHG FLOW. b. Open both MVG-8107 and MVG-8108, CHG LINE ISOL. c. Adjust FCV-122, CHG FLOW, to obtain 70 gpm Charging flow. d. Close both MVG-8801A(B), HI HEAD TO COLD LEG INJ. EOP-1.2 RO 4 Control FCV-122, CHG FLOW, to maintain PZR level. EOP-1.2 RO 5 Check if RHR Pumps should be stopped: a. Check if any RHR Pump is running with suction aligned to the RWST. b. Stop any RHR Pump which is running with suction aligned to the RWST and place in Standby. EOP-1.2 RO 6 Verify SI flow is NOT required: a. RCS subcooling on TI-499A(B), A(B) TEMP °F, is GREATER THAN 52.5°F. b. PZR level is GREATER THAN 10%. EOP-1.2 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # 8 Page: 46 of 49 Event

Description:

LOCA Outside the Reactor Building -Start Failure. Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 RO 7 Check if Letdown can be established: a. Verify PZR level is GREATER THAN 22%. b. Establish Normal Letdown: 1) Adjust FCV-122, CHG FLOW, to obtain 70 gpm Charging flow. 2) Set PCV-145, LO PRESS LTDN, to 70%. 3) Open TCV-144, CC TO LTDN HX. 4) Open PVT-8152, LTDN LINE ISOL. 5) Open both LCV-459 and LCV-460, LTDN LINE ISOL. 6) Open desired Orifice Isolation Valve(s) to obtain 60 gpm to 120 gpm: PVT-8149A, LTDN ORIFICE A ISOL (45 gpm). PVT-8149B, LTDN ORIFICE B ISOL (60 gpm). PVT-8149C, LTDN ORIFICE C ISOL (60 gpm). 7) Adjust FCV-122, CHG FLOW, to maintain TI-140, REGEN HX OUT TEMP °F, between 250°F and 350°F while maintaining PZR level. 8) Adjust PCV-145, LO PRESS LTDN, to maintain PI-145, LO PRESS LTDN PRESS PSIG, between 300 psig and 400 psig. 9) Place PCV-145, LO PRESS LTDN, in AUTO. 10) Place TCV-144, CC TO LTDN HX, in AUTO. EOP-1.2 EVALUATOR NOTE: Terminate scenario after normal charging and letdown is established.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # NA Page: 47 of 49 Event

Description:

SI Equipment Verification (ATTACHMENT 3) Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 EVALUATOR NOTE: , Pumps and Valves will need to be manually positioned to their required condition. BOP 1 Ensure EFW Pumps are running: a. Ensure both MD EFW Pumps are running. b. Verify the TD EFW Pump is running if necessary to maintain SG levels. EOP-1.0 Attachment 3 BOP 2 Ensure the following EFW valves are open: FCV-3531(3541)(3551), MD EFP TO SG A(B)(C). FCV-3536(3546)(3556), TD EFP TO SG A(B)(C). MVG-2802A(B), MS LOOP B(C) TO TD EFP. Attachment 3 BOP 3 Verify total EFW flow is GREATER THAN 450 gpm. Attachment 3 BOP 4 Ensure FW Isolation: a. Ensure the following are closed: FW Flow Control, FCV-478(488)(498). FW Isolation, PVG-1611A(B)(C). FW Flow Control Bypass, FCV-3321(3331)(3341). SG Blowdown, PVG-503A(B)(C). SG Sample, SVX-9398A(B)(C). b. Ensure all Main FW Pumps are tripped. Attachment 3 EVALUATOR NOTE: It is a critical task , if it has not been previously started, to provide High Head Safety Injection. CRITICAL TASK BOP 5 Ensure SI Pumps are running: Two Charging Pumps are running. Both RHR Pumps are running. Attachment 3 BOP 6 Ensure two RBCU Fans are running in slow speed (one per train). Attachment 3 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # NA Page: 48 of 49 Event

Description:

SI Equipment Verification (ATTACHMENT 3) Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOP 7 Verify Service Water to the RBCUs: a. Ensure two Service Water Pumps are running. b. Verify both Service Water Booster Pumps A(B) are running. c. Verify GREATER THAN 2000 gpm flow for each train on: FI-4466, SWBP A DISCH FLOW GPM. FI-4496, SWBP B DISCH FLOW GPM. Attachment 3 BOP 8 Verify two CCW Pumps are running. Attachment 3 BOP 9 Ensure two Chilled Water Pumps and Chillers are running. Attachment 3 BOP 10 Verify both trains of Control Room Ventilation are running in Emergency Mode. Attachment 3 BOP 11 Check if Main Steamlines should be isolated: a. Check if any of the following conditions are met: RB pressure GREATER THAN 6.35 psig. OR Steamline pressure LESS THAN 675 psig. OR Steamline flow GREATER THAN 1.6 MPPH AND Tavg LESS THAN 552°F. b. Ensure all the following are closed: MS Isolation Valves, PVM-2801A(B)(C). MS Isolation Bypass Valves, PVM-2869A(B)(C). Attachment 3 BOP 12 Ensure Excess Letdown Isolation Valves are closed: PVT-8153, XS LTDN ISOL. PVT-8154, XS LTDN ISOL. Attachment 3 BOP 13 Verify ESF monitor lights indicate Phase A AND Containment Ventilation Isolation on XCP-6103, 6104, and 6106. REFER TO ATTACHMENT 4, CONTAINMENT ISOLATION VALVE MCB STATUS LIGHT LOCATIONS, as needed. Attachment 3 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 1 Event # NA Page: 49 of 49 Event

Description:

SI Equipment Verification (ATTACHMENT 3) Time Position NRC 2015 - Scenario 1 NUREG -1021 R9 S1 BOP 14 Verify proper SI alignment: a. Verify SI valve alignment by verifying SAFETY INJECTION/PHASE A ISOL monitor lights are bright on XCP-6104. Attachment 3 BOP b. Verify all SAFETY INJECTION monitor lights are dim on XCP-6106. c. Verify SI flow on FI-943, CHG LOOP B CLD/HOT LG FLOW GPM. d. Check if RCS pressure is LESS THAN 325 psig. (NO) ALTERNATIVE ACTION d. ATTACHMENT 3, SI EQUIPMENT VERIFICATION, is complete. Attachment 3 BOP Report completion of Attachment 3. EVALUATOR NOTE: ATTACHMENT 3 is complete.

TURNOVER NOTES (read at the start of the scenario) Turnover Notes Mode 2 // 10E-3% Power // Work Week B1 // 2 Trains VU // EOOS: Yellow (LOSP x 2 Thunderstorms and ASI) // Grid Risk: Red // FEP Risk: Green The plant has completed a Mid-Cycle outage to repair a steam leak on the turbine. The Reactor is critical at 10-3 % power. Critical Data has been recorded. Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. The secondary has been warmed and the MSIVs are open. The National Weather Service has issued a Severe Weather Warning for Richland, Fairfield and Kershaw counties for the next one hour. Current RCS Boron Concentration by chemistry sample is 652 ppm. The BOP will be directed to start the 2B RBCU and then stop the 1B RBCU following turnover due to a request from Engineering. GOP-3, Reactor Startup From Hot Standby To Startup (Mode 3 to Mode 2) to step 3.14 GOP-4A Power Operation (Mode 1 - Ascending) has been started. Continue the Reactor Startup.

OAP-100.6 ATTACHMENT VIII PAGE 1 OF 2 REVISION 4 CONTROL ROOM SUPERVISOR RELIEF CHECKLIST DATE/TIME: today RELIEF SECTION Turnover Notes Mode 2 // 10E-3% Power // Work Week B1 // 2 Trains VU // EOOS: Yellow (LOSP x 2 Thunderstorms and ASI) // Grid Risk: Red // FEP Risk: Green The plant has completed a Mid-Cycle outage to repair a steam leak on the turbine. The Reactor is critical at 10-3 % power. Critical Data has been recorded. Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. The secondary has been warmed and the MSIVs are open. The National Weather Service has issued a Severe Weather Warning for Richland, Fairfield and Kershaw counties for the next one hour. Current RCS Boron Concentration by chemistry sample is 652 ppm. The BOP will be directed to start the 2B RBCU and then stop the 1B RBCU following turnover due to a request from Engineering. GOP-3, Reactor Startup From Hot Standby To Startup (Mode 3 to Mode 2) to step 3.14 GOP-4A Power Operation (Mode 1 - Ascending) has been started. Continue the Reactor Startup. Offgoing Control Room Supervisor Operations in progress (GOPs, SOPs, load changes, etc.): GOP-3, Reactor Startup From Hot Standby To Startup (Mode 3 to Mode 2) to Step 3.14 with additional steps completed as possible GOP-4A, Power Operation (Mode 1 - Ascending), has been started Operations scheduled for oncoming shifts: Direct the BOP to start the 2B RBCU and then stop the 1B RBCU following turnover due to a request from Engineering. Continue up-power in accordance with the reactivity plan Plant safeguard systems in degraded status: Initials In the Control Room, all books are replaced, the desk and console tops are clear, and all trash is properly disposed of. CRS Station Log completed. CRS OAP-100.6 ATTACHMENT VIII PAGE 2 OF 2 REVISION 4 C02 To the best of my knowledge, I am fully qualified to assume this watch taking into consideration fitness for duty, requalification status, and minimum watchstanding qualification. Shift relief completed: Oncoming Control Room Supervisor Offgoing Control Room Supervisor CR Supervisor Shift Supervisor review Oncoming Control Room Supervisor Initials Oncoming watch has reviewed the VCS Switchgear mailbox for switching orders. Plant Status (to be completed prior to turnover): Plant ESF System Status: Component Cooling System Service water System Reactor Building Cooling System Reactor Building Spray System Accumulator Tanks RHR System Charging/Safety Injection System Emergency Feedwater System Accumulator Tanks Diesel Generator Chilled Water System Control Room Ventilation System Position indications, power availability, and annunciator alarms are normal for present plant conditions. Plant Parameters Limit Reactor Power 0-100% RCS Tavg 589.2°F per loop RCS Pressure <2385 psig RCS Flow >100% per loop RCS Subcooling Normal All parameters within allowable limits for plant conditions. If not, what actions are being taken to correct conditions: Review of Logs: Station Log Removal and Restoration Log Tagout Log Special Orders Shift Turnover (to be completed during turnover): Briefing on plant conditions by offgoing Control Room Supervisor. Review of SPDS and BISI displays. Discussion of Protected Equipment. Identification of in-progress procedures including their present status and locations.

OAP-100.6 ATTACHMENT IX PAGE 1 OF 2 REVISION 4 REACTOR OPERATOR RELIEF CHECKLIST DATE/TIME: today LOG SECTION Date Entry RELIEF SECTION Entry Turnover Notes Mode 2 // 10E-3% Power // Work Week B1 // 2 Trains VU // EOOS: Yellow (LOSP x 2 Thunderstorms and ASI) // Grid Risk: Red // FEP Risk: Green The plant has completed a Mid-Cycle outage to repair a steam leak on the turbine. The Reactor is critical at 10-3 % power. Critical Data has been recorded. Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. The secondary has been warmed and the MSIVs are open. The National Weather Service has issued a Severe Weather Warning for Richland, Fairfield and Kershaw counties for the next one hour. Current RCS Boron Concentration by chemistry sample is 652 ppm. The BOP will be directed to start the 2B RBCU and then stop the 1B RBCU following turnover due to a request from Engineering. GOP-3, Reactor Startup From Hot Standby To Startup (Mode 3 to Mode 2) to step 3.14 GOP-4A Power Operation (Mode 1 - Ascending) has been started. Continue the Reactor Startup. Offgoing Reactor Operator Initials Main Control Board (Reactor Operator portion) properly aligned for the applicable mode. RO Housekeeping is satisfactory in the Reactor Operator area of responsibility. RO Discussion of Protected Equipment. RO Oncoming Reactor Operator Initials Review of HVAC Panel. Review of Station Log. Review of Removal & Restoration Log. Review of Main Control Board Panels. System Alignment A B C Train aligned to Reasons for any inoperable equipment Service Water Pumps X X A Component Cooling Pumps X A Charging Pumps X A HVAC Chillers X X A Reactor Building Spray Pumps RHR Pumps TDEFP OAP-100.6 ATTACHMENT IX PAGE 2 OF 2 REVISION 4 Emergency Feedwater Pumps Inoperable Radiation Monitors C02 To the best of my knowledge, I am fully qualified to assume this watch taking into consideration fitness for duty, requalification status, and minimum watchstanding qualification. Shift relief completed: Oncoming Reactor Operator Offgoing Reactor Operator Reactor Operator Shift Supervisor review OAP-100.6 ATTACHMENT X PAGE 1 OF 1 REVISION 4 BALANCE OF PLANT RELIEF CHECKLIST Date Entry RELIEF SECTION Entry Turnover Notes Mode 2 // 10E-3% Power // Work Week B1 // 2 Trains VU // EOOS: Yellow (LOSP x 2 Thunderstorms and ASI) // Grid Risk: Red // FEP Risk: Green The plant has completed a Mid-Cycle outage to repair a steam leak on the turbine. The Reactor is critical at 10-3 % power. Critical Data has been recorded. Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. The secondary has been warmed and the MSIVs are open. The National Weather Service has issued a Severe Weather Warning for Richland, Fairfield and Kershaw counties for the next one hour. Current RCS Boron Concentration by chemistry sample is 652 ppm. The BOP will be directed to start the 2B RBCU and then stop the 1B RBCU following turnover due to a request from Engineering. GOP-3, Reactor Startup From Hot Standby To Startup (Mode 3 to Mode 2) to step 3.14 GOP-4A Power Operation (Mode 1 - Ascending) has been started. Continue the Reactor Startup. Offgoing Reactor Operator Initials Main Control Board (Reactor Operator portion) properly aligned for the applicable mode. BOP Housekeeping is satisfactory in the Reactor Operator area of responsibility. BOP Discussion of Protected Equipment. BOP Oncoming Reactor Operator Initials Review of Main Control Room Panels. Review of Station Log. Review of Removal & Restoration Log. Test annunciator lights (with Offgoing operator concurrence). C02 To the best of my knowledge, I am fully qualified to assume this watch taking into consideration fitness for duty, requalification status, and minimum watchstanding qualification. Shift relief completed: Oncoming Balance of Plant Offgoing Balance of Plant Balance of Plant Shift Supervisor review OAP-100.6 ATTACHMENT IB PAGE 1 OF 2 REVISION 3 REACTIVITY MANAGEMENT BRIEF MODES 1 - 3 NOTE PART 1 REACTIVITY MANAGEMENT TURNOVER should be read at Shift Turnover Meeting. PART 2 REACTOR STATUS should be discussed between the NROATC, BOP, and CRS.

PART 1 REACTIVITY MANAGEMENT TURNOVER: Date of last Automatic or Manual Make-Up: Is Auto Makeup expected this shift (circle)? YES NO Expected Boric Acid total gallons on a normal Auto Makeup based on current BAT in service: gallons FCV 113 A&B, pot setting for current RCS boron concentration: Expected Boric Acid flowrate for VCT makeup: Total gallons Diluted Borated (Last Shift)

Last evolution (circle one): Borate / Dilute / Blended Expected Borations, Dilutions, or Blended changes to the RCS: List Reactivity Concerns in progress or planned and action(s) necessary (i.e. Steam or Feed Flow transmitter in test, Steam Generator Blowdown out of service, Calorimetric inputs in service, etc.).

OAP-100.6 ATTACHMENT IB PAGE 2 OF 2 REVISION 3 REACTIVITY MANAGEMENT BRIEF MODES 1 - 3 (Cont'd) PART 2 REACTOR STATUS: (circle one below) Delta I on Target (+ 2%)? YES NO Not in Mode 1 If NO is circled, identify plan to re-establish target band: Xenon Trend: Stable Building In Burning Out Demineralizers: Mixed Bed in service: A B PRC01 Y / N Standby Demineralizer: Filled Borated Empty PRC01 Cation Bed: Date last in service Boron Concentration when in service ATTACHMENT IA reviewed and current: YES NO Midnight Boron Concentration and Date when CHG/SI pump was secured: CB A Date CB B Date CB C Date CHG G REP-102.001 ATTACHMENT I PAGE 1 OF 1 REVISION 6 CYCLE 22 PLAN# 2015 - TRNG REACTIVITY MANAGEMENT PLAN VERIFICATION BEACON Filenames: Model Input filename training Summary Results filename training Calibration filename training Power Profile filename training Sign and date steps below to document performance. Step Number Signature Date *3.0 Prerequisites Signature 1 today *7.36 Verify 9.0 Criteria Signature 1 today *7.38 RE Verifier RE Signature today *7.39 Operations Reviewer Ops Signature today COMMENTS REP-102.001 ATTACHMENT II PAGE 1 OF 1 REVISION 6 CYCLE 22 PLAN# 2015 - TRNG REACTIVITY MANAGEMENT PLAN VERIFICATION OPERATIONS GUIDELINES FOR CYCLE 22 MID-CYCLE STARTUP INITIAL CONDITIONS AND ASSUMPTIONS: Reactor is at 2% RTP Burnup is approximately 20,000 MWD/MTU RCS Boron is approximately 652 ppm D Bank is approximately 110 steps TRANSIENT ASSUMPTIONS: Change power per Attachment II schedule PREDICTION CONSTRAINTS: Use control bank D and boron for reactivity compensation. Maintain Control Bank D position at least 15 steps above RIL. Note: See attached predictive trends. (BEACON predicted xenon will NOT match the xenon displayed on the plant computer.) Contact the following if there are questions about this guidance: Reactor Engineering office pager Mike Strickland 54625 251-5767 Damon Bryson 54814 733-7618 Nate Smith 54733 758-8590 Bill Herwig 54414 540-9111 REP-102.001 ATTACHMENT II PAGE 1 OF 1 REVISION 6 CYCLE 22 PLAN# 2015 - TRNG REACTIVITY MANAGEMENT PLAN VERIFICATION PROPOSED POWER MANEUVER Comments (e.g. control rod or boron issues, Time after Start Reactor Power activities to be performed, holds, etc.) 00:00 2% Begin startup 31:00 100% 100% power COMMENTS list power plateau activities, unusual operational restraints, contingency plans, alternate power history variations to address, time periods to avoid boration, etc. This plan increases power continuously at 3% per hour until 100% power.

HoursDTotalTotalRAOCRAOCXenonRILAfterRxBankBoronBoron WaterBoron WaterDelta-IBandBandWorthLimitStartPowerPosPPM(gal)(gal)(gal)(gal)(%)LowHigh(pcm)(steps)0.002%110652.200001.03-22.020.0000.254%110646.4044804482.04-22.020.0-100.506%126646.40004482.47-22.020.0-420.758%134646.40004483.06-22.020.0-761.0010%143646.40004483.60-22.020.0-10101.2511%149646.40004484.26-22.020.0-15131.5013%155646.40004484.96-22.020.0-21171.7515%161646.40004485.68-22.020.0-29212.0015%162646.40004485.69-22.020.0-37212.2516%164646.40004486.02-22.020.0-46222.5017%167646.40004486.37-22.020.0-56242.7517%170646.40004486.80-22.020.0-67253.0018%173646.40004487.30-22.020.0-79273.2519%159627.301513019615.82-22.020.0-88283.5020%140608.801514034754.46-22.020.0-98303.7520%118590.701514049894.16-22.020.0-111314.0021%90573.201515065045.29-22.020.0-129334.2522%78556.201515080194.30-22.020.0-139344.5022%69539.701516095353.15-22.020.0-149364.7523%59523.6015170110522.23-22.020.0-159375.0024%54512.7010500121011.97-22.020.0-173395.2525%55510.202460123471.73-22.020.0-188405.5025%57509.101080124551.47-22.020.0-203425.7526%58506.802330126881.27-22.020.0-219436.0027%60505.601170128061.09-22.020.0-236456.2528%61503.202380130440.98-22.020.0-253466.5028%63501.901280131710.78-22.020.0-271486.7529%64499.302590134300.66-22.020.0-290497.0030%66498.001350135650.51-22.020.0-309517.2530%67495.302690138330.39-22.020.0-328527.5031%69493.901440139770.26-22.020.0-348547.7532%70491.202740142510.20-22.020.0-368558.0033%72489.701570144080.09-22.020.0-389568.2533%73486.802880146960.03-22.020.0-410588.5034%75485.20173014869-0.04-22.020.0-432608.7535%76482.20309015177-0.10-22.020.0-454619.0036%77479.10324015501-0.30-22.020.0-475639.2536%79477.10201015702-0.36-22.020.0-498649.5037%80474.00335016037-0.55-22.020.0-519659.7538%82472.10198016235-0.67-22.020.0-5426710.0038%83468.80348016583-0.71-22.020.0-5656810.2539%85466.50245016828-0.84-22.020.0-5877010.5040%86463.00377017206-0.90-22.020.0-6117110.7541%88460.50273017479-1.11-22.020.0-6337311.0041%89457.10368017847-1.39-22.020.0-6567411.2542%91454.10331018178-1.77-22.020.0-6787611.5043%92450.20425018603-2.04-22.020.0-7017711.7544%94447.20338018941-2.52-22.020.0-7237912.0044%95443.30426019367-2.62-22.020.0-7478012.2545%97440.20357019724-3.16-22.020.0-7688212.5046%98436.30445020170-3.56-22.020.0-79183Cycle 22 Simulator 20k MWD/MTU Startup 2-100%Tech Spec Ref: N/AProcedure Ref: REP-102.001Figure Ref: N/A HoursDTotalTotalRAOCRAOCXenonRILAfterRxBankBoronBoron WaterBoron WaterDelta-IBandBandWorthLimitStartPowerPosPPM(gal)(gal)(gal)(gal)(%)LowHigh(pcm)(steps)Cycle 22 Simulator 20k MWD/MTU Startup 2-100%12.7546%112440.12802820170-4.47-22.020.0-8088513.0047%124442.82004820170-4.46-22.020.0-8278613.2548%136446.62807620170-3.46-22.020.0-8498813.5049%136442.404697620639-3.51-22.020.0-8738913.7549%136438.004937621132-3.57-22.020.0-8969114.0050%136433.704967621628-3.71-22.020.0-9199214.2551%148438.939011521628-1.42-21.819.8-9439414.5052%148434.3053011522158-1.60-21.619.7-9669514.7552%148429.9051011522667-1.86-21.419.5-9889715.0053%148425.4052411523191-1.92-21.219.3-10119815.2554%151422.6032811523520-1.90-21.019.1-103410015.5054%151418.7046011523980-1.96-20.818.9-105610115.7555%152415.0043711524417-1.99-20.618.8-107910316.0056%153411.4043611524853-2.04-20.418.6-110110416.2557%154407.8042811525281-2.06-20.218.4-112310616.5057%155404.3043111525712-2.03-19.918.2-114510716.7558%155400.1052511526236-2.06-19.718.1-116710917.0059%156396.5044311526680-2.09-19.517.9-118911017.2560%156392.3052411527204-2.13-19.317.7-121111217.5060%157388.7045411527658-2.23-19.117.5-123311317.7561%158385.4043111528089-2.26-18.917.4-125411518.0062%159381.9044811528537-2.28-18.717.2-127511618.2563%160378.3046411529001-2.30-18.517.0-129611818.5063%161374.9044811529449-2.31-18.316.8-131711918.7564%162371.5045311529902-2.32-18.116.7-133812119.0065%163367.8049111530393-2.33-17.916.5-135812219.2565%164364.5044811530842-2.30-17.716.3-137912319.5066%165361.0047011531312-2.31-17.516.1-139912519.7567%165357.2053311531845-2.35-17.315.9-141912720.0068%166353.7048311532328-2.50-17.115.8-143812820.2568%167350.5043911532767-2.53-16.915.6-145813020.5069%168347.2046511533232-2.56-16.715.4-147713120.7570%169343.9047011533703-2.56-16.415.2-149613321.0071%170340.8045511534158-2.58-16.215.1-151413421.2571%171337.6046411534622-2.54-16.014.9-153313621.5072%172334.3047111535093-2.55-15.814.7-155113721.7573%173331.2046311535556-2.49-15.614.5-156913922.0074%173327.4056211536118-2.53-15.414.4-158714022.2574%174324.1049611536614-2.57-15.214.2-160514222.5075%175321.0047911537093-2.67-15.014.0-162214322.7576%176318.0045311537546-2.70-14.813.8-163914423.0077%177315.0047211538017-2.73-14.613.7-165614623.2577%178312.0047311538490-2.74-14.413.5-167314823.5078%179309.0046811538958-2.75-14.213.3-168914923.7579%180306.1046611539424-2.70-14.013.1-170615124.0079%181303.2045911539883-2.70-13.812.9-172215224.2580%181299.6057711540460-2.75-13.612.8-173815324.5081%182296.5050611540966-2.85-13.412.6-175315524.7582%183293.7047011541436-2.89-13.212.4-176915725.0082%184290.8047611541912-2.96-12.912.2-178315825.2583%185288.0047211542385-2.98-12.712.1-1799160Tech Spec Ref: N/AProcedure Ref: REP-102.001Figure Ref: N/A HoursDTotalTotalRAOCRAOCXenonRILAfterRxBankBoronBoron WaterBoron WaterDelta-IBandBandWorthLimitStartPowerPosPPM(gal)(gal)(gal)(gal)(%)LowHigh(pcm)(steps)Cycle 22 Simulator 20k MWD/MTU Startup 2-100%25.5084%186285.2048111542865-3.00-12.511.9-181316125.7585%187282.4048211543347-2.99-12.311.7-182816326.0085%188279.6048011543827-3.00-12.111.5-184216426.2586%189277.0046311544291-2.94-11.911.4-185616526.5087%190274.3046811544759-2.95-11.711.2-187016726.7588%191271.6048411545243-2.89-11.511.0-188316927.0088%191268.3059111545834-2.93-11.310.8-189717027.2589%192265.5050911546343-2.99-11.110.7-191117227.5090%193262.8048711546830-3.12-10.910.5-192417327.7590%194260.1049611547325-3.21-10.710.3-193717428.0091%195257.4050311547828-3.29-10.510.1-195017628.2592%196254.8049011548318-3.33-10.39.9-196217728.5093%197252.2049711548815-3.39-10.19.8-197517928.7593%198249.7048111549295-3.42-9.99.6-198718129.0094%199247.2049911549794-3.45-9.79.4-199918229.2595%200244.6049811550292-3.48-9.49.2-201118429.5096%201242.0051811550810-3.51-9.29.1-202318529.7596%203240.0039011551200-3.38-9.08.9-203418630.0097%204237.4052311551723-3.21-8.88.7-204518830.2598%205235.0049811552221-3.22-8.68.5-205619030.5099%206232.4053611552758-3.25-8.48.4-206819130.7599%208230.4041711553174-3.21-8.28.2-207819331.00100%209227.8053511553710-3.20-8.08.0-2089194Tech Spec Ref: N/AProcedure Ref: REP-102.001Figure Ref: N/A OAP-102.1 ATTACHMENT II PAGE 1 OF 1 REVISION 7 SCHEDULED WORK APPROVAL/DENIAL Scheduled Work/Activity Date Description of Work/Activity to be performed: I. This Moderate Risk, Elevated Risk, High Risk, or Cross Train activity is approved for work provided the required plant conditions are available on the scheduled due date. OR This specific activity has been reviewed for EOOS Risk Reassessment. Set EOOS Environmental Variance __________________________ Set Risk at Times The following items were considered for making this approval: Operations Supervisor (Moderate Risk or Cross Train) In the absence of the Operations Supervisor: Operations Scheduling, Shift Supervisor GMNPO/MDS (Elevated Risk) PSRC (High Risk) II. This work activity/package cannot be performed on the scheduled date due to the following reason(s): SRO (WCC or On Shift) Operations Scheduling Supervisor III. Recommended re-schedule date or plant conditions:

SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION NUCLEAR OPERATIONS NUCLEAR OPERATIONS COPY NO. GENERAL OPERATING PROCEDURE GOP-3 REACTOR STARTUP FROM HOT STANDBY TO STARTUP (MODE 3 TO MODE 2) REVISION 13 SAFETY RELATED

RECORD OF CHANGES CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE A P 01/25/10 B P 06/19/12 C P 07/02/12 D P 04/26/14 E P 06/30/14 F P 11/14/14 CONTINUOUS USE Continuous Use of Procedure Required. Read Each Step Prior to Performing.

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GOP-3 PAGE i REVISION 13 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE/SCOPE 1 2.0 INITIAL CONDITIONS 2 3.0 INSTRUCTIONS 4

4.0 REFERENCES

21

ATTACHMENTS Attachment I - Sign-off Identification List

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GOP-3 REVISION 13 PAGE 1 OF 22 1.0 PURPOSE/SCOPE 1.1 This procedure provides instructions for Reactor Startup, from Hot Standby to Startup.

1.2 The following governing regulations apply to this procedure: a. 10CFR50.59.

b. 10CFR50, Appendix B.

c. SAP-630, Procedure/Commitment Accountability Program.

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GOP-3 REVISION 13 PAGE 2 OF 22 NOTE 2.0 and 3.0 a. All personnel who sign off steps in this procedure must enter their names and initials on Attachment I. b. Each step should be initialed and dated when all its substeps are either completed and checked-off or marked N/A and initialed. NOTE 2.0 If this procedure must be initiated under conditions other than those in Section 2.0, INITIAL CONDITIONS, the Shift Supervisor or Control Room Supervisor will review Sections 2.0, INITIAL CONDITIONS, and 3.0, INSTRUCTIONS. Steps that are not applicable due to plant conditions will be marked N/A and initialed by the Shift Supervisor or Control Room Supervisor. All other items will require sign-off or check-off. 2.0 INITIAL CONDITIONS INITIALS/DATE 2.1 RCS status is as follows: /

a. System temperature is being maintained between 555°F and 559°F using the Bank 1 Condenser Steam Dumps or Steamline PORVs.

b. System pressure is being maintained between 2230 psig and 2240 psig in AUTO control.

c. All Reactor Coolant Pumps are in operation.

d. Pressurizer level is being maintained at 25% in AUTO control. 2.2 All Safety Injection Systems are aligned and operable. /

2.3 Excore NIs are aligned for critical operation per SOP-404, / Excore Nuclear Instrumentation System.

2.4 The Reactor is shutdown with all Control Bank Rods fully / inserted.

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GOP-3 REVISION 13 PAGE 3 OF 22 INITIALS/DATE 2.5 Shutdown Margin is being maintained for Mode 3 conditions / per STP-134.001, Shutdown Margin Verification. 2.6 Reactor Makeup Control is in AUTO and set for blended flow / equal to the existing boron concentration.

2.7 Secondary Plant status is as follows: /

a. The Main Turbine is on the Turning Gear per SOP-215, Main Turbine Lube Oil Supply System.

b. The Main Feedwater Pumps are on their Turning Gears per SOP-209, Feedwater Turbine Lube Oil System.

c. Narrow Range Steam Generator levels are being maintained between 60% and 65% with chemistry within specification using the following:

1) Blowdown per SOP-212, Steam Generator Blowdown.

2) Emergency Feedwater per SOP-211, Emergency Feedwater System. d. Main Steam is being warmed per SOP-201, Main Steam System.

e. Feedwater is being warmed per SOP-210, Feedwater System.

f. Condensate is in operation per SOP-208, Condensate System.

g. Circulating Water is in operation per SOP-207, Circulating Water. 2.8 The Rod Control and Position Indicating Systems are in / operation per SOP-403, Rod Control And Position Indicating System.

2.9 The Control Rod Drive Mechanism Ventilation System is in / operation per SOP-114, Reactor Building Ventilation System.

2.10 GOP Appendix A review has been completed. /

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 4 OF 22 3.0 INSTRUCTIONS INITIALS/DATE 3.1 Shut down and isolate BTRS as follows: /

a. Place HCV-387, BTRS BYP FLOW, in BYP.

b. Place BTRS SELECT Switch in OFF. 3.2 Verify RCS Chemistry control for startup: / a. Contact Chemistry to ensure RCS Chemistry control is satisfactory for startup per CP-625, Chemistry Refueling Shutdown And Startup Plan.

b. Record current Boron concentration:

ppm 3.3 Perform the following if an RB entry is in progress or will occur / during the reactor startup:

a. Obtain the approval of the General Manager, Nuclear Plant Operations, for personnel to be in the RB during the reactor startup

b. Notify Health Physics that a reactor startup is about to commence and dose rates in the RB could change rapidly. CHG F GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 5 OF 22 INITIALS/DATE 3.4 Align Excore NIs for Reactor Startup as follows: / Z005 a. Ensure INI00033-NI, REMOTE SOURCE RANGE MONITOR, is de-energized with fuses removed per SOP-404, Excore Nuclear Instrumentation System, Section IV.F. Z007 b. Ensure the following Nuclear Instrumentation Channels are in operation per SOP-404, Excore Nuclear Instrumentation System,Section III.A and tested per the applicable STPs:

1) Two Source Range Channels.

2) Two Intermediate Range Channels.

3) At least three Power Range Channels. c. Verify both Source Range Channels are indicating a minimum of two counts per second.

d. Perform either of the following to monitor Source and Intermediate Range Channels as follows:

1) Select the highest reading Source Range Channel and either Intermediate Range Channel on recorder NR-45, NIS RECORDER.

2) Monitor the highest reading Source Range Channel and either Intermediate Range Channel using computer display NR45 in FAST SPEED. CHG D GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 6 OF 22 INITIALS/DATE Step 3.4 continued NOTE 3.4.e Audio Count Rate is not required to be operable. e. At the AUDIO COUNT RATE CHANNEL drawer, perform the following:

1) Select the highest reading Source Range Channel on the CHANNEL SELECTOR Switch.

2) Adjust the AUDIO MULTIPLIER Switch as necessary to maintain a distinguishable audio countrate.

3) Place the SR COUNTER/SCALER, POWER switch in the POWER position. 3.5 Complete Attachment III.A, Prior to Closing Reactor Trip / Breakers in Modes 3, 4 & 5, of GTP-702. C01 3.6 Ensure the P-4 trip actuating device operational test is / N01 performed and Reactor Trip breakers are closed per STP-345.039, Reactor Trip P-4 Trip Actuating Device Operational Test. Z008 3.7 Ensure both Rod Control MG sets are supplying load to / to Rod Control per SOP-403, Rod Control and Position Indicating System,Section III.A. CHG B CHG D GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 7 OF 22 INITIALS/DATE 3.8 If necessary, withdraw the Shutdown Banks as follows: /

a. Verify Shutdown Margin Boron Concentration is satisfactory by performing STP-134.001, Shutdown Margin Verification for Mode 3 with S/D Banks OUT

b. Place ROD CNTRL START UP RESET Switch in START UP. CAUTION 3.8.c To minimize the possibility of binding at the full in position, rods should not be driven below the 000 indication on the Group Demand Step Counters.

c. Ensure the Step Counters indicate zero (000) steps. Z009 d. Update Rod Bank positions on the IPCS, refer to OAP-107.1, Control of IPCS Functions, Step 6.2.b.

e. Ensure IZM01200, DRPI Main Control Board Display Monitor, and IZM01201, DRPI Main Control Board Display Monitor, indicate RB.

f. Momentarily depress the ROD CNTRL ALARM RESET Pushbutton.

g. Verify ROD CNTRL SYS FAIL URGENT (XCP-620 5-1) and ROD CNTRL SYS FAIL NON-URGENT (XCP-620 5-5) alarms cleared. CHG C CHG D CHG F GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 8 OF 22 INITIALS/DATE Step 3.8 continued CAUTION 3.8.h To prevent any inadvertent inward rod motion the ROD CNTRL BANK SEL Switch should not be placed in or pass through AUTO. NOTE 3.8.h Reactor Coolant System temperature is being maintained between 555°F and 559°F using the Bank 1 Condenser Steam Dumps or Steamline PORVs. h. Place ROD CNTRL BANK SEL Switch in SBA. CAUTION 3.8.i 12 steps should NOT be exceeded until Rod Bottom lights are off. If all Shutdown Bank A Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered.

i. Using the ROD CONTROL ROD MOTION Lever, perform the following:

1) Withdraw Shutdown Bank A to ten Steps.

2) Verify that all RB lights for Shutdown Bank A are off.

3) Using the ROD CONTROL ROD MOTION Lever, withdraw SBA to 230 steps.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 9 OF 22 INITIALS/DATE Step 3.8 continued CAUTION 3.8.j To prevent any inadvertent inward rod motion the ROD CNTRL BANK SEL Switch should not be placed in or pass through AUTO.

j. Place ROD CNTRL BANK SEL Switch in SBB. CAUTION 3.8.k 12 steps should NOT be exceeded until Rod Bottom lights are off. If all Shutdown Bank B Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered.

k. Using the ROD CONTROL ROD MOTION Lever, perform the following:

1) Withdraw Shutdown Bank B to ten steps.

2) Verify that all RB lights for Shutdown Bank B are off.

3) Using the ROD CONTROL ROD MOTION Lever, withdraw SBB to 230 steps.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 10 OF 22 INITIALS/DATE 3.9 Contact Reactor Engineering for recommended rod heights / and Estimated Critical Condition information. NOTE 3.10 Reactor Coolant System temperature is being maintained between 555°F and 559°F using the Bank 1 Condenser Steam Dumps or Steamline PORVs.

3.10 Perform a Shutdown Margin verification per STP-134.001, / Shutdown Margin Verification, using Estimated Critical Condition boron, desired RCS temperature, and expected xenon.

STTS # NOTE 3.11 through 3.13 For initial criticality following refueling, REP-107.001, Controlling Procedure For Refueling Startup And Power Ascension Testing, is the controlling document for Reactor Startup. Appropriate steps of GOP-3 should be initialed as they are performed.

3.11 Prepare for Reactor Startup as follows: /

a. Adjust Boron concentration as required by Estimated Critical Condition calculation as follows: Z003 1) Borate or dilute per SOP-106, Z010 Reactor Makeup Water System, Z017 Sections III.D, III.E, or III.F.

2) When complete, direct Chemistry to sample the RCS and the Pressurizer for boron. CHG D GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 11 OF 22 INITIALS/DATE Step 3.11 continued

b. Block HIGH FLUX AT SHUTDOWN as follows:

1) Disable the IPCS High Flux At Shutdown alarm function as follows:

a) Type the Turn-On-Code HFAS.

b) Verify OPERATOR DISABLED is indicated above the ENABLE CALCS box.

c) If OPERATOR ENABLED is indicated, select DISABLE CALCS.

2) Place HIGH FLUX AT SHUTDOWN Switch for SOURCE RANGE N-31 in BLOCK.

3) Place HIGH FLUX AT SHUTDOWN Switch for SOURCE RANGE N-32 in BLOCK.

4) Verify SR HI FLUX AT SHUTDN BLOCK (XCP-620 4-4) annunciator alarms.

c. Review Estimated Critical Condition calculation within four hours prior to criticality, verifying predicted rod height is above the Rod Insertion Limit per Tech Spec 4.1.1.1.1.c. Time CHG D GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 12 OF 22 INITIALS/DATE Step 3.11 continued

d. Review the following for current status and limitations for Mode escalation:

1) Removal and Restoration Log.

2) Danger Tag Log.

3) 31 Day Surveillance Book.

4) Ensure completion of Attachment II.F, Operational Mode Change Plant Startup - Entering Mode 2, of GTP-702.

5) Ensure SAP-116, PLANT TRIP/SAFETY INJECTION PLANT RECOVERY, is completed, if necessary. Z011 e. Perform OAP-100.4, Communication, Attachment I, Mode Change Brief Checklist. f. Update the IPCS Plant Mode indicator to indicate Mode 2 as the current Plant Mode as follows:

1) Type the Turn-On-Code MODE to display the PLANT MODE CHANGE DISPLAY window

2) Select the SET MODE 2 Pushbutton.

3) Verify the selected Mode is displayed on the left end of the top toolbar.

g. Verify all Shutdown Bank Rods fully withdrawn within 15 minutes of commencing Control Bank Rod withdrawal. Time CHG D GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 13 OF 22 INITIALS/DATE Step 3.11 continued NOTE 3.11.h Reactor Coolant System temperature is being maintained between 555°F and 559°F using the Bank 1 Condenser Steam Dumps or Steamline PORVs. h. Obtain the Shift Supervisor's permission to commence a Reactor Startup.

i. Announce Reactor Startup over the page system. j. If used, place NR-45 CHART in HI speed. k. Initiate REP-109.002, Inverse Count Rate Ratio Plot. Time

l. If performing an initial cycle startup, refer to REP-107.001, Controlling Procedure For Refueling Startup And Power Ascension Testing, for additional actions.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 14 OF 22 INITIALS/DATE 3.12 Achieve Reactor criticality as follows: / a. Review GOP Appendix A, Generic Operating Precautions, for Reactor Startup. CAUTION 3.12.b To prevent any inadvertent inward rod motion the ROD CNTRL BANK SEL Switch should not be placed in or pass through AUTO. b. Place the ROD CNTRL BANK SEL Switch in MAN. NOTE 3.12.c A stable Startup Rate of one decade per minute should NOT be exceeded.

c. Using ROD CONTROL ROD MOTION lever, commence Control Bank Rod withdrawal to ten steps on Bank A. Time CAUTION 3.12.d 12 steps should NOT be exceeded until all Rod Bottom lights are off. If all Control Bank A Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered.

d. At ten steps on Control Bank A, stop and verify:

1) Bank A RB lights clear.

2) ONE ROD ON BOTTOM (XCP-621 3-1) annunciator clears.

3) RODS ON BOTTOM (XCP-621 3-2) annunciator clears. e. Recommence withdrawing rods while observing that the groups sequence properly.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 15 OF 22 INITIALS/DATE Step 3.12 continued CAUTION 3.12.f 12 steps should NOT be exceeded until all Rod Bottom lights are off. If all Control Bank B Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered.

f. At ten steps on Control Bank B, stop and verify Bank B RB lights clear. g. Recommence withdrawing rods while observing that the groups sequence properly.

h. Verify 102 step Bank Overlap between Control Bank A and Control Bank B. CAUTION 3.12.i 12 steps should NOT be exceeded until all Rod Bottom lights are off. If all Control Bank C Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered.

i. At ten steps on Control Bank C, stop and verify Bank C RB lights clear.

j. Recommence withdrawing rods while observing that the groups sequence properly.

k. Verify 102 step Bank Overlap between Control Bank B and Control Bank C.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 16 OF 22 INITIALS/DATE Step 3.12 continued CAUTION Step 3.12.l Reactor startup should be stopped and I&C notified if the CRB INSERT LMT LO-LO (XCP-621 1-1) annunciator fails to clear between 118 steps and 134 steps on Bank C.

l. Verify CRB INSERT LMT LO-LO (XCP-621 1-1) annunciator clears between 118 steps and 134 steps on Bank C. Steps CAUTION 3.12.m 12 steps should NOT be exceeded until all Rod Bottom lights are off. If all Control Bank D Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered. NOTE 3.12.m Reactor Coolant System temperature is being maintained between 555°F and 559°F using the Bank 1 Condenser Steam Dumps or Steamline PORVs.

m. At ten steps on Control Bank D, stop and verify Bank D RB lights clear

n. Recommence withdrawing rods while observing that the groups sequence properly.

o. Verify the CRB INSERT LMT LO (XCP-621 1-2) annunciator clears between 138 steps and 144 steps on Bank C. Steps

p. Verify 102 step Bank Overlap between Control Bank C and Control Bank D. CHG A GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 17 OF 22 INITIALS/DATE Step 3.12 continued q. Within 15 minutes before achieving criticality, verify Tavg greater than or equal to 551°F. Time Tave r. Announce criticality over the page system. Time s. Verify critical rod position is above the Rod Insertion Limit per Tech Spec 3.1.3.6.

t. Maintain as close to 0 SUR as reasonably achievable.

u. At the AUDIO COUNT RATE CHANNEL drawer, place the following switches in OFF: 1) AUDIO MULTIPLIER. 2) CHANNEL SELECTOR.

3) SR COUNTER/SCALER, POWER switch. (Toggle down)

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 18 OF 22 INITIALS/DATE 3.13 Increase Reactor Power to 10-3% as follows: / a. Establish a stable Startup Rate of less than one decade per minute. b. At 7.5x10-6%, perform the following: 1) Verify P6 Permissive energizes to bright. 2) Verify a minimum of one decade overlap between Source Range Channels and Intermediate Range Channels. c. Prior to 105 CPS, perform the following:

1) Momentarily place SR TRAIN A Switch in BLOCK. 2) Verify SR A TRIP BLCK Permissive energizes to bright. 3) Momentarily place SR TRAIN B Switch in BLOCK.

4) Verify SR B TRIP BLCK Permissive energizes to bright. d. Perform one of the following for continued monitoring of Intermediate and Power Range instrument:

1) If available for use, select one Intermediate Range Channel and one Power Range Channel on NR-45, NIS RECORDER.

2) Ensure at least one Intermediate Range and at least one Power Range instrument are selected for continuous monitoring using computer display NR45. e. Stabilize Reactor Power at 10-3%.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 19 OF 22 INITIALS/DATE Step 3.13 continued f. Record the following Critical Data:

1) RCS pressure: psig

2) Tavg: °F

3) Bank at steps

4) Boron Concentration: ppm

5) Time:

6) Stable Power:  %

g. If performing an initial cycle startup, refer to REP-107.001, Controlling Procedure For Refueling Startup And Power Ascension Testing, for physics testing instructions.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 20 OF 22 INITIALS/DATE CAUTION 3.14 While operating with a positive Moderator Temperature Coefficient: a. All reactivity additions should be slow and controlled. b. A stable Startup Rate of 0.3 decade per minute should not be exceeded. c. Rods should be moved in 1/2 step increments until the effect of rod motion has been evaluated. NOTE 3.14 Ensure sufficient Emergency Feedwater Flow exists prior to raising power.

3.14 Increase Reactor Power to between 1% and 3%. /

3.15 At the Point of Adding Heat, if NR-45, NIS RECORDER, / had previously been selected to HI speed place the recorder in LO speed. CAUTION 3.16 a. Adjustment of Tavg with the Rod Control System must not be attempted with the ROD CNTRL BANK SEL Switch in any position other than MAN. b. Manual rod control is required to establish equilibrium conditions, since C-5 blocks automatic rod withdrawal.

3.16 Maintain Tavg between 555°F and 559°F. /

3.17 Complete Attachment II.G, Operational Mode Change / Plant Startup - Entering Mode 1, of GTP-702.

3.18 Proceed to GOP-4A, Power Operation (Mode 1 - Ascending). /

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GOP-3 REVISION 13 PAGE 21 OF 22

4.0 REFERENCES

4.1 CP-625, Chemistry Refueling Shutdown And Startup Plan.

4.2 FSAR Section 5.0.

4.3 GOP Appendix A.

4.4 GOP-4A, Power Operation (Mode 1 - Ascending).

4.5 GTP-702, Operational Mode Change and Contingency Surveillance Requirements.

4.6 OAP-100.4, Communication.

4.7 REP-107.001, Controlling Procedure For Refueling Startup And Power Ascension Testing.

4.8 REP-109.002, Inverse Count Rate Ratio Plot.

4.9 SAP-630, Procedure / Commitment Accountability Program.

4.10 SOP-103, Boron Thermal Regeneration System.

4.11 SOP-106, Reactor Makeup Water System.

4.12 SOP-114, Reactor Building Ventilation System.

4.13 SOP-201, Main Steam System.

4.14 SOP-205, Turbine Sealing Steam System.

4.15 SOP-206, Main and Auxiliary Condenser Air Removal System.

4.16 SOP-207, Circulating Water.

4.17 SOP-208, Condensate System.

4.18 SOP-209, Feedwater Turbine Lube Oil System.

4.19 SOP-210, Feedwater System.

4.20 SOP-211, Emergency Feedwater System.

4.21 SOP-212, Steam Generator Blowdown. CHG F

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GOP-3 REVISION 13 PAGE 22 OF 22 4.22 SOP-215, Main Turbine Lube Oil Supply System.

4.23 SOP-403, Rod Control And Position Indicating System.

4.24 SOP-404, Excore Nuclear Instrumentation System.

4.25 STP-134.001, Shutdown Margin Verification.

4.26 STP-345.039, Reactor Trip P-4 Trip Actuating Device Operational Test.

4.27 V.C. Summer Precautions, Limitations, and Setpoints.

4.28 V.C. Summer Reactor Engineering Procedures.

4.29 V.C. Summer Tech Specs.

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GOP-3 ATTACHMENT I PAGE 1 OF 1 REVISION 13 SIGN-OFF IDENTIFICATION LIST PERSONNEL NAME (PRINTED) PERSONNEL NAME (SIGNATURE) PERSONNEL INITIALS

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SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION NUCLEAR OPERATIONS NUCLEAR OPERATIONS COPY NO.

GENERAL OPERATING PROCEDURE GOP-4A POWER OPERATION (MODE 1 - ASCENDING) REVISION 2 SAFETY RELATED

RECORD OF CHANGES CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE A P 10/31/11 F P 06/30/14 B P 04/25/12 G P 07/20/14 C P 11/01/12 H P 07/21/14 D P 05/01/14 E P 05/23/14 CONTINUOUS USE Continuous Use of Procedure Required. Read Each Step Prior to Performing.

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GOP-4A PAGE i REVISION 2 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE/SCOPE 1 2.0 INITIAL CONDITIONS 1 3.0 INSTRUCTIONS 5

4.0 REFERENCES

49 ENCLOSURES Enclosure A - Estimated Generator Capability Enclosure B - DA Low Power Temperature Curve ATTACHMENTS Attachment I - Sign-off Identification List Attachment II - Required System Alignment Verification

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GOP-4A REVISION 2 PAGE 1 OF 50 1.0 PURPOSE/SCOPE 1.1 The purpose of this procedure is to provide the steps required to be performed to startup the plant from the Point of Adding Heat to 100% Reactor Power.

1.2 10CFR50 Appendix B, SAP-630, and 10CFR50.59 apply to this procedure.

NOTE 2.0 through 4.0 a. If this procedure must be initiated under conditions other than those in Section 2.0, INITIAL CONDITIONS, the Shift Supervisor or Control Room Supervisor will review Sections 2.0, INITIAL CONDITIONS, and 3.0, INSTRUCTIONS. Steps that are not applicable due to plant conditions will be marked N/A and initialed by the Shift Supervisor or Control Room Supervisor. All other items will require sign-off or check-off. b. All personnel who sign off steps in this procedure must enter their names and initials on Attachment I. c. Each step should be initialed and dated when all its substeps are either completed and checked-off or marked as N/A and initialed. 2.0 INITIAL CONDITIONS INITIALS/DATE 2.1 RCS status is as follows: / a. System temperature is being maintained between 555°F and 559°F using the Steam Dump System or Steamline PORVs. b. System pressure is being maintained between 2220 psig and 2250 psig in AUTO control. c. All Reactor Coolant Pumps are in operation. d. Pressurizer level is being maintained at 25% in AUTO control. 2.2 All Safety Injection Systems are aligned and operable. /

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GOP-4A REVISION 2 PAGE 2 OF 50 INITIALS/DATE 2.3 Excore NIs are aligned for Power Operation per SOP-404, / Excore Nuclear Instrumentation System. 2.4 Reactor Power is being maintained between 1% and 3%. / 2.5 For Mode 2, with no untrippable or dropped Control Rods, / Shutdown Margin requirements are satisfied once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verification of Control Rods above the Rod Insertion Limit. 2.6 Reactor Makeup Control is in AUTO and set for the existing / RCS boron concentration. 2.7 The Rod Control and Position Indicating Systems are in operation / per SOP-403, Rod Control And Position Indicating System. 2.8 Secondary Plant status is as follows: / a. The Main Turbine is on the Turning Gear per SOP-215, Main Turbine Lube Oil Supply System. b. The Main Feedwater Pumps are on their Turning Gears per SOP-209, Feedwater Turbine Lube Oil System, or otherwise rotating via system flow.

c. Narrow Range Steam Generator levels are being maintained between 60% and 65% with chemistry within specification using the following: 1) Steam Generator Blowdown per SOP-212, Steam Generator Blowdown if desired, with Condensate return temperature maintained less than or equal to DA temperature. 2) Emergency Feedwater per SOP-211, Emergency Feedwater System. d. Main Steam heatup is complete per SOP-201, Main Steam System. e. Feedwater is being warmed per SOP-210, Feedwater System.

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GOP-4A REVISION 2 PAGE 3 OF 50 INITIALS/DATE f. Condensate is in operation per SOP-208, Condensate System. g. Circulating Water is in operation per SOP-207, Circulating Water. h. Condenser Vacuum is established per SOP-205, Turbine Sealing Steam System, and SOP-206, Main and Auxiliary Condenser Air Removal System in the following: 1) Main Condenser. 2) Auxiliary Condensers. 2.9 The following controller setpoints are aligned as follows: / a. LCV 3235, DEAER START UP DRAIN CNTRL AUTO with setpoint potentiometer set at 7.1

b. Feedwater Pumps: 1) PUMP A SPEED CONTROL AUTO with setpoint potentiometer set at 0.25. 2) PUMP B SPEED CONTROL AUTO with setpoint potentiometer set at 0.50. 3) PUMP C SPEED CONTROL AUTO with setpoint potentiometer set at 0.75. c. IFK3136, FLOW TO DEAERATOR AUTO with setpoint potentiometer set at 5.0. d. TURB OIL TEMP AUTO with setpoint potentiometer set at 2.0 - 2.66. e. EHC HYDRO OIL AUTO with setpoint potentiometer set at 2.4 - 7.1. f. H2 GAS TEMP AUTO. g. ALT COOLER TEMP AUTO. CHG D

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GOP-4A REVISION 2 PAGE 4 OF 50 INITIALS/DATE NOTE 2.9.h. IPV-2231, MS/PEGGING STM TO DEAERATOR, should be adjusted to maintain DA temperature between 130°F and 150°F. h. IPV-2231, MS/PEGGING STM TO DEAERATOR MAN or AUTO. 2.10 Reactor Engineering has verified the LEFM constants are / removed per SAP-119, Control Of The Station Calorimetric Computer Program. 2.11 Reactor Engineering has provided a Reactivity Management Plan / for the Turbine Startup and power ascension per SAP-0155, Reactivity Management. 2.12 A Pre-job brief has been conducted, including a review of / GOP-Appendix A and the Reactivity Management Plan. 2.13 IPCS is available to monitor Heat Up Rate during Startup when / Moderator Temperature Coefficient is near zero or positive. 2.14 Initiate a work order for Electrical Maintenance to perform / Thermography per steps 3.12.d.2 and 3.16.h. of this procedure.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 5 OF 50 3.0 INSTRUCTIONS INITIALS/DATE NOTE 3.1 through 3.11 Steps 3.1 through 3.11 raise Reactor Power from 1% to 25%. 3.1 Consult the following for current system status, Mode escalation / limitations, and Reactor Power level restrictions: a. Complete Attachment II for required system alignments. b. Removal and Restoration Log. c. Danger Tag Log. d. Reactor Engineering. e. Chemistry. f. Ensure completion of GTP-702 Attachment II.G, Category "C1" Operational Mode Change Plant Startup - Entering Mode 1. Z156 g. Initiate STP-120.003, Emergency Feedwater Valve Verification. (Free of air section 6.1 only.) 3.2 Perform OAP-100.4, Communication, Attachment I, Mode Change Brief Checklist. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 6 OF 50 INITIALS/DATE N013.3 Ensure the following disconnects are properly closed: / a. The associated disconnect switch mechanical operator is locked in position: 1) Manual Disconnect 8891. 2) Manual Disconnect 8893. 3) Manual Disconnect 8901. 4) Manual Disconnect 8903. b. Inform the Electrical Department that disconnects have been closed and initiate a work order to have Electrical perform thermography when disconnect is energized: 1) Manual Disconnect 8891. 2) Manual Disconnect 8893. 3) Manual Disconnect 8901. 4) Manual Disconnect 8903. CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 7 OF 50 INITIALS/DATE 3.4 Align Steam Dump control for Automatic operation as follows: / a. Transfer Steamline PORVs to Automatic operation as follows: 1) Place the Steamline PWR RELIEF A(B)(C) SETPT Controller(s) in MAN. 2) Adjust the PWR RELIEF SETPT Controllers to 8.4 (1092 psig). 3) Place the Steamline Power Relief Mode Switches in AUTO. 4) Place the PWR RELIEF SETPT Controllers in AUTO. b. Transfer STM DUMP CNTRL to Automatic operation as follows: 1) Place the STM DUMP CNTRL Controller in MAN. 2) Adjust the STM DUMP CNTRL setpoint to 8.4 (1092 psig). 3) Place the STM DUMP CNTRL Controller in AUTO. c. If necessary, reset C-7A and C-7B by taking STM DUMP MODE SELECT to RESET and return to STM PRESS.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 8 OF 50 INITIALS/DATE NOTE 3.5 Main Steamline GP-A and GP-B drains should be cycled open at least once every thirty minutes when left closed to minimize cooldown, in order to ensure moisture removal. 3.5 Complete Secondary Plant warm-up as follows: / Z137 a. Secure the Feedwater Recirculation flowpath per SOP-210, Feedwater System,Section III.B., Feedwater System Header Warming, Step 2.3. Z130 b. Perform SOP-204, Extraction Steam, Reheat Steam, Heater Vents and Drains,Section III.A, Start Up Of Extraction Steam, Reheat Steam, Heater Vents and Drains Step 2.1. Z149 c. Complete Main Turbine warm-up per SOP-214, Main Turbine And Controls,Section III.A, Turbine Startup, starting at Step 2.8.q. Z131 d. At the completion of the Main Turbine warm-up, continue with SOP-204, Extraction Steam, Reheat Steam, Heater Vents and Drains,Section III.A, Start Up Of Extraction Steam, Reheat Steam, Heater Vents and Drains, Step 2.2.

e. Maintain DA temperature between 130°F and 150°F as follows: 1) As required, adjust IPV-2231, MS/PEGGING STM TO DEAERATOR, in Automatic or Manual. 2) As required, LCV 3235, DEAER START UP DRAIN CNTRL, may be used to raise flow through the DA. f. Verify Feedwater and Condensate System chemistry is in specification per CP-615, Feedwater And Condensate Chemistry Control. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 9 OF 50 INITIALS/DATE 3.6 Prepare the Secondary Plant for power ascension as follows: / a. Ensure the Main Turbine is reset. Z158 b. Ensure the Main Generator Breaker Disconnect Switch 89 is Closed per SOP-303, Main Generator Breaker And Isophase Bus Duct Cooling,Section IV.B, Closing The Main Generator Breaker Disconnect Switch 89 c. Place 43-TS12, UNDER FREQ. TRIP CONTROL SW, in OFF. d. Direct Electricians to verify proper phase-phase voltages between the Main Generator Breaker and the low side of the Main Transformer as indicated by the G6 set of potential transformer voltage readings in XPN6222 on fuse block 2BU (FU3) (Reference 210-121, Sheet 3 ) are approximately 120 Vac and are balanced: 1) V2 - V4

2) V2 - V6

3) V4 - V6 e. Verify at least two Circulating Water Pumps are operating. f. Verify that at least one Condensate Pump is operating per SOP-208, Condensate System. g. Ensure Sparging Steam to the Deaerator is secured. Z139 h. Ensure all available Feedwater Booster Pumps are operating per SOP-210, Feedwater System, Section III.D, Feedwater Booster Pump Startup. i. Ensure Deaerator temperature is between 130°F and 150°F. CHG A CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 10 OF 50 INITIALS/DATE Step 3.6 continued j. Raise condensate flow to the Blowdown Heat Exchangers to between 150 gpm and 250 gpm per Steam Generator (450-750 gpm total) on FI-3061, CONDENSATE BLOWDOWN COOLERS FLOW IND, using the following controllers in MANUAL (AB-436):

1) ITV-3062A, BD COOLER A CDSTE OUT TEMP.

2) ITV-3062B, BD COOLER B CDSTE OUT TEMP.

3) ITV-3062C, BD COOLER C CDSTE OUT TEMP. 3.7 Align the Feedwater System for power ascension as follows: / C01 a. Perform PTP-102.005, Main Feedwater Pump Turbine Z128 Checks, quarterly portion Steps 6.1 through 6.12.

PMTS#_________

b. Ensure the following are MAN/CLOSED: 1) PVT-478, SG A FWF 2) PVT-488, SG B FWF 3) PVT-498, SG C FWF Z141 c. Start one Main Feedwater Pump per SOP-210, Feedwater System,Section III.E, Feedwater Pump Startup.

d. Reset the Feedwater Isolation signal by momentarily turning the following switches to the right: 1) FW ISOL TRAIN A RESET. 2) FW ISOL TRAIN B RESET. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 11 OF 50 INITIALS/DATE Step 3.7 continued CAUTION 3.7.e 1) Feedwater Header pressure should be maintained on program prior to opening Feedwater Isolation Valves to minimize water hammer. 2) Annunciator Point XCP-625 3-3 (FIV A/B/C ACCUM PRESS LO) should be verified clear or pressure locally verified greater than 500 psi prior to Mode 1 entry to ensure Feedwater Isolation Valve operability. (ref. Tech Spec 3.7.1.6) e. Open the following: 1) PVG-1611A, A ISOL. 2) PVG-1611B, B ISOL. 3) PVG-1611C, C ISOL. NOTE 3.7.f Use MANUAL control only if the Master Speed Controller is unable to control in AUTO. Z140 f. Ensure the MASTER SPEED CNTRL (MCB M/A station) is in Automatic per SOP-210, Feedwater System, Section III.E, Feedwater Pump Startup, Step 2.8. Z197 g. Prepare the Main Generator for startup per SOP-301, MAIN GENERATOR SYSTEM, SECTION III.A, Startup, Steps 2.1 and 2.2. h. Contact Reactor Engineering to verify LEFM constants are removed per SAP-119, Control Of The Station Calorimetric Computer Program. CHG D CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 12 OF 50 INITIALS/DATE CAUTION 3.8 Reactor Power must be maintained less than or equal to 10% until Emergency Feedwater is aligned per STP-120.003, Emergency Feedwater Valve Verification. (refer to Tech Spec 4.7.1.2.a.4) 3.8 Prepare for power ascension as follows: / a. Verify the accumulator pressure for each Feedwater Isolation Valve is greater than 500 psi as indicated by either of the following: 1) XCP-625 3-3 (FIV A/B/C ACCUM PRESS LO) is clear. 2) Accumulator pressure locally verified is greater than 500 psi for each valve. b. Commence Reactor Power increase to between 6% and 9% (Target 8% power, at a reasonably achievable ramp rate up to 1/2%/minute). c. Log the time and date the plant entered Mode 1: Mode 1 Entry: / Time Date GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 13 OF 50 INITIALS/DATE Step 3.8 continued NOTE 3.8.d Maintaining Main Feedwater Pump discharge pressure 50 psi -150 psi greater than Main Steam header pressure will maintain Steam Generator levels until Main Feedwater Pump speed control is placed in Automatic. d. If the MASTER SPEED CNTRL will NOT control in AUTO, perform the following: 1) Place the MASTER SPEED CNTRL in Manual. 2) Adjust Main Feedwater Pump speed as necessary to maintain Main Feedwater Pump discharge pressure 50 psi to 150 psi greater than Main Steam header pressure. Z142 e. Perform SOP-210, Feedwater System,Section III.F, Transferring Emergency Feed Flow To The Main Feed Reg Valves (Preferred method). f. Transfer Feedwater Flow from (Alternate method): Z143 1) Emergency Feed to the Bypass Valves per SOP-210, Feedwater System,Section IV.A, Transferring Feedwater Flow From The Emergency Feed To The Bypass Valves. Z144 2) The Bypass Valves to the Main Feed Reg Valves per SOP-210, Feedwater System,Section IV.B, Transferring Feedwater Flow From The Bypass Valves To The Main Feed Reg Valves. Z140 g. Establish automatic Feedwater Pump speed control per SOP-210, Feedwater System, Feedwater System, Section III.E, Feedwater Pump Startup, Step 2.8. C03 CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 14 OF 50 INITIALS/DATE Step 3.8 continued h. Update the IPCS Plant Mode indicator to indicate Mode 1 as follows: 1) Type the Turn-On-Code MODE to display the PLANT MODE CHANGE DISPLAY window. 2) Select the SET MODE 1 Pushbutton. 3) Verify POWER OPER is displayed on the left end of the top toolbar. CAUTION 3.8.i 1) Moisture Separator/Reheater temperature changes and Main Turbine vibration levels must be monitored closely while placing the MSRs in service. 2) To minimize stress in the Low Pressure Turbines, Hot Reheat Steam temperature changes must be limited to 125F/hr. Z134 i. Start up MSR A and B, in RAMP (TEMP CONTROL) mode, per SOP-204, Extraction Steam, Reheat Steam, Heater Vents And Drains,Section III.D, Normal Startup And Operation Of The MSRs. j. When less than 15% power, ensure the following valves are open: 1) XVT02072A-HD, REHEAT A 4TH-PASS DUMP TO CNDSR THROT. 2) XVT02072B-HD, REHEAT B 4TH-PASS DUMP TO CNDSR THROT. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 15 OF 50 INITIALS/DATE Step 3.8 continued k. Ensure the following as Reactor Power increases to no more than 9%: 1) DA temperature is being maintained between 130°F and 150F by adjusting IPV-2231, MS/PEGGING STM TO DEAERATOR, as necessary. 2) Narrow Range Steam Generator levels are maintained between 55% and 65%. 3) Condensate flow to the Deaerator increases. Z145 l. Secure Emergency Feedwater per SOP-211, Emergency Feedwater System,Section III.B, Motor Driven Emergency Feedwater Pump Shutdown. m. Complete STP-120.003, Emergency Feedwater Valve Verification (Valve Position Verification portion). STTS# ___________ NOTE 3.9 RCS TAVG - TREF DEV HI/LO (XCP-615 2-5) is expected to alarm as TAVG is increased and TREF remains constant. Compensatory actions should be taken per the ARP for this alarm. 3.9 When Emergency Feedwater is aligned for power operation, / prepare to synchronize and load the Main Generator as follows: a. Slowly raise Reactor Power to between 12% and 15% while continuing with this procedure. b. At 10% Reactor Power, perform the following: 1) Verify P10, NIS PR, permissive energizes to bright. 2) Verify P7, REACTOR TRIP BLOCKED, permissive de-energizes to dim. 3) Verify normal Power Range Channel indication. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 16 OF 50 INITIALS/DATE Step 3.9.b continued 4) Monitor the highest indicating Power Range Channel and Delta Flux on either of the following: a) NR-45, NIS RECORDER. b) Computer display NR45. 5) Ensure REGULATOR CORE 1 ALARM and REGULATOR CORE 2 ALARM (XCP-633) are reset.

c. Stabilize Reactor Power to establish and maintain the following conditions prior to and during the Main Turbine rollup to 1800 RPM: 1) Reactor Power between 12% and 15%. 2) Steam Dump Demand between 8% and 14% as indicated on TI-408, SD CNTRL S/G %. 3) Main Steam Header Pressure less than 1120 psig. d. If not completed previously, perform the following per SOP-210, Feedwater System: Z140 1) Establish automatic Feedwater Pump speed control, Section III.E, Feedwater Pump Startup, Step 2.8. Z142 2) Transfer Feedwater from the Main Feed Bypass Valves to the Main Feed Regulating Valves,Section IV.F, Transferring Feedwater Flow From The Bypass Valves To The Main Feed Reg Valves. Z135 e. Transfer Gland Sealing Steam to Main Steam per SOP-205, Turbine Sealing Steam System,Section III.A, Startup Of The Turbine Sealing Steam System Using Main Steam. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 17 OF 50 INITIALS/DATE Step 3.9 continued f. Momentarily place the following RESET-BLOCK Switches in BLOCK: 1) IR TRAIN A. 2) IR TRAIN B. 3) PR LOW SP TRAIN A. 4) PR LOW SP TRAIN B.

g. Verify the following status lights energize to bright: 1) IR A TRIP BLCK. 2) IR B TRIP BLCK. 3) PR A TRIP BLCK. 4) PR B TRIP BLCK. Z147 h. Roll the Main Turbine to 1800 RPM, per SOP-214, Main Turbine And Controls,Section III.A, Turbine Startup, Step 2.13. i. Ensure 0.5 scfh flow through FLOW METER FOR GAS ANALYZER (XPN-7201, HYDROGEN AND STATOR COOLING WTR CNT PNL) by adjusting XVT12205-HY, MACHINE GAS ANALYZER INLET ISOL VALVE (TB-412). j. Obtain a Switching Order from the System Controller. CHG D CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 18 OF 50 INITIALS/DATE CAUTION 3.10 a. Thermal Power changes of greater than 15% in any one hour period require completion of GTP-702 Attachment III.H. b. VCS DDS Report, POWER CHANGE SEARCH, should be periodically performed to ensure a thermal power change of greater than 15% in any one hour period is detected. c. Prolonged operation at low loads (less than 150 MWe) may result in Turbine rubs and elevated bearing vibration caused by low Exhaust Hood temperatures. d. To prevent equipment damage, Step 3.10 should be completed as conditions allow. This is especially true when a Turbine load increase is stopped prior to reaching 150 MWe. NOTE 3.10 through 3.18 a. IFK3136, FLOW TO DEAERATOR, AUTO setpoint should be adjusted during power changes to maintain LI-3136, DEAER STOR TK NR LVL, between 2.5 feet and 5.0 feet as LCV 3235, DEAR START UP DRAIN CNTRL, is closed. b. Acknowledging dialog boxes is considered a "skill of the craft". 3.10 Synchronize and load the Main Generator to as follows: / a. Adjust Reactor Power as necessary to maintain between 8% and 14% Steam Dump Demand as indicated on TI-408, SD CNTRL S/G %, while continuing with this procedure. CHG B CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 19 OF 50 INITIALS/DATE Step 3.10 continued NOTE 3.10.b When the Main Generator Breaker is closed the Generator icon will swap from speed (rpm's) indication to load (MW) indication. Z152 b. Synchronize and load the Main Generator to 50 MWe per SOP-301, Main Generator System,Section III.A, Startup, Step 2.3. c. Monitor Exhaust Hood temperature using any of the following: 1) On the EHC HMI select Monitor/LP Hoods 2) Computer display TURBRG. 3) Computer points T3058A, EXHAUST SPRAY HOOD A TEMP, and T3068A, EXHAUST SPRAY HOOD B TEMP. d. Raise Turbine load to 150 MWe as follows: 1) Verify Exhaust Hood temperature is less than 175°F as indicated on the EHC HMI, Monitor/LP Hoods screen. 2) Using the EHC HMI, Control/Load screen, on Load Set, select Ramp Rate and enter desired rate of 1% or less. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 20 OF 50 INITIALS/DATE Step 3.10 continued 3) Increase Turbine load by one of the following methods: a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 13.59%. (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 21 OF 50 INITIALS/DATE Step 3.10 continued CAUTION 3.10.e Do not stop the Turbine load increase with Exhaust Hood temperature less than 80°F as indicated on the EHC HMI, Monitor/LP Hoods. e. If necessary, stop the Turbine load increase by one of the following methods: 1) Depress the HOLD button. 2) Release the Raise Pushbutton on the MCB. f. Maintain Exhaust Hood temperature greater than 80°F as indicated on the EHC HMI, Monitor/LP Hoods screen, by Turbine load adjustments.

g. If necessary, re-commence the Turbine load increase by one of the following methods: 1) Using the EHC HMI, Control/Load screen, on Load Set, select Ramp Rate and enter desired rate of 1% or less.

2) Manually using the raise/lower pushbuttons: a) Select the desired ramp rate of 1%/min or less. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe). CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 22 OF 50 INITIALS/DATE Step 3.10 continued h. As Turbine load increases, perform the following: 1) Ensure 1st Stage Shell Inner heatup rate as indicated on EHC HMI, Aux/Metal Temps screen does not exceed 150°F/hr. 2) Maintain DA temperature as follows: a) Adjust IPV-2231, MS/PEGGING STM TO DEAERATOR, as necessary, to maintain DA temperature per Enclosure B, DA Low Power Temperature Curve. b) If required, LCV 3235, DEAER START UP DRAIN CNTRL, may be used to raise flow through the DA. Z138 c) Ensure Feedwater Booster Pump warm-up criteria are maintained with DA temperature changes per SOP-210, Feedwater System, Section III.D, Step 2.1. d) Maintain Blowdown Heat Exchanger condensate outlet temperatures at least 30F below DA temperature. i. Adjust Reactor Power as necessary to maintain between 2% and 14% Steam Dump Demand as indicated on TI-408, SD CNTRL S/G % while continuing with this procedure. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 23 OF 50 INITIALS/DATE Step 3.10 continued Z148 j. Close GP-A drain valves per SOP-214, Main Turbine And Controls,Section III.A, Turbine Startup, Step 2.24. k. Place 43-TS12, UNDER FREQ. TRIP CONTROL SW, in ON. l. Reset VOLT. UNBAL. RELAY 60G. Z132 m. Perform the 50 MWe Main Control Board Extraction Drain Switch alignment per SOP-204, Extraction Steam, Reheat Steam, Heater Vents and Drains,Section III.A, Startup Of Extraction Steam, Reheat Steam, Heater Vents And Drains, Step 2.3. n. Perform PTP-102.003, Main Generator Temperature Monitoring. PMTS# __________

o. When the Turbine Load is greater than 10% (100 MWe),

as indicated on any DCS graphic screen or EHC HMI, perform the following: 1) Open MVG-1212, EXTR STM TO DEAER ISOL. 2) Adjust the IPV-2231, MS/PEGGING STM TO DEAERATOR, setpoint to a setting of 7.0 in AUTO.

3) Ensure the following are closed (TB-412): a) XVG02075-HD, HP FW HEATER 2A DRAIN TO DEAER HDR ISOLATION. b) XVG02074-HD, HP FW HEATER 2B DRAIN TO DEAER HDR ISOLATION. 4) Place the following in NORMAL (GRAPHIC 101 and 102 or 110 screens - I icons): a) FW HTR 2A OPRTR SELECT ISOLATION. b) FW HTR 2B OPRTR SELECT ISOLATION. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 24 OF 50 INITIALS/DATE Step 3.10 continued p. Verify P13, 1st STAGE PRESSURE, permissive de-energizes to dim. q. When Turbine Load has stabilized at 150 MWe, perform the following: 1) Close the following Main Steam GP-B drain valves: a) MVG-2899A, X-AROUND DRN VLV-1. b) MVG-2899B, X-AROUND DRN VLV-2. c) MVG-2899C, X-AROUND DRN VLV-3. d) MVG-2899D, X-AROUND DRN VLV-4. Z133 2) Close heater startup vents and bypass valves for 15% Turbine Load, SOP-204, Extraction Steam, Reheat Steam, Heater Vents and Drains,Section III.A, Startup Of Extraction Steam, Reheat Steam, Heater Vents And Drains, Step 2.4.

3) Perform the following (TB-412): a) Throttle XVT02072A-HD, REHEAT A 4TH-PASS DUMP TO CNDSR THROT, to 2.0 turns open. b) Throttle XVT02072B-HD, REHEAT B 4TH-PASS DUMP TO CNDSR THROT, to 3.25 turns open.

4) If desired, secure the Auxiliary Boiler per SOP-506, Auxiliary Boiler Operation: Z153 a)Section III.B, Shutdown Z154 b)Section IV.E, Operation Of The Temporary Boiler. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 25 OF 50 INITIALS/DATE Step 3.10.q continued 5) Perform PTP-102.003, Main Generator Temperature Monitoring. PMTS# Z155 6) Align Extraction Steam to Auxiliary Building Steam per SOP-507, Auxiliary Steam System,Section III.A, Supplying Auxiliary Building Steam From Extraction Steam NOTE 3.10.q.7) If returning from a power level of greater than 75 %, per Reactor Engineering, the LEFM constants are not required to be adjusted (i.e. quarterly valve testing). 7) Contact Reactor Engineering to determine if LEFM constants need to be re-determined for current power history. 3.11 Raise Reactor Power to 25%, as follows:

a. Maintain DA temperature during load increases as follows: 1) Adjust IPV-2231, MS/PEGGING STM TO DEAERATOR, as necessary, to maintain DA temperature per Enclosure B, DA Low Power Temperature Curve. Z138 2) Ensure Feedwater Booster Pump warm-up criteria are maintained with DA temperature changes per SOP-210, Feedwater System,Section III.D, Feedwater Booster Pump Startup, Step 2.1. 3) Maintain Blowdown Heat Exchanger condensate outlet temperatures at least 30F below DA temperature. b. Ensure the A(B)(C) FPT SETPOINT RAMP LIMIT (A icon) RV value for operating Feedwater Pump A(B)(C) is set to 3000 rpm per minute. CHG D CHG D CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 26 OF 50 INITIALS/DATE Step 3.11 continued c. Verify Exhaust Hood temperature is less than 175°F as indicated on the EHC HMI, select Monitor/LP Hoods. d. Using the EHC HMI, Control/Load screen, on Load Set, select Ramp Rate and enter desired rate of 1% or less. e. Increase Turbine load to 300 MWe by one of the following methods: 1) Slowly Raise Turbine load automatically as follows (preferred method): (a) Select the Load pushbutton (a dialog box opens). (b) Enter 27.17%. (c) Confirm setpoint. (d) Verify proper plant response. 2) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2%

(20 MWe) (utilizes previously selected ramp rate) 3) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 27 OF 50 INITIALS/DATE Step 3.11.f continued f. As Turbine load increases, perform the following: 1) Ensure First Stage Shell heatup rate as indicated on EHC HMI, Aux/Metal Temps screen is maintained less than 150°F/hr. 2) Adjust Reactor Power as necessary to maintain between 2% and 14% Steam Dump Demand as indicated on TI-408, SD CNTRL S/G %.

3) When C5 (15% Turbine Load), 1st STAGE PRESSURE, permissive de-energizes to dim, perform the following in the order listed: a) Hold Reactor Power constant. NOTE 3.11.f.3)b) Establishing approximately a 0°F/hr to 30°F/hr cooldown rate will allow Tavg to slowly approach Tref.) b) Continue raising turbine load to match Tavg and Tref. c) Verify Tref is within 1°F of Tavg. d) Place the ROD CNTRL BANK SEL Switch in AUTO. e) Ensure the STM DUMP CNTRL auto setpoint is set to 8.4 (1092 psig). f) Transfer Steam Dump control to the Tavg mode as follows by placing the STM DUMP MODE SELECT Switch in TAVG.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 28 OF 50 INITIALS/DATE Step 3.11. continued g. When Turbine load has stabilized, perform the following: 1) Perform STP-102.002, NIS Power Range Heat Balance. STTS# C02 2) As a second check on Nuclear Instrumentation, compare RCS Loop T to the results of STP-102.002. 3) Perform PTP-102.003, Main Generator Temperature Monitoring. PMTS#

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 29 OF 50 INITIALS/DATE Step 3.11 continued h. Raise Turbine load to attain 25% Reactor Power as follows: 1) Select Ramp Rate and enter desired rate of 1% or less. 2) Raise Turbine Load by one of the following methods:

a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 18.2% (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 30 OF 50 INITIALS/DATE Step 3.11 continued i. At 25% Reactor Power, verify the following status lights energize to bright: 1) CHAN I IR FLUX HI. 2) CHAN II IR FLUX HI. 3) CHAN I PR FLUX LO SET PT. 4) CHAN II PR FLUX LO SET PT. 5) CHAN III PR FLUX LO SET PT. 6) CHAN IV PR FLUX LO SET PT. NOTE 3.12 and 3.13 Steps 3.12 and 3.13 raise Reactor Power from 25% to 48%. 3.12 Raise Reactor Power to 38% as follows: / a. Contact Chemistry to verify there are no 30% power Chemistry holds.

b. Ensure the following are closed (TB-412): 1) XVG02075-HD, HP FW HEATER 2A DRAIN TO DEAER HDR ISOLATION. 2) XVG02074-HD, HP FW HEATER 2B DRAIN TO DEAER HDR ISOLATION.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 31 OF 50 INITIALS/DATE Step 3.12 continued c. Raise Turbine load to attain 38% Reactor Power as follows: 1) Select Ramp Rate and enter desired rate of 1% or less. 2) Raise Turbine Load by one of the following methods:

a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 29.9% (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.)

d. At 250 MWe perform the following: 1) Ensure all Extraction Drain Valves are latched. 2) Contact Electrical Maintenance to perform thermography on manual disconnects 8901 and 8903. CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 32 OF 50 INITIALS/DATE Step 3.12 continued e. At 300 MWe, perform the following to start filling the drain lines from the 2A and 2B Heaters to the DA: 1) Open XVT12083-HD, 1" BYPASS VALVE FOR XVG-02075 (TB-412) (requires ladder). 2) Open XVT12085-HD, 1" BYPASS VALVE FOR XVG-02074 (TB-412). 3) Throttle XVT02018A-HD, FW HTR 2A DRN TO DEAER LVL CONT VLV BYP, ten turns off the closed seat (TB-463). 4) Throttle XVT02018B-HD, FW HTR 2B DRN TO DEAER LVL CONT VLV BYP, ten turns off the closed seat (TB-463). Z136 f. Place a second Condensate Pump in service per SOP-208, Condensate System,Section III.B, Condensate Pump Startup when total Condensate flow approaches 9000 gpm as indicated on the following:

1) FI 3026, PUMP A DISCH FLOW.

2) FI 3036, PUMP B DISCH FLOW.

3) FI 3046, PUMP C DISCH FLOW. g. Between 30% and 35% Reactor Power, perform the following: Z138 1) Ensure Feedwater Booster Pump warm-up criteria are maintained with DA temperature changes per SOP-210, Feedwater System, Section III.D, Feedwater Booster Pump Startup, Step 2.1. Z139 2) Ensure at least three Feedwater Booster Pumps are in service per SOP-210, Feedwater System, Section III.D, Feedwater Booster Pump Startup. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 33 OF 50 INITIALS/DATE Step 3.12 continued h. At 38% Reactor Power verify P8, REACTOR TRIP BLOCKED, de-energizes to dim. Z146 i. Establish automatic Steam Generator blowdown temperature control per SOP-212, Steam Generator Blowdown,Section III.A, Steam Generator Blowdown System Startup And Operation Steps 2.19 and 2.20. Z141 j. Between 35% and 48% Reactor Power, place a second Main Feedwater Pump in service per, SOP-210, Feedwater System,Section III.E, Feedwater Pump Startup. CHG D CHG E GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 34 OF 50 INITIALS/DATE Step 3.12 continued k. When 40% is indicated in the INTERMEDIATE PRESSURE block on the PLU test Screen perform the following: 1) On the EHC HMI, Control/Load Screen, select HOLD.

2) Perform the Power Load Unbalance (PLU) test as follows: (a) Verify P9, REACTOR TRIP BLOCKED, permissive is BRIGHT. (b) On the EHC HMI, select Tests/PLU test. (c) Verify all PLU Status indicate OFF. (d) Select PLU Test ON. (e) Select OK.

(f) Verify the following: (1) Test initiation on all 6 status indicators. (2) Final status indication for all 6 indicators PLU TEST FOR R/S/T COMPLETE. (g) Select PLU Test OFF (h) Select OK. (i) Select desired Ramp Rate, %/min increase on Control/Load screen. l. Determine the GOP Appendix A recommended power ascension rate. CHG G CHG H GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 35 OF 50 INITIALS/DATE 3.13 Raise Reactor Power to 48% at the GOP Appendix A / recommended power ascension rate, as follows: a. Verify Steam Generator chemistry is in specification per CP-613, Steam Generator Chemistry Control. b. Raise Turbine load to attain 48% Reactor Power as follows: 1) Select Ramp Rate and enter the recommended Load Ramp Rate.

2) Raise Turbine Load by one of the following methods: a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 39.5.%. (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 36 OF 50 INITIALS/DATE Step 3.13 continued c. When Turbine Load is greater than 40% (385 MWe), verify C20, 1st STG PRESS, de-energizes to dim. d. Monitor the following for proper operation: 1) Stator Water Cooling. 2) Hydrogen Seal Oil. e. Between 400 MWe and 450 MWe, open the following valves to align the 2A and 2B Heaters to the DA (TB-412): 1) XVG02075-HD, HP FW HEATER 2A DRAIN TO DEAER HDR ISOLATION. 2) XVG02074-HD, HP FW HEATER 2B DRAIN TO DEAER HDR ISOLATION. Z129 f. Secure the Condensate Polishing per SOP-203, Condensate Polishing System,Section III.F, Removing The Condensate Polishing System From Service. Z157 g. Maintain the following SOP-401, Reactor Protection And Control System,Section III.B, Load Variations With Manual Or Automatic Reactor Control parameters using Z017 SOP-106, Reactor Makeup Water System Section III.E, Z003 Alternate Dilute Operations or Section III.F, Borate Operations: 1) I within limits. 2) Control Rods above the Rod Insertion Limit. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 37 OF 50 INITIALS/DATE Step 3.13 continued NOTE 3.13.h Above 40% Turbine load MSR 4th pass drains to the Condenser close and the MSR 4th pass drains to the #1 Feedwater Heaters open. h. Using the DCS Computer Graphic Screens 101 and 102, verify MSR 4th pass drain valves have repositioned: 1) XVT-2071A indicates closed. 2) XVT-2071B indicates closed. 3) XVT-02068A indicates open. 4) XVT-02068B indicates open. i. When Reactor Power is stable at or below 48%,

perform the following: 1) STP-102.002, NIS Power Range Heat Balance. STTS# C02 2) As a second check on Nuclear Instrumentation, compare RCS Loop T to the results of STP-102.002.

3) Determine the operability of the Axial Flux Difference alarm: a) Perform STP-133.001, Axial Flux Difference Calculation. STTS# b) Verify Annunciator Point XCP-620 2-4 (CMPTR FLUX LMT EXCEEDS) is clear. 4) PTP-102.003, Main Generator Temperature Monitoring. PMTS# CHG C GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 38 OF 50 NOTE 3.14 through 3.16 Steps 3.14 through 3.16 raise Reactor Power from 48% to 90%. 3.14 At the GOP Appendix A recommended power ascension rate, / continue the Reactor Power ascension above 50% as follows: a. Ensure I, Axial Flux Difference, is within limits per V.C.Summer Curve Book, Figure I-4.1 prior to increasing Reactor Power above 50% per Tech Spec 3.2.1. b. When greater than 50% Rated Thermal Power perform STP-108.001, Quadrant Power Tilt Ratio. STTS#

c. Raise Turbine load to attain 90% Reactor Power as follows: 1) Select Ramp Rate and enter the recommended. Load Ramp Rate.

2) Raise Turbine Load by one of the following methods: a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 81.52% (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG B CHG C GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 39 OF 50 INITIALS/DATE Step 3.14 continued d. As power increases, verify the following annunciators clear: 1) PR UP DET FLUX HI DEV AUTO DEFEAT (XCP-620 1-5). 2) PR LOW DET FLUX HI DEV AUTO DEFEAT (XCP-620 1-6). e. Adjust Megavars using GEN FIELD VOLT ADJ as requested by the System Controller and within the Estimated Generator Capability Curve (Enclosure A). Z157 f. Maintain the following SOP-401, Reactor Protection And Control System,Section III.B, Load Variations With Manual Or Automatic Reactor Control parameters using Z017 SOP-106, Reactor Makeup Water System Section III.E, Z003 Alternate Dilute Operations or Section III.F, Borate Operations: 1) I within limits. 2) Control Rods above the Rod Insertion Limit. g. Verify P9, REACTOR TRIP BLOCKED, permissive de-energizes to dim. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 40 OF 50 INITIALS/DATE 3.15 When Turbine Load is greater than 50%, as indicated on any / DCS graphic screen, place IZY03783A and IZY03793A, MANUAL OPERATOR STATIONs in Automatic as follows (GRAPHIC 101 and 102 or 110 screens): a. Ensure the following are open (TB-412): 1) XVG02075-HD, HP FW HEATER 2A DRAIN TO DEAER HDR ISOLATION. 2) XVG02074-HD, HP FW HEATER 2B DRAIN TO DEAER HDR ISOLATION. b. Ensure HTR 2A&B DRN VLV TO DEAERATOR CLOSE is selected to CNTRL (C icon). c. Ensure both FW HTR 2A OPRTR SELECT ISOLATION and FW HTR 2B OPRTR SELECT ISOLATION are in NORMAL (I icon). NOTE 3.15.d The #2 Heaters utilize the lowest controller signal to maintain level, regardless of the M/A station output (in Manual or Automatic). As a result, manipulation of the M/A station in manual may not yield the anticipated result. d. Slowly raise OUT on both ILY03783A and ILY03793A MANUAL OPERATOR STATIONs (M/A - icons) until the associated IZY03783B and IZY03793B (2A/2B Heater drain to the Condenser) are fully closed. e. When FEEDWATER HEATER 2A(B) levels are stable and slightly below setpoint, place ILY03783A and ILY03793A MANUAL OPERATOR STATIONs (M/A - icons) in Automatic. f. Adjust IFK3136, FLOW TO DEAERATOR, AUTO setpoint as necessary to ensure LI-3136, DEAER STOR TK NR LVL is between 2.5 feet and 5.0 feet with LCV 3235, DEAR START UP DRAIN CNTRL closed.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 41 OF 50 INITIALS/DATE Step3.15 continued g. Close the following bypass valves: 1) XVT12083-HD, 1" BYPASS VALVE FOR XVG-02075 (TB-412) (requires ladder). 2) XVT12085-HD, 1" BYPASS VALVE FOR XVG-02074 (TB-412). 3) XVT02018A-HD, FW HTR 2A DRN TO DEAER LVL CONT VLV BYP (TB-463). 4) XVT02018B-HD, FW HTR 2B DRN TO DEAER LVL CONT VLV BYP (TB-463). 3.16 As the Reactor Power ascension to 90% continues perform / the following:

a. Between 60% and 65% Reactor Power, perform the following: Z138 1) Ensure Feedwater Booster Pump warm-up criteria are maintained with DA temperature changes per SOP-210, Feedwater System,Section III.D Feedwater Booster Pump Startup, Step 2.1. Z139 2) Ensure four Feedwater Booster Pumps are in service per SOP-210, Feedwater System, Section III.D, Feedwater Booster Pump Startup. Z141 3) Start a third Main Feedwater Pump per SOP-210. Feedwater System,Section III.E, Feedwater Pump Startup. b. At 65% Reactor Power, perform PTP-102.003, Main Generator Temperature Monitoring. PMTS# CHG D CHG D CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 42 OF 50 INITIALS/DATE Step 3.16 continued c. When Reactor Power is between 60% and 75%, perform PTP-102.001, Main Turbine Tests (Power Operated Extraction System Check Valve portion only). PMTS# d. At 75% Reactor Power, perform STP-102.002, NIS Power Range Heat Balance. STTS# e. At 80% Reactor Power, align Control Valve drain valves as follows:

1) Ensure PVG-2898B, DV-4, is open as follows: a) Verify Control Valve #4 is closed. b) Verify PVG-2898B, DV-4, is open. c) If both PVG-2898B, DV-4, and Control Valve #4 are closed, open PVG-2898B, DV-4, by opening MVG-2898D, STM LEAD DRN FOR CV-1. 2) Open MVG-2897, COMB CNTRL VLV BSD. f. When Control Valve #4 indicates greater than 5% open, perform the following: 1) Ensure PVG-2898B, DV-4, is CLOSED. 2) Close MVG-2897, COMB CNTRL VLV BSD. g. At 85% Reactor Power, perform PTP-102.003, Main Generator Temperature Monitoring. PMTS# ___________ CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 43 OF 50 INITIALS/DATE Step 3.16 continued h. Contact Electrical Maintenance to perform thermography on manual disconnects 8901 and 8903. i. Adjust Megavars using GEN FIELD VOLT ADJ as requested by the System Controller and within the Estimated Generator Capability Curve (Enclosure A). j. If desired stabilize Reactor Power at 90%, otherwise proceed to Step 3.17. NOTE 3.17 and 3.18 Steps 3.17 and 3.18 raise Reactor Power from 90% to 100%. 3.17 At the GOP Appendix A recommended power ascension rate, / increase Reactor Power from 90% to 95% as follows: a. Raise Turbine load to attain 95% Reactor Power. 1) Select Ramp Rate and enter the recommended. Load Ramp Rate.

2) Raise Turbine Load by one of the following methods:

a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 87.41%. (3) Confirm setpoint. (4) Verify proper plant response. CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 44 OF 50 INITIALS/DATE Step 3.17.a.2) continued b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.)

b. Place the following Blowdown Temperature Controllers in AUTOMATIC (AB-436): 1) ITV-3062A, BD COOLER A CDSTE OUT TEMP. 2) ITV-3062B, BD COOLER B CDSTE OUT TEMP. 3) ITV-3062C, BD COOLER C CDSTE OUT TEMP. c. Adjust Megavars using GEN FIELD VOLT ADJ as requested by the System Controller and within the Estimated Generator Capability Curve (Enclosure A). Z157 d. Maintain the following SOP-401, Reactor Protection And Control System,Section III.B, Load Variations With Manual Or Automatic Reactor Control parameters using Z017 SOP-106, Reactor Makeup Water System Section III.E, Z003 Alternate Dilute Operations or Section III.F, Borate Operations: 1) I within limits. 2) Control Rods above the Rod Insertion Limit. CHG B CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 45 OF 50 INITIALS/DATE Step 3.17 continued e. Stabilize Reactor Power at 95% and perform the following: 1) STP-102.002, NIS Power Range Heat Balance. STTS# ___________ 2) During the first power ascension following refueling, contact Reactor Engineering to continue the power ascension per REP-107.001, Controlling Procedure For Refueling Startup And Power Ascension Testing. 3.18 Slowly increase Reactor Power to 100% as follows: /

a. If the IPCS is available, verify the NSSS CRT is displaying the following computer points: 1) SHIFT AVG POWER (U9002). 2) QCORE 1 (C1M) (U9003). b. Raise Turbine load to attain 100% Reactor Power per the GOP Appendix A recommended power ascension rate, while continuing with this procedure. 1) Select Ramp Rate and enter the recommended. Load Ramp Rate.

2) Raise Turbine Load by one of the following methods: a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 92.12%. (3) Confirm setpoint. (4) Verify proper plant response. CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 46 OF 50 INITIALS/DATE Step 3.18.b.2) continued b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.)

c. Adjust Turbine load to attain 100% Reactor Power, while continuing with this procedure. 1) Using the EHC HMI, Control/Load screen, on Load Set, select Ramp Rate and enter desired rate of 1% or less.

2) Adjust Turbine Load as follows: a) Select the Load pushbutton (a dialog box opens). b) Adjust the setpoint in incremental values not to exceed 0.2%. c) Confirm setpoint. d) Verify proper plant response. d. Adjust Megavars using GEN FIELD VOLT ADJ as requested by the System Controller and within the Estimated Generator Capability Curve (Enclosure A). CHG B CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 47 OF 50 INITIALS/DATE Step 3.18 continued e. Monitor the following for proper operation: Z099 1) Stator Water per SOP-218, Stator Cooling Water System,Section III.A Z150 2) Hydrogen Seal Oil per SOP-216, Seal Oil System, Section III.A. Z151 3) Generator Gas per SOP-217, Generator Gas And Vent System,Section III.A f. Stabilize at 100% Reactor Power and perform STP-102.002, NIS Power Range Heat Balance. STTS# ___________ g. If desired, place the Load Limit circuit in service as follows: 1. Select desired Ramp Rate on Load Limit. (usually Normal). 2. Select Setpoint on Load Limit, (a dialog box opens). 3. Enter the desired setpoint (must be less than the indicated Load Reference). 4. Confirm setpoint. 5. Verify the Load Limit status indicates LIMITING. 6. Verify proper system response. CHG B CHG B CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 48 OF 50 INITIALS/DATE Step 3.18 continued NOTE 3.18.h f returning from a power level of greater than 75 %, per Reactor Engineering, the LEFM constants are not required to be adjusted (i.e. quarterly valve testing). h. Contact Reactor Engineering to determine if LEFM constants need to be re-determined for current power history. i. Adjust Reactor Power to 100% Rated Thermal Power, and perform the following: 1) STP-102.002, NIS Power Range Heat Balance. STTS# ___________ 2) PTP-102.003, Main Generator Temperature Monitoring. PMTS# ___________ j. Maintain operation as close to 100% of licensed core power (2900 MWt) as possible, per OAP-100.6, Control Room Conduct and Control of Shift Activities. k. Notify Reactor Engineering to evaluate the requirements for performing STP-201.001, Monthly Reactor Engineering Surveillances. NOTE 3.18.l For purposes of record, this procedure is complete when all steps through 3.18.l are initialed and dated. It should then be routed to the Operations Supervisor. l. 100% Reactor Power achieved: 1) Date / / 2) Time CHG C

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GOP-4A REVISION 2 PAGE 49 OF 50

4.0 REFERENCES

4.1 CP-613, Steam Generator Chemistry Control.

4.2 CP-615, Feedwater And Condensate Chemistry Control.

4.3 ES560.120, Feedwater Flow Rate And Temperature Normalization Surveillance.

4.4 FSAR Section 5.0.

4.5 GOP-Appendix A. C03 4.6 LER 97002. N01 4.7 MRB 9501. 4.8 OAP-100.4, Communication.

4.9 PTP-102.002, Main Turbine Monthly Oil System Test.

4.10 PTP-102.003, Main Generator Temperature Monitoring.

4.11 PTP-102.005, Main Feedwater Pump Turbine Checks.

4.12 PTP-102.008, Main Turbine Overspeed Testing.

4.13 SAP-119, Control Of The Station Calorimetric Computer Program. C01 4.14 SER 880024.

C02 4.15 SOER 90-3. 4.16 SOP-102, Chemical And Volume Control System.

4.17 SOP-106, Reactor Makeup Water System.

4.18 SOP-201, Main Steam System.

4.19 SOP-203, Condensate Polishing System.

4.20 SOP-204, Extraction Steam, Reheat Steam, Heater Vents And Drains.

4.21 SOP-205, Turbine Sealing Steam System.

4.22 SOP-206, Main and Auxiliary Condenser Air Removal System.

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GOP-4A REVISION 2 PAGE 50 OF 50 4.23 SOP-207, Circulating Water. 4.24 SOP-208, Condensate System. 4.25 SOP-209, Feedwater Turbine Lube Oil System. 4.26 SOP-210, Feedwater System. 4.27 SOP-211, Emergency Feedwater System. 4.28 SOP-212, Steam Generator Blowdown. 4.29 SOP-214, Main Turbine And Controls. 4.30 SOP-215, Main Turbine Lube Oil Supply System. 4.31 SOP-216, Seal Oil System. 4.32 SOP-217, Generator Gas And Vent System. 4.33 SOP-218, Stator Cooling Water System. 4.34 SOP-301, Main Generator System. 4.35 SOP-403, Rod Control And Position Indicating System. 4.36 SOP-404, Excore Nuclear Instrumentation System. 4.37 SOP-506, Auxiliary Boiler Operation. 4.38 SOP-507, Auxiliary Steam System. 4.39 STP-102.002, Nis Power Range Heat Balance. 4.40 STP-108.001, Quadrant Power Tilt Ratio.

4.41 STP-120.003, Emergency Feedwater Valve Verification. 4.42 STP-133.001, Axial Flux Difference Calculation. 4.43 STP-134.001, Shutdown Margin Verification. 4.44 STP-201.001, Monthly Reactor Engineering Surveillances. 4.45 V.C. Summer Precautions, Limitations, and Setpoints.

4.46 V.C. Summer Tech Specs.

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GOP-4A ENCLOSURE A PAGE 1 OF 1 REVISION 2 ESTIMATED GENERATOR CAPABILITY -1000100200300400 500010020030040050060070080090010001100MEGAWATTSMEGAVARSBBAACCCurve AB Limited By Field HeatingCurve BC Limited By Armature HeatingLAGLEAD45 PSIG60 PSIGMVAR LIMIT+325ADMIN LIMIT ONLYCONTACT SYSTEM CONTROLLER IF EXCEEDED 484.95 P.F..90 P.F.UEL SETPOINTAT 20.9 KV UEL SETPOINT AT 22.0 KV

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GOP-4A ENCLOSURE B PAGE 1 OF 1 REVISION 2 DA Low Power Temperature Curve100120140160180200220240020406080100120140MegaWatts ElectricDA Temperature, Deg F150130MINIMUM TEMPERATURE FOR PRESSURIZATIONDesired Operating BandCAUTIONDA operation in the regionabove this curve may result inwaterhammer.

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GOP-4A ATTACHMENT I PAGE 1 OF 1 REVISION 2 SIGN-OFF IDENTIFICATION LIST PERSONNEL NAME (PRINTED) PERSONNEL NAME (SIGNATURE) PERSONNEL INITIALS

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GOP-4A ATTACHMENT II PAGE 1 OF 1 REVISION 2 REQUIRED SYSTEM ALIGNMENT VERIFICATION PROCEDURE NUMBER PROCEDURE TITLE Date of Last Alignment Has the System been 00S > 14 days Has the System undergone Significant Maintenance Does the System Require a Complete new Alignment Date of Record for this Procedure Verification by Shift Supervisor Yes No Yes No Yes No Initials/Date GOP-2 Required Systems Alignments Current and Completed NA NA NA NA NA NA NA / LIST OTHERS REQUIRED / / / / /

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Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 2 NUREG-1021 R9 S1 Facility: VC SUMMER Scenario No: 2 Op Test No: NRC-ILO-13-01 Examiners: Operators: CRS: RO: BOP: Initial Conditions:

• 60% Power MOL.

• B1 Train Work Week.

• The National Weather Service has declared a Severe Weather Warning for Richland, Fairfield, and Kershaw counties for the next four (4) hours.

• Alternate Seal Injection is OOS.

• is not running.

• Investigating small Turnover:

• MFW Pump for operation. Critical Tasks:

• Maintain SG Level without causing a Reactor/Turbine trip.

• Close "A" or "B" MSIV Prior to Orange path on Integrity or Containment.

• Isolate EFW to the faulted SG prior to Orange path reached on Integrity. Event No. Malf No. Event Type* Event Description 1 CVC008 C-RO, CRS TS-CRS Isolable Letdown Line Leak Inside Reactor Building - 50 gpm. 2 MS005O I-BOP, CRS TS-CRS FT-494 ( Steam Flow Transmitter) fails LOW. 3 NA N-BOP, CRS R-RO Rapid downpower due to B MFW pump vibration. 4 CRF007H14 C-RO, CRS TS-CRS Rod H14 stuck but trippable (blown fuse). 5 FW017O I-BOP, CRS PT-508 (MFW Pump Discharge Header Pressure) Fails LOW. (Manual control of MFW Pp speed)

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 2 NUREG-1021 R9 S1 6 MSS003A AUX009AA AUX009AB AUX009AC M-ALL Main Steamline Break inside the RB due to a seismic event. EPS013 Main Generator and Voltage Regulator Breakers Fail to Trip. MSS006A MSS006B * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 2 NUREG-1021 R9 S1 The following notation is used in the ES-D- IOA designates Immediate Operator Action steps.

• designates Continuous Action steps. The crew will assume the watch having been pre-briefed on the Initial Conditions, the plan for this shift and any related operating procedures. GOP-4A, Power ascension was halted at 60% due to a mechanical problem Pump. The current power level has been maintained for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Xenon is stable. The crew MFW Pump for operation. EVENT 1: Isolable Letdown Line Leak Inside Reactor Building - 50 gpm. TRIGGER 1 o MAL-CVC008 LETDOWN LINE LEAK INSIDE CONTAINMENT FINAL=30% On cue from the Examiner, a 50 gpm leak will be inserted on the letdown line inside the Reactor Building. The crew will identify that a leak exists and implement AOP-101.1, Loss of Reactor Coolant Not Requiring SI. The RO will isolate the leak by isolating letdown. The RO will then place excess letdown in service. The CRS will refer to Technical Specification 3.4.6.2, Operational Leakage. EVENT 2: FT- TRIGGER 2 o XMT-MS005O IFT00494 SG C STEAM FLOW FAIL TO POSN FINAL=0 On cue from the Examiner, FT-Flow transmitter for SG level control. FCV-498 will stroke closed causing SG level to lower. The crew will implement AOP-401.3 Steam Flow - Feedwater Flow Protection Channel Failure and select the operable Steam Flow Transmitter for control. This is a Technical Specification transmitter. The CRS will refer to Technical Specification 3.3.1, Reactor Trip System Instrumentation, Action 6 and 3.3.2, Engineered Safety Feature Actuation System Instrumentation, Action 24.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 2 NUREG-1021 R9 S1 EVENT 3: Rapid downpower due to B MFW pump vibration. The Shift Supervisor will call the CRS and report that the B Main Feedwater Pump vibration is in alert. He will instruct to crew to lower power to less than 45% at 3% per minute in accordance with GOP-4C Rapid Power Reduction and leave the pump running for engineering to evaluate. The RO will utilize boration and/or rod control to lower power while coordinating the downpower with the BOP who will be controlling turbine demand. EVENT 4: Rod H14 stuck but trippable (blown fuse). AUTO-TRIGGER 3 FNISPR(2) < 56 (N-42 indicates < 56% Power) o MAL-CRF007H14 STUCK ROD H-14 FAIL TO: TRIPPABLE TRIGGER 4 o MAL-CRF007H14 STUCK ROD H-14 FAIL TO: TRIPPABLE DELETE=00:00:01 Removes failure to allow rod recovery This event will occur when power is reduced to less than 56% or earlier if directed by the Examiner. Control Rod H-14 in Control Bank D will stop moving. This event must be inserted early enough in the downpower so that the failure will be apparent as power is lowered. The RO will realign the control rods in accordance with AOP-403.5 Stuck or Misaligned Control Rod. The CRS will refer to Technical Specification 3.1.3.1, Movable Control Assemblies. EVENT 5: PT-508 (MFW Pump Discharge Header Pressure) Fails LOW. (Manual control of MFW Pp speed). TRIGGER 5 o XMT-FW017O IPT00508 FW PP DSCHG HDR PRESS PI-508 FAIL TO POSN FINAL=200 On cue from the Examiner, a Main Feedwater Header Pressure transmitter will fail LOW causing the MFW Pump speed to increase and raise SG level. The operators will respond to annunciators and implement AOP-210.3, Feedwater Pump Malfunction. The BOP will take manual control of the master Speed control and adjust speed to maintain Feedwater Pump discharge pressure 150 to 250 psi greater than Main Steam Header pressure and restore SG levels.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 2 NUREG-1021 R9 S1 eamline Break inside the RB due to a seismic event. TRIGGER 6 o MAL-AUX009AB SEISMIC EVENT EARTHQUAKE 2/3 O.B.E.(UP/DOWN VERTICAL) Final Value = 2.1 Delay = 0 o MAL-AUX009AA SEISMIC EVENT EARTHQUAKE 2/3 O.B.E.(NORTH/SOUTH HORIZONTAL) Final Value = 3.2 Delay = 1 sec o MAL-AUX009AC SEISMIC EVENT EARTHQUAKE 2/3 O.B.E.(EAST/WEST HORIZONTAL) Final Value = 4.2 Delay = 1 sec o MAL-MSS003A STEAMLINE S/G A BREAK INSIDE CONTAINMENT Final Value = 3E+6 lbm/hr Delay = 10 sec o MAL-EPS013 GENERATOR BREAKER FAILS TO TRIP o MAL-MSS006A MAIN STEAM ISOLATION VALVE (S/G A) FAILURE Fail to: FAILURE TO CLOSE o MAL-MSS006B MAIN STEAM ISOLATION VALVE (S/G B) FAILURE Fail to: FAILURE TO CLOSE AUTO-TRIGGER 7 o MAL-MSS006A MAIN STEAM ISOLATION VALVE (S/G A) FAILURE Delete = 00:00:01 AUTO-TRIGGER 8 o MAL-MSS006B MAIN STEAM ISOLATION VALVE (S/G B) FAILURE Delete = 00:00:01 On cue from the Examiner, seismic monitors will indicate a seismic event has occurred. Ten (10) seconds later a steamline break inside the Reactor Building will be inserted. The Reactor will trip and the crew will implement EOP-1.0 (E-0) Reactor Trip/Safety Injection Actuation. The crew will identify that at least one Steam generator is faulted and transition to EOP-3.0 (E-2), Faulted Steam Generator Isolation. When the faulted SG is isolated the crew will transition to EOP-1.2 (ES-1.1), Safety Injection Termination.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 2 NUREG-1021 R9 S1 The malfunction will be apparent after the Reactor Trip. The crew will identify that the Main Generator Output Breaker failed to automatically trip and the BOP will manually open the breaker from the control board. The BOP must identify that the MSIVs are open and manually close them from the control board to isolate the faulted Steam Generator and prevent over-pressurization of the Reactor Building. TERMINATION: The scenario can be terminated after the crew has transitioned to EOP-1.2 (ES-1.1), Safety Injection Termination, and terminated Safety Injection or at the discretion of the Examiner. Scenario Attributes Events Total Malfunctions (5-8) 7

• Letdown Line Leak Inside Reactor Building - 50 gpm.

• FT-Flow Transmitter) fails LOW

• Rod H14 stuck but it can be tripped (blown fuse).

• PT-508 (MFW Discharge Header Pressure) Fails LOW

• MSLB inside Reactor Building

• Main Gen and Voltage Regulator Breakers Fail to Trip

• AUTO. Malfunctions after EOP entry (1-2) 2

• Main Generator and Voltage Regulator breakers fail to trip.

• A and B MSIVs fail to close in auto. Abnormal Events (2-4) 4

• Letdown Line Leak Inside Reactor Building - 50 gpm.

• FT- fails LOW

• Rod H14 stuck but it can be tripped (blown fuse).

• PT-508 (MFW Discharge Header Pressure) Fails LOW Major Transient (1-2) 1

• Faulted Steam Generator (MSLB inside RB) EOPs Entered (1-2) 2

• EOP-3.0 (E-2), Faulted Steam Generator Isolation

• EOP-1.2 (ES-1.1), Safety Injection Termination EOP Contingencies (0-2) 0 Critical Tasks (2-3) 3

• Maintain SG Level without a Reactor/Turbine trip.

• Close "A" or "B" MSIV Prior to Orange path on Integrity or Containment.

• Isolate EFW to the faulted SG prior to Orange path reached on Integrity.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 2 NUREG-1021 R9 S1 SIMULATOR SCENARIO SETUP INITIAL CONDITIONS: IC Set 291 60% Power Rod Position = 173 FCV-113 Pot Setting = 4.81 Boron = 1122 ppm Xe = - 2815 pcm Burnup = 10,000 MWD/MTU Prior to the scenario, crew should pre-brief on conditions and expectations for the Shift (maintain power, repairs estimated to be complete well before LCO action time expires.) PRE-EXERCISE: Ensure simulator has been checked for hardware problems (DORT, burnt out light bulbs, switch malfunctions, chart recorders, etc.) VCS-TQP-0807 Attachment I-A, Unit 1 Booth Instructor Checklist, has been completed. Hang Tags for equipment out of service. o Hang Caution Tag on HCV-186 due to ASI being OOS Mark up procedures in use with Sapplicable. o GOP-4A, POWER OPERATION (MODE 1 - ASCENDING) marked to step 3.16 Prepare a turnover sheet for each position. Conduct two-minute drill. The simulator may be left running at turnover (stable plant conditions). Ensure SIPCS rod position is matched to DRPI indication. PRELOAD: STANDARD SIMULATOR SETUP: PMP-LD003P, LEAK DETECTION SUMP PMP LOSS OF POWER VLV-FW028W, FW HDR RECIRC ISOL VLV LOSS OF POWER VLV-FW029W, FW HTR RECIRC ISO VLV LOSS OF POWER VLV-CS052W, RCP A SEAL LEAKOFF VLV LOSS OF POWER VLV-CS053W, RCP B SEAL LEAKOFF VLV LOSS OF POWER VLV-CS054W, RCP C SEAL LEAKOFF VLV LOSS OF POWER Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 2 NUREG-1021 R9 S1 SCENARIO RELATED ALTERNATE SEAL INJECTION OUT-OF-SERVICE ANN-CS044, ALT SEAL INJ PUMP TRBL Fail to: ON ANN-CS046, ALT SEAL D/G TRBL Fail to: ON MAL-CVS027, ASI D/G FAIL TO START MAL-CVC029, ASI PUMP FAIL TO START Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 1 Page: 9 of 51 Event

Description:

Isolable Letdown Line Leak Inside Reactor Building - 50 gpm. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOOTH OPERATOR: When directed, initiate Event 1 (TRIGGER 1). EVALUATOR NOTE: On cue from the Examiner, a 50 gpm leak will be inserted on the letdown line inside the Reactor Building. The crew will identify that a leak exists and implement AOP-101.1, Loss of Reactor Coolant Not Requiring SI. The RO will isolate the leak by isolating letdown. The RO will then place excess letdown in service. Indications Available: Charging increasing with no change in Letdown Changing RB environmental conditions XCP-606/607-2-2, RBCU Drain Flow alarms XCP-614-5-1, CHG LINE FLO HI/LO XCP-615-3-6, RCS Leak Calculation alarm RM-A2 HI RAD EVALUATOR NOTE: If the BOP responds to the HVAC alarms the BOOTH OPERATOR will ENSURE that an INSTRUCTOR notifies the BOP that the Control Building Operator will handle all future HVAC alarms. BOOTH OPERATOR: If necessary direct an Instructor to relieve the BOP as the Control Building Operator. Inform the BOP that you will handle all future HVAC alarms. When the HVAC panel annunciates acknowledge the alarm. Report as the Control Building Operator (Unit 4) High Temperature alarm in REACTOR COMPARTMENT COOLING SYSTEM. EVALUATOR NOTE: If Primary containment average air temperature exceeds 120°F then Technical Specification 3.6.1.5 Action applies: average air temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 1 Page: 10 of 51 Event

Description:

Isolable Letdown Line Leak Inside Reactor Building - 50 gpm. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 RO CORRECTIVE ACTIONS: 3. Monitor LT-112A and LT-115, % LEVEL, to verify proper VCT level. 4. Monitor FI-122A, CHG FLOW GPM. SUPPLEMENTAL ACTIONS: 6. If RCS leakage is indicated, determine the leak rate and refer to Technical Specification 3.4.6.2. XCP-614 5-1 CRS Determines than RCS leakage is indicated and implements AOP-101.1, LOSS OF REACTOR COOLANT NOT REQUIRING SI. NOTE If a Reactor Trip occurs AND SI is NOT required, this procedure should be continued after the actions of EOP-1.1, Reactor Trip Recovery, are completed. As valves are isolated, it may be necessary to monitor RCS pressure for a period of time to determine if the leak is isolated. AOP-101.1

• RO 1 Verify PZR level is at or trending to program level. AOP-101.1

• RO 2 Check if SI is required: (NO) a. Check if any of the following criteria are met: PZR level is decreasing with Charging maximized and Letdown minimized. OR PZR level is approaching 8%. OR PZR pressure is approaching 1870 psig. OR VCT level is approaching 5%. ALTERNATIVE ACTION a. GO TO Step 3. AOP-101.1 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 1 Page: 11 of 51 Event

Description:

Isolable Letdown Line Leak Inside Reactor Building - 50 gpm. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 NOTE - Step 3 Conditions for implementing Emergency Plan Procedures should be evaluated using EPP.001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN. AOP-101.1 CRS 3 Determine RCS leak rate: a. Estimate the RCS leak rate: REFER TO IPCS CHGNET. REFER TO IPCS 4RW1. b. Estimate the RCS leak rate using IPCS VCT level. (14 gal/percent) c. If necessary, calculate the RCS leak rate. REFER TO STP-114.002, OPERATIONAL LEAK TEST. AOP-101.1 CRS 4 Check if the RCS leak rate is GREATER THAN Technical Specification 3.4.6.2 limits. AOP-101.1 CRS 5 Comply with the applicable Technical Specification 3.4.6.2 action statement. AOP-101.1 CRS TS 3.4.6.2 Reactor Coolant System operational leakage shall be limited to: b. 1 GPM UNIDENTIFIED LEAKAGE, ACTION b. With any operational Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, primary-to-secondary leakage, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. TS 3.4.6.2 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 1 Page: 12 of 51 Event

Description:

Isolable Letdown Line Leak Inside Reactor Building - 50 gpm. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 RO 6 Verify RCS pressure is GREATER THAN 2210 psig. AOP-101.1 EVALUATOR NOTE: Step 7 isolates the letdown leak. RO 7 Close all Letdown Isolation Valves: a. PVT8149A, LTDN ORIFICE A ISOL. b. PVT8149B, LTDN ORIFICE B ISOL. c. PVT8149C, LTDN ORIFICE C ISOL. d. LCV459, LTDN LINE ISOL. e. LCV460, LTDN LINE ISOL. AOP-101.1 RO 8 Check if the leak has been isolated: a. Evaluate the following: IPCS CHGNET IPCS 4RW1 Pressurizer level VCT Level Reactor Building Conditions b. If necessary, calculate the RCS leak rate. REFER TO STP114.002, OPERATIONAL LEAK TEST. AOP-101.1 BOOTH OPERATOR: If directed to investigate Relay Room Alarms - report there are Reactor 0 GPM. RO 9 Place Letdown in service using Attachment 1, Establishing Excess Letdown. AOP-101.1 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 1 Page: 13 of 51 Event

Description:

Isolable Letdown Line Leak Inside Reactor Building - 50 gpm. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 RO 1 Place Excess Letdown in service: a. Close FCV122, CHG FLOW. b. Maintain Core Power LESS THAN 2898 MWt prior to and during Excess Letdown operation. c. Close all Letdown Isolation Valves: 1) PVT8149A(B)(C), LTDN ORIFICE A(B)(C) ISOL. 2) LCV459, LTDN LINE ISOL. 3) LCV460, LTDN LINE ISOL. d. Isolate Charging by closing either of the following: MVG8107, CHG LINE ISOL. MVG8108, CHG LINE ISOL. e. Reduce Seal Injection flow to 7 gpm per RCP as indicated on the following: FI130A, RCP A INJ FLO GPM. FI127A, RCP B INJ FLO GPM. FI124A, RCP C INJ FLO GPM. f. Ensure HCV137, XS LTDN HX, is closed. AOP-101.1 ATTACH 1 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 1 Page: 14 of 51 Event

Description:

Isolable Letdown Line Leak Inside Reactor Building - 50 gpm. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 g. Place PVM8143, XS LTDN TO VCT OR RCDT, in VCT position. h. Ensure the following are open: MVT8100, SEAL WTR RTN ISOL. MVT8112, SEAL WTR RTN ISOL. i. Ensure MVG9583, FROM XS LTDN HX, is open. j. Open the following: PVT8153, XS LTDN ISOL. PVT8154, XS LTDN ISOL. k. Establish Excess Letdown flow: 1) Slowly throttle open HCV137, XS LTDN HX. 2) Maintain temperature on TI139, XS LETDOWN HX OUT TEMP °F, LESS THAN 165°F. l. Monitor the following to ensure flow between 0.2 gpm and 5.0 gpm: FR154A, RCP SL LKOFF HI RANGE. FR154B, RCP SL LKOFF LO RANGE. AOP-101.1 ATTACH 1 10 GO TO Step 42. AOP-101.1 42 Evaluate Plant status: a. Maintain stable plant conditions. b. Consult with the Shift Supervisor to determine further actions. AOP-101.1 43 RETURN TO Procedure and Step in effect. AOP-101.1 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 1 Page: 15 of 51 Event

Description:

Isolable Letdown Line Leak Inside Reactor Building - 50 gpm. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 EVALUATOR NOTE: The next event may be inserted after excess letdown is placed in service.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 2 Page: 16 of 51 Event

Description:

FT-Steam Flow Transmitter) fails LOW. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOOTH OPERATOR: When directed, initiate Event 2 (TRIGGER 2). EVALUATOR NOTE: On cue from the Examiner, FT-Flow transmitter for SG level control. FCV-498 will stroke closed causing SG level to lower. The crew will implement AOP-401.3 Steam Flow - Feedwater Flow Protection Channel Failure and select the operable Steam Flow Transmitter for control. This is a Technical Specification transmitter. Technical Specifications 3.3.1, Reactor Trip System Instrumentation, Action 6 and 3.3.2, Engineered Safety Feature Actuation System Instrumentation, Action 24 requires that the inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Indications Available: XCP-624 3-5, SG C LVL DEV XCP-624 6-4, SG C FWF>STF MISMATCH CRS Implements AOP-401.3, Steam Flow - Feedwater Flow Protection Channel Failure NOTE Throughout this procedure, "AFFECTED" refers to any SG experiencing level control problems. AOP-401.3 IOA BOP 1 Verify the failed channel is the controlling channel. AOP-401.3 NOTE - Step 2 FW AND STEAM CONTROL CHANNEL SEL Switches for a SG should be selected to the same direction (both to the left or both to the right). AOP-401.3 IOA BOP 2 Select the operable flow channel: Place FW CONTROL CHANNEL SEL Switch to the operable channel. Place STEAM CONTROL CHANNEL SEL Switch to the operable channel. AOP-401.3 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 2 Page: 17 of 51 Event

Description:

FT-Steam Flow Transmitter) fails LOW. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 NOTE - Step 3 CTRL+ALT+S on either EHC HMI is equivalent to 50 MWe, and is the preferred method to accomplish a rapid load reduction. AOP-401.3 IOA BOP 3 Verify Turbine Load is LESS THAN 950 MWe. AOP-401.3 IOA BOP 4 Verify only one SG is AFFECTED. AOP-401.3 IOA BOP 5 Adjust the Feedwater Flow Control Valve as necessary to restore feed flow to the AFFECTED SG. AOP-401.3 IOA BOP 6 Check if Feedwater Pump speed control is operating properly: Feedwater Header pressure is GREATER THAN Main Steam Header pressure. Feed flow is normal for steam flow and power level. All operating Feedwater Pump speeds and flows are balanced. AOP-401.3 BOP 7 Verify Narrow Range levels in all SGs are between 60% and 65%. AOP-401.3 BOP 8 Restore the AFFECTED SG control systems to normal: Place the Feedwater Flow Control Valve in AUTO. Place the Feedwater Pump Speed Control System in AUTO. REFER TO SOP-210, FEEDWATER SYSTEM. AOP-401.3 NOTE - Step 9 Steam flow transmitters FT-474, FT-484, FT-494, FT-475, FT-485, and FT-495 are density compensated by steam pressure transmitters PT-475, PT-485, PT-495, PT-476, PT-486, and PT-496. AOP-401.3 CRS 9 Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, place the failed channel protection bistables in a tripped condition: a. Identify the associated bistables for the failed channel. REFER TO Attachment 1. AOP-401.3 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 2 Page: 18 of 51 Event

Description:

FT-Steam Flow Transmitter) fails LOW. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 AOP-401.3 Attachment 1 CRS Identifies FB-474A and FB-478B as the affected Bistables. AOP-401.3 Attachment 1 CRS b. Record the following for each associated bistable on SOP-401, REACTOR PROTECTION AND CONTROL SYSTEM, Attachment I: Instrument. Associated Bistable. Bistable Location. STPs. AOP-401.3 CRS Refers to Technical Specifications: Table 3.3-1 item 14 (Action 6 within 72 hrs) - The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Table 3.3-3 item 4.d (Action 24 within 72 hrs) - The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Tech Specs BOOTH OPERATOR: Acknowledge requests for support in troubleshooting and placing channel in trip. CRS c. Notify the I&C Department to place the identified bistables in trip. AOP-401.3 EVALUATE NOTE: The next event may be initiated after the applicable Technical Specification Actions have been identified.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 3 Page: 19 of 51 Event

Description:

Rapid downpower due to B MFW pump vibration. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOOTH OPERATOR: When directed by the Evaluator Call the Control Room as the Shift Supervisor unusual. Engineering and Mechanical Maintenance. Reduce power to less than 45% at 3% per minute IAW GOP-4C, Rapid Power Reduction. . GOP-4C Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 3 Page: 20 of 51 Event

Description:

Rapid downpower due to B MFW pump vibration. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 EVALUATOR NOTE: The crew will lower power from 60% in accordance with GOP-4C, Rapid Power Reduction. The RO will utilize boration and/or rod control to lower power while coordinating the downpower with the BOP who will be controlling turbine demand. NOTE 3.0 If time allows, load reductions should be discussed with the Load Dispatcher. GOP-4C CAUTION 3.1 through 3.12 a. Thermal Power changes of greater than 15% in any one-hour period requires completion of GTP-702 Attachment III.H. b. VCS PID Report, POWER CHANGE SEARCH, should be periodically performed to ensure a thermal power change of greater than 15% in any one-hour period is detected. GOP-4C RO 3.1 Commence rapid Plant Shutdown as follows: a. Energize all Pressurizer Heaters. GOP-4C NOTE 3.1.b Setting FCV-113A&B, BA FLOW SET PT to 8.3 will yield 33 gpm Boration flow rate. GOP-4C RO b. Maintain the following with rod motion or boron concentration changes: 1) Tavg within 10°F and trending to Tref. 3) Control Rods above the rod insertion limit. GOP-4C RO c. Maintain Tavg within the control band by Control Rod motion or boron concentration changes. GOP-4C EVALUATOR NOTE: Applicable portions of SOP-106 are attached. RO 3.1.d. Borate or dilute per SOP-106, Reactor Makeup Water System, to maintain the following parameters: 1) 2) Control Rods above the Rod Insertion Limit. GOP-4C Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 3 Page: 21 of 51 Event

Description:

Rapid downpower due to B MFW pump vibration. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 NOTE 3.2 a. Step 3.2 lowers Reactor Power from 90% to 48%. b. While the plant is being maneuvered, total condensate flow through the Blowdown Heat Exchangers must be maintained greater than 450 gpm, which should maintain condensate outlet temperature at least 30°F below the DA temperature. GOP-4C BOP 3.2 Reduce Reactor Power to 48% as follows: GOP-4C EVALUATOR NOTE: Applicable portions of SOP-214 are attached. BOP a. Using the EHC HMI, Control/Load screen, reduce load per SOP-214 at a rate of 3% per minute or less. GOP-4C NOTE 3.2.b The System Controller should be notified prior to manually changing MVARs by more than 50 MVARs in a five minute period, unless the change is needed to prevent equipment damage. GOP-4C BOP b. As load decreases, adjust Megavars using GEN FIELD VOLT ADJ as requested by the System Controller and within the Estimated Generator Capability curve (Enclosure A). GOP-4C BOP c. When Reactor Power is between 60% and 80%, reduce to the following pumps in service per SOP-210, Feedwater System: 1) Two Main Feedwater Pumps. 2) Three Feedwater Booster Pumps. GOP-4C BOP d. When Reactor Power is between 60% and 75%, perform PTP-102.001, Main Turbine Tests (Power Operated Extraction System Check Valve portion only). NA Power was never increased above 60%, GOP-4C EVALUATOR NOTE: Event 4 (Stuck Rod) will be auto-triggered at 56% power so that it is apparent that a rod is not moving.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 4 Page: 22 of 51 Event

Description:

Rod H14 stuck but trippable (blown fuse). Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOOTH OPERATOR: NO ACTION REQUIRED. Event 4 (TRIGGER 3) will auto-initiate when power is below 56%. EVALUATOR NOTE: This event will auto-actuate when power is reduced below 56%. Control Rod H-14 in Control Bank D will stop moving. The RO will realign the control rods in accordance with AOP-403.5 Stuck or Misaligned Control Rod. Indications available: IPCS alarm, XCP-620 2-5 CMPTR ROD DEV CRS Refer to Alarm Response Procedure ARP-001-XCP-620 2-5 CORRECTIVE ACTIONS: 1. Observe the Digital Rod Position Indication display for proper rod positions. 2. Determine if the cause is a dropped or misaligned rod. 3. If DRPI ALARM URGENT is in refer to ARP-001-XCP-621, 2-1. XCP-620 2-5 SUPPLEMENTAL ACTIONS: 1. If a rod is misaligned, refer to AOP-403.5, Stuck or Misaligned Rod. 3. Operate the Rod Control System in MAN as described in SOP-403 until proper automatic Rod Control in restored. 4. Refer to Technical Specification 3.1.3.1. XCP-620 2-5 CRS Implement AOP-403.5, Stuck or Misaligned Rod.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 4 Page: 23 of 51 Event

Description:

Rod H14 stuck but trippable (blown fuse). Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 CRS Apply Technical Specification 3.1.3.1 Action d d. With one full length rod inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than12 steps (indicated position), POWER OPERATION may continue provided that within one hour either: 1. The rod is restored to OPERABLE status within the above alignment requirements, or 2. The remainder of the rods in the group with the inoperable rod are aligned to within 12 steps of the inoperable rod within one hour while maintaining the rod sequence and insertion limits specified in the CORE OPERATING LIMITS REPORT (COLR); the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or TS-3.1.3.1 Action d IOA RO 1 Place ROD CNTRL BANK SEL Switch in MAN. AOP-403.5 IOA RO 2 Check if Reactor Power is GREATER THAN OR EQUAL TO 5%. AOP-403.5 IOA BOP 3 Stabilize Main Turbine load/Steam Dumps demand. AOP-403.5

• CREW 4 Maintain Tavg within 5°F of Tref using the following: Main Turbine load or Steam Dumps demand adjustment. RCS Boration or Dilution. REFER TO SOP-106, REACTOR MAKEUP WATER SYSTEM. AOP-403.5 NOTE - Steps 5 through 16 Throughout the following steps, "AFFECTED" refers to any Rod Bank which contains a misaligned Control Rod. AOP-403.5 RO 5 Record the misaligned Control Rod and AFFECTED Bank: Misaligned Rod: _________. AFFECTED Bank: _________. AOP-403.5 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 4 Page: 24 of 51 Event

Description:

Rod H14 stuck but trippable (blown fuse). Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 NOTE - Step 6 Computer rod positions can be found at Group Display DRPIRODS. AOP-403.5 RO 6 Record the Control Rod positions and Group Step Counter demands: Control Bank A DRPI: __________. Computer: __________. Demand Group 1: __________. Demand Group 2: __________. Control Bank B DRPI: __________. Computer: __________. Demand Group 1: __________. Demand Group 2: __________. Control Bank C DRPI: __________. Computer: __________. Demand Group 1: __________. Demand Group 2: __________. Control Bank D DRPI: __________. Computer: __________. Demand Group 1: __________. Demand Group 2: __________. Shutdown Bank A DRPI: __________. Computer: __________. Demand Group 1: __________. Demand Group 2: __________. Shutdown Bank B DRPI: __________. Computer: __________. Demand Group 1: __________. Demand Group 2: __________. AOP-403.5 CRS 7 Notify the following plant personnel prior to moving Control Rods: Management Duty Supervisor. Rod Control System Engineer. Reactor Engineering. AOP-403.5 CRS 8 Notify the I&C Department to investigate the cause of the Control Rod misalignment. AOP-403.5 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 4 Page: 25 of 51 Event

Description:

Rod H14 stuck but trippable (blown fuse). Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOOTH OPERATOR: Acknowledge request for I&C support Wait 3 minutes Report as a blown lift Request permission to replace blown fuse. Insert TRIGGER 4 - Removes the blown fuse failure. Report as I&C that the CAUTION - Step 9 IF rod alignment could result in a mode change or a subcritical Reactor reaching criticality, then the plant shall be shut down to Mode 3. AOP-403.5 RO 9 Verify the misaligned Control Rod is NOT located on bottom of core. AOP-403.5 CRS 10 Provide Reactor Engineering with the following information: Time Control Rod noticed to be AFFECTED: _________. AFFECTED Control Rod location: _________. Initial Reactor power level: _________. Current Reactor power level: _________. Current QPTR: _________. AOP-403.5 Acknowledge request for Reactor Engineering support Wait 2 minutes after being provided with the information from step 10. Notify the CRS o Perform rod recovery at the current power level. o There is no restriction on rod withdrawal speed. NOTE - Step 11 This Step must be completed before continuing with Step 12. AOP-403.5 CRS 11 Obtain the following information from Reactor Engineering: Power level at which recovery is to be performed: _________. Rate of Control Rod movement during recovery: _________. AOP-403.5 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 4 Page: 26 of 51 Event

Description:

Rod H14 stuck but trippable (blown fuse). Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 RO 12 Rotate the ROD CNTRL BANK SEL Switch clockwise to the AFFECTED Bank position. AOP-403.5 RO 13 Check if the misaligned Control Rod can be moved: a. Monitor DRPI. b. Using the rate of Control Rod movement determined in Step 11, move the AFFECTED Bank six steps in the direction of the misaligned rod. c. Using the rate of Control Rod movement determined in Step 11, move the AFFECTED Bank six to eight steps in the direction of its original position. AOP-403.5 NOTE - Step 13.d Technical Specification 3.1.3.1 requires plant shutdown if a Control Rod can NOT be moved due to excessive friction or mechanical interference in Mode 1 or 2 OR is known to be untrippable. AOP-403.5 RO d. Check if the misaligned Control Rod moved. e. Using the rate of Control Rod movement determined in Step 11, return the AFFECTED Bank to its original position. AOP-403.5 RO 14 If necessary, reduce Reactor power to the power level determined in Step 11. REFER TO GOP-4B, Power Operation (Mode 1 - Descending) Or Gop-4c, Rapid Power Reduction. AOP-403.5 RO 15 Align the misaligned Control Rod with the AFFECTED Bank: a. At the CONTROL ROD DISCONNECT SWITCH BOX inside the MCB, place all Lift Coil Disconnect Switches for the AFFECTED Bank, except the switch for the misaligned Control Rod, to the ROD DISCONNECTED position. b. Dispatch an operator with the Rod Control Cabinets Key to the Rod Control Cabinet room (IB-463). AOP-403.5 NOTE - Step 15.c This step is only applicable for Control Banks. AOP-403.5 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 4 Page: 27 of 51 Event

Description:

Rod H14 stuck but trippable (blown fuse). Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOOTH OPERATOR: Acknowledge request for a field operator. Report you are located at the Rod Control Cabinet. If directed to report to the Control Room for the key REPORT I&C is with you and they have a key. Use INSIGHT ZCRFPA(6), REAL PA VALUE to PCSROD, for the P/A converter reading for the bank D. NA c. Locally at XCA4-CR, P/A CONVERTER CABINET (IB-463), record the P/A CONVERTER reading for the AFFECTED Bank: ___________. AOP-403.5 NOTE - Step 15.d ROD CNTRL SYS FAIL URGENT (XCP-620 5-1), annunciator will alarm when a misaligned rod is moved in this step. AOP-403.5 RO d. Using the rate of Control Rod movement determined in Step 11, move the misaligned Control Rod six steps in the direction of the AFFECTED Bank. e. Verify only the misaligned Control Rod moved. f. Using the rate of Control Rod movement determined in Step 11, continue moving the misaligned Control Rod until it is realigned with the AFFECTED Bank. AOP-403.5 RO 16 Reset the Group Step Counters and P/A CONVERTER: a. Reset the Bank Group Step Counters to indicate the Group Demands recorded in Step 6. AOP-403.5 NOTE - Step 16.b This step is only applicable for Control Banks. AOP-403.5 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 4 Page: 28 of 51 Event

Description:

Rod H14 stuck but trippable (blown fuse). Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOOTH OPERATOR: When directed - reset P/A converter using Remote Functions Insert LOA-CRF005, P/A MAN HEIGHT VALUE (USE BEFORE SETTING LOA CRF1), to the value recorded in step 15. Insert LOA-CRF001, P/A MAN BANK SELECT (USE AFTER SETTING LOA CRF5), to CB D. b. Locally at XCA4-CR, P/A CONVERTER CABINET (IB-463), reset the P/A CONVERTER as follows: 1) Ensure the Bank Position Display Switch is in the AFFECTED Bank position. 2) Place MANUAL/AUTOMATIC Switch in MANUAL. 3) Depress the UP or DOWN Pushbutton to reset the P/A CONVERTER to the reading recorded in Step 15. 4) Place the MANUAL/AUTOMATIC Switch in AUTOMATIC. 5) Place the Bank Position Display Switch to DISPLAY OFF. AOP-403.5 NOTE - Step 17 If the Control Rods are near the All Rods Out position, SOP-403, ROD CONTROL AND POSITION INDICATING SYSTEM, should be used for final alignment. AOP-403.5 BOP 17 Restore the Rod Control System to normal alignment: a. Place all Lift Coil Disconnect Switches to the ROD CONNECTED position. b. Rotate the ROD CNTRL BANK SEL Switch counter-clockwise to MAN. c. Depress the ROD CNTRL ALARM RESET Pushbutton. d. Verify the ROD CNTRL SYS FAIL URGENT (XCP-620 5-1), annunciator clears. AOP-403.5 EVALUATOR NOTE: Initiate the next event after rod alignment has been restored.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 5 Page: 29 of 51 Event

Description:

PT-508 (MFW Pump Discharge Header Pressure) Fails LOW. (Manual control of MFW Pp speed) Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOOTH OPERATOR: When directed, insert Event 5 (TRIGGER 5) EVALUATOR NOTE: On cue from the Examiner, a Main Feedwater Header Pressure transmitter will fail LOW causing the MFW Pump speed to increase and raise SG level. The operators will respond to annunciators and implement AOP-210.3, Feedwater Pump Malfunction. The BOP will take manual control of the master Speed control and adjust speed to maintain Feedwater Pump discharge pressure 150 to 250 psi greater than Main Steam Header pressure and restore SG levels. Indications Available: XCP-624 1-5; 2-5; 3-5, SG LVL DEV XCP-624 4-4, 5-4, 6-4, FWF>STF MISMATCH CREW Responds to multiple SG LVL DEV alarms and/or change in feedwater flow. EVALUATOR NOTE: The crew may first enter an ARP but could go directly to AOP-210.3, FEEDWATER PUMP MALFUNCTION, based on multiple alarms or early diagnosis. CRS Enters ARP-001-XCP-624 1-5 or 2-5 or 3-5 BOP CORRECTIVE ACTIONS: 1. If required, restore Steam Generator A level to between 60% and 65% by performing either or both of the following: a. Manually control PVT-478, SG A FWF, as required. b. Manually control Feedwater Pump speed as follows: 1) Place the Feedwater Pump MASTER SPEED CNTRL in MAN. 2) Adjust the differential pressure between Feedwater Pump discharge header pressure and Main Steam header pressure, as required, to restore Steam Generator water level. XCP-624 1-5 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 5 Page: 30 of 51 Event

Description:

PT-508 (MFW Pump Discharge Header Pressure) Fails LOW. (Manual control of MFW Pp speed) Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOP 2. Evaluate SG A Narrow Range level indicators LI-474, LI-475, and LI-476: a. For increasing level: 1) At 70% Narrow Range level: (a) During startups (below 15% power) close the Feed Regulating valves with the B Train Switches. (b) When above 15% power take manual control of PVT-478, SG A FWF. (c) Ensure Feed Flow is 200 kbh to 400 kbh less than Steam Flow. 2) At 75% Narrow Range level: (a) Trip the Reactor if above 15% power. (b) Close the Feed Isolation valves. (c) Trip the Turbine. (d) Trip the Feed Pumps. (e) Close the Feedwater Regulating valves, if not closed earlier. (f) If the Reactor has NOT been tripped, reduce power to between 1% and 3% (g) Reestablish Emergency Feed. XCP-624 1-5 CRS 3. If FCV-478, A FCV, malfunctioned go to AOP-210.1, Feedwater Flow Control Valve Failure. (NO) XCP-624 1-5 CRS 4. If a Main Feedwater Pump has tripped or is malfunctioning go to AOP-210.3, Feedwater Pump Malfunction. XCP-624 1-5 CRS Implements AOP-210.3, Feedwater Pump Malfunction.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 5 Page: 31 of 51 Event

Description:

PT-508 (MFW Pump Discharge Header Pressure) Fails LOW. (Manual control of MFW Pp speed) Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 REFERENCE PAGE FOR AOP210.3 1 Manual Control of Main Feedwater Regulating Valves is permissible at any time as deemed necessary during the performance of this procedure. If a Main Feedwater Regulating Valve has been placed in Manual it should be returned to Automatic as soon as possible. 2 IF only one Main Feedwater Pump is operating and cannot be controlled THEN trip the Main Turbine and go to AOP214.1, TURBINE TRIP. 3 IF Narrow Range SG level decreases to LESS THAN 40%, THEN Trip the reactor and enter EOP1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION. 4 IF Reactor Power is GREATER THAN 15% and NR Steam Generator level exceeds 75%, THEN Trip the reactor and enter EOP1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION. AOP-210.3 IOA BOP 1 Verify at least one Feedwater Pump is running. AOP-210.3 IOA BOP 2 Check if a Feedwater Pump trip occurred. (NO) ALTERNATIVE ACTION 2 GO TO Step 4. AOP-210.3 IOA BOP 4 Check Main Feedwater Pump operation. a. Verify all Main Feedwater Pumps are affected. b. Place the MCB MASTER SPEED CNTRL in MAN. and adjust the MCB MASTER SPEED CNTRL as necessary to match Steam Flow and Feedwater Flow. AOP-210.3 BOP 5 If necessary, place the Main Feed Regulating valves in Manual. AOP-210.3 NOTE - Step 6 Due to the slow operation of the Main Feedwater Pump Recirculation Valves, a constant Main Feedwater Pump speed should be maintained until the recirculation valves have become relatively stable while adjusting Feedwater Flow. AOP-210.3 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 5 Page: 32 of 51 Event

Description:

PT-508 (MFW Pump Discharge Header Pressure) Fails LOW. (Manual control of MFW Pp speed) Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1

• 6 Maintain Narrow Range Steam Generator Water level between 60% and 65% AOP-210.3 EVALUATOR NOTE: Manual speed control will be maintained for the remainder of the scenario. EVALUATOR NOTE: The NEXT EVENT may be initiated after the Main Feedwater Pump Master speed controller is in manual. 7 WHEN conditions allow, Place Main Feed Regulating valves in AUTO. AOP-210.3 NOTE - Step 8 Main Feedwater Program P should be established using the following as available: PI508, FW PP DISCH HDR PRESS PSIG. Any operating Main Feedwater Pump Discharge Pressure. PI464C, MS HDR PRESS PSIG. Any available MCB Main Steam Header Pressure. IPCS (ZZMENU S/G SU Trend or FW Start) AOP-210.3 8 Restore Feedwater Pump D/P to program. a. Using the Feedwater Pump Speed Control method established in Step 4, slowly adjust Feedwater Pump discharge header pressure to within the limits of ATTACHMENT 1, FEEDWATER PUMP D/P LIMITS. b. Adjust PUMP A(B)(C) SPEED CNTRL (MCB M/A Stations) as necessary to balance all operating Feedwater Pumps speed to within 120 rpm of each other. AOP-210.3 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 33 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOOTH OPERATOR: When directed, initiate Event 6 (TRIGGER 6). EVALUATOR NOTE (EVENT 6): On cue from the Examiner, a steamline break inside the Reactor Building will be inserted. The Reactor will trip and the crew will implement EOP-1.0 (E-0) Reactor Trip/Safety Injection Actuation. The crew will identify that at least one Steam generator is faulted and transition to EOP-3.0 (E-2), Faulted Steam Generator Isolation. When the faulted SG is isolated the crew will transition to EOP-1.2 (ES-1.1), Safety Injection Termination. The crew will identify that the Main Generator Output Breaker failed to automatically trip and the BOP will manually open the breaker from the control board. The BOP must identify that the MSIVs are open and manually close them from the control board to isolate the faulted Steam Generator and prevent over-pressurization of the Reactor Building. Indications Available: XCP-621 3-2 RODS ON BOTTOM XCP-624 2-4 STMLN PRESS LO XCP-626 4-1 STM PRESS LO SI XCP-626 6-1 RB PRESS HI-1 SIl Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 34 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 EOP-1.0 NOTE Steps 1 through 5 are Immediate Operator Actions. The EOP REFERENCE PAGE should be monitored throughout the use of this procedure. Conditions for implementing Emergency Plan Procedures should be evaluated using EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN. EOP-1.0 IOA RO 1 Verify Reactor Trip: Trip the Reactor using either Reactor Trip Switch. Verify all Reactor Trip and Bypass Breakers are open. Verify all Rod Bottom Lights are lit. Verify Reactor Power level is decreasing. EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 35 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 IOA BOP 2 Verify Turbine/Generator Trip: a. Verify all Turbine STM STOP VLVs are closed. EOP-1.0 EVALUATOR NOTE: The GEN BKR and GEN FIELD BKR are failed and need to be manually opened. IOA BOP b. Ensure Generator Trip (after 30 second delay): 1) Ensure the GEN BKR is open. (NO) 2) Ensure the GEN FIELD BKR is open. (NO) 3) Ensure the EXC FIELD CNTRL is tripped. EOP-1.0 IOA BOP 3 Verify both ESF buses are energized. EOP-1.0 IOA RO 4 Check if SI is actuated: a. Check if either: SI ACT status light is bright on XCP-6107 1-1. OR Any red first-out SI annunciator is lit on XCP-626 top row. b. Actuate SI using either SI ACTUATION Switch. C. GO TO Step 6. EOP-1.0 EVALUATOR NOTE: ATTACHMENT 3, SI EQUIPMENT VERIFICATION is included in a separate section. BOP 6 Initiate ATTACHMENT 3, SI EQUIPMENT VERIFICATION. EOP-1.0 CREW 7 Announce plant conditions over the page system. EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 36 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1

• RO 8 Verify RB pressure has remained LESS THAN 12 psig on PR-951, RB PSIG (P-951), red pen. (NO) ALTERNATIVE ACTION 8 Perform the following: a) Verify both the following annunciators are lit: XCP-612 3-2 (RB SPR ACT). XCP-612 4-2 (PHASE B ISOL). IF either annunciator is NOT lit, THEN actuate RB Spray by placing the following switches to ACTUATE: Both CS-SGA1 and CS-SGA2. OR Both CS-SGB1 and CS-SGB2. b) Verify Phase B Isolation by ensuring RB SPRAY/PHASE B ISOL monitor lights are bright on XCP-6105. c) Ensure the following are open: MVG-3001A(B), RWST TO SPRAY PUMP A(B) SUCT. MVG-3002A(B), NAOH TO SPRAY PUMP A(B) SUCT. MVG-3003A(B), SPRAY HDR ISOL LOOP A(B). d) Ensure both RB Spray Pumps are running. e) Verify RB Spray flow is GREATER THAN 2500 gpm for each operating train on: FI-7368, SPR PP A DISCH FLOW GPM. FI-7378, SPR PP B DISCH FLOW GPM. f) Stop all RCPs. EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 37 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1

• RO 9 Check RCS temperature: With any RCP running, RCS Tavg is stable at OR trending to 557°F. OR With no RCP running, RCS Tcold is stable at OR trending to 557°F. ALTERNATIVE ACTION 9 IF RCS temperature is LESS THAN 557°F AND decreasing, THEN stabilize temperature by performing the following as required: a) Close IPV-2231, MS/PEGGING STM TO DEAERATOR. b) Perform one of the following: IF Narrow Range SG level is LESS THAN 41% in all SGs, THEN reduce EFW flow as necessary to stop cooldown, while maintaining total EFW flow GREATER THAN 450 gpm. OR WHEN Narrow Range SG level is GREATER THAN 41% in at least one SG, THEN control EFW flow as necessary to stabilize RCS temperature at 557°F. c) Initiate ATTACHMENT 6, STEAM VALVE ISOLATION, while continuing with this procedure. EOP-1.0 EVALUATOR NOTE: Close "A" or "B" MSIV Prior to Orange path on Integrity or Containment.

• Critical Task RO d) IF RCS cooldown continues, THEN close: MS Isolation Valves, PVM-2801A(B)(C). MS Isolation Bypass Valves, PVM-2869A(B)(C). EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 38 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 RO 10 Check PZR PORVs and Spray Valves: a. PZR PORVs are closed. b. PZR Spray Valves are closed. c. Verify power is available to at least one PZR PORV Block Valve: MVG-8000A, RELIEF 445 A ISOL. MVG-8000B, RELIEF 444 B ISOL. MVG-8000C, RELIEF 445 B ISOL. d. Verify at least one PZR PORV Block Valve is open. EOP-1.0 NOTE - Step 11 Seal Injection flow should be maintained to all RCPs. EOP-1.0 RO 11 Check if RCPs should be stopped: a. Check if either of the following criteria is met: Annunciator XCP-612 4-2 is lit (PHASE B ISOL). OR RCS pressure is LESS THAN 1418 psig AND SI flow is indicated on FI-943, CHG LOOP B CLD/HOT LG FLOW GPM. b. Stop all RCPs. EOP-1.0 RO 12 Verify no SG is FAULTED: No SG pressure is decreasing in an uncontrolled manner. No SG is completely depressurized. ALTERNATIVE ACTION 12 GO TO EOP-3.0, FAULTED STEAM GENERATOR ISOLATION, Step 1. EOP-1.0 CRS Implement EOP-3.0, Faulted Steam Generator Isolation, Step 1.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 39 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 EVALUATOR NOTE: If EFW pump runout protection has occurred, the CRS may enter and exit EOP-15.0, Response To Loss Of Secondary Heat Sink. CAUTION At least one SG must be maintained available for RCS cooldown. Any FAULTED SG or secondary break should remain isolated during subsequent recovery actions unless needed for RCS cooldown, to prevent reinitiating the break. EOP-3.0 NOTE Conditions for implementing Emergency Plan Procedures should be evaluated using EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN. EOP-3.0 BOP 1 Ensure all the following are closed: MS Isolation Valves, PVM-2801A(B)(C). MS Isolation Bypass Valves, PVM-2869A(B)(C). EOP-3.0 BOP 2 Check if any SG is NON-FAULTED: Pressure in any SG is stable OR increasing. Any SG is NOT completely depressurized. EOP-3.0 BOP 3 Identify any FAULTED SG(s): Any SG pressure decreasing in an uncontrolled manner. OR Any SG completely depressurized. EOP-3.0 BOP 4 Close the following for each FAULTED SG: FW Flow Control, FCV-478. FW Isolation, PVG-1611A. SG Blowdown, PVG-503A. FW Flow Control Bypass, FCV-3321. EOP-3.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 40 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOP 5 Complete the isolation of each FAULTED SG: a. Close SG Chemical Feed Isolation, MVK-1633A. b. Close MS Drain Isolation, PVT-2843A. c. Close MS Drain Isolation, PVT-2877A for SG A PVT-2877B for SG C. d. Place the Steamline PWR RELIEF A SETPT Controller(s) in MAN and closed. e. Place the Steamline Power Relief A Mode Switch(s) in PWR RLF. EOP-3.0 CRITICAL TASK BOP f. Close FCV-3531, MD EFP TO SG A. g. Close FCV-3536, TD EFP TO SG A. EOP-3.0 CAUTION - Step 5.h If the TD EFW Pump is the only available source of feed flow, the steam supply to the TD EFW Pump must be maintained from at least one SG, to maintain a secondary heat sink. EOP-3.0 BOP h. Close and locally deenergize the appropriate valve if SG B or SG C is FAULTED: (NA) EOP-3.0 NOTE - Step 6 Any high radiation level received on a radiation monitor that was unisolated at event initiation, may be considered a valid alarm. EOP-3.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 41 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 CRS 6 Check if Secondary radiation levels are normal: a. Check radiation levels normal on all unisolated radiation monitors: RM-G19A(B)(C), STMLN HI RNG GAMMA. RM-L3, STEAM GENERATOR BLOWDOWN LIQUID MONITOR. RM-L10, SG BLOWDOWN CW DISCHARGE LIQUID MONITOR. RM-A9, CNDSR EXHAUST GAS ATMOS MONITOR. b. Notify Chemistry to sample all SG secondary sides, and screen samples for abnormal activity using a frisker. EOP-3.0 CRS 7 Check if SI flow should be reduced: a. RCS subcooling on TI-499A(B), A(B) TEMP °F, is GREATER THAN 52.5°F [67.5°F]. b. Secondary Heat Sink is adequate: Total EFW flow to INTACT SGs is GREATER THAN 450 gpm. OR Narrow Range level is GREATER THAN 41% in at least one INTACT SG. c. RCS pressure is stable OR increasing. d. PZR level is GREATER THAN 28%. EOP-3.0 RO 8 Reset both SI RESET TRAIN A(B) Switches. EOP-3.0 RO 9 Reset Containment Isolation: RESET PHASE A - TRAIN A(B) CNTMT ISOL. RESET PHASE B - TRAIN A(B) CNTMT ISOL. EOP-3.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 42 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOP 10 Place both ESF LOADING SEQ A(B) RESETS to: a. NON-ESF LCKOUTS. b. AUTO-START BLOCKS. EOP-3.0 BOP 11 Establish Instrument Air to the RB: a. Start one Instrument Air Compressor and place the other in Standby. b. Open PVA-2659, INST AIR TO RB AIR SERV. c. Open PVT-2660, AIR SPLY TO RB. EOP-3.0 CRS 12 GO TO EOP-1.2, SAFETY INJECTION TERMINATION, Step 1. EOP-3.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 43 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 EOP-1.2 NOTE The EOP REFERENCE PAGE should be monitored throughout the use of this procedure. EOP-1.2 RO 1 Stop all but one Charging Pump and place in Standby. EOP-1.2 RO 2 Verify RCS pressure is stable OR increasing. EOP-1.2 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # 6 Page: 44 of 51 Event

Description:

. Generator breaker failure. A and B MSIVs fail to close. Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 RO 3 Establish Normal Charging: a. Close FCV-122, CHG FLOW. b. Open both MVG-8107 and MVG-8108, CHG LINE ISOL. c. Adjust FCV-122, CHG FLOW, to obtain 70 gpm Charging flow. d. Close both MVG-8801A(B), HI HEAD TO COLD LEG INJ. EOP-1.2 EVALUATOR NOTE: The scenario may be terminated now the Safety Injection has been terminated (i.e. normal charging restored).

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # NA Page: 45 of 51 Event

Description:

SOP-106, BORATE OPERATIONS Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 NOTE 2.0 1. Energizing additional Pressurizer Heaters will enhance mixing. 2. LCV-115A, LTDN DIVERT TO HU-TK, will begin to modulate to the HU-TK position at 70% level on LI-115, VCT LEVEL %. SOP-106 RO 2.1 Ensure at least one Reactor Coolant Pump is running. SOP-106 RO 2.2 Place RX COOL SYS MU switch to STOP. SOP-106 RO 2.3 Place RX COOL SYS MU MODE SELECT switch to BOR. SOP-106 RO 2.4 Set FIS-113, BA TO BLNDR FLOW, batch integrator to the desired volume. SOP-106 RO 2.5 Place RX COOL SYS MU switch to START. SOP-106 NOTE 2.6 Step 2.6 may be omitted when borating less than 10 gallons. SOP-106 RO 2.6 Place FCV-113 A&B, BA FLOW, controller in AUTO. SOP-106 NOTE 2.7 The AUTO setpoint dial for FCV-113A&B, BA FLOW, controller may be adjusted slowly to obtain the desired flow rate. SOP-106 RO 2.7 Verify the desired Boric Acid flow rate on FR-113, BA TO BLNDR GPM (F-113). SOP-106 RO 2.8 When the preset volume of boric acid has been reached, perform the following: a. Place FCV-113A&B, BA flow controller in MAN. b. Verify boration stops. SOP-106 RO 2.9 Place RX COOL SYS MU switch to STOP. SOP-106 NOTE 2.10 a. If plant conditions require repeated borations, Step 2.10 may be omitted. b. The volume in the piping between the blender and the VCT outlet is approximately 3.8 gallons. SOP-106 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # NA Page: 46 of 51 Event

Description:

SOP-106, BORATE OPERATIONS Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 RO 2.10 Alternate Dilute 4 to 6 gallons of Reactor Makeup Water to flush the line down stream of the blender by performing the following: a. Place RX COOL SYS MU MODE SELECT switch to ALT DIL. b. Adjust FCV-168, TOTAL MU FLOW SET PT, to desired flow rate. c. Set FIS-168, TOTAL MU FLOW, batch integrator to desired volume. d. Place RX COOL SYS MU switch to START. e. Verify desired flow rate on FR-113, TOTAL MU GPM (F-168). f. Verify alternate dilution stops when preset volume is reached on FIS-168, TOTAL MU FLOW, batch integrator. g. Place RX COOL SYS MU switch to STOP. SOP-106 RO 2.11 Place RX COOL SYS MU MODE SELECT switch to AUTO. SOP-106 RO 2.12 Adjust FCV-168, TOTAL MU FLOW SET PT, to 7.5 (120 gpm). SOP-106 RO 2.13 In MAN, adjust FCV-113 A&B, BA FLOW OUTPUT, to the required position which will ensure proper Boric Acid addition for subsequent Automatic Makeup operations. SOP-106 RO 2.14 Adjust FCV-113A&B, BA FLOW SET PT, to the desired position to ensure proper boric acid addition for subsequent Automatic Makeup operations. SOP-106 RO 2.15 Place RX COOL SYS MU switch to START. SOP-106 RO 2.16 Perform the following: a. Start XPP-13A(B), BA XFER PP A(B), for the in-service Boric Acid Tank. b. If necessary, start XPP-13A(B), BA XFER PP A(B), for the Boric Acid Tank on recirculation. SOP-106 END OF SECTION SOP-106 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # NA Page: 47 of 51 Event

Description:

SOP-214, Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOP 2.1 Ensure the Control/Load screen is selected. SOP-214 NOTE 2.2 The turbine will come off the limiter and turbine load will lower once Load Set Reference is less than Load Limit Reference. SOP-214 BOP 2.2 To lower Turbine Load using Load Set, perform the following: a. If directed by Operations Management, disable the Turbine Vibration Trips per Section III. b. Select (or enter) the desired Rate %/min on Load Set. c. Select Load on Load Set (a dialog box will open). d. Enter the desired load and confirm. e. Verify proper system response. f. If during a load reduction, it is desired to stop the load reduction, perform the following: 1) Select Hold on Load Set. 2) Select the desired Rate %/min to resume load reduction. 3) If desired, place LOAD LIMIT in service per Section III. SOP-214 BOP 2.3 For rapid load shedding of 50 MWe, on an HMI keypad select Ctrl + Alt + S. SOP-214 EVALUATOR NOTE: The remainder of this section deals with actions after Turbine Load is below 15%.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # NA Page: 48 of 51 Event

Description:

EOP-1.0, Attachment 3 Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOP 1 Ensure EFW Pumps are running: a. Ensure both MD EFW Pumps are running. b. Verify the TD EFW Pump is running if necessary to maintain SG levels. EOP-1.0 Attachment 3 BOP 2 Ensure the following EFW valves are open: FCV-3531(3541)(3551), MD EFP TO SG A(B)(C). FCV-3536(3546)(3556), TD EFP TO SG A(B)(C). MVG-2802A(B), MS LOOP B(C) TO TD EFP. Attachment 3 BOP 3 Verify total EFW flow is GREATER THAN 450 gpm. Attachment 3 BOP 4 Ensure FW Isolation: a. Ensure the following are closed: FW Flow Control, FCV-478(488)(498). FW Isolation, PVG-1611A(B)(C). FW Flow Control Bypass, FCV-3321(3331)(3341). SG Blowdown, PVG-503A(B)(C). SG Sample, SVX-9398A(B)(C). b. Ensure all Main FW Pumps are tripped. Attachment 3 BOP 5 Ensure SI Pumps are running: Two Charging Pumps are running. Both RHR Pumps are running. Attachment 3 BOP 6 Ensure two RBCU Fans are running in slow speed (one per train). Attachment 3 BOP 7 Verify Service Water to the RBCUs: a. Ensure two Service Water Pumps are running. b. Verify both Service Water Booster Pumps A(B) are running. c. Verify GREATER THAN 2000 gpm flow for each train on: FI-4466, SWBP A DISCH FLOW GPM. FI-4496, SWBP B DISCH FLOW GPM. Attachment 3 BOP 8 Verify two CCW Pumps are running. Attachment 3 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # NA Page: 49 of 51 Event

Description:

EOP-1.0, Attachment 3 Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOP 9 Ensure two Chilled Water Pumps and Chillers are running. Attachment 3 BOP 10 Verify both trains of Control Room Ventilation are running in Emergency Mode. Attachment 3 EVALUATOR NOTE: It is a critical task to close the "A" or "B" MSIV Prior to Orange path on Integrity or RB pressure. CRITICAL TASK BOP 11 Check if Main Steamlines should be isolated: a. Check if any of the following conditions are met: RB pressure GREATER THAN 6.35 psig. OR Steamline pressure LESS THAN 675 psig. OR Steamline flow GREATER THAN 1.6 MPPH AND Tavg LESS THAN 552°F. b. Ensure all the following are closed: MS Isolation Valves, PVM-2801A(B)(C). MS Isolation Bypass Valves, PVM-2869A(B)(C). Attachment 3 BOP 12 Ensure Excess Letdown Isolation Valves are closed: PVT-8153, XS LTDN ISOL. PVT-8154, XS LTDN ISOL. Attachment 3 BOP 13 Verify ESF monitor lights indicate Phase A AND Containment Ventilation Isolation on XCP-6103, 6104, and 6106. REFER TO ATTACHMENT 4, CONTAINMENT ISOLATION VALVE MCB STATUS LIGHT LOCATIONS, as needed. Attachment 3 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # NA Page: 50 of 51 Event

Description:

EOP-1.0, Attachment 3 Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOP 14 Verify proper SI alignment: a. Verify SI valve alignment by verifying SAFETY INJECTION/PHASE A ISOL monitor lights are bright on XCP-6104. b. Verify all SAFETY INJECTION monitor lights are dim on XCP-6106. c. Verify SI flow on FI-943, CHG LOOP B CLD/HOT LG FLOW GPM. d. Check if RCS pressure is LESS THAN 325 psig. Attachment 3 BOP Report completion of Attachment 3. EVALUATOR NOTE: ATTACHMENT 3 is complete.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 2 Event # NA Page: 51 of 51 Event

Description:

EOP-1.0, STEAM VALVE ISOLATION, Attachment 6 Time Position NRC 2015 Scenario 2 NUREG-1021 R9 S1 BOP 1 Close Feedwater Pump TURB DRN VLVs: MOV-1-5A. MOV-1-5B. MOV-1-5C. EOP-1.0 Attachment 6 BOP 2 Close the following turbine drain valves: MVG-2896A, SV-1 BSD. MVG-2896B, SV-2 BSD. MVG-2896C, SV-3 BSD. MVG-2896D, SV-4 BSD. EOP-1.0 Attachment 6 BOP 3 Ensure the following drain valves are in AUTO: PVT-2851A,B,C,D, MS LINES TO TURB DRN. PVT-2713A,B,C,D, STM DUMP DRN BYP. PVT-2870, TO MSR A & B DRN. PVT-2875, TO MSR A & B DRN. PVT-2845A,B,C, PVT-2824, PVT-2879A,B, LINE DRN. PVT-2838A,B, HDR DRNS. EOP-1.0 Attachment 6 BOP 4 Place the STM DUMP CNTRL Controller in MAN and CLOSED. EOP-1.0 Attachment 6 BOP 5 Place the STM DUMP MODE SELECT Switch in STM PRESS. EOP-1.0 Attachment 6 BOP 6 Place the STM DUMP CNTRL Controller in AUTO. EOP-1.0 Attachment 6

• BOP 7 WHEN the Condenser is NOT available, THEN perform the following: a. Place the Steamline Power Relief A(B)(C) Mode Switches in PWR RLF. b. Adjust the PWR RELIEF A(B)(C) SETPT Controllers as necessary to control RCS temperature. EOP-1.0 Attachment 6 BOP 8 Verify proper response of all Steamline PORVs and Condenser Steam Dumps for existing plant conditions. EOP-1.0 Attachment 6 BOP 9 Ensure SG Blowdown Valves, PVG-503A(B)(C), are closed. EOP-1.0 Attachment 6 BOP 10 If desired, drain valves may be aligned per Shift Supervisor discretion based on current and expected plant status. EOP-1.0 Attachment 6 TURNOVER NOTES (read at the start of the scenario) Turnover Notes Mode 1 // 60% Power // Work Week B1 // 2 Trains VU // EOOS: Yellow (LOSP x 2 Thunderstorms and ASI) // Grid Risk: Red // FEP Risk: Green Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. The National Weather Service has issued a Severe Weather Warning for Richland, Fairfield and Kershaw counties for the next four hours. Maintenance is investigating a GOP-4A is in progress to step 3.16. The power ascension was hal Reactor Engineering will provide an updated Reactivity Plan prior to continuing the up-power. The plant has been at the current power level for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Xenon is stable.

OAP-100.6 ATTACHMENT VIII PAGE 1 OF 2 REVISION 4 CONTROL ROOM SUPERVISOR RELIEF CHECKLIST DATE/TIME: today RELIEF SECTION Turnover Notes Mode 1 // 60% Power // Work Week B1 // 2 Trains VU // EOOS: Yellow (LOSP x 2 Thunderstorms and ASI) // Grid Risk: Red // FEP Risk: Green Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. The National Weather Service has issued a Severe Weather Warning for Richland, Fairfield and Kershaw counties for the next four hours. Maintenance is investigating a GOP-4A is in progress to step 3 Reactor Engineering will provide an updated Reactivity Plan prior to continuing the up-power. The plant has been at the current power level for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Xenon is stable. Offgoing Control Room Supervisor Operations in progress (GOPs, SOPs, load changes, etc.): GOP-4A is in progress to step 3.16 Operations scheduled for oncoming shifts: In Plant safeguard systems in degraded status: Initials In the Control Room, all books are replaced, the desk and console tops are clear, and all trash is properly disposed of. CRS Station Log completed. CRS OAP-100.6 ATTACHMENT VIII PAGE 2 OF 2 REVISION 4 C02 To the best of my knowledge, I am fully qualified to assume this watch taking into consideration fitness for duty, requalification status, and minimum watchstanding qualification. Shift relief completed: Oncoming Control Room Supervisor Offgoing Control Room Supervisor CR Supervisor Shift Supervisor review Oncoming Control Room Supervisor Initials Oncoming watch has reviewed the VCS Switchgear mailbox for switching orders. Plant Status (to be completed prior to turnover): Plant ESF System Status: Component Cooling System Service water System Reactor Building Cooling System Reactor Building Spray System Accumulator Tanks RHR System Charging/Safety Injection System Emergency Feedwater System Accumulator Tanks Diesel Generator Chilled Water System Control Room Ventilation System Position indications, power availability, and annunciator alarms are normal for present plant conditions. Plant Parameters Limit Reactor Power 0-100% RCS Tavg 589.2°F per loop RCS Pressure <2385 psig RCS Flow >100% per loop RCS Subcooling Normal All parameters within allowable limits for plant conditions. If not, what actions are being taken to correct conditions: Review of Logs: Station Log Removal and Restoration Log Tagout Log Special Orders Shift Turnover (to be completed during turnover): Briefing on plant conditions by offgoing Control Room Supervisor. Review of SPDS and BISI displays. Discussion of Protected Equipment. Identification of in-progress procedures including their present status and locations.

OAP-100.6 ATTACHMENT IX PAGE 1 OF 2 REVISION 4 REACTOR OPERATOR RELIEF CHECKLIST DATE/TIME: today LOG SECTION Date Entry RELIEF SECTION Entry Turnover Notes Mode 1 // 60% Power // Work Week B1 // 2 Trains VU // EOOS: Yellow (LOSP x 2 Thunderstorms and ASI) // Grid Risk: Red // FEP Risk: Green Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. The National Weather Service has issued a Severe Weather Warning for Richland, Fairfield and Kershaw counties for the next four hours. Maintenance is investigating a GOP-4A is in progress to step 3.16. The power ascensio Reactor Engineering will provide an updated Reactivity Plan prior to continuing the up-power. The plant has been at the current power level for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Xenon is stable. Offgoing Reactor Operator Initials Main Control Board (Reactor Operator portion) properly aligned for the applicable mode. RO Housekeeping is satisfactory in the Reactor Operator area of responsibility. RO Discussion of Protected Equipment. RO Oncoming Reactor Operator Initials Review of HVAC Panel. Review of Station Log. Review of Removal & Restoration Log. Review of Main Control Board Panels. System Alignment A B C Train aligned to Reasons for any inoperable equipment Service Water Pumps X X A Component Cooling Pumps X A Charging Pumps X A HVAC Chillers X X A Reactor Building Spray Pumps RHR Pumps TDEFP Emergency Feedwater Pumps Inoperable Radiation Monitors OAP-100.6 ATTACHMENT IX PAGE 2 OF 2 REVISION 4 C02 To the best of my knowledge, I am fully qualified to assume this watch taking into consideration fitness for duty, requalification status, and minimum watchstanding qualification. Shift relief completed: Oncoming Reactor Operator Offgoing Reactor Operator Reactor Operator Shift Supervisor review OAP-100.6 ATTACHMENT X PAGE 1 OF 1 REVISION 4 BALANCE OF PLANT RELIEF CHECKLIST DATE/TIME: today Date Entry RELIEF SECTION Entry Turnover Notes Mode 1 // 60% Power // Work Week B1 // 2 Trains VU // EOOS: Yellow (LOSP x 2 Thunderstorms and ASI) // Grid Risk: Red // FEP Risk: Green Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. The National Weather Service has issued a Severe Weather Warning for Richland, Fairfield and Kershaw counties for the next four hours. is OOS for maintenance. Maintenance is investigating a GOP- Reactor Engineering will provide an updated Reactivity Plan prior to continuing the up-power. The plant has been at the current power level for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Xenon is stable. Offgoing Reactor Operator Initials Main Control Board (Reactor Operator portion) properly aligned for the applicable mode. BOP Housekeeping is satisfactory in the Reactor Operator area of responsibility. BOP Discussion of Protected Equipment. BOP Oncoming Reactor Operator Initials Review of Main Control Room Panels. Review of Station Log. Review of Removal & Restoration Log. Test annunciator lights (with Offgoing operator concurrence). C02 To the best of my knowledge, I am fully qualified to assume this watch taking into consideration fitness for duty, requalification status, and minimum watchstanding qualification. Shift relief completed: Oncoming Balance of Plant Offgoing Balance of Plant Balance of Plant Shift Supervisor review OAP-100.6 ATTACHMENT IA PAGE 1 OF 2 REVISION 4 REACTIVITY CONTROL PARAMETERS NOTE This information should be recalculated every Sunday Dayshift (when the plant is in Mode 1) to be available for Reactor Engineering review Monday morning or following work day. RCS Boron Concentration (CRCS) ppm Burnup______________ MWD/MTU (Check BAT in Service) CB "A" BAT ppm CB "B" BAT ppm Moderator Temperature Coefficient (MTC) (Fig. II-3.7, HFP) pcm/°F Differential Boron Worth (DBW) (Fig. II-7.2, HFP) pcm/ppm Gallons of Boric Acid or Reactor Makeup Water required to change RCS average temperature by one (1) degree: MTC/DBW = / = (B) ppm Boron Change/°F gal. Acid/°F = From Fig. III-2: gal. Acid/°F = 49640 ln (()()) gal. RMW/°F = From Fig. III-3: gal. RMW/°F = 49640 ln (() ) Power Defect (PD) for 10% power change (100% to 90%) (Fig. II-2): _______ PD @ 100% RTP - ________ PD @ 90% RTP = Gallons of Boric Acid only to reduce reactor power from 100% to 90%: ________ ppm Boron (Fig. III-2) _________ gal. Boric Acid/10% RTP Final rod height using rods only to reduce reactor power from 100% to 90%: (Assume ARO) (IRW) = ___ pcm (Fig. II-10) final rod height Bank D OAP-100.6 ATTACHMENT IA PAGE 2 OF 2 REVISION 4 REACTIVITY CONTROL PARAMETERS NOTE For a 10% reduction in load, 1/2 of the calculated boric acid should be used and 1/2 the calculated Control Rod motion. For a 100% to 90% load reduction: Use ___________ gallons boric acid (1/2 the gallons calculated above), and expect the rods to be at approximately ___________ steps on bank D (Fig. II-10 series, 1/2 the IRW, NOT 1/2 the steps). To change TAVG by 1o F: ____________ gallons Boric Acid/°F ____________ gallons Reactor Makeup Water/°F For a 100% to 90% load reduction: Use ___________ gallons boric acid and expect ___________ steps on bank D NOTE: This calculation is to provide a second check to the batch integrator setting to establish continuity between the setting and actual make-up results. FCV 113 A&B, pot setting for current RCS boron concentration __________ Expected Boric Acid flowrate for VCT makeup ________

Expected Boric Acid total gallons on an Auto Makeup based on current BAT in service: Current RCS CB _____ x _____ gallons* = _______________ CB for BAT in service

• Normal Auto Makeup is 267 to 275 gallons Calculation and Auto Makeup pot settings by Signature / Date Calculation and Auto Makeup pot settings verified by Signature / Date Reactor Engineering Review _________________________ Date______________

OAP-100.6 ATTACHMENT IB PAGE 1 OF 2 REVISION 4 REACTIVITY MANAGEMENT BRIEF MODES 1 - 3 NOTE PART 1 REACTIVITY MANAGEMENT TURNOVER should be read at Shift Turnover Meeting. PART 2 REACTOR STATUS should be discussed between the NROATC, BOP, and CRS.

PART 1 REACTIVITY MANAGEMENT TURNOVER: Date of last Automatic or Manual Make-Up:

Is Auto Makeup expected this shift (circle)? YES NO Expected Boric Acid total gallons on a normal Auto Makeup based on current BAT in service: gallons FCV 113 A&B, pot setting for current RCS boron concentration:

Expected Boric Acid flowrate for VCT makeup:

Total gallons Diluted Borated (Last Shift)

Last evolution (circle one): Borate / Dilute / Blended Expected Borations, Dilutions, or Blended changes to the RCS: List Reactivity Concerns in progress or planned and action(s) necessary (i.e. Steam or Feed Flow transmitter in test, Steam Generator Blowdown out of service, Calorimetric inputs in service, etc.).

OAP-100.6 ATTACHMENT IB PAGE 2 OF 2 REVISION 4 REACTIVITY MANAGEMENT BRIEF MODES 1 - 3 (Cont'd) PART 2 REACTOR STATUS: (circle one below) Delta I on Target (+ 2%)? YES NO Not in Mode 1 If NO is circled, identify plan to re-establish target band: Xenon Trend: Stable Building In Burning Out Demineralizers: Mixed Bed in service: A B PRC01 Y / N Standby Demineralizer: Filled Borated Empty PRC01 Cation Bed: Date last in service Boron Concentration when in service ATTACHMENT IA reviewed and current: YES NO Midnight Boron Concentration and Date when CHG/SI pump was secured:

CB A Date CB B Date CB C Date REP-102.001 ATTACHMENT I PAGE 1 OF 1 REVISION 6 CYCLE _______ PLAN# _____-______ REACTIVITY MANAGEMENT PLAN VERIFICATION BEACON Filenames: Model Input filename Summary Results filename Calibration filename Power Profile filename Sign and date steps below to document performance.

Step Number Signature Date

• 3.0 Prerequisites *7.36 Verify 9.0 Criteria *7.38 RE Verifier *7.39 Operations Reviewer COMMENTS

REP-102.001 ATTACHMENT II PAGE 1 OF 1 REVISION 6 CYCLE _______ PLAN# _____-______ REACTIVITY MANAGEMENT PLAN INPUTS PROPOSED POWER MANEUVER Reactor Power activities to be performed, holds, etc.)

COMMENTS - list power plateau activities, unusual operational restraints, contingency plans, alternate power history variations to address, time periods to avoid boration, etc. Comments (e.g. control rod or boron issues, Date/Time HoursDTotalTotalRAOCRAOCXenonRILAfterRxBankBoronBoron WaterBoron WaterDelta-IBandBandWorthLimitStartPowerPosPPM(gal)(gal)(gal)(gal)(%)LowHigh(pcm)(steps)0.0060%1731121.60000-2.91-19.217.6-28391120.2561%1741122.68080-2.93-19.017.4-28281140.5061%1751123.580160-2.76-18.817.3-28171150.7562%1751123.500160-2.78-18.617.1-28051171.0063%1751123.500160-2.82-18.416.9-27921181.2564%1751124.580240-2.85-18.216.7-27781201.5064%1751124.500240-2.90-18.016.6-27641211.7565%1761126.1130370-2.94-17.816.4-27501232.0066%1771127.4110480-2.95-17.616.2-27351242.2566%1771128.270550-2.99-17.416.1-27201262.5067%1781129.5100650-3.01-17.215.9-27041272.7568%1781130.480720-3.04-17.015.7-26891293.0069%1791131.7110830-3.05-16.815.5-26731303.2569%1791132.670900-3.08-16.615.4-26581313.5070%1791133.360970-3.12-16.415.2-26421333.7571%1801134.5901060-3.15-16.215.0-26271344.0071%1801135.5801140-3.19-16.014.9-26111364.2572%1811136.61001240-3.21-15.814.7-25961374.5073%1821138.01101350-3.21-15.614.5-25801394.7574%1821138.9701420-3.24-15.414.3-25651405.0074%1821138.9001420-3.28-15.214.2-25511425.2575%1831140.51301560-3.31-15.014.0-25361435.5076%1831141.3701620-3.36-14.813.8-25221445.7576%1841142.3801700-3.39-14.613.7-25081466.0077%1851143.4901800-3.23-14.413.5-24941476.2578%1851143.4001800-3.27-14.213.3-24811496.5079%1851144.3701870-3.32-14.013.1-24681506.7579%1851144.3001870-3.39-13.813.0-24561527.0080%1861145.71201990-3.53-13.612.8-24431537.2581%1871145.7001990-3.57-13.412.6-24321557.5081%1881147.11102100-3.59-13.212.5-24201567.7582%1891147.9702170-3.57-13.012.3-24091588.0083%1891147.9002170-3.60-12.812.1-23981598.2584%1891147.9002170-3.65-12.611.9-23881618.5084%1911149.11002270-3.63-12.411.8-23771628.7585%1911149.1002270-3.65-12.211.6-23681639.0086%1911149.1002270-3.69-12.011.4-23591659.2586%1921149.1002270-3.73-11.811.3-23501669.5087%1921149.1002270-3.78-11.611.1-23421689.7588%1931149.1002270-3.83-11.410.9-233416910.0089%1941149.1002270-3.84-11.210.7-232617110.2589%1941149.1002270-3.88-11.010.6-231917210.5090%1951149.1002270-3.91-10.810.4-231217410.7591%1961149.1002270-3.94-10.610.2-230617511.0091%1961149.1002270-3.98-10.410.1-230017611.2592%1971148.303622736-4.06-10.29.9-229417811.5093%1981148.30022736-4.09-10.09.7-228918011.7594%1991148.30022736-4.05-9.89.5-228318112.0094%1991147.304422780-4.09-9.69.4-227918212.2595%2001147.30022780-4.13-9.49.2-227518412.5096%2011146.6031227110-4.16-9.29.0-2271185Cycle 22 Simulator 10k MWD/MTU Startup 60-100%Tech Spec Ref: N/AProcedure Ref: REP-102.001Figure Ref: N/A HoursDTotalTotalRAOCRAOCXenonRILAfterRxBankBoronBoron WaterBoron WaterDelta-IBandBandWorthLimitStartPowerPosPPM(gal)(gal)(gal)(gal)(%)LowHigh(pcm)(steps)Cycle 22 Simulator 10k MWD/MTU Startup 60-100%12.7596%2021146.600227110-4.16-9.08.9-226618713.0097%2021145.2061227172-4.20-8.88.7-226318813.2598%2031145.200227172-4.24-8.68.5-226019013.5099%2041144.0053227224-4.29-8.48.3-225819113.7599%2061144.000227224-4.41-8.28.2-225519314.00100%2071143.1036227260-4.42-8.08.0-225319414.25100%2081143.100227260-4.32-8.08.0-225219414.50100%2091144.180235260-3.28-8.08.0-225019414.75100%2071143.4031235292-3.42-8.08.0-225119415.00100%2061142.5040235332-3.79-8.08.0-225319415.25100%2061142.500235332-3.82-8.08.0-225519415.50100%2061142.500235332-3.84-8.08.0-225819415.75100%2061141.6039235371-3.87-8.08.0-226119416.00100%2061141.600235371-3.89-8.08.0-2264194Tech Spec Ref: N/AProcedure Ref: REP-102.001Figure Ref: N/A OAP-102.1 ATTACHMENT II PAGE 1 OF 1 REVISION 7 SCHEDULED WORK APPROVAL/DENIAL Scheduled Work/Activity Date Description of Work/Activity to be performed: I. This Moderate Risk, Elevated Risk, High Risk, or Cross Train activity is approved for work provided the required plant conditions are available on the scheduled due date. OR This specific activity has been reviewed for EOOS Risk Reassessment. Set EOOS Environmental Variance __________________________ Set Risk at Times The following items were considered for making this approval: Operations Supervisor (Moderate Risk or Cross Train) In the absence of the Operations Supervisor: Operations Scheduling, Shift Supervisor GMNPO/MDS (Elevated Risk) PSRC (High Risk) II. This work activity/package cannot be performed on the scheduled date due to the following reason(s): SRO (WCC or On Shift) Operations Scheduling Supervisor III. Recommended re-schedule date or plant conditions:

SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION NUCLEAR OPERATIONS NUCLEAR OPERATIONS COPY NO.

GENERAL OPERATING PROCEDURE GOP-4A POWER OPERATION (MODE 1 - ASCENDING) REVISION 2 SAFETY RELATED

RECORD OF CHANGES CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE A P 10/31/11 F P 06/30/14 B P 04/25/12 G P 07/20/14 C P 11/01/12 H P 07/21/14 D P 05/01/14 E P 05/23/14 CONTINUOUS USE Continuous Use of Procedure Required. Read Each Step Prior to Performing.

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GOP-4A PAGE i REVISION 2 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE/SCOPE 1 2.0 INITIAL CONDITIONS 1 3.0 INSTRUCTIONS 5

4.0 REFERENCES

49 ENCLOSURES Enclosure A - Estimated Generator Capability Enclosure B - DA Low Power Temperature Curve ATTACHMENTS Attachment I - Sign-off Identification List Attachment II - Required System Alignment Verification

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GOP-4A REVISION 2 PAGE 1 OF 50 1.0 PURPOSE/SCOPE 1.1 The purpose of this procedure is to provide the steps required to be performed to startup the plant from the Point of Adding Heat to 100% Reactor Power.

1.2 10CFR50 Appendix B, SAP-630, and 10CFR50.59 apply to this procedure.

NOTE 2.0 through 4.0 a. If this procedure must be initiated under conditions other than those in Section 2.0, INITIAL CONDITIONS, the Shift Supervisor or Control Room Supervisor will review Sections 2.0, INITIAL CONDITIONS, and 3.0, INSTRUCTIONS. Steps that are not applicable due to plant conditions will be marked N/A and initialed by the Shift Supervisor or Control Room Supervisor. All other items will require sign-off or check-off. b. All personnel who sign off steps in this procedure must enter their names and initials on Attachment I. c. Each step should be initialed and dated when all its substeps are either completed and checked-off or marked as N/A and initialed. 2.0 INITIAL CONDITIONS INITIALS/DATE 2.1 RCS status is as follows: / a. System temperature is being maintained between 555°F and 559°F using the Steam Dump System or Steamline PORVs. b. System pressure is being maintained between 2220 psig and 2250 psig in AUTO control. c. All Reactor Coolant Pumps are in operation. d. Pressurizer level is being maintained at 25% in AUTO control. 2.2 All Safety Injection Systems are aligned and operable. /

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GOP-4A REVISION 2 PAGE 2 OF 50 INITIALS/DATE 2.3 Excore NIs are aligned for Power Operation per SOP-404, / Excore Nuclear Instrumentation System. 2.4 Reactor Power is being maintained between 1% and 3%. / 2.5 For Mode 2, with no untrippable or dropped Control Rods, / Shutdown Margin requirements are satisfied once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verification of Control Rods above the Rod Insertion Limit. 2.6 Reactor Makeup Control is in AUTO and set for the existing / RCS boron concentration. 2.7 The Rod Control and Position Indicating Systems are in operation / per SOP-403, Rod Control And Position Indicating System. 2.8 Secondary Plant status is as follows: / a. The Main Turbine is on the Turning Gear per SOP-215, Main Turbine Lube Oil Supply System. b. The Main Feedwater Pumps are on their Turning Gears per SOP-209, Feedwater Turbine Lube Oil System, or otherwise rotating via system flow.

c. Narrow Range Steam Generator levels are being maintained between 60% and 65% with chemistry within specification using the following: 1) Steam Generator Blowdown per SOP-212, Steam Generator Blowdown if desired, with Condensate return temperature maintained less than or equal to DA temperature. 2) Emergency Feedwater per SOP-211, Emergency Feedwater System. d. Main Steam heatup is complete per SOP-201, Main Steam System. e. Feedwater is being warmed per SOP-210, Feedwater System.

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GOP-4A REVISION 2 PAGE 3 OF 50 INITIALS/DATE f. Condensate is in operation per SOP-208, Condensate System. g. Circulating Water is in operation per SOP-207, Circulating Water. h. Condenser Vacuum is established per SOP-205, Turbine Sealing Steam System, and SOP-206, Main and Auxiliary Condenser Air Removal System in the following: 1) Main Condenser. 2) Auxiliary Condensers. 2.9 The following controller setpoints are aligned as follows: / a. LCV 3235, DEAER START UP DRAIN CNTRL AUTO with setpoint potentiometer set at 7.1

b. Feedwater Pumps: 1) PUMP A SPEED CONTROL AUTO with setpoint potentiometer set at 0.25. 2) PUMP B SPEED CONTROL AUTO with setpoint potentiometer set at 0.50. 3) PUMP C SPEED CONTROL AUTO with setpoint potentiometer set at 0.75. c. IFK3136, FLOW TO DEAERATOR AUTO with setpoint potentiometer set at 5.0. d. TURB OIL TEMP AUTO with setpoint potentiometer set at 2.0 - 2.66. e. EHC HYDRO OIL AUTO with setpoint potentiometer set at 2.4 - 7.1. f. H2 GAS TEMP AUTO. g. ALT COOLER TEMP AUTO. CHG D

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GOP-4A REVISION 2 PAGE 4 OF 50 INITIALS/DATE NOTE 2.9.h. IPV-2231, MS/PEGGING STM TO DEAERATOR, should be adjusted to maintain DA temperature between 130°F and 150°F. h. IPV-2231, MS/PEGGING STM TO DEAERATOR MAN or AUTO. 2.10 Reactor Engineering has verified the LEFM constants are / removed per SAP-119, Control Of The Station Calorimetric Computer Program. 2.11 Reactor Engineering has provided a Reactivity Management Plan / for the Turbine Startup and power ascension per SAP-0155, Reactivity Management. 2.12 A Pre-job brief has been conducted, including a review of / GOP-Appendix A and the Reactivity Management Plan. 2.13 IPCS is available to monitor Heat Up Rate during Startup when / Moderator Temperature Coefficient is near zero or positive. 2.14 Initiate a work order for Electrical Maintenance to perform / Thermography per steps 3.12.d.2 and 3.16.h. of this procedure.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 5 OF 50 3.0 INSTRUCTIONS INITIALS/DATE NOTE 3.1 through 3.11 Steps 3.1 through 3.11 raise Reactor Power from 1% to 25%. 3.1 Consult the following for current system status, Mode escalation / limitations, and Reactor Power level restrictions: a. Complete Attachment II for required system alignments. b. Removal and Restoration Log. c. Danger Tag Log. d. Reactor Engineering. e. Chemistry. f. Ensure completion of GTP-702 Attachment II.G, Category "C1" Operational Mode Change Plant Startup - Entering Mode 1. Z156 g. Initiate STP-120.003, Emergency Feedwater Valve Verification. (Free of air section 6.1 only.) 3.2 Perform OAP-100.4, Communication, Attachment I, Mode Change Brief Checklist. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 6 OF 50 INITIALS/DATE N013.3 Ensure the following disconnects are properly closed: / a. The associated disconnect switch mechanical operator is locked in position: 1) Manual Disconnect 8891. 2) Manual Disconnect 8893. 3) Manual Disconnect 8901. 4) Manual Disconnect 8903. b. Inform the Electrical Department that disconnects have been closed and initiate a work order to have Electrical perform thermography when disconnect is energized: 1) Manual Disconnect 8891. 2) Manual Disconnect 8893. 3) Manual Disconnect 8901. 4) Manual Disconnect 8903. CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 7 OF 50 INITIALS/DATE 3.4 Align Steam Dump control for Automatic operation as follows: / a. Transfer Steamline PORVs to Automatic operation as follows: 1) Place the Steamline PWR RELIEF A(B)(C) SETPT Controller(s) in MAN. 2) Adjust the PWR RELIEF SETPT Controllers to 8.4 (1092 psig). 3) Place the Steamline Power Relief Mode Switches in AUTO. 4) Place the PWR RELIEF SETPT Controllers in AUTO. b. Transfer STM DUMP CNTRL to Automatic operation as follows: 1) Place the STM DUMP CNTRL Controller in MAN. 2) Adjust the STM DUMP CNTRL setpoint to 8.4 (1092 psig). 3) Place the STM DUMP CNTRL Controller in AUTO. c. If necessary, reset C-7A and C-7B by taking STM DUMP MODE SELECT to RESET and return to STM PRESS.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 8 OF 50 INITIALS/DATE NOTE 3.5 Main Steamline GP-A and GP-B drains should be cycled open at least once every thirty minutes when left closed to minimize cooldown, in order to ensure moisture removal. 3.5 Complete Secondary Plant warm-up as follows: / Z137 a. Secure the Feedwater Recirculation flowpath per SOP-210, Feedwater System,Section III.B., Feedwater System Header Warming, Step 2.3. Z130 b. Perform SOP-204, Extraction Steam, Reheat Steam, Heater Vents and Drains,Section III.A, Start Up Of Extraction Steam, Reheat Steam, Heater Vents and Drains Step 2.1. Z149 c. Complete Main Turbine warm-up per SOP-214, Main Turbine And Controls,Section III.A, Turbine Startup, starting at Step 2.8.q. Z131 d. At the completion of the Main Turbine warm-up, continue with SOP-204, Extraction Steam, Reheat Steam, Heater Vents and Drains,Section III.A, Start Up Of Extraction Steam, Reheat Steam, Heater Vents and Drains, Step 2.2.

e. Maintain DA temperature between 130°F and 150°F as follows: 1) As required, adjust IPV-2231, MS/PEGGING STM TO DEAERATOR, in Automatic or Manual. 2) As required, LCV 3235, DEAER START UP DRAIN CNTRL, may be used to raise flow through the DA. f. Verify Feedwater and Condensate System chemistry is in specification per CP-615, Feedwater And Condensate Chemistry Control. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 9 OF 50 INITIALS/DATE 3.6 Prepare the Secondary Plant for power ascension as follows: / a. Ensure the Main Turbine is reset. Z158 b. Ensure the Main Generator Breaker Disconnect Switch 89 is Closed per SOP-303, Main Generator Breaker And Isophase Bus Duct Cooling,Section IV.B, Closing The Main Generator Breaker Disconnect Switch 89 c. Place 43-TS12, UNDER FREQ. TRIP CONTROL SW, in OFF. d. Direct Electricians to verify proper phase-phase voltages between the Main Generator Breaker and the low side of the Main Transformer as indicated by the G6 set of potential transformer voltage readings in XPN6222 on fuse block 2BU (FU3) (Reference 210-121, Sheet 3 ) are approximately 120 Vac and are balanced: 1) V2 - V4

2) V2 - V6

3) V4 - V6 e. Verify at least two Circulating Water Pumps are operating. f. Verify that at least one Condensate Pump is operating per SOP-208, Condensate System. g. Ensure Sparging Steam to the Deaerator is secured. Z139 h. Ensure all available Feedwater Booster Pumps are operating per SOP-210, Feedwater System, Section III.D, Feedwater Booster Pump Startup. i. Ensure Deaerator temperature is between 130°F and 150°F. CHG A CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 10 OF 50 INITIALS/DATE Step 3.6 continued j. Raise condensate flow to the Blowdown Heat Exchangers to between 150 gpm and 250 gpm per Steam Generator (450-750 gpm total) on FI-3061, CONDENSATE BLOWDOWN COOLERS FLOW IND, using the following controllers in MANUAL (AB-436):

1) ITV-3062A, BD COOLER A CDSTE OUT TEMP.

2) ITV-3062B, BD COOLER B CDSTE OUT TEMP.

3) ITV-3062C, BD COOLER C CDSTE OUT TEMP. 3.7 Align the Feedwater System for power ascension as follows: / C01 a. Perform PTP-102.005, Main Feedwater Pump Turbine Z128 Checks, quarterly portion Steps 6.1 through 6.12.

PMTS#_________

b. Ensure the following are MAN/CLOSED: 1) PVT-478, SG A FWF 2) PVT-488, SG B FWF 3) PVT-498, SG C FWF Z141 c. Start one Main Feedwater Pump per SOP-210, Feedwater System,Section III.E, Feedwater Pump Startup.

d. Reset the Feedwater Isolation signal by momentarily turning the following switches to the right: 1) FW ISOL TRAIN A RESET. 2) FW ISOL TRAIN B RESET. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 11 OF 50 INITIALS/DATE Step 3.7 continued CAUTION 3.7.e 1) Feedwater Header pressure should be maintained on program prior to opening Feedwater Isolation Valves to minimize water hammer. 2) Annunciator Point XCP-625 3-3 (FIV A/B/C ACCUM PRESS LO) should be verified clear or pressure locally verified greater than 500 psi prior to Mode 1 entry to ensure Feedwater Isolation Valve operability. (ref. Tech Spec 3.7.1.6) e. Open the following: 1) PVG-1611A, A ISOL. 2) PVG-1611B, B ISOL. 3) PVG-1611C, C ISOL. NOTE 3.7.f Use MANUAL control only if the Master Speed Controller is unable to control in AUTO. Z140 f. Ensure the MASTER SPEED CNTRL (MCB M/A station) is in Automatic per SOP-210, Feedwater System, Section III.E, Feedwater Pump Startup, Step 2.8. Z197 g. Prepare the Main Generator for startup per SOP-301, MAIN GENERATOR SYSTEM, SECTION III.A, Startup, Steps 2.1 and 2.2. h. Contact Reactor Engineering to verify LEFM constants are removed per SAP-119, Control Of The Station Calorimetric Computer Program. CHG D CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 12 OF 50 INITIALS/DATE CAUTION 3.8 Reactor Power must be maintained less than or equal to 10% until Emergency Feedwater is aligned per STP-120.003, Emergency Feedwater Valve Verification. (refer to Tech Spec 4.7.1.2.a.4) 3.8 Prepare for power ascension as follows: / a. Verify the accumulator pressure for each Feedwater Isolation Valve is greater than 500 psi as indicated by either of the following: 1) XCP-625 3-3 (FIV A/B/C ACCUM PRESS LO) is clear. 2) Accumulator pressure locally verified is greater than 500 psi for each valve. b. Commence Reactor Power increase to between 6% and 9% (Target 8% power, at a reasonably achievable ramp rate up to 1/2%/minute). c. Log the time and date the plant entered Mode 1: Mode 1 Entry: / Time Date GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 13 OF 50 INITIALS/DATE Step 3.8 continued NOTE 3.8.d Maintaining Main Feedwater Pump discharge pressure 50 psi -150 psi greater than Main Steam header pressure will maintain Steam Generator levels until Main Feedwater Pump speed control is placed in Automatic. d. If the MASTER SPEED CNTRL will NOT control in AUTO, perform the following: 1) Place the MASTER SPEED CNTRL in Manual. 2) Adjust Main Feedwater Pump speed as necessary to maintain Main Feedwater Pump discharge pressure 50 psi to 150 psi greater than Main Steam header pressure. Z142 e. Perform SOP-210, Feedwater System,Section III.F, Transferring Emergency Feed Flow To The Main Feed Reg Valves (Preferred method). f. Transfer Feedwater Flow from (Alternate method): Z143 1) Emergency Feed to the Bypass Valves per SOP-210, Feedwater System,Section IV.A, Transferring Feedwater Flow From The Emergency Feed To The Bypass Valves. Z144 2) The Bypass Valves to the Main Feed Reg Valves per SOP-210, Feedwater System,Section IV.B, Transferring Feedwater Flow From The Bypass Valves To The Main Feed Reg Valves. Z140 g. Establish automatic Feedwater Pump speed control per SOP-210, Feedwater System, Feedwater System, Section III.E, Feedwater Pump Startup, Step 2.8. C03 CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 14 OF 50 INITIALS/DATE Step 3.8 continued h. Update the IPCS Plant Mode indicator to indicate Mode 1 as follows: 1) Type the Turn-On-Code MODE to display the PLANT MODE CHANGE DISPLAY window. 2) Select the SET MODE 1 Pushbutton. 3) Verify POWER OPER is displayed on the left end of the top toolbar. CAUTION 3.8.i 1) Moisture Separator/Reheater temperature changes and Main Turbine vibration levels must be monitored closely while placing the MSRs in service. 2) To minimize stress in the Low Pressure Turbines, Hot Reheat Steam temperature changes must be limited to 125F/hr. Z134 i. Start up MSR A and B, in RAMP (TEMP CONTROL) mode, per SOP-204, Extraction Steam, Reheat Steam, Heater Vents And Drains,Section III.D, Normal Startup And Operation Of The MSRs. j. When less than 15% power, ensure the following valves are open: 1) XVT02072A-HD, REHEAT A 4TH-PASS DUMP TO CNDSR THROT. 2) XVT02072B-HD, REHEAT B 4TH-PASS DUMP TO CNDSR THROT. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 15 OF 50 INITIALS/DATE Step 3.8 continued k. Ensure the following as Reactor Power increases to no more than 9%: 1) DA temperature is being maintained between 130°F and 150F by adjusting IPV-2231, MS/PEGGING STM TO DEAERATOR, as necessary. 2) Narrow Range Steam Generator levels are maintained between 55% and 65%. 3) Condensate flow to the Deaerator increases. Z145 l. Secure Emergency Feedwater per SOP-211, Emergency Feedwater System,Section III.B, Motor Driven Emergency Feedwater Pump Shutdown. m. Complete STP-120.003, Emergency Feedwater Valve Verification (Valve Position Verification portion). STTS# ___________ NOTE 3.9 RCS TAVG - TREF DEV HI/LO (XCP-615 2-5) is expected to alarm as TAVG is increased and TREF remains constant. Compensatory actions should be taken per the ARP for this alarm. 3.9 When Emergency Feedwater is aligned for power operation, / prepare to synchronize and load the Main Generator as follows: a. Slowly raise Reactor Power to between 12% and 15% while continuing with this procedure. b. At 10% Reactor Power, perform the following: 1) Verify P10, NIS PR, permissive energizes to bright. 2) Verify P7, REACTOR TRIP BLOCKED, permissive de-energizes to dim. 3) Verify normal Power Range Channel indication. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 16 OF 50 INITIALS/DATE Step 3.9.b continued 4) Monitor the highest indicating Power Range Channel and Delta Flux on either of the following: a) NR-45, NIS RECORDER. b) Computer display NR45. 5) Ensure REGULATOR CORE 1 ALARM and REGULATOR CORE 2 ALARM (XCP-633) are reset.

c. Stabilize Reactor Power to establish and maintain the following conditions prior to and during the Main Turbine rollup to 1800 RPM: 1) Reactor Power between 12% and 15%. 2) Steam Dump Demand between 8% and 14% as indicated on TI-408, SD CNTRL S/G %. 3) Main Steam Header Pressure less than 1120 psig. d. If not completed previously, perform the following per SOP-210, Feedwater System: Z140 1) Establish automatic Feedwater Pump speed control, Section III.E, Feedwater Pump Startup, Step 2.8. Z142 2) Transfer Feedwater from the Main Feed Bypass Valves to the Main Feed Regulating Valves,Section IV.F, Transferring Feedwater Flow From The Bypass Valves To The Main Feed Reg Valves. Z135 e. Transfer Gland Sealing Steam to Main Steam per SOP-205, Turbine Sealing Steam System,Section III.A, Startup Of The Turbine Sealing Steam System Using Main Steam. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 17 OF 50 INITIALS/DATE Step 3.9 continued f. Momentarily place the following RESET-BLOCK Switches in BLOCK: 1) IR TRAIN A. 2) IR TRAIN B. 3) PR LOW SP TRAIN A. 4) PR LOW SP TRAIN B.

g. Verify the following status lights energize to bright: 1) IR A TRIP BLCK. 2) IR B TRIP BLCK. 3) PR A TRIP BLCK. 4) PR B TRIP BLCK. Z147 h. Roll the Main Turbine to 1800 RPM, per SOP-214, Main Turbine And Controls,Section III.A, Turbine Startup, Step 2.13. i. Ensure 0.5 scfh flow through FLOW METER FOR GAS ANALYZER (XPN-7201, HYDROGEN AND STATOR COOLING WTR CNT PNL) by adjusting XVT12205-HY, MACHINE GAS ANALYZER INLET ISOL VALVE (TB-412). j. Obtain a Switching Order from the System Controller. CHG D CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 18 OF 50 INITIALS/DATE CAUTION 3.10 a. Thermal Power changes of greater than 15% in any one hour period require completion of GTP-702 Attachment III.H. b. VCS DDS Report, POWER CHANGE SEARCH, should be periodically performed to ensure a thermal power change of greater than 15% in any one hour period is detected. c. Prolonged operation at low loads (less than 150 MWe) may result in Turbine rubs and elevated bearing vibration caused by low Exhaust Hood temperatures. d. To prevent equipment damage, Step 3.10 should be completed as conditions allow. This is especially true when a Turbine load increase is stopped prior to reaching 150 MWe. NOTE 3.10 through 3.18 a. IFK3136, FLOW TO DEAERATOR, AUTO setpoint should be adjusted during power changes to maintain LI-3136, DEAER STOR TK NR LVL, between 2.5 feet and 5.0 feet as LCV 3235, DEAR START UP DRAIN CNTRL, is closed. b. Acknowledging dialog boxes is considered a "skill of the craft". 3.10 Synchronize and load the Main Generator to as follows: / a. Adjust Reactor Power as necessary to maintain between 8% and 14% Steam Dump Demand as indicated on TI-408, SD CNTRL S/G %, while continuing with this procedure. CHG B CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 19 OF 50 INITIALS/DATE Step 3.10 continued NOTE 3.10.b When the Main Generator Breaker is closed the Generator icon will swap from speed (rpm's) indication to load (MW) indication. Z152 b. Synchronize and load the Main Generator to 50 MWe per SOP-301, Main Generator System,Section III.A, Startup, Step 2.3. c. Monitor Exhaust Hood temperature using any of the following: 1) On the EHC HMI select Monitor/LP Hoods 2) Computer display TURBRG. 3) Computer points T3058A, EXHAUST SPRAY HOOD A TEMP, and T3068A, EXHAUST SPRAY HOOD B TEMP. d. Raise Turbine load to 150 MWe as follows: 1) Verify Exhaust Hood temperature is less than 175°F as indicated on the EHC HMI, Monitor/LP Hoods screen. 2) Using the EHC HMI, Control/Load screen, on Load Set, select Ramp Rate and enter desired rate of 1% or less. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 20 OF 50 INITIALS/DATE Step 3.10 continued 3) Increase Turbine load by one of the following methods: a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 13.59%. (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 21 OF 50 INITIALS/DATE Step 3.10 continued CAUTION 3.10.e Do not stop the Turbine load increase with Exhaust Hood temperature less than 80°F as indicated on the EHC HMI, Monitor/LP Hoods. e. If necessary, stop the Turbine load increase by one of the following methods: 1) Depress the HOLD button. 2) Release the Raise Pushbutton on the MCB. f. Maintain Exhaust Hood temperature greater than 80°F as indicated on the EHC HMI, Monitor/LP Hoods screen, by Turbine load adjustments.

g. If necessary, re-commence the Turbine load increase by one of the following methods: 1) Using the EHC HMI, Control/Load screen, on Load Set, select Ramp Rate and enter desired rate of 1% or less.

2) Manually using the raise/lower pushbuttons: a) Select the desired ramp rate of 1%/min or less. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe). CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 22 OF 50 INITIALS/DATE Step 3.10 continued h. As Turbine load increases, perform the following: 1) Ensure 1st Stage Shell Inner heatup rate as indicated on EHC HMI, Aux/Metal Temps screen does not exceed 150°F/hr. 2) Maintain DA temperature as follows: a) Adjust IPV-2231, MS/PEGGING STM TO DEAERATOR, as necessary, to maintain DA temperature per Enclosure B, DA Low Power Temperature Curve. b) If required, LCV 3235, DEAER START UP DRAIN CNTRL, may be used to raise flow through the DA. Z138 c) Ensure Feedwater Booster Pump warm-up criteria are maintained with DA temperature changes per SOP-210, Feedwater System, Section III.D, Step 2.1. d) Maintain Blowdown Heat Exchanger condensate outlet temperatures at least 30F below DA temperature. i. Adjust Reactor Power as necessary to maintain between 2% and 14% Steam Dump Demand as indicated on TI-408, SD CNTRL S/G % while continuing with this procedure. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 23 OF 50 INITIALS/DATE Step 3.10 continued Z148 j. Close GP-A drain valves per SOP-214, Main Turbine And Controls,Section III.A, Turbine Startup, Step 2.24. k. Place 43-TS12, UNDER FREQ. TRIP CONTROL SW, in ON. l. Reset VOLT. UNBAL. RELAY 60G. Z132 m. Perform the 50 MWe Main Control Board Extraction Drain Switch alignment per SOP-204, Extraction Steam, Reheat Steam, Heater Vents and Drains,Section III.A, Startup Of Extraction Steam, Reheat Steam, Heater Vents And Drains, Step 2.3. n. Perform PTP-102.003, Main Generator Temperature Monitoring. PMTS# __________

o. When the Turbine Load is greater than 10% (100 MWe),

as indicated on any DCS graphic screen or EHC HMI, perform the following: 1) Open MVG-1212, EXTR STM TO DEAER ISOL. 2) Adjust the IPV-2231, MS/PEGGING STM TO DEAERATOR, setpoint to a setting of 7.0 in AUTO.

3) Ensure the following are closed (TB-412): a) XVG02075-HD, HP FW HEATER 2A DRAIN TO DEAER HDR ISOLATION. b) XVG02074-HD, HP FW HEATER 2B DRAIN TO DEAER HDR ISOLATION. 4) Place the following in NORMAL (GRAPHIC 101 and 102 or 110 screens - I icons): a) FW HTR 2A OPRTR SELECT ISOLATION. b) FW HTR 2B OPRTR SELECT ISOLATION. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 24 OF 50 INITIALS/DATE Step 3.10 continued p. Verify P13, 1st STAGE PRESSURE, permissive de-energizes to dim. q. When Turbine Load has stabilized at 150 MWe, perform the following: 1) Close the following Main Steam GP-B drain valves: a) MVG-2899A, X-AROUND DRN VLV-1. b) MVG-2899B, X-AROUND DRN VLV-2. c) MVG-2899C, X-AROUND DRN VLV-3. d) MVG-2899D, X-AROUND DRN VLV-4. Z133 2) Close heater startup vents and bypass valves for 15% Turbine Load, SOP-204, Extraction Steam, Reheat Steam, Heater Vents and Drains,Section III.A, Startup Of Extraction Steam, Reheat Steam, Heater Vents And Drains, Step 2.4.

3) Perform the following (TB-412): a) Throttle XVT02072A-HD, REHEAT A 4TH-PASS DUMP TO CNDSR THROT, to 2.0 turns open. b) Throttle XVT02072B-HD, REHEAT B 4TH-PASS DUMP TO CNDSR THROT, to 3.25 turns open.

4) If desired, secure the Auxiliary Boiler per SOP-506, Auxiliary Boiler Operation: Z153 a)Section III.B, Shutdown Z154 b)Section IV.E, Operation Of The Temporary Boiler. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 25 OF 50 INITIALS/DATE Step 3.10.q continued 5) Perform PTP-102.003, Main Generator Temperature Monitoring. PMTS# Z155 6) Align Extraction Steam to Auxiliary Building Steam per SOP-507, Auxiliary Steam System,Section III.A, Supplying Auxiliary Building Steam From Extraction Steam NOTE 3.10.q.7) If returning from a power level of greater than 75 %, per Reactor Engineering, the LEFM constants are not required to be adjusted (i.e. quarterly valve testing). 7) Contact Reactor Engineering to determine if LEFM constants need to be re-determined for current power history. 3.11 Raise Reactor Power to 25%, as follows:

a. Maintain DA temperature during load increases as follows: 1) Adjust IPV-2231, MS/PEGGING STM TO DEAERATOR, as necessary, to maintain DA temperature per Enclosure B, DA Low Power Temperature Curve. Z138 2) Ensure Feedwater Booster Pump warm-up criteria are maintained with DA temperature changes per SOP-210, Feedwater System,Section III.D, Feedwater Booster Pump Startup, Step 2.1. 3) Maintain Blowdown Heat Exchanger condensate outlet temperatures at least 30F below DA temperature. b. Ensure the A(B)(C) FPT SETPOINT RAMP LIMIT (A icon) RV value for operating Feedwater Pump A(B)(C) is set to 3000 rpm per minute. CHG D CHG D CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 26 OF 50 INITIALS/DATE Step 3.11 continued c. Verify Exhaust Hood temperature is less than 175°F as indicated on the EHC HMI, select Monitor/LP Hoods. d. Using the EHC HMI, Control/Load screen, on Load Set, select Ramp Rate and enter desired rate of 1% or less. e. Increase Turbine load to 300 MWe by one of the following methods: 1) Slowly Raise Turbine load automatically as follows (preferred method): (a) Select the Load pushbutton (a dialog box opens). (b) Enter 27.17%. (c) Confirm setpoint. (d) Verify proper plant response. 2) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2%

(20 MWe) (utilizes previously selected ramp rate) 3) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 27 OF 50 INITIALS/DATE Step 3.11.f continued f. As Turbine load increases, perform the following: 1) Ensure First Stage Shell heatup rate as indicated on EHC HMI, Aux/Metal Temps screen is maintained less than 150°F/hr. 2) Adjust Reactor Power as necessary to maintain between 2% and 14% Steam Dump Demand as indicated on TI-408, SD CNTRL S/G %.

3) When C5 (15% Turbine Load), 1st STAGE PRESSURE, permissive de-energizes to dim, perform the following in the order listed: a) Hold Reactor Power constant. NOTE 3.11.f.3)b) Establishing approximately a 0°F/hr to 30°F/hr cooldown rate will allow Tavg to slowly approach Tref.) b) Continue raising turbine load to match Tavg and Tref. c) Verify Tref is within 1°F of Tavg. d) Place the ROD CNTRL BANK SEL Switch in AUTO. e) Ensure the STM DUMP CNTRL auto setpoint is set to 8.4 (1092 psig). f) Transfer Steam Dump control to the Tavg mode as follows by placing the STM DUMP MODE SELECT Switch in TAVG.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 28 OF 50 INITIALS/DATE Step 3.11. continued g. When Turbine load has stabilized, perform the following: 1) Perform STP-102.002, NIS Power Range Heat Balance. STTS# C02 2) As a second check on Nuclear Instrumentation, compare RCS Loop T to the results of STP-102.002. 3) Perform PTP-102.003, Main Generator Temperature Monitoring. PMTS#

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 29 OF 50 INITIALS/DATE Step 3.11 continued h. Raise Turbine load to attain 25% Reactor Power as follows: 1) Select Ramp Rate and enter desired rate of 1% or less. 2) Raise Turbine Load by one of the following methods:

a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 18.2% (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 30 OF 50 INITIALS/DATE Step 3.11 continued i. At 25% Reactor Power, verify the following status lights energize to bright: 1) CHAN I IR FLUX HI. 2) CHAN II IR FLUX HI. 3) CHAN I PR FLUX LO SET PT. 4) CHAN II PR FLUX LO SET PT. 5) CHAN III PR FLUX LO SET PT. 6) CHAN IV PR FLUX LO SET PT. NOTE 3.12 and 3.13 Steps 3.12 and 3.13 raise Reactor Power from 25% to 48%. 3.12 Raise Reactor Power to 38% as follows: / a. Contact Chemistry to verify there are no 30% power Chemistry holds.

b. Ensure the following are closed (TB-412): 1) XVG02075-HD, HP FW HEATER 2A DRAIN TO DEAER HDR ISOLATION. 2) XVG02074-HD, HP FW HEATER 2B DRAIN TO DEAER HDR ISOLATION.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 31 OF 50 INITIALS/DATE Step 3.12 continued c. Raise Turbine load to attain 38% Reactor Power as follows: 1) Select Ramp Rate and enter desired rate of 1% or less. 2) Raise Turbine Load by one of the following methods:

a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 29.9% (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.)

d. At 250 MWe perform the following: 1) Ensure all Extraction Drain Valves are latched. 2) Contact Electrical Maintenance to perform thermography on manual disconnects 8901 and 8903. CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 32 OF 50 INITIALS/DATE Step 3.12 continued e. At 300 MWe, perform the following to start filling the drain lines from the 2A and 2B Heaters to the DA: 1) Open XVT12083-HD, 1" BYPASS VALVE FOR XVG-02075 (TB-412) (requires ladder). 2) Open XVT12085-HD, 1" BYPASS VALVE FOR XVG-02074 (TB-412). 3) Throttle XVT02018A-HD, FW HTR 2A DRN TO DEAER LVL CONT VLV BYP, ten turns off the closed seat (TB-463). 4) Throttle XVT02018B-HD, FW HTR 2B DRN TO DEAER LVL CONT VLV BYP, ten turns off the closed seat (TB-463). Z136 f. Place a second Condensate Pump in service per SOP-208, Condensate System,Section III.B, Condensate Pump Startup when total Condensate flow approaches 9000 gpm as indicated on the following:

1) FI 3026, PUMP A DISCH FLOW.

2) FI 3036, PUMP B DISCH FLOW.

3) FI 3046, PUMP C DISCH FLOW. g. Between 30% and 35% Reactor Power, perform the following: Z138 1) Ensure Feedwater Booster Pump warm-up criteria are maintained with DA temperature changes per SOP-210, Feedwater System, Section III.D, Feedwater Booster Pump Startup, Step 2.1. Z139 2) Ensure at least three Feedwater Booster Pumps are in service per SOP-210, Feedwater System, Section III.D, Feedwater Booster Pump Startup. CHG D CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 33 OF 50 INITIALS/DATE Step 3.12 continued h. At 38% Reactor Power verify P8, REACTOR TRIP BLOCKED, de-energizes to dim. Z146 i. Establish automatic Steam Generator blowdown temperature control per SOP-212, Steam Generator Blowdown,Section III.A, Steam Generator Blowdown System Startup And Operation Steps 2.19 and 2.20. Z141 j. Between 35% and 48% Reactor Power, place a second Main Feedwater Pump in service per, SOP-210, Feedwater System,Section III.E, Feedwater Pump Startup. CHG D CHG E GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 34 OF 50 INITIALS/DATE Step 3.12 continued k. When 40% is indicated in the INTERMEDIATE PRESSURE block on the PLU test Screen perform the following: 1) On the EHC HMI, Control/Load Screen, select HOLD.

2) Perform the Power Load Unbalance (PLU) test as follows: (a) Verify P9, REACTOR TRIP BLOCKED, permissive is BRIGHT. (b) On the EHC HMI, select Tests/PLU test. (c) Verify all PLU Status indicate OFF. (d) Select PLU Test ON. (e) Select OK.

(f) Verify the following: (1) Test initiation on all 6 status indicators. (2) Final status indication for all 6 indicators PLU TEST FOR R/S/T COMPLETE. (g) Select PLU Test OFF (h) Select OK. (i) Select desired Ramp Rate, %/min increase on Control/Load screen. l. Determine the GOP Appendix A recommended power ascension rate. CHG G CHG H GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 35 OF 50 INITIALS/DATE 3.13 Raise Reactor Power to 48% at the GOP Appendix A / recommended power ascension rate, as follows: a. Verify Steam Generator chemistry is in specification per CP-613, Steam Generator Chemistry Control. b. Raise Turbine load to attain 48% Reactor Power as follows: 1) Select Ramp Rate and enter the recommended Load Ramp Rate.

2) Raise Turbine Load by one of the following methods: a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 39.5.%. (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 36 OF 50 INITIALS/DATE Step 3.13 continued c. When Turbine Load is greater than 40% (385 MWe), verify C20, 1st STG PRESS, de-energizes to dim. d. Monitor the following for proper operation: 1) Stator Water Cooling. 2) Hydrogen Seal Oil. e. Between 400 MWe and 450 MWe, open the following valves to align the 2A and 2B Heaters to the DA (TB-412): 1) XVG02075-HD, HP FW HEATER 2A DRAIN TO DEAER HDR ISOLATION. 2) XVG02074-HD, HP FW HEATER 2B DRAIN TO DEAER HDR ISOLATION. Z129 f. Secure the Condensate Polishing per SOP-203, Condensate Polishing System,Section III.F, Removing The Condensate Polishing System From Service. Z157 g. Maintain the following SOP-401, Reactor Protection And Control System,Section III.B, Load Variations With Manual Or Automatic Reactor Control parameters using Z017 SOP-106, Reactor Makeup Water System Section III.E, Z003 Alternate Dilute Operations or Section III.F, Borate Operations: 1) I within limits. 2) Control Rods above the Rod Insertion Limit. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 37 OF 50 INITIALS/DATE Step 3.13 continued NOTE 3.13.h Above 40% Turbine load MSR 4th pass drains to the Condenser close and the MSR 4th pass drains to the #1 Feedwater Heaters open. h. Using the DCS Computer Graphic Screens 101 and 102, verify MSR 4th pass drain valves have repositioned: 1) XVT-2071A indicates closed. 2) XVT-2071B indicates closed. 3) XVT-02068A indicates open. 4) XVT-02068B indicates open. i. When Reactor Power is stable at or below 48%,

perform the following: 1) STP-102.002, NIS Power Range Heat Balance. STTS# C02 2) As a second check on Nuclear Instrumentation, compare RCS Loop T to the results of STP-102.002.

3) Determine the operability of the Axial Flux Difference alarm: a) Perform STP-133.001, Axial Flux Difference Calculation. STTS# b) Verify Annunciator Point XCP-620 2-4 (CMPTR FLUX LMT EXCEEDS) is clear. 4) PTP-102.003, Main Generator Temperature Monitoring. PMTS# CHG C GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 38 OF 50 NOTE 3.14 through 3.16 Steps 3.14 through 3.16 raise Reactor Power from 48% to 90%. 3.14 At the GOP Appendix A recommended power ascension rate, / continue the Reactor Power ascension above 50% as follows: a. Ensure I, Axial Flux Difference, is within limits per V.C.Summer Curve Book, Figure I-4.1 prior to increasing Reactor Power above 50% per Tech Spec 3.2.1. b. When greater than 50% Rated Thermal Power perform STP-108.001, Quadrant Power Tilt Ratio. STTS#

c. Raise Turbine load to attain 90% Reactor Power as follows: 1) Select Ramp Rate and enter the recommended. Load Ramp Rate.

2) Raise Turbine Load by one of the following methods: a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 81.52% (3) Confirm setpoint. (4) Verify proper plant response. b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.) CHG B CHG C GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 39 OF 50 INITIALS/DATE Step 3.14 continued d. As power increases, verify the following annunciators clear: 1) PR UP DET FLUX HI DEV AUTO DEFEAT (XCP-620 1-5). 2) PR LOW DET FLUX HI DEV AUTO DEFEAT (XCP-620 1-6). e. Adjust Megavars using GEN FIELD VOLT ADJ as requested by the System Controller and within the Estimated Generator Capability Curve (Enclosure A). Z157 f. Maintain the following SOP-401, Reactor Protection And Control System,Section III.B, Load Variations With Manual Or Automatic Reactor Control parameters using Z017 SOP-106, Reactor Makeup Water System Section III.E, Z003 Alternate Dilute Operations or Section III.F, Borate Operations: 1) I within limits. 2) Control Rods above the Rod Insertion Limit. g. Verify P9, REACTOR TRIP BLOCKED, permissive de-energizes to dim. CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 40 OF 50 INITIALS/DATE 3.15 When Turbine Load is greater than 50%, as indicated on any / DCS graphic screen, place IZY03783A and IZY03793A, MANUAL OPERATOR STATIONs in Automatic as follows (GRAPHIC 101 and 102 or 110 screens): a. Ensure the following are open (TB-412): 1) XVG02075-HD, HP FW HEATER 2A DRAIN TO DEAER HDR ISOLATION. 2) XVG02074-HD, HP FW HEATER 2B DRAIN TO DEAER HDR ISOLATION. b. Ensure HTR 2A&B DRN VLV TO DEAERATOR CLOSE is selected to CNTRL (C icon). c. Ensure both FW HTR 2A OPRTR SELECT ISOLATION and FW HTR 2B OPRTR SELECT ISOLATION are in NORMAL (I icon). NOTE 3.15.d The #2 Heaters utilize the lowest controller signal to maintain level, regardless of the M/A station output (in Manual or Automatic). As a result, manipulation of the M/A station in manual may not yield the anticipated result. d. Slowly raise OUT on both ILY03783A and ILY03793A MANUAL OPERATOR STATIONs (M/A - icons) until the associated IZY03783B and IZY03793B (2A/2B Heater drain to the Condenser) are fully closed. e. When FEEDWATER HEATER 2A(B) levels are stable and slightly below setpoint, place ILY03783A and ILY03793A MANUAL OPERATOR STATIONs (M/A - icons) in Automatic. f. Adjust IFK3136, FLOW TO DEAERATOR, AUTO setpoint as necessary to ensure LI-3136, DEAER STOR TK NR LVL is between 2.5 feet and 5.0 feet with LCV 3235, DEAR START UP DRAIN CNTRL closed.

GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 41 OF 50 INITIALS/DATE Step3.15 continued g. Close the following bypass valves: 1) XVT12083-HD, 1" BYPASS VALVE FOR XVG-02075 (TB-412) (requires ladder). 2) XVT12085-HD, 1" BYPASS VALVE FOR XVG-02074 (TB-412). 3) XVT02018A-HD, FW HTR 2A DRN TO DEAER LVL CONT VLV BYP (TB-463). 4) XVT02018B-HD, FW HTR 2B DRN TO DEAER LVL CONT VLV BYP (TB-463). 3.16 As the Reactor Power ascension to 90% continues perform / the following:

a. Between 60% and 65% Reactor Power, perform the following: Z138 1) Ensure Feedwater Booster Pump warm-up criteria are maintained with DA temperature changes per SOP-210, Feedwater System,Section III.D Feedwater Booster Pump Startup, Step 2.1. Z139 2) Ensure four Feedwater Booster Pumps are in service per SOP-210, Feedwater System, Section III.D, Feedwater Booster Pump Startup. Z141 3) Start a third Main Feedwater Pump per SOP-210. Feedwater System,Section III.E, Feedwater Pump Startup. b. At 65% Reactor Power, perform PTP-102.003, Main Generator Temperature Monitoring. PMTS# CHG D CHG D CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 42 OF 50 INITIALS/DATE Step 3.16 continued c. When Reactor Power is between 60% and 75%, perform PTP-102.001, Main Turbine Tests (Power Operated Extraction System Check Valve portion only). PMTS# d. At 75% Reactor Power, perform STP-102.002, NIS Power Range Heat Balance. STTS# e. At 80% Reactor Power, align Control Valve drain valves as follows:

1) Ensure PVG-2898B, DV-4, is open as follows: a) Verify Control Valve #4 is closed. b) Verify PVG-2898B, DV-4, is open. c) If both PVG-2898B, DV-4, and Control Valve #4 are closed, open PVG-2898B, DV-4, by opening MVG-2898D, STM LEAD DRN FOR CV-1. 2) Open MVG-2897, COMB CNTRL VLV BSD. f. When Control Valve #4 indicates greater than 5% open, perform the following: 1) Ensure PVG-2898B, DV-4, is CLOSED. 2) Close MVG-2897, COMB CNTRL VLV BSD. g. At 85% Reactor Power, perform PTP-102.003, Main Generator Temperature Monitoring. PMTS# ___________ CHG G GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 43 OF 50 INITIALS/DATE Step 3.16 continued h. Contact Electrical Maintenance to perform thermography on manual disconnects 8901 and 8903. i. Adjust Megavars using GEN FIELD VOLT ADJ as requested by the System Controller and within the Estimated Generator Capability Curve (Enclosure A). j. If desired stabilize Reactor Power at 90%, otherwise proceed to Step 3.17. NOTE 3.17 and 3.18 Steps 3.17 and 3.18 raise Reactor Power from 90% to 100%. 3.17 At the GOP Appendix A recommended power ascension rate, / increase Reactor Power from 90% to 95% as follows: a. Raise Turbine load to attain 95% Reactor Power. 1) Select Ramp Rate and enter the recommended. Load Ramp Rate.

2) Raise Turbine Load by one of the following methods:

a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 87.41%. (3) Confirm setpoint. (4) Verify proper plant response. CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 44 OF 50 INITIALS/DATE Step 3.17.a.2) continued b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.)

b. Place the following Blowdown Temperature Controllers in AUTOMATIC (AB-436): 1) ITV-3062A, BD COOLER A CDSTE OUT TEMP. 2) ITV-3062B, BD COOLER B CDSTE OUT TEMP. 3) ITV-3062C, BD COOLER C CDSTE OUT TEMP. c. Adjust Megavars using GEN FIELD VOLT ADJ as requested by the System Controller and within the Estimated Generator Capability Curve (Enclosure A). Z157 d. Maintain the following SOP-401, Reactor Protection And Control System,Section III.B, Load Variations With Manual Or Automatic Reactor Control parameters using Z017 SOP-106, Reactor Makeup Water System Section III.E, Z003 Alternate Dilute Operations or Section III.F, Borate Operations: 1) I within limits. 2) Control Rods above the Rod Insertion Limit. CHG B CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 45 OF 50 INITIALS/DATE Step 3.17 continued e. Stabilize Reactor Power at 95% and perform the following: 1) STP-102.002, NIS Power Range Heat Balance. STTS# ___________ 2) During the first power ascension following refueling, contact Reactor Engineering to continue the power ascension per REP-107.001, Controlling Procedure For Refueling Startup And Power Ascension Testing. 3.18 Slowly increase Reactor Power to 100% as follows: /

a. If the IPCS is available, verify the NSSS CRT is displaying the following computer points: 1) SHIFT AVG POWER (U9002). 2) QCORE 1 (C1M) (U9003). b. Raise Turbine load to attain 100% Reactor Power per the GOP Appendix A recommended power ascension rate, while continuing with this procedure. 1) Select Ramp Rate and enter the recommended. Load Ramp Rate.

2) Raise Turbine Load by one of the following methods: a) Slowly Raise Turbine load automatically as follows (preferred method): (1) Select the Load pushbutton (a dialog box opens). (2) Enter 92.12%. (3) Confirm setpoint. (4) Verify proper plant response. CHG B GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 46 OF 50 INITIALS/DATE Step 3.18.b.2) continued b) Manually by pushing and holding the Raise Pushbutton on the MCB to increase Turbine load in increments of less than or equal to 2% (20 MWe) (utilizes previously selected ramp rate) c) Under Manual Adj momentarily select Raise to increase Turbine load in increments of 0.1-0.2% to a total of 2% (20 MWe). (one cycle utilizes 10%/min ramp rate and returns to previously selected ramp rate.)

c. Adjust Turbine load to attain 100% Reactor Power, while continuing with this procedure. 1) Using the EHC HMI, Control/Load screen, on Load Set, select Ramp Rate and enter desired rate of 1% or less.

2) Adjust Turbine Load as follows: a) Select the Load pushbutton (a dialog box opens). b) Adjust the setpoint in incremental values not to exceed 0.2%. c) Confirm setpoint. d) Verify proper plant response. d. Adjust Megavars using GEN FIELD VOLT ADJ as requested by the System Controller and within the Estimated Generator Capability Curve (Enclosure A). CHG B CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 47 OF 50 INITIALS/DATE Step 3.18 continued e. Monitor the following for proper operation: Z099 1) Stator Water per SOP-218, Stator Cooling Water System,Section III.A Z150 2) Hydrogen Seal Oil per SOP-216, Seal Oil System, Section III.A. Z151 3) Generator Gas per SOP-217, Generator Gas And Vent System,Section III.A f. Stabilize at 100% Reactor Power and perform STP-102.002, NIS Power Range Heat Balance. STTS# ___________ g. If desired, place the Load Limit circuit in service as follows: 1. Select desired Ramp Rate on Load Limit. (usually Normal). 2. Select Setpoint on Load Limit, (a dialog box opens). 3. Enter the desired setpoint (must be less than the indicated Load Reference). 4. Confirm setpoint. 5. Verify the Load Limit status indicates LIMITING. 6. Verify proper system response. CHG B CHG B CHG D GOP- 4A REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. Axial Flux Difference, I, should be maintained within limits per V.C. Summer Curve Book, Figure I-4.1 during Reactor Power Operation above 50% per Tech Spec 3.2.1. C. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. D. If time allows, all load changes should be discussed with the System Controller prior to commencing the load change. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient: 1) Power and temperature changes should be slow and will require constant operator attention. 2) Tavg should be maintained within 0.5F of Tref unless Tavg is being increased in preparation for Turbine startup. 3) All power and load changes should be performed in small increments. B. Reactor Power increases should be made in accordance with the guidelines established in GOP Appendix A. The recommended rate of power increase is 1/2% per minute and need not be continuous. C. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. TURBINE CONTROL A. If during load changes, plant stabilization is required, under the Turbine HMI: Control/Load screen, select HOLD. B. To resume power ascension select the recommended Load Ramp Rate. C. Turbine Load values are approximate and provided as initial starting points for load changes. When desired Reactor or Turbine parameters are achieved stabilize (if necessary) and proceed as directed. MSR CONTROL A. Do not exceed 50F T between the inlets to the Low Pressure Turbine. B. When in Manual, do not exceed 25F per half-hour temperature change rate for the tube side of the Moisture Separator/Reheater. CHG B GOP-4A REVISION 2 PAGE 48 OF 50 INITIALS/DATE Step 3.18 continued NOTE 3.18.h f returning from a power level of greater than 75 %, per Reactor Engineering, the LEFM constants are not required to be adjusted (i.e. quarterly valve testing). h. Contact Reactor Engineering to determine if LEFM constants need to be re-determined for current power history. i. Adjust Reactor Power to 100% Rated Thermal Power, and perform the following: 1) STP-102.002, NIS Power Range Heat Balance. STTS# ___________ 2) PTP-102.003, Main Generator Temperature Monitoring. PMTS# ___________ j. Maintain operation as close to 100% of licensed core power (2900 MWt) as possible, per OAP-100.6, Control Room Conduct and Control of Shift Activities. k. Notify Reactor Engineering to evaluate the requirements for performing STP-201.001, Monthly Reactor Engineering Surveillances. NOTE 3.18.l For purposes of record, this procedure is complete when all steps through 3.18.l are initialed and dated. It should then be routed to the Operations Supervisor. l. 100% Reactor Power achieved: 1) Date / / 2) Time CHG C

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GOP-4A REVISION 2 PAGE 49 OF 50

4.0 REFERENCES

4.1 CP-613, Steam Generator Chemistry Control.

4.2 CP-615, Feedwater And Condensate Chemistry Control.

4.3 ES560.120, Feedwater Flow Rate And Temperature Normalization Surveillance.

4.4 FSAR Section 5.0.

4.5 GOP-Appendix A. C03 4.6 LER 97002. N01 4.7 MRB 9501. 4.8 OAP-100.4, Communication.

4.9 PTP-102.002, Main Turbine Monthly Oil System Test.

4.10 PTP-102.003, Main Generator Temperature Monitoring.

4.11 PTP-102.005, Main Feedwater Pump Turbine Checks.

4.12 PTP-102.008, Main Turbine Overspeed Testing.

4.13 SAP-119, Control Of The Station Calorimetric Computer Program. C01 4.14 SER 880024.

C02 4.15 SOER 90-3. 4.16 SOP-102, Chemical And Volume Control System.

4.17 SOP-106, Reactor Makeup Water System.

4.18 SOP-201, Main Steam System.

4.19 SOP-203, Condensate Polishing System.

4.20 SOP-204, Extraction Steam, Reheat Steam, Heater Vents And Drains.

4.21 SOP-205, Turbine Sealing Steam System.

4.22 SOP-206, Main and Auxiliary Condenser Air Removal System.

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GOP-4A REVISION 2 PAGE 50 OF 50 4.23 SOP-207, Circulating Water. 4.24 SOP-208, Condensate System. 4.25 SOP-209, Feedwater Turbine Lube Oil System. 4.26 SOP-210, Feedwater System. 4.27 SOP-211, Emergency Feedwater System. 4.28 SOP-212, Steam Generator Blowdown. 4.29 SOP-214, Main Turbine And Controls. 4.30 SOP-215, Main Turbine Lube Oil Supply System. 4.31 SOP-216, Seal Oil System. 4.32 SOP-217, Generator Gas And Vent System. 4.33 SOP-218, Stator Cooling Water System. 4.34 SOP-301, Main Generator System. 4.35 SOP-403, Rod Control And Position Indicating System. 4.36 SOP-404, Excore Nuclear Instrumentation System. 4.37 SOP-506, Auxiliary Boiler Operation. 4.38 SOP-507, Auxiliary Steam System. 4.39 STP-102.002, Nis Power Range Heat Balance. 4.40 STP-108.001, Quadrant Power Tilt Ratio.

4.41 STP-120.003, Emergency Feedwater Valve Verification. 4.42 STP-133.001, Axial Flux Difference Calculation. 4.43 STP-134.001, Shutdown Margin Verification. 4.44 STP-201.001, Monthly Reactor Engineering Surveillances. 4.45 V.C. Summer Precautions, Limitations, and Setpoints.

4.46 V.C. Summer Tech Specs.

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GOP-4A ENCLOSURE A PAGE 1 OF 1 REVISION 2 ESTIMATED GENERATOR CAPABILITY -1000100200300400 500010020030040050060070080090010001100MEGAWATTSMEGAVARSBBAACCCurve AB Limited By Field HeatingCurve BC Limited By Armature HeatingLAGLEAD45 PSIG60 PSIGMVAR LIMIT+325ADMIN LIMIT ONLYCONTACT SYSTEM CONTROLLER IF EXCEEDED 484.95 P.F..90 P.F.UEL SETPOINTAT 20.9 KV UEL SETPOINT AT 22.0 KV

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GOP-4A ENCLOSURE B PAGE 1 OF 1 REVISION 2 DA Low Power Temperature Curve100120140160180200220240020406080100120140MegaWatts ElectricDA Temperature, Deg F150130MINIMUM TEMPERATURE FOR PRESSURIZATIONDesired Operating BandCAUTIONDA operation in the regionabove this curve may result inwaterhammer.

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GOP-4A ATTACHMENT I PAGE 1 OF 1 REVISION 2 SIGN-OFF IDENTIFICATION LIST PERSONNEL NAME (PRINTED) PERSONNEL NAME (SIGNATURE) PERSONNEL INITIALS

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GOP-4A ATTACHMENT II PAGE 1 OF 1 REVISION 2 REQUIRED SYSTEM ALIGNMENT VERIFICATION PROCEDURE NUMBER PROCEDURE TITLE Date of Last Alignment Has the System been 00S > 14 days Has the System undergone Significant Maintenance Does the System Require a Complete new Alignment Date of Record for this Procedure Verification by Shift Supervisor Yes No Yes No Yes No Initials/Date GOP-2 Required Systems Alignments Current and Completed NA NA NA NA NA NA NA / LIST OTHERS REQUIRED / / / / /

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Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 3 NUREG -1021 R9 S1 Facility: VC SUMMER Scenario No: 3 Op Test No: NRC-ILO-13-01 Examiners: Operators: CRS: RO: BOP: Initial Conditions:

• 100% MOL.

• B Train Work Week.

• Alternate Seal Injection is OOS.

• Thermography of transformer disconnects is in progress in the switchyard. Turnover:

• Maintain 100% power. Critical Tasks:

• Start Back-up EHC Pump prior to a Rx/Turbine Trip on Low EHC Pressure.

• t a Rx/Turbine Trip on SG Level.

• Manually trip the Reactor prior to completion of Immediate Actions of EOP-1.0 (E-0).

• Establish feed flow to at least one SG before RCS feed and bleed criteria is met. Event No. Malf No. Event Type* Event Description 1 TUR012A I-RO, CRS TS-CRS PT-446 (Turbine First Stage Pressure) fails LOW. (Rods Drive In) 2 NA N-BOP, CRS R-RO Rapid Power Reduction due to overheating of main generator disconnects. 3 EH001T EH002F C-BOP Running EHC Pump Trip. (Standby EHC must be manually started). 4 MS020O I-BOP, CRS TS-CRS LT-496 ails HIGH. (Manually control feedwater to C 5 CVC005C C-RO, CRS Progressive failure of #2 Seal 6 RCS003A PCS008A PCS008B FW025P C-RO, CRS RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip)

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 3 NUREG -1021 R9 S1 7 MSS015 EF001S EF002T M-ALL Loss Of Heat Sink (EFW) After Reactor Trip. FWM001A FWM001B FWM001C All Main Feedwater Pumps Trip. (Feed with Condensate) * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 3 NUREG -1021 R9 S1 The following notation is used in the ES-D- IOA designates Immediate Operator Action steps.

• designates Continuous Action steps. The crew will assume the watch having been pre-briefed on the Initial Conditions, the plan for this shift and any related operating procedures. EVENT 1: PT-446 (Turbine First Stage Pressure) fails LOW. (Rods Drive In) TRIGGER 1 o MAL-TUR012A TURBINE IMPULSE PRESSURE TRANSMITTER PT-446 FAILURE FINAL=0 TRIGGER 2 o ANN-MI007 AMSAC GENERAL WARNING Fail To: ON On cue from the Examiner, the selected turbine first stage pressure transmitter (PT-446) will fail LOW. The failure causes a Tave -Tref mismatch resulting in rods inserting at the maximum speed. The crew will enter AOP-401.7, Turbine First Stage Pressure Channel Failure. The RO will respond to the rod insertion by placing rod control in manual and restoring Tave to within 1 degree of Tref. The crew will then select the operable 1st stage pressure channel for control. The RO may restore automatic rod control after the operable channel is selected. The BOP will place the STM DUMP MODE SELECT in STM PRESS. The CRS will refer to Technical Specification Table 3.3-1 Items 19.B, E and Table 3.3-3 Item 4.d. EVENT 2: Rapid Power Reduction due to overheating of main generator disconnects. On cue from the Examiner, the Booth Operator as the Shift Supervisor will direct the CRS to lower power 10% within the next 15 minutes in accordance with GOP-4C Rapid Power Reduction, due to a report that the transformer disconnects are overheating. The RO will lower Reactor power with boration and/or rod motion. The BOP will reduce turbine load using the Turbine controls.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 3 NUREG -1021 R9 S1 EVENT 3: Running EHC Pump Trip. (Standby EHC must be manually started). TRIGGER 3 o PMP-EH001T XPP0003-PP1 HFM PMP A TRIP ON COMMAND o PMP-EH002F XPP0003-PP2 HFM PMP B FAIL TO START On cue from the Examiner, the running EHC pump will trip and the backup pump will not start in auto. The BOP will respond to annunciator XCP-631 1-4, EHC PP A MOTOR OVRLD, determine the cause of the event, and take corrective action by starting the backup EHC pump to prevent turbine stop valves from closing. A Turbine trip will occur within 2 minutes if the event is not mitigated. EVENT 4: LT- (Manually control feedwater to TRIGGER 4 o XMT-MS020O ILT00496 SG C NR LVL LI-496 FAIL TO POSN FINAL VALUE = 100 RAMP = 00:00:10 transmitter will fail HIGH. The BOP will identify the failure and take manual contrSG level to between 60% and 65% and prevent a reactor trip. The crew will enter AOP-401.11, Steam Generator Level Control and Protection Channel Failure, and remove the channel from service. The CRS will refer to Technical Specifications 3.3-1, Item 13 (Action 6) and 3.3-3, Items 5, and 6c (Action 24) .

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 3 NUREG -1021 R9 S1 EVENT 5: TRIGGER 5 o MAL-CVC005A, RCP 1 NUMBER 2 SEAL FAILURE TRIGGER 6 o VLV-CS052W, XVT08141A-CS RCP A SEAL LEAKOFF VLV LOSS OF POWER DELETE: 1 second On cue from the Examiner, RCP A Seal Number 2 will begin a ramped failure. The crew will respond to annunciator XCP-617 2-4, RCP A STNDPIP LVL HI/LO. NOTE: The Annunciator will alarm within 4 minutes after the event is triggered and will not clear. The RO will fill the standpipe for 2 minutes to determine that either the #1 or #2 seal is failing. The crew will implement AOP-101.2, Reactor Coolant Pump Seal Failure and determine that a reactor trip is not required. The RO will continue to monitor the RCP for further seal degradation. EVENT 6: RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) TRIGGER 7 o MAL-PCS009AB REACTOR TRIP BREAKER A FAILURE (FAIL TO OPEN) FAIL TO: AUTO o MAL-PCS009BB REACTOR TRIP BREAKER B FAILURE (FAIL TO OPEN) FAIL TO: AUTO o PCS008B FAILURE OF MANUAL REACTOR TRIP SWITCH CS-CR01A o MAL-RCS003A REACTOR COOLANT PUMP 1 TRIP FAIL TO: TRIP o VLV-FW025P XVG01611A-FW FEEDWTR ISO VLV A FAIL POSITION Final = 0 Delay = 00:00:02 EVENT TRIGGER 8 o VLV-FW025P XVG01611A-FW FEEDWTR ISO VLV A FAIL POSITION Delete = 00:00:01 X07D033M < 26 Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 3 NUREG -1021 R9 S1 On cue from the Examiner, RCP A will trip but the Reactor will not trip. The crew will enter EOP-1.0 (E-0), Reactor Trip/Safety Injection Actuation. The RO will manually trip the Reactor. Only the Manual Reactor trip switch the RO normally operates is functional. PVG-1611A) spuriously fails closed to lower level above 26% required to force the crew to EOP-15 (FR-H.1), Response to Loss of Secondary Heat Sink. The PVG-1611A failure is auto-removed aft EVENT 7: Loss Of Heat Sink (EFW) After Reactor Trip. EVENT TRIGGER 9 L52RTAO == 1 RX TRIP BRKR RTA OPEN = TRUE OR L52RTBO == 1 RX TRIP BRKR RTB OPEN = TRUE o MAL-MSS015 STEAM FAILURE TO EFW TURBINE PRELOAD o PMP-EF001S XPP0021A MOTOR DRIVEN EFW PMP A SHEARED SHAFT PRELOAD o PMP-EF002T XPP0021B MOTOR DRIVEN EFW PMP B TRIP ON COMMAND o MAL-FWM001A MAIN FEEDWATER PUMP A TRIP o MAL-FWM001B MAIN FEEDWATER PUMP B TRIP o MAL-FWM001C MAIN FEEDWATER PUMP C TRIP TRIGGER 10 o LOA-FWM040 SS-FW61A XVG01611A,B,C KEY SWITCH Position To: BYPASS TRIGGER 11 o LOA-FWM041 SS-FW81A1 IFV03321,3331,3341 TRAIN A KEY SWITCH Position To: BYPASS Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 3 NUREG -1021 R9 S1 TRIGGER 12 o LOA-FWM042 SS-FW81B1 IFV03321,3331,3341 TRAIN B KEY SWITCH Position To: BYPASS This event is automatically triggered when the Reactor Trip Breakers open. The crew will continue in EOP-1.0 (E-0) Reactor Trip/Safety Injection Actuation and identify that there is no Emergency Feedwater flow to the Steam Generators. The crew will then transition to EOP-15.0 (FR-H.1) Response to Loss of Secondary Heat Sink. All Main Feedwater Pumps will trip when the Reactor Trip Breakers open. The BOP will depressurize one Steam Generator, reset the Safety Injection actuation, and attempt to establish Main Feedwater flow to one SG. The Main Feedwater pumps cannot be reset so the success path is to continue in EOP-15 (FR-H.1) and utilize Condensate flow to restore SG level. Trigger 10, 11, 12 places local key-switches in bypass so that Feedwater Valves can be opened to restore flow to one steam generator using Condensate and Feedwater Booster pumps. CRITICAL TASKS: It is a critical task to:

• Start Back-up EHC Pump prior to a Rx/Turbine Trip on Low EHC Pressure. *

• Insert a manual Reactor Trip prior to completion of Immediate Actions of EOP-1.0 (E-0).

• Establish feed flow to at least one SG before RCS Feed and Bleed criteria is met (WR level in any two SGs is less than 12% or PZR pressure is greater than 2330 psig due to the loss of secondary heat sink) is met. TERMINATION: The scenario can be terminated after the crew has established feedwater to one Steam Generator or at the Examiners discretion.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 3 NUREG -1021 R9 S1 Scenario Attributes Events Total Malfunctions (5-8) 9

• PT-446 (Turbine First Stage Pressure) fails LOW.

• Running EHC Pump Trip.

• LT- *

• RCP Trip.

• RX trip breakers fail to open.

• FW Isol Valve 1611A Fails Closed.

• EFW pumps trip.

• MFW pumps trip. Malfunctions after EOP entry (1-2) 3

• Loss of MFW.

• Loss of MD EFW.

• Loss of TD EFW. Abnormal Events (2-4) 4

• PT-446 (Turbine First Stage Pressure) fails LOW.

• Running EHC Pump Trip.

• LT-

• Major Transient (1-2) 1

• Loss of all Feedwater MFW, MD EFW and TD EFW. EOPs Entered (1-2) 1

• EOP-15.0 (FR-H.1) Response to Loss of Secondary Heat Sink. EOP Contingencies (0-2) 1

• EOP-15.0 (FR-H.1) Response to Loss of Secondary Heat Sink. Critical Tasks (2-3) 4

• Start Back-up EHC Pump prior to a Rx/Turbine Trip on Low EHC Pressure.

• Level.

• Manual Trip prior to completion of Immediate Actions of EOP-1.0 (E-0).

• Establish feed flow to at least one SG before RCS feed and bleed criteria is met.

Appendix D Scenario Outline Form ES-D-1 NRC 2015 Scenario 3 NUREG -1021 R9 S1 SIMULATOR SCENARIO SETUP INITIAL CONDITIONS: IC Set 292 100% Power MOL Rod Position = 230 FCV-113 Pot Setting = 4.31 Boron = 1005 ppm Xe = - 2700 pcm Burnup = 10001 MWD/MTU Prior to the scenario, crew should pre-brief on conditions and expectations for the Shift (maintain power, repairs estimated to be complete well before LCO action time expires.) PRE-EXERCISE: Ensure simulator has been checked for hardware problems (DORT, burnt out light bulbs, switch malfunctions, chart recorders, etc.) VCS-TQP-0807 Attachment I-A, Unit 1 Booth Instructor Checklist, has been completed. Hang Tags for equipment out of service. o Hang Caution Tag on HCV-186 due to ASI being OOS Mark up procedures in use with . A turnover sheet has been prepared for each position. Conduct two-minute drill. PRE-LOAD: STANDARD SIMULATOR SETUP: PMP-LD003P, XPP0138 LEAK DETECTION SUMP PMP LOSS OF POWER VLV-FW028W, XVG01676-FW FW HDR RECIRC ISOL VLV LOSS OF POWER VLV-FW029W, XVG01679-FW FW HTR RECIRC ISO VLV LOSS OF POWER VLV-CS052W, XVT08141A-CS RCP A SEAL LEAKOFF VLV LOSS OF POWER VLV-CS054W, XVT08141C-CS RCP C SEAL LEAKOFF VLV LOSS OF POWER VLV-CS053W, XVT08141B-CS RCP B SEAL LEAKOFF VLV LOSS OF POWER ANN-TA030, GEN AUX PNL TRBL SCENARIO RELATED: ANN-TA030 , GEN AUX PNL TRBL FAIL TO: OFF ANN-CS044, ALT SEAL INJ PUMP TRBL FAIL TO: ON MAL-CVC027, ALT SEAL INJ D/G FAIL TO START MAL-CVC029, ALT SEAL INJ PUMP FAIL TO START Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 1 Page: 10 of 40 Event

Description:

PT-446 (Turbine First Stage Pressure) fails LOW. (Rods Drive In) Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 EVALUATOR NOTE: On cue from the Examiner, the selected turbine first stage pressure transmitter (PT-446) will fail LOW. The failure causes a Tave -Tref mismatch resulting in rods inserting at the maximum speed. The crew will enter AOP-401.7, Turbine First Stage Pressure Channel Failure. The RO will respond to the rod insertion by placing rod control in manual and restoring Tave to within 1 degree of Tref. The crew will then select the operable 1st stage pressure channel for control. The RO may restore automatic rod control after the operable channel is selected. The BOP will place the STM DUMP MODE SELECT in STM PRESS. The Failed instrument is addressed in Technical Specification Table 3.3-1 Items 19.B, E and Table 3.3-3 Item 4.d. Within one hour verify the P7 and P13 permissives are dim, trip the affected bistables within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and place AMSAC in Bypass. BOOTH OPERATOR: When directed, insert Event 1 (TRIGGER 1) Indications available: Uncontrolled Rod Motion XCP-615, 2-5, RCS TAVG-TREF HI/LO; XCP-624-4-2, 5-2, 6-2; SG A, B, C STM FLO HI EVALUATOR NOTE: The crew could enter the ARP but it is likely that they will recognize the entry condition for AOP-401.7, Turbine First Stage Pressure Channel Failure. CRS Enters AOP-401.7, Turbine First Stage Pressure Channel Failure AOP-401.7 EVALUATOR NOTE: If XCP-621 1-1 CRB INSERT LMT LO-LO is received the RO will immediately Emergency Borate per AOP-106.1, Emergency Boration until Shutdown Margin is restored. IOA RO 1 Place Rod Control Bank Select Switch to MANUAL AOP-401.7 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 1 Page: 11 of 40 Event

Description:

PT-446 (Turbine First Stage Pressure) fails LOW. (Rods Drive In) Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 RO 2 Ensure TREF 1ST STG PRESS switch is positioned to the operable channel: P446, CH III. (FAILED) OR PT-447, CH IV AOP-401.7 RO 3 Adjust Control Rods until Tavg is within 1.0° F of Tref. AOP-401.7 BOP 4 Check if Main Turbine load is greater than 10% AOP-401.7 CRS 5 Within one hour, verify the following permissives are dim: P-13, 1st STG PRESS P-7, REACTOR TRIP BLOCKED AOP-401.7 EVALUATOR NOTE: Due to the windup (integral) characteristic of the Rod Control function, the crew may not immediately place rods back in automatic. The crew may elect to restore rods to their previous position. RO 6 Restore automatic rod control. a. Check if automatic rod control is desired. b. Verify Reactor power is GREATER THAN 15% (C-5 status light dim). c. Verify Tavg is within 1.0°F of Tref . d. Place ROD CNTRL BANK SEL Switch in AUTO. AOP-401.7 BOP 7 Place Steam Dump Mode Select Switch in STM PRESS. AOP-401.7 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 1 Page: 12 of 40 Event

Description:

PT-446 (Turbine First Stage Pressure) fails LOW. (Rods Drive In) Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 BOOTH OPERATOR: Acknowledge requests for support. Wait 3 minutes Call for permission to proceed. Use TRIGGER 2 to place AMSAC in BYPASS Report that AMSAC is in Bypass CRS 8 Notify I&C to place AMSAC in BYPASS. AOP-401.7 CRS 9 Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, place the failed channel protection bistables in a tripped condition: AOP-401.7 CRS a. Identify the associated bistables for the failed channel. REFER TO Attachment 1. AOP-401.7 AOP-401.7 Attachment 1 CRS b. Record the following for each associated bistable on SOP-401, REACTOR PROTECTION AND CONTROL SYSTEM, Attachment I: Instrument Associated Bistable. Bistable Location. STPs. AOP-401.7 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 1 Page: 13 of 40 Event

Description:

PT-446 (Turbine First Stage Pressure) fails LOW. (Rods Drive In) Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 CRS Table 3.3-1 Action 7 Refers to Technical Specification Table 3.3-1 and within one hour determines by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition. Tech Spec 3.3.1 CRS Table 3.3-3 Action 24 Refers to Technical Specification Table 3.3-3 and determines that the inoperable channel must be placed in a tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Tech Spec 3.3.2 BOOTH OPERATOR: Acknowledge requests for assistance and inform the crew that support personnel will be assigned.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 1 Page: 14 of 40 Event

Description:

PT-446 (Turbine First Stage Pressure) fails LOW. (Rods Drive In) Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 CRS c Notify I&C to place the failed channel protection bistables in a tripped condition within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s: FB-474A FB-484A FB-494A d Initiate a 30 day R&R for placing AMSAC in BYPASS. AOP-401.7 EVALUATOR NOTE: The next event may be initiated after I&C is called to trip the bistables and Technical Specifications have been addressed.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 2 Page: 15 of 40 Event

Description:

Rapid Power Reduction due to overheating of main generator disconnects. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 EVALUATOR NOTE: On cue from the Examiner, the Booth Operator as the Shift Supervisor will direct the CRS to lower power to less than 90% within the next 15 minutes in accordance with GOP-4C Rapid Power Reduction, due to a report that the transformer disconnects are overheating. The RO will lower Reactor power with boration and/or rod motion. The BOP will reduce turbine load using the Turbine controls. BOOTH OPERATOR: No triggers for this event. When directed, call the Control Room as the Shift Supervisor. Notify the CRS that thermography in the switchyard indicates a problem with the main Generator disconnects. Direct the CRS to lower power 10% within the next 15 minutes in accordance with GOP-4C, Rapid Power Reduction. CRS Direct the crew to reduce power 10% in accordance with GOP-4C Rapid Power Reduction. NOTE 2.0 through 3.0 a. If this procedure must be initiated under conditions other than those in Section 2.0, INITIAL CONDITIONS, the Shift Supervisor or Control Room Supervisor will review Sections 2.0, INITIAL CONDITIONS, and 3.0, INSTRUCTIONS. Steps that are not applicable due to plant conditions will be marked N/A and initialed by the Shift Supervisor or Control Room Supervisor. All other items will require sign-off or check-off. b. All personnel who sign off steps in this procedure must enter their names and initials on Attachment I. c. Each step should be initialed and dated when all its substeps are either completed and checked-off or marked as N/A and initialed. GOP-4C Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 2 Page: 16 of 40 Event

Description:

Rapid Power Reduction due to overheating of main generator disconnects. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 GOP- 4C REFERENCE PAGE GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. After any Thermal Power change of greater than 15% within any one hour, Attachment III.H. of GTP-702 must be completed. C. If Reactor Power is stabilized during this procedure for the purpose of raising power per GOP-4A, a Power Range Heat Balance shall be performed. D. Once a Rapid Power Reduction has begun, every effort should be made to prevent the ize the plant. REACTOR CONTROL A. During operation with a positive Moderator Temperature Coefficient, power and temperature changes will require constant operator attention. B. Rod Control should be maintained in Automatic if any Pressurizer PORV is isolated. C. If at any time, power decreases unexpectedly below 0.1% on any Power Range NI (computer indication available) OR below 1.0% on any Power Range NI control board indication (computer not available): 1) No positive reactivity will be added by rods or dilution. 2) A complete reactor shutdown shall be performed per GOP-5. 3) A controlled reactor startup may be commenced per GOP-3 once the event has been reviewed by Reactor Engineering. REACTOR TRIP CRITERIA DURING RAPID LOAD REDUCTION A. If any of the following conditions occur, trip the Reactor and implement EOP-1.0: 1) RCS Tavg is less than 551°F for greater than 15 minutes. 2) Tavg/Tref mismatch exceeds 10°F. 3) Pressurizer pressure approaches 1870 psig. 4) Power reduction at 5% per minute is not sufficient to mitigate the event. GOP-4C NOTE 3.0 If time allows, load reductions should be discussed with the Load Dispatcher. GOP-4C Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 2 Page: 17 of 40 Event

Description:

Rapid Power Reduction due to overheating of main generator disconnects. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 CAUTION 3.1 through 3.12 a. Thermal Power changes of greater than 15% in any one-hour period requires completion of GTP-702 Attachment III.H. b. VCS PID Report, POWER CHANGE SEARCH, should be periodically performed to ensure a thermal power change of greater than 15% in any one-hour period is detected. GOP-4C RO 3.1 Commence rapid Plant Shutdown as follows: a. Energize all Pressurizer Heaters. GOP-4C NOTE 3.1.b Setting FCV-113A&B, BA FLOW SET PT to 8.3 will yield 33 gpm Boration flow rate. GOP-4C RO b. Maintain the following with rod motion or boron concentration changes: 1) Tavg within 10°F and trending to Tref. 2) 3) Control Rods above the rod insertion limit. GOP-4C BOP c. Using the Turbine HMI, Control/Load screen, reduce to the desired load, as low as 5% (50 MWe), as follows: 1) Under Rate %/min, select desired ramp rate up to 5% per minute. 2) Select Load (a dialog box opens). 3) Enter desired load. 4) Select OK. 5) Confirm setpoint. 6) Select OK. 7) Verify proper plant response. GOP-4C CREW Stabilize the unit at 10% reduced power. EVALUATOR NOTE: The next event may be initiated after a significant power change has been observed.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 3 Page: 18 of 40 Event

Description:

Running EHC Pump Trip. (Standby EHC must be manually started). Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 EVALUATOR NOTE: On cue from the Examiner, the running EHC pump will trip and the backup pump will not start in auto. The BOP will respond to annunciator XCP-631 1-4, EHC PP A MOTOR OVRLD, determine the cause of the event, and take corrective action by starting the backup EHC pump to prevent turbine stop valves from closing. A Turbine trip will occur within 2 minutes if the event is not mitigated. BOOTH OPERATOR: When directed, insert Event 3 (TRIGGER 3) Indications available: Control Switch Red and Green lights XCP-631, 1-4, EHC PP A MOTOR OVRLD XCP-631, 1-2, EHC FLUID PRESS LO BOP Enters ARP-001-XCP-631 1-4, EHC PP A MOTOR OVRLD XCP-631 1-4 CORRECTIVE ACTIONS: XCP-631 1-4 BOP 1. If EHC PUMP A is still running, verify high amps. (NO) XCP-631 1-4 BOP 2. Start EHC PUMP B and observe motor amps. XCP-631 1-4 BOP 3. If EHC PUMP A is still running with higher amps than EHC PUMP B, secure EHC PUMP A and continue to monitor EHC PUMP B (NO) XCP-631 1-4 BOOTH OPERATOR: Acknowledge request to check for EHC Leaks. 3 minutes later report no leaks. If called to investigate the pump and/or breaker, wait 3 minutes and report the breaker for the BOP 4. Dispatch an operator to check for EHC System leaks. XCP-631 1-4 BOP 5. If EHC PUMP B is drawing high amps with EHC PUMP A tripped, attempt to restart EHC PUMP A and run both pumps until an external leak is located or a low level in the EHC fluid tank alarm is received. (NO) XCP-631 1-4 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 3 Page: 19 of 40 Event

Description:

Running EHC Pump Trip. (Standby EHC must be manually started). Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 BOP 6. If EHC PUMP B overload annunciator is received after starting, commence a Turbine Runback at 5% per minute per GOP-4C. (NO) XCP-631 1-4 EVALUATOR NOTE: The failure of the backup EHC pump to auto-start results in EHC pressure continuing to decrease. The Low Pressure alarm will alert operators to the failure if not previously discovered however this alarm provides no additional operator actions BOP Respond to alarm EHC FLUID PRESS LO (XCP-631, 1-2) EHC FLUID PRESS LO EVALUATOR NOTE: The next event may be initiated after the B EHC pump is started.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 4 Page: 20 of 40 Event

Description:

LT- SG) Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 EVALUATOR NOTE: to maintain/restore SG level to between 60% and 65% and prevent a reactor trip. The crew will enter AOP-401.11, Steam Generator Level Control and Protection Channel Failure, and remove the channel from service. The CRS will refer to Technical Specifications 3.3-1, Item 13 (Action 6) and 3.3-3, Items 5, and 6c (Action 24) to determine that the protection bistables for the failed channels must be placed in the TRIPPED condition with 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. BOOTH OPERATOR: When directed, insert Event 4 (TRIGGER 4) Indication Available: XCP-624 3-4, SG C LVL DEV XCP-624 6-4, SG C FWF>STF MISMATCH BOP Responds to alarms. BOP Diagnoses/reports LT-496 failed. CRS Enters AOP-401.11. IOA BOP 1 Adjust the Feedwater Flow Control Valve as necessary to restore Narrow Range level in the AFFECTED SG to between 60% and 65%. AOP-401.11 CRS 2 Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, place the failed channel protection bistables in a tripped condition: a. Identify the associated bistables for the failed channel. REFER TO Attachment 1. AOP-401.11 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 4 Page: 21 of 40 Event

Description:

LT- SG) Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 AOP-401.11 Excerpt from Attachment 1 BOP b. Record the following for each associated bistable on SOP-401, REACTOR PROTECTION AND CONTROL SYSTEM, Attachment I: Instrument. Associated Bistable. Bistable Location. STPs. AOP-401.11 EVALUATOR NOTE: The scenario does not allow time any bistables to be put in trip or bypass. BOOTH OPERATOR: Acknowledge request to troubleshoot failure and place bistables in trip. CRS c. Notify the I&C Department to place the identified bistables in trip. AOP-401.11 CRS d. For channels LT-474, LT-485, and LT-496, initiate a 30 day R&R for placing AMSAC in BYPASS. AOP-401.11 Excerpt from Tech Spec Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 4 Page: 22 of 40 Event

Description:

LT- SG) Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 Excerpt from Tech Spec CRS Refers to: T.S. 3.4.3.1 Reactor Trip System Instrumentation - Table 3.3-1 Action 6 T.S. 3.4 3.2 Engineered Safety Feature Actuation System Instrumentation - Table 3.3-3 Action 24 EVALUATOR NOTE: The next event may be initiated after Technical Specifications have been addressed.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 5 Page: 23 of 40 Event

Description:

Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 EVALUATOR NOTE: On cue from the Examiner, RCP A Seal Number 2 will begin a ramped failure. The crew will respond to annunciator XCP-617 2-4, RCP A STNDPIP LVL HI/LO. NOTE: The Annunciator will alarm within 4 minutes after the event is triggered and will not clear. The RO will fill the standpipe for 2 minutes to determine that either the #1 or #2 seal is failing. The crew will implement AOP-101.2, Reactor Coolant Pump Seal Failure and determine that a reactor trip is not required. The RO will continue to monitor the RCP for further seal degradation. BOOTH OPERATOR: When directed, insert Event 5 (TRIGGER 5) Indications Available: XCP-617 2-4, RCP A STNDPIP LVL HI/LO. RO 1. Determine which seal failed as follows: XCP-617 2-4 RO a. Attempt to fill the standpipe as follows: 1) Ensure Reactor Makeup Water System Non-Essentials are aligned. 2) Open PVD-8028, PRT RMWST MU. 3) Open PVD-8168A, RX MU WTR TO STNDPIPE A. XCP-617 2-4 RO 4) When one of the following occurs, close PVD-8168A, RX MU WTR TO STNDPIPE A: a) RCP A STNDPIP LVL HI/LO alarm clears and re-annunciates on a standpipe high level. b) RCP A STNDPIP LVL HI/LO alarm does not clear within two minutes. 5) Close PVD-8028, PRT RMWST MU. 6) Monitor radiation levels in the Reactor Building. XCP-617 2-4 EVALUATOR NOTE: The alarm will NOT clear by filling the standpipe.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 5 Page: 24 of 40 Event

Description:

Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 RO b. If the standpipe alarm clears by filling, assume #2 or #3 Seal failure. c. If the standpipe alarm does not clear by filling, assume #1 or #2 Seal failure and go to AOP-101.2, Reactor Coolant Pump Seal Failure. XCP-617 2-4 CRS Diagnose #1 or #2 Seal failure. CRS Implement AOP-101.2, Reactor Coolant Pump Seal Failure. CAUTION PVT-8141A(B)(C), A(B)(C) SEAL LKOFF, should be closed between three minutes and five minutes after the affected Reactor Coolant Pump is secured. Reactor Coolant System Controlled Leakage should be limited to 33 gpm per Technical Specification 3.4.6.2 in Modes 1,2,3, and 4. AOP-101.2 BOOTH OPERATOR: Wait 3 minutes after being directed to install the fuses Use Trigger 6 to install XVT-8141A-FU-CS75 for RCP A Report power has been restored to the Seal Leakoff Valve for RCP A RO 1 While continuing with this procedure, have an operator install the pre-staged fuses for the AFFECTED RCP's Seal Leakoff Valve in Main Control Board Panel XCP-6109 Subpanel #5: XVT-8141A-FU-CS75. XVT-8141B-FU-CS76. XVT-8141C-FU-CS77. AOP-101.2 NOTE - Step 2 IF Seal Injection flow has been throttled to optimize RCP Seal performance, THEN Step 2 does not need to be performed. AOP-101.2 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 5 Page: 25 of 40 Event

Description:

Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 RO 2 Ensure seal injection flow is GREATER THAN 8 gpm for the affected Reactor Coolant Pump on FI-130A(127A)(124A), RCP A(B)(C) INJ FLO GPM. AOP-101.2 RO 3 Ensure Component Cooling Water flow to the affected Reactor Coolant Pump thermal barrier is between 35 gpm (50%) and 60 gpm (87.5%) on FM-7138(7158)(7178), RCP THERM BAR A(B)(C) (MODUFLASH M2 CC POINTS 19,18, and 20). AOP-101.2

• RO 4 Check the following conditions for the affected Reactor Coolant Pump on the IPCS: Bearing water temperature (LOWER SEAL WTR BRG T) on T0417A (T0437A)(T0457A) is LESS THAN 225°F and NOT significantly increasing. AND #1 seal leakoff temperature (SEAL WTR OUT TEMP) on T0181A (T0182A)(T0183A) is LESS THAN 235°F and NOT significantly increasing. AOP-101.2 RO 5 GO TO Step 11. AOP-101.2 RO 11 Check total #1 seal flow (#1 seal leakoff plus #2 seal leakoff) for the affected Reactor Coolant Pump from the following: a. Check if #1 seal leakoff flow is LESS THAN 6 gpm on FR-154A, RCP SL LKOFF HI RANGE. AOP-101.2 RO b. Determine total #1 seal flow (#1 seal leakoff plus #2 seal leakoff) for the affected Reactor Coolant Pump from the following: 1) #1 seal leakoff flow by observing FR-154B, RCP SL LKOFF LO RANGE, and FR-154A, RCP SL LKOFF HI RANGE, or by having I&C install a temporary flow transmitter with readout on the IPCS per ICP-340.050, TEMPORARY INSTRUMENT INSTALLATION FOR RCP SEAL LEAKOFF MONITORING. 2) #2 seal leakoff flow by monitoring RCDT inleakage per the applicable portion of STP-114.002, OPERATIONAL LEAKAGE TEST, for any increase from the previous leak rate. AOP-101.2 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 5 Page: 26 of 40 Event

Description:

Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 RO 12 IF total #1 seal flow is GREATER THAN 0.8 gpm AND LESS THAN 6 gpm, THEN perform the following: Contact Plant Support Engineering for evaluation. Continue to monitor for further seal degradation. AOP-101.2 EVALUATOR NOTE: The next event may be initiated after #1 seal flow is identified as greater than 0.8 gpm. CRS 13 Return to Procedure and Step in effect. AOP-101.2 EVALUATOR NOTE: If this event is run for greater than 15 minutes XCP-617 2-1, RCP A #1 SL LKOFF FLO HI/LO, will alarm indicating that seal flow is less than 0.8 gpm. The Alternative Action for Step 12 would then require completion of steps 14-16. CRS 14 Within eight hours, stop the affected Reactor Coolant Pump. REFER TO SOP-101, REACTOR COOLANT SYSTEM. AOP-101.2 BOOTH OPERATOR: Acknowledge requests for support. CRS 15 Contact Plant Support Engineering for evaluation. AOP-101.2 RO 16 Continue to monitor for further seal degradation. AOP-101.2 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 27 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 EVALUATOR NOTE: EVENT 6 On cue from the ExaminerEOP-1.0 (E-0) Reactor Trip/Safety Injection Actuation. The RO will manually trip the Reactor. Only the Manual Reactor trip switch the RO normally operates is functional. EVENT 7 This event is automatically triggered when the Reactor Trip Breakers open. The crew will continue in EOP-1.0 (E-0) Reactor Trip/Safety Injection Actuation and identify that there is no Emergency Feedwater flow to the Steam Generators. The crew will then transition to EOP-15.0 (FR-H.1) Response to Loss of Secondary Heat Sink. The BOP will depressurize one Steam Generator, reset the Safety Injection actuation, and attempt to establish Main Feedwater flow to one SG. The Main Feedwater pumps cannot be reset so the success path is to continue in EOP-15 (FR-H.1) and utilize Condensate flow to restore SG level. BOOTH OPERATOR: When directed, insert Event 6 (TRIGGER 7) Indications Available: Indication of a Turbine trip w/o a reactor trip.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 28 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 EOP-1.0 NOTE Steps 1 through 5 are Immediate Operator Actions. The EOP REFERENCE PAGE should be monitored throughout the use of this procedure. Conditions for implementing Emergency Plan Procedures should be evaluated using EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN. EOP-1.0 EVALUATOR NOTE: The Reactor will not automatically trip and the RO Trip switch will not work. The BOP must use his switch to manually trip the reactor.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 29 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 IOA RO 1 Verify Reactor Trip: (NO) Trip the Reactor using either Reactor Trip Switch. Verify all Reactor Trip and Bypass Breakers are open. Verify all Rod Bottom Lights are lit. Verify Reactor Power level is decreasing. EOP-1.0 IOA CRITICAL TASK BOP ALTERNATIVE ACTION 1 Trip the Reactor using both Reactor Trip Switches. If the Reactor is NOT subcritical, THEN GO TO EOP-13.0, RESPONSE TO ABNORMAL NUCLEAR POWER GENERATION, Step 1. EOP-1.0 IOA BOP 2 Verify Turbine/Generator a. Verify all Turbine VLVs are closed. b. Ensure Generator Trip (after 30 second delay): 1) Ensure the GEN BKR is open. 2) Ensure the GEN FIELD BKR is open. 3) Ensure the EXC FIELD CNTRL is tripped. EOP-1.0 IOA BOP 3 Verify both ESF buses are energized. EOP-1.0 IOA RO 4 Check if SI is actuated: (NO) a. Check if either: SI ACT status light is bright on XCP-6107 1-1. OR Any red first-out SI annunciator is lit on XCP-626 top row. b. Actuate SI using either SI ACTUATION Switch. c. GO TO Step 6. ALTERNATIVE ACTION a. GO TO Step 5. EOP-1.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 30 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 RO 5 Check if SI is required: (NO) a. Check if any of the following conditions exist: PZR pressure LESS THAN 1850 psig. OR RB pressure GREATER THAN 3.6 psig. OR Steamline pressure LESS THAN 675 psig. OR Steamline differential pressure GREATER THAN 97 psid. EOP-1.0 EVALUATOR NOTE: EOP-1.0 directs a transition to EOP-1.1 however a Red Path on Heat Sink requires implementation of EOP-15.0, Response To Loss of Secondary Heat Sink. CRS ALTERNATIVE ACTION a. GO TO EOP-1.1, REACTOR TRIP RECOVERY, Step 1. EOP-1.0 CAUTION If total EFW flow is LESS THAN 450 gpm due to operator action, this procedure should NOT be performed, since these actions are NOT appropriate if 450 gpm EFW flow is available. If a NON-FAULTED SG is available, feed flow should NOT be reestablished to any FAULTED SG, to prevent thermal shock to SG tubes. EOP-15.0 NOTE Conditions for implementing Emergency Plan Procedures should be evaluated using EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN. EOP-15.0 RO 1 Check if a secondary heat sink is required: a. Verify RCS pressure is GREATER THAN any NON-FAULTED SG pressure. b. Verify RCS Thot is GREATER THAN 350°F. EOP-15.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 31 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 RO 2 Verify power is available to all PZR PORV Block Valves: a. MVG-8000A, RELIEF 445 A ISOL. b. MVG-8000B, RELIEF 444 B ISOL. c. MVG-8000C, RELIEF 445 B ISOL. EOP-15.0 RO 3 Open the Block Valve for any PZR PORV that has been isolated due to excessive seat leakage: MVG-8000A, RELIEF 445 A ISOL. MVG-8000B, RELIEF 444 B ISOL. MVG-8000C, RELIEF 445 B ISOL. EOP-15.0 CAUTION - Steps 4 through 16 If Wide Range level in any two SGs is LESS THAN 12% OR PZR pressure is GREATER THAN 2330 psig due to loss of secondary heat sink, Steps 17 through 24 should be immediately initiated for bleed and feed cooling. EOP-15.0 RO 4 Ensure the following valves are closed: SG Blowdown, PVG-503A(B)(C). SG Sample, SVX-9398A(B)(C). EOP-15.0 NOTE - Step 5 If EFW flow control can NOT be reestablished from the Control Room, this procedure should be continued while local operator action is in progress to restore EFW flow. EOP-15.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 32 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 BOP 5 Try to establish EFW flow to at least one SG: a. Check Control Room indications for the cause of EFW failure: 1) Verify no EFW annunciators are lit: XCP-621 3-5 (EFP SUCT HDR PRESS LO XFER TO SW). Any alarm on XCP-622. Any alarm on XCP-623. 2) Verify CST level is GREATER THAN 5 ft. 3) Ensure power is available to both MD EFW Pumps. EOP-15.0 CAUTION - Step 5.a.4) EFW valves should NOT be opened to SGs with Wide Range level LESS THAN 12%. If Wide Range level in all SGs is LESS THAN 12%, EFW valves should be open to only one SG, until RCS temperatures are decreasing, to limit any failure to one SG. EOP-15.0 BOP 4) Ensure all EFW valves are open: FCV-3531(3541)(3551), MD EFP TO SG A(B)(C). FCV-3536(3546)(3556), TD EFP TO SG A(B)(C). MVG-2802A(B), MS LOOP B(C) TO TD EFP. PVG-2030, STM SPLY TO TD EFP TRN A(B). EOP-15.0 BOP b. Try to restore any EFW flow. EOP-15.0 BOOTH OPERATOR: Acknowledge requests to investigate the EFW problems. The following conditions exist: The TD EFW Pp Steam Supply Valves have failed closed and cannot be opened. After 5 minutes report to the control room that attempts to correct the problems are unsuccessful.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 33 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 BOP c. Check total EFW flow to SGs GREATER THAN 450 gpm. ALTERNATIVE ACTION c. IF any feed flow to at least one SG verified, THEN perform the following: (NO) IF feed flow NOT verified, THEN perform the following: 1) Locally restore EFW flow. 2) GO TO Step 7. EOP-15.0 RO 7 Stop all RCPs. EOP-15.0 CAUTION - Step 8 Deaerator Storage Tank level should be monitored closely and maintained between 2.5 ft and 10.5 ft on LI-3135, DEAER STOR TK WR LVL FEET, to prevent tripping Condensate and Feedwater Booster Pumps. EOP-15.0 BOP 8 Align the MCB for establishing feed flow: a. Ensure one Condensate Pump is running. b. Ensure two Feedwater Booster Pumps are running. c. Ensure Main FW Control Valves are closed: FCV-478, A FCV. FCV-488, B FCV. FCV-498, C FCV. d. Place all Main FW Bypass Valve Controllers in MAN and closed: FCV-3321,LOOP A MAIN FW BYP. FCV-3331,LOOP B MAIN FW BYP. FCV-3341,LOOP C MAIN FW BYP. EOP-15.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 34 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 BOOTH OPERATOR: When called to place local key-operated switches in bypass: Acknowledge request. Wait 3 minutes. Insert TRIGGER 10, TRIGGER 11, and TRIGGER 12. Wait 1 minute. Report that all switches are in BYPASS. BOP e. Locally place the following key switches in BYPASS (CB-448): XVG01611A,B,C (XPN 7114). IFV03321,3331,3341 TRAIN A (XPN 7115). IFV03321,3331,3341 TRAIN B (XPN 7121). EOP-15.0 BOP f. Verify XCP-612 2-1 is NOT lit (RB PRESS HI-2 STM LINE ISOL). EOP-15.0 NOTE - Step 8.g SG B or C is preferred, so that a steam supply for the TD EFP will be restored as soon as possible. Before the Low Steamline Pressure SI signal is blocked, Main Steam Isolation will occur if the Low Steam Pressure rate setpoint is exceeded. EOP-15.0 BOP g. Align the MS Isolation Valves to depressurize only one SG: 1) Verify the MS Isolation Valve, PVM-2801A(B)(C), is open for the SG to be depressurized. 2) Ensure the remaining two MS Isolation Valves, PVM-2801A(B)(C), are closed. EOP-15.0 EVALUATOR NOTE: The Alternative Actions are only required if the Low Steam Pressure rate setpoint is exceeded before the Low Steamline Pressure SI signal is blocked as stated in Step Note 8.g. BOP ALTERNATIVE ACTION g. Open the MS Isolation Bypass Valve for one SG:

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 35 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 1) IF RCS Tavg is LESS THAN P-12(552°F), THEN place STMLN SI TRAIN A(B) Switches to BLOCK. 2) Depress both MAIN STEAM ISOL VALVES RESET TRAIN A(B). 3) Open the MS Isolation Bypass Valve, PVM-2869A(B)(C), for only the SG to be depressurized. 4) Ensure the remaining two MS Isolation Bypass Valves, PVM-2869A(B)(C), are closed. h. Place the following switches in AUTO: PVG-1611A(B)(C), A(B)(C) ISOL. FCV-3321,3331,3341, FW CNTRL BYP VLVS, Train A Switch. FCV-3321,3331,3341, MAIN FW BYPASS VLVS, Train B Switch. BOP 9 Reset both SI RESET TRAIN A(B) Switches. EOP-15.0 EVALUATOR NOTE: The Main Feedwater Pumps will trip immediately if they are reset due to failure inserted in scenario. BOP 10 Establish Main Feedwater flow to the unisolated SG: a. Verify PERMISV C-9 status light is bright on XCP-6114 1-3. b. Open MOV-1-5A(B)(C), TURB DRN VLV. c. Ensure Feedwater Pump to be started is RESET (MCB or DCS (T ICON)). (NO If any pump resets it will trip immediately) ALTERNATIVE ACTION 10 GO TO Step 11. Observe the NOTE prior to Step 11. EOP-15.0 NOTE - Step 11 Step 11 should NOT be performed as long as the Main Feed Pump is supplying sufficient flow to increase SG level. Before the Low Steamline Pressure SI signal is blocked, Main Steam Isolation will occur if the Low Steam Pressure rate setpoint is exceeded. EOP-15.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 36 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 BOP 11 WHEN the Main Feed Pump will NOT supply adequate flow to the SG, THEN depressurize one SG to establish Condensate flow: a. WHEN RCS Tavg is LESS THAN P-12 (552°F), THEN: Place both STM DUMP INTERLOCK Switches to BYP INTLK. Place STMLN SI TRAIN A(B) Switches to BLOCK. EOP-15.0 NOTE - Step 11.b SG B or C is preferred, so that a steam supply for the TD EFP will be restored as soon as possible. EOP-15.0 BOP b. Open FCV-3321(3331)(3341), LOOP A(B)(C) MAIN FW BYP, to the SG to be depressurized. EOP-15.0 BOP c. Dump steam to the Condenser at the maximum rate: 1) Verify PERMISV C-9 status light is bright on XCP-6114 1-3. 2) Place the STM DUMP MODE SELECT Switch in STM PRESS. 3) Adjust the STM DUMP CNTRL Controller to fully open the Bank 1 Steam Dump Valves. EOP-15.0 CRITICAL TASK BOP d. Adjust Condensate flow to restore SG Narrow Range level to between 26% and 60%. EOP-15.0 EVALUATOR NOTE: The scenario may be terminated after Condensate flow is established to one Steam Generator and SG Level increases. BOP 12 Reset Containment Isolation: RESET PHASE A - TRAIN A(B) CNTMT ISOL. RESET PHASE B - TRAIN A(B) CNTMT ISOL. EOP-15.0 BOP 13 Place both ESF LOADING SEQ A(B) RESETS to: EOP-15.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 37 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 a. NON-ESF LCKOUTS. b. AUTO-START BLOCKS. BOP 14 Establish Instrument Air to the RB: a. Start one Instrument Air Compressor and place the other in Standby. b. Open PVA-2659, INST AIR TO RB AIR SERV. c. Open PVT-2660, AIR SPLY TO RB. EOP-15.0 NOTE - Steps 15 and 16 The specified SG level range (Narrow Range OR Wide Range) must be used in the following Steps. EOP-15.0 EVALUATOR NOTE: If this scenario is not terminated the crew may continue Step 15 until Narrow Range level is GREATER THAN 26%. EOP-15.0 BOP 15 Check SG levels: a. Verify Narrow Range level is GREATER THAN 26% in at least one SG. (NO) ALTERNATIVE ACTION a. IF feed flow to at least one SG is verified by: Core exit TC temperatures decreasing, OR Wide Range SG level increasing, THEN maintain flow to restore Narrow Range SG level to GREATER THAN 26%. RETURN TO Step 15.a. IF flow is NOT verified to any SG, THEN GO TO Step 16. (NA) EOP-15.0 CRS b. RETURN TO the Procedure and Step in effect. EOP-15.0 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # 6, 7 Page: 38 of 40 Event

Description:

RCP Trip, ATWS, FW Isol Valve 1611A Fails Closed. (Manual Reactor trip) Loss of Heat Sink (EFW), Trip of Main FW pumps. Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 EVALUATOR NOTE: If the scenario continues to this point the crew will exit EOP-15.0 and implement EOP-1.1, Reactor Trip Recovery.

Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # NA Page: 39 of 40 Event

Description:

SOP-106, BORATE OPERATIONS Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 NOTE 2.0 1. Energizing additional Pressurizer Heaters will enhance mixing. 2. LCV-115A, LTDN DIVERT TO HU-TK, will begin to modulate to the HU-TK position at 70% level on LI-115, VCT LEVEL %. SOP-106 RO 2.1 Ensure at least one Reactor Coolant Pump is running. SOP-106 RO 2.2 Place RX COOL SYS MU switch to STOP. SOP-106 RO 2.3 Place RX COOL SYS MU MODE SELECT switch to BOR. SOP-106 RO 2.4 Set FIS-113, BA TO BLNDR FLOW, batch integrator to the desired volume. SOP-106 RO 2.5 Place RX COOL SYS MU switch to START. SOP-106 NOTE 2.6 Step 2.6 may be omitted when borating less than 10 gallons. SOP-106 RO 2.6 Place FCV-113 A&B, BA FLOW, controller in AUTO. SOP-106 NOTE 2.7 The AUTO setpoint dial for FCV-113A&B, BA FLOW, controller may be adjusted slowly to obtain the desired flow rate. SOP-106 RO 2.7 Verify the desired Boric Acid flow rate on FR-113, BA TO BLNDR GPM (F-113). SOP-106 RO 2.8 When the preset volume of boric acid has been reached, perform the following: a. Place FCV-113A&B, BA flow controller in MAN. b. Verify boration stops. SOP-106 RO 2.9 Place RX COOL SYS MU switch to STOP. SOP-106 NOTE 2.10 a. If plant conditions require repeated borations, Step 2.10 may be omitted. b. The volume in the piping between the blender and the VCT outlet is approximately 3.8 gallons. SOP-106 Appendix D Operator Actions Form ES-D-2 Op Test No: NRC-ILO-13-01 Scenario # 3 Event # NA Page: 40 of 40 Event

Description:

SOP-106, BORATE OPERATIONS Time Position NRC 2015 Scenario 3 NUREG -1021 R9 S1 RO 2.10 Alternate Dilute 4 to 6 gallons of Reactor Makeup Water to flush the line downstream of the blender by performing the following: a. Place RX COOL SYS MU MODE SELECT switch to ALT DIL. b. Adjust FCV-168, TOTAL MU FLOW SET PT, to desired flow rate. c. Set FIS-168, TOTAL MU FLOW, batch integrator to desired volume. d. Place RX COOL SYS MU switch to START. e. Verify desired flow rate on FR-113, TOTAL MU GPM (F-168). f. Verify alternate dilution stops when preset volume is reached on FIS-168, TOTAL MU FLOW, batch integrator. g. Place RX COOL SYS MU switch to STOP. SOP-106 RO 2.11 Place RX COOL SYS MU MODE SELECT switch to AUTO. SOP-106 RO 2.12 Adjust FCV-168, TOTAL MU FLOW SET PT, to 7.5 (120 gpm). SOP-106 RO 2.13 In MAN, adjust FCV-113 A&B, BA FLOW OUTPUT, to the required position which will ensure proper Boric Acid addition for subsequent Automatic Makeup operations. SOP-106 RO 2.14 Adjust FCV-113A&B, BA FLOW SET PT, to the desired position to ensure proper boric acid addition for subsequent Automatic Makeup operations. SOP-106 RO 2.15 Place RX COOL SYS MU switch to START. SOP-106 RO 2.16 Perform the following: a. Start XPP-13A(B), BA XFER PP A(B), for the in-service Boric Acid Tank. b. If necessary, start XPP-13A(B), BA XFER PP A(B), for the Boric Acid Tank on recirculation. SOP-106 END OF SECTION SOP-106 OAP-100.6 ATTACHMENT VIII PAGE 1 OF 1 REVISION 4 TURNOVER NOTES (read at the start of the scenario) Turnover Notes Mode 1 // 100% Power // Work Week B1 // EOOS: Green // Grid Risk: Green // FEP Risk: Green // Switchyard thermography is in progress. Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. Xenon concentration is at equilibrium. Midnight RCS Boron Concentration is 1005 ppm.

OAP-100.6 ATTACHMENT VIII PAGE 1 OF 2 REVISION 4 CONTROL ROOM SUPERVISOR RELIEF CHECKLIST DATE/TIME: today RELIEF SECTION Turnover Notes Mode 1 // 100% Power // Work Week B1 // EOOS: Green // Grid Risk: Green // FEP Risk: Green // Switchyard thermography is in progress. Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. Xenon concentration is at equilibrium. Midnight RCS Boron Concentration is 1005 ppm. Offgoing Control Room Supervisor Operations in progress (GOPs, SOPs, load changes, etc.): Operations scheduled for oncoming shifts: Plant safeguard systems in degraded status: Initials In the Control Room, all books are replaced, the desk and console tops are clear, and all trash is properly disposed of. CRS Station Log completed. CRS OAP-100.6 ATTACHMENT VIII PAGE 2 OF 2 REVISION 4 C02 To the best of my knowledge, I am fully qualified to assume this watch taking into consideration fitness for duty, requalification status, and minimum watchstanding qualification. Shift relief completed: Oncoming Control Room Supervisor Offgoing Control Room Supervisor CR Supervisor Shift Supervisor review Oncoming Control Room Supervisor Initials Oncoming watch has reviewed the VCS Switchgear mailbox for switching orders. Plant Status (to be completed prior to turnover): Plant ESF System Status: Component Cooling System Service water System Reactor Building Cooling System Reactor Building Spray System Accumulator Tanks RHR System Charging/Safety Injection System Emergency Feedwater System Accumulator Tanks Diesel Generator Chilled Water System Control Room Ventilation System Position indications, power availability, and annunciator alarms are normal for present plant conditions. Plant Parameters Limit Reactor Power 0-100% RCS Tavg 589.2°F per loop RCS Pressure <2385 psig RCS Flow >100% per loop RCS Subcooling Normal All parameters within allowable limits for plant conditions. If not, what actions are being taken to correct conditions: Review of Logs: Station Log Removal and Restoration Log Tagout Log Special Orders Shift Turnover (to be completed during turnover): Briefing on plant conditions by offgoing Control Room Supervisor. Review of SPDS and BISI displays. Discussion of Protected Equipment. Identification of in-progress procedures including their present status and locations.

OAP-100.6 ATTACHMENT IX PAGE 1 OF 1 REVISION 4 REACTOR OPERATOR RELIEF CHECKLIST DATE/TIME: today LOG SECTION Date Entry RELIEF SECTION Entry Turnover Notes Mode 1 // 100% Power // Work Week B1 // EOOS: Green // Grid Risk: Green // FEP Risk: Green // Switchyard thermography is in progress. Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. Xenon concentration is at equilibrium. Midnight RCS Boron Concentration is 1005 ppm. Offgoing Reactor Operator Initials Main Control Board (Reactor Operator portion) properly aligned for the applicable mode. RO Housekeeping is satisfactory in the Reactor Operator area of responsibility. RO Discussion of Protected Equipment. RO Oncoming Reactor Operator Initials Review of HVAC Panel. Review of Station Log. Review of Removal & Restoration Log. Review of Main Control Board Panels. System Alignment A B C Train aligned to Reasons for any inoperable equipment Service Water Pumps X X A Component Cooling Pumps X A Charging Pumps X A HVAC Chillers X X A Reactor Building Spray Pumps RHR Pumps TDEFP Emergency Feedwater Pumps Inoperable Radiation Monitors C02 To the best of my knowledge, I am fully qualified to assume this watch taking into consideration fitness for duty, requalification status, and minimum watchstanding qualification. Shift relief completed: Oncoming Reactor Operator Offgoing Reactor Operator Reactor Operator Shift Supervisor review OAP-100.6 ATTACHMENT X PAGE 1 OF 1 REVISION 4 BALANCE OF PLANT RELIEF CHECKLIST DATE/TIME: today Date Entry RELIEF SECTION Entry Turnover Notes Mode 1 // 100% Power // Work Week B1 // EOOS: Green // Grid Risk: Green // FEP Risk: Green // Switchyard thermography is in progress. Alternate Seal Injection is OOS for planned maintenance. It has been OOS for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and is expected back in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. A fire watch has been established IAW SOP-102. Xenon concentration is at equilibrium. Midnight RCS Boron Concentration is 1005 ppm. Offgoing Reactor Operator Initials Main Control Board (Reactor Operator portion) properly aligned for the applicable mode. BOP Housekeeping is satisfactory in the Reactor Operator area of responsibility. BOP Discussion of Protected Equipment. BOP Oncoming Reactor Operator Initials Review of Main Control Room Panels. Review of Station Log. Review of Removal & Restoration Log. Test annunciator lights (with Offgoing operator concurrence). C02 To the best of my knowledge, I am fully qualified to assume this watch taking into consideration fitness for duty, requalification status, and minimum watchstanding qualification. Shift relief completed: Oncoming Balance of Plant Offgoing Balance of Plant Balance of Plant Shift Supervisor review OAP-100.6 ATTACHMENT IA PAGE 1 OF 2 REVISION 4 REACTIVITY CONTROL PARAMETERS NOTE This information should be recalculated every Sunday Dayshift (when the plant is in Mode 1) to be available for Reactor Engineering review Monday morning or following work day. RCS Boron Concentration (CRCS) ppm Burnup______________ MWD/MTU (Check BAT in Service) CB "A" BAT ppm CB "B" BAT ppm Moderator Temperature Coefficient (MTC) (Fig. II-3.7, HFP) pcm/°F Differential Boron Worth (DBW) (Fig. II-7.2, HFP) pcm/ppm Gallons of Boric Acid or Reactor Makeup Water required to change RCS average temperature by one (1) degree: MTC/DBW = / = (B) ppm Boron Change/°F gal. Acid/°F = From Fig. III-2: gal. Acid/°F = 49640 ln (()()) gal. RMW/°F = From Fig. III-3: gal. RMW/°F = 49640 ln (() ) Power Defect (PD) for 10% power change (100% to 90%) (Fig. II-2): _______ PD @ 100% RTP - ________ PD @ 90% RTP = Gallons of Boric Acid only to reduce reactor power from 100% to 90%: ________ ppm Boron (Fig. III-2) _________ gal. Boric Acid/10% RTP Final rod height using rods only to reduce reactor power from 100% to 90%: (Assume ARO) (IRW) = ___ pcm (Fig. II-10) final rod height Bank D OAP-100.6 ATTACHMENT IA PAGE 2 OF 2 REVISION 4 REACTIVITY CONTROL PARAMETERS NOTE For a 10% reduction in load, 1/2 of the calculated boric acid should be used and 1/2 the calculated Control Rod motion. For a 100% to 90% load reduction: Use ___________ gallons boric acid (1/2 the gallons calculated above), and expect the rods to be at approximately ___________ steps on bank D (Fig. II-10 series, 1/2 the IRW, NOT 1/2 the steps). To change TAVG by 1o F: ____________ gallons Boric Acid/°F ____________ gallons Reactor Makeup Water/°F For a 100% to 90% load reduction: Use ___________ gallons boric acid and expect ___________ steps on bank D NOTE: This calculation is to provide a second check to the batch integrator setting to establish continuity between the setting and actual make-up results. FCV 113 A&B, pot setting for current RCS boron concentration __________ Expected Boric Acid flowrate for VCT makeup ________

Expected Boric Acid total gallons on an Auto Makeup based on current BAT in service: Current RCS CB _____ x _____ gallons* = _______________ CB for BAT in service

• Normal Auto Makeup is 267 to 275 gallons Calculation and Auto Makeup pot settings by Signature / Date Calculation and Auto Makeup pot settings verified by Signature / Date Reactor Engineering Review _________________________ Date______________

OAP-100.6 ATTACHMENT IB PAGE 1 OF 2 REVISION 4 REACTIVITY MANAGEMENT BRIEF MODES 1 - 3 NOTE PART 1 REACTIVITY MANAGEMENT TURNOVER should be read at Shift Turnover Meeting. PART 2 REACTOR STATUS should be discussed between the NROATC, BOP, and CRS.

PART 1 REACTIVITY MANAGEMENT TURNOVER: Date of last Automatic or Manual Make-Up:

Is Auto Makeup expected this shift (circle)? YES NO Expected Boric Acid total gallons on a normal Auto Makeup based on current BAT in service: gallons FCV 113 A&B, pot setting for current RCS boron concentration:

Expected Boric Acid flowrate for VCT makeup:

Total gallons Diluted Borated (Last Shift)

Last evolution (circle one): Borate / Dilute / Blended Expected Borations, Dilutions, or Blended changes to the RCS: List Reactivity Concerns in progress or planned and action(s) necessary (i.e. Steam or Feed Flow transmitter in test, Steam Generator Blowdown out of service, Calorimetric inputs in service, etc.).

OAP-100.6 ATTACHMENT IB PAGE 2 OF 2 REVISION 4 REACTIVITY MANAGEMENT BRIEF MODES 1 - 3 (Cont'd) PART 2 REACTOR STATUS: (circle one below) Delta I on Target (+ 2%)? YES NO Not in Mode 1 If NO is circled, identify plan to re-establish target band: Xenon Trend: Stable Building In Burning Out Demineralizers: Mixed Bed in service: A B PRC01 Y / N Standby Demineralizer: Filled Borated Empty PRC01 Cation Bed: Date last in service Boron Concentration when in service ATTACHMENT IA reviewed and current: YES NO Midnight Boron Concentration and Date when CHG/SI pump was secured:

CB A Date CB B Date CB C Date OAP-102.1 ATTACHMENT II PAGE 1 OF 1 REVISION 7 SCHEDULED WORK APPROVAL/DENIAL Scheduled Work/Activity Date Description of Work/Activity to be performed: I. This Moderate Risk, Elevated Risk, High Risk, or Cross Train activity is approved for work provided the required plant conditions are available on the scheduled due date. OR This specific activity has been reviewed for EOOS Risk Reassessment. Set EOOS Environmental Variance __________________________ Set Risk at Times The following items were considered for making this approval: Operations Supervisor (Moderate Risk or Cross Train) In the absence of the Operations Supervisor: Operations Scheduling, Shift Supervisor GMNPO/MDS (Elevated Risk) PSRC (High Risk) II. This work activity/package cannot be performed on the scheduled date due to the following reason(s): SRO (WCC or On Shift) Operations Scheduling Supervisor III. Recommended re-schedule date or plant conditions:

JPM NO:NJPSF-141A2013 and 2015 NRC Sim a; RO, SRO-I & SRO-U: Continuous Rod WithdrawalV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Page 1 of TASK:000-006-05-01PREFERRED EVALUATION METHODPERFORMTASK STANDARD:The reactor is tripped per AOP-403.3, CONTINUOUS CONTROL ROD MOTION, to terminate the transient prior to rods withdrawing to the point of adding heat (10e0 on Intermediate Range) and immediate actions of EOP-1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION, are completed.PREFERRED EVALUATION LOCATIONSIMULATORRESPOND TO CONTINUOUS ROD MOTION

REFERENCES:

TERMINATING CUE:Immediate actions of EOP-1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION are complete.Curve BookREP-109.001REP-109.002GOP Appendix A EOP-1.0 AOP-403.3GOP-3Station Curve BookCalculation of Estimated Critical ConditionsInverse Count Rate Ratio PlotGeneric Operating PrecautionsREACTOR TRIP/SAFETY INJECTION ACTUATIONCONTINUOUS CONTROL ROD MOTIONREACTOR STARTUP FROM HOT STANDBY TO STARTUP(MODE 3 TO MODE 2)INDEX NO.ROSROK/A NO.0000012413Knowledge of crew roles and 4.04.62.4.13000001A105AA1.05Reactor trip switches4.34.2Page 2 of EVALUATION TIME30TIME CRITICALNoTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45.a.3TOOLS:GOP Appendix A, Generic Operating PrecautionsNJPSF-141A Handout 1; GOP-3, Reactor Startup from Hot Standby to Startup (MODE 3 TO MODE 2), marked up through step 3.12.kNJPSF-141A Handout 2; REP-109.002, Enclosure 9.2, Recommended Rod Positions for ICRR marked through Bank C at 129 steps. AOP-403.3, Continuous Control Rod MotionEOP-1.0, Reactor Trip/Safety Injection ActuationCopy of REP-190.002, Inverse Count Rate Ratio PlotCANDIDATE:

INITIATING CUES:Complete the Reactor Start up and increase reactor power to 10-3% per GOP-3 starting at Step 3.12.l.INITIAL CONDITION:A reactor start up is in progress after a short mini-outage. GOP-3, REACTOR STARTUP FROM HOT STANDBY TO STARTUP (MODE 3 TO MODE 2), has been completed through step 3.12.k. The Rod Insertion Limit at 0% power is 118 steps on Control Bank C.The CRB INSERT LMT LO-LO (XCP-621 1-1) annunciator is NOT clear.Control bank C is at 129 steps with Control Bank D at 1 step.The estimated critical position is 100 steps on bank "D". The Minimum rod height for criticality (-500 pcm equivalent) is 38 steps on Bank D. The Maximum rod height for criticality (+500 pcm equivalent) is 185 steps on Bank D.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Page of STEPSProcedure Caution: Reactor startup should be stopped and I&C notified if the CRB INSERT LMT LO-LO (XCP-621 1-1) annunciator fails to clear between 118 steps and 134 steps on Bank C.Step 3.12.l; Verify CRB INSERT LMT LO-LO (XCP-621 1-1) annunciator clears between 118 steps and134 steps on Bank C.. Steps________________Evaluator note: Provide Examinee with copies of NJPA-141A Handout 1 (GOP-3.0 markup) and NJPA-141A Handout 2 (REP-109.002 Enclosure 9.2 mark up) following initial conditions brief.Surrogate cue: Once Examinee is ready (on evaluator prompt) provide the following direction "Pull to 6 steps on Control Bank D or until the LO-LO Insertion Limit Alarm Clears whichever occurs first"Evaluator note: Expect alarm to clear at 3 steps withdrawn on Control Bank D. Surrogate cue: Once Examinee stops and verifies LO-LO insertion limit annunciator is clear, provide the following direction "Pull to 10 steps on Control Bank D."YesVerifies LO-LO insertion Limit Annunciator clears (XCP-621 1-1)CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Page of Procedure Caution: 12 steps should NOT be exceeded until all Rod Bottom lights are off. If all Control Bank D Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered.Step 3.12.m; At ten steps on Control Bank D, stop and verify Bank D RB lights clear.Surrogate cue: After Examinee stops rod pull give the following direction "Inform me when counts are stable"Surrogate cue: After "ICRR is 0.4 and criticality is predicted at 120 steps withdrawn on Control Bank D." Then give the following direction "Pull Control Bank D to 16 steps withdrawn or until LO Insertion Limit annunciator clears whichever comes first."YesStops at 10 steps withdrawn on Control Bank D.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:2Page of Step 3.12.n; Recommence withdrawing rods while observing that the groups sequence properly.Surrogate cue: Once Examinee stops and verifies LO insertion limit annunciator is clear, provide the following direction "Pull to 53 steps on Control Bank D."YesContinues to pull rods and stops at 16 steps withdrawn on Control Bank D or when the LO Insertion Limit Annunciator clears .CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:3Recommence withdrawing rods while observing that the groups sequence properly.Evaluator note: When rods are >51 steps on Bank D the continuous rod motion malfunction inserts. When examinee stops pull at 53 steps continuous rod motion occurs. This is the point that the JPM becomes alternate path.YesContinues rod withdrawal.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:4Page of Enters AOP-403.3, CONTINUOUS CONTROL ROD MOTION.Evaluator note: The examinee is not expected to pull out the procedure, but may perform the actions of this procedure from memory and trip the unit.YesEnters AOP-403.3, CONTINUOUS CONTROL ROD MOTION.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:5Step 1; Verify rod motion is NOT required: Tavg is within 1.5 °F of Tref.AND No load rejection has occurred (C7A OR C7B).Evaluator note: This is an immediate operator action from AOP-403.3 and is expected to be performed from memory.YesNotes Tavg and Tref matched and Status lights for C7A and C7B are dim: rod motion is not required.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:6Page of Step 2; Place ROD CNTRL BANK SEL Switch in MAN.YesRods are already in manual no action required.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:7Step 3; Verify rod motion is stopped.Evaluator note: This is an immediate operator action from AOP-403.3 and is expected to be performed from memory.YesNotes rods out light lit, step counters clicking and DRPI showing Bank D withdrawing, concludes that rod motion has NOT stopped.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:8Page of Step 3 Alternative Action: Perform the following:a)Trip the Reactor.b)GO TO EOP-1.0. REACTOR TRIP/SAFETY INJECTIONACTUATION.Evaluator note: This is an immediate operator action from the Alternative Action of AOP-403.3 step 3.The Point of Adding Heat was noted at approximately 170 steps withdrawn on Control Bank Dduring development.YesTurns one of the two reactor trip switches to trip prior to rods withdrawing to the Point of Adding Heat (10e0 on Intermediate Range instrumentation).CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:9Page of Procedure Note:-Steps 1 through 5 are Immediate Operator Actions. -The EOP REFERENCE PAGE should be monitored throughout the use of this procedure.- Conditions for implementing Emergency Plan Procedures should be evaluated using EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN.Step 1. Verify Reactor Trip: - Trip the Reactor using either Reactor Trip Switch. - Verify all Reactor Trip and Bypass Breakers are open. - Verify all Rod Bottom Lights are lit. - Verify Reactor Power level is decreasing.Evaluator note: This is an immediate operator action from EOP-1.0 and is expected to be performed from memory.

Evaluator cue: Direct Examinee to perform all Immediate actions from EOP-1.0 (both RO and BOP actions).YesVerifies: -Reactor Trip and Bypass Breakers indicate Green light ON Red light OFF.-Rod Bottom Lights are lit.-Reactor Power level is decreasing.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:10Page 1 of Step 2; Verify Turbine/Generator Trip:a.Verify all Turbine STM STOP VLVs are closed.b.Ensure Generator Trip (after 30 second delay):1)Ensure the GEN BKR is open.2)Ensure the GEN FIELD BKR is open.3)Ensure the EXC FIELD CNTRL is tripped.Evaluator note: This is an immediate operator action from EOP-1.0 and is expected to be performed from memory.YesVerifies:a.All Turbine STM STOP VLV indicate closed, status light for each valve is bright.b.GEN BKR, GEN FIELD BKR, and EXC FIELD CNTRL indicate Green light ON and Red light OFF.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:11Page 1 of Step 3; Verify both ESF buses are energized.Evaluator note: This is an immediate operator action from EOP-1.0 and is expected to be performed from memory.YesVerifies potential lights on 1DA and 1DB are ON for all three phases on both buses.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:12Page 1 of Step 4; Check if SI is actuated:a.Check if either:SI ACT status light is bright on XCP-6107 1-1. OR Any red first-out SI annunciator is lit on XCP-626 top row.Alternative Action go to Step 5.Evaluator note: This is an immediate operator action from EOP-1.0 and is expected to be performed from memory.YesVerifies: status light dim and no SI first out lit, goes to Step 5.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:13Page 1 of Examiner ends JPM at this point.Step 5; Check if SI is required: a.Check if any of the following conditions exist:PZR pressure LESS THAN 1850 psig. OR RB pressure GREATER THAN 3.6 psig. OR Steamline pressure LESS THAN 675 psig. OR Steamline differential pressure GREATER THAN 97 psid. Alternative action GO TO EOP-1.1, REACTOR TRIP RECOVERY, Step 1.YesVerifies: PZR pressure is greater than 1850 psigRB pressure less than 3.6 psigAll steam line pressures greater than 675 psigAll steam line pressures within 97 psi.Transitions to EOP-1.1, REACTOR TRIP RECOVERY.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:14Page 1 of JPM NO:NJPSF-141ADESCRIPTION:2013 and 2015 NRC Sim a; RO, SRO-I & SRO-U: Continuous Rod WithdrawalIC SET:310INSTRUCTIONS:If IC 310 is designated for this JPM reset to IC 310.1.RUN2.Set up Audio Count Rate per GOP-3, step 3.4.e3.Turn on Digital Reactivity Function of the IPCS per REP-109.002 step 7.6. set updisplay at ROATC SIPCS station using ZZREAC or RX STRT off the ZZ Menu.4.Place HIGH FLUX AT SHUTDOWN in block per GOP-3, step 3.11.b.5.Set SIPCS to MODE 2 per GOP-3, step 3.11.f6.Set NR-45 to HI speed.7.RUN until the Heat up or cooldown history clears on SIPCS. This may take 10-15 minutes on theinitial reset. 8.FREEZE9.When Examinee is ready (on Evaluator cue) go to RUN.If IC 310 is not designated for this JPM then initial conditions may be established by resetting to IC 15 and following the below directions:1.Go to RUN and withdraw Control Rods to 129 steps on Control bank C (1 step on Control Bank D).2.Insert: MAL-PCS009AB REACTOR TRIP BREAKER A FAILURE (FAIL TO OPEN) Delay = 0, Fail To = AUTO (UV)2.Insert MAL-PCS009BB REACTOR TRIP BREAKER B FAILURE (FAIL TO OPEN) Delay = 0, Fail To = AUTO (UV)4.Set Event #1 as Mcrfpa(11) >515.Insert: MAL-CRF006B UNCONTROLLED MANUAL ROD MOTION, Delay=0, set to event #16.Set up Audio Count Rate per GOP-3, step 3.4.e7.Place HIGH FLUX AT SHUTDOWN in block per GOP-3, step 3.11.bJPM SETUP SHEETPage 1 of 8.Set SIPCS to MODE 2 per GOP-3, step 3.11.f9.Turn on Digital Reactivity Function of the IPCS per REP-109.002 step 7.6.set up display at ROATCSIPCS station using ZZREAC or RX STRT off the ZZ Menu.10.Set NR-45 to HI speed.

11.RUN until the Heat up or cooldown history clears on SIPCS. This may take 10-15 minutes on theinitial reset.12.FREEZE13.When examinee is ready: RUNCOMMENTS:Provide a surrogate in the role of CRS to simulate performing REP-109.002, Inverse Count Rate RatioPlot and to provide cues for start up process.During development, the Point of Adding Heat (10e0 on Intermediate Range ) was observed at approximately 170 steps withdrawn on Control Bank D when the continuous rod withdrawal was allowed to proceed with auto trips failed malfunction in place. When the continuous rod motion malfunction was run without the auto trips blocked it took 2 minutes to reach the Source Range High Flux trip setpoint (10e5 CPS) and Control Bank D was at 147 steps withdrawn..Thursday, , 201Page 1 of INITIATING CUES:Complete the Reactor Start up and increase reactor power to 10-3% per GOP-3 starting at Step 3.12.l.INITIAL CONDITION:A reactor start up is in progress after a short mini-outage. GOP-3, REACTOR STARTUP FROM HOT STANDBY TO STARTUP (MODE 3 TO MODE 2), has been completed through step 3.12.k. The Rod Insertion Limit at 0% power is 118 steps on Control Bank C.The CRB INSERT LMT LO-LO (XCP-621 1-1) annunciator is NOT clear.Control bank C is at 129 steps with Control Bank D at 1 step.The estimated critical position is 100 steps on bank "D". The Minimum rod height for criticality (-500 pcm equivalent) is 38 steps on Bank D. The Maximum rod height for criticality (+500 pcm equivalent) is 185 steps on Bank D.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:JPM BRIEFING SHEETHAND THIS PAPER BACKTO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION NUCLEAR OPERATIONS NUCLEAR OPERATIONS COPY NO. GENERAL OPERATING PROCEDURE GOP-3 REACTOR STARTUP FROM HOT STANDBY TO STARTUP (MODE 3 TO MODE 2) REVISION 13 SAFETY RELATED RECORD OF CHANGES CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE A P 01/25/10 B P 06/19/12 C P 07/02/12 D P 04/26/14 E P 06/30/14 F P 11/14/14 CONTINUOUS USE Continuous Use of Procedure Required. Read Each Step Prior to Performing.

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GOP-3 PAGE i REVISION 13 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE/SCOPE 1 2.0 INITIAL CONDITIONS 2 3.0 INSTRUCTIONS 4

4.0 REFERENCES

21 ATTACHMENTS Attachment I - Sign-off Identification List This page Intentionally left blank. For printing 2 sided sheets.

GOP-3 REVISION 13 PAGE 1 OF 22 1.0 PURPOSE/SCOPE 1.1 This procedure provides instructions for Reactor Startup, from Hot Standby to Startup. 1.2 The following governing regulations apply to this procedure: a.10CFR50.59.b.10CFR50, Appendix B.c.SAP-630, Procedure/Commitment Accountability Program.

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GOP-3 REVISION 13 PAGE 2 OF 22 NOTE 2.0 and 3.0 a.All personnel who sign off steps in this procedure must enter their names andinitials on Attachment I.b.Each step should be initialed and dated when all its substeps are eithercompleted and checked-off or marked N/A and initialed.NOTE 2.0 If this procedure must be initiated under conditions other than those in Section 2.0, INITIAL CONDITIONS, the Shift Supervisor or Control Room Supervisor will review Sections 2.0, INITIAL CONDITIONS, and 3.0, INSTRUCTIONS. Steps that are not applicable due to plant conditions will be marked N/A and initialed by the Shift Supervisor or Control Room Supervisor. All other items will require sign-off or check-off. 2.0 INITIAL CONDITIONS INITIALS/DATE 2.1 RCS status is as follows: / a.System temperature is being maintained between555°F and 559°F using the Bank 1 Condenser SteamDumps or Steamline PORVs.b.System pressure is being maintained between 2230 psigand 2240 psig in AUTO control.c.All Reactor Coolant Pumps are in operation.d.Pressurizer level is being maintained at 25%in AUTO control.2.2 All Safety Injection Systems are aligned and operable. / 2.3 Excore NIs are aligned for critical operation per SOP-404, / Excore Nuclear Instrumentation System. 2.4 The Reactor is shutdown with all Control Bank Rods fully / inserted.

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GOP-3 REVISION 13 PAGE 3 OF 22 INITIALS/DATE 2.5 Shutdown Margin is being maintained for Mode 3 conditions / per STP-134.001, Shutdown Margin Verification. 2.6 Reactor Makeup Control is in AUTO and set for blended flow / equal to the existing boron concentration. 2.7 Secondary Plant status is as follows: / a.The Main Turbine is on the Turning Gear per SOP-215,Main Turbine Lube Oil Supply System.b.The Main Feedwater Pumps are on their Turning Gearsper SOP-209, Feedwater Turbine Lube Oil System.c.Narrow Range Steam Generator levels are beingmaintained between 60% and 65% with chemistry withinspecification using the following:1)Blowdown per SOP-212, Steam GeneratorBlowdown.2)Emergency Feedwater per SOP-211,Emergency Feedwater System.d.Main Steam is being warmed per SOP-201,Main Steam System.e.Feedwater is being warmed per SOP-210,Feedwater System.f.Condensate is in operation per SOP-208,Condensate System.g.Circulating Water is in operation per SOP-207,Circulating Water.2.8 The Rod Control and Position Indicating Systems are in / operation per SOP-403, Rod Control And Position Indicating System. 2.9 The Control Rod Drive Mechanism Ventilation System is in / operation per SOP-114, Reactor Building Ventilation System. 2.10 GOP Appendix A review has been completed. /

GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 4 OF 22 3.0 INSTRUCTIONS INITIALS/DATE 3.1 Shut down and isolate BTRS as follows: / a.Place HCV-387, BTRS BYP FLOW, in BYP.b.Place BTRS SELECT Switch in OFF.3.3 Perform the following if an RB entry is in progress or will occur during the reactor startup: a.Obtain the approval of the General Manager,Nuclear Plant Operations, for personnel to be in the RBduring the reactor startupb.Notify Health Physics that a reactor startup is about tocommence and dose rates in the RB could change rapidly.

3.2 Verify RCS Chemistry control for startup: a.Contact Chemistry to ensure RCS Chemistry control issatisfactory for startup per CP-625, Chemistry RefuelingShutdown And Startup Plan.b.Record current Boron concentration:CHG F /

GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 5 OF 22 INITIALS/DATE 3.4 Align Excore NIs for Reactor Startup as follows: / Z005 a.Ensure INI00033-NI, REMOTE SOURCE RANGEMONITOR, is de-energized with fuses removedper SOP-404, Excore Nuclear Instrumentation System,Section IV.F.Z007 b.Ensure the following Nuclear Instrumentation Channelsare in operation per SOP-404, Excore NuclearInstrumentation System,Section III.A and testedper the applicable STPs:1)Two Source Range Channels.2)Two Intermediate Range Channels.3)At least three Power Range Channels.c.Verify both Source Range Channels are indicating aminimum of two counts per second.d.Perform either of the following to monitor Source andIntermediate Range Channels as follows:1)Select the highest reading Source RangeChannel and either Intermediate Range Channelon recorder NR-45, NIS RECORDER.2)Monitor the highest reading Source RangeChannel and either Intermediate Range Channelusing computer display NR45 in FAST SPEED.CHG D GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 6 OF 22 INITIALS/DATE Step 3.4 continued NOTE 3.4.e Audio Count Rate is not required to be operable. e.At the AUDIO COUNT RATE CHANNEL drawer,perform the following:1)Select the highest reading Source RangeChannel on the CHANNEL SELECTOR Switch.2)Adjust the AUDIO MULTIPLIER Switch asnecessary to maintain a distinguishable audiocountrate.3)Place the SR COUNTER/SCALER,POWER switch in the POWER position.3.5 Complete Attachment III.A, Prior to Closing Reactor Trip / Breakers in Modes 3, 4 & 5, of GTP-702. C013.6 Ensure the P-4 trip actuating device operational test is / N01performed and Reactor Trip breakers are closed per STP-345.039, Reactor Trip P-4 Trip Actuating Device Operational Test. Z008 3.7 Ensure both Rod Control MG sets are supplying load to / to Rod Control per SOP-403, Rod Control and Position Indicating System,Section III.A. CHG B CHG D GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 7 OF 22 INITIALS/DATE 3.8 If necessary, withdraw the Shutdown Banks as follows: / a.Verify Shutdown Margin Boron Concentration issatisfactory by performing STP-134.001, ShutdownMargin Verification for Mode 3 with S/D Banks OUTb.Place ROD CNTRL START UP RESET Switch inSTART UP.CAUTION 3.8.c To minimize the possibility of binding at the full in position, rods should not be driven below the 000 indication on the Group Demand Step Counters. c.Ensure the Step Counters indicate zero (000) steps.Z009 d.Update Rod Bank positions on the IPCS, refer toOAP-107.1, Control of IPCS Functions, Step 6.2.b.f.Momentarily depress the ROD CNTRL ALARM RESETPushbutton.g.Verify ROD CNTRL SYS FAIL URGENT (XCP-620 5-1)and ROD CNTRL SYS FAIL NON-URGENT(XCP-620 5-5) alarms cleared.CHG C CHG D e.Ensure IZM01200, DRPI Main Control Board DisplayMonitor, and IZM01201, DRPI Main Control BoardDisplay Monitor, indicate RB.CHG F GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 8 OF 22 INITIALS/DATE Step 3.8 continued CAUTION 3.8.h To prevent any inadvertent inward rod motion the ROD CNTRL BANK SEL Switch should not be placed in or pass through AUTO. NOTE 3.8.h Reactor Coolant System temperature is being maintained between 555°F and 559°F using the Bank 1 Condenser Steam Dumps or Steamline PORVs. h.Place ROD CNTRL BANK SEL Switch in SBA.CAUTION 3.8.i 12 steps should NOT be exceeded until Rod Bottom lights are off. If all Shutdown Bank A Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered. i.Using the ROD CONTROL ROD MOTION Lever,perform the following:1)Withdraw Shutdown Bank A to ten Steps.2)Verify that all RB lights for Shutdown Bank A areoff.3)Using the ROD CONTROL ROD MOTION Lever,withdraw SBA to 230 steps.

GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 9 OF 22 INITIALS/DATE Step 3.8 continued CAUTION 3.8.j To prevent any inadvertent inward rod motion the ROD CNTRL BANK SEL Switch should not be placed in or pass through AUTO. j.Place ROD CNTRL BANK SEL Switch in SBB.CAUTION 3.8.k 12 steps should NOT be exceeded until Rod Bottom lights are off. If all Shutdown Bank B Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered. k.Using the ROD CONTROL ROD MOTION Lever,perform the following:1)Withdraw Shutdown Bank B to ten steps.2)Verify that all RB lights for Shutdown Bank Bare off.3)Using the ROD CONTROL ROD MOTION Lever,withdraw SBB to 230 steps.

GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 10 OF 22 INITIALS/DATE 3.9 Contact Reactor Engineering for recommended rod heights / and Estimated Critical Condition information. NOTE 3.10 Reactor Coolant System temperature is being maintained between 555°F and 559°F using the Bank 1 Condenser Steam Dumps or Steamline PORVs. 3.10 Perform a Shutdown Margin verification per STP-134.001, / Shutdown Margin Verification, using Estimated Critical Condition boron, desired RCS temperature, and expected xenon. STTS # NOTE 3.11 through 3.13 For initial criticality following refueling, REP-107.001, Controlling Procedure For Refueling Startup And Power Ascension Testing, is the controlling document for Reactor Startup. Appropriate steps of GOP-3 should be initialed as they are performed. 3.11 Prepare for Reactor Startup as follows: / a.Adjust Boron concentration as required by EstimatedCritical Condition calculation as follows:Z003 1)Borate or dilute per SOP-106,Z010 Reactor Makeup Water System,Z017Sections III.D, III.E, or III.F.2)When complete, direct Chemistry to samplethe RCS and the Pressurizer for boron.CHG D GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 11 OF 22 INITIALS/DATE Step 3.11 continued b.Block HIGH FLUX AT SHUTDOWN as follows:1)Disable the IPCS High Flux At Shutdown alarmfunction as follows:a)Type the Turn-On-Code HFAS.b)Verify OPERATOR DISABLED isindicated above the ENABLE CALCS box.c)If OPERATOR ENABLED is indicated,select DISABLE CALCS.2)Place HIGH FLUX AT SHUTDOWN Switch forSOURCE RANGE N-31 in BLOCK.3)Place HIGH FLUX AT SHUTDOWN Switch forSOURCE RANGE N-32 in BLOCK.4)Verify SR HI FLUX AT SHUTDN BLOCK(XCP-620 4-4) annunciator alarms.c.Review Estimated Critical Condition calculationwithin four hours prior to criticality, verifyingpredicted rod height is above the Rod InsertionLimit per Tech Spec 4.1.1.1.1.c.TimeCHG D GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 12 OF 22 INITIALS/DATE Step 3.11 continued d.Review the following for current status and limitationsfor Mode escalation:1)Removal and Restoration Log.2)Danger Tag Log.3)31 Day Surveillance Book.4)Ensure completion of Attachment II.F,Operational Mode Change PlantStartup - Entering Mode 2, of GTP-702.5)Ensure SAP-116, PLANT TRIP/SAFETYINJECTION PLANT RECOVERY, is completed,if necessary.Z011 e.Perform OAP-100.4, Communication, Attachment I,Mode Change Brief Checklist.f.Update the IPCS Plant Mode indicator to indicateMode 2 as the current Plant Mode as follows:1)Type the Turn-On-Code MODE to display thePLANT MODE CHANGE DISPLAY window2)Select the SET MODE 2 Pushbutton.3)Verify the selected Mode is displayed onthe left end of the top toolbar.g.Verify all Shutdown Bank Rods fully withdrawn within15 minutes of commencing Control Bank Rod withdrawal.TimeCHG D GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 13 OF 22 INITIALS/DATE Step 3.11 continued NOTE 3.11.h Reactor Coolant System temperature is being maintained between 555°F and 559°F using the Bank 1 Condenser Steam Dumps or Steamline PORVs. h.Obtain the Shift Supervisor's permission to commencea Reactor Startup.i.Announce Reactor Startup over the page system.j.If used, place NR-45 CHART in HI speed.k.Initiate REP-109.002, Inverse Count Rate Ratio Plot.Timel.If performing an initial cycle startup, refer toREP-107.001, Controlling Procedure ForRefueling Startup And Power Ascension Testing,for additional actions.

GOP 3 REFERENCE PAGE 1.GENERAL NOTESA.Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2.REACTOR CONTROLA.Shutdown Bank Control: 1)The Shutdown Banks must be fully withdrawn whenever reactivityadditions are being made by dilution, Xenon, Tavg, or control rods unless one of the following conditions exists: a)The RCS is borated to Cold Shutdown concentration andverified by sample.b)Tavg is 557°F and the RCS is borated to the hot, Xenon-freeconcentration and verified by sample.2)If the count rate on any source range channel increases by morethan a factor of two during any increment of Shutdown Bankwithdrawal, rod withdrawal shall be stopped and the ShutdownBank reinserted. Until Reactor Engineering has made a satisfactoryevaluation of the situation, rod withdrawal shall not resume.B. Source Range Control: 1)Source Range Counts and Digital Rod Position indication shouldbe monitored during any Shutdown and Control Bank withdrawal orinsertion.2)While in the Source Range, positive reactivity may be changed byonly one controlled method.C. Anticipate criticality anytime: 1)During rod motion.2)Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 14 OF 22 INITIALS/DATE 3.12 Achieve Reactor criticality as follows: / a.Review GOP Appendix A, Generic OperatingPrecautions, for Reactor Startup.CAUTION 3.12.b To prevent any inadvertent inward rod motion the ROD CNTRL BANK SEL Switch should not be placed in or pass through AUTO. b.Place the ROD CNTRL BANK SEL Switch in MAN.NOTE 3.12.c A stable Startup Rate of one decade per minute should NOT be exceeded. c.Using ROD CONTROL ROD MOTION lever, commenceControl Bank Rod withdrawal to ten steps on Bank A.TimeCAUTION 3.12.d 12 steps should NOT be exceeded until all Rod Bottom lights are off. If all Control Bank A Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered. d.At ten steps on Control Bank A, stop and verify:1)Bank A RB lights clear.2)ONE ROD ON BOTTOM (XCP-621 3-1)annunciator clears.3)RODS ON BOTTOM (XCP-621 3-2)annunciator clears.e.Recommence withdrawing rods while observing thatthe groups sequence properly.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 15 OF 22 INITIALS/DATE Step 3.12 continued CAUTION 3.12.f 12 steps should NOT be exceeded until all Rod Bottom lights are off. If all Control Bank B Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered.

f. At ten steps on Control Bank B, stop and verify Bank B RB lights clear. g. Recommence withdrawing rods while observing that the groups sequence properly.

h. Verify 102 step Bank Overlap between Control Bank A and Control Bank B. CAUTION 3.12.i 12 steps should NOT be exceeded until all Rod Bottom lights are off. If all Control Bank C Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered.

i. At ten steps on Control Bank C, stop and verify Bank C RB lights clear.

j. Recommence withdrawing rods while observing that the groups sequence properly.

k. Verify 102 step Bank Overlap between Control Bank B and Control Bank C.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 16 OF 22 INITIALS/DATE Step 3.12 continued CAUTION Step 3.12.l Reactor startup should be stopped and I&C notified if the CRB INSERT LMT LO-LO (XCP-621 1-1) annunciator fails to clear between 118 steps and 134 steps on Bank C. l.Verify CRB INSERT LMT LO-LO (XCP-621 1-1)annunciator clears between 118 steps and134 steps on Bank C.Steps CAUTION 3.12.m 12 steps should NOT be exceeded until all Rod Bottom lights are off. If all Control Bank D Rod Bottom lights are NOT off at ten steps, AOP-403.5, Stuck Or Misaligned Control Rod, should be entered. NOTE 3.12.m Reactor Coolant System temperature is being maintained between 555°F and 559°F using the Bank 1 Condenser Steam Dumps or Steamline PORVs. m.At ten steps on Control Bank D, stop and verifyBank D RB lights clearn.Recommence withdrawing rods while observing thatthe groups sequence properly.o.Verify the CRB INSERT LMT LO (XCP-621 1-2)annunciator clears between 138 steps and144 steps on Bank C.Steps p.Verify 102 step Bank Overlap between Control Bank Cand Control Bank D.CHG A GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 17 OF 22 INITIALS/DATE Step 3.12 continued q. Within 15 minutes before achieving criticality, verify Tavg greater than or equal to 551°F. Time Tave r. Announce criticality over the page system. Time s. Verify critical rod position is above the Rod Insertion Limit per Tech Spec 3.1.3.6.

t. Maintain as close to 0 SUR as reasonably achievable.

u. At the AUDIO COUNT RATE CHANNEL drawer, place the following switches in OFF: 1) AUDIO MULTIPLIER. 2) CHANNEL SELECTOR.

3) SR COUNTER/SCALER, POWER switch. (Toggle down)

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 18 OF 22 INITIALS/DATE 3.13 Increase Reactor Power to 10-3% as follows: / a. Establish a stable Startup Rate of less than one decade per minute. b. At 7.5x10-6%, perform the following: 1) Verify P6 Permissive energizes to bright. 2) Verify a minimum of one decade overlap between Source Range Channels and Intermediate Range Channels. c. Prior to 105 CPS, perform the following:

1) Momentarily place SR TRAIN A Switch in BLOCK. 2) Verify SR A TRIP BLCK Permissive energizes to bright. 3) Momentarily place SR TRAIN B Switch in BLOCK.

4) Verify SR B TRIP BLCK Permissive energizes to bright. d. Perform one of the following for continued monitoring of Intermediate and Power Range instrument:

1) If available for use, select one Intermediate Range Channel and one Power Range Channel on NR-45, NIS RECORDER.

2) Ensure at least one Intermediate Range and at least one Power Range instrument are selected for continuous monitoring using computer display NR45. e. Stabilize Reactor Power at 10-3%.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 19 OF 22 INITIALS/DATE Step 3.13 continued f. Record the following Critical Data:

1) RCS pressure: psig

2) Tavg: °F

3) Bank at steps

4) Boron Concentration: ppm

5) Time:

6) Stable Power:  %

g. If performing an initial cycle startup, refer to REP-107.001, Controlling Procedure For Refueling Startup And Power Ascension Testing, for physics testing instructions.

GOP 3 REFERENCE PAGE 1. GENERAL NOTES A. Procedure steps should normally be performed in sequence. However, it is acceptable to perform steps in advance after thorough evaluation of plant conditions and impact by the Shift Supervisor or Control Room Supervisor. B. At least two licensed operators, one of whom is SRO licensed, must be present in the Control Room during Reactor Startup. 2. REACTOR CONTROL A. Shutdown Bank Control: 1) The Shutdown Banks must be fully withdrawn whenever reactivity additions are being made by dilution, Xenon, Tavg , or control rods unless one of the following conditions exists: a) The RCS is borated to Cold Shutdown concentration and verified by sample. b) Tavg is 557°F and the RCS is borated to the hot, Xenon-free concentration and verified by sample. 2) If the count rate on any source range channel increases by more than a factor of two during any increment of Shutdown Bank withdrawal, rod withdrawal shall be stopped and the Shutdown Bank reinserted. Until Reactor Engineering has made a satisfactory evaluation of the situation, rod withdrawal shall not resume. B. Source Range Control: 1) Source Range Counts and Digital Rod Position indication should be monitored during any Shutdown and Control Bank withdrawal or insertion. 2) While in the Source Range, positive reactivity may be changed by only one controlled method. C. Anticipate criticality anytime: 1) During rod motion. 2) Boron dilution is in progress.

GOP-3 REVISION 13 PAGE 20 OF 22 INITIALS/DATE CAUTION 3.14 While operating with a positive Moderator Temperature Coefficient: a. All reactivity additions should be slow and controlled. b. A stable Startup Rate of 0.3 decade per minute should not be exceeded. c. Rods should be moved in 1/2 step increments until the effect of rod motion has been evaluated. NOTE 3.14 Ensure sufficient Emergency Feedwater Flow exists prior to raising power.

3.14 Increase Reactor Power to between 1% and 3%. /

3.15 At the Point of Adding Heat, if NR-45, NIS RECORDER, / had previously been selected to HI speed place the recorder in LO speed. CAUTION 3.16 a. Adjustment of Tavg with the Rod Control System must not be attempted with the ROD CNTRL BANK SEL Switch in any position other than MAN. b. Manual rod control is required to establish equilibrium conditions, since C-5 blocks automatic rod withdrawal.

3.16 Maintain Tavg between 555°F and 559°F. /

3.17 Complete Attachment II.G, Operational Mode Change / Plant Startup - Entering Mode 1, of GTP-702.

3.18 Proceed to GOP-4A, Power Operation (Mode 1 - Ascending). /

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GOP-3 REVISION 13 PAGE 21 OF 22

4.0 REFERENCES

4.1 CP-625, Chemistry Refueling Shutdown And Startup Plan.

4.2 FSAR Section 5.0.

4.3 GOP Appendix A.

4.4 GOP-4A, Power Operation (Mode 1 - Ascending).

4.5 GTP-702, Operational Mode Change and Contingency Surveillance Requirements.

4.6 OAP-100.4, Communication.

4.7 REP-107.001, Controlling Procedure For Refueling Startup And Power Ascension Testing.

4.8 REP-109.002, Inverse Count Rate Ratio Plot.

4.9 SAP-630, Procedure / Commitment Accountability Program.

4.10 SOP-103, Boron Thermal Regeneration System.

4.11 SOP-106, Reactor Makeup Water System.

4.12 SOP-114, Reactor Building Ventilation System.

4.13 SOP-201, Main Steam System.

4.14 SOP-205, Turbine Sealing Steam System.

4.15 SOP-206, Main and Auxiliary Condenser Air Removal System.

4.16 SOP-207, Circulating Water.

4.17 SOP-208, Condensate System.

4.18 SOP-209, Feedwater Turbine Lube Oil System.

4.19 SOP-210, Feedwater System.

4.20 SOP-211, Emergency Feedwater System.

4.21 SOP-212, Steam Generator Blowdown. CHG F

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GOP-3 REVISION 13 PAGE 22 OF 22 4.22 SOP-215, Main Turbine Lube Oil Supply System.

4.23 SOP-403, Rod Control And Position Indicating System.

4.24 SOP-404, Excore Nuclear Instrumentation System.

4.25 STP-134.001, Shutdown Margin Verification.

4.26 STP-345.039, Reactor Trip P-4 Trip Actuating Device Operational Test.

4.27 V.C. Summer Precautions, Limitations, and Setpoints.

4.28 V.C. Summer Reactor Engineering Procedures.

4.29 V.C. Summer Tech Specs.

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REP-109.002 ENCLOSURE 9.2 PAGE 1 OF 1 REVISION 1 Recommended Rod Positions for ICRR Bank A Bank B Bank C Bank D Purpose 0 Start 10Rod Bottom lights 53ICRR 103 ICRR 129 1 Overlap 138 10 ICRR, RB lights 181 53 ICRR 230 103 ICRR, Overlap 129 1 Overlap 138 10 ICRR, RB lights 181 53 ICRR 230 103ICRR, Overlap 118-134*LO-LO Alarm Clear 1291 Overlap 118-134*0-6 LO-LO Alarm Clear138 10 ICRR, RB lights, 138-144 10-16 LO Alarm Clear 181 53 ICRR 206 78 ICRR (If < 0.2) 230 103 ICRR, Overlap 128 ICRR (If < 0.2) 153 ICRR 178 ICRR 203 ICRR = placekeeping checkbox *LO-LO Alarm should clear in the 118-134 range and overlap should be checked at 129.

2015 NRC Sim b SRO & RO: Steam Generator Tube Rupture (Depressurize RCS to < Ruptured S/G Pressure)V.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:JPM NO:NJPSF-007APage 1 of TASK:000-038-05-01PREFERRED EVALUATION METHODPERFORMEVALUATION TIME10TIME CRITICALNoTASK STANDARD:RCS pressure is reduced to less than ruptured S/G pressure with PZR level > 10% or PZR level 76% or RCS subcooling < 52.5°F. The use of applicable Human Performance Tools (3-way communications, self checking, peer checking, phonetic alphabet, etc) and industrial safety practices meets expectations. This JPM is related to PRA event OAP2 " Depressurize RCS to stop leakage into ruptured S/G"PREFERRED EVALUATION LOCATIONSIMULATORTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)6RESPOND TO STEAM GENERATOR TUBE RUPTURETOOLS:Copy of EOP-4.0 marked for current plant conditions (tube rupture on S/G "C") up to step 2.

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TERMINATING CUE:RCS depressurization completed with task standard met and the chosen PORV Block valve closed.CANDIDATE:EOP-4.0STEAM GENERATOR TUBE RUPTUREINDEX NO.ROSROK/A NO.010000A203PORV failures4.14.2A2.03Page 2 of INITIATING CUES:Control Room Supervisor directs you as ROATC to depressurize the RCS using PZR Spray, per EOP-4.0, Step 2.INITIAL CONDITION:A Steam Generator Tube Rupture is in progress. S/G "C" has been isolated per EOP-4.0. An operator initiated cooldown has been performed according to EOP-4.0, through Step 2.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Page 3 of STEPSAttempt to depressurize the RCS using normal spray valve PCV-444D.Evaluator note: Give candidate 1-2 minutes to become familiar with control board indications and status of procedure implementation. Evaluator note: With "B" and "C" RCP secure candidate should NOT attempt to open PCV-444C. Proceds to Step 2 based on alternative action Step 2 a.YesPlaces PZR Spray Valve PCV-444D controller in MANUAL and increases output to 100% demand.

Determines that PCV-444D did not open based on Red light OFF and Green light ON for PCV-444D.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Verifies at least one PZR PORV is available.YesNotes all three PZR PORVs are available by observing Green lights ON for all PORV position indicators.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:2Page 4 of Opens one PZR PORV until any termination criteria is met; RCS pressure < 'C' (ruptured) S/G pressure and PZR level > 10%; or PZR level >76; or RCS subcooling <52.5°F.Evaluator note: Using the MCB indicators it is most likely that candidate will terminate on RCS pressure < Ruptured ('C') S/G pressure and PZR level > 10%, but if using IPCS values it is possible that they will terminate on PZR level >76. Both termination criteria occur at about the same time and terminating on either one is satisfactory.YesSelected PORV indicates Red light ON, Green light OFF. RCS pressure decreases.

Recognizes one of the following from MCB indications: -RCS pressure < 'C' S/G pressure with PZR level >10% or,-PZR level >76% or,-RCS subcooling <52.5°F.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:3Page 5 of Examiner ends JPM at this point.Closes Selected PORV.YesTakes PORV Control Switch for the Selected PORV to the CLOSE position. - Notes Selected PORV position indicates Red light ON, Green light OFF and RCS pressure still decreasing. - Candidate notes that selected PORV failed to close.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:4Stops RCS depressurization by closing block valve for associated PORV.YesPlaces associated PORV Block Valve (MVG-8000A/B/C) to close observes Red light OFF, Green light ON.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:5Page 6 of JPM NO:NJPSF-007ADESCRIPTION:2015 NRC Sim b SRO & RO: Steam Generator Tube Rupture (Depressurize RCS to < Ruptured S/G Pressure)IC SET:311INSTRUCTIONS:If IC-311 is designated for this JPM then reset to IC-311, leaving simulator in FREEZE. Set IPSC to display 2PS1 at the RO station.Mark up EOP-4.0 for current plant conditions (tube rupture on SG "C") up to step 2.When Examinee is ready, (on evaluator cue) go to RUN.If IC-311 is not designated for this JPM then initial conditions may be established by reseting to IC-10 and following the below directions:1.Insert: MAL-RCS002C Final Value = 600 Ramp = 0 (S/G Tube Rupture on 'C' S/G) MAL-PRS003B Final Value = 0 (PRESSURIZER SPRAY VALVE 444D FAILURE)2.SetEvent #1asX05i386o >0Set Event #2 as X05i387o >0Set Event #3 as X05i388o >03.Insert: VLV-RC004P Final Value = 100 (PCV-445A STUCK OPEN), set to event #1 VLV-RC001P Final Value = 100 (PCV-444B STUCK OPEN), set to event #2 VLV-RC005P Final Value = 100 (PCV-445B STUCK OPEN), set to event #3 .RUN 180 seconds.Manual SI and perform actions of EOP-1.0 & EOP-4.0 up through step ..Throttle EFW to 'C' S/G when level > 40%..FREEZE.Insert: LOA-MSS033 Position To = RACK OUT, (RACK OUT BKR FOR MVG-2802B (STM SUPPLY TO TDEFP)).RUN.Trip RCPs "B" and "C"..Perform actions of steps to step 2 of EOP-4.0.JPM SETUP SHEET, , 201Page 7 of INITIATING CUES:Control Room Supervisor directs you as ROATC to depressurize the RCS using PZR Spray, per EOP-4.0, Step 2.INITIAL CONDITION:A Steam Generator Tube Rupture is in progress. S/G "C" has been isolated per EOP-4.0. An operator initiated cooldown has been performed according to EOP-4.0, through Step 2.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEET JPM NO:NJPS-0652015 NRC Sim c RO:Establish Hot Leg Injection During Loss of RHR at Mid-Loop ConditionsV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Thursday, Page 1 of TASK:000-083-05-01PREFERRED EVALUATION METHODPERFORMEVALUATION TIME10TIME CRITICALNoTASK STANDARD:SI flow verified on FI-940, CHG LOOP A CLD/HOT LG FLOW and hot leg level increasing.PREFERRED EVALUATION LOCATIONSIMULATORTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:41(b)10RESPOND TO LOSS OF RESIDUAL HEAT REMOVAL SYSTEM WHILAT MID-LOOP CONDITIONSTOOLS:AOP-115.5 marked to match initial conditions.

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TERMINATING CUE:SI flow verified on FI-940 and hot leg level increasing.CANDIDATE:AOP-115.1RHR PUMP VORTEXINGAOP-115.5LOSS OF RHR WITH THE RCS NOT INTACT (MODES 5 AND 6)INDEX NO.ROSROK/A NO.000025K301Shift to alternate flowpath3.13.4AK3.01Page 2 of INITIATING CUES:The CRS directs you as NROATC to establish Hot Leg Injection per INITIAL CONDITION:The plant was in Mode 5 with RCS at Mid-loop conditions with the The 'A' RHR loop was the in-service loop. Due to lowering hot leg level, the Crew entered AOP-115.1 and then AOP-115.5. The present conditions are : - RCS hot leg level is off scale low. - Step 17 of AOP-115.5 has been reached and core exit TC temperatures are >200°F and increasing. - The 'B' Charging pump is in service.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Page 3 of STEPSStep 1. Check if a Charging Pump is available.Evaluator cue: Inform Examinee that a surrogate will answer all annunciators that are not related to the task.Evaluator note: Assure that SIPCS screens are set up per the JPM Setup Instructions.YesVerifies 'B' charging pump is available by observing Red light ON above pump control switch and amps indicated on ammeter.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Step 2. Stop any running Charging PumpYesStops 'B' Charging pump by placing control switch to STOP.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:2Page of Step 3. Close MVG-8107 and MVG-8108, CHG LINE ISOL.Evaluator note: At least one of the valves must be closed to satisfy step.YesPositions MVG-8107 & 8108, CHG LINE ISOLs, to closed position, Red light OFF and Green light ON for each valve.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:3Step 4.a. Ensure all of the following are closed MVG-8885, CHG LP A ALT TO COLD LEGS. MVG-8801A(B), HI HEAD TO COLD LEG INJ.YesVerifies MVG-8885, CHG LP A ALT TO COLD LEGS, indicates Red light OFF and Green light ON.Verifies MVG-8801A & B, HI HEAD TO COLD LEG INJ, indicates Red light OFF and Green light ON.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:4Page of Step 4.b. Open MVG-8884, CHG LP A TO HOT LEGS..Evaluator note: The procedure does not describe operation of the Power Lockout Switch but this action is required to get MVG-8884 to change position.YesPlaces TRN A PWR LCKOUT switch to ON. Positions MVG-8884, CHG LP A ALT TO HOT LEGS, to Open position; Red light ON and Green light OFF.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:5Step 4.c. Close MVG-8106, CHG PP, Miniflow Isolation.Evaluator note; MVG-8106 operation rels on the same power lockout switch as MVG-8884.YesPositions MVG-8106 CHG PP, Miniflow Isolation to Closed position; Red light OFF and Green light ONPlaces TRN A LCKOUT switch to OFF.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:6Page of Step 4.d. Close MVT-8105, SEAL WTR INJ ISOL.YesPositions MVT-8105, SEAL WTR INJ ISOL, to Closed position; Red light OFF and Green light ON.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:7Step 5. Start one Charging Pump.YesStarts 'B' Charging pump by placing control switch to START position. Pump indicates Red light ON and Green light OFF with normal pump amps.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:8Page of Examiner ends JPM at this point.Step 6. Verify SI flow on FI-940, CHG LOOP A CLD/HOT LG FLOW GPM.YesSI flow verified on FI-940, CHG LOOP A CLD/HOT LG FLOW.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:9Restores Hot Leg levelEvaluator cue: End JPM when level begins to increase.Evaluator note: The RCS Mid Loop Level Monitoring system LR-1330/1331 indicates in inches above the bottom of the Hot Leg and 15.5" is the desired indication. The Mansell indication is in feet elevation and 430' 10" is the desired reading.YesHot Leg level > 15.5"CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:10Page of JPM NO:NJPS-065DESCRIPTION:2015 NRC Sim c RO: Establish Hot Leg Injection During Loss of RHR at Mid-Loop ConditionsIC SET:312INSTRUCTIONS:If IC 312 is designated for this JPM then reset to IC 312, leaving the simulator in FREEZE.1.Set one SIPCS screen to HALFPIPE from the Map Menu selections and another to ZZSHTDWN,Shutdown off the ZZ Menu pad.2.Set up Mansell Level monitoring and display on CRS SIPCS screen by typing MLMSA or MLMSBfrom any SIPCS screen NOTE: NEED to verify that SIM group has enabled the Mansell Functionwith exam security set. Discussed with Jody Lawter on 6/16/14 and Sim group is aware andworking on this. If the SIPCS function is NOT enabled set up the lap top for Mansell Indication.3.When Examinee is ready (on evaluator cue) go to RUNIf IC-312 is not designated for this JPM then initial conditions may be established by reseting to IC-20 and following the below directions:1.Insert: MAL-RCS006C Final Value = 4000 (RCS Cold leg leak) OVR-AA028 Override To = True (Ann acknowledge) LOA-RCS053 Final Value = POWER_ON (Mid-loop Monitor Disconnect Switch)2.Set one SIPCS screen to HALFPIPE from the Map Menu selections and another to ZZSHTDWN,Shutdown off the ZZ Menu pad..3.Set up Mansell Level monitoring and display on CRS SIPCS screen by typing MLMSA or MLMSBfrom any SIPCS screen NOTE: NEED to verify that SIM group has enabled the Mansell Functionwith exam security set. Discussed with Jody Lawter on 6/16/14 and Sim group is aware and working on this. If the SIPCS function is NOT enabled set up the lap top for Mansell Indication.4.RUN5.Perform actions of AOP-115.1, step 1 waiting for break flow to require Alternative Action 1 d.6.Perform actions of AOP-115.5 steps 1-4 and steps 11 - 17.

7.When core exit TC temperature is >200°F, with LT1330/1331 < 15.5" and Mansell < 430' 10"modify MAL-RCS006C to 2,000.9.FREEZE8.When Examinee is ready (on evaluator cue): RUNJPM SETUP SHEETPage of COMMENTS:

INITIATING CUES:The CRS directs you as NROATC to establish Hot Leg Injection per INITIAL CONDITION:The plant was in Mode 5 with RCS at Mid-loop conditions with the The 'A' RHR loop was the in-service loop. Due to lowering hot leg level, the Crew entered AOP-115.1 and then AOP-115.5. The present conditions are : - RCS hot leg level is off scale low. - Step 17 of AOP-115.5 has been reached and core exit TC temperatures are >200°F and increasing. - The 'B' Charging pump is in service.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEET JPM NO:NJPSF-019A2015 NRC Sim d RO & SRO-U: Manually Initiate Reactor Building SprayV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Page 1 of 1 TASK:026-005-01-01PREFERRED EVALUATION METHODPERFORMEVALUATION TIME5TIME CRITICALNOTASK STANDARD:At least one train of containment spray is manually actuated with >2500 gpm per EOP-1.0 and RCPs are secured PRIOR to damaging RCP due to loss of CCW as evident from Motor Bearing temperature exceeding 195°F or Lower Seal Water Bearing temperature exceeding 225°F or SeWater Outlet temperature exceeding 235°F.PREFERRED EVALUATION LOCATIONSIMULATORTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(b)(8)MANUALLY INITIATE REACTOR BUILDING SPRAYTOOLS:EOP-1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION marked through step 7.

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TERMINATING CUE:RB Spray initiated.CANDIDATE:EOP-1.0REACTOR TRIP/SAFETY INJECTION ACTUATIONINDEX NO.ROSROK/A NO.026000A401CSS controls4.54.3A4.01Page 2 of 1 INITIATING CUES:The CRS directs you as the ROATC to perform Step 8 of EOP-1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION.INITIAL CONDITION:The reactor has tripped from 100% power and an SI has occurred.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Page 3 of 1 STEPSStep 8; Verify RB pressure has remained LESS THAN 12 psig on PR-951, RB PSIG (P-951), red penYesVerifies RB pressure >12 PSIG, moves to Alternative Action column for step 8.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Page 4 of 1 AA Step 8 a); Verify both the following annunciators are lit: XCP-612 3-2 (RB SPR ACT). XCP-612 4-2 (PHASE B ISOL). IF either annunciator is NOT lit, THEN actuate RB Spray by placing the following switches to ACTUATE: Both CS-SGA1 and CS-SGA2. OR Both CS-SGB1 and CS-SGB2.Evaluator note: These switches require two hand operation to turn both switches at once.Evaluator note: Examinee may try both trains of switches. If "A" train switches are used they will fail to work and starting individual components becomes critical. If "B" train switches are used they will cause all spray sytem functions to occur EXCEPT the Train "A" RB Spray pump discharge valve, MVG-3003A, will not automatically open and must be manually opened. The JPM becomes alternate path once the Examinee begins manual realignment actions.YesVerifies both annunciators are NOT lit.Places (CS-SGA1 and CS-SGA2) or (CS-SGB1 and CS-SGB2) to the ACTUATE position.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:2Page 5 of 1 AA Step 8 b); Verify Phase B Isolation by ensuring RB SPRAY/PHASE B ISOL monitor lights arebright on XCP-6105.Evaluator cue: If told as the SS that Phase B monitor lights are not bright on XCP-6105 then direct Examinee to ensure valves are aligned as required for Phase B.Evaluator note: If only the "A" train switches were used then PHASE B lights will not turn bright.YesPHASE B Isol monitor lights are bright on XCP-6105.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:3Page 6 of 1 AA Step 8c); Ensure the following are open: MVG-3001A(B), RWST TO SPRAY PUMP A(B) SUCT. MVG-3002A(B), NAOH TO SPRAY PUMP A(B) SUCT. MVG-3003A(B), SPRAY HDR ISOL LOOP A(B).YesMVG-3001A(B), RWST TO SPRAY PUMP A(B) SUCT. Indicates Red light ON, Green light OFF.

MVG-3002A(B), NAOH TO SPRAY PUMP A(B) SUCT. Indicates Red light ON, Green light OFF.MVG-3003A(B), SPRAY HDR ISOL LOOP A(B). Indicates Red light ON, Green light OFF.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:4Page 7 of 1 AA Step 8 d); Ensure both RB Spray Pumps are running.YesVerifies 'A' and 'B' RB Spray Pumps are running by Red light ON indication and normal running amps.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:5AA Step 8 e); Verify RB Spray flow is GREATER THAN 2500 gpm for each operating train on:FI-7368, SPR PP A DISCH FLOW GPM.FI-7378, SPR PP B DISCH FLOW GPM.YesFI-7368, SPR PP A DISCH FLOW, and FI-7378, SPR PP B DISCH FLOW, are > 2500 gpm.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:6Page 8 of 1 Examiner ends JPM at this point.AA Step 8 f); Stop all RCPs.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • Booth Operator note: Do not reset the simulator until Evaluator is satisfied that RCP temperatures were not exceeded. Provide information from SIPCS in the booth.********************************************************************************************************************YesPlaces 'A', 'B', & 'C' RCP switches in Stop; Red light OFF, Green light ON, flow decreasing and 0 running amps PRIOR to damaging RCP due to loss of CCW as evident from Motor Bearing temperature exceeding 195°F or Lower Seal Water Bearing temperature exceeding 225°F or Seal Water Outlet temperature exceeding 235°F.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:7Page 9 of 1 JPM NO:NJPSF-019ADESCRIPTION:2015 NRC Sim d RO & SRO-U: Manually Initiate Reactor Building SprayIC SET:313INSTRUCTIONS:If IC-313 is designated for this JPM then reset to IC 313 leaving the simulator in FREEZE.1.RUN, silence annunciators and FREEZE promptly.2.Silence DCS speaker.3.Set up SIPCS in Booth with RCP temperatures; Motor Bearings, Lower Seal Water Bearingand Seal Watrer Outlet.4.When Examinee is ready (on Evaluator cue) go to run.If IC-313 is not designated for this JPM then initial conditions may be established by reseting to IC-10 and following the below directions:1.Insert: LOA-PCS109 Position To = AS IS (HI-3 Channel 1 fail as is) LOA-PCS110 Position To = AS IS (HI-3 Channel 2 fail as is) LOA-PCS116 Position To = AS IS (HI-3 Channel 4 fail as is) MAL-RHR008A Reactor Building Spray Pump "A" discharge valve (3003A) fail MAL-MSS003A Final Value = 1.2E7, (Steamline break inside containment)

OVR-SG011 Override To = FALSE, (Fail RB Spray actuation switch) CS-SGA1 (Train A) OVR-SG012 Override To = FALSE, (Fail RB Spray actuation switch) CS-SGA2 (Train A) 2.Set event #1 as x02i101o = = 1 (Allows manual opening of 3003A when 101 switch taken to open)3.Insert a "new" MAL-RHR008A, set to Event #1, set Delete in = 1 second.4 .RUN until RB pressure >12 psig and ESF loading sequencer is complete (approximately 60 seconds). Leave RCPs running.5.FREEZE6.Ensure RCS pressure is greater than the 1418 psig RCP trip criteria then modify MAL-MSS003Ato final value = 1.8E67.RUN, silence annunciators and FREEZE promptly.8.Silence DCS speaker.9.Set up SIPCS in Booth with RCP temperatures; Motor Bearings, Lower Seal Water Bearingand Seal Watrer Outlet.JPM SETUP SHEETPage 10 of 1 10.When Examinee is ready (on evaluator cue): RUNCOMMENTS:Failing 1/2 RB Spray Actuation switches in a train will disable that function.*************************************************CAUTION:*******************************************************Booth Operator do NOT reset simulator until Evaluator has verified RCP Temperatures did NOT exceed critical standard of Motor Bearing > 195°F, Lower Seal Water Bearing > 225°F and Seal Water Outlet >235°F. ********************************************************************************************************************Thursday, Page 11 of 1 INITIATING CUES:The CRS directs you as the ROATC to perform Step 8 of EOP-1.0, REACTOR TRIP/SAFETY INJECTION ACTUATION.INITIAL CONDITION:The reactor has tripped from 100% power and an SI has occurred.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEET JPM NO:NJPSF-025A2015 NRC Sim e RO: Start and Load "A" Emergency Diesel GeneratorV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Page 1 of 1 TASK:064-003-01-01PREFERRED EVALUATION METHODPERFORMEVALUATION TIME15TIME CRITICALNoTASK STANDARD:The "A" Diesel Generator is tripped from the MCB upon exceeding the high lube oil temperature trip setpoint and prior to parallel. The use of applicable Human Performance Tools (3-way communications, self checking, peer checking, phonetic alphabet, etc) and industrial safety practices meets expectations.PREFERRED EVALUATION LOCATIONSIMULATORTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)8LOAD THE DIESEL GENERATORTOOLS:NJPSF-025A Handout 1; SOP-306 Section IV.A, Operation of Diesel Generator A from the Control Room in the Test Start Mode, marked up through step 2.2..

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TERMINATING CUE:"A" D/G tripped at the MCB.CANDIDATE:ARP-001-XCP-636ANNUNCIATOR RESPONSE PROCEDURE (PANEL XCP-636)SOP-306EMERGENCY DIESEL GENERATORINDEX NO.ROSROK/A NO.064000A401Local and remote operation of the ED/G4.04.3A4.01064000A101ED/G lube oil temperature and pressure3.03.1A1.01Page 2 of 1 INITIATING CUES:The CRS directs you as an RO to start and load "A" D/G per SOP-306,Section IV.A, Steps 2.2.j thru 2.5.INITIAL CONDITION:The plant is operating at 100% power.Normal and Alternate AC power available to buses 1DA and 1DB. It is an A2 Maintenance Work Week. Relay testing is in progress on 1DA which requires the removal of 1DA from both NORMAL and ALTERNATE feed. Station and Operations Management have given approval for this work due to recent OE concerning maintenance of these relays. "A" D/G is to be started and loaded onto bus 1DA. Bus 1DA will then be divorced from its NORMAL and ALTERNATE power sources until completion of testing. All pre-start check steps have been completed.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Page 3 of 1 STEPSStep 2.2 j. At the Main Control Board, perform the following:1)Depress the GEN RELAYS RESET Pushbutton.2)Momentarily place the EXCITER Switch, to RESET.3)Ensure XCP-636 1-2 (DG A AUTOSTART NOT READY) is NOT in alarm.Evaluator cue: Provide NJPSF-025A Handout 1, -306 Section IV.A marked up to the Examinee at conclusion of JPM brief.YesDG A GEN RELAYS RESET Pushbutton is depressed.DG A EXCITER Switch taken to RESET.DG A AUTOSTART NOT READY annunciator is clear.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Page of 1 Step 2.3 a. To start Diesel Generator A from the Main Control Board, perform the following:a.Ensure the diesel is ready to be started as indicated by the following:1)XCP-636 1-2 (DG A AUTOSTART NOT READY) is NOT in alarm.2)The READY FOR AUTO START Light is lit at the Diesel Generator A LocalControl Panel (DB-436).Booth Operator cue: When requested, as the IB operator, inform the Examinee that the "READY FOR AUTO START" light is ON at the "A" D/G Local Control Panel.YesVerifies DG A AUTOSTART NOT READY annunciator is clear.Calls the IB operator and verifies the "READY FOR AUTO START" light is ON at the "A" D/G Local Control Panel.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:2Page of 1 Step 2.3 b. Momentarily place the Diesel Generator A TEST Switch to START.Evaluator note: Examinee should request a peer check.YesMomentarily rotates "A" Diesel Generator TEST switch to the START position.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:3Step 2.3 c. Verify the Diesel Generator starts and stabilizes between the following:1)58.9 Hz and 61.1 Hz.2)6800 volts and 7600 volts.YesDG A VOLTS indicates 6800-7600 volts and FREQUENCY indicates 58.9 - 61.1 Hertz.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:4Page of 1 Step 2.3 d. Reset the tripped Diesel Generator 'A' relay flags at the local panel (XCX-5201, DB-436).Booth Operator cue: When requested, as the IB operator, inform the operator that the Diesel Generator 'A' relay flags have been reset.Evaluator cue: s CRS direct Examinee to load "A" DG perNote 2.4.YesCalls the IB operator and verifies the Diesel Generator 'A' relay flags are reset at the Local Control Panel.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:5Page of 1 Procedure note: Prior to closing the Diesel Generator Breaker, the Diesel should be run atno-load for at least ten minutes.Step 2.4 If the Diesel Generator is to be loaded, perform the following:a.Ensure the VOLT REG Switch is in AUTO.b.Place the DG A SYNC SEL Switch in DSL.Evaluator cue: Evaluator note: YesCommences ten minute wait Ensures VOLT REG Switch is in AUTOPlaces the DG A SYNC SEL Switch in DSL.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:6Page of 1 Contact IB Operator to identify cause of Annunciator 636, 6-3, DG A ENG TEMP TRBL when it alarms.Booth Operator cue: When contacted as IB Operator report that Lube Oil temperature locally is indicating 178°F and rising.YesCalls IB Operator with request to identify cause for Temperature alarm on "A" DG.References Annunciator response procedure for XCP 636 6-3.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:7Page of 1 Examiner ends JPM at this point.Ensure AUTOMATIC ACTIONS of Annunciator 636, 6-3 since the alarm is due to High Temp.Booth Operator cue: If called as IB operator regarding LO temp report "A" DG Lube Oil Temperaturelocally indicated 185°F and rising.Booth Operator cue: If called as IB operator to locally trip the DG reply that you left the DG roomdue to safety concerns since the Diesel was making loud noises.

Evaluator note: Examinee should recognize that "A" DG has exceeded trip setpoint and has not automatically tripped. It is critical that the Operator take action to trip the "A" EDG at this point.The JPM becomes alternate path once lube oil temperature exceeds 175°F and the auto trip has not occurred.YesOperator takes "A" DG TEST Switch to STOP.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:8Page 1 of 1 JPM NO:NJPSF-025ADESCRIPTION:2015 NRC Sim e RO: Start and Load "A" Emergency Diesel GeneratorIC SET:314INSTRUCTIONS:If IC-314 is designated for this JPM then reset to IC-314 leaving the simulator in FREEZE.1.When Examinee is ready (on Evaluator cue) go to RUN.If IC-314 is not designated for this JPM then initial conditions may be established by reseting to IC-10 and following the below directions:1.Set event #1 as x13i101d==1 (DG 'A' Synch Selector Switch in DSL) 2.Insert: ANN-DG014 Delay = 0 Fail To = ON (DG A ENG TEMP TRBL), set to event #13.When Examinee is ready; RUNCOMMENTS:JPM SETUP SHEETThursday, Page 1 of 1 INITIATING CUES:The CRS directs you as an RO to start and load "A" D/G per SOP-306,Section IV.A, Steps 2.2.j thru 2.5.INITIAL CONDITION:The plant is operating at 100% power.Normal and Alternate AC power available to buses 1DA and 1DB. It is an A2 Maintenance Work Week. Relay testing is in progress on 1DA which requires the removal of 1DA from both NORMAL and ALTERNATE feed. Station and Operations Management have given approval for this work due to recent OE concerning maintenance of these relays. "A" D/G is to be started and loaded onto bus 1DA. Bus 1DA will then be divorced from its NORMAL and ALTERNATE power sources until completion of testing. All pre-start check steps have been completed.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEET SOP-306 REVISION 19 PAGE 16 OF 136 IV.INFREQUENT OPERATIONSA. OPERATION OF DIESEL GENERATOR A FROM THE CONTROL ROOM IN THE TEST START MODE 1.0 INITIAL CONDITIONS 1.1 A Pre-Job Brief has been conducted per OAP-100.3. 1.2 The Precautions of Section II have been reviewed. 1.3 Diesel Generator A is prepared to start per Section III. 1.4 Enclosure F, Tech Spec/EOOS/Functionality Review has been reviewed. 2.0 INSTRUCTIONS 2.1 If XTF0004, UNIT 1 ENGINEERED SAFEGUARD TRANSFORMER is in service and Diesel Generator A will be paralleled to the 115KV line, perform one of the following (YD-380 SSW): NOTE 2.1.a Immediately prior to and during the time the XTF0006, XTF0004 7.2KV VOLTAGE REGULATOR, AUTO-OFF-MANUAL Switch is placed in MANUAL or OFF, the 115KV and the 7.2KV Bus voltages being supplied from XTF0006 should be monitored continuously. 115KV Bus voltage should be verified and recorded to be within the limits specified in OAP-106.1 for the present transformer configuration with the regulator out of service. If the OAP limits are exceeded, the System Controller should be notified to restore 115KV Bus voltage to within the limits. a.With XTF0006, XTF0004 7.2KV VOLTAGE REGULATOR in service,perform the following:1)Monitor ESF XFMR FEED KV (MCB) voltage and either 1DAVOLTS and/or 1DB VOLTS (MCB) Bus voltage being suppliedfrom XTF0006.2)Using the Generic Log attachment from OAP-106.1, OperatingRounds, record an initial ESF XFMR FEED KV (MCB) voltageand either 1DA VOLTS and/or 1DB VOLTS (MCB) Bus voltage and then record hourly thereafter.

SOP-306 REVISION 19 PAGE 17 OF 136 Step 2.1 continued 3)Place the XTF0006, XTF0004 7.2KV VOLTAGE REGULATOR,AUTO-OFF-MANUAL Switch, in OFF.b.With XTF0006, XTF0004 7.2KV VOLTAGE REGULATOR out ofservice, ensure XES0008, 7.2KV TRANSFER & DISC SWITCHESTransfer Switch is in the LOAD position.NOTE 2.2 a.If Diesel Generator A has been run in the last twenty-four (24) hours, Step 2.2 maybe omitted.2.2 Verify the Diesel Generator A cylinders are free of fluid as follows: a.At the Main Control Board, momentarily place the EXCITER Switch,in SHUTDN. (PEER )b.Place the REMOTE/LOCAL/MAINT Switch, in MAINT (DB-436).(PEER )c.Verify the fuel injection racks move freely (DB-436).d.Open all twelve cylinder test cocks (DB-436).e.Place the fuel rack in the NO FUEL position by one of the followingmethods (DB-436):1)Hold the Stop Lever in the STOP position.2)Place the Stop Lever in the STOP position and install the FuelRack Stop Lever Blocking Device.C01 SOP-306 REVISION 19 PAGE 18 OF 136 Step 2.2 continued CAUTION 2.2.f Personnel should stand clear of both sides of Diesel Generator A when barring the engine due to the exhaust of high pressure air from the test cocks. NOTE 2.2.f Some discharge from the cylinder test cocks, such as a spray or mist, is to be expected. Excessive discharge which results in accumulation of fluid in the area indicates a potential coolant leak in the cylinders. If excessive fluid is found in one or more cylinders, the Diesel Generator must be declared Inoperable and the appropriate actions taken. f. While observing the cylinder test cocks to detect the possibility of fluid leakage into the cylinders, bar the engine over by one of the following methods (DB-436):

1) Starting air by momentarily depressing the TEST START Pushbutton.

2) Starting air by using the spanner wrench on the top of one of the Main Air Start Valves on the engine.

3) Barring device motor.

4) Manually, using a wrench attached to the shaft end.

g. Remove the Stop Lever from STOP by one of the following methods (DB-436): 1) If the Stop Lever was held in the STOP position, release the Stop Lever from the STOP position.

2) If the Fuel Rack Stop Lever Blocking Device was installed, perform the following: a) Remove the Fuel Rack Stop Lever Blocking Device. b) Release the Stop Lever from the STOP position. h. Close all twelve cylinder test cocks (DB-436). C01 SOP-306 REVISION 19 PAGE 19 OF 136 Step 2.2 continued i. Place the REMOTE/LOCAL/MAINT Switch, in REMOTE (DB-436). (PEER ) j. At the Main Control Board, perform the following: 1) Depress the GEN RELAYS RESET Pushbutton. 2) Momentarily place the EXCITER Switch, to RESET. 3) Ensure XCP-636 1-2 (DG A AUTOSTART NOT READY) is NOT in alarm. CAUTION 2.3 through 2.7 The REMOTE/LOCAL/MAINT Switch should NOT be operated anytime the Diesel Generator is running.

2.3 To start Diesel Generator A from the Main Control Board, perform the following:

a. Ensure the diesel is ready to be started as indicated by the following: 1) XCP-636 1-2 (DG A AUTOSTART NOT READY) is NOT in alarm. 2) The READY FOR AUTO START Light is lit at the Diesel Generator A Local Control Panel (DB-436). b. Momentarily place the Diesel Generator A TEST Switch to START. (PEER )

c. Verify the Diesel Generator starts and stabilizes between the following: 1) 58.9 Hz and 61.1 Hz. 2) 6800 volts and 7600 volts. d. Reset the tripped Diesel Generator A relay flags at the local panel (XCX-5201, DB-436).

SOP-306 REVISION 19 PAGE 20 OF 136 NOTE 2.4 If time permits, the following guidelines should be utilized to achieve the desired load: a. Prior to closing the Diesel Generator Breaker, the Diesel should be run at no-load for at least ten minutes. b. Once the Diesel Generator Breaker is closed, load should be adjusted to between 850 KW and 1000 KW and maintained for at least ten minutes. c. Load should be adjusted to between 2250 KW and 2550 KW and maintained for at least ten minutes. d. Load should be adjusted to between 3250 KW and 3550 KW and maintained for at least ten minutes. e. Load should be adjusted to between 4150 KW and 4250 KW and maintained for at least ten minutes. 2.4 If the Diesel Generator is to be loaded, perform the following: a. Ensure the VOLT REG Switch is in AUTO. (PEER ) b. Place the DG A SYNC SEL Switch in DSL. (PEER ) c. Using the VOLT REG RAISE-LOWER Switch adjust Diesel Generator A SYNC VOLTS to slightly higher than 1DA SYNC VOLTS. (PEER ) d. Using the SPEED Switch, adjust Diesel Generator A frequency to cause the SYNCHROSCOPE to rotate slowly in the FAST direction (clockwise). (PEER ) e. When the SYNCHROSCOPE passes 11 o'clock and slowly approaches 12 o'clock, close BUS 1DA DG FEED Breaker. (PEER )

SOP-306 REVISION 19 PAGE 21 OF 136 Step 2.4 continued NOTE 2.4.f Limits per Enclosure B, Diesel Generator Power Factor, should be maintained. f. Using the SPEED Switch adjust load as necessary while monitoring the following: 1) KILOWATTS Meter. 2) AMPS Meters. 3) KILOVARS Meter. g. Place the DG A SYNC SEL Switch in OFF. (PEER ) h. Using the VOLT REG RAISE-LOWER Switch adjust KILOVARS. CAUTION 2.5 While operation in this configuration is not prohibited by Tech Specs, the time spent separated from Offsite Power should be limited to that required for troubleshooting.

2.5 If it is desired to divorce XSW1DA from Offsite Power, perform the following: a. Utilizing Enclosure C estimate the present load on XSW1DA. b. Using the SPEED Switch, adjust Diesel Generator A load until the estimated XSW1DA load is being carried by Diesel Generator A. c. Open one of the following as appropriate for the Offsite Power source currently in parallel with the Diesel Generator: (PEER )

1) BUS 1DA NORM FEED Breaker.

2) BUS 1DA ALT FEED Breaker. d. Using the SPEED Switch, adjust Diesel Generator A as necessary to maintain frequency between 59.5 Hz and 60.5 Hz.

SOP-306 REVISION 19 PAGE 22 OF 136 Step 2.5 continued e. Using the VOLT REG RAISE-LOWER Switch adjust Diesel Generator A as necessary to maintain voltage between 6800 VAC and 7600 VAC. f. When time permits, perform the following:

1) Direct I&C to connect a Fluke 45 DMM to the back of Main Control Board meter DG A VOLTS (V-DGA) with the following settings (inside MCB): a) AC volts. b) AUTO. c) Medium rate. 2) Using the VOLT REG RAISE-LOWER Switch, adjust Diesel Generator A as necessary to maintain voltage between 114.67 VAC and 122.90 VAC by Fluke 45 indication connected at the MCB (between 6880.1 VAC and 7373.8 VAC).

2.6 If the Diesel Generator Breaker is closed and Diesel Generator A is no longer required as a source of power, perform one of the following:

a. If the Diesel Generator is the only power source supplying XSW1DA, perform the following to parallel with Offsite Power: 1) Place the DG A SYNC SEL Switch in one of the following positions as appropriate: (PEER )

a) NORM - allows paralleling with the 115 KV offsite source.

b) EMERG - allows paralleling with the 230 KV offsite source. 2) Using the VOLT REG RAISE-LOWER Switch, adjust Diesel Generator A 1DA SYNC VOLTS to slightly lower than SYNC VOLTS. (PEER ) 3) Using the SPEED Switch adjust Diesel Generator A frequency to cause the SYNCHROSCOPE to rotate slowly in the SLOW direction (counter-clockwise). (PEER )

SOP-306 REVISION 19 PAGE 23 OF 136 Step 2.6 continued 4) When the SYNCHROSCOPE indicator passes 1 o'clock and slowly approaches 12 o'clock, close one of the following as appropriate for the synchroscope position selected: (PEER ) a) BUS 1DA NORM FEED Breaker.

b) BUS 1DA ALT FEED Breaker. 5) Place the DG A SYNC SEL Switch in OFF. (PEER ) NOTE 2.6.b If time permits, the following guidelines should be utilized to unload the Diesel Generator: 1) Load should be reduced to between 2150 KW and 2550 KW and maintained for three to five minutes. 2) Load should be reduced to between 850 KW and 1250 KW and maintained for three to five minutes. 3) Load should be reduced to 50 KW.

b. If the Diesel Generator is running in parallel with an Offsite Power source, perform the following: 1) Unload Diesel Generator A by holding the SPEED Switch in LOWER until load is 50 KW. 2) Using the VOLT REG RAISE-LOWER Switch, reduce KILOVARS to minimum. 3) Open BUS 1DA DG FEED Breaker. (PEER ) 4) Ensure DG A VOLTS indicates between 6800 volts and 7600 volts. 5) Momentarily place the EXCITER Switch, in SHUTDN. (PEER )

SOP-306 REVISION 19 PAGE 24 OF 136 Step 2.6.b continued NOTE 2.6.b.6) The VOLT REG RAISE-LOWER Switch should not be adjusted for the remainder of this procedure. 6) Verify the steady-state, no-load, voltage for Diesel Generator A as follows: a) Momentarily depress the EMERG START Pushbutton. (PEER ) b) Verify DG A VOLTS indicates between 6800 volts and 7600 volts. c) Momentarily depress the EMERG START OVRRIDE Pushbutton. (PEER ) d) Momentarily place the Diesel Generator A TEST Switch, in START. (PEER )

2.7 To return Diesel Generator A to standby status perform the following: a. Momentarily place the EXCITER Switch in SHUTDN. (PEER ) b. Momentarily place the TEST Switch in STOP. (PEER ) c. Unless otherwise directed, prepare Diesel Generator A for automatic/manual operation by performing the appropriate steps of Section III.

2.8 If Diesel Generator A has been run for greater than or equal to an hour, perform the following steps to check for and remove any accumulated water in XTK0020A DG, DG FUEL OIL DAY TANK A (DB-436): a. If required, install a drain hose between XVT30957-DG, HI ISOL VLV FOR TEST CONNECTION, and a suitable container. b. Throttle open XVT30957-DG, HI ISOL VLV FOR TEST CONNECTION. c. Unlock and throttle open XVT00990A-DG, DG FUEL OIL DAY TANK A DRAIN VALVE.

SOP-306 REVISION 19 PAGE 25 OF 136 Step 2.8 continued d. When Diesel Generator A Day Tank is free of water, perform the following: 1) Close XVT00990A-DG, DG FUEL OIL DAY TANK A DRAIN VALVE. 2) Lock XVT00990A-DG, DG FUEL OIL DAY TANK A DRAIN VALVE. 3) Close XVT30957-DG, HI ISOL VLV FOR TEST CONNECTION. 4) If necessary, remove the drain hose from XVT30957-DG, HI ISOL VLV FOR TEST CONNECTION. NOTE 2.9 a. XTF0005, UNIT 2 ENGINEERED SAFEGUARD TRANSFORMER, must be in standby prior to placing XTF0006, XTF0004 7.2KV VOLTAGE REGULATOR, in AUTO. b. If the Band Indicator HIGH or LOW light is lit, the Voltage Regulator will step immediately when placed in AUTO. 2.9 If AUTO operation is desired, place the XTF0006, XTF0004 7.2KV VOLTAGE REGULATOR, AUTO-OFF-MANUAL Switch in AUTO and stop recording hourly Bus Voltage readings (YD-380 SSW). 2.10 If previously installed, direct I&C to disconnect the Fluke 45 DMM from V-DGA (inside MCB) END OF SECTION JPM NO:NJPS-10002015 NRC Sim f RO: Respond to Steam Generator Pressure Channel FailureV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Thursday, August 28, 2014Page 1 of 1 TASK:000-103-05-01PREFERRED EVALUATION METHODPERFORMEVALUATION TIME10TIME CRITICALTASK STANDARD:PREFERRED EVALUATION LOCATIONSIMULATORTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:10CFR55.4Respond to Excessive Feedwater Increase per AOP-401.3TOOLS:AOP-401.3, Steam Flow - Feedwater Flow Protection Channel Failure

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TERMINATING CUE:SG "B" level is under control and CANDIDATE:SOP-401REACTOR PROTECTION AND CONTROL SYSTEMAOP-401.3STEAM FLOW-FEEDWATER FLOW PROTECTION CHANNEL FAILUREINDEX NO.ROSROK/A NO.059000A211Failure of feedwater control system3.03.3A2.11Thursday, August 28, 2014Page 2 of 1 INITIATING CUES:Respond to developing plant conditions.INITIAL CONDITION:The plant is operating at 100% power with all controls in automatic.READ TO OPERATOR:SAFETY CONSIDERATIONS:NoneWHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Thursday, August 28, 2014Page 3 of 1 STEPSStep 1. Verify the failed channel is the controlling channel.Evaluator cue: hen Examinee has accepted turnover and has completed board walk down signal Booth Operator to activate Event #1. Evaluator note: This is an immediate operator action.YesFI-484 indicates ~ 5 MPPH, PI-485 indicates ~ 1300 psig. Examinee notes FI-484 is the controlling channel.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Thursday, August 28, 2014Page 4 of 1 Step 2. Select the operable flow channel: Place FW CONTROL CHANNEL SEL Switch to the operable channel. Place STEAM CONTROL CHANNEL SEL Switch to the operable channel.Evaluator note: This is an immediate operator action.

Evaluator note: This step is critical to remove the failed channel from control and to restore "B" SG level to program value (60-65%).YesPlaces FW CONTROL CHANNEL SEL switch and STEAM CONTROL CHANNEL SEL switch to the opposite position.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Thursday, August 28, 2014Page 5 of 1 Step 3. Verify turbine load LESS THAN 950 MWE.Evaluator note: This is an immediate operator action.Procedure note: CTRL+ALT+S on either EHC HMI is equivalent to 50 Me, and is the method to accomplish a rapid load reduction.YesNotes > 950 MWE and per Alternative Action 3, using any available method, reduces turbine load by 40 MWE to 50 MWE.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:3Step 4. Verify only one SG AFFECTED.Evaluator note: This is an immediate operator action.Evaluator note: The failed channel does input to Feedwater Pump Speed Control and thus may have some slight impact on "A" and "C" SG level as well as "B" SG. It is not expected that there will be any appreciable impact on "A" or "C" SG levels.YesVerifies only "B" SG affected.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:4Thursday, August 28, 2014Page 6 of 1 Step 5. Adjust the Feedwater Flow Control Valve as necessary to restore feed flow to the AFFECTEDSG.Evaluator note: This is an immediate operator action.Evaluator note: Typically this step does not require any operator action once an operable channel is selected. Examinee may place Feedwater Flow Control Valve in manual and lower flow to obtain program SG level.

Evaluator note: This step is critical if the examinee is slow in selecting the operable SF and FF channels and SG NR level has exceeded 70% (approaching Hi Hi level Turbine trip)YesManually controls the SG "B" FWC controller as necessary to restore SG "B" level.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:5Thursday, August 28, 2014Page 7 of 1 Step 6. Check if Feedwater Pump speed control is operating properly: Feedwater Header pressure is GREATER THAN Main Steam Header pressure. Feed flow is normal for steam flow and power level. All operating Feedwater Pump speeds and flows are balanced.Evaluator note: This is an immediate operator action.Evaluator cue: As CRS provide a copy of AOP-401.3 and direct Examinee to complete AOP-401.3 steps 7 through 9. Examinee should reference the AOP for remaining actions. Evaluator note: The failed Steam Pressure Channel affects the controlling Steam Flow Channel which in turn feeds into the program value for Main Feedwater Pump Delta P. Once the examinee has selected the non-failed Steam Flow channel the program Delta P will return to normal and Main Feed Pump speed should restore to normal without any Operator Action. Examinee may place Main Feedpump Speed Control in manual and lower FW Flow and FW Header pressure to obtain program SG level.YesVerifies:-FW Header Pressure > Main Steam Header Pressure.-FW flow is normal.-All operating FWP speeds and flows are balanced.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:6Thursday, August 28, 2014Page 8 of 1 Step 7. Verify Narrow Range levels in all SGs are between 60% and 65%. Alternative Action 7. Adjust the Feedwater Flow Control Valve as necessary to restore feed flow to the AFFECTED SG(s).Evaluator note: It is not expected that candidate will need to take manual control of the Feedwater Flow Control Valve or Feedwater Pump Speed control for this failure.YesRestores and maintains "B" SG level to between 60% and 65% in Manual as necessary.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:7Step 8. Restore the AFFECTED SG control systems to normal: Place the Feedwater Flow Control Valve in AUTO. Place the Feedwater Pump Speed Control System in AUTO. REFER TO SOP-210, FEEDWATER SYSTEM.Evaluator note: It is expected that the Feedwater Flow Control Valve and the Feedpump Speed Control will have remained in automatic for this failure.YesEnsures Feedwater Control Valve is in automatic and Feedwater Pump Speed Control is in automaticCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:8Thursday, August 28, 2014Page 9 of 1 Examiner ends JPM at this point.STEP:9Step 9 a. Identify the associated bistables for the failed channel. REFER TO AOP 401.3, Attachment 1.Evaluator Cue: Have examinee identify the failed channel by pointing out the correct instrument number (PT-485) on AOP-401.3 Attachment 1. Evaluator note: The task of completing the bistable tripping data sheet (SOP-401 Attachment 1) is performed by SROs with a Shift Engineer review.YesExaminee identifies instrument PT-485 (Compensates FT-484).CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoThursday, August 28, 2014Page 10 of 1 JPM NO:NJPS-1000DESCRIPTION:2015 NRC Sim f RO: Respond to Steam Generator Pressure Channel FailureIC SET:315INSTRUCTIONS:If IC 315 is designated for this JPM then reset to IC-315 leaving the simulator in FREEZE.1.When Examinee is ready (on Evaluator cue) go to RUN2.On evaluator cue activate Event #1If IC 315 is not designated for this JPM then initial conditions may be established by reseting to IC 10 and following the below directions:1.Insert: MAL-MSS001E Final Value = 1300, Ramp = 3 sec (SG PT 485 Fail) set to Event #12.When Examinee is ready: RUN3.On evaluator cue activate Event #1COMMENTS:JPM SETUP SHEETThursday, August 28, 2014Page 11 of 1 INITIATING CUES:Respond to developing plant conditions.INITIAL CONDITION:The plant is operating at 100% power with all controls in automatic.SAFETY CONSIDERATIONS:NoneOPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEET JPM NO:NJPS-0842015 NRC Sim g RO: Restore Spent Fuel Pool Level During RefuelingV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Page 1 of 1 TASK:000-140-05-01PREFERRED EVALUATION METHODPERFORMEVALUATION TIME10TIME CRITICALNoTASK STANDARD:Spent Fuel Pool Level greater than or equal to 460 ft 6 inches on LI-7431 and LI-7433.PREFERRED EVALUATION LOCATIONSIMULATORTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)(7)RESPOND TO DECREASING WATER LEVEL IN THE SPENT FUEL POOL OR REFUELING CAVITYTOOLS:AOP-123.1 marked up through step 10.

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TERMINATING CUE:Spent Fuel Pool Level greater than or equal to 460 ft 6 inches on LI-7431and LI-7433.CANDIDATE:AOP-123.1DECREASING LEVEL IN THE SPENT FUEL POOL OR REFUELING CAVITY DURING REFUELINGINDEX NO.ROSROK/A NO.033000A203Abnormal spent fuel pool water level or loss of water level3.13.5A2.03Page 2 of 1 INITIATING CUES:The CRS has directed you as the ROATC, to respond to a decreasing level in the Spent Fuel Pool in accordance with AOP-123.1 starting with Step 10.INITIAL CONDITION:The Plant is in MODE 6 with Core Off Load in Progress. The 'A' RHR Loop is in service providing Core Cooling. AOP-123.1 has been entered due to decreasing level in the Spent Fuel Pool. The leakage was isolated in step 8.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Page 3 of 1 STEPSStep 10 a: Check if the operating RHR train is intact.Booth Operator cue: If examinee calls building operator regarding status of "A" RHR Train, report that it has been verified as intact.YesVerifies normal pump amps and flow on 'A' RHR pump.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Step 10 b: Open MVG8809A(B), RWST TO RHR PP A(B).Evaluator note:This step is critical because opening MVG-8809A assures that the "A" RHR pump has a suction source and that RHR can add inventory to the RCS.YesOpens MVG-8809A, RWST TO RHR PP A, and verifies Red light ON and Green light OFF.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Page 4 of 1 Step 10 c: Close MVG8701A(B), RCS LP A(C) TO PUMP A(B).Evaluator note: This step is critical because closing MVG-8701A or MVG-8702A assures that the RHR suction source is the RWST and not the Refueling Cavity.YesCloses MVG-8701A, RCS LP A TO PUMP A and verifies Red light OFF and Green light ONCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:3Step 10 d: Close MVG8702A(B), RCS LP A(C) TO PUMP A(B).Evaluator note: This step is critical because closing MVG-8701A or MVG-8702A assures that the RHR suction source is the RWST and not the Refueling Cavity.YesCloses MVG-8702A, RCS LP A TO PUMP A and verifies Red light OFF and Green light ONCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:4Page 5 of 1 Step 10 e: Close HCV603A(B), A(B) OUTLET.YesCloses HCV-603A, A OUTLET, by turning to zero (0).CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:5Step 10 f: Adjust FCV605A(B), A(B) BYP, as necessary to establish the desired refueling water level.YesTakes manual control of FCV-605A and controls flow to raise SFP level.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:STEP:6Page 6 of 1 Step 10 g: Ensure the associated Component Cooling train is operating. REFER TO SOP118, COMPONENT COOLING WATER.YesVerifies 'A' CCW pump is running with normal pump amps and flow. Verifies 'A' CCW is the active loop by verifying MVB-9524A/9526A, LP A NON-ESSEN LOAD ISOL, and MVB-9687A/9525A, LP A NON-ESSEN LOAD ISOL, are open, and by verifying MVB-9524B/9526B, LP B NON-ESSEN LOAD ISOL, and MVB-9687B/9525B, LP B NON-ESSEN LOAD ISOL, are closedCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:7Step 10 h: Verify CCW flow through the RHR Pump A(B) Seal Cooler: FM7245, A RHR PP (IUR14400, M3/SW 53). FM7255, B RHR PP (IUR14401, M4/SW 53).YesVerifies flow indicated on FM7245, flow recorder for RHR Pump "A"CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:8Page 7 of 1 Step 11: Locally align the Spent Fuel Cooling System to fill the Spent Fuel Pool via Spent Fuel Cooling Pump B. REFER TO SOP123, SPENT FUEL COOLING SYSTEM.Booth Operator cue: Inform Examinee as the building operator that you acknowledge the order to align Spent Pool Fuel Cooling to fill the Spent Fuel Pool.YesDirects building operator to align Spent Fuel Cooling to fill the SFP from the RWST per SOP-123.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:9STEP:10Verify that Refueling Cavity AND Spent Fuel Pool level is recovering.Booth Operator cue: If examinee calls a building operator to verify Refueling Cavity Level, report that it is rising provided Examinee has observed Spent Fuel Pool level as rising on LI-7431/7433.YesVerifies that Spent Fuel Pool level on LI-7431/7433 is increasing.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoPage 8 of 1 Examiner ends JPM at this point.STEP:11Check if Refueling Cavity AND Spent Fuel Pool level is adequate:Refueling Cavity level is GREATER THAN OR EQUAL TO 460 ft 6 inches.Fuel Pool level is GREATER THAN OR EQUAL TO 460 ft 6 inchesBooth Operator cue: If examinee calls a building operator to verify Refueling Cavity Level, report that it is 460 ft 6 inches once Spent Fuel Pool level is observed as 460 ft 6 inches on LI-7431/7433.YesVerifies that Spent Fuel Pool level on LI-7431/7433 is GREATER THAN OR EQUAL TO 460 ft 6 inches.Verifies that Refueling Cavity level on LI-7403 and or Mansell is GREATER THAN OR EQUAL TO 460 ft 6 inchesCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoPage 9 of 1 JPM NO:NJPS-084DESCRIPTION:2015 NRC Sim g RO: Restore Spent Fuel Pool Level During RefuelingIC SET:316INSTRUCTIONS:Will also need to make sure that Mansell indication can be employed in exam mode. If Mansell is OK in exam mode then include directions for set up and incorporate into body of JPM. If IC-316 is designated for this JPM then reset to IC-316 leaving the simulator in FREEZE.1.Select 'shutdown' from ZZMENU on MCB1 IPCS screen.2.Set up the Mansell level monitoring laptop and if available load MLMS on SIPCS.

3.Place red tags on RB spray pumps, PZR Back up heaters4.When Examinee is ready (on Evaluator cue) go to RUN.If IC-316 is not designated for this JPM then initial conditions may be established by reseting to IC-379 and followingthe below directions:1.Insert: MAL-RHR005A Final Value = 3000 (RHR bypass line leak)OVR-AA028 Override To = True (Override Radiation Monitoring Panel Annunciators)2.Verify: LOA-FHB001 Final Value = 1 (Fuel Transfer Tube Isolation - Open) LOA-FHB002 Final Value = 1 (Spent Fuel Gate to Transfer Tube - Open)3.RUN 4.When refueling cavity/SFP levels indicate < 460 feet, FREEZE5.Set MAL-RHR005A Final Value = 106.Select 'shutdown' from ZZMENU on MCB1 IPCS screen7.Set up the Mansell level monitoring laptop and if available load MLMS on SIPCS.8.Place red tags on RB spray pumps, PZR Back up heaters and whatever else Steve thinks wouldbe appropriate.9.When Examinee is ready (on Evaluator cue): RUNJPM SETUP SHEETPage 10 of 1COMMENTS:

INITIATING CUES:The CRS has directed you as the ROATC, to respond to a decreasing level in the Spent Fuel Pool in accordance with AOP-123.1 starting with Step 10.INITIAL CONDITION:The Plant is in MODE 6 with Core Off Load in Progress. The 'A' RHR Loop is in service providing Core Cooling. AOP-123.1 has been entered due to decreasing level in the Spent Fuel Pool. The leakage was isolated in step 8.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEET JPM NO:NJPS-10012015 NRC Sim h RO: Establish Reactor Building Purge Supply and ExhaustV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Friday, January 16, 2015Page 1 of 1 TASK:088-505-01-04PREFERRED EVALUATION METHODPERFORMEVALUATION TIME30TIME CRITICALNOTASK STANDARD:The RB Purge System is in service with both Purge Exhaust Fans and no more than one Purge Supply Fan started.PREFERRED EVALUATION LOCATIONSIMULATORTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:10CFR55:45(a)(8)Perform Line ups of the Reactor Building Ventilation Systems.TOOLS:NJPS-1001 Handout 1; Marked up copy of SOP-114, Reactor Building Ventilation SystemNJPS-1001 Handout 2; HPP-709 Attachment VI, Reactor Building Purge Release PermitCopy of HPP-709, Sampling and Release of Radioactive Gaseous Effluents.

Four yellow plastic Test in progress tags for Plant Status labeling on RM-A2 and RM-A4.

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TERMINATING CUE:The RB purge system is in service.CANDIDATE:OAP-100.5GUIDELINES FOR CONFIGURATION CONTROL AND OPERATION OF PLANT EQUIPMENTHPP-709Sampling and Release of Radioactive Gaseous EffluentsSOP-114REACTOR BUILDING VENTILATION SYSTEMINDEX NO.ROSROK/A NO.029000A201Maintenance or other activity taking place inside containment2.93.6A2.01Friday, January 16, 2015Page 2 of 1EXAMINER:SIGNATUREDATE INITIATING CUES:You are being directed to place Reactor Building Purge in service using SOP-114, Reactor Building Ventilation System Section III.C . All applicable procedure Initial Conditions are completed.INITIAL CONDITION:The plant is in Mode 5 with preparations for a refueling outage in The equipment hatch is open. The RB atmosphere sample analysis has been completed.The RM-A2 and RM-A4 setpoints have been adjusted for this release and source checks are completed on both channels.Reactor Building Purge had been in service but was shutdown on the previous shift.READ TO OPERATOR:SAFETY CONSIDERATIONS:NoneWHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Friday, January 16, 2015Page of 1 STEPSStep 2.1; Ensure RMA0004, ATM GASEOUS IODINE-RB PURGE EXHAUST (gas channel) is in service (Rad Monitoring Panel).Evaluator cue: Once the examinee acknowledges the initiating cue provide them the marked copy of SOP-114, Reactor Building Ventilation System (NJPS-1001 Handout 1) and HPP-709, Attachment VI Reactor Building Purge Release Permit (NJPS-1001 Handout 2).Evaluator cue: Provide yellow plastic "Test in Progress" tags for Examinee to place on RM-A2 and RM-A4 pump switches and Gas channel power supply switches. This activity is described in procedure note 2.0 and tag usage is described in OAP-100.5.YesEnsures RMA-4 is in service by checking for power and indication at Rad Monitoring Panel.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Friday, January 16, 2015Page of 1 Step 2.2; If core alterations are in progress, ensure RMA0002, ATM GASEOUS IODINE RB SAMPLE LINE (gas channel), is in service (Rad Monitoring Panel).Evaluator note: Since Unit is NOT in Mode 6 and no core alterations are in progress Steps 2.2 (Check of RMA-2) is N/A. Examinee may mark step complete as RM-A2 is in service and will be required in service once core alterations begin.YesMarks step N/A and proceeds to step 2.3.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:2Friday, January 16, 2015Page of 1 Step 2.3; If RB atmosphere sample analysis dictates, place RB Charcoal Cleanup System in service per Section III.a.XFN-66A, FAN A (RB CHAR CLEANUP).b.XFN-66B, FAN B (RB CHAR CLEANUP).Evaluator cue: Inform Examinee as HP that RB atmosphere sample analysis does NOT dictate Charcoal Cleanup.YesMarks step N/A and proceeds to step 2.4.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:3Friday, January 16, 2015Page of 1 Step 2.4; Align RMA0004 sample point for RB Purge Exhaust Fan operation as follows (AB-485):a.Open XVA00006-AH, RMA0004 SAMPLE INLET ISOLATION VALVE.b.Close XVA00005-AH, RMA0004 SAMPLE INLET ISOLATION VALVE.Booth Operator cue: As Building Operator acknowledge request for sample valve alignment or verification and report task completed per the request. Use time compression for response.Evaluator note: Since Purge had previously been in service the building operator may only be asked to verify sample valve alignment correct.YesCalls Building operator and directs OPEN XVA-6-AH and CLOSE XVA-5-AH.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:4Friday, January 16, 2015Page of 1 Step 2.5; Ensure the following radiation monitors high radiation alarm setpoints are adjusted per Reactor Building Purge Release Permit:a.RMA0002, ATM GASEOUS IODINE RB SAMPLE LINE.b.RMA0004, RB PURGE EXH GAS ATMOS MONITOR.Evaluator cue: Provide simulated Reactor Building Purge Release Permit, HPP-709, Attachment VI, (NJPS-1001 Handout) if not already done. Evaluator note: The simulator does not model setpoint changes. Provide following cue. Evaluator cue: s CRS state that the alarm setpoints have been verified by another RO.YesEnsures setpoint on RMA-2 and RMA-4 at Rad monitor panel match the Purge Release Permit values.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:5Friday, January 16, 2015Page of 1 Step 2.6; Prior to placing RB Purge System in operation for the first time during an outage, perform STP-130.005B, AH Valve Operability Testing (Mode 5).Evaluator note: Initiating cue provided information that RB Purge had been in service previously.YesMarks step N/A and proceeds to step 2.6.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:6Friday, January 16, 2015Page of 1 Step 2.7; Unlock and open the following:a.XVB00001A-AV2-AH, IA HDR ISOLATION VLV FOR XVB00001A-AH (FB-479).b.XVB00001B-AV2-AH, IA HDR ISOLATION VLV FOR XVB00001B-AH (RB-463).c.XVB00002B-AV2-AH, IA HDR ISOLATION VALVE FOR XVB00002B-AH (RB-463).d.XVB00002A-AV2-AH, IA HDR ISOLATION VLV FOR XVB00002A-AH (FB-479).Booth Operator cue: As Building Operator acknowledge requests for valve alignment or verification and report task completed per the request. Use time compression for response.Evaluator note: Since Purge had previously been in service the building operator may only be asked to verify air header isolation valve alignment correct.YesCalls Building operators and directs opening XVB-1A(B) and XVB-2A(B) air header isolations.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:7Friday, January 16, 2015Page 1 of 1 Step 2.8; Ensure a Reactor Building Purge Release Permit has been issued per HPP-709.Evaluator note: The NJPS-1001 Handout 2 (Release Permit) already indicates a SAT source check.Evaluator note: Have a Copy of HPP-709 available for Examinee to refer to if they ask for it.YesEnsures permit is current and less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> old.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:8Step 2.9; Start Reactor Building Purge as follows:a.Open PVB-2A, CNTMT EXH ISOLEvaluator note: This task is critical in order to ensure that the Reactor Building atmosphere is exchanged with fresh air.YesPlaces control switch to OPEN and holds in OPEN until Red light ON and Green light OFF.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:9Friday, January 16, 2015Page 1 of 1 Procedure NOTE 2.9.b; If both trains of RB Purge are to be run, both exhaust fans should be started simultaneously.Step 2.9 b; Hold PVB-2B, CNTMT EXH ISOL, to OPEN while simultaneously holding one or both of the following fan control switches in the START position: 1)XFN-13A, EXH FAN A.2)XFN-13B, EXH FAN B.Evaluator cue: As CRS direct that both Exh Fan A (XFN-13A) and Exh Fan B (XFN-13B) should be started. Provide surrogate operator to manipulate whichever fan switch the Examinee directs. Examinee should ask for a peer check.Evaluator note: This task is critical in order to ensure that the Reactor Building atmosphere is exchanged with fresh air. Completion of the Purge Release Permit Data is NOT critical.YesPlaces control switch for PVB-2B to OPEN and Holds in OPEN. Places control switches for both XFN-13A and XFN-13B in START and holds in START until Red light ON and Green light OFF for fans and PVB-2B.CompletesSection II, Actual Release Data on Purge Release Permit:1.Release Start Date and Time (current date and time)2.Start Readings on RM-A2 and RM-A4 in cps.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:10Friday, January 16, 2015Page 1 of 1 Step 2.9 c; Verify the following:1)XFN-13A(B)-AH inlet damper opens.2)XFN-13A(B)-AH outlet damper opens.Evaluator note: Fan inlet and outlet damper indications are on the mimic board above the control switches.YesVerifies White light ON for the INLET and OUTLET damper for the Fan that was started (XFN-13A or XFN-13B)CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:11Step 2.9 d; Open XDP-28, INTAKE DMPR.YesPlaces control Switch for XDP-28 to OPEN and verifies Red light ON and Green light OFF.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:12Friday, January 16, 2015Page 1 of 1 Step 2.9 e; Open the following:1)PVB-1A, CNTMT SPLY ISOL.2)PVB-1B, CNTMT SPLY ISOL.YesPlaces control Switch for PVB-1A to OPEN and holds until Red light ON and Green light OFF.Places control Switch for PVB-1A to OPEN and holds until Red light ON and Green light OFF.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:13Friday, January 16, 2015Page 1 of 1 Examiner ends JPM at this point. Step 2.9 f; Start one or both of the following, as necessary:1)XFN-11A, SPLY FAN A.2)XFN-11B, SPLY FAN B.Evaluator cue: If asked as CRS which supply fan to start, state "Operate supply fans as required by SOP-114". Evaluator note: There is a procedure note prior to the step which starts the Purge Supply Fans. The note informs the Operator that in order to maintain a negative pressure on the RB with the Equipment Hatch open, fewer Supply Fans than Exhaust Fans should be operated. In this case no supply fans or one supply fan should be started.Evaluator note: Since the equipment hatch is open no more than ONE supply fan should be started. This step is critical because the Examinee must maintain negative pressure on the RB.YesStarts no more than ONE Purge Supply Fan.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:14Friday, January 16, 2015Page 1 of 1 JPM NO:NJPS-1001DESCRIPTION:2015 NRC Sim h RO: Establish Reactor Building Purge Supply and ExhaustIC SET:317INSTRUCTIONS:If IC 317 is designated for this JPM then reset to IC-317 leaving the simulator in FREEZE. 1.When Examinee is ready (on Evaluaor cue) go to RUN.If IC 317 is not designated for this JPM then initial conditions may be established by reseting to IC 3 and following the below directions:1.RUN2.Perform the following at the HVAC Control Panel, XCP-6210:Place 101 switch for RB Purge Supply Fan A in STOPPlace 101 switch for RB Purge Supply Fan B in STOPPlace 101 switch for RB Purge Exhaust Fan A in STOPPlace 101 switch for RB Purge Exhaust Fan B in STOPPlace 101 switch for XDP-28, Intake Damper to CLOSE3.FREEZE4.When examinee is ready: RUNCOMMENTS:JPM SETUP SHEETFriday, January 16, 2015Page 1 of 1 INITIATING CUES:You are being directed to place Reactor Building Purge in service using SOP-114, Reactor Building Ventilation System Section III.C . All applicable procedure Initial Conditions are completed.INITIAL CONDITION:The plant is in Mode 5 with preparations for a refueling outage in The equipment hatch is open. The RB atmosphere sample analysis has been completed.The RM-A2 and RM-A4 setpoints have been adjusted for this release and source checks are completed on both channels.Reactor Building Purge had been in service but was shutdown on the previous shift.SAFETY CONSIDERATIONS:NoneOPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEET HPP-0709 ATTACHMENT VI PAGE 1 OF 1 REVISION 12 GWRP No. REACTOR BUILDING PURGE RELEASE PERMIT 6-INCH (any mode) 36-INCH (modes 5, 6 or defueled) I. RELEASE AUTHORIZATION (Count Room)Estimated release Duration, (date/time) From: To: Maximum release rate, (cfm) RM-A2 gas channel, (cpm) BKG. *Alarm SetpointRM-A4 gas channel, (cpm) BKG. Alarm Setpoint *Do not adjust when in modes 1-4. RMA-2 is part of the Leak Detection System.Comments: Count Room: Date/Time: II.ACTUAL RELEASE DATA (Operations)Release Approved, SS/CRS: Date/Time: Release START, Date/Time: RM-A2 (cpm) RM-A4 (cpm) INITIALS Alarm Set Point (cpm) Source Check Sat/Unsat Reading @ Release Start (cpm) Reading @ End of Release (cpm) Alarm Set Point returned to 2 x ni Daily Verification of High Radiation Alarm Setpoint: Date/Time RM-A2 gas channel (cpm) RM-A4 gas channel (cpm) Release Rate, (cfm) Comment: Release TERMINATED, Date/Time: _____ Reason: Release Completed Authorization Expired RM-A4 High Radiation Alarm, (cpm): Release Continued on New Permit ____________________________________________ Other Operations Review: Date/Time: Updated by: Date/Time: Count Room SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION NUCLEAR OPERATIONS NUCLEAR OPERATIONS COPY NO. SYSTEM OPERATING PROCEDURE SOP-114 REACTOR BUILDING VENTILATION SYSTEM REVISION 21 SAFETY RELATED RECORD OF CHANGES CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE CHANGE LETTER TYPE CHANGE APPROVAL DATE CANCELLATION DATE A P 02/25/14 B P 07/27/14 CONTINUOUS USE Continuous Use of Procedure Required. Read Each Step Prior to Performing.

SOP-114 PAGE i REVISION 21 TABLE OF CONTENTS SECTION PAGE I. PURPOSE/SCOPE 1 II. PRECAUTIONS 1 III. NORMAL OPERATIONS 3 A. OPERATING REACTOR BUILDING COOLING 3 UNITS B. OPERATING REACTOR BUILDING CHARCOAL 6 CLEANUP UNITS C. STARTUP REACTOR BUILDING PURGE 7 SUPPLY AND EXHAUST D. SHUTDOWN REACTOR BUILDING PURGE 11 SUPPLY AND EXHAUST E. SWAPPING REACTOR BUILDING PURGE SUPPLY 14 AND EXHAUST FANS OR STARTING A SECOND REACTOR BUILDING PURGE SUPPLY OR EXHAUST FAN F. STARTUP AND SHUTDOWN REACTOR 15 BUILDING REFUELING WATER SURFACE VENTILATION G. STARTUP AND SHUTDOWN REACTOR 17 COMPARTMENT COOLING FANS H. STARTUP SECONDARY COMPARTMENT 19 COOLING FANS I. SHUTDOWN SECONDARY COMPARTMENT 21 COOLING FANS J. FILL AND VENT CONTROL ROD DRIVE 22 MECHANISM COOLING WATER SYSTEM SOP-114 PAGE ii REVISION 21 TABLE OF CONTENTS (cont.) SECTION PAGE K. STARTUP OF THE CONTROL ROD DRIVE MECHANISM 26 SHROUD VENTILATION COOLING FANS L. STARTUP CONTROL ROD DRIVE MECHANISM 28 COOLING WATER SYSTEM M. SHUTDOWN (AND DRAIN IF LONG TERM) 30 CONTROL ROD DRIVE MECHANISM COOLING SYSTEM N. SWAPPING CRDM SHROUD EXHAUST FANS 36 O. RAISING REACTOR BUILDING PRESSURE 38 USING NORMAL PRESSURE CONTROL P. LOWERING REACTOR BUILDING PRESSURE 40 USING NORMAL PRESSURE CONTROL Q. STARTUP REACTOR BUILDING ALTERNATE 43 PURGE SUPPLY AND EXHAUST R. SHUTDOWN REACTOR BUILDING ALTERNATE 46 PURGE SUPPLY AND EXHAUST S. OPERATING DIGITAL ROD POSITION 48 INDICATION CABINET COOLING SYSTEM T. SWAPPING INSERVICE CRDM COOLING 50 SPRAY AND COOLING WATER PUMPS IV. INFREQUENT OPERATIONS 52 A. FILLING CONTROL ROD DRIVE MECHANISM 52 COOLING SYSTEM EXPANSION TANK FROM DEMINERALIZED WATER B. FILLING CONTROL ROD DRIVE MECHANISM 53 COOLING SYSTEM EXPANSION TANK FROM FILTERED WATER C. SUPPLYING NITROGEN TO OPERATE REACTOR 54 BUILDING PURGE SUPPLY AND EXHAUST VALVES SOP-114 PAGE iii REVISION 21 TABLE OF CONTENTS (cont.) SECTION PAGE D. SHIFTING REACTOR BUILDING COOLING UNITS 60 FAN SPEED V. OFF NORMAL CONDITIONS 63 A. LOSS OF ANY REACTOR BUILDING 63 VENTILATION FAN B. HIGH TEMPERATURE IN DIGITAL ROD 65 POSITION INDICATION CABINET C. HIGH TEMPERATURE ON REACTOR 66 BUILDING CHARCOAL CLEANUP FILTER PLENUM OR REACTOR BUILDING PURGE EXHAUST FILTER PLENUM VI. REFERENCES 68 ENCLOSURES Enclosure A - System Information ATTACHMENTS Attachment IA - Reactor Building Ventilation System Valve Lineup Attachment IB - CRDM Cooling Water System Valve Lineup Attachment IIA - Reactor Building Ventilation System Electrical Lineup Attachment IIB - CRDM Cooling Water System Electrical Lineup Attachment III - Reactor Building Ventilation System Instrument Lineup Attachment IV - Reactor Building Ventilation System Control Panel Lineup Attachment V - CRDM Shutdown and Cooling Drain Electrical Lineup SOP-114 REVISION 21 PAGE 1 OF 68 I. PURPOSE/SCOPE 1. This procedure outlines the steps involved in operating the Reactor Building Ventilation System.

2. 10CFR50 Appendix B, 10CFR50.59, SAP-630 and the ODCM apply to this procedure.

3. 10CFR50.65a(4) applies to Section IV.C.

II. PRECAUTIONS

1. The following valves must be open for PI-8254, RB NR PRESS PSI, to accurately indicate Reactor Building (RB) pressure: a. SVX-6054, RB NR PRESS CNTMT ISOL.

b. SVX-6050A, POST ACCID H2 LOOP A (IRB). 2. RB cooling fans must be operated during plant heatup. This is to limit concrete temperature around Reactor Vessel supports and nozzles to less than 150°F.

3. RBCU fan starting duties for Normal Speed Fan motors are as follows: a. Cold, two immediate restarts are allowed. b. Hot, one immediate restart allowed. c. Subsequent restart is allowed after 30 minutes minimum run time.

d. Subsequent restart is allowed after 60 minutes minimum with motor standing idle. e. Two bump starts are considered 1 normal start.

4. CRDM Shroud Ventilation Cooling Fans must be operated to limit temperature in vicinity of CRDMs to less than 170°F. In order to accomplish this, the fans must be operated if Control Rod Drive Mechanisms are energized, or if RCS temperature is greater than 170°F.

5. DRPI Data Cabinet Cooling System should be operated continuously to limit space temperature in the cabinet area to less than 95°F to prevent acceleration in the aging process of the system electronic components. Operation during excursions above this temperature will diminish the life expectancy of the system electronic components.

SOP-114 REVISION 21 PAGE 2 OF 68 6. CRDM Cooling Water System is normally operating to cool the exhaust air from the CRDM's and reduce RB heat load. 7. When running RB Purge System, if only one RB Purge Supply Fan is running, then only one RB Purge Exhaust Fan should be running, unless RB Equipment Hatch is open, then two RB Purge Exhaust Fans may be run.

8. To prevent overflow into Refueling Cavity ventilation ductwork due to manometer effect when fuel transfer tube is open, level in Spent Fuel Pool and Refueling Cavity must be monitored closely when shifting Reactor Building Ventilation.

9. Chemistry should be contacted prior to draining from any of the following locations (YD-125'E):

a. XVT07514A-AC, HIGH ROOT TO IPX05576.

b. XVT07514B-AC, HIGH ROOT TO IPX05578.

c. XVT07514C-AC, HIGH ROOT TO IPX05582.

d. XVT07514D-AC, HIGH ROOT TO IPX05584.

e. XVT07515A-AC, HIGH ROOT TO IPX05577.

f. XVT07515B-AC, HIGH ROOT TO IPX05579.

g. XVT07515C-AC, HIGH ROOT TO IPX05583.

h. XVT07515D-AC, HIGH ROOT TO IPX05585.

i. East Manway cover.

j. West Manway cover. 10. XSW1A3 - 1C3 Tie Breaker closure affects CRDM Shroud Exhaust fan operation as follows:

a. If both XFN0067A and XFN0067B are available, then regardless of operating fan combination, XFN0067C and XFN0067D will be tripped and XFN0067A and XFN0067B will start when tie breaker is closed.

b. If either XFN0067A or XFN0067B is not available, then regardless of operating fan combination, neither XFN0067C nor XFN0067D will be tripped and XFN0067A and XFN0067B will not be started when tie breaker is closed. OA 9218 SOP-114 REVISION 21 PAGE 3 OF 68 III. NORMAL OPERATIONS Z063 A. OPERATING REACTOR BUILDING COOLING UNITS 1.0 INITIAL CONDITIONS 1.1 Electrical lineup is complete per Attachment IIA. 1.2 Control Panel lineup is complete per Attachment IV. 2.0 INSTRUCTIONS 2.1 Ensure one of the following is in-service and aligned to provide cooling water to the RBCUs: a. Industrial Cooling Water per SOP-125,Section III.A, "Startup and Operation of the Industrial Cooling Water System." b. Service Water per SOP-117.Section IV.E, "Supplying Service Water to the Train A Reactor Building Cooling Units," or Section IV.F, "Supplying Service Water to the Train B Reactor Building Cooling Units." NOTE 2.2 a. Due to eddy current brakes, RBCU control switches must be held in START position until the red breaker closed light is lit and starting current is indicated on appropriate meter. b. Normal and preferred lineup is three RBCUs running in NORM (fast speed). c. To increase stay times for teams entering containment, four RBCUs may be placed in service in NORM (fast speed).

2.2 Place RBCUs in service by starting three or four RBCUs in SLOW or NORM as follows: a. For XFN0064A-AH, REACTOR BLDG COOLING UNIT 1A EMERG FAN, start one of the following: 1) XFN 0064A-AH, 1A NORM.

2) XFN 0064A-AH, 1A SLOW. CHG B CHG B SOP-114 REVISION 21 PAGE 4 OF 68 Step 2.2 continued b. For XFN0064B-AH, REACTOR BLDG COOLING UNIT 1B EMERG FAN, start one of the following: 1) XFN 0064B-AH, 1B NORM. 2) XFN 0064B-AH, 1B SLOW. c. For XFN0065A-AH, REACTOR BLDG COOLING UNIT 2A EMERG FAN, start one of the following: 1) XFN 0065A-AH, 2A NORM.

2) XFN 0065A-AH, 2A SLOW. d. For XFN0065B-AH, REACTOR BLDG COOLING UNIT 2B EMERG FAN, start one of the following: 1) XFN 0065B-AH, 2B NORM.

2) XFN 0065B-AH, 2B SLOW. NOTE 2.2.e If RBCU fan motor amps exceed the values given, PSE should be contacted to evaluate. e. Verify RBCU Fan motor amps return to normal operating range: 1) For fast speed operation, 250 amps to 280 amps.

2) For slow speed operation, 55 amps to 70 amps. CHG B SOP-114 REVISION 21 PAGE 5 OF 68 Step 2.2 continued NOTE 2.2.f The RBCU TRAIN A (B) EMERG switch must be selected to an operable RBCU. f. Verify the following switches are in the desired position: 1) XFN-64A/XFN 65A - RBCU TRAIN A EMERG. 2) XFN-64B/XFN 65B - RBCU TRAIN B EMERG. 2.3 Shut down RBCUs by placing appropriate switch(es) in STOP: a. XFN 0064A-AH, 1A NORM. b. XFN 0064A-AH, 1A SLOW.

c. XFN 0064B-AH, 1B NORM.

d. XFN 0064B-AH, 1B SLOW.

e. XFN 0065A-AH, 2A NORM.

f. XFN 0065A-AH, 2A SLOW.

g. XFN 0065B-AH, 2B NORM.

h. XFN 0065B-AH, 2B SLOW. END OF SECTION SOP-114 REVISION 21 PAGE 6 OF 68 B. OPERATING REACTOR BUILDING CHARCOAL CLEANUP UNITS 1.0 INITIAL CONDITIONS 1.1 Electrical lineup is complete per Attachment IIA. 1.2 Control Panel lineup is complete per Attachment IV. 1.3 Fire Service System is in service per SOP-509. NOTE 2.0 All operations are performed at HVAC Control Panel, XCP-6210. 2.0 INSTRUCTIONS 2.1 Place RB Charcoal Cleanup Units in service by placing one or both of the following in START: a.XFN-66A, FAN A (RB CHAR CLEANUP).b.XFN-66B, FAN B (RB CHAR CLEANUP). 2.2 Shut down running RB Charcoal Cleanup Units by placing the following in STOP: a.XFN-66A, FAN A (RB CHAR CLEANUP).b.XFN-66B, FAN B (RB CHAR CLEANUP).END OF SECTION SOP-114 REVISION 21 PAGE 7 OF 68 C. STARTUP REACTOR BUILDING PURGE SUPPLY AND EXHAUST 1.0 INITIAL CONDITIONS 1.1 The Plant is in Mode 5, Mode 6, or defueled. 1.2 Valve lineup is complete per Attachment IA. 1.3 Electrical lineup is complete per Attachment IIA. 1.4 Control Panel lineup is complete per Attachment IV. 1.5 Fire Service System is in service per SOP-509. NOTE 2.0 a. Unless otherwise noted, all operations are performed at HVAC Control Panel, XCP-6210. b. Plant Status Labels should be placed on radiation monitor equipment required for RB ventilation operability. 2.0 INSTRUCTIONS 2.1 Ensure RMA0004, ATM GASEOUS IODINE-RB PURGE EXHAUST (gas channel) is in service (Rad Monitoring Panel). 2.2 If core alterations are in progress, ensure RMA0002, ATM GASEOUS IODINE RB SAMPLE LINE (gas channel), is in service (Rad Monitoring Panel). 2.3 If RB atmosphere sample analysis dictates, place RB Charcoal Cleanup System in service per Section III. CHG B SOP-114 REVISION 21 PAGE 8 OF 68 2.4 Align RMA0004 sample point for RB Purge Exhaust Fan operation as follows (AB-485): a. Open XVA00006-AH, RMA0004 SAMPLE INLET ISOLATION VALVE. b. Close XVA00005-AH, RMA0004 SAMPLE INLET ISOLATION VALVE.

2.5 Ensure the following radiation monitors high radiation alarm setpoints are adjusted per Reactor Building Purge Release Permit: a. RMA0002, ATM GASEOUS IODINE RB SAMPLE LINE. b. RMA0004, RB PURGE EXH GAS ATMOS MONITOR. 2.6 Prior to placing RB Purge System in operation for the first time during an outage, perform STP-130.005B, AH Valve Operability Testing (Mode 5). NOTE 2.7 RB Purge Supply and Exhaust Isolation Valves may only be opened in Mode 5 or Mode 6 per Technical Specifications 3.6.1.7.

2.7 Unlock and open the following: a. XVB00001A-AV2-AH, IA HDR ISOLATION VLV FOR XVB00001A-AH (FB-479). b. XVB00001B-AV2-AH, IA HDR ISOLATION VLV FOR XVB00001B-AH (RB-463-300-62). c. XVB00002B-AV2-AH, IA HDR ISOLATION VALVE FOR XVB00002B-AH (RB-463-015-62). d. XVB00002A-AV2-AH, IA HDR ISOLATION VLV FOR XVB00002A-AH (FB-479).

SOP-114 REVISION 21 PAGE 9 OF 68 2.8 Ensure a Reactor Building Purge Release Permit has been issued per HPP-709. 2.9 Start Reactor Building Purge as follows: a. Open PVB-2A, CNTMT EXH ISOL. NOTE 2.9.b If both trains of RB Purge are to be run, both exhaust fans should be started simultaneously. b. Hold PVB-2B, CNTMT EXH ISOL, to OPEN while simultaneously holding one or both of the following fan control switches in the START position: (Peer ) 1) XFN-13A, EXH FAN A. 2) XFN-13B, EXH FAN B. c. Verify the following: 1) XFN-13A(B)-AH inlet damper opens. 2) XFN-13A(B)-AH outlet damper opens. d. Open XDP-28, INTAKE DMPR. e. Open the following: 1) PVB-1A, CNTMT SPLY ISOL. 2) PVB-1B, CNTMT SPLY ISOL. N01 SOP-114 REVISION 21 PAGE 10 OF 68 NOTE 2.9.f 1) If both trains of RB Purge are to be run, both supply fans should be started simultaneously. 2) When the Reactor Building Equipment Hatch is open, only one RB Purge supply fan should be in operation. 3) When equipment hatch is open, negative pressure will be maintained in the RB by operating fewer supply fans than exhaust fans. 4) Both supply fans may be secured with the equipment hatch open and two exhaust fans running to control radiological conditions. f. Start one or both of the following, as necessary: 1) XFN-11A, SPLY FAN A. 2) XFN-11B, SPLY FAN B. g. Verify the following: 1) XFN-11A(B)-AH inlet damper opens fully. 2) XFN-11A(B)-AH outlet damper opens fully. 3) The following selected fan(s) starts: a) XFN-11A, SPLY FAN A.

b) XFN-11B, SPLY FAN B. 2.10 Perform STP-118.006 REACTOR BUILDING PURGE SUPPLY AND EXHAUST VALVE TEST if this is the first time RB Purge is started for the current Refueling Outage. 2.11 When RB atmosphere sample analysis allows, remove the RB Charcoal Cleanup System from service per Section III. END OF SECTION N01 SOP-114 REVISION 21 PAGE 11 OF 68 D. SHUTDOWN REACTOR BUILDING PURGE SUPPLY AND EXHAUST 1.0 INITIAL CONDITIONS 1.1 The RB Purge Supply and Exhaust System is in operation per Section III. NOTE 2.0 Unless otherwise noted, all operations are performed at HVAC Control Panel, XCP-6210. 2.0 INSTRUCTIONS 2.1 Stop one of the following: a. XFN-11A, SPLY FAN A.

b. XFN-11B, SPLY FAN B. 2.2 Verify the following: a. XFN-11A(B)-AH inlet damper is closed. b. XFN-11A(B)-AH outlet damper is closed. 2.3 Remove the remaining supply fan from service per Steps 2.1 and 2.2, if required.

2.4 Close the following: a. PVB-1A, CNTMT SPLY ISOL. b. PVB-1B, CNTMT SPLY ISOL. 2.5 Close XDP-28, INTAKE DMPR. 2.6 Stop the following running exhaust fans: a. XFN-13A, EXH FAN A.

b. XFN-13B, EXH FAN B.

SOP-114 REVISION 21 PAGE 12 OF 68 2.7 Verify the following: a. XFN-13A(B)-AH inlet damper is closed. b. XFN-13A(B)-AH outlet damper is closed. 2.8 Remove the remaining exhaust fan from service per Steps 2.6 and 2.7, if required.

2.9 Close the following: a. PVB-2A, CNTMT EXH ISOL. b. PVB-2B, CNTMT EXH ISOL.

2.10 Align RMA0004 sample point for RB Alternate Purge operation as follows (AB-485): a. Open XVA00005-AH, RMA0004 SAMPLE INLET ISOLATION VALVE. b. Close XVA00006-AH, RMA0004 SAMPLE INLET ISOLATION VALVE.

2.11 Close the following instrument air isolation valves for the purge exhaust valves: a. XVB00001A-AV2-AH, IA HDR ISOLATION VLV FOR XVB00001A-AH (FB-479). b. XVB00002A-AV2-AH, IA HDR ISOLATION VLV FOR XVB00002A-AH (FB-479). c. XVB00001B-AV2-AH, IA HDR ISOLATION VLV FOR XVB00001B-AH (RB-463-300-62). d. XVB00002B-AV2-AH, IA HDR ISOLATION VALVE FOR XVB00002B-AH (RB-463-015-62).

SOP-114 REVISION 21 PAGE 13 OF 68 NOTE 2.12 Performance of STP-118.004 is not required for temporary shutdowns of Reactor Building Purge Supply and Exhaust during Mode 5. 2.12 Perform STP-118.004, Reactor Building Purge Isolation Verification, prior to entry into Mode 4. END OF SECTION SOP-114 REVISION 21 PAGE 14 OF 68 E. SWAPPING REACTOR BUILDING PURGE SUPPLY AND EXHAUST FANS OR STARTING A SECOND REACTOR BUILDING PURGE SUPPLY OR EXHAUST FAN 1.0 INITIAL CONDITIONS 1.1 One train of RB Purge is in operation per Section III. 2.0 INSTRUCTIONS 2.1 Shut down RB Purge Fan(s) as required per Section III. 2.2 Start up RB Purge Fan(s) as required per Section III. END OF SECTION N01 N01 SOP-114 REVISION 21 PAGE 15 OF 68 F. STARTUP AND SHUTDOWN REACTOR BUILDING REFUELING WATER SURFACE VENTILATION 1.0 INITIAL CONDITIONS 1.1 One of the following is in service per Section III: a. XFN-13A, EXH FAN A. b. XFN-13B, EXH FAN B. 1.2 Electrical lineup is complete per Attachment IIA. 1.3 Control Panel lineup is complete per Attachment IV.

NOTE 2.0 All operations are performed at local control panel XPN-5170 (RB-463-340-48). 2.0 INSTRUCTIONS 2.1 Start up RB Refueling Water Surface Ventilation Fans as follows: a. Place the control switch for RB REFUELING WTR SURFACE FANS in ON.

b. Verify the following fans start: 1) XFN0007A-AH, RB REFUELING WTR SURFACE SUPPLY FAN A. 2) XFN0007B-AH, RB REFUELING WTR SURFACE SUPPLY FAN B. 3) XFN0008-AH, RB REFUELING WTR SURFACE EXH FAN.

SOP-114 REVISION 21 PAGE 16 OF 68 2.2 Shut down RB Refueling Water Surface Ventilation Fans as follows: a. Place the control switch for RB REFUELING WTR SURFACE FANS in OFF.

b. Verify the following fans stop: 1) XFN0007A-AH, RB REFUELING WTR SURFACE SUPPLY FAN A. 2) XFN0007B-AH, RB REFUELING WTR SURFACE SUPPLY FAN B. 3) XFN0008-AH, RB REFUELING WTR SURFACE EXH FAN. END OF SECTION SOP-114 REVISION 21 PAGE 17 OF 68 G. STARTUP AND SHUTDOWN REACTOR COMPARTMENT COOLING FANS 1.0 INITIAL CONDITIONS 1.1 Electrical lineup is complete per Attachment IIA. 1.2 Control Panel lineup is complete per Attachment IV. CAUTION 2.0 To minimize starting current, running fan should be shutdown for a minimum of 30 seconds prior to starting an idle fan. NOTE 2.0 a. All operations are performed at HVAC Control Panel, XCP-6210. b. Each fan can supply 100% of required flow. 2.0 INSTRUCTIONS 2.1 Start up Reactor Compartment Cooling Fan by placing one of the following in START: a. XFN-9A, FAN A.

b. XFN-9B, FAN B. N01 SOP-114 REVISION 21 PAGE 18 OF 68 CAUTION 2.2 and 2.3 During Modes 1 through 3 maintaining at least one Reactor Compartment Cooling Fan in operation limits the temperature of the concrete around the Reactor Vessel 2.2 Shut down running Reactor Compartment Cooling Fan by placing the following in STOP: a. XFN-9A, FAN A.

b. XFN-9B, FAN B. 2.3 If directed by the Shift Supervisor to secure Reactor Compartment Cooling when RCS temperature is greater establish the following conditions within sixteen hours of securing Reactor Compartment cooling flow: a. Establish natural circulation flow by blocking the Incore Pit door (RB-101) fully open with an Engineering approved device within sixteen hours securing Reactor Compartment Cooling flow. b. Monitor performance of Excore Detectors to ensure that increasing temperature in the Incore area will not affect Excore Detectors.

c. Monitor Incore Instrument Chase temperature: 1) T9332, RB INCORE INSTRUMENT CHASE TEMP. 2) T9336, RB INCORE INSTRUMENT CHASE TEMP. d. If conditions do not allow opening door to Incore Pit (RB-101), Power Operations may continue for up to twelve hours, after which a shutdown must commence to place the plant in Mode 4 within the next twenty hours. END OF SECTION SOP-114 REVISION 21 PAGE 19 OF 68 H. STARTUP SECONDARY COMPARTMENT COOLING FANS 1.0 INITIAL CONDITIONS 1.1 Electrical lineup is complete per Attachment IIA. 1.2 Control Panel lineup is complete per Attachment IV. CAUTION 2.0 To minimize starting current, running fans in each compartment should be shut down for a minimum of 30 seconds prior to starting an idle fan for that compartment. NOTE 2.0 All operations are performed at HVAC Control Panel, XCP-6210. 2.0 INSTRUCTIONS 2.1 Start up Steam Generator Compartment A Cooling Fans by simultaneously placing two of the following in START: a. XFN-68A, SG A FAN A.

b. XFN-68B, SG A FAN B.

c. XFN-68C, SG A FAN C. 2.2 Start up Steam Generator Compartment B Cooling Fans by simultaneously placing two of the following in START: a. XFN-69A, SG B FAN A.

b. XFN-69B, SG B FAN B.

c. XFN-69C, SG B FAN C. N01 SOP-114 REVISION 21 PAGE 20 OF 68 2.3 Start up Steam Generator Compartment C Cooling Fans by simultaneously placing two of the following in START: a. XFN-70A, SG C FAN A. b. XFN-70B, SG C FAN B.

c. XFN-70C, SG C FAN C. END OF SECTION SOP-114 REVISION 21 PAGE 21 OF 68 I. SHUTDOWN SECONDARY COMPARTMENT COOLING FANS 1.0 INITIAL CONDITIONS 1.1 The Secondary Compartment Cooling Fans are in service per Section III. NOTE 2.0 a. All operations are performed at HVAC Control Panel, XCP-6210. b. Two Secondary Compartment Cooling Fans are required for each loop during normal operation. 2.0 INSTRUCTIONS 2.1 Shut down Steam Generator Compartment A Cooling Fans by placing the following in STOP as necessary: a. XFN-68A, SG A FAN A.

b. XFN-68B, SG A FAN B.

c. XFN-68C, SG A FAN C. 2.2 Shut down Steam Generator Compartment B Cooling Fans by placing the following in STOP as necessary: a. XFN-69A, SG B FAN A.

b. XFN-69B, SG B FAN B.

c. XFN-69C, SG B FAN C. 2.3 Shut down Steam Generator Compartment C Cooling Fans by placing the following in STOP as necessary: a. XFN-70A, SG C FAN A.

b. XFN-70B, SG C FAN B.

c. XFN-70C, SG C FAN C. END OF SECTION SOP-114 REVISION 21 PAGE 22 OF 68 J. FILL AND VENT CONTROL ROD DRIVE MECHANISM COOLING WATER SYSTEM 1.0 INITIAL CONDITIONS 1.1 System is shutdown and fully or partially drained. 2.0 INSTRUCTIONS 2.1 Perform system lineup per Attachment IB.

2.2 Energize and open the following Containment Isolation valves (MCB): a. MVG-7501, TO CRDM CLR ISOL (ORB). (XMC1DA2X 06AD). b. MVG-7502, TO CRDM CLR ISOL (IRB). (XMC1DB2X 07IM). c. MVG-7503, FR CRDM CLR ISOL (IRB). (XMC1DA2X 11IM). d. MVG-7504, FR CRDM CLR ISOL (ORB). (XMC1DB2X 07AD). CAUTION 2.3 through 2.6 CRDM Cooling Water System pressure should be monitored continuously when XVG07520-AC, CRDM COOLING WATER RETURN HDR FILL VLV, is open to prevent lifting system relief valves inside containment. 2.3 Throttle open XVG07520-AC, CRDM COOLING WATER RETURN HDR FILL VLV (YD-125'E), to raise system pressure to a maximum of 25 psig as indicated on any of the following (YD-125'E): a. IPI05590-AC, AC SYS INDUSTRIAL CLR OUTLET PRESS IND.

b. IPI05592-AC, AC SYS INDUSTRIAL CLR INLET PRESS IND.

c. IPI05569-AC, AC SYS INDUSTRIAL CLR INLET PRESS IND.

d. IPI05570-AC, AC SYS INDUSTRIAL CLR OUTLET PRESS IND.

SOP-114 REVISION 21 PAGE 23 OF 68 2.4 Vent XCI0004-AC, CRDM COOLING WATER INDUSTRIAL COOLER, at the following points. a. XVT07514A-AC, HIGH ROOT TO IPX05576 (YD-125'E). b. XVT07514B-AC, HIGH ROOT TO IPX05578 (YD-125'E). c. XVT07514C-AC, HIGH ROOT TO IPX05582 (YD-125'E). d. XVT07514D-AC, HIGH ROOT TO IPX05584 (YD-125'E). e. XVT07515A-AC, HIGH ROOT TO IPX05577 (YD-125'E). f. XVT07515B-AC, HIGH ROOT TO IPX05579 (YD-125'E). g. XVT07515C-AC, HIGH ROOT TO IPX05583 (YD-125'E). h. XVT07515D-AC, HIGH ROOT TO IPX05585 (YD-125'E). i. XVT07547-AC, CRDM COOLER AC SUPPLY HEADER VENT VALVE (IB-436).

2.5 Vent Spray Pump B and D casings from the following vents (YD-125'E): a. XVM07570B-AC, XPP0156B PUMP CASING VENT VALVE. b. XVM07570D-AC, XPP0156D PUMP CASING VENT VALVE.

2.6 Vent XTK0151-AC, CRDM COOLING WATER EXPANSION TANK (IB-436), as follows: a. Open XVG17513-AC, AC SYSTEM VENT ISOLATION VALVE (IB-436). b. As water appears in lower portion of ILI05552, CRDM COOLING WATER EXP TANK LEVEL IND, close XVG17513-AC, AC SYSTEM VENT ISOLATION VALVE (IB-436). c. When expansion tank is 1/2 full, close XVG07520-AC, CRDM COOLING WATER RETURN HDR FILL VLV (YD-125'E). 2.7 For remainder of venting process, open XVG07520-AC, CRDM COOLING WATER RETURN HDR FILL VLV (YD-125'E), as necessary to maintain expansion tank level between 1/4 full and 1/2 full.

SOP-114 REVISION 21 PAGE 24 OF 68 2.8 Vent CRDM supply and return headers at the following points: a. XVT07543-AC, INDUSTRIAL CLR AC RETURN HDR TEST CONN (RB-436-120-60). b. XVT07540-AC, CRDM COOLER AC SUPPLY HEADER TEST CONN (RB-436-135-60).

2.9 Vent XCE0021-AH, CONTROL ROD DRIVE MECHANISM COOLER, at the following points: a. XVT07508A-AC, HIGH ROOT TO IPX05561 (RB-436-120-40). b. XVT07508B-AC, HIGH ROOT TO IPX05560 (RB-436-120-40). c. XVT07508C-AC, HIGH ROOT TO IPX05563 (RB-436-110-45). d. XVT07508D-AC, HIGH ROOT TO IPX05562 (RB-436-110-45).

2.10 Upon completion of venting, establish normal expansion tank pressure and level as follows (IB-436): a. Throttle open XVG17513-AC, AC SYSTEM VENT ISOLATION VALVE, until IPI05553-AC, AC PUMPS SUCTION HEADER PRESSURE IND, indicates 6 psig to 8 psig. b. Open XVT17514-AC, AC SYSTEM EXPANSION TANK DRAIN ISOLATION VALVE, as necessary to maintain expansion tank level between 1/4 full and 1/2 full. 2.11 Perform the electrical lineup per Attachment IIB. CAUTION 2.12 If the operating pump cavitates or loses suction pressure, the pump should be stopped immediately to prevent internal damage. 2.12 Select a pump by placing XPP157A AND B, CRDM COOLING WATER PUMP, in A or B position. 2.13 Run selected pump for one minute.

SOP-114 REVISION 21 PAGE 25 OF 68 2.14 Place XPP157A AND B, CRDM COOLING WATER PUMP, in OFF. NOTE 2.15 If no air is vented while performing Steps 2.3 through 2.14, then Step 2.15 may be omitted. 2.15 Repeat Steps 2.4 through 2.14 until all air has been vented from system. 2.16 Verify the following (IB-436): a. IPI05553-AC, AC PUMPS SUCTION HEADER PRESSURE IND, indicates 6 psig to 8 psig. b. Expansion Tank is between 1/4 full and 1/2 full. CAUTION 2.17 If the operating pump cavitates or loses suction pressure, the pump should be stopped immediately to prevent internal damage. 2.17 Select the other pump by placing XPP157A AND B, CRDM COOLING WATER PUMP, in A or B position. 2.18 Monitor selected pump for signs of cavitation. 2.19 If it appears air is still present, shut down pump and perform Steps 2.3 through 2.14, as necessary. 2.20 Ensure the valves identified on Attachment IB, NOTE 1 are closed and capped. 2.21 Contact I&C to align IFI05575-AC, INDUSTRIAL CLR AC RETURN HDR FLOW IND, per Attachment III. END OF SECTION SOP-114 REVISION 21 PAGE 26 OF 68 K. STARTUP OF THE CONTROL ROD DRIVE MECHANISM SHROUD VENTILATION COOLING FANS 1.0 INITIAL CONDITIONS 1.1 Control Panel lineup is complete per Attachment IV. CAUTION 2.0 CRDM Shroud Ventilation Cooling Fans must be operated to limit temperature in vicinity of CRDMs to less than 170°F. 2.0 INSTRUCTIONS CAUTION 2.1 a. XFN-67A, FAN A, and XFN-67B, FAN B should NOT be paired together as this causes unacceptably high vibration levels in XFN-67A, FAN A. b. XFN-67C, FAN C, and XFN-67D, FAN D should NOT be paired together as this causes unacceptably high vibration levels in XFN-67C, FAN C. NOTE 2.1 For each fan started, the switch must be held in START for five seconds or until the associated RUN light is energized. 2.1 At XCP-6210, HVAC Control Panel, start one of the following pairs of CRDM Shroud Exhaust fans:

a. XFN-67A, FAN A, and XFN-67D, FAN D (preferred combination).

b. XFN-67B, FAN B, and XFN-67C, FAN C (preferred combination).

c. XFN-67A, FAN A, and XFN-67C, FAN C.

d. XFN-67B, FAN B, and XFN-67D, FAN D.

e. XFN-67A, FAN A, and XFN-67B, FAN B (non-preferred combination).

f. XFN-67C, FAN C, and XFN-67D, FAN D (non-preferred combination).

SOP-114 REVISION 21 PAGE 27 OF 68 2.2 After starting above fan units verify the following (HVAC Panel): a. HIGH TEMP (LCB1 Point 9-9) is clear. b. CRDM COOLING WATER TROUBLE (LCB1 Point 9-11) is clear. c. TI-9341, CRDM SHROUD OUT TEMP °F, indicates between 70°F and 117°F. END OF SECTION SOP-114 REVISION 21 PAGE 28 OF 68 L. STARTUP CONTROL ROD DRIVE MECHANISM COOLING WATER SYSTEM 1.0 INITIAL CONDITIONS 1.1 CRDM Cooling Water System is filled and vented per Section III. 1.2 Control Panel lineup is complete per Attachment IV. 2.0 INSTRUCTIONS 2.1 Verify the following Containment Isolation valves are open (MCB): a. MVG-7501, TO CRDM CLR ISOL (ORB). b. MVG-7502, TO CRDM CLR ISOL (IRB). c. MVG-7503, FR CRDM CLR ISOL (IRB). d. MVG-7504, FR CRDM CLR ISOL (ORB).

NOTE 2.2 through 2.6 Operations are performed at XPN5409, LOC CONT INDUST COOLER CRDM UNIT (YD-125'E).

2.2 Place the following in ON position: a. XFN-145A, INDUSTRIAL COOLER CRDM COIL FAN. b. XFN-145B, INDUSTRIAL COOLER CRDM COIL FAN. c. XFN-145C, INDUSTRIAL COOLER CRDM COIL FAN. d. XFN-145D, INDUSTRIAL COOLER CRDM COIL FAN. CHG A SOP-114 REVISION 21 PAGE 29 OF 68 2.3 Place the following in ON position: a. XPP-156A, AC SYS INDUSTRIAL COOLER SPRAY PUMP A. b. XPP-156B, AC SYS INDUSTRIAL COOLER SPRAY PUMP B. c. XPP-156C, AC SYS INDUSTRIAL COOLER SPRAY PUMP C. d. XPP-156D, AC SYS INDUSTRIAL COOLER SPRAY PUMP D.

2.4 Ensure proper operation of CRDM Industrial Cooler as follows: a. Monitor shutdown spray pumps for reverse rotation. b. Verify adequate spray flow. c. Observe fan damper indication for proper operation. d. Adjust blow down as necessary by throttling (YD-425'E):

1) XVA17524A-AC, AC SYSTEM INDUSTRIAL COOLER BLOW DOWN VALVE.

2) XVA17524B-AC, AC SYSTEM INDUSTRIAL COOLER BLOW DOWN VALVE. e. Check discharge pipe for normal blow down flow. NOTE 2.5 Depressing PB-AC 14, ITS05573 OVER-RIDE, will bypass the high temperature trip for five minutes to allow mixing of expected hot water slug. 2.5 Depress and release PB-AC 14, ITS05573 OVER-RIDE, pushbutton. 2.6 Place XPP157A AND B, CRDM COOLING WATER PUMP, in A or B position. END OF SECTION SOP-114 REVISION 21 PAGE 30 OF 68 M. SHUTDOWN (AND DRAIN IF LONG TERM) CONTROL ROD DRIVE MECHANISM COOLING SYSTEM 1.0 INITIAL CONDITIONS 1.1 Plant is in Mode 5 or Containment Integrity is relaxed, prior to draining. NOTE 2.0 Unless otherwise noted, all operations are performed at XPN5409, LOC CONT INDUST COOLER CRDM UNIT (YD-125'E). 2.0 INSTRUCTIONS 2.1 Place XPP157A AND B, CRDM COOLING WATER PUMP, in OFF position.

2.2 Place the following in OFF position: a. XFN-145A, INDUSTRIAL COOLER CRDM COIL FAN. b. XFN-145B, INDUSTRIAL COOLER CRDM COIL FAN. c. XFN-145C, INDUSTRIAL COOLER CRDM COIL FAN. d. XFN-145D, INDUSTRIAL COOLER CRDM COIL FAN.

2.3 Place the following in OFF position: a. XPP-156A, AC SYS INDUSTRIAL COOLER SPRAY PUMP A. b. XPP-156B, AC SYS INDUSTRIAL COOLER SPRAY PUMP B. c. XPP-156C, AC SYS INDUSTRIAL COOLER SPRAY PUMP C. d. XPP-156D, AC SYS INDUSTRIAL COOLER SPRAY PUMP D.

SOP-114 REVISION 21 PAGE 31 OF 68 NOTE 2.4 a. If XSW1A3-1C3 Tie Breaker is closed and XFN-67A and XFN-67B are available, XFN-67C and XFN-67D will be locked out and XFN-67A and XFN-67B will have a locked in Auto-Start signal. b. Prior to shutdown of CRDM shroud exhaust fans, the CRDMs must be deenergized and RCS temperature must be less than 170°F. 2.4 At HVAC Control Panel, XCP-6210, shut down running CRDM Shroud Exhaust fans by placing the following in STOP position: a. XFN-67A, FAN A.

b. XFN-67B, FAN B.

c. XFN-67C, FAN C.

d. XFN-67D, FAN D. 2.5 If the CRDM Cooling System is being shutdown for an extended period of time (i.e. an outage) perform Attachment V.

2.6 Perform the following steps to place the CRDM Cooling System in long term shutdown for freeze protection or Maintenance purposes: a. Inform Chemistry of the intent to drain CRDM Cooling Water System to the IB sumps. The IB sump discharge will be required to be aligned to Pond 008.

b. Ensure the following valves are open: 1) MVG-7501, TO CRDM CLR ISOL (ORB). 2) MVG-7502, TO CRDM CLR ISOL (IRB). 3) MVG-7503, FR CRDM CLR ISOL (IRB). 4) MVG-7504, FR CRDM CLR ISOL (ORB). OA 9218 SOP-114 REVISION 21 PAGE 32 OF 68 Step 2.6 continued c. Ensure closed (or Tag if applicable) the following make up valves: 1) XVG07520-AC, CRDM COOLING WATER RETURN HDR FILL VLV (YD-125'E). 2) XVT28762-DN, MAKEUP TO CRDM COOLING SYSTEM FILL VALVE (IB-436, E PEN). CAUTION 2.6.d Draining the CRDM Cooling Water to the East Penetration Access Area Sump will result in Sodium Meta Silicate intrusion of Liquid Waste. Draining CRDM Cooling to the IB sump will require Chemistry to swap the IB sump discharge alignment to Pond 008. d. Drain CRDM Cooling Water to the IB sumps as follows:

1) Attach drain hoses to the following valves (412 E Pen in overhead, Northeast corner): a) XVT07548-AC, AC SYS INDUSTRIAL CLR OUT HDR DRAIN VLV. b) XVT07549-AC, INDUSTRIAL CLR AC RETURN HDR DRAIN VLV. 2) Direct the attached drain hoses from the 412 East Penetration area, through the door (DRPA/103), to the 412 IB sump.

3) Open the following valves to begin draining (412 E Pen overhead of Northeast corner): a) XVT07548-AC, AC SYS INDUSTRIAL CLR OUT HDR DRAIN VLV. b) XVT07549-AC, INDUSTRIAL CLR AC RETURN HDR DRAIN VLV.

SOP-114 REVISION 21 PAGE 33 OF 68 Step 2.6 continued e. Uncap and open the following (YD-125'E): 1) XVT07514A-AC, HIGH ROOT TO IPX05576. 2) XVT07514B-AC, HIGH ROOT TO IPX05578. 3) XVT07514C-AC, HIGH ROOT TO IPX05582. 4) XVT07514D-AC, HIGH ROOT TO IPX05584. 5) XVT07515A-AC, HIGH ROOT TO IPX05577. 6) XVT07515B-AC, HIGH ROOT TO IPX05579. 7) XVT07515C-AC, HIGH ROOT TO IPX05583. 8) XVT07515D-AC, HIGH ROOT TO IPX05585.

SOP-114 REVISION 21 PAGE 34 OF 68 Step 2.6 continued CAUTION 2.6.f Opening vent valves when water is being drained to the RB and Containment Integrity is required will result in a loss of Containment Integrity. NOTE 2.6.f It may be necessary to drain more to the 412 IB until no water comes out in the RB. Valves are only open to provide a vent path for draining the system to the IB-412. f. When the system drain rate decreases, open the following vent valves (RB 436): 1) XVT07508A-AC, HIGH ROOT TO IPX05561. 2) XVT07509A-AC, HIGH ROOT TO IPX05557. 3) XVT07508B-AC, HIGH ROOT TO IPX05560. 4) XVT07509B-AC, HIGH ROOT TO IPX05556. 5) XVT07508C-AC, HIGH ROOT TO IPX05563. 6) XVT07509C-AC, HIGH ROOT TO IPX05567. 7) XVT07508D-AC, HIGH ROOT TO IPX05562. 8) XVT07509D-AC, HIGH ROOT TO IPX05566.

g. Ensure the following: 1) Draining is complete to the 412 IB and report it to the Duty Shift Supervisor or Tagging Desk. 2) Remove the drain hoses and close door DRPA/103. 3) Notify Chemistry the CRDM draining is complete.

SOP-114 REVISION 21 PAGE 35 OF 68 Step 2.6 continued h. Protect Containment Integrity by closing and de-energizing one of the following sets of valves:

1) For A TRAIN: a) MVG-7501, TO CRDM CLR ISOL (ORB). (XMC1DA2X 06AD). b) MVG-7503, FR CRDM CLR ISOL (IRB). (XMC1DA2X 11IM). OR 2) For B TRAIN: a) MVG-7502, TO CRDM CLR ISOL (IRB). (XMC1DB2X 07IM). b) MVG-7504, FR CRDM CLR ISOL (ORB). (XMC1DB2X 07AD). END OF SECTION SOP-114 REVISION 21 PAGE 36 OF 68 N. SWAPPING CRDM SHROUD EXHAUST FANS 1.0 INITIAL CONDITIONS 1.1 CRDM Cooling System is in operation per Section III. NOTE 2.0 All operations are performed at the HVAC Control Panel, XCP-6210. 2.0 INSTRUCTIONS 2.1 Shut down running CRDM Shroud Exhaust fans by placing the following in STOP position: a. XFN-67A, FAN A.

b. XFN-67B, FAN B.

c. XFN-67C, FAN C.

d. XFN-67D, FAN D.

SOP-114 REVISION 21 PAGE 37 OF 68 CAUTION 2.2 a. To minimize starting current, running fans should be shutdown for a minimum of 30 seconds prior to starting idle fans. b. XFN-67A, FAN A, and XFN-67B, FAN B should NOT be paired together as this causes unacceptably high vibration levels in XFN-67A, FAN A. c. XFN-67C, FAN C, and XFN-67D, FAN D should NOT be paired together as this causes unacceptably high vibration levels in XFN-67C, FAN C. NOTE 2.2 a. For each fan started, the switch must be held in START for five seconds or until the associated RUN light is energized. b. If XSW1A3-1C3 Tie Breaker is closed and XFN-67A and XFN-67B are available, XFN-67C and XFN-67D will be locked out and XFN-67A and XFN-67B will have a locked in Auto-Start signal. 2.2 Start one of the following pairs of CRDM Shroud Exhaust fans: a. XFN-67A, FAN A, and XFN-67D, FAN D (preferred combination). b. XFN-67B, FAN B, and XFN-67C, FAN C (preferred combination).

c. XFN-67A, FAN A, and XFN-67C, FAN C.

d. XFN-67B, FAN B, and XFN-67D, FAN D.

e. XFN-67A, FAN A, and XFN-67B, FAN B (non-preferred combination).

f. XFN-67C, FAN C, and XFN-67D, FAN D (non-preferred combination).

2.3 After operating fans have been swapped, verify the following (HVAC Panel): a. HIGH TEMP (LCB1 Point 9-9) is clear. b. CRDM COOLING WATER TROUBLE (LCB1 Point 9-11) is clear. c. TI-9341, CRDM SHROUD OUT TEMP °F, indicates less than 170 F. END OF SECTION N01 SOP-114 REVISION 21 PAGE 38 OF 68 O. RAISING REACTOR BUILDING PRESSURE USING NORMAL PRESSURE CONTROL 1.0 INITIAL CONDITIONS 1.1 Raising Reactor Building pressure is required/desired. NOTE 2.0 Unless otherwise noted, all operations are performed at HVAC Control Panel, XCP-6210. 2.0 INSTRUCTIONS 2.1 Ensure RMA0002, RB SAMPLE LINE GAS ATMOS MONITOR is in service (Rad Monitoring Panel).

2.2 At the Main Control Board, ensure the following valves are open to indicate pressure on PI-8254, RB NR PRESS PSI: a. SVX-6054, RB NR PRESS CNTMT ISOL. b. SVX-6050A, POST ACCID H2 LOOP A (IRB).

2.3 Raise RB pressure as follows: a. Ensure MVB-6063, H2 REMOVAL ALT PUR THROT, is closed. CAUTION 2.3.b To prevent damage to Alternate Purge System ducting, Alternate Purge must not be placed in service if RB pressure is greater than 3 psi.

b. Open the following: 1) PVG-6056, ALT PUR SPLY ISOL VLV. 2) PVG-6057, ALT PUR SPLY ISOL VLV.

SOP-114 REVISION 21 PAGE 39 OF 68 Step 2.3 continued CAUTION 2.3.c If flow is noted on FI-8251, H2 PURGE FLOW CFM, then XFN-95, 6 INCH RB PUR INTAKE FAN, should be tripped and MVB-6063, H2 REMOVAL ALT PUR THROT, should be checked for leakage. c. Start XFN-95, 6 INCH RB PUR INTAKE FAN. 2.4 If RB pressure is being raised to support door maintenance, simultaneous opening of both personnel hatch doors or opening of equipment hatch door, perform the following: a. Monitor PI-8254, RB NR PRESS PSI. b. When RB pressure is between negative 0.05 psi and positive 0.05 psi, stop XFN-95, 6 INCH RB PUR INTAKE FAN. 2.5 If RB pressure is being raised for any other reason, perform the following: a. Monitor PI-8254, RB NR PRESS PSI. b. When RB pressure is between 0.2 psi and 1.0 psi, stop XFN-95, 6 INCH RB PUR INTAKE FAN.

2.6 Close the following: a. PVG-6056, ALT PUR SPLY ISOL VLV. b. PVG-6057, ALT PUR SPLY ISOL VLV. END OF SECTION SOP-114 REVISION 21 PAGE 40 OF 68 P. LOWERING REACTOR BUILDING PRESSURE USING NORMAL PRESSURE CONTROL 1.0 INITIAL CONDITIONS 1.1 Lowering Reactor Building pressure is required/desired. NOTE 2.0 Unless otherwise noted, all operations are performed at HVAC Control Panel, XCP-6210. 2.0 INSTRUCTIONS 2.1 Ensure a Reactor Building Purge Release Permit has been issued per HPP-709.

2.2 Ensure the following Radiation Monitors are in service (Rad Monitoring Panel): a. RMA0004, RB PURGE EXH GAS ATMOS MONITOR. b. RMA0002, RB SAMPLE LINE GAS ATMOS MONITOR.

2.3 On Main Control Board, ensure the following valves are open to indicate pressure on PI-8254, RB NR PRESS PSI: a. SVX-6054, RB NR PRESS CNTMT ISOL. b. SVX-6050A, POST ACCID H2 LOOP A (IRB). 2.4 Ensure RMA0004, RB PURGE EXH GAS ATMOS MONITOR, high radiation alarm setpoint is adjusted per Reactor Building Release Permit. OA 9843 SOP-114 REVISION 21 PAGE 41 OF 68 2.5 Reduce RB pressure as follows: CAUTION 2.5.a To prevent damage to Alternate Purge System ducting, Alternate Purge must not be placed in service if RB pressure is greater than 3 psi.

a. Open the following: 1) PVG-6066, CNTMT PUR EXH ISOL VLV. 2) PVG-6067, CNTMT PUR EXH ISOL VLV. b. Start XFN-96, ALT PUR EXH FAN. (Peer )

c. Record time release was started and flow rate as indicated on FI-8252, H2 PURGE FLOW CFM, in the following places: 1) Reactor Building Purge Release Permit. 2) Station Log Book.

2.6 If RB pressure is being lowered to support door maintenance, simultaneous opening of both personnel hatch doors or opening of equipment hatch door, perform the following: a. Monitor PI-8254, RB NR PRESS PSI. b. When RB pressure is between negative 0.05 psi and positive 0.05 psi, stop XFN-96, ALT PUR EXH FAN.

2.7 If RB pressure is being lowered for any other reason, perform the following: a. Monitor PI-8254, RB NR PRESS PSI. b. When RB pressure is between 0.2 psi and 1.0 psi, stop XFN-96, ALT PUR EXH FAN.

2.8 Close the following: a. PVG-6066, CNTMT PUR EXH ISOL VLV. b. PVG-6067, CNTMT PUR EXH ISOL VLV.

SOP-114 REVISION 21 PAGE 42 OF 68 2.9 Record time release was stopped in the following places: a. Reactor Building Purge Release Permit. b. Station Log Book. END OF SECTION SOP-114 REVISION 21 PAGE 43 OF 68 Q. STARTUP REACTOR BUILDING ALTERNATE PURGE SUPPLY AND EXHAUST 1.0 INITIAL CONDITIONS 1.1 Valve lineup is complete per Attachment IA. 1.2 Electrical lineup is complete per Attachment IIA. 1.3 Control Panel lineup is complete per Attachment IV. NOTE 2.0 Unless otherwise noted, all operations are performed at HVAC Control Panel, XCP-6210. 2.0 INSTRUCTIONS 2.1 Ensure a Reactor Building Purge Release Permit has been issued per HPP-709.

2.2 Ensure the following Radiation Monitors are in service (Rad Monitoring Panel): a. RMA0004, RB PURGE EXH GAS ATMOS MONITOR. b. RMA0002, RB SAMPLE LINE GAS ATMOS MONITOR. c. Hang Main Control Board Status Indicators, per OAP-100.5, on RMA0002 Sample Pump controls and Channel Modules identifying that Purge is in progress. d. Hang Main Control Board Status Indicators, per OAP-100.5, on RMA0004 Sample Pump controls and Channel Modules identifying that Purge is in progress.

2.3 On Main Control Board, ensure the following valves are open to indicate pressure on PI-8254, RB NR PRESS PSI: a. SVX-6054, RB NR PRESS CNTMT ISOL. b. SVX-6050A, POST ACCID H2 LOOP A (IRB). OA 9843 SOP-114 REVISION 21 PAGE 44 OF 68 2.4 Ensure RMA0004, RB PURGE EXH GAS ATMOS MONITOR high radiation alarm setpoint is adjusted per Reactor Building Purge Release Permit. CAUTION 2.5 To prevent damage to Alternate Purge System ducting, Alternate Purge must not be placed in service if RB pressure is greater than 3 psi.

2.5 Initiate Reactor Building Alternate Purge as follows: NOTE 2.5.a and 2.5.b If RB pressure is near the upper limit, Steps 2.5.a and 2.5.b may be reversed to prevent exceeding Technical Specifications limit. a. Start up Alternate Purge Supply System as follows: 1) Ensure MVB-6063, H2 REMOVAL ALT PUR THROT, is closed. 2) Open PVG-6056, ALT PUR SPLY ISOL VLV. 3) Open PVG-6057, ALT PUR SPLY ISOL VLV. 4) Start XFN-95, 6 INCH RB PUR INTAKE FAN. b. Start up Alternate Purge Exhaust System as follows: 1) Open PVG-6066, CNTMT PUR EXH ISOL VLV. 2) Open PVG-6067, CNTMT PUR EXH ISOL VLV. 3) Start XFN-96, ALT PUR EXH FAN. (Peer ) c. Verify flow on FI-8252, H2 PURGE FLOW CFM. 2.6 Record purge start time.

SOP-114 REVISION 21 PAGE 45 OF 68 2.7 Establish RB pressure in one of the following pressure bands: a. Adjust MVB-6063, H2 REMOVAL ALT PUR THROT, to maintain RB pressure between 0.2 psi and 1.0 psi, as indicated on PI-8254, RB NR PRESS PSI.

b. To support door maintenance, simultaneous opening of both personnel hatch doors or opening of equipment hatch door, adjust MVB-6063, H2 REMOVAL ALT PUR THROT, to maintain RB pressure between negative 0.05 psi and positive 0.05 psi, as indicated on PI-8254, RB NR PRESS PSI. NOTE 2.8 Flow, as read on FI-8251, H2 PURGE FLOW CFM, shall be maintained less than or equal to 100 cfm, while throttling MVB-6063, H2 REMOVAL ALT PUR THROT. 2.8 If required, adjust MVB-6063, H2 REMOVAL ALT PUR THROT, to maintain RB pressure in desired pressure band, as indicated on PI-8254, RB NR PRESS PSI, as follows:

a. As pressure increases, throttle MVB-6063, H2 REMOVAL ALT PUR THROT, open, as necessary, to lower pressure.

b. As pressure decreases, throttle MVB-6063, H2 REMOVAL ALT PUR THROT, closed, as necessary, to raise pressure. 2.9 If MVB-6063, H2 REMOVAL ALT PUR THROT, is fully closed and pressure continues to decrease, stop and start XFN-96, ALT PUR EXH FAN, as necessary, to maintain pressure in desired pressure band, as indicated on PI-8254, RB NR PRESS PSI. END OF SECTION SOP-114 REVISION 21 PAGE 46 OF 68 R. SHUTDOWN REACTOR BUILDING ALTERNATE PURGE SUPPLY AND EXHAUST 1.0 INITIAL CONDITIONS 1.1 RB Alternate Purge Supply and Exhaust System is in operation per Section III. NOTE 2.0 Unless otherwise noted, all operations are performed at HVAC Control Panel, XCP-6210. 2.0 INSTRUCTIONS 2.1 Monitor PI-8254, RB NR PRESS PSI, to ensure pressure is between 0.2 psi and 1.0 psi. NOTE 2.2 and 2.3 If RB pressure is near the upper limit, Steps 2.2 and 2.3 may be reversed to prevent exceeding Technical Specifications limit.

2.2 Shut down Alternate Purge Exhaust System as follows: a. Stop XFN-96, ALT PUR EXH FAN. b. Close PVG-6066, CNTMT PUR EXH ISOL VLV. c. Close PVG-6067, CNTMT PUR EXH ISOL VLV.

2.3 Shut down Alternate Purge Supply System as follows: a. Close MVB-6063, H2 REMOVAL ALT PUR THROT. b. Stop XFN-95, 6 INCH RB PUR INTAKE FAN. c. Close PVG-6056, ALT PUR SPLY ISOL VLV. d. Close PVG-6057, ALT PUR SPLY ISOL VLV.

SOP-114 REVISION 21 PAGE 47 OF 68 2.4 Record purge stop time. 2.5 Remove Main Control Board Status Indicators on RMA0002 and RMA0004 Sample Pump controls and Channel Modules identifying that Purge is in progress. END OF SECTION SOP-114 REVISION 21 PAGE 48 OF 68 S. OPERATING DIGITAL ROD POSITION INDICATION CABINET COOLING SYSTEM 1.0 INITIAL CONDITIONS 1.1 Electrical lineup is complete per Attachment IIA. 1.2 Control Panel lineup is complete per Attachment IV. 1.3 Industrial Cooling Water is in service per SOP-125. 1.4 Service Water System is aligned and in service per SOP-117. NOTE 2.0 All operations are performed at HVAC Control Panel, XCP-6210. 2.0 INSTRUCTIONS 2.1 Open the following: a. PVT-3164, DRPI CLG UNIT COIL ISOL. b. PVT-3169/PVT-3165, DRPI CLG UNIT COIL ISOL. 2.2 If available, start XPP-149, DRPI CLG UNIT BSTR PP. 2.3 Start XFN-107, RPI CABINET CLG FAN.

2.4 If the DRPI Booster pump is available, verify the following annunciators have cleared: a. FAN PUMP TRIP (LCB2 Point 10-25). b. HIGH TEMP (LCB2 Point 11-25).

SOP-114 REVISION 21 PAGE 49 OF 68 2.5 Return system to standby when DRPI Cooling System operation is no longer required, as follows: a. If desired, stop XFN-107, RPI CABINET CLG FAN. b. If previously started, stop XPP-149, DRPI CLG UNIT BSTR PP. c. Close the following: 1) PVT-3169/PVT-3165, DRPI CLG UNIT COIL ISOL. 2) PVT-3164, DRPI CLG UNIT COIL ISOL. END OF SECTION SOP-114 REVISION 21 PAGE 50 OF 68 T. SWAPPING INSERVICE CRDM COOLING SPRAY AND COOLING WATER PUMPS 1.0 INITIAL CONDITIONS 1.1 CRDM Cooling System is in service per Section III. NOTE 2.0 All operations are performed at XCI0004-AC, CRDM COOLING WATER INDUSTRIAL COOLER, (YD-125'E) unless otherwise noted. 2.0 INSTRUCTIONS 2.1 Swap in-service CRDM Spray Pumps as follows: a. At XPN5409, LOC CONT INDUST COOLER CRDM UNIT, position XPP-156A, B, C AND D, INDUSTRIAL COOLER CRDM COIL NORMAL AND STANDBY PUMPS, Switch to A-C (B-D). b. Vent off any entrapped air from the casing of the running CRDM Spray Pumps by performing the following: 1) Throttle the appropriate discharge valves closed for the running spray pumps: a) XVB17502, AC IND CLR SPRAY PUMP A DISCH VALVE. b) XVB17508, AC IND CLR SPRAY PUMP C DISCH VALVE. c) XVB17505, AC IND CLR SPRAY PUMP B DISCH VALVE. d) XVB17511, AC IND CLR SPRAY PUMP D DISCH VALVE. 2) Vent each running spray pump casing until a steady stream of water is discharged. 3) Open the discharge valves that were throttled in Step 2.1.b.1). 4) Verify that observed spray flow is adequate. CHG B SOP-114 REVISION 21 PAGE 51 OF 68 Step 2.1 continued c. Ensure proper operation of CRDM Industrial Cooler as follows: 1) Monitor shutdown spray pumps for reverse rotation. 2) Observe fan damper indication for proper operation. 3) Adjust blow down as necessary by throttling: a) XVA17524A-AC, AC SYSTEM INDUSTRIAL COOLER BLOW DOWN VALVE.

b) XVA17524B-AC, AC SYSTEM INDUSTRIAL COOLER BLOW DOWN VALVE. 4) Check discharge pipe for normal blow down flow.

2.2 Swap CRDM Cooling Water Pumps by placing XPP157A AND B, CRDM COOLING WATER PUMP, in the A or B position, as desired. END OF SECTION CHG B CHG B SOP-114 REVISION 21 PAGE 52 OF 68 IV. INFREQUENT OPERATIONS A. FILLING CONTROL ROD DRIVE MECHANISM COOLING SYSTEM EXPANSION TANK FROM DEMINERALIZED WATER 1.0 INITIAL CONDITIONS 1.1 CRDM Cooling Water System has been filled and vented per Section III. 1.2 Filling of CRDM Cooling Water System Expansion Tank is required/desired. NOTE 2.0 All operations are performed in the IB-436 East Penetration Area. 2.0 INSTRUCTIONS 2.1 Remove cap from XVG17513-AC, AC SYSTEM VENT ISOLATION VALVE. 2.2 Throttle open XVT28762-DN, DN MAKEUP TO CRDM COOLING SYSTEM FILL VALVE, to begin filling expansion tank. 2.3 Throttle open XVG17513-AC, AC SYSTEM VENT ISOLATION VALVE, as necessary to maintain CRDM Cooling Water System suction header pressure between 3 psig and 8 psig as indicated on IPI05553-AC, AC PUMPS SUCTION HEADER PRESSURE IND. 2.4 Close XVT28762-DN, DN MAKEUP TO CRDM COOLING SYSTEM FILL VALVE when expansion tank is at desired level, between 1/4 full and 3/4 full. 2.5 Close and cap XVG17513-AC, AC SYSTEM VENT ISOLATION VALVE. END OF SECTION SOP-114 REVISION 21 PAGE 53 OF 68 B. FILLING CONTROL ROD DRIVE MECHANISM COOLING SYSTEM EXPANSION TANK FROM FILTERED WATER 1.0 INITIAL CONDITIONS 1.1 CRDM Cooling Water System has been filled and vented per Section III. 1.2 CRDM Cooling Tower Spray Sump is at normal level with no makeup in progress. 1.3 Filling of CRDM Cooling Water System Expansion Tank is required/desired. 2.0 INSTRUCTIONS 2.1 Notify Control Room prior to filling expansion tank. NOTE 2.2 Maintaining IPI05592-AC, AC SYS INDUSTRIAL CLR INLET PRESS IND (YD-125'E), less than 20 psig prevents lifting of XVR07510A,B,C,D, CRDM COOLER AC RETURN HEADER RELIEF VLV (RB-436), at 50 psig. 2.2 Throttle open XVG07520-AC, CRDM COOLING WATER RETURN HDR FILL VLV (YD-125'E), to begin filling expansion tank while maintaining less than 20 psig on IPI05592-AC, AC SYS INDUSTRIAL CLR INLET PRESS IND. 2.3 Remove cap from XVG17513-AC, AC SYSTEM VENT ISOLATION VALVE. 2.4 Throttle open XVG17513-AC, AC SYSTEM VENT ISOLATION VALVE, as necessary to maintain CRDM Cooling Water System suction header pressure between 3 psig and 8 psig as indicated on IPI05553-AC, AC PUMPS SUCTION HEADER PRESSURE IND. 2.5 Close XVG07520-AC, CRDM COOLING WATER RETURN HDR FILL VLV (YD-125'E), when expansion tank is at desired level, between 1/4 full and 3/4 full. 2.6 Notify Control Room after filling expansion tank. END OF SECTION SOP-114 REVISION 21 PAGE 54 OF 68 C. SUPPLYING NITROGEN TO OPERATE REACTOR BUILDING PURGE SUPPLY AND EXHAUST VALVES 1.0 INITIAL CONDITIONS 1.1 The Plant is Defueled. 1.2 Reactor Building Purge and Exhaust is secured. 1.3 Valve lineup is complete per Attachment IA. 1.4 Electrical lineup is complete per Attachment IIA. 1.5 Control Panel lineup is complete per Attachment IV. 1.6 Fire Service System is in service per SOP-509. 1.7 Nitrogen bottles with regulators are staged in the RB. NOTE2.0 Installation of Nitrogen to operate the RB Purge and Exhaust valves makes the RB Purge system inoperable. 2.0 INSTRUCTIONS 2.1 Initiate an MWR, per EIR 81067, to direct I&C to connect nitrogen bottles to: a. XVB00001B-AH, RB PURGE SUPPLY ISOLATION VALVE (IRC).

b. XVB00002B-AH, RB PURGE EXHAUST ISOLATION VALVE (IRC).

SOP-114 REVISION 21 PAGE 55 OF 68 2.2 When I&C completes installation of the nitrogen bottles, then start Reactor Building Purge as follows: a. Open PVB-2A, CNTMT EXH ISOL. NOTE 2.2.b If both trains of RB Purge are to be run, both exhaust fans should be started simultaneously. b. Hold PVB-2B, CNTMT EXH ISOL, to OPEN while simultaneously holding one or both of the following fan control switches in the START position: (Peer ) 1) XFN-13A, EXH FAN A. 2) XFN-13B, EXH FAN B. c. Verify the following: 1) XFN-13A(B)-AH inlet damper opens. 2) XFN-13A(B)-AH outlet damper opens. d. Open XDP-28, INTAKE DMPR. e. Open the following: 1) PVB-1A, CNTMT SPLY ISOL. 2) PVB-1B, CNTMT SPLY ISOL. N01 SOP-114 REVISION 21 PAGE 56 OF 68 Step 2.2 continued NOTE 2.2.f 1) If both trains of RB Purge are to be run, both supply fans should be started simultaneously. 2) When the Reactor Building Equipment Hatch is open, only one RB Purge supply fan should be in operation. 3) When equipment hatch is open, negative pressure will be maintained in the RB by operating fewer supply fans than exhaust fans. 4) Both supply fans may be secured with the equipment hatch open and two exhaust fans running to control radiological conditions. f. Start one or both of the following, as necessary: 1) XFN-11A, SPLY FAN A. 2) XFN-11B, SPLY FAN B. g. Verify the following: 1) XFN-11A(B)-AH inlet damper opens fully. 2) XFN-11A(B)-AH outlet damper opens fully. 3) The following selected fan(s) starts: a) XFN-11A, SPLY FAN A.

b) XFN-11B, SPLY FAN B. 2.3 When it is desired for I&C to restore Instrument Air to the Reactor Building Purge and Exhaust Valves, proceed to shut down the Reactor Building Purge and Exhaust per Steps 2.4 through 2.12. 2.4 Stop one of the following: a. XFN-11A, SPLY FAN A.

b. XFN-11B, SPLY FAN B. N01 SOP-114 REVISION 21 PAGE 57 OF 68 2.5 Verify the following: a. XFN-11A(B)-AH inlet damper is closed. b. XFN-11A(B)-AH outlet damper is closed. 2.6 Remove remaining supply fan from service per Steps 2.4 and 2.5, if required. 2.7 Close the following: a. PVB-1A, CNTMT SPLY ISOL. b. PVB-1B, CNTMT SPLY ISOL. 2.8 Close XDP-28, INTAKE DMPR. 2.9 Stop the following running exhaust fans: a. XFN-13A, EXH FAN A.

b. XFN-13B, EXH FAN B. 2.10 Verify the following: a. XFN-13A(B)-AH inlet damper is closed. b. XFN-13A(B)-AH outlet damper is closed. 2.11 Remove remaining exhaust fan from service per Steps 2.9 and 2.10, if required.

2.12 Close the following: a. PVB-2A, CNTMT EXH ISOL. b. PVB-2B, CNTMT EXH ISOL. 2.13 Direct I&C (per EIR 81067/MWR step), to reconnect instrument air to:

a. XVB00001B-AH, RB PURGE SUPPLY ISOLATION VALVE (IRC).

b. XVB00002B-AH, RB PURGE EXHAUST ISOLATION VALVE (IRC). 2.14 Perform STP-130.005B to verify Reactor Building Purge and Exhaust Valve operability.

SOP-114 REVISION 21 PAGE 58 OF 68 2.15 Start Reactor Building Purge as follows: a. Open PVB-2A, CNTMT EXH ISOL. NOTE 2.15.b If both trains of RB Purge are to be run, both exhaust fans should be started simultaneously. b. Hold PVB-2B, CNTMT EXH ISOL, to OPEN while simultaneously holding one or both of the following fan control switches in the START position: (Peer ) 1) XFN-13A, EXH FAN A. 2) XFN-13B, EXH FAN B. c. Verify the following: 1) XFN-13A(B)-AH inlet damper opens. 2) XFN-13A(B)-AH outlet damper opens. d. Open XDP-28, INTAKE DMPR. e. Open the following: 1) PVB-1A, CNTMT SPLY ISOL. 2) PVB-1B, CNTMT SPLY ISOL. N01 SOP-114 REVISION 21 PAGE 59 OF 68 Step 2.15 continued NOTE 2.15.f 1) If both trains of RB Purge are to be run, both supply fans should be started simultaneously. 2) When the Reactor Building Equipment Hatch is open, only one RB Purge supply fan should be in operation. 3) When equipment hatch is open, negative pressure will be maintained in the RB by operating fewer supply fans than exhaust fans. 4) Both supply fans may be secured with the equipment hatch open and two exhaust fans running to control radiological conditions. f. Start one or both of the following, as necessary: 1) XFN-11A, SPLY FAN A. 2) XFN-11B, SPLY FAN B. g. Verify the following: 1) XFN-11A(B)-AH inlet damper opens fully. 2) XFN-11A(B)-AH outlet damper opens fully. 3) The following selected fan(s) starts: a) XFN-11A, SPLY FAN A.

b) XFN-11B, SPLY FAN B. END OF SECTION N01 SOP-114 REVISION 21 PAGE 60 OF 68 D. SHIFTING REACTOR BUILDING COOLING UNITS FAN SPEED 1.0 INITIAL CONDITIONS 1.1 None. 2.0 INSTRUCTIONS 2.1 Shutdown RBCUs by placing appropriate switch(es) in STOP: a. XFN 0064A-AH, 1A NORM. b. XFN 0064A-AH, 1A SLOW. c. XFN 0064B-AH, 1B NORM. d. XFN 0064B-AH, 1B SLOW. e. XFN 0065A-AH, 2A NORM. f. XFN 0065A-AH, 2A SLOW. g. XFN 0065B-AH, 2B NORM. h. XFN 0065B-AH, 2B SLOW. 2.2 Ensure the RBCU dampers are in BYP: a. XDP-110A, RBCU 64A HEPA FLTR BYP DMPR. b. XDP-111A, RBCU 65A HEPA FLTR BYP DMPR. c. XDP-110B, RBCU 64B HEPA FLTR BYP DMPR d. XDP-111B, RBCU 65B HEPA FLTR BYP DMPR. CHG A SOP-114 REVISION 21 PAGE 61 OF 68 NOTE 2.3 a. Due to eddy current brakes, RBCU control switches must be held in START position until the red breaker closed light is lit and starting current is indicated on appropriate meter. b. Normal and preferred lineup is three RBCUs running in NORM (fast speed). c. To increase stay times for teams entering containment, four RBCUs may be placed in service in NORM (fast speed). 2.3 Place RBCUs in service by starting desired number of RBCUs in SLOW or NORM as follows: a. For XFN0064A-AH, REACTOR BLDG COOLING UNIT 1A EMERG FAN, start one of the following:

1) XFN 0064A-AH, 1A NORM.

2) XFN 0064A-AH, 1A SLOW. b. For XFN0064B-AH, REACTOR BLDG COOLING UNIT 1B EMERG FAN, start one of the following: 1) XFN 0064B-AH, 1B NORM.

2) XFN 0064B-AH, 1B SLOW. c. For XFN0065A-AH, REACTOR BLDG COOLING UNIT 2A EMERG FAN, start one of the following: 1) XFN 0065A-AH, 2A NORM.

2) XFN 0065A-AH, 2A SLOW. d. For XFN0065B-AH, REACTOR BLDG COOLING UNIT 2B EMERG FAN, start one of the following: 1) XFN 0065B-AH, 2B NORM.

2) XFN 0065B-AH, 2B SLOW.

SOP-114 REVISION 21 PAGE 62 OF 68 NOTE 2.4 If RBCU fan motor amps exceed the values given, PSE should be contacted to evaluate. 2.4 Verify RBCU Fan motor amps return to normal operating range: a. For fast speed operation: 275 amps to 300 amps.

b. For slow speed operation: 55 amps to 70 amps. NOTE 2.5 The RBCU TRAIN A (B) EMERG switch must be selected to an operable RBCU.

2.5 Verify the following switches are in the desired position: a. XFN-64A/XFN 65A - RBCU TRAIN A EMERG. b. XFN-64B/XFN 65B - RBCU TRAIN B EMERG. END OF SECTION SOP-114 REVISION 21 PAGE 63 OF 68 V. OFF NORMAL CONDITIONS A. LOSS OF ANY REACTOR BUILDING VENTILATION FAN 1.0 ENTRY CONDITIONS 1.1 High vibration, high temperature, high smoke, or fan trip alarm on any RB ventilation fan (HVAC Control Panel or MCB). 1.2 One of the following has tripped on a high vibration: a. Reactor Compartment Cooling Fan. b. Secondary Compartment Cooling Fan.

NOTE 2.0 XSW1A3 - 1C3 Tie Breaker closure affects CRDM Shroud Exhaust fan operation as follows: a. If both XFN0067A and XFN0067B are available, then regardless of operating fan combination, XFN0067C and XFN0067D will be tripped and XFN0067A and XFN0067B will start when tie breaker is closed. b. If either XFN0067A or XFN0067B is not available, then regardless of operating fan combination, neither XFN0067C nor XFN0067D will be tripped and XFN0067A and XFN0067B will not be started when tie breaker is closed. 2.0 CORRECTIVE ACTIONS 2.1 If cause is high temperature or high smoke, stop affected fan. CAUTION 2.2 A fan that has tripped due to high vibration, high temperature, or high smoke alarm should NOT be restarted. 2.2 If permissible, restart tripped fan per applicable subsection of Section III.

SOP-114 REVISION 21 PAGE 64 OF 68 2.3 Realign remaining fans in affected compartment per applicable subsection of Section III. 2.4 Refer to Technical Specifications, if applicable, for LCO and Surveillance Requirements.

2.5 If no Reactor Compartment Cooling fans are available, proceed as follows: a. To continue Power Operations indefinitely, perform the following: 1) Block door to Incore Pit (RB-101) fully open with an Engineering approved device within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of loss of Reactor Compartment Cooling. 2) Operations will monitor performance of Excore Detectors in support of the evaluation to ensure that increasing temperature in the Incore area will not affect Excore Detectors. b. If conditions do not allow opening door to Incore Pit (RB-101), Power Operations may continue for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, after which a shutdown must commence to place the plant in Mode 4 within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. END OF SECTION SOP-114 REVISION 21 PAGE 65 OF 68 B. HIGH TEMPERATURE IN DIGITAL ROD POSITION INDICATION CABINET 1.0 ENTRY CONDITIONS 1.1 HIGH TEMP (LCB2 Point 11-25) alarm on XCP6210, HVAC Control Board.

NOTE 2.0 All operations are performed at HVAC Control Panel, XCP-6210. 2.0 CORRECTIVE ACTIONS 2.1 Open PVT-3164, DRPI CLG UNIT COIL ISOL. 2.2 Open PVT-3169/PVT-3165, DRPI CLG UNIT COIL ISOL. 2.3 If available, start XPP-149, DRPI CLG UNIT BSTR PP. 2.4 Start XFN-107, RPI CABINET CLG FAN.

2.5 Verify the following annunciators have cleared: a. FAN PUMP TRIP (LCB2 Point 10-25). b. HIGH TEMP (LCB2 Point 11-25).

2.6 Correct cause of high temperature condition and restore DRPI Cooling System to a standby condition, if allowed, as follows: a. Stop XFN-107, RPI CABINET CLG FAN. b. Stop XPP-149, DRPI CLG UNIT BSTR PP. c. Close PVT-3169/PVT-3165, DRPI CLG UNIT COIL ISOL. d. Close PVT-3164, DRPI CLG UNIT COIL ISOL. END OF SECTION SOP-114 REVISION 21 PAGE 66 OF 68 C. HIGH TEMPERATURE ON REACTOR BUILDING CHARCOAL CLEANUP FILTER PLENUM OR REACTOR BUILDING PURGE EXHAUST FILTER PLENUM 1.0 ENTRY CONDITIONS 1.1 HI/HI-HI BED TEMP alarms on RB Charcoal Cleanup System Filter Plenum, on XCP6210, HVAC Control Board. 1.2 HI/HI-HI BED TEMP alarms on RB Purge Exhaust System Filter Plenum, on XCP6210, HVAC Control Board. NOTE 2.0 Unless otherwise noted, all operations are performed at HVAC Control Panel, XCP-6210. 2.0 CORRECTIVE ACTIONS 2.1 Ensure MVG-6797, FIRE SERV CNTMT ISOL, is open (MCB). 2.2 Shut down operating fan(s) for RB Charcoal Cleanup System, if affected: a. XFN-66A, FAN A:

b. XFN-66B, FAN B. 2.3 Shut down operating fan(s) for RB Purge Exhaust System, if affected: a. XFN-13A, EXH FAN A.

b. XFN-13B, EXH FAN B. 2.4 Open XVM-6795, PLEN DELUGE to Initiate deluge spray to RB Charcoal Cleanup Plenum if affected.

SOP-114 REVISION 21 PAGE 67 OF 68 2.5 Open the following to initiate deluge spray to RB Purge Exhaust Plenum if affected: a. XVM-6760, EXH PLEN DELUGE. CAUTION 2.5.b XVG06759-FS, RB PURGE EXH PLENUM SPR DELUGE VLV IN (AB-463), is closed to prevent inadvertent flooding of RB Charcoal Plenum. This valve must be manually opened to initiate deluge spray, but only in case of fire involving this plenum. b. XVG06759-FS, RB PURGE EXH PLENUM SPR DELUGE VLV IN (AB-463). 2.6 Verify high temperature alarm clears for affected filter plenum.

2.7 Close deluge valve(s) to affected plenum when temperature alarm clears as follows: a. Close XVM-6795, PLEN DELUGE to RB Charcoal Cleanup Plenum. b. Close the following to RB Purge Exhaust Plenum: 1) XVM-6760, EXH PLEN DELUGE. 2) XVG06759-FS, RB PURGE EXH PLENUM SPR DELUGE VLV IN (AB-463). 2.8 Notify Maintenance to inspect filter. END OF SECTION SOP-114 REVISION 21 PAGE 68 OF 68 VI. REFERENCES 1. G/C Final System Description.

2. FSAR Section 3.8.1.5.1.2, 6.2.2, and 9.4.8.

3. D-912-102, RB Cooling System.

4. D-912-103, RB Purge Supply and Exhaust.

5. D-912-104, RB Charcoal Clean-up, Secondary, Compt Cooling and Reactor Compt Cooling.

6. D-912-105, RB Refueling Water Surface System.

7. D-912-106, RB CRDM Shroud Ventilation System.

8. D-302-861, Post Accident H2 Removal and Alternate Purge System.

9. D-302-222, Service Water Cooling.

10. D-302-852, CRDM Cooling Water.

11. V.C. Summer Tech Specs 3.6.1.4, 3.6.1.5, 3.6.1.7, and 3.6.2.3.

12. Design Basis Document for HVAC.

13. CGSS-03-2840-NO.

14. EIR-80375.

15. Offsite Dose Calculation Manual.

16. HPP-709, Sampling and Release of Radioactive Gaseous Effluents.

17. MMP-500.002, Reactor Building Personnel Airlock Maintenance And Operation.

18. B-208-004, AH-216 through AH-298.

19. B-208-016, AC-1 through 15 and AC-50.

20. B-208-108 VL-53, Reactor Building Elevator Machine Room Exhaust Fan (XFN-135).

21. STP-130.005B, AH Valve Operability Testing.

SOP-114 ENCLOSURE A PAGE 1 OF 2 REVISION 21 SYSTEM INFORMATION 1. CRDM cooling pumps will trip on high temperature (120°F) from Reactor Building cooling coils. This trip is intended to prevent degradation of concrete around the RB penetration. If pumps trip, they can be restarted but will only run for five minutes while temperature remains above 120°F. When the temperature is below 120°F, the pumps will run normally. The CRDM Industrial Cooler (XCI-4) spray pumps, and spray fans are manually actuated by control switches on XPN-5409. Cooling tower outlet temperature as measured by ITE5555, provides a variable resistance output proportional to process fluid temperature to the primary damper controller, XDP5022D, which in turn modulates cooling tower dampers (XDP5022A, XDP5022B, and XDP5022C) to control industrial cooler (XCI-

2. Reactor Building Cooling Units (each)

Normal Accident a. Air Flow (cfm) 123,000 60,270 b. Fan Motor Horsepower 275 100 c. Heat Removal Rate (x106 BTU/hr) 4.16 117.5 d. Water Flow (gpm) 665 2000

3. Reactor Building Charcoal Clean-up Units (each)

a. Air Flow (cfm) 12,000 b. Fan Motor Horsepower 40 c. Roughing Filters 9 d. HEPA Filters 18(9 per bank) e. Charcoal Filters 6

4. Purge Supply System (each)

a. Max Filter Plenum Capacity (cfm) 12,000 b. Fan Capacity (cfm) 10,000 c. Fan Motor Horsepower 20 d. Roughing Filters 6 e. Heating Coil (KW) 95 SOP-114 ENCLOSURE A PAGE 2 OF 2 REVISION 21 SYSTEM INFORMATION (Cont'd) 5. Purge Exhaust System

a. Max Filter Plenum Capacity (cfm) 21,000 b. Roughing Filters 15 c. HEPA Filters 30(15 per bank) d. Charcoal Filters 11 e. Fan Capacity (cfm each) 10,000 f. Fan Motor Horsepower 25

6. Reactor Compartment Cooling Fans (each) a. Air Flow (cfm) 35,000 b. Fan Motor Horsepower 75 7. Secondary Compartment Cooling Fans (each) a. Air Flow (cfm) 45,000 b. Fan Motor Horsepower 25 1) XFN-68C 40 8. CRDM Cooling Fans (each)

a. Air Flow (cfm) 33,000 b. Fan Motor Horsepower 125

9. DRPI Data Cabinet Cooling

a. Air Flow (cfm) 7,000 b. Fan Motor Horsepower 10 10. XSW1A3 - 1C3 Tie Breaker closure affects CRDM Shroud Exhaust fan operation as follows: a. If both XFN0067A and XFN0067B are available, then regardless of operating fan combination, XFN0067C and XFN0067D will be tripped and XFN0067A and XFN0067B will start when tie breaker is closed. b. If either XFN0067A or XFN0067B is not available, then regardless of operating fan combination, neither XFN0067C nor XFN0067D will be tripped and XFN0067A and XFN0067B will not be started when tie breaker is closed.

SOP-114 ATTACHMENT IA PAGE 1 OF 3 REVISION 21 Persons completing checklist (print) Initials REACTOR BUILDING VENTILATION SYSTEM VALVE LINEUP Reviewed by SS/CRS Date/Time Date/Time started / / Date/Time completed / Valve Lineup Initial Conditions Positioning the following components to the REQUIRED POSITION aligns the system as follows: a. RB Purge Supply and Exhaust is secured and isolated. b. RB Alternate Purge Supply and Exhaust is aligned ready for startup. COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS 485' AUXILIARY BUILDING XDP0029A-AV1 IA ISOLATION VLV FOR XDP0029A&XDP0030A OPEN XDP0029B-AV1 IA ISOLATION VLV FOR XDP0029B&XDP0030B OPEN XVB06068-HR ALTERNATE PURGE EXHAUST FAN DISCH VALVE OPEN XVD06093B-HR ALT PURGE VENTL EXH FAN DISCH HEADER TC CLOSED XVA00005-AH RMA0004 SAMPLE INLET ISOLATION VALVE OPEN XVA00006-AH RMA0004 SAMPLE INLET ISOLATION VALVE CLOSED XDP0028-AV1 IA ISOLATION VALVE FOR XDP0028 OPEN XDP0031A-AV1 IA ISOLATION VLV FOR XDP0031A&XDP0032A OPEN XDP0031B-AV1 IA ISOLATION VLV FOR XDP0031B&XDP0032B OPEN ITC09261-AV1-AH IA ISOLATION VALVE FOR ITC9261 OPEN SOP-114 ATTACHMENT IA PAGE 2 OF 3 REVISION 21 Valve Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS 463' AUXILIARY BUILDING XVD16051-HR ALTERNATE PURGE HEADER TEST CONNECTION CLOSED XVG06067-AV1-HR IA ISOLATION VALVE FOR XVG06067-HR OPEN 463' FUEL HANDLING BUILDING XVB00002A-AV1-AH IA ISOLATION VALVE FOR XVB00002A-AH OPEN XVB00002A-AV2-AH (INSIDE BOX) NOTE 1 IA HDR ISOLATION VLV FOR XVB00002A-AH VALVE CLOSED (LVP) XVB00001A-AV2-AH (INSIDE BOX) NOTE 1 IA HDR ISOLATION VLV FOR XVB00001A-AH VALVE CLOSED (LVP) XVB00001A-AV1-AH IA ISOLATION VALVE FOR XVB00001A-AH OPEN XVD16050-HR BACK-UP PURGE HEADER TEST CONNECTION CLOSED XVB06092-HR ALTERNATE PURGE VENT FAN DISCHARGE VLV THROTTLED (LVP) NOTE 2 XVD06093A-HR ALT PURGE VENT FAN DISCH HDR TEST CONN CLOSED XVT06069-HR RB PURGE EXH PLENUM SUCT HDR TEST CONN CLOSED/ CAPPED XVT06070-HR RB PURGE FANS DISCH HEADER TEST CONN CLOSED/ CAPPED XVG06057-AV1-HR IA ISOLATION VALVE FOR XVG06057-HR OPEN REACTOR BUILDING XVG06056-AV1-HR (RB-463-015-62) IA ISOLATION VALVE FOR XVG06056-HR OPEN XVB00002B-AV1-AH (RB-463-015-62) IA ISOLATION VALVE FOR XVB00002B-AH OPEN XVB00002B-AV2-AH (RB-463-015-62) (INSIDE BOX) IA HDR ISOLATION VALVE FOR XVB00002B-AH VALVE CLOSED (LVP) XVB00001B-AV1-AH (RB-463-300-62) IA ISOLATION VALVE FOR XVB00001B-AH OPEN XVB00001B-AV2-AH (RB-463-300-62) (INSIDE BOX) IA HDR ISOLATION VLV FOR XVB00001B-AH VALVE CLOSED (LVP) NOTE 1: XVB00001A-AV2-AH and XVB00002A-AV2-AH are located in a common locked box. NOTE 2: Adjusted by HVAC Department to maintain 600 to 800 CFM.

SOP-114 ATTACHMENT IA PAGE 3 OF 3 REVISION 21 Valve Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS REACTOR BUILDING (Cont'd) IFS01900A-TV1-LD (RB-463-150-47) TEST VALVE FOR IFS1900A OPEN (LVP) IFS01900A-TV2-LD (RB-463-150-47) TEST VALVE FOR IFS1900A CLOSED IFS01900B-TV1-LD (RB-463-335-56) TEST VALVE FOR IFS1900B OPEN (LVP) IFS01900B-TV2-LD (RB-463-335-56) TEST VALVE FOR IFS1900B CLOSED XVG06066-AV1-HR (RB-463-256-62) IA ISOLATION VALVE FOR XVG06066-HR OPEN IFT75025-HR-LD (RB-436-90-35) IFT75025 INLET ISOLATION VALVE OPEN IFT75025-LR-LD (RB-436-90-35) IFT75025 OUTLET ISOLATION VALVE OPEN IFT75025-HT-LD (RB-436-90-35) IFT75025 TEST CONNECTION VALVE CLOSED/ CAPPED IFT75025-HD-LD (RB-436-90-35) IFT75025 SAMPLE/DRAIN VALVE CLOSED/ CAPPED IFT75025-LD-LD (RB-436-90-35) IFT75025 DISCHARGE DRAIN VALVE CLOSED/ CAPPED REACTOR BUILDING (At RBCU Cooling Units) XDP0110A-AV1 (RB-514-144-58) IA ISOLATION VALVE FOR XDP0110A OPEN XDP0110B-AV1 (RB-514-019-58) IA ISOLATION VALVE FOR XDP0110B OPEN XDP0111A-AV1 (RB-514-144-50) IA ISOLATION VALVE FOR XDP0111A OPEN XDP0111B-AV1 (RB-514-341-58) IA ISOLATION VALVE FOR XDP0111B OPEN SOP-114 ATTACHMENT IB PAGE 1 OF 6 REVISION 21 Persons completing checklist (print) Initials CRDM COOLING WATER SYSTEM VALVE LINEUP Reviewed by SS/CRS Date/Time Date/Time started / / Date/Time completed / Valve Lineup Initial Conditions Positioning the following components to the REQUIRED POSITION aligns the system for operation. COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS 436' INTERMEDIATE BUILDING (East Pen) XVT07522-AC INDUSTRIAL CLR AC RETURN HDR DRAIN VLV CLOSED/ CAPPED XVT07516-AC CRDM COOLER AC SUPPLY HEADER DRAIN VLV CLOSED/ CAPPED XVT07547-AC CRDM COOLER AC SUPPLY HEADER VENT VALVE CLOSED/ CAPPED NOTE 1 XVB07500A-AC CRDM COOLING WATER PUMP A DISCH VALVE OPEN XVT07532A-AC HIGH ROOT TO IPS05572 CLOSED XVT07531A-AC HIGH ROOT TO IPI05554 OPEN XVB07513A-AC CRDM COOLING WATER PUMP A SUCTION VALVE OPEN XVB07500B-AC CRDM COOLING WATER PUMP B DISCH VALVE OPEN XVT07532B-AC HIGH ROOT TO IPS05588 CLOSED XVT07531B-AC HIGH ROOT TO IPI05589 OPEN NOTE 1 - The indicated valves should remain uncapped prior to filling and venting the system. Upon completion of fill and vent the procedure directs capping the vent points.

SOP-114 ATTACHMENT IB PAGE 2 OF 6 REVISION 21 Valve Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS 436' INTERMEDIATE BUILDING (East Pen) (Cont'd) XVB07513B-AC CRDM COOLING WATER PUMP B SUCTION VALVE OPEN XVT28762-DN DN MAKEUP TO CRDM COOLING SYSTEM FILL VALVE CLOSED XVG07525-AC CHEMICAL FEED TANK AC OUTLET VALVE CLOSED XVG07524-AC CHEMICAL FEED TANK AC INLET VALVE CLOSED XVT07527-AC CHEMICAL FEED TANK CHEMICAL FILL VALVE CLOSED XVT07526-AC CHEMICAL FEED TANK DRAIN VALVE CLOSED XVG07550-AC CHEM FEED TK VENT VALVE CLOSED XVT07537-AC HIGH ROOT TO IPI05553 OPEN XVT07528-AC AC EXPANSION TANK SURGE ISOL VALVE OPEN XVT17514-AC AC SYSTEM EXPANSION TANK DRAIN ISOLATION VALVE CLOSED/ CAPPED NOTE 1 XVT07529-AC AC SYSTEM EXPANSION TANK ISOLATION VALVE OPEN XVT07530-AC AC SYSTEM EXPANSION TANK VENT VALVE OPEN ILI05552-HR-AC HIGH ROOT TO ILT05552 OPEN ILI05552-LR-AC LOW ROOT TO ILT05552 OPEN XVT07539-AC AC SYS EXP TANK INST RACK HIGH ISOL VLV OPEN XVT07538-AC AC SYS EXP TANK INST RACK LOW ISOL VLV OPEN XVG17513-AC AC SYSTEM VENT ISOLATION VALVE CLOSED/ CAPPED NOTE 1 412' INTERMEDIATE BUILDING (East Pen) XVT07533-AC HIGH ROOT TO IFI05575 OPEN XVT07534-AC LOW ROOT TO IFI05575 OPEN XVT07549-AC INDUSTRIAL CLR AC RETURN HDR DRAIN VLV CLOSED/ CAPPED XVT07548-AC AC SYS INDUSTRIAL CLR OUT HDR DRAIN VLV CLOSED/ CAPPED NOTE 1 - The indicated valves should remain uncapped prior to filling and venting the system. Upon completion of fill and vent the procedure directs capping the vent points.

SOP-114 ATTACHMENT IB PAGE 3 OF 6 REVISION 21 Valve Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS YD-125'E (CRDM Cooling Tower) XVG07517-AC AC SYSTEM FI SUPPLY HEADER ISOL VALVE OPEN XVG07520-AC CRDM COOLING WATER RETURN HDR FILL VLV CLOSED XVT07546-AC AC SYSTEM INDUSTRIAL COOLER DRAIN VALVE CLOSED XVA17517-AC ISOLATION VLV FOR TEMP CRDM COOLER INSTALL CLOSED/ FLANGED XVA17518-AC ISOLATION VLV FOR TEMP CRDM COOLER INSTALL OPEN XVA17519-AC RETURN ISOLATION VLV FOR TEMP CRDM COOLER CLOSED/ FLANGED XVA17520-AC ISOLATION VLV FOR TEMP CRDM COOLER INSTALL OPEN XVB17506-AC AC IND CLR SPRAY PUMP C SUCTION VALVE OPEN XVB17508-AC AC IND CLR SPRAY PUMP C DISCH VALVE OPEN XVA17524A-AC AC SYSTEM INDUSTRIAL COOLER BLOW DOWN VALVE CLOSED NOTE 2 XVB17511-AC AC IND CLR SPRAY PUMP D DISCH VALVE OPEN XVB17509-AC AC IND CLR SPRAY PUMP D SUCTION VALVE OPEN XVT07535A-AC HIGH ROOT TO IPI05569 OPEN XVG07511A-AC AC SYS INDUSTRIAL CLR INLET ISOL VALVE OPEN XVT07514A-AC HIGH ROOT TO IPX05576 CLOSED/ CAPPED NOTE 1 XVT07512A-AC AC SYS INDUSTRIAL CLR OUTLET ISOL VALVE OPEN XVT07515A-AC HIGH ROOT TO IPX05577 CLOSED/ CAPPED NOTE 1 XVG07511B-AC AC SYS INDUSTRIAL CLR INLET ISOL VALVE OPEN XVT07514B-AC HIGH ROOT TO IPX05578 CLOSED/ CAPPED NOTE 1 NOTE 1 - The indicated valves should remain uncapped prior to filling and venting the system. Upon completion of fill and vent the procedure directs capping the vent points. NOTE 2 - This valve will be throttled open per the procedure after system startup.

SOP-114 ATTACHMENT IB PAGE 4 OF 6 REVISION 21 Valve Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS YD-125'E (CRDM Cooling Tower) (Cont'd) XVB17502-AC AC IND CLR SPRAY PUMP A DISCH VALVE OPEN XVA17524B-AC AC SYSTEM INDUSTRIAL COOLER BLOW DOWN VALVE CLOSED NOTE 2 XVB17500-AC AC IND CLR SPRAY PUMP A SUCTION VALVE OPEN XVA17522-AC ISOLATION VLV FOR TEMP CRDM COOLER INSTALL OPEN XVA17521-AC ISOLATION VLV FOR TEMP CRDM COOLER INSTALL OPEN XVB17505-AC AC IND CLR SPRAY PUMP B DISCH VALVE OPEN XVB17503-AC AC IND CLR SPRAY PUMP B SUCTION VALVE OPEN XVA17525-AC AC SYSTEM INDUSTRIAL COOLER FI MAKE-UP VALVE TO CHEMICAL ADDITION CLOSED/ CAPPED XVT07512B-AC AC SYS INDUSTRIAL CLR OUTLET ISOL VALVE OPEN XVT07515B-AC HIGH ROOT TO IPX05579 CLOSED/ CAPPED NOTE 1 XVT07536A-AC HIGH ROOT TO IPI05570 OPEN XVT07536B-AC HIGH ROOT TO IPI05590 OPEN XVT07512D-AC AC SYS INDUSTRIAL CLR OUTLET ISOL VALVE OPEN XVT07515D-AC HIGH ROOT TO IPX05585 CLOSED/ CAPPED NOTE 1 XVG07511D-AC AC SYS INDUSTRIAL CLR INLET ISOL VALVE OPEN XVT07514D-AC HIGH ROOT TO IPX05584 CLOSED/ CAPPED NOTE 1 NOTE 1 - The indicated valves should remain uncapped prior to filling and venting the system. Upon completion of fill and vent the procedure directs capping the vent points. NOTE 2 - This valve will be throttled open per the procedure after system startup.

SOP-114 ATTACHMENT IB PAGE 5 OF 6 REVISION 21 Valve Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS YD-125'E (CRDM Cooling Tower) (Cont'd) XVT07512C-AC AC SYS INDUSTRIAL CLR OUTLET ISOL VALVE OPEN XVT07515C-AC HIGH ROOT TO IPX05583 CLOSED/ CAPPED NOTE 1 XVG07511C-AC AC SYS INDUSTRIAL CLR INLET ISOL VALVE OPEN XVT07514C-AC HIGH ROOT TO IPX05582 CLOSED/ CAPPED NOTE 1 XVT07535B-AC HIGH ROOT TO IPI05592 OPEN XVM07570B-AC XPP0156B PUMP CASING VENT VALVE CLOSED XVM07570D-AC XPP0156D PUMP CASING VENT VALVE CLOSED NOTE 1 - The indicated valves should remain uncapped prior to filling and venting the system. Upon completion of fill and vent the procedure directs capping the vent points.

SOP-114 ATTACHMENT IB PAGE 6 OF 6 REVISION 21 Valve Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS REACTOR BUILDING XVG07507D-AC (RB-436-110-45) CRDM COOLER AC SUPPLY HEADER ISOL VALVE OPEN XVT07509D-AC (RB-436-110-45) HIGH ROOT TO IPX05566 CLOSED/ CAPPED XVT07506D-AC (RB-436-110-45) CRDM COOLER AC RETURN HEADER ISOL VLV OPEN XVG07507C-AC (RB-436-110-45) CRDM COOLER AC SUPPLY HEADER ISOL VALVE OPEN XVT07508D-AC (RB-436-110-45) HIGH ROOT TO IPX05562 CLOSED/ CAPPED NOTE 1 XVT07509C-AC (RB-436-110-45) HIGH ROOT TO IPX05567 CLOSED/ CAPPED XVT07506C-AC (RB-436-110-45) CRDM COOLER AC RETURN HEADER ISOL VLV OPEN XVT07508C-AC (RB-436-110-45) HIGH ROOT TO IPX05563 CLOSED/ CAPPED NOTE 1 XVG07507B-AC (RB-436-120-40) CRDM COOLER AC SUPPLY HEADER ISOL VALVE OPEN XVT07509B-AC (RB-436-120-40) HIGH ROOT TO IPX05556 CLOSED/ CAPPED XVT07506B-AC (RB-436-120-40) CRDM COOLER AC RETURN HEADER ISOL VLV OPEN XVT07508B-AC (RB-436-120-40) HIGH ROOT TO IPX05560 CLOSED/ CAPPED NOTE 1 XVG07507A-AC (RB-436-120-40) CRDM COOLER AC SUPPLY HEADER ISOL VALVE OPEN XVT07509A-AC (RB-436-120-40) HIGH ROOT TO IPX05557 CLOSED/ CAPPED XVT07506A-AC (RB-436-120-40) CRDM COOLER AC RETURN HEADER ISOL VLV OPEN XVT07508A-AC (RB-436-120-40) HIGH ROOT TO IPX05561 CLOSED/ CAPPED NOTE 1 XVT07545-AC (RB-436-120-60) INDUSTRIAL CLR AC RETURN HDR DRAIN VLV CLOSED/ CAPPED XVT07543-AC (RB-436-120-60) INDUSTRIAL CLR AC RETURN HDR TEST CONN CLOSED/ CAPPED NOTE 1 XVT07542-AC (RB-436-135-60) CRDM COOLER AC SUPPLY HEADER DRAIN VLV CLOSED/ CAPPED XVT07540-AC (RB-436-135-60) CRDM COOLER AC SUPPLY HEADER TEST CONN CLOSED/ CAPPED NOTE 1 NOTE 1 - The indicated valves should remain uncapped prior to filling and venting the system. Upon completion of fill and vent the procedure directs capping the vent points.

SOP-114 ATTACHMENT IIA PAGE 1 OF 5 REVISION 21 Persons completing checklist (print) Initials REACTOR BUILDING VENTILATION SYSTEM ELECTRICAL LINEUP Reviewed by SS/CRS Date/Time Date/Time started / / Date/Time completed / Electrical Lineup Initial Conditions Positioning the following components to the REQUIRED POSITION aligns the system for operation with electrical power available to essential components. COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XMC1A3X 463' AUXILIARY BUILDING XMC1A3X 02EF POST ACCIDENT H2 REMOVAL PURGE EXHAUST FAN XFN0096-HR ON XMC1A3X 09AD SECONDARY COMPT LOOP C COOLING FAN A XFN0070A-AH ON XMC1A3X 10CF SECONDARY CMPT LOOP B COOLING FAN A XFN0069A-AH ON XMC1A3X 10GK RB CHARCOAL CLEAN-UP UNIT A XFN0066A-AH ON XMC1A3X 11AD DRPI CLG UNIT BOOSTER PUMP XPP0149-SW ON XMC1A3X 11EH SECONDARY CMPT LOOP A COOLING FAN A XFN0068A-AH ON XMC1B3X 463' AUXILIARY BUILDING XMC1B3X 03ABL EDDY CURRENT BRAKE FOR XFN64A & 65A ON XMC1B3X 03ABR EDDY CURRENT BRAKE FOR XFN64B & 65B ON XMC1B3X 09AD REFUELING WATER SURFACE SUPPLY FAN B XFN0007B-AH ON XMC1B3X 10FJ RB CHARCOAL CLN-UP UNIT FAN B XFN0066B-AH ON SOP-114 ATTACHMENT IIA PAGE 2 OF 5 REVISION 21 Electrical Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XMC1C3X 463' AUXILIARY BUILDING XMC1C3X 02AD CONTR ROD POSITION DATA CAB CLG FAN XFN0107-VL ON XMC1C3X 02EH SEC COMPT LOOP B COOLING FAN C XFN0069C-AH ON XMC1C3X 02IL SEC COMPT LOOP C COOLING FAN C XFN0070C-AH ON XMC1C3X 03EH REFUELING WATER SURFACE SUPPLY FAN A XFN0007A-AH ON XMC1C3X 03IL REFUELING WATER SURFACE EXHAUST FAN XFN0008-AH ON XMC1C3X 08AE SEC COMPT LOOP A COOLING FAN C XFN0068C-AH ON XMC1B3Y 463' FUEL HANDLING BUILDING XMC1B3Y 02IL REACTOR BLDG PURGE EXH FAN B XFN0013B-AH ON XMC1B3Y 03AD SECONDARY COMPARTMENT LOOP A COOLING FAN B XFN0068B-AH ON XMC1B3Y 03EH SECONDARY COMPARTMENT LOOP B COOLING FAN B XFN0069B-AH ON XMC1B3Y 03IL SECONDARY COMPARTMENT LOOP C COOLING FAN B XFN0070B-AH ON XMC1B3Y 09CD REACTOR BLDG PURGE SUPPLY FAN B XFN0011B-AH ON XMC1A2X 485' AUXILIARY BUILDING XMC1A2X 01FG RB PURGE SUPPLY FAN A XFN0011A-AH ON XMC1A2X 01HI RB PURGE EXHAUST FAN A XFN0013A-AH ON XSW1A2 485' AUXILIARY BUILDING XSW1A2 05C RB PURGE SUPPLY HTG COIL A XHC0012A-AH CLOSED XSW1B2 485' AUXILIARY BUILDING XSW1B2 02C RB PURGE SUPPLY HTG COIL B XHC0012B-AH CLOSED SOP-114 ATTACHMENT IIA PAGE 3 OF 5 REVISION 21 Electrical Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XSW1DB1 463' AUXILIARY BUILDING XSW1DB1 06C RBCU MFN0097D XFN0065B SLOW SPEED MOTOR RACKED IN XSW1DB1 06C CCP CLOSING CNTRL PWR MFN 97D XFN 65B ON XSW1DB1 06C TCP TRIPPING CNTRL PWR MFN 97D XFN 65B ON XSW1DB1 06C CHARGING POWER ON XSW1DB1 06D RBCU MFN0097B XFN0064B SLOW SPEED MOTOR RACKED IN XSW1DB1 06D CCP CLOSING CNTRL PWR MFN 97B XFN 64B ON XSW1DB1 06D TCP TRIPPING CNTRL PWR MFN 97B XFN 64B ON XSW1DB1 06D CHARGING POWER ON XSW1DB1 07B RBCU MFN0096B XFN0064B FAST SPEED MOTOR RACKED IN XSW1DB1 07B CCP CLOSING CNTRL PWR MFN 96B XFN 64B ON XSW1DB1 07B TCP TRIPPING CNTRL PWR MFN 96B XFN 64B ON XSW1DB1 07B CHARGING POWER ON XSW1DB1 07C RBCU MFN0096D XFN0065B FAST SPEED MOTOR RACKED IN XSW1DB1 07C CCP CLOSING CNTRL PWR MFN 96D XFN 65B ON XSW1DB1 07C TCP TRIPPING CNTRL PWR MFN 96D XFN 65B ON XSW1DB1 07C CHARGING POWER ON APN01DB1 463' AUXILIARY BUILDING APN01DB1 09 XFN0064B/65B EDDY CURRENT BRAKE ON XSW1A3 436' AUXILIARY BUILDING XSW1A3 02A REACTOR COMP COOLING FAN A XFN0009A-AH RACKED IN XSW1A3 02A CCP CLOSING CNTRL PWR XFN0009A-AH ON XSW1A3 02A TCP TRIPPING CNTRL PWR XFN0009A-AH ON XSW1A3 02A CHARGING POWER ON SOP-114 ATTACHMENT IIA PAGE 4 OF 5 REVISION 21 Electrical Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XSW1B3 436' AUXILIARY BUILDING XSW1B3 03A RB REACTOR COMPT CLG FAN B XFN0009B-AH RACKED IN XSW1B3 03A CCP CLOSING CNTRL PWR XFN0009B-AH ON XSW1B3 03A TCP TRIPPING CNTRL PWR XFN0009B-AH ON XSW1B3 03A CHARGING POWER ON XSW1DA1 463' INTERMEDIATE BUILDING XSW1DA1 05B RBCU MFN0096C XFN0065A FAST SPEED MOTOR RACKED IN XSW1DA1 05B CCP CLOSING CNTRL PWR XFN 65A MFN 96C ON XSW1DA1 05B TCP TRIPPING CNTRL PWR XFN 65A MFN 96C ON XSW1DA1 05B CHARGING POWER ON XSW1DA1 05C RBCU MFN0097A XFN0064A SLOW SPEED MOTOR RACKED IN XSW1DA1 05C CCP CLOSING CNTRL PWR XFN 64A MFN 97A ON XSW1DA1 05C TCP TRIPPING CNTRL PWR XFN 64A MFN 97A ON XSW1DA1 05C CHARGING POWER ON XSW1DA1 06B RBCU MFN0096A XFN0064A FAST SPEED MOTOR RACKED IN XSW1DA1 06B CCP CLOSING CNTRL PWR XFN 64A MFN 96A ON XSW1DA1 06B TCP TRIPPING CNTRL PWR XFN 64A MFN 96A ON XSW1DA1 06B CHARGING POWER ON XSW1DA1 06C RBCU MFN0097C XFN0065A SLOW SPEED MOTOR RACKED IN XSW1DA1 06C CCP CLOSING CNTRL PWR XFN 65A MFN 97C ON XSW1DA1 06C TCP TRIPPING CNTRL PWR XFN 65A MFN 97C ON XSW1DA1 06C CHARGING POWER ON APN01DA1 463' INTERMEDIATE BUILDING APN01DA1 09 EDDY CURR BRAKES - RBCU FANS XFN64A & 65A ON SOP-114 ATTACHMENT IIA PAGE 5 OF 5 REVISION 21 Electrical Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XMC1A1X 412' TURBINE BUILDING XMC1A1X 09KL POST ACC H2 REMOVAL EXH PUR FAN XFN0095-HR ON APN-4005 REACTOR BUILDING-463-325-52 APN-4005 26, 28, 30 SPACE HEATERS FOR FANS XFN68C, 9A & 9B ON APN-4005 31, 33, 35 SPACE HEATERS FOR FANS XFN66A & 66B ON APN-5005 REACTOR BUILDING-463-325-52 APN-5005 05 SPACE HTR FOR XFN0064A-AH (MFN0096A-AH) ON APN-5005 06 SPACE HTR FOR XFN0064B-AH (MFN0097B-AH) ON APN-5005 07 SPACE HTR FOR XFN0064A-AH (MFN0097A-AH) ON APN-5005 08 SPACE HTR FOR XFN0064B-AH (MFN0096B-AH) ON APN-5005 09 SPACE HTR FOR XFN0065A-AH (MFN0096C-AH) ON APN-5005 10 SPACE HTR FOR XFN0065B-AH (MFN0096D-AH) ON APN-5005 11 SPACE HTR FOR XFN0065A-AH (MFN0097C-AH) ON APN-5005 12 SPACE HTR FOR XFN0065B-AH (MFN0097D-AH) ON APN-5005 15 XFN0135-ELEVATOR MACHINE ROOM EXHAUST FAN ON SOP-114 ATTACHMENT IIB PAGE 1 OF 3 REVISION 21 Persons completing checklist (print) Initials CRDM COOLING WATER SYSTEM ELECTRICAL LINEUP Reviewed by SS/CRS Date/Time Date/Time started / / Date/Time completed / Electrical Lineup Initial Conditions Positioning the following components to the REQUIRED POSITION aligns the system for operation with electrical power available to essential components. COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XSW1C3 436' AUXILIARY BUILDING XSW1C3 02D CRDM COOLING SYS FAN C XFN0067C-AH RACKED IN XSW1C3 02D CCP CLOSING CNTRL PWR XFN0067C-AH ON XSW1C3 02D TCP TRIPPING CNTRL PWR XFN0067C-AH ON XSW1C3 02D CHARGING POWER ON XSW1A3 436' AUXILIARY BUILDING XSW1A3 01C CRDM COOLING SYS FAN A XFN0067A-AH RACKED IN XSW1A3 01C CCP CLOSING CNTRL PWR XFN0067A-AH ON XSW1A3 01C TCP TRIPPING CNTRL PWR XFN0067A-AH ON XSW1A3 01C CHARGING POWER ON XSW1A3 03A CRDM COOLING SYS FAN D XFN0067D-AH RACKED IN XSW1A3 03A CCP CLOSING CNTRL PWR XFN0067D-AH ON XSW1A3 03A TCP TRIPPING CNTRL PWR XFN0067D-AH ON XSW1A3 03A CHARGING POWER ON SOP-114 ATTACHMENT IIB PAGE 2 OF 3 REVISION 21 Electrical Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XSW1B3 436' AUXILIARY BUILDING XSW1B3 02D CRDM COOLING SYS FAN B XFN0067B-AH RACKED IN XSW1B3 02D CCP CLOSING CNTRL PWR XFN0067B-AH ON XSW1B3 02D TCP TRIPPING CNTRL PWR XFN0067B-AH ON XSW1B3 02D CHARGING POWER ON APN01C3 436' AUXILIARY BUILDING APN01C3 05, 07 CRDM COOLING FAN BRAKE POWER SUPPLY ON XMC1A2X 485' AUXILIARY BUILDING XMC1A2X 01LM IND CLG WTR HT TRACE CONTR PNL XPN2015 ON XPN 5533 436' AUXILIARY BUILDING (ACROSS FROM XSW1A3) XPN 5533 DSW-A ON XPN 5533 DSW-B ON XPN 5533 DSW-C ON XPN 5533 DSW-D ON APN4005 REACTOR BUILDING-463-325-52 APN4005 19, 21, 23 SPACE HEATERS FOR FANS XFN-67A-67D ON XMC1DA2X 463' INTERMEDIATE BUILDING XMC1DA2X 06AD CRDM COOLING WATER INLET VALVE XVG7501-AC ON XMC1DA2X 11IM CRDM COOLING WATER OUTLET XVG7503-AC ON XMC1DB2X 436' INTERMEDIATE BUILDING XMC1DB2X 07AD CRDM CLG WTR OUTLET VLV XVG7504-AC ON XMC1DB2X 07IM CRDM CLG WTR INLET VLV XVG7502-AC ON CHG B CHG A SOP-114 ATTACHMENT IIB PAGE 3 OF 3 REVISION 21 Electrical Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XMC1B1X 436' TURBINE BUILDING XMC1B1X 05CD IND COOLER CRDM COIL FAN D XFN0145D-AC ON XMC1B1X 06AC IND CLR IMMERSION HEATERS A&B XHC0112-AC ON XMC1B1X 09IJ INDUST COOLER CRDM COIL FAN XFN0145B-AC ON XMC1B1X 09KL INDUST COOLER CRDM COIL FAN XFN0145C-AC ON XMC1B1X 10CD IND CLR CRDM COIL SPR NORM PP XPP0156A-AC ON XMC1B1X 11GH IC CRDM COIL SPRAY STNDBY PP XPP0156B-AC ON XMC1B1X 11KL IND CLR CRDM COILS SPR NORM PP XPP0156C-AC ON XMC1B1X 13CD IND COOLER CRDM COIL FAN XFN0145A-AC ON XMC1B1X 15LM IC CRDM COIL SPR STNDBY PP XPP0156D-AC ON APN 5016 436' TURBINE BUILDING APN 5016 17 XTF 0146 ON XMC1A1X 412' TURBINE BUILDING XMC1A1X 04EG CRDM COOLING WATER PUMP B XPP0157B-AC ON XMC1A1X 12IK CRDM COOLING WTR PP A XPP0157A-AC ON YD-125'E (CRDM Cooling Tower) XDS0037-AC CRDM CLG TWR LVL CNTRL DISC SWTCH CLOSED XDS0225-AC IND COOLER IMMERSION HTR A DISC SW CLOSED XDS0226-AC IND COOLER IMMERSION HTR B DISC SW CLOSED CHG B SOP-114 ATTACHMENT III PAGE 1 OF 1 REVISION 21 REACTOR BUILDING VENTILATION SYSTEM INSTRUMENT LINEUP Instrument Lineup Initial Conditions Aligning the following components ensures the Reactor Building Ventilation System Instrumentation is filled, vented and aligned for service. SS/CRS may authorize alignment of more than one instrument. COMPONENT: IFI05575 INDUSTRIAL CLR AC RETURN HDR FLOW IND LOCATION: IB-436 (East Penetration by back door) SS/CRS AUTHR:________ EQUIPMENT AFFECTED: Local indicator Initials Control Room contacted: Initials Ensure valved in & vent: Verified by: Name Name Time/Date completed: /

SOP-114 ATTACHMENT IV PAGE 1 OF 4 REVISION 21 Persons completing checklist (print) Initials REACTOR BUILDING VENTILATION SYSTEM CONTROL PANEL LINEUP Reviewed by SS/CRS Date/Time Date/Time Started / Date/Time Completed / Control Panel Lineup Initial Conditions Positioning the following components to the REQUIRED POSITION aligns the Reactor Building Ventilation System ready for startup or for automatic initiation if required. COMPONENT DESCRIPTION POWER AVAILABLE REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS YES NO MAIN CONTROL BOARD XFN0064A-AH 1A NORM AFTER STOP XFN0064A-AH 1A SLOW AFTER STOP XFN0065A-AH 2A NORM AFTER STOP XFN0065A-AH 2A SLOW AFTER STOP XFN0064B-AH 1B NORM AFTER STOP XFN0064B-AH 1B SLOW AFTER STOP XFN0065B-AH 2B NORM AFTER STOP XFN0065B-AH 2B SLOW AFTER STOP RBCU TRAIN A EMERG XFN64A XFN65A NOTE 1 XDP-110A RBCU 64A HEPA FLTR BYP DMPR AUTO/ BYP XDP-111A RBCU 65A HEPA FLTR BYP DMPR AUTO/ BYP NOTE 1: The RBCU TRAIN A(B) EMERG Switch must be selected to an operable RBCU.

SOP-114 ATTACHMENT IV PAGE 2 OF 4 REVISION 21 Control Panel Lineup (Cont'd) COMPONENT DESCRIPTION POWER AVAILABLE REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS YES NO MAIN CONTROL BOARD (Cont'd) RBCU TRAIN B EMERG XFN64B XFN65B NOTE 1 XDP-110B RBCU 64B HEPA FLTR BYP DMPR AUTO/ BYP XDP-111B RBCU 65B HEPA FLTR BYP DMPR AUTO/ BYP MVG-7501 TO CRDM CLR ISOL (ORB) AUTO/ OPEN MVG-7502 TO CRDM CLR ISOL (IRB) AUTO/ OPEN MVG-7503 FR CRDM CLR ISOL (IRB) AUTO/ OPEN MVG-7504 FR CRDM CLR ISOL (ORB) AUTO/ OPEN HVAC CONTROL PANEL XVM-6795 PLEN DELUGE (RB Char Cleanup) STOP XFN-66A FAN A (RB Char Cleanup) STOP XFN-66B FAN B (RB Char Cleanup) STOP XFN-67A FAN A (CRDM Shroud Exhaust) AFTER STOP XFN-67B FAN B (CRDM Shroud Exhaust) AFTER STOP XFN-67C FAN C (CRDM Shroud Exhaust) AFTER STOP XFN-67D FAN D (CRDM Shroud Exhaust) AFTER STOP XFN-68A SG A FAN A (SG Compart Cooling) STOP XFN-68B SG A FAN B (SG Compart Cooling) STOP XFN-68C SG A FAN C (SG Compart Cooling) STOP XFN-69A SG B FAN A (SG Conpart Cooling) STOP XFN-69B SG B FAN B (SG Compart Cooling) STOP XFN-69C SG B FAN C (SG Compart Cooling) STOP XFN-70A SG C FAN A (SG Compart Cooling) STOP XFN-70B SG C FAN B (SG Compart Cooling) STOP NOTE 1: The RBCU TRAIN A(B) EMERG Switch must be selected to an operable RBCU.

SOP-114 ATTACHMENT IV PAGE 3 OF 4 REVISION 21 Control Panel Lineup (Cont'd) COMPONENT DESCRIPTION POWER AVAILABLE REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS YES NO HVAC CONTROL PANEL (Cont'd) XFN-70C SG C FAN C (SG Compart Cooling) STOP XFN-95 6 INCH RB PUR INTAKE FAN STOP MVB-6063 H2 REMOVAL ALT PUR THROT CLOSED XFN-96 ALT PUR EXH FAN STOP PVG-6066 CNTMT PUR EXH ISOL VLV AUTO/ CLOSED PVG-6067 CNTMT PUR EXH ISOL VLV AUTO/ CLOSED PVG-6056 ALT PUR SPLY ISOL VLV AUTO/ CLOSED PVG-6057 ALT PUR SPLY ISOL VLV AUTO/ CLOSED PVT-3164 DRPI CLG UNIT COIL ISOL AUTO/ CLOSED PVT-3165 PVT-3169 DRPI CLG UNIT COIL ISOL AUTO/ CLOSED XFN-107 RPI CABINETS CLG FAN STOP XPP-149 DRPI CLG UNIT BSTR PP STOP XFN-9A FAN A (RB Cooling) AFTER STOP XFN-9B FAN B (RB Cooling) AFTER STOP XFN-11A SPLY FAN A (RB Purge) STOP XFN-13A EXH FAN A (RB Purge) STOP XFN-11B SPLY FAN B (RB Purge) STOP XFN-13B EXH FAN B (RB Purge) STOP PVB-1A CNTMT SPLY ISOL AUTO/ CLOSED PVB-2A CNTMT EXH ISOL AUTO/ CLOSED PVB-1B CNTMT SPLY ISOL AUTO/ CLOSED PVB-2B CNTMT EXH ISOL AUTO/ CLOSED XDP-28 INTAKE DMPR (RB Purge) CLOSE XVM-6760 EXH PLEN DELUGE OFF SOP-114 ATTACHMENT IV PAGE 4 OF 4 REVISION 21 Control Panel Lineup (Cont'd) COMPONENT DESCRIPTION POWER AVAILABLE REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS YES NO CRDM COOLING TOWER CONTROL PANEL (YD-125'E) XPP0156A AC SYS INDUSTRIAL COOLER SPRAY PUMP A OFF XPP0156B AC SYS INDUSTRIAL COOLER SPRAY PUMP B OFF XPP0156C AC SYS INDUSTRIAL COOLER SPRAY PUMP C OFF XPP0156D AC SYS INDUSTRIAL COOLER SPRAY PUMP D OFF XPP0156A, B, C, AND D INDUSTRIAL COOLER CRDM COIL NORMAL AND STANDBY PUMP'S N/A N/A NOTE 2 XFN-145A INDUSTRIAL COOLER CRDM COIL FAN OFF XFN-145B INDUSTRIAL COOLER CRDM COIL FAN OFF XFN-145C INDUSTRIAL COOLER CRDM COIL FAN OFF XFN-145D INDUSTRIAL COOLER CRDM COIL FAN OFF XPP-157A AND B CRDM COOLING WATER PUMP OFF XDS0027 (412' Reactor Building) SWITCH SPACE HEATER CONT FOR XFN-68C N/A N/A ON XDS0028 (412' Reactor Building) SWITCH SPACE HEATER CONT FOR XFN-9A & B N/A N/A ON XDS0034 (436' Reactor Building) (Under C MS Line) SWITCH SPACE HEATER CONT FOR XFN-67A & B N/A N/A ON XDS0035 (436' Reactor Building) SWITCH SPACE HEATER CONT FOR XFN-67C & D N/A N/A ON XDS0026 (RB-515-300-50) (At XFN0066B) SWITCH XFN0066A&B SPACE HTR DISCONNECT SW N/A N/A ON RB-463-340-48 XPN-5170-AH RB REFUELING WTR SURFACE FANS N/A N/A OFF XPN5236-VL (RB-463-350-40) (Above elevator/requires GGMK Key) XFN0135-VL RB ELEVATOR MACH ROOM EXHAUST FAN N/A N/A AUTO NOTE 2: Positioned to A-C or B-D position. CHG B SOP-114 ATTACHMENT V PAGE 1 OF 3 REVISION 21 Persons completing checklist (print) Initials CRDM SHUTDOWN AND COOLING DRAIN ELECTRICAL LINEUP Reviewed by SS/CRS Date/Time Date/Time started / / Date/Time completed / Electrical Lineup Initial Conditions Positioning the following components to the REQUIRED POSITION aligns the system for shutdown and draining. COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XSW1C3 436' AUXILIARY BUILDING XSW1C3 02D CRDM COOLING SYS FAN C XFN0067C-AH RACKED OUT XSW1C3 02D CCP CLOSING CNTRL PWR XFN0067C-AH OFF XSW1C3 02D TCP TRIPPING CNTRL PWR XFN0067C-AH OFF XSW1C3 02D CHARGING POWER OFF XSW1A3 436' AUXILIARY BUILDING XSW1A3 01C CRDM COOLING SYS FAN A XFN0067A-AH RACKED OUT XSW1A3 01C CCP CLOSING CNTRL PWR XFN0067A-AH OFF XSW1A3 01C TCP TRIPPING CNTRL PWR XFN0067A-AH OFF XSW1A3 01C CHARGING POWER OFF XSW1A3 03A CRDM COOLING SYS FAN D XFN0067D-AH RACKED OUT XSW1A3 03A CCP CLOSING CNTRL PWR XFN0067D-AH OFF XSW1A3 03A TCP TRIPPING CNTRL PWR XFN0067D-AH OFF XSW1A3 03A CHARGING POWER OFF SOP-114 ATTACHMENT V PAGE 2 OF 3 REVISION 21 Electrical Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS XSW1B3 436' AUXILIARY BUILDING XSW1B3 02D CRDM COOLING SYS FAN B XFN0067B-AH RACKED OUT XSW1B3 02D CCP CLOSING CNTRL PWR XFN0067B-AH OFF XSW1B3 02D TCP TRIPPING CNTRL PWR XFN0067B-AH OFF XSW1B3 02D CHARGING POWER OFF APN01C3 436' AUXILIARY BUILDING APN01C3 05, 07 CRDM COOLING FAN BRAKE POWER SUPPLY OFF XPN 5533 436' AUXILIARY BUILDING XPN 5533 DSW-A OFF XPN 5533 DSW-B OFF XPN 5533 DSW-C OFF XPN 5533 DSW-D OFF APN4005 REACTOR BUILDING-463-325-52 APN4005 19, 21, 23 SPACE HEATERS FOR FANS XFN-67A-67D OFF XMC1B1X 436' TURBINE BUILDING XMC1B1X 05CD IND COOLER CRDM COIL FAN D XFN0145D-AC OFF XMC1B1X 06AC IND CLR IMMERSION HEATERS A&B XHC0112-AC OFF XMC1B1X 09IJ INDUST COOLER CRDM COIL FAN XFN0145B-AC OFF XMC1B1X 09KL INDUST COOLER CRDM COIL FAN XFN0145C-AC OFF XMC1B1X 10CD IND CLR CRDM COIL SPR NORM PP XPP0156A-AC OFF XMC1B1X 11GH IC CRDM COIL SPRAY STNDBY PP XPP0156B-AC OFF XMC1B1X 11KL IND CLR CRDM COILS SPR NORM PP XPP0156C-AC OFF XMC1B1X 13CD IND COOLER CRDM COIL FAN XFN0145A-AC OFF XMC1B1X 15LM IC CRDM COIL SPR STNDBY PP XPP0156D-AC OFF SOP-114 ATTACHMENT V PAGE 3 OF 3 REVISION 21 Electrical Lineup (Cont'd) COMPONENT DESCRIPTION REQUIRED POSITION ACTUAL POSITION INITIALS VERIFIERS INITIALS APN 5016 436' TURBINE BUILDING APN 5016 17 XTF 0146 OFF XMC1A1X 412' TURBINE BUILDING XMC1A1X 04EG CRDM COOLING WATER PUMP B XPP0157B-AC OFF XMC1A1X 12IK CRDM COOLING WTR PP A XPP0157A-AC OFF

JPM NO:NJPP-4022015 NRC InPlant i RO & SRO-U: Locally Dilute the Boric Acid TanksV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Page 1 of 1 TASK:000-055-05-01PREFERRED EVALUATION METHODSIMULATEEVALUATION TIME15TIME CRITICALNOTASK STANDARD:The 'A' BAT has been drained to 50%, then refilled to 90-95%. XVD08324A-CS is closed and both the drain rig and fill rig have been removed.PREFERRED EVALUATION LOCATIONPLANTTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)(6)RESPOND TO LOSS OF OFF SITE AND ON SITE POWERTOOLS:NJPP-402 Handout - EOP-6.0, Attachment 6

REFERENCES:

TERMINATING CUE:The "A" BAT has been diluted.CANDIDATE:EOP-6.0LOSS OF ALL ESF AC POWERINDEX NO.ROSROK/A NO.000024K302Actions contained in EOP for emergency boration4.24.4AK3.02Page 2 of 1 INITIATING CUES:CRS directs diluting the "A" BAT per EOP-6.0, Attachment 6.INITIAL CONDITION:The plant has experienced an ESF Bus Blackout with the CRS implementing EOP-6.0. Annunciator "BAT A TEMP HI/LO" has been received and local verification indicates that temperature is 68°F in 'A' BAT room.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!FOR ELECTRICAL MANIPULATIONS, AT NO TIME ARE YOU TO BREAK THE PLANE OF THE ELECTRICAL PANEL!AT NO TIME ARE YOU TO OPERATE ANY PLANT EQUIPMENT!Page 3 of 1 STEPSStep 1a. Connects the drain rig stored in the BAT room to XVD08324A-CS, BORIC ACID TANK A DRAIN ISOL VALVE, and to the 2-inch capped pipe which penetrates the floor.Evaluator cue: Provide a copy of NJPP-402 Handout (EOP-6.0 Attachment 6)Evaluator cue: Provide key to Examinee once they explain where they would obtain it. Evaluator note: Requires obtaining Key G1A from the control room key box hook 45 or key KA1 on a set of rover keys from the SS key box. The tool box has a lock on either side and can be difficult to open due to the nature of the lock mechanism. Evaluator note: Examinee must identify hoses and fittings and demonstrate where each should be installed and actual layouts. Does not require removal of components from locker. Following is a description of fittings and connection points: 1.Using a pipe wrench remove the 2" stainless steel pipe cap from the nippledownstream of XVD08324A-CS2.Thread on the Male Quick disconnect fitting labeled "Hook to Tank Drain"at the 8324A nipple. Tighten with wrench. Fitting has 2" threads and a 1.5" quickconnect for a cam lock connector.3.Using a pipe wrench remove the 2" stainless steel pipe cap from the verticalpipe stub rising from the floor next to valve 8323A.4.Thread on the Male Quick disconnect fitting labeled "Hook to Floor Drain" atthe vertical pipe stub rising from the floor next to valve 8323A. Tighten with wrench.Fitting has 2" threads and a 1.5" male quick connect for a cam lock connector.5.Connect the red rubber hose labeled "Drain Rig" to the male fittings just installed.Place one female quick connect cam lock on the tank drain and the other one on thefloor drain connection. Connections are made by placing the male fitting inside thefemale fitting with the tabs on the female fitting perpendicular to the fitting then locking iton by moving the tabs 90° to parallel to the fitting.Evaluator cue: Once demonstration by Examinee is complete state: "The drain rig is connected."YesShows location of drain rig and XVD08324A and the floor penetration. Identifies necessary tools and equipment and explains how to make the connection.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:1Page 4 of 1 Step 1b. Opens XVD08324A-CS, BORIC ACID TANK A DRAIN ISOL VALVE.Evaluator cue: State, "Handwheel rotated CCW and stem is out".Evaluator note: This step is critical because a sufficient volume of the tank contents must be removed to accomplish the desired outcome.YesSimulates opening XVD08324A, by operating handwheel counter clockwise a few turns until resistance is felt.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Step 1c. Coordinate with the Control Room and drain the BAT to 50% level.Evaluator cue: As the NROATC, report that BAT level is 49%.Evaluator note: This step is critical because a sufficient volume of the tank contents must be removed to accomplish the desired outcome.YesContacts the CR by radio or the plant page system.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:3Page 5 of 1 Step 1d. Close XVD08324A-CS, BORIC ACID TANK A DRAIN ISOL VALVE.Evaluator cue: State, "Handwheel rotated CW and stem is in".YesSimulates closing XVD08324A by operating the handwheel a few turns in the clockwise direction until it stops turning.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:4Step 1e. Remove the drain rig.Evaluator note: Following is a description of necessary steps to disconnect the rig: Disconnect the red rubber hose labeled "Drain Rig" from the male fittings at the tankdrain and the floor drain. Disconnect is accomplished by raising the tabs on the female fittings 90° to the position and then pulling off the quick connect. Evaluator cue: Once Examinee completes the description of rig removal state: "Drain rig is removed."YesDescribes how to disconnect the rig from XVD08324A and the penetration from the floor.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:5Page 6 of 1 Step 2a. Open the nearest Fire Hose Reel Isolation Valve and flush the fire hose to the floor drain until the water is clear.Evaluator cue: When valve is open and the nozzle is rotated CCW, state: "The water is clear."YesSimulates operating the reel isolation valve counter clockwise until in line with pipe. Simulates operating the nozzle counter clockwise.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:6Step 2b. Close the Fire Hose Reel Isolation Valve.Evaluator cue: State, "Handle is perpendicular to pipe."Yes operating the reel isolation valve in the clockwise direction until perpendicular with pipeCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:7Page 7 of 1 Step 2c. Connect the fill rig to XVD08324A-CS BORIC ACID TANK A DRAIN ISOL VALVE, and to the fire hose from the hose reel.Evaluator cue: State, "Fill rig is connected".Evaluator note: Following is a description of necessary steps to connect the rig: 1.Take the red 1.5" fire hose labeled "fill rig" and connect the 1.5" female fire hosecoupling labeled "Hook to Hose Reel" to the male threaded coupling on the fire hosefrom the reel station.2.Connect the female cam lock connection labeled "To Boric Acid Tank" on the oppositeend of the fill rig to the previously installed male fitting at the tank drain (valve 8324A). Connectionmade by placing the male fitting inside the female fitting with the tabs on the female fittingperpendicular to the fitting then locking it on by moving the tabs 90° to parallel to the fitting.YesExplains how to make the connection.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:8Page 8 of 1 Step 2d. Opens XVD08324A-CS, BORIC ACID TANK A DRAIN ISOL VALVE.Evaluator cue: State, "Handwheel rotated CCW and stem is out".YesSimulates opening XVD08324A, by operating handwheel counterclockwise a few turns until resistance is felt.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:9STEP:10Step 2e. Open the Fire Hose Reel Valve to fill the BAT.Evaluator cue: State, "Handle is in-line with pipe".YesOperates the hose reel isolation valve counter-clockwise until in-line with pipe.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesPage 9 of 1 Step 2f. WHEN BAT level is between 90% and 95%, THEN close the Fire Hose Reel Isolation Valve.Evaluator cue: As NROATC, report that 'A' BAT level is 92%. Evaluatorcue: State handle is perpendicular with pipe.YesOperates the reel isolation valve 90° to shut.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:11Step 2g. Close XVD08324A-CS, BORIC ACID TANK A(B) DRAIN ISOL VALVE.Evaluator cue: State handwheel rotated CW and stem is in.YesOperates XVD08324A in the clockwise direction a few turns until resistance is felt.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:12Page 10 of 1 Examiner ends JPM at this point.Step 2h. Remove the fill rig.Evaluator cue: State fill rig is removed.Evaluator note: Following is a description of fill rig removal:1.Removes cam lock female quick connect at tank drain by moving tabs 90° to to fitting and lifts off coupling.2. fire hose coupling connecting fill rig to hose reel hose.2.Removes male quick disconnects from tank drain and floor drain with wrench.3.Reinstalls pipe caps on tank drain and floor drain.YesDisconnects the rig from the fire hose and from XVD08324A. Stores the rig.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:13Page 11 of 1 JPM NO:NJPP-402DESCRIPTION:2015 NRC InPlant i RO & SRO-U: Locally Dilute the Boric Acid TanksIC SET:NAINSTRUCTIONS:COMMENTS:JPM SETUP SHEETPage 12 of 1 INITIATING CUES:CRS directs diluting the "A" BAT per EOP-6.0, Attachment 6.INITIAL CONDITION:The plant has experienced an ESF Bus Blackout with the CRS implementing EOP-6.0. Annunciator "BAT A TEMP HI/LO" has been received and local verification indicates that temperature is 68°F in 'A' BAT room.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEETAT NO TIME ARE YOU TO OPERATE ANY PLANT EQUIPMENT!

JPM NO:NJPPF-0492011 InPlant i and 2015 InPlant j NRC RO &SRO-U: Control Room Evacuation (Duties of BOP Operator)V.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Page 1 of 1 TASK:000-068-05-01PREFERRED EVALUATION METHODSIMULATEEVALUATION TIME14TIME CRITICALNoTASK STANDARD:AOP-600.1 Attachment 2 performed with the following complete: 1.All MFPs have been tripped2.Rod Drive MG set "B" feeder breaker has been tripped3.RCP "B" is left running ('A' and 'C' RCP are tripped already).

4.Two condensate pumps have been tripped5.Three FWBP's have been tripped.The use of applicable Human Performance Tools (3-way communications, self checking, peer checking, phonetic alphabet, etc) and industrial safety practices meets expectations.PREFERRED EVALUATION LOCATIONPLANTTIME START:TIME FINISH:PERFORMANCE TIME:10CFR55:45(a)13PERFORM CONTROL ROOM EVACUATIONTOOLS:NJPPF-049 Handout 1; AOP-600.1, Attachment 2NJPPF-049 Handout 2; Picture of the inside of a 7.2 KV breaker.NJPPF Handout 3; SOP-313 Section IV.C, Local Operation of a Reactor Trip Breaker NJPPF Handout 4; SOP-313 Section IV.J, Local Operation of a 7.2KV Breaker.

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TERMINATING CUE:Step 12 of Attachment 2 is complete or when examinee returns procedureto examiner.SOP-313LOCAL SWITCHGEAR BREAKER OPERATIONSISP-027ELECTRICAL SAFETYAOP-600.1CONTROL ROOM EVACUATIONINDEX NO.ROSROK/A NO.0000682130Ability to locate and operate components, including local controls.4.44.02.1.30Page 2 of 1PERFORMANCE RATING:CANDIDATE:EXAMINER:SAT:UNSAT:SIGNATUREDATE INITIATING CUES:The Control Room Supervisor directs you as the BOP Operator to perform Attachment 2 of AOP-600.1, Steps 10 through 12.INITIAL CONDITION:The plant is operating at 100% power, with all controls in automatic. A call has been received that a bomb has been placed in the control room. The SS has directed a control room evacuation. AC power is available to both ESF Buses. The reactor has been tripped by the Reactor Operator.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!FOR ELECTRICAL MANIPULATIONS, AT NO TIME ARE YOU TO BREAK THE PLANE OF THE ELECTRICAL PANEL!AT NO TIME ARE YOU TO OPERATE ANY PLANT EQUIPMENT!Page of 1 STEPSProcedure CAUTION - Step 10 "Reactor Trip should be verified with the Reactor Operator prior to securing the Main Feedwater Pumps."Verifies reactor has been tripped.Evaluator note: nitial conditions have indicated that the RO has already tripped the reactor. Evaluator cue: f the Examinee calls the Reactor Operator to verify the reactor trip respond as the Reactor Operator that the reactor has been tripped.YesCalls the Reactor Operator and verifies reactor has been tripped.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Step 10. Locally trip all Main Feedwater Pumps (436' TB).Evaluator note: ripping MFPs is critical because leaving them in service would result in excessive RCS cooldown and positive reactivity addition.YesPulls MFP "PULL TO TRIP" handle on front standard for MFP's "A" "B" & "C". Verifies trip by noting RPM decrease locally OR trips MFPs from local DCS station.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Page of 1 Step 11 a. Locally at XSW1A Switchgear Room (TB-436): Trips XSW-1B1 06C - ROD DRIVE M/G SET "B".Evaluator cue: f asked as the Shift Supervisor if ISP-027, ELECTRICAL SAFETY INDUSTRIAL SAFETY PROCEDURE, requirements can be waived respond that the requirements can be waived. This waiver will be applied to the rest of the task. If Examinee does not wish to waive ISP-027 requirements then the following are required: Hard hat; safety glasses, hearing protection; Fire Retardant Pants and shirt or Fire Retardant coveralls. Evaluator note: copy of the applicable procedure is shown in NJPPF-049 Handout 3 (SOP-313 Section IV.C). If examinee describes the correct procedure and states they would obtain a copy then provide Handout 3. Evaluator cue: nform Examinee that MG Set "B" breaker cubicle has a Green "OPEN" flag with Red light OFF and Green light ON.Evaluator cue: YesTrips rod drive MG set "B" bkr 06C at XSW-1B1 by pushing on red TRIP pushbutton on left side on front of breaker. Verifies a Green "OPEN" flag results and Red light OFF, Green light ON.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:3Page of 1 Step 11 b. Check status of XSW1A 06 FD WTR BOOSTER PUMP "A" breaker. (TB-436)Evaluator cue: nform Examinee that "A" FWBP, breaker cubicle has the Red light ON.YesVerifies that the "A" FWBP, bkr 06 is closed by observing red light on outside of cubicle door.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:4Step 11 c. Checks status of XSW1A 09, RX COOLANT PUMP "A" breaker.(TB-436)Evaluator cue: nform Examinee that RCP "A" breaker cubicle has the Green light ON. Evaluator note: his will "setup" alternate path portion of this JPM. Examinee will have to leave 'B' RCP running in Step 12.c.)YesChecks RCP "A" breaker at XSW1A 09. Verifies breaker cubicle door has Green light ON.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:5Page of 1 Step 11 d. Check status of XSW1A 07, COND PUMP "A" breaker. (TB-436)Evaluator cue: nform Examinee that "A" condensate pump breaker cubicle has the Red light ON.YesVerifies that the "A" condensate pump bkr 07 is closed by observing Red light ON outside of cubicle door.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:6Page of 1 Step 12 Locally at XSW1B and XSW1C Switchgear Room (TB-412): Step 12 a. If Condensate Pump "A" is running THEN trip both of the following: Trip XSW1B 09, COND PUMP "B" breaker.(TB-412)Evaluator cue: f asked as the Shift Supervisor if ISP-027, ELECTRICAL SAFETY INDUSTRIAL SAFETY PROCEDURE, requirements can be waived respond that the requirements can be waived.

If Examinee does NOTwish to waive ISP-027 requirements then the following are required: 25 Cal/cm2, arc flash suit and hood (use of an arc flash hood without a hard hat in an area with overhead work in progress will require manager approval. Otherwise no hard had is required when in an arc flash hood). Short sleeve natural fiber shirt, voltage rated gloves with leather, safety glasses, earmuffs are the preferred hearing protection when an arc flash suit is being worn, however earplugs may be used. Fire Retardant coveralls or Fire Retardant Shirt (tucked in) & Pants. A 10' flash protection boundary is established.Evaluator Note: o not let Examinee open the breaker door. A picture has been included of the inside of a 7.2 breaker (NJPPF-049 Handout 2). Evaluator note: copy of the applicable procedure is shown in NJPPF-049 Handout 4 (SOP-313 Section IV.J). If examinee describes the correct procedure and states they would obtain a copy then provide Handout 4. This handout is to all of the 7.2 KV breaker local operations. Evaluator cue: IF correct action is described inform the Examinee Condensate Pump "B" breaker cubicle door has the Green light ON. IF correct actions are NOT described inform the Examinee that the Red light is ON. Evaluator note: ripping Condensate Pumps is critical because leaving them in service would result in excessive RCS cooldown and positive reactivity addition.YesTrips breaker XSW1B 09 for Cond Pump "B" by pushing the "MANUAL TRIP" lever on front of breaker (inside cubicle door). Verifies breaker cubicle door has the Green light ON.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:7Page of 1 Step 12 a. Checks status of XSW1C 06, COND PUMP "C" breaker. (TB-412)Evaluator cue: Inform Examinee that COND PUMP "C" breaker cubicle has the Green light ON.YesChecks COND PUMP "C" Breaker, XSW01C 06. Verifies breaker cubicle door has the Green light ON.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:8Page of 1 Step 12 b. If Feedwater Booster Pump "A" is running, THEN trip all of the following: Trips XSW1B 06, FD WTR BOOSTER PUMP "B" breaker. (TB-412)Evaluator note: Same ISP-027, ELECTRICAL SAFETY INDUSTRIAL SAFETY PROCEDURE, requirements as for the Condensate pumps.Evaluator Note: Do not let Examinee open the breaker door. A picture has been included of the inside of a 7.2 breaker (NJPPF-049 Handout 2). Evaluator cue: IF correct action is described inform the Examinee that Feed Water Booster Pump "B" breaker cubicle door has the Green light ON. IF correct actions are NOT described inform the Examinee that the Red light is ON. Evaluator note: Tripping FW Booster Pumps is critical because leaving them in service would result in excessive RCS cooldown and positive reactivity addition.YesTrips the FWBP "B" bkr 06 manually at XSW-1B by pushing the "MANUAL TRIP" lever on front of breaker (inside cubicle door). Verifies a green light on outside of cubicle door results.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:9Page 1 of 1 Step 12 b. Trips XSW1B 13, FD WTR BOOSTER PUMP "D" breaker. (TB-412)Evaluator note: Same ISP-027, ELECTRICAL SAFETY INDUSTRIAL SAFETY PROCEDURE, requirements as for the Condensate pumps.Evaluator cue: IF correct action is described inform the Examinee that Feed Water Booster Pump "D" breaker cubicle door has the Green light ON. IF correct actions are NOT described inform the Examinee that the Red light is ONEvaluator Note: Do not let Examinee open the breaker door. A picture has been included of the inside of a 7.2 breaker (NJPPF-049 Handout 2). Evaluator note: Tripping FW Booster Pumps is critical because leaving them in service would result in excessive RCS cooldown and positive reactivity addition.YesTrips the FWBP "D" bkr 13 manually at XSW-1B by pushing the "MANUAL TRIP" lever on front of breaker (inside cubicle door). Verifies breaker door has the Green light ON.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:10Page 1 of 1 Step 12 b. Trips XSW1C 08, FD WTR BOOSTER PUMP "C" XPP0028C-FW breaker. (TB-412)Evaluator note: Same ISP-027, ELECTRICAL SAFETY INDUSTRIAL SAFETY PROCEDURE, requirements as for the Condensate pumps.Evaluator Note: Do not let Examinee open the breaker door. A picture has been included of the inside of a 7.2 breaker (NJPPF-049 Handout 2). Evaluator cue: IF correct action is described inform the Examinee that Feed Water Booster Pump "C" breaker cubicle door has the Green light ON. IF correct actions are NOT described inform the Examinee that the Red light is ONEvaluator note: Tripping FW Booster Pumps is critical because leaving them in service would result in excessive RCS cooldown and positive reactivity addition.YesTrips the FWBP "C" bkr 08 manually at XSW-1C by pushing the "MANUAL TRIP" lever on front of breaker (inside cubicle door). Verifies breaker cubicle door has the Green light ON.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:11Page 1 of 1 Step 12 c. If RCP A is running (NO), goes to RNO. Ensure one of the following is open: XSW1B 07, RX COOLANT PUMP "B" breaker OR XSW1C 03, RX COOLANT PUMP "C" breaker.Evaluator cue: Inform Examinee, when looking at breaker positions, that XSW1B 07, RX COOLANT PUMP B breaker cubicle has the Red light ON and that XSW1C 03, RX COOLANT PUMP C breaker cubicle has the Green light ON.Evaluator note: This is the alternative path portion of this JPM. It is critical that the "B" RCP be left running since both the "A" and "C" pumps are already tripped in this JPM.Evaluator note: Same ISP-027, ELECTRICAL SAFETY INDUSTRIAL SAFETY PROCEDURE, requirements as the Condensate pumps.YesDoes NOT trip the RCP "B" bkr 07 at XSW-1B (because RCP "C" bkr 03 at XSW-1C is already open).CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:STEP:12Page 1 of 1 Examiner ends JPM at this point.Step 12 d. Ensure XSW 1C 02 Press Heater Transformer breaker is closed.Evaluator cue: Inform Examinee that PZR Heater Transformer breaker cubicle door has the Red light ON, Green light OFF.Evaluator note: Same ISP-027 considerations as Condensate pumps if it was to be operated but since only verifying proper position there are no ISP-027 requirements.YesVerifies that the PZR Heater Transformer Breaker 02 at XSW-1C is closed by observing red light on outside of cubicle door or a red "closed" flag on front of breaker.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:13Page 1 of 1 JPM NO:NJPPF-049DESCRIPTION:2011 InPlant i and 2015 InPlant j NRC RO &SRO-U: Control Room Evacuation (Duties of BOP Operator)IC SET:INSTRUCTIONS:COMMENTS:JPM SETUP SHEETday, J 1, 201Page 1 of 1 INITIATING CUES:The Control Room Supervisor directs you as the BOP Operator to perform Attachment 2 of AOP-600.1, Steps 10 through 12.INITIAL CONDITION:The plant is operating at 100% power, with all controls in automatic. A call has been received that a bomb has been placed in the control room. The SS has directed a control room evacuation. AC power is available to both ESF Buses. The reactor has been tripped by the Reactor Operator.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEETAT NO TIME ARE YOU TO OPERATE ANY PLANT EQUIPMENT!

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NJPPF- 049 Handout 2 SOP-313 REVISION 5 PAGE 34 OF 53 C. LOCAL OPERATION OF A REACTOR TRIP BREAKER 1.0 INITIAL CONDITIONS 1.1 The Control Room has been informed of the local operations to be performed. NOTE 2.0 These steps shall only be performed when directed by the Shift Supervisor. 2.0 INSTRUCTIONS 2.1 Verify proper Personal Protective Equipment and Approach Boundaries per ISP-027, Enclosures 8.4 and 8.2. 2.2 To locally close a Reactor Trip Breaker, depress the PUSH TO CLOSE button. 2.3 To locally open a Reactor Trip Breaker, depress the red TRIP button on the front of the breaker face. END OF SECTION CHG C SOP-313 REVISION 5 PAGE 45 OF 53 J. LOCAL OPERATION OF A 7.2 KV BREAKER 1.0 INITIAL CONDITIONS 1.1 The Control Room has been informed of all local operations to be performed. 1.2 One of the following conditions have been met: a.The SS/CRS has entered the cubicle number and nomenclature for thebreaker to be operated in Step 2.1.a, below.b.Personnel racking the breaker have a controlling document in the field(i.e. LOTO or SOP lineup) that identifies the correct component bylisting both the breaker cubicle number and nomenclature.NOTE 2.0 These steps shall only be performed when directed by the Shift Supervisor. 2.0 INSTRUCTIONS 2.1 Ensure you are at the correct breaker by one of the following: a.Per SS/CRS:b.Per controlling document in accordance with Step 1.2.b. 2.2 Verify proper Personal Protective Equipment and Approach Boundaries per ISP-027, Enclosures 8.4 and 8.2. 2.3 Inform the Control Room of the component affected, by the breaker to be operated. N01 CHG A CHG A CHG C SOP-313 REVISION 5 PAGE 46 OF 53 2.4 Locally close the breaker by performing one of the following: a. For breakers that allow manual operation from inside the cubicle: 1) Depress the MANUAL CLOSE button on the lower, center portion of the breaker face. 2) Verify the breaker is closed, as indicated by the CLOSED flag being visible.

b. For breakers that are cannot be operated inside the cubicle and local use of the pistol grip switch is desired: 1) Place the pistol grip handle to the CLOSED position with a crisp hand motion. 2) Verify the breaker is closed, as indicated by the CLOSED flag being visible. CAUTION 2.5 When possible, the breaker should be opened locally under minimal load. 2.5 Locally open the breaker by performing one of the following:

a. For breakers that allow manual operation from inside the cubicle: 1) Depress the MANUAL TRIP lever through the opening on the lower, right hand corner of the breaker face. 2) Verify the breaker is open, as indicated by the OPEN flag being visible.

b. For breakers that are cannot be operated inside the cubicle and local use of the pistol grip switch is desired: 1) Place the pistol grip handle to the TRIP position with a crisp hand motion. 2) Verify the breaker is open, as indicated by the CLOSED flag being visible.

END OF SECTION CHG D CHG D JPM NO:NJPP-0402015 NRC In-Plant k RO: Transfer a Vital 120 Volt Instrument Power SupplyV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Tuesday, January 27, 2015Page 1 of TASK:062-010-01-04PREFERRED EVALUATION METHODSIMULATEEVALUATION TIME10TIME CRITICALNoTASK STANDARD:XIT-5901 is shutdown with APN-5901 supplied by alternate power from 1FA via manual bypass switch. The use of applicable Human Performance Tools (3-way communications, self checking, peer checking, phonetic alphabet, etc) and industrial safety practices meets expectations.PREFERRED EVALUATION LOCATIONPLANTTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)6REMOVE ENGINEERING SAFETY FEATURES VITAL INVERTER FROM SERVICETOOLS:NJPP-040 Handout; SOP-310 Section IV.E Placing Inverter XIT5901 in an Alternate AC Lineup, marked through step 1.4 with steps 2.10 and 2.11 marked N/A.

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TERMINATING CUE:Examinee returns SOP-310 to examiner.CANDIDATE:SOP-310ENGINEERED SAFETY FEATURES 120 VAC INSTRUMENT AND CONTROL SYSTEMINDEX NO.ROSROK/A NO.062000A203Consequences of improper sequencing when transferring to or from an inverter2.93.4A2.03Tuesday, January 27, 2015Page 2 of INITIATING CUES:Shift Supervisor directs that XIT-5901 be removed from service and APN-5901 be placed on alternate power in accordance with SOP-310,Section IV.E for inverter PMs. Initial conditions are completed through step 1.4.STP-506.005, DEENERGIZATION OF THE ESF LOAD SEQUENCER LOSS OF VOLTAGE AND DEGRADED VOLTAGE RELAYS has been completed.INITIAL CONDITION:Plant is at 100% power. 'A1' Train maintenance work. XIT-5901 is scheduled for preventive maintenance.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!FOR ELECTRICAL MANIPULATIONS, AT NO TIME ARE YOU TO BREAK THE PLANE OF THE ELECTRICAL PANEL!AT NO TIME ARE YOU TO OPERATE ANY PLANT EQUIPMENT!Tuesday, January 27, 2015Page 3 of STEPSStep 1.5 On XIT5901 the following breakers are closed:a.ALT. AC SOURCE.b.BACKUP SOURCE.c.NORMAL AC SOURCE.Evaluator cue: Provide a copy of NJPP-040 Handout, SOP-310 Section IV.EEvaluator note: All 3 breakers should be closed during normal full power operation.YesAlternate source, Backup Source and Normal AC source breakers are all closed.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Step 1.6 The MAN BYPASS Switch is in NORMAL.Evaluator note: Manual Bypass should be in NORMAL during normal full power operation.YesManual Bypass switch is in NORMALCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:2Tuesday, January 27, 2015Page 4 of 1.7 The TEST TRANSFER Switch is in the CENTER Position.Evaluator note: Test Transfer switch should be in the CENTER position during normal full power operation.YesThe TEST TRANSFER Switch is in the CENTER Position.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:31.8 The SYNC MONITOR light is NOT lit.YesVisually observes SYNC MONITOR Light OFF. (CB-436). (center of panel)CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:4Tuesday, January 27, 2015Page 5 of Procedure CAUTION 2.0STP-506.005, De-Energization Of The ESF Load Sequencer Loss Of Voltage AndDegraded Voltage Relays A Train, XEG0001A, should be completed on Train A priorto removing Inverter XIT5901 from service. This should be tracked on the DieselGenerator A R&R. Procedure NOTE 2.0a.All steps are performed in the Relay Room (CB-436) unless otherwise stated.b.XCP-636 1-5 (INV 1/2 TROUBLE) and XCP-636 1-6 (INV 1/2 AC INPUT LOSS)will be locked in alarm at the completion of this procedure.Step 2.1 Ensure Alternate Source voltage is acceptable for transfer:a.Place the SOURCE SELECTOR Switch in LINE.b.Verify the Alternate AC Source voltage is acceptable as indicatedby an AC OUTPUT voltage indication between 115 VAC and 125 VAC.c.Place the SOURCE SELECTOR Switch in OUTPUT.Evaluator cue: When Examinee selects line point to AC Output voltage meter indicating 120 Volts.YesSimulates placing Source Selector to line, verifies AC Source Voltage and returns Source Selector switch to Output.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:5Tuesday, January 27, 2015Page 6 of Procedure NOTE 2.2Placing the TEST TRANSFER Switch in the ALT Position forces the Static Switch toreceive power from the Alternate Source and deliver it to APN5901. This will alsocause the following:a)XCP-636 1-5 (INV 1/2 TROUBLE) annunciates (MCB).b)ON ALTERNATE light illuminates.c)ON INVERTER light is extinguishes.Step 2.2; Place the TEST TRANSFER Switch to the ALT Position.Evaluator cue: Inform Examinee that the TEST TRANSFER Switch is in ALT position and the "ON ALTERNATE" light illuminates and the "ON INVERTER" light is extinguishes.Evaluator cue: If Examinee contacts the control room respond that annunciator XCP-636 1-5 (INV 1/2 TROUBLE) did annunciate at the Main Control Board.Evaluator note: Step 6 is critical is critical as the alternate source must be placed in service to accomplish the task standard. YesCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:6Tuesday, January 27, 2015Page 7 of Step 2.3; Verify the ON ALTERNATE light is lit.Evaluator cue: Inform Examinee that the "ON ALTERNATE" Light is ON.YesVerifies the ON ALTERNATE light illuminated.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:7Tuesday, January 27, 2015Page 8 of Procedure CAUTION Step 2.4If the ON ALTERNATE light is not illuminated, then it cannot be verified that theStatic Switch has aligned itself to receive power from the Alternate Source.Proceeding with this procedure may result in a loss of APN5901. Procedure NOTE 2.4 and 2.5: The SYNC MONITOR Light should illuminate when the INVERTER STOP Pushbutton is pressed.Step 2.4; Momentarily depress the INVERTER STOP Pushbutton and verify the SYNC MONITOR Light is lit.Evaluator cue: Inform Examinee that the SYNC MONITOR Light is ON after the INVERTER STOP Pushbutton is depressed.Evaluator note: Examinee should proceed to step 2.6.Evaluator note: Step 8 is critical as the inverter is to be removed from service and this step accomplishes that.YesCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:8Tuesday, January 27, 2015Page 9 of Procedure NOTE 2.6The transfer performed with the MAN BYPASS switch is a Make-Before-Breakoperation which provides a momentary paralleling of power sources.Step 2.6; Align Alternate AC power through the Static Switch to APN5901 by rotating the MAN BYPASS switch clockwise to the BYP TO ALT Position.Evaluator note: Step 9 is critical to align the alternate source to 1FA.YesCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:9Step 2.7; Open the BACKUP SOURCE Breaker on the Inverter front.Evaluator note: Step 10 is critical to isolate inverter power per task standardYesCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:10Tuesday, January 27, 2015Page 10 of Examiner ends JPM at this point.Step 2.8; Open the NORMAL AC SOURCE Breaker on the Inverter front.Evaluator note: Step 11 is critical to isolate inverter power per task standardYesCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:11Step 2.9: Verify XCP-636 1-6 (INV 1/2 AC INPUT LOSS) is in alarm (MCB).Evaluator cue: Respond as control room that XCP-636 1-6 did alarm.NoCalls control room to verify XCP-636 1-6 is in alarm.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:12Tuesday, January 27, 2015Page 11 of JPM NO:NJPP-040DESCRIPTION:2015 NRC In-Plant k RO: Transfer a Vital 120 Volt Instrument Power SupplyIC SET:NAINSTRUCTIONS:COMMENTS:JPM SETUP SHEETTuesday, January 27, 2015Page 12 of INITIATING CUES:Shift Supervisor directs that XIT-5901 be removed from service and APN-5901 be placed on alternate power in accordance with SOP-310,Section IV.E for inverter PMs. Initial conditions are completed through step 1.4.STP-506.005, DEENERGIZATION OF THE ESF LOAD SEQUENCER LOSS OF VOLTAGE AND DEGRADED VOLTAGE RELAYS has been completed.INITIAL CONDITION:Plant is at 100% power. 'A1' Train maintenance work. XIT-5901 is scheduled for preventive maintenance.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEETAT NO TIME ARE YOU TO OPERATE ANY PLANT EQUIPMENT!

E. PLACING INVERTER XIT5901 IN AN ALTERNATE AC LINEUPPre-Job Brief

NOTEND OF SECTION

JPM NO:NJPA-021A2015 NRC RO/SRO Common A1-a: Perform Boric Acid Dilution Volume DeterminationV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Tuesday, September 30, 2014Page 1 of TASK:004-006-01-01PREFERRED EVALUATION METHODPERFORMEVALUATION TIME10TIME CRITICALNOTASK STANDARD:Examinee determines that about 9359 gals (interpolated value) will be required for dilution. A range of 9250 to 9450 (~ +/- 100 gal or ~ 1%) would be acceptable. If the examinee elects to calculate vice interpolate, the calculated value is the same range.PREFERRED EVALUATION LOCATIONCLASSROOMTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)(12)TOOLS:Curve Book (Accessible via Desk Top Computers)CalculatorNJPA-021A Dilution Calculation hand out.

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TERMINATING CUE:CANDIDATE:SOP-106REACTOR MAKEUP WATER SYSTEMINDEX NO.ROSROK/A NO.1940012125Ability to interpret reference materials such as graphs, curves, tables, etc.3.94.22.1.251940012137Knowledge of procedures, guidelines, or limitations associated with reactivity management.4.34.62.1.37Tuesday, September 30, 2014Page 2 of INITIATING CUES:The SS directs you to determine the dilution volume to establish the estimated critical boron concentration of 1688 ppm. The latest RCS sample indicated boron concentration was 2038 ppm. Show all work on the NJPA-021A Dilution Calculation hand out provided.INITIAL CONDITION:The plant is in MODE 3 preparing for reactor startup in accordance with GOP-3.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Tuesday, September 30, 2014Page 3 of STEPSExaminee determines that the amount of dilution required.Evaluator note: Ci and Cf values are given in the initiating cue.YesDetermines Ci - Cf = 350 ppm dilution required.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:1Selects correct curve from curve book.Evaluator note: Do NOTprompt Examinee on location of the dilution tables.YesThe examinee refers to Figure III-3, RCS Dilution Gallons (Vw) of Dilution Water Required, in the Curve Book.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Tuesday, September 30, 2014Page 4 of Examiner ends JPM at this point.Examinee interpolates or calculates the volume of water required and reports the dilution volume required to the CRS.Evaluator note: The volume reported to the CRS should be the interpolated value or greater. A range of 9250 to 9450 (~+/- 100 gal) would be acceptable. If the examinee elects to calculate vice interpolate the same range applies. See JPA-021A Key for calculations.YesExaminee determines that 9359 gals (interpolated value) will be required for dilution.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:3Tuesday, September 30, 2014Page 5 of JPM NO:NJPA-021ADESCRIPTION:2015 NRC RO/SRO Common A1-a: Perform Boric Acid Dilution Volume DeterminationIC SET:NAINSTRUCTIONS:COMMENTS:JPM SETUP SHEETTuesday, September 30, 2014Page 6 of INITIATING CUES:The SS directs you to determine the dilution volume to establish the estimated critical boron concentration of 1688 ppm. The latest RCS sample indicated boron concentration was 2038 ppm. Show all work on the NJPA-021A Dilution Calculation hand out provided.INITIAL CONDITION:The plant is in MODE 3 preparing for reactor startup in accordance with GOP-3.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEET JPM NO:NJPA-10002015 NRC RO/SRO Common A1-b: Calculate Work hour limitations.V.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:Tuesday, September , 2014Page 1 of TASK:341-038-03-02PREFERRED EVALUATION METHODPERFORMEVALUATION TIME0TIME CRITICALNOTASK STANDARD:Determines that work hours would be exceeded by RO-1, due to not having a 34 hour3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> break in th9 days from 1/29 through 2/6 and by RO-2 due to not having an average of 2.5 days offin the 5 week fixed cycle.PREFERRED EVALUATION LOCATIONCLASSROOMTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:41(b) (10)INTERPRET AND ENSURE COMPLIANCE WITH PLANT ADMINISTRATIVE PROCEDURES DURING NORMAL AND OFF NORMAL PLANT OPERATIONSTOOLS:NJPA-1000 Handout.

NJPA-1000 Schedules.

SAP-152 (Available on Desk Top Computer)

Calculator

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TERMINATING CUE:Reviews work history and determines that RO- is eligible while RO-1 anRO- are NOT eligible to work the requested overtime shift without a waiver.CANDIDATE:SAP-152FATIGUE MANAGEMENT AND WORK HOUR LIMITSINDEX NO.ROSROK/A NO.1940012105Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.2.93.92.1.5Tue, September , 2014Page 2 of INITIATING CUES:Given the work schedules provided (NJPA-1000 Schedules) determine if any of three available ROs (RO-1, RO-2 or RO-3) are eligible to work the day shift beginning at 0700 on February 6 without reliance on a waiver or change to the scheduled OT. NJPA-1000 Schedules includes all hours actually worked by all three ROs after December 31 AND all hours projected to be worked through February 28. If any of the ROs is NOT eligible, identify the criteria that supports your determination.

Use the VCS fixed shift cycle method which begins on the first night of the 3 night portion of the schedule. Assume RO-1, RO-2 and RO-3 are fully qualified. Show all work on NJPA-1000 Handout provided.INITIAL CONDITION:The Unit is at 100% power currently and has been at power for the last 3 months.The regular on-line 5 shift rotation is in effect.The Control Room has been informed on February 5 night shift that an RO scheduled for the February 6 day shift cannot work due to illness.You are asked to identify if any of three available individuals are eligible to work as an RO for Thursday February 6 day shift.READ TO OPERATOR:SAFETY CONSIDERATIONS:None.WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Tuesday, September , 2014Page 3 of STEPSReviews RO-1 work history to determine if RO-1 is eligible to work the day shift on February 6.Evaluator cue: Provide a copy of NJPA-1000 Handout and NJPA-1000 Schedules to the examinee. Instruct examinee to put their name and the date on NJPA-1000 Handout and to return it to you when they are finished. Instruct Examinee to show all work on the Handout. Evaluator note: SAP-152, Fatigue Management and Work Hour Limits should be referenced while reviewing NJPA-1000 Schedules.

Evaluator note: The cycle for RO-1 runs from Jan 7 through Feb 10 (fixed five weeks).Evaluator note: Refer to NJPA-1000 Key.YesCompares work history for RO-1 to SAP-152 criteria and notes that RO-1 is not eligible to work without reliance on a waiver.Identifies that RO-1 would not have had the required 34 hour3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> break within the previous 9 days. He would have had only a single 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> break (0700 on 2/4 to 0700 on 2/5) in the 9 days from 1/29 through 2/6.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:1Tuesday, September , 2014Page of Reviews RO-2 work history to determine if RO-2 is eligible to work the day shift on February 6.Evaluator note: The cycle for RO-2 runs from Jan 14 through Feb 17 (fixed five weeks).Evaluator note: Refer to NJPA-1000 Key.YesCompares work history for RO-2 to SAP-152 criteria and notes that RO-2 is NOT eligible to work without reliance on a waiver.Identifies that RO- would not have had an average 2.5 days off per week averaged over the shift cycle. If RO-2 works 2/6 he would have had 12 days off in the 5 week cycle, an average of 2.4 days per weekCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Tuesday, September , 2014Page of Examiner ends JPM at this point.Reviews RO-3 work history to determine if RO-3 is eligible to work the day shift on February 6.Evaluator cue: Inform the examinee that the JPM has ended when they return NJPA-1000 Handout to you.Evaluator note: Refer to NJPA-1000 Key.YesCompares work history for RO- to SAP-152 criteria and notes that RO- is eligible to work without reliance on a waiver.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:3Tuesday, September , 2014Page of JPM NO:NJPA-1000DESCRIPTION:2015 NRC RO/SRO Common A1-b: Calculate Work hour limitations.IC SET:NAINSTRUCTIONS:COMMENTS:JPM SETUP SHEETTuesday, September , 2014Page of INITIATING CUES:Given the work schedules provided (NJPA-1000 Schedules) determine if any of three available ROs (RO-1, RO-2 or RO-3) are eligible to work the day shift beginning at 0700 on February 6 without reliance on a waiver or change to the scheduled OT. NJPA-1000 Schedules includes all hours actually worked by all three ROs after December 31 AND all hours projected to be worked through February 28. If any of the ROs is NOT eligible, identify the criteria that supports your determination.Use the VCS fixed shift cycle method which begins on the first night of the 3 night portion of the schedule. Assume RO-1, RO-2 and RO-3 are fully qualified. Show all work on NJPA-1000 Handout provided.INITIAL CONDITION:The Unit is at 100% power currently and has been at power for the last 3 months.The regular on-line 5 shift rotation is in effect.The Control Room has been informed on February 5 night shift that an RO scheduled for the February 6 day shift cannot work due to illness.You are asked to identify if any of three available individuals are eligible to work as an RO for Thursday February 6 day shift.SAFETY CONSIDERATIONS:None.OPERATOR INSTRUCTIONS:JPM BRIEFING SHEETHAND THIS PAPER BACKTO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

SSMTWThFSFeb14SMTWThFSSSMTWThFSMTWThFShift110111213141516171819202122232425262728RO1NNNOTDOTDNNNTTTOTDNNNNOTNOTDDDDDOTNNNNDDDDRO2DDDRO3NNNOTDNNNOTNDDDDTTTTTWThFSSSMTWThFJan14SMTWThFSSMTWThFSSMTWThFShift110111213141516171819202122232425262728293031TTTDDDOTNOTNNNNNOTNNNNNOTNDDDDTTOTDTTRO1OTNRO2RO3OTNNNNDDDDOTNNNNOTNOTDDDDDTTTTTDDDOTDOTNN

ProposedovertimeshiftTTTT NJPA-1000 JPM A1-b RO and SRO-U - 2015 Page1of2Examinee Name: ____________________________ Date: _________________Note: 1.NJPA-1000 Schedules includes all hours actually worked by RO-1, RO-2 and RO-3 after December 31all hours projected to be worked through Feb 2.2.Day shift is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; 0700 to 1900 and Night shift is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1900 to 0700 and Training is 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 0730 to 1730.1.Place a check mark in the appropriate box for RO-1, RO-2 and RO-3.2.Provide basis if a worker is NOT eligibleRO-1 is eligible to work 0700-1900 on 2/6.RO-1 is NOT eligible to work 0700-1900 on 2/6.IF RO-1 is NOT eligible to work the overtime STATE the basis:RO-2 is eligible to work 0700-1900 on 2/6.RO-2 is NOT eligible to work 0700-1900 on 2/6.IF RO-2 is NOT eligible to work the overtime STATE the basis: RO-3 is eligible to work 0700-1900 on 2/6.RO-3 is NOT eligible to work 0700-1900 on 2/6.IF RO-3 is NOT eligible to work the overtime STATE the basis:

NJPA-1000 JPM A1-b RO and SRO-U - 2015 Page2of2Additional work:

NJPA-1000 JPM A1-b RO and SRO-U - 2015 Page1of2Examinee Name: ____________________________ Date: _________________Note: 1.NJPA-1000 Schedules ncludes all hours actually worked by RO-1, RO-2 and RO-3 after December 31all hours projected to be worked through Feb 2.2.Day shift is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; 0700 to 1900 and Night shift is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1900 to 0700 and Training is 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 0730 to 1730.1.Place a check mark in the appropriate box for RO-1, RO-2 and RO-3.2.Provide basis if a worker is NOT eligibleRO-1 is eligible to work 0700-1900 on 2/6.RO-1 is NOT eligible to work 0700-1900 on 2/6.IF RO-1 is NOT eligible to work the overtime STATE the basis:RO-2 is eligible to work 0700-1900 on 2/6.RO-2 is NOT eligible to work 0700-1900 on 2/6.IF RO-2 is NOT eligible to work the overtime STATE the basis: RO-3 is eligible to work 0700-1900 on 2/6.RO-3 is NOT eligible to work 0700-1900 on 2/6.IF RO-3 is NOT eligible to work the overtime STATE the basis:

2015 NRC RO A2: Determine Surveillance Requirements due to loss of Main Control Board AnnunciatorsV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:JPM NO:NJPA-1006THIS JPM IS NOT APPROVEDWednesday, September 10, 2014Page 1 of TASK:000-170-05-01PREFERRED EVALUATION METHODPERFORMEVALUATION TIME5TIME CRITICALNOTASK STANDARD:Identifies the following surveillance requirements from Attachment 3:XCP-615, 2-5; GTP-702 Att IV.GXCP-615, 3-3; GTP-702 Att VI.V-3 XCP-615, 3-6; OAP-106.1, RB Sump Level and STP-114.002 XCP-620, 1-5; GTP-702, Att IV.EXCP-620, 1-6; GTP-702, Att IV.EXCP-620, 2-4; GTP-702, Att IV.D XCP-620, 2-5; GTP-702, Att IV.B XCP-620, 4-2; GTP-702, Att VI.L-2XCP-620, 4-3; GTP-702, Att VI.L-2XCP-621, 1-1; GTP-702, Att IV.CPREFERRED EVALUATION LOCATIONCLASSROOMTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)(3)Respond to loss of Main Control Board annunciators per AOP-100.5.TOOLS:NJPA-1006 Handout 1; AOP-100.5, Att 2 NJPA-1006 Handout 2; AOP-100.5, Att 3.

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TERMINATING CUE:Identifies required Surveillance Attachments associated with XCP-615, XCP-620 and XCP-621 from AOP 100.5 Attachment 3.CANDIDATE:-1Loss of Main Control Board AnnunciatorsINDEX NO.ROSROK/A NO.1940012214Knowledge of the process for controlling equipment configuration or status.3.94.32.2.14Wednesday, September 10, 2014Page 2 of INITIATING CUES:I&C has verified that DPN 1HX1 01 has tripped. The CRS has directed you to any surveillance requirements associated with Main Control Board Annunciators that have lost power. Indicate any required surveillances by highlighting the NJPA-1006 Handout 2 to show the applicable surveillances and attachments.INITIAL CONDITION:The plant is at 100% power. A loss of multiple Main Control Board annunciators has been experienced due to a loss of power.READ TO OPERATOR:SAFETY CONSIDERATIONS:NAWHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Wednesday, September 10, 2014Page of STEPSRefers to AOP-100.5 Attachment 2 to determine which Annunciator panels are associatedwith DPN 1HX1 01Evaluator cue: Hand examinee a copy of NJPA-1006 Handout 1 and Handout 2 (AOP-100.5, Att 2 and Att 3). Tell examinee to put their name on Handout 2 and mark which surveillances will be required.Evaluator note: Examinee will identify applicable panels using AOP-100.5 Attachment 2 and the initiating cue of "panel DPN 1HX1 circuit 01 has tripped".YesIdentifies that the following Annunciator panels are affected:XCP-610, XCP-610, 611, 612, 613, 614, 615, 616, 617, 618, 619, 620, 621, 624 and 626CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:1Wednesday, September 10, 2014Page of Examiner ends JPM at this point.REFER TO ATTACHMENT 3, SURVEILLANCE ANNUNCIATORS, for annunciators that have surveillance requirements.Evaluator note: refer to NJPA-1006 Key for expected response.YesIdentifies the following surveillance requirements from Attachment 3:XCP-615, 2-5; GTP-702 Att IV.GXCP-615, 3-3; GTP-702 Att VI.V-3XCP-615, 3-6; OAP-106.1, RB Sump Level and STP-114.002XCP-620, 1-5; GTP-702, Att IV.E XCP-620, 1-6; GTP-702, Att IV.E XCP-620, 2-4; GTP-702, Att IV.DXCP-620, 2-5; GTP-702, Att IV.BXCP-620, 4-2; GTP-702, Att VI.L-2 XCP-620, 4-3; GTP-702, Att VI.L-2 XCP-621, 1-1; GTP-702, Att IV.CCUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Wednesday, September 10, 2014Page of JPM NO:NJPA-1006DESCRIPTION:2015 NRC RO A2: Determine Surveillance Requirements due to loss of Main Control Board AnnunciatorsIC SET:NAINSTRUCTIONS:COMMENTS:JPM SETUP SHEETWednesday, September 10, 2014Page of INITIATING CUES:I&C has verified that DPN 1HX1 01 has tripped. The CRS has directed you to any surveillance requirements associated with Main Control Board Annunciators that have lost power. Indicate any required surveillances by highlighting the NJPA-1006 Handout 2 to show the applicable surveillances and attachments.INITIAL CONDITION:The plant is at 100% power. A loss of multiple Main Control Board annunciators has been experienced due to a loss of power.SAFETY CONSIDERATIONS:NAOPERATOR INSTRUCTIONS:HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.JPM BRIEFING SHEET

(#178994:(#178994:)-3+"+#,!9)-3+"+#,!9(%%('&0-,%!P(%%('&0-,%!P1(/-!8!?F!81(/-!8!?F!8)-"%#)(%+#,!#;!;(+*-6!0(+,!'#,%)#*!5#()6)-"%#)(%+#,!#;!;(+*-6!0(+,!'#,%)#*!5#()6(,,$,'+(%#)"(,,$,'+(%#)"A. Make a list of the annunciator panels lost.A. Make a list of the annunciator panels lost.B. Contact I&C to verify the source of power to the annunciatorB. Contact I&C to verify the source of power to the annunciator panel per the following table. panel per the following table.DPN 1HA2 04DPN 1HA2 04XPN6091XPN6091XCP-601, 604, 606, 608, 622, 629, 636XCP-601, 604, 606, 608, 622, 629, 63613.7% MCB ANNUNCIATORS13.7% MCB ANNUNCIATORS13.7% MCB ANNUNCIATORS TOTAL13.7% MCB ANNUNCIATORS TOTALDPN 1HX1 02DPN 1HX1 02XPN6092XPN6092XCP-603, 628, 630, 631, 632, 633XCP-603, 628, 630, 631, 632, 63320.3% MCB ANNUNCIATORS20.3% MCB ANNUNCIATORSXPN6093XPN6093XCP-625, 627, 634, 635, 638XCP-625, 627, 634, 635, 63818.3% MCB ANNUNCIATORS18.3% MCB ANNUNCIATORS38.6% MCB ANNUNCIATORS TOTAL38.6% MCB ANNUNCIATORS TOTALDPN 1HB 02DPN 1HB 02XPN6094XPN6094XCP-602, 605, 607, 609, 623, 637XCP-602, 605, 607, 609, 623, 63712.4% MCB ANNUNCIATORS12.4% MCB ANNUNCIATORS12.4% MCB ANNUNCIATORS TOTAL12.4% MCB ANNUNCIATORS TOTALDPN 1HX1 01DPN 1HX1 01XPN6095XPN6095XCP-610, 611, 612, 613, 614, 615, 616XCP-610, 611, 612, 613, 614, 615, 61618.3% MCB ANNUNCIATORS18.3% MCB ANNUNCIATORSXPN6096XPN6096XCP-617, 618, 619, 620, 621, 624, 626XCP-617, 618, 619, 620, 621, 624, 62617.0% MCB ANNUNCIATORS17.0% MCB ANNUNCIATORS35.3% MCB ANNUNCIATORS TOTAL35.3% MCB ANNUNCIATORS TOTALC. Determine if the actions taken in response to a failed annunciatorC. Determine if the actions taken in response to a failed annunciator are subject to OAP-113.1, Operator Workaround and Dark Board are subject to OAP-113.1, Operator Workaround and Dark Board Program. Program.1(/-!R!#;!T1(/-!R!#;!T

(#178994:(#178994:)-3+"+#,!9)-3+"+#,!9(%%('&0-,%!>(%%('&0-,%!>1(/-!8!?F!81(/-!8!?F!8"$)3-+**(,'-!(,,$,'+(%#)""$)3-+**(,'-!(,,$,'+(%#)"NOTENOTEThis matrix is to aid in identifying failed annunciators which have surveillance requirements.This matrix is to aid in identifying failed annunciators which have surveillance requirements.The applicable ARP shoud be utilized when performing the surveillance.The applicable ARP shoud be utilized when performing the surveillance.PANELPANELPANELWINDOWWINDOWWINDOWSURVEILLANCESURVEILLANCESURVEILLANCEATTACHMENTATTACHMENTATTACHMENTXCP-615XCP-6152-52-5GTP-702GTP-702IV.GIV.GXCP-615XCP-6153-33-3GTP-702GTP-702VI.V-3VI.V-3XCP-615XCP-6153-63-6OAP-106.1OAP-106.1RB SUMP LEVELRB SUMP LEVELSTP-114.002STP-114.002N/AN/AXCP-620XCP-6201-51-5GTP-702GTP-702IV.EIV.EXCP-620XCP-6201-61-6GTP-702GTP-702IV.EIV.EXCP-620XCP-6202-42-4GTP-702GTP-702IV.DIV.DXCP-620XCP-6202-52-5GTP-702GTP-702IV.BIV.BXCP-620XCP-6204-24-2GTP-702GTP-702VI.L-2VI.L-2XCP-620XCP-6204-34-3GTP-702GTP-702VI.L-2VI.L-2XCP-621XCP-6211-11-1GTP-702GTP-702IV.CIV.CXCP-632XCP-6326-56-5GTP-702GTP-702IV.B, IV.D, IV.E,IV.B, IV.D, IV.E,VI.KK, VI.NNVI.KK, VI.NNOAP-106.1OAP-106.1RB TEMPSRB TEMPSRB SUMP LEVELRB SUMP LEVELMW/KV/MVARSMW/KV/MVARSGENERIC LOG SR NIGENERIC LOG SR NIOAP-100.6OAP-100.6OPERATION ATOPERATION ATLICENSED LIMITLICENSED LIMITOAP-107.1OAP-107.1RESTORATION OFRESTORATION OFIPCS FUNCTIONSIPCS FUNCTIONSXCP-638XCP-6381-41-4OAP-106.1OAP-106.1MW/KV/MVARSMW/KV/MVARSXCP-638XCP-6382-42-4OAP-106.1OAP-106.1MW/KV/MVARSMW/KV/MVARS1(/-!T!#;!T1(/-!T!#;!T

(#178994:(#178994:)-3+"+#,!9)-3+"+#,!9(%%('&0-,%!>(%%('&0-,%!>1(/-!8!?F!81(/-!8!?F!8"$)3-+**(,'-!(,,$,'+(%#)""$)3-+**(,'-!(,,$,'+(%#)"NOTENOTEThis matrix is to aid in identifying failed annunciators which have surveillance requirements.This matrix is to aid in identifying failed annunciators which have surveillance requirements.The applicable ARP shoud be utilized when performing the surveillance.The applicable ARP shoud be utilized when performing the surveillance.PANELPANELPANELWINDOWWINDOWWINDOWSURVEILLANCESURVEILLANCESURVEILLANCEATTACHMENTATTACHMENTATTACHMENTXCP-615XCP-6152-52-5GTP-702GTP-702IV.GIV.GXCP-615XCP-6153-33-3GTP-702GTP-702VI.V-3VI.V-3XCP-615XCP-6153-63-6OAP-106.1OAP-106.1RB SUMP LEVELRB SUMP LEVELSTP-114.002STP-114.002N/AN/AXCP-620XCP-6201-51-5GTP-702GTP-702IV.EIV.EXCP-620XCP-6201-61-6GTP-702GTP-702IV.EIV.EXCP-620XCP-6202-42-4GTP-702GTP-702IV.DIV.DXCP-620XCP-6202-52-5GTP-702GTP-702IV.BIV.BXCP-620XCP-6204-24-2GTP-702GTP-702VI.L-2VI.L-2XCP-620XCP-6204-34-3GTP-702GTP-702VI.L-2VI.L-2XCP-621XCP-6211-11-1GTP-702GTP-702IV.CIV.CXCP-632XCP-6326-56-5GTP-702GTP-702IV.B, IV.D, IV.E,IV.B, IV.D, IV.E,VI.KK, VI.NNVI.KK, VI.NNOAP-106.1OAP-106.1RB TEMPSRB TEMPSRB SUMP LEVELRB SUMP LEVELMW/KV/MVARSMW/KV/MVARSGENERIC LOG SR NIGENERIC LOG SR NIOAP-100.6OAP-100.6OPERATION ATOPERATION ATLICENSED LIMITLICENSED LIMITOAP-107.1OAP-107.1RESTORATION OFRESTORATION OFIPCS FUNCTIONSIPCS FUNCTIONSXCP-638XCP-6381-41-4OAP-106.1OAP-106.1MW/KV/MVARSMW/KV/MVARSXCP-638XCP-6382-42-4OAP-106.1OAP-106.1MW/KV/MVARSMW/KV/MVARS1(/-!T!#;!T1(/-!T!#;!T JPM NO:NJPA-210A2015 NRC SRO A2: Determine Administrative Actions to Place 1DB on Alternate Feed.V.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:SRO ONLYThursday, January 29, 2015Page 1 of TASK:341-038-03-02PREFERRED EVALUATION METHODPERFORMTASK STANDARD:Section 1 and 2 of the Removal and Restoration Checklist is completed satisfactorily (see Key foNJPA-210A). It is critical to identify as an action R&R, that the Train is "B", the Equipment ID is XSW1DB 16, Equipment name is Bus 1DB Normal Incoming Breaker, to indicate that the TS is 3.8.1.1.a (the a is optional), that 3.0.4 does apply, that the restraining mode is 4 and that the mode discovered is 1, Compensatory Requirements are GTP-702, Att. VI.Y-1 and "Other" with some statement about recording bus voltage readings hourly.PREFERRED EVALUATION LOCATIONCLASSROOMINTERPRET AND ENSURE COMPLIANCE WITH PLANT ADMINISTRATIVE PROCEDURES DURING NORMAL AND OFF NORMAL PLANT OPERATIONSTOOLS:NJPA-210A handout (hardcopy of SAP-205, Attachment I, REMOVAL AND RESTORATION CHECKSHEET.)Electronic access to the following: SOP-304, 115KV/7.2KV OPERATIONS

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TERMINATING CUE:After the Removal and Restoration Checksheet is provided to the Evaluator this JPM is complete.OAP-106.1Operating RoundsSTP-125.001Electric Power Systems Weekly TestSAP-205STATUS CONTROL AND REMOVAL AND RESTORATIONSOP-304115KV/7.2KV OPERATIONST.S.Technical SpecificationsINDEX NO.ROSROK/A NO.1940012214Knowledge of the process for controlling equipment configuration or status.3.94.32.2.14Thursday, January 29, 2015Page 2 of Technical Specification 3.8.1.GTP-702, SURVEILLANCE ACTIVITY TRACKING AND TRIGGERINGSAP-205, STATUS CONTROL AND REMOVAL AND OAP-106.1, OPERATING ROUNDSEVALUATION TIME20TIME CRITICALNOTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)(13)CANDIDATE:

INITIATING CUES:As the CRS, complete Section 1 and Section 2 of a Removal and Restoration Checksheet against XSW1DB 16, BUS 1DB NORMAL INCOMING BKR to track all the requirements associated with transferring Bus 1DB to XTF-4/6 including any necessary compensatory actions.The R&R number 150333 has been assigned. Record your answers in section 1 and section 2 of the provided SAP-205, Attachment I, REMOVAL AND RESTORATION CHECKSHEET (NJPA-210A handout).Note: The required by date and time for compensatory requirements if necessary will be filled in when the breaker is declared inoperable. You are to leave that field blank for this JPM.INITIAL CONDITION:100% power. B1 Maintenance Week is in progress.ESF Bus 1DB must be transferred to XTF-4/6 to allow XSW1DB 16, BUS 1DB NORMAL INCOMING BKR to be replaced. The Integrated Fire Computer is being fed from Train "A".XAC-12-IA, SUPP INST AIR COMPRESSOR is NOT running.The work is expected to take approximately two hours. ESF Bus 1DA will remain on the normal source during the work and alarm setpoints will NOT be adjusted since the alignment will only be in effect for approximately two hours. The BOP operator is preparing a pre-job brief for the evolution.READ TO OPERATOR:SAFETY CONSIDERATIONS:None.WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORHAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!Thursday, January 29, 2015Page of STEPSComplete section 1 of SAP-205, STATUS CONTROL AND REMOVAL AND RESTORATION, Attachment I, REMOVAL AND RESTORATION CHECKSHEET.Evaluator Note: Refer to NJPA-210A Key.YesThe section 1 data are:Type: Action (critical)Service Impact: Removed From Service (NOT critical)

Train: 'B' Train (critical)

R&R Number: 130333 (NOT critical)System: ES (NOT critical)Equipment ID: XSW1DB 16 (critical)

Equipment Name: Bus 1DB Normal Incoming Bkr (critical)Reason Inoperable: Something to the effect of breaker replacement (NOT critical).CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:1Thursday, January 29, 2015Page of Examiner ends JPM at this point.Complete section 2 of SAP-205, STATUS CONTROL AND REMOVAL AND RESTORATION, Attachment I, REMOVAL AND RESTORATION CHECKSHEET.Evaluator note: Refer to NJPA-210A key.Evaluator Note: Redundant Equipment Operable is typically used for things like charging pumps that have a swing component, but in this case there is an alternate feed.YesThe section 2 data are:Compensatory Requirements: GTP-702, Att. VI.Y-1 and "Other" and some statement about recording bus voltage readings hourly. (critical.)Technical Specifications: TS 3.8.1.1.a (critical). ("a" may be left off)Tech Spec 3.0.4 applies: Yes (critical)

Restraining Mode: 4 (critical)

Mode Discovered: 1 (critical) Redundant Equipment Operable: Yes or No (not critical)CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Thursday, January 29, 2015Page of JPM NO:NJPA-210ADESCRIPTION:2015 NRC SRO A2: Determine Administrative Actions to Place 1DB on Alternate Feed.IC SET:NAINSTRUCTIONS:COMMENTS:NJPA-210A folder in the Exam Data folder contains answer keyJPM SETUP SHEETThursday, January 29, 2015Page of INITIATING CUES:As the CRS, complete Section 1 and Section 2 of a Removal and Restoration Checksheet against XSW1DB 16, BUS 1DB NORMAL INCOMING BKR to track all the requirements associated with transferring Bus 1DB to XTF-4/6 including any necessary compensatory actions.The R&R number 150333 has been assigned. Record your answers in section 1 and section 2 of the provided SAP-205, Attachment I, REMOVAL AND RESTORATION CHECKSHEET (NJPA-210A handout).Note: The required by date and time for compensatory requirements if necessary will be filled in when the breaker is declared inoperable. You are to leave that field blank for this JPM.INITIAL CONDITION:100% power. B1 Maintenance Week is in progress.ESF Bus 1DB must be transferred to XTF-4/6 to allow XSW1DB 16, BUS 1DB NORMAL INCOMING BKR to be replaced. The Integrated Fire Computer is being fed from Train "A".XAC-12-IA, SUPP INST AIR COMPRESSOR is NOT running.The work is expected to take approximately two hours. ESF Bus 1DA will remain on the normal source during the work and alarm setpoints will NOT be adjusted since the alignment will only be in effect for approximately two hours. The BOP operator is preparing a pre-job brief for the evolution.SAFETY CONSIDERATIONS:None.OPERATOR INSTRUCTIONS:JPM BRIEFING SHEETHAND THIS PAPER BACKTO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

SAP-205ATTACHMENT I PAGE 1 OF 1 REVISION 10REMOVAL AND RESTORATION CHECKSHEETSection 1-Summary Data TYPE: ActionTracking SERVICE IMPACT: Removed From Service Restricted Service TRAIN:'A' Train 'X' Train 'B' Train N/AR&R NUMBER: SYSTEM:EQUIPMENTID:EQUIPMENT NAME: REASON INOPERABLE: Section 2-Removal Requirements COMPENSATORY REQUIREMENTS:Required By Date/Time Completed Date/Time TECHNICAL SPECIFICATIONS: NoneTrip/Bypass Bistables? / / Backup Fire Suppression? / / TECH. SPEC. 3.0.4 APPLIES: YesNo Restraining Mode:___ Mode Discovered:___ REDUNDANT EQUIPMENT OPERABLE: YesNo N/ARoving Fire Watch? / / Continuous Fire Watch? / / Alternate Radiation Monitoring? / / SUPPORTING DOCUMENTATION: Smoke Detectors Operable? / / GTP-702 Att. ____________ / / Other: / / REMOVAL COMMENTS: Section 3-Restoration Req./Related Documents RESTORATION REQUIREMENTS: RELATED DOCUMENTS: Operable STP STTS # Completed Date/Time Document Type* Document #Completed Initials/Date Comments // // // //All compensatory requirements restored or terminated? YesNo N/AECR Operability Form?YesNo N/A//Continued on Attachment VII. *ECR, MWR, NCN, PMTS, RTO, STTS, WPO, etc.RESTORATION COMMENTS: Section 4-Removal/Restoration Status REMOVAL/RESTORATION STATUS: SS Authorization OATC Concurrence Date/Time Updated MCBBISIEOOSDeclared Inoperable / Yes NoYes NoYes NoTime Limit to Declare Operable Restoration Required By / Downgraded to:Tracking / Restricted Service / Yes NoDeclared Operable / Yes NoYes NoYes NoTotal Time: Inoperable Non-Functional COMMENTS:

SAP-205ATTACHMENT I PAGE 1 OF 1 REVISION 10REMOVAL AND RESTORATION CHECKSHEETSection 1-Summary Data TYPE: ActionTracking SERVICE IMPACT: Removed From Service Restricted Service TRAIN:'A' Train 'X' Train 'B' Train N/AR&R NUMBER: SYSTEM:EQUIPMENTID:EQUIPMENT NAME: REASON INOPERABLE: Section 2-Removal Requirements COMPENSATORY REQUIREMENTS:Required By Date/Time Completed Date/Time TECHNICAL SPECIFICATIONS: NoneTrip/Bypass Bistables? / / Backup Fire Suppression? / / TECH. SPEC. 3.0.4 APPLIES: YesNoRestraining Mode:___ Mode Discovered:REDUNDANT EQUIPMENT OPERABLE: YesNo N/ARoving Fire Watch? / / Continuous Fire Watch? / / Alternate Radiation Monitoring? / / SUPPORTING DOCUMENTATION: Smoke Detectors Operable? / / GTP-702 Att. ____________ / / Other: / / REMOVAL COMMENTS: Section 3-Restoration Req./Related Documents RESTORATION REQUIREMENTS: RELATED DOCUMENTS: Operable STP STTS # Completed Date/Time Document Type* Document #Completed Initials/Date Comments // // // //All compensatory requirements restored or terminated? YesNo N/AECR Operability Form?YesNo N/A//Continued on Attachment VII. *ECR, MWR, NCN, PMTS, RTO, STTS, WPO, etc.RESTORATION COMMENTS: Section 4-Removal/Restoration Status REMOVAL/RESTORATION STATUS: SS Authorization OATC Concurrence Date/Time Updated MCBBISIEOOSDeclared Inoperable / Yes NoYes NoYes NoTime Limit to Declare Operable Restoration Required By / Downgraded to:Tracking / Restricted Service / Yes NoDeclared Operable / Yes NoYes NoYes NoTotal Time: Inoperable Non-Functional COMMENTS:

2011 and 2015 NRC Admin A3 RO & SRO: Apply Facility ALARA Principles to a Specific Task and Determine Overall DoseV.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:JPM NO:NJPA-083A(R1)THIS JPM IS NOT APPROVEDPage 1 of TASK:000-061-05-0PREFERRED EVALUATION METHODPERFORMEVALUATION TIME15TIME CRITICALNOTASK STANDARD:Calculate dose for each case. Determines exposure as 590 to 670 mR with a respirator and 537to 538 mR without a respirator. Determines that working WITHOUT respirator is the best option.PREFERRED EVALUATION LOCATIONCLASSROOMTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)(10)RESPOND TO AREA RADIATION MONITORING SYSTEM ALARMSTOOLS:HPP-0153 and HPP-0155 (Available via desk top computer)CalculatorNJPA-083A(R1) Handout 1 (Worksheet)

NJPA-083A(R1) Handout 2 (Hardcopy of HPP-155)

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TERMINATING CUE:Options have been prioritized and provided to the Evaluator.CANDIDATE:HPP-0155Control of Airborne Radiation Exposure (DAC-HRS)HPP-0153Administrative Exposure LimitsINDEX NO.ROSROK/A NO.1940012312Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.3.23.72.3.12Page 2 of INITIATING CUES:You have been assigned to calculate the expected dose for the two INITIAL CONDITION:A hydrogen explosion in the waste gas system has resulted in a The crew is performing actions of ARP-019 XCP-644; 2-1, GAS DECAY TK AREA RM-G10 HI RAD. The leak will continue until several manual valves are manipulated to isolate the leak. Access to the valves requires climbing a ladder in a tight space which has an ambient temperature of 100°F.The general area radiation level where the work will be performed is 1000 mR/hour.Airborne activity in the work area is estimated at a Weighted Derived Air Concentration of 30 DAC.There are two options for performing the work:- One person without a respirator = 30 minutes or- One person with an SCBA = 36 minutes.READ TO OPERATOR:SAFETY CONSIDERATIONS:WHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORPage 3 of HAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!

STEPSAccess necessary reference material.Evaluator cue: Provide a copy of NJPA-083A (R1) Handout 1 (worksheet) and NJPA-083A(R1) Handout 2 (copy of HPP-155)Evaluator note: Procedures that are applicable are HPP-0153, Administrative Exposure Limits and HPP-0155, Control of Airborne Radiation Exposures (DAC-HRS). These are available via the desktopcomputers.YesReviews conditions and refers to procedures for respirator factors and DAC conversion.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:NoSTEP:1Page of Calculate dose for each option - One person without respirator.Evaluator note: Acceptable range is 537 to 538 mR. Step is critical since an accurate calculation is required in order to make correct ALARA decision.YesExternal exposure = (1000 mr/hr) (1 hr/60 minutes) (30 minutes) Internal exposure = (30 DAC) (30 minutes) (1 hr/60 minutes) (2.5 mr/DAC-hr) = 37.5 mR.Total Exposure without respirator = 537.5 mR.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Page of Examiner ends JPM at this point.Calculate dose for each option - One person with a respirator.Evaluator note: Acceptable range is 590 - 670 mR. The acceptable range allows for failure to apply the respirator inefficiency factor of 1.1. This factor is not common knowledge. Step is critical since an accurate calculation is required in order to make correct ALARA decision.YesExternal dose = (1000 mr/hr) (1 hr/60 minutes) (36 minutes) (1.1) = 660 mRInternal Dose = (30 DAC) (36 min) (1 hr/60 min) (2.5 mr/DAC-hr) (1.1) / (1000 protection factor)= 0495 mR.

Total Exposure with respirator = 660.05 mR.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:3Determines best option IAW the lowest total dose.Evaluator note: Step is critical since the correct comparison must be made in order to make correct ALARA decision.YesBest option is: One person WITHOUTrespirator.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:4Page of JPM NO:NJPA-083A(R1DESCRIPTION:2011 and 2015 NRC Admin A3 RO & SRO: Apply Facility ALARA Principles to a Specific Task and Determine Overall DoseIC SET:N/AINSTRUCTIONS:COMMENTS:JPM SETUP SHEET, , 201Page of INITIATING CUES:You have been assigned to calculate the expected dose for the two the best option between them according to the VC Summer ALARA philosophy. For the purposes of the JPM, assume that no dose is received in transit and there is no additional external exposure due to respiratory equipment setup. Place your name on the NJPA-083A(R1) handout and write your answer in the space provided. Show all work.INITIAL CONDITION:A hydrogen explosion in the waste gas system has resulted in a The crew is performing actions of ARP-019 XCP-644; 2-1, GAS DECAY TK AREA RM-G10 HI RAD. The leak will continue until several manual valves are manipulated to isolate the leak. Access to the valves requires climbing a ladder in a tight space which has an ambient temperature of 100°F.The general area radiation level where the work will be performed is 1000 mR/hour.Airborne activity in the work area is estimated at a Weighted Derived Air Concentration of 30 DAC.There are two options for performing the work:- One person without a respirator = 30 minutes or- One person with an SCBA = 36 minutes.SAFETY CONSIDERATIONS:OPERATOR INSTRUCTIONS:JPM BRIEFING SHEETHAND THIS PAPER BACKTO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

NJPA-083A(R1) Application of ALARA Principles - Handout Examinee Name __________________________ Dose Calculation - One Person without a respirator: Dose Calculation - One Person with a respirator: _________________________________________

HPP-0155CONTROL OF AIRBORNE RADIATION EXPOSURE (DAC-HRS)Revision 13SAFETY RELATEDINFORMATION USEProcedure May Be Performed from Memory.User retains Accountability For Proper Performance.South Carolina Electric and Gas CompanyVirgil C. Summer Nuclear StationNUCLEAR OPERATIONSCOPY NO.

HPP-0155PAGE iREVISION 13TABLE OF CONTENTSSECTIONPAGE1.0PURPOSE/SCOPE

12.0REFERENCES

13.0RESPONSIBILITY24.0LIMITS AND PRECAUTIONS25.0PROCEDURE46.0ENCLOSURES97.0RECORDS98.0REVISION SUMMARY9ATTACHMENTSAttachment I-DeletedAttachment II-Deleted Attachment III-Deleted Attachment IV-Airborne Area Entry Tracking Log Attachment V-DAC-Hr Tracking and Calculation Attachment VI-Deleted Attachment VII-Deleted Attachment VIII-Xe-133 Skin Dose Calculation Attachment IX-TEDE ALARA Respirator Evaluation HPP-0155REVISION 13Page 1of 91.0PURPOSE/SCOPE1.1This procedure provides the requirements and methodology for:1.1.1Entry into Airborne Radioactivity Areas, 1.1.2Tracking the accumulation of DAC-hours, 1.1.3Assigning the skin dose equivalent due to noble gases, and 1.1.4Performing ALARA assessment of airborne radioactivity protection methods.1.2This procedure implements portions of 10CFR20 and meets the intent of FSAR sections 12.3.1.2 and 12.3.2.3 for maintaining airborne activity monitoring and controlling personnel airborne radioactivity exposure. This procedure implements the intent of managerial or administrative procedures governing the conduct of facility operations (subject to the control of 10CFR50, Appendix B). The PCAP'ed sections of this procedure are governed by SAP-630. A 10CFR50.59 review is not required for this procedure.

2.0REFERENCES

2.110 CFR Part 20, "Standards for Protection Against Radiation"2.2FSAR, "Final Safety Analysis Report", Section 12.3 2.3HPP-0151, "Use of the Radiation Work Permit" 2.4HPP-0152, "Radiation Control Area Access Control" 2.5HPP-0515, "Interpretation of Bioassay Analyses" 2.6SAP-500, "Health Physics Manual" 2.7NUREG-0041, "Manual of Respiratory Protection Against Airborne Radioactive Materials"2.8Technical Work Record 98-008, "Skin Dose From Noble Gas Revisited", October 7, 19892.9HPP-0825, "Weighted DAC Determination for Airborne Alpha Radioactivity" 2.10SentinelUser's Manual HPP-0155REVISION 13 Page 2of 9 2.11Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection"2.12SAP-999, "Corrective Action Program" 2.13EPRI Alpha Monitoring Guidelines for Operating Nuclear Power Plants, Report # 3002000409 August 20133.0RESPONSIBILITY3.1It is the responsibility of all Managers and Supervisors to ensure that each individual within their organization complies with the requirements outlined in this procedure.3.2It is the responsibility of each individual entering a known Airborne Radioactivity Area to record his/her name, HPID number, work location, time in, and time out on Attachment IV or on the computerized access control system.3.3It is the responsibility of Health Physics to:3.3.1Control access into Airborne Radioactivity Areas, 3.3.2Provide primary radiological coverage for entries into areas where the DAC, due to airborne particulate, radioiodine and tritium, is greater than or equal to 1.0,C013.3.3Record the DAC-hours received in a timely manner to ensure compliance with Reference 2.1, (10CFR20, Section 20.1202),3.3.4Schedule whole body counts when required, 3.3.5Ensure skin dose is assigned properly when the noble gas concentration due to Xe-133 is >1 DAC.3.4TEDE ALARA assessments require some degree of professional judgment and should therefore be performed by a Health Physics representative qualified to provide primary radiological coverage.4.0LIMITS AND PRECAUTIONS4.1Noble gases are considered to be an external radiation hazard and the deep dose equivalent is measured by the TLD. Reference 2.8 documents that the TLD does not detect Xe-133 betas. Xe-133 skin exposure should be evaluated in accordance with this procedure.

HPP-0155REVISION 13 Page 3of 9 4.2DAC-hours are calculated and documented for individuals exposed to airborne particulate, radioiodine, gas, and tritium exceeding 0.25 WDAC using Attachments IV and V. DAC-hours due to gas do not need to be tracked in the SENTINEL database.Skin dose is calculated and documented for individuals exposed to Xe-133at levels equal to or greater than 1 WDAC using Attachments IV and VIII.4.3Health Physics shall be notified prior to:4.3.1Entry into any posted Airborne Radioactivity Area, or 4.3.2Entry into any area where airborne radioactivity, in excess of 0.25 DAC, is suspected.4.4Whole body counts should be performed under the following conditions:4.4.1When a worker has accumulated 40 DAC-hrs beta/gamma (excluding Noble Gas) since their last whole body count,within the current year.4.4.2Any worker receiving 4 DAC-hrs Alpha plus Beta/Gamma in 7 consecutive days (corresponding to CEDE internal dose of 10 mrem, planned or unplanned).C014.5Personnel with positive whole body counts should receive timely assessment of internal dose for appropriate exposure record update.4.6Process and engineering controls, such as ventilation, containment devices, and decontamination shall be used, as practical, to minimize exposure of personnel to airborne radioactivity.4.7When personnel are exposed to airborne radioactivity, increased monitoring is required and TEDE is to be maintained ALARA. Respiratory protection may be used when the total risk to the worker is reduced. In some cases, use of respiratory protection may increase worker risk (either radiological or industrial risk). In these cases, the lower risks associated with not using respiratory protection may be acceptable. When ALARA assessments indicate that the lowest risks are obtained by allowing unprotected exposure to airborne radioactivity, document this decision on the RWP and inform the Manager, Health Physics and Safety Services/Designee.4.8Individuals likely to receive internal dose will have an administrative internal dose guideline based on the individual administrative threshold value in Enclosure C or the estimate calculated on Attachment IX (without respiratory equipment),

depending on the TEDE ALARA evaluation method used. A Condition Report HPP-0155REVISION 13 Page 4of 9 will be initiated if an individual receives an internal dose in excess of the administrative internaldose guideline. 4.9A High Significance CRwill be initiated if an individual's TEDE dose (external and internal dose) exceeds the administrative (external and internal) dose guidelines by 100 mrem.4.10Attachment V may be computer generated provided it contains, at a minimum, the information required by the attachment in this procedure.4.11In vitro bioassay analysis may be required, if an alpha uptake is suspected. These samples will be collected in accordance with the requirements of Reference 2.5 (HPP-0515).5.0PROCEDURE5.1Manual logging of personnel entry/exit in Airborne Radioactivity Areas using Attachment IV.5.1.1Complete the information as required on Attachment IV.

5.1.2This information isused to determine entryduration.

5.1.3After all the information is input into the SENTINELsystemthe attachment can be discarded.5.2Manual calculation and tracking of WDAC-hrs Using Attachment V5.2.1Transfer RWP, HPID, name, and entry and exit times to Attachment V.

5.2.2Calculate elapsed time to the nearest tenth of an hour and record.

5.2.3Record respirator type, protection factor and sample ID number.

5.2.4Calculate and record weighted DAC-hr as follows:WDAC-hr = (Activityi)x Elapsed TimeDACi5.2.5Calculate and record effective DAC-hr as follows:EDAC-hr = WDAC-hrProtection factor HPP-0155REVISION 13 Page 5of 9 5.2.6Calculate and record the dose receivedas follows:Dose (mrem) = EDAC-hr x 2.5 mrem per EDAC-hr5.3Manual calculation and tracking of Xe-133 skin exposure Using Attachment VIII5.3.1Transfer RWP, HPID and name, to Attachment VIII from Attachment IV.

5.3.2Referring to Attachment IV calculate elapsed time to the nearest tenth of an hour and record on Attachment VIII.5.3.3Record sample ID number andactivity on Attachment VIII.

5.3.4Calculate and record DAC-hr as follows:DAC-hrXe-133=Sample ActivityXe-133X ElapsedTimeDACXe-1335.3.5Calculate skin dose as follows:Xe-133 Skin Dose = Total DAC-hrXe-133x 3.5mrem per DAC-hrXe-1335.3.6Complete Attachment VIII and forward to Dosimetry for update of the affected individual's exposure record.5.4To add or review Xe-133 Skin Dose records in Sentinel, open theDose Analysis module.5.4.1Open the desired worker and select the External Dose tab.

5.4.2Select File, then New to initiate the dose analysis.

5.4.3Ensure that Whole Body Dose is selected as the Type of dose analysis.

5.4.4The default date may be accepted for the Date Initiated field.

5.4.5Complete the Initiator's HPID field.

5.4.6Select 'EXT -XE-133 SKIN DOSE' as the Dose Analysis Reason.

5.4.7The RWP Number, Task Number, and Comment fields shouldbecompleted for reference purposes.5.4.8The entry and exit dates & times from Attachment IV should be entered for the dose analysis Begin and End fields.

HPP-0155REVISION 13 Page 6of 9 5.4.9Enter the serial number of the TLD assigned to the worker during the time of the airborne area entry.5.4.10Select 'Adding' in the dropdown field next to the TLD No. field.

5.4.11Enter a zero into the DDE and Lens dose fields.

5.4.12Enter the dose calculated on Attachment VIII into the Shallow dose field.5.4.13Save the dose analysis to assign the dose to the worker as estimated dose.5.4.14The record will remain as estimated dose until the dose analysis has been approved by someone other than the Initiator.5.4.15Use the User Utilities function to generate a printout of the dose analysis.5.4.16Once signed by the Initiator and the Approver, the dose analysis will be placed in the worker's exposure file.5.5To add or review DAC-hr records in Sentinel, open the Dose Analysis module.5.5.1Open the desired worker and select the DAC Hourstab.NOTE:5.5.2DAC-hrs due tonoble gas (other than Xe-133) do not need to be tracked in SENTINEL. Skin dose assessment from exposure to Xe-133will be done per section5.3.5.5.2Select File New to enable the fields and add all required data.5.5.3The Air Sample ID field is a reference field where you may enter an air sample ID number for later reference.5.5.4Enter the Performer's HPIDand date initiated (or accept the default date).5.5.5Select 'DAC -AIRBORNE AREA ENTRY' as the Dose Analysis Reason 5.5.6You may enter RWP and Task numbers for reference purposes.

5.5.7The Comment field may be completed for reference purposes.

HPP-0155REVISION 13 Page 7of 9 5.5.8Enter the required begin and end dates and times.5.5.9The dose blockcontains fields for DAC-hr Values, estimated dose and resolution dose.5.5.10The DAC-hrs Value is multiplied by 2.5 (2.5 mrem per DAC-hr) and displayed in the estimated dose field.5.5.11Savethe dose analysisto assign the dose to the selected workeras estimated dose.NOTE: 5.5.12External dosedue to exposure to airborne radioactivity (other than Xe-133) is accounted for when the worker's TLD is processed. .5.5.12The record will remain as estimated dose until a resolutiondose is assigned and approved.5.5.13When theestimated dose is equal to or greater than 10 millirem the estimated dose shall be assigned as the Resolution Dose. The resolution Dose shall be assigned as zero millirem when the estimated dose is less than 10 millirem.5.5.14The Approval Information sub tab must be completed and savedby someone other than the Initiator, for theresolution dose to be assigned.5.6The attachment sub tab is used to link any supporting file.5.6.1 The attachment is a link to the actual file and if the file is moved, the link will be broken.5.6.2Select "attach file"and browse to the desired file, double click or choose "open to attach."5.6.3A file may be removed by selecting "detach"and"yes"at the prompt.5.7TEDE ALARA Evaluations 5.7.1TEDE ALARA evaluations of airborne radioactivity protection methods shall be performed and documented, using the guidance in Enclosure A and the calculations in Attachment IX, if the following deep-dose equivalent thresholds are expected to be exceeded:

HPP-0155REVISION 13 Page 8of 9 Administrative Threshold ValuesFor an individual per entry10 DAC-hrs or 25 mrem DDECollective dose for a task20 DAC-hrs or 50 mrem DDE 5.7.2TEDE ALARA evaluations do nothave tobeperformed or documented if:A.The administrative threshold values are not expected to be exceeded, or B.The degree of uncertainty, regarding the assumptions used for the calculations in Attachment IX, is sufficient enough that the estimates can not be calculated. (For example, there is no historical data from a previous/similar job, or there is no data due to the job being new to the station.)5.7.3Health Physics may perform a qualitative assessmentusing a mental process and professional judgment for jobs that are not expected to exceed the administrative threshold values but that may require respiratory protection. The guidance in Enclosure Cshould be used.5.7.4TEDE ALARA assessments should show that TEDE dose for the job will be ALARA. That is, the internal dose avoided by using respiratory protection would be greater than any additional external dosethat may result from factors like respirator-induced worker inefficiency.5.7.5When assessment results do not show a clear dose savings or obvious indication to use or not use respiratory protection, the evaluator will make the determination using professional judgment. 5.7.6Industrial safety considerations may outweigh radiological (ALARA) considerations and must be included in any evaluation. 5.7.7When changes to the original assessment occur, the reasons should be documented within the applicable RWP. (See step 4.7)

HPP-0155REVISION 13 Page 9of 9 6.0ENCLOSURES6.1Enclosure A-Total Risk Assessment Guide for Use of Respiratory Equipment6.2Enclosure B-Respiratory Protection Factors 6.3Enclosure C-Qualitative Assessment Guide7.0RECORDS7.1Attachment V will be stored in the Dosimetry records office until it is transmitted to Nuclear Records for permanent retention.7.2Attachment VIII will be placed in the affected individual's Dosimetry file.7.3Attachment IX will be placed into the associated RWP Package.8.0REVISION SUMMARY8.1Revision 13 implements all changes to Revision 12; deletes old Reference 2.13;and adds new 2.13; deletes original 4.4.2 and makes 4.4.3 new 4.4.2; adds 4.11 to implement new EPRI Alpha Guideline and delete HPP-031; deletes "may" adds "should" in 5.4.7; deletes last sentence in Note 5.5.12; adds new 5.5.13 and renumbers old to 5.5.14 to implement new Sentinel.

HPP-0155ENCLOSURE A PAGE 1OF 2 REVISION 13TOTAL RISK ASSESSMENT GUIDE FOR USE OF RESPIRATORY EQUIPMENT1.Determine if Engineering Controls can be utilized to reduce or eliminate airborne radioactivity. (Engineering Controls are preferred over use of respiratory equipment).2.If Engineering Controls cannot be utilized, Determine what other controls may be used to limit exposure; like restricting access, limiting stay times, etc.3.A Task Evaluation (using available historical data or reliable estimates ) is done to determine:who is performing the work, what work activity is to be performed,where work activity will be performed,when the work activity will be performed,expected duration of job, work area radiological conditions (loose and/or fixed contamination levels, general area dose rates, potential alpha contamination, expected DAC),what protective clothing/gear is required and how it will affect worker comfort/efficiency,will the respiratory equipment increase the worker's industrial safety risk thus outweighing the benefit in TEDE reduction? (increase heat stress, limit vision range while climbing, etc.)will the respiratory equipment decrease the worker's industrial safety risk? (decrease heat stress, etc.)is the respiratory equipment needed for industrial hazards?are there postjob negative impacts to the worker? (personnel contamination, unplanned intake, etc.)NOTE:4.5.6If large uncertainties are present in the radioactiveconcentration determination or estimation due to a new job or no historical data, the calculations in Attachment IX do not need to be performed.

Instead, a qualitative assessment based on professional judgment will be made using Enclosure C4.Determine if the use of respiratory equipment will be TEDE ALARA using the information gathered above andthe calcualtions in Attachment IX.

HPP-0155ENCLOSURE A PAGE 2OF 2 REVISION 135.Attachment IX of this procedure will be used to document TEDE-ALARA evaluations and calculated estimates. The following factors are used in the calculations: 2.5 mrem/DAC-hr (Conversion Factor for DAC-hr to mrem)1.10 (Respirator-induced worker inefficiency factor, represents the expected lengthening of a work activity due to wearing a respirator) Any exposure from setting up breathing air equipment should be added to the External exposure total.Respirator Protection Factors are found in Enclosure B.6.Compare total exposure with and without a respirator. Respirator use is indicated when TEDE is lower with a respirator than without one. Note that industrial safety considerations may outweigh ALARA considerations and must be included in any evaluation. Continue with step 8.7.Evaluation results will be documented in the applicable RWP.

8.When a change occurs from the initial evaluation or protective requirement, the change should also be documented in the applicable RWP. Examples of changes include:Respiratory equipment needs to be worn for industrial safety reasons, Respiratory protection will not be worn because TEDE is expected to increase due to the use of respiratory protection, Respiratory protection will be worn to reduce risk of an intake because peak airborne concentration cannot be reasonably determined, and is potentially very high, Respiratory protection will be worn due to a worker's perceived need for respiratory protection,Respiratory equipment is not worn due to safety concernsand an intake is possible, or due to high dose rates when an intake is possible.

HPP-0155ENCLOSURE B PAGE 1 OF 1 REVISION 13RESPIRATORY PROTECTION FACTORSDEVICEPARTICULATEIODINETRITIUMNOBLE GASMSA FULL FACENEG. PRESSURE50111MSA FULL FACEDUO-FLOW50111MSA FULL FACEPRESSURE DEMAND1000111MSAFULL FACECONTINUOUS FLOW 1000111SCBA1000100011HOOD1000100011MURUROA V4 F1 BUBBLE SUIT5000500001 HPP-0155ENCLOSURE C PAGE1OF1 REVISION 13QUALITATIVE ASSESSMENT GUIDEA qualitative assessment using a mental process and professional judgment should be used for jobs that are not expected to exceed the administrative threshold values. AdministrativeThreshold ValuesFor an individual per entry10 DAC-hrs or 25 mrem DDECollective dose for a task20 DAC-hrs or 50 mrem DDEThe following key points should be considered: Who is performing the work (do they have previous experience)What work acitivity is to be performed (does the activity have the potential to generate airborne radioactivity)Where the work will be performed (what are the environmental conditions)When the work will be performed (what other factors can affect the job)Expected duration of the jobRadiological conditions (loose surface and fixed contamination levels, dose rates, potential alpha contamination levels, potential DAC) What process or engineering controls can be used (filtered ventilation, filtered vacuum, decontamination)Will the probability of occurrence(based on the work activity)/potential radioactive airborne concentration (based on contamination levels) be Low/Low, High/Low, Low/High, or High/High?A.Rate the work as having a "High" or "Low" probability for creating airborne radioactivity.B.Rate the potential radioactive airborne concentration as "High" or "Low" based on the work area contamination levels and the work activity.C.Combine the assessments from A and B to determine, in addition to other factors listed in this enclosure, the need for respiratory protection. Examples of Low/Low conditions include non-abrasive work, minor valve work, hand tool use, or deconatamination in an area with loose surface contamination levels <100,000 dpm/100 cm2. Low/Low conditions do not normally require respiratory protection.Examples of High/Low conditions include abrasive work, welding, cutting, or machining in an area with loose surface contamination levels <100,000 dpm/100 cm2. High/Low conditions may require respiratory protection and will be at the professional judgment of the evaluator.Examples of Low/High conditions include non-abrasive work, hand tool use, or wiring in an area with loose surface contamination levels >100,000 dpm/100 cm2 . Low/High conditions may require respiratory protection and will be at the professional judgment of the evaluator.Examples of High/High conditions include abrasive work, cutting, machining, or decontamination in an area with loose surface contamination levels >100,000 dpm/100 cm2. High/High conditions will normally require respiratory protection.The qualitative assessment results should be noted in the applicable RWP using a simple statement regarding the probability of occurrence/potential radioactive concentration.

HPP-0155ATTACHMENT IV PAGE 1 OF 1 REVISION 13AIRBORNE AREA ENTRY TRACKINGLOGDATE:HPIDLAST NAMEWORKSTATIONRESP TYPEENTRYTIMEEXIT TIMEAIR SAMPLE NUMBERRWPRESP TYPE:N-NONEF-FULL FACEA-AIRLINES-SCBAH-HOOD HPP-155ATTACHMENT V PAGE 1 OF 1 REVISION 13RWPLASTENTRYEXITELAPSEDRESP.PROT.ISOSAMPLEISO TYPEWDACEDACDOSENUMBERHPIDNAMEDATE / TIMEDATE / TIMEHOURSTYPEFACT.TYPEIDWDAC*HOURSHOURSmrem *** FROM HPP0303 attachment IIICALCULATED BYDATE** EDAC HOURS x 2.5 mrem/EDAC hrINPUT BYDATEREVIEWED BYDATESOUTH CAROLINA ELECTRIC & GAS CO. - V.C. SUMMER NUCLEAR STATIONDAC-HR TRACKING AND CALCULATIONS HPP-0155ATTACHMENT VIII PAGE 1 OF 1 REVISION 13Xe-133 SKIN DOSE CALCULATIONNAME: HPID: RWP: Exposure Date / TimeElapsed TimeSample IDSample ActivityDAC-hrXe-133Total DAC-hrXe-133DAC-hrXe-133=Sample ActivityXe-133X Elapsed TimeDACXe-133Xe-133 Skin Dose = Total DAC-hrXe-133X 3.5 mrem / DAC-hrXe-133=mremPerformed by: NameDateReviewed by: NameDate*Records Updated by: NameDate*Forward to Dosimetry for update andinclusion into individual's exposure records.

HPP-0155ATTACHMENT IX PAGE 1OF 2 REVISION 13TEDE ALARA Respirator EvaluationRWP #: ___________________________Task Evaluated: ____________________________________________________________________________________________________________________________1.Radiological Considerations for Respirator UseAverage Expected Dose Rate in Work Areamrem/hr(A)Job Duration Expectedhours(B)Weighted DAC (WDAC) in Work AreaDAC(C)WDAC-hr = (Activityi)x Elapsed TimeDACiRespirator Protection Factor (Enclosure B)*____________(D)Respirator-induced Inefficiency Factor (EnclosureA)1.10(E)Conversion Factor (EnclosureA)2.5mrem/DAC-hr(F)TEDE = Total Exposure = External Exposure + Internal Exposurea.Without Respiratory Equipment:External = A x B = x=mrem(G)Internal = C x B x F = xx 2.5 = mrem(H)TEDE = G + H = +=mremb.With Respiratory Equipment:External = A x B x E + Respext*=xx+=mrem (I)Internal = (C / D) x B x 1.10 x 2.5= (/)xx 1.10 x 2.5 = mrem (J)TEDE = I + J = +=mrem*Respext= External exposure due to respiratory equipment set-up, if applicable. Contact Respiratory Services if assistance is needed to estimate equipment set-up time or to choose appropriate equipment for task.These values are to be used to determine whether the use of respiratory protection will be detrimental to maintaining the work process ALARA.2.Will wearing respiratory equipment increasethe worker's industrial safety risk?or decreasethe worker's industrial safety risk?

HPP-0155ATTACHMENT IX PAGE 2OF 2 REVISION 13Comments: __________________________________________________________________________

__________________________________________________________________________3.Would notwearing respiratory protection lead to post job negative implications?a.Personnel decontaminationb.Skin dose assessmentc.Portal monitor alarmsd.Extensive bioassay evaluationComments: ___________________________________________________________________________

___________________________________________________________________________

___________________________________________________________________________4. Other considerations: ______________________________________________________________________________________________________________________________________

____________________________________________________________________________

____________________________________________________________________________5. Overall EvaluationRespirator use isindicatedRespirator use is notindicatedEvaluator _________________________________ Date ______________________

JPM NO:NJPA-10032015 NRC SRO A4: Classify Emergency (Simulator - SAE - Inadequate Core Cooling) (ENF)V.C. SUMMER NUCLEAR STATIONJOB PERFORMANCE MEASURECANDIDATE:EXAMINER:SRO ONLYTIME CRITICAL JPMWednesday, January 28, 2015Page 1 of TASK:344-019-03-02PREFERRED EVALUATION METHODPERFORMTASK STANDARD:Emergency classification evaluated as a SITE AREA EMERGENCY per Fission Product BarriersEAL number FS1.1. Classification based on Potential Loss of Fuel Clad Barrier (Item B Core Exit TCs > 700°F) and Loss of Reactor Coolant System Barrier (Item D.2 - RCS leak rate > available make up capacity as indicated by a loss of RCS subcooling). This is a time critical JPMand the declaration must be made within 15 minutes after the emergency condition exists, and successful completion of EPP-002, Communication and Notification, Attachment I, Nuclear PowePlant Notification Form, must be made within 15 minutes after the emergency declaration (see key for this JPM).PREFERRED EVALUATION LOCATIONSIMULATORCLASSIFY EMERGENCY EVENTS REQUIRING EMERGENCY PLAN IMPLEMENTATIONTOOLS:EPP-002, Communication and Notification, Attachment I, Nuclear Power Plant Notification Form (NJPA-1003 Handout)Stopwatch or other suitable timepiece EPP-001, Attachment 1 EAL Classification Matrix (available via desk

REFERENCES:

TERMINATING CUE:Successful completion of EPP-002, Communication and Notification, Attachment I, Nuclear Power Plant Notification Form.EPP-106EMERGENCY PREPAREDNESS PERFORMANCE INDICATOR PROCEDUREEOP-12.0MONITORING OF CRITICAL SAFETY FUNCTIONSEPP-001ACTIVATION AND IMPLEMENTATION OF THE EMERGENCY PLANEPP-002COMMUNICATION AND NOTIFICATIONINDEX NO.ROSROK/A NO.1940012441Knowledge of the emergency action level thresholds and classifications.2.94.62.4.411940012440Knowledge of the SRO's responsibilities in emergency plan implementation.2.74.52.4.40Wednesday, January 28, 2015Page 2 of EVALUATION TIME20TIME CRITICALYESTIME START:TIME FINISH:PERFORMANCE RATING:SAT:UNSAT:PERFORMANCE TIME:EXAMINER:SIGNATUREDATE10CFR55:45(a)(11)CANDIDATE:

INITIATING CUES: 1. Your task is to perform the following: a.Classify the event. b.Complete the required INITIAL NOTIFICATION FORM. 2.Do not use SS Judgment as the basis for your classification. 3.The simulator will remain in freeze during the JPM. 4.There are two Time Critical elements for this JPM; Time ofEvent Classification and completion of the Notification Form.Inform the Evaluator atthe completion of each Time Criticalelement.5.For the purposes of this JPM, initial event time zero and the starttime for classification will begin when the Evaluator tells you to begin.6.Reactor Trip time will be 30 minutes prior to T-0 announced by theINITIAL CONDITION:1.The plant was in MODE 1.2.Weekend night shift, only routine evolutions in progress.3.The 'C' Charging pump was tagged out for maintenance on theprevious shift. No other equipment is out of service.4.An event occurred resulting in an automatic Reactor Trip and Safety5.The crew entered EOP-1.0, REACTOR TRIP OR SAFETY 6.The CRS has received the following reports from the crew.a."Reactor Trip."b."Turbine Trip."

c."Safety Injection" d."ESFLS complete on Train A and Train B"e."The ' B' Charging pump failed to start"f."RCS pressure is less than 1418 psig with flow on FI-943,7.The crew transitioned from EOP-1.0 to "EOP-2.0, LOSS OFREACTOR OR SECONDARY COOLANT"8.The CRS was subsequently notified that the 'A' Charging pump has9.The crew is taking action using EOP-2.0, at step 13.10.The simulator was frozen at this point.READ TO OPERATOR:SAFETY CONSIDERATIONS:NoneWHEN I TELL YOU TO BEGIN, YOU ARE TO PERFORM THE ACTIONS AS DIRECTED IN THE INITIATING CUES. I WILL DESCRIBE THE GENERAL CONDITIONS UNDER WHICH THIS TASK IS TO BE PERFORMED AND PROVIDE THE NECESSARY TOOLS WITH WHICH TO PERFORM THIS TASK. BEFORE STARTING, I WILL EXPLAIN THE INITIAL CONDITIONS, WHICH STEPS TO SIMULATE OR DISCUSS, AND PROVIDE INITIATING CUES. WHEN YOU COMPLETE THE TASK SUCCESSFULLY, THIS JOB PERFORMANCE MEASURE WILL BE SATISFIED.INSTRUCTIONS TO OPERATORWednesday, January 28, 2015Page of Record all work on the NJPA-1003 HandoutTHIS IS A TIME CRITICAL JPM!HAND JPM BRIEFING SHEET TO OPERATOR AT THIS TIME!

STEPSDetermine that conditions require classification as SITE AREA EMERGENCYEvaluator cue: Announce the Initial Event time (T-0) based on Simulator Clock. Record time T-0 for future reference. Announce that reactor trip was 30 minutes prior to T-0.Evaluator note: Examinee has 15 minutes from the end of the plant conditions brief (end of brief = T-0)to arrive at event classification. Record the simulator time when Examinee classifies as theclassification time (Line 10) on EPP-002 Attachment 1 (NJPA-1003 Handout). Hand Examinee the EPP-002 Attachment 1 (NJPA-1003 Handout).YesClassifies event as a SITE AREA EMERGENCY based on:EAL FS1.1 Loss or Potential Loss of any two barriers (Table F-1)CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:1Wednesday, January 28, 2015Page of Examiner ends JPM at this point.Complete EPP-002, Attachment 1, Nuclear Power Plant Emergency Notification Form and provide basis for classification.Evaluator cue: If Examinee appears to be using SIPCS for Rx trip time repeat initiating cue that trip was 30 minutes prior to T-0 for the JPM. Evaluator cue: If asked provide 803-334-1234 as the confirmation phone number.Evaluator cue: Examinee must explain basis for classification. Ask Examinee for basis if it is not offered. If the basis is not correct, this constitutes failure even if the classification was correct and within 15 minutes. Step is critical since proper classification must be made within 15 minutes. Evaluator note: Record Simulator time when Examinee hands you the EPP-002 Attachment as the Approved by time (line 17 on the EPP-002 Attachment 1). Must be within 15 minutes of time when their classification had been completed. Refer to NJPA-1003 Key.YesCorrectly completes EPP-002 Attachment 1 within 15 minutes of declaring event classification. See key for correct manner of completing the attachment.

Classification Basis: 1.Core exit TCs>700°F - Potential Loss of Fuel Clad Barrier, Item B.2.2.RCS leak rate > available make up capacity as indicated by a loss of RCS subcooling - Loss ofReactor Coolant System Barrier, Item D.2.CUES:SEQUENCED:STEP STANDARD:COMMENTS:SATUNSATCRITICAL:YesSTEP:2Wednesday, January 28, 2015Page of JPM NO:NJPA-1003DESCRIPTION:2015 NRC SRO A4: Classify Emergency (Simulator - SAE - Inadequate Core IC SET:318INSTRUCTIONS:If IC-318 is designated for this JPM then reset to IC-318 leaving the simulator in FREEZE.1.Place Danger Tag on 'C' Charging pump for Maintenance.If IC-318 is not designated for this JPM then initial conditions may be established by reseting to IC-10 and following the below directions:

1.Place Danger Tag on 'C' Charging pump for Maintenance.2.Insert: MAL-RCS006A Final Value = 10000 Delay =10 (RCS loop 'A' LOCA) MAL-CVC017A Delay = 120 ('A' Charging Pump Trip) PMP-CS006F (Charging Pump 'B' fail to start) XMT-MI016F 10 Meter Wind Direction Fail As Is XMT-MI008F 10 Meter Wind speed Fail As Is XMT-MI015F 61 Meter Wind Direction Fail As Is XMT-MI007F 61 Meter Wind speed Fail As Is3.RUN 4.Manually trip RCPs when RCS pressure <1400 psig.5.Perform the following actions >1 minute after SI is initiated: Reset SI Reset Phase A Reset Phase B Reset the ESFLS Establish IA to the RB6.Place 'A' and 'B' Charging pumps in pull to lock once they have stopped.7.Ensure steps of EOP-1.0 and in particular EOP-1.0 attachment 3 have been fully and correctly implemented.8.Align EFW for normal operation and throttle to approximatly 200 gpm per Steam Generator.9.When RVLIS NR Level is <45%, reduce RCS leak to 500 GPM.10.When Core Exit Thermocouples >715°F and <725°F with RVLIS >40%: FREEZEJPM SETUP SHEETWednesday, January 28, 2015Page of 11.If necessary adjust RVLIS NR to approximately 45% using: (RV NR Level LI-1311 Fail to Position) XMT-MI002O Final Value = 44XMT-MI005O Final Value = 45 (RV NR Level LI-1321 Fail to Position)12.Record met data from SIPCS, need Wind Direction, Wind Speed and Stability Class for theCOMMENTS:

INITIATING CUES: 1. Your task is to perform the following: a.Classify the event. b.Complete the required INITIAL NOTIFICATION FORM. 2.Do not use SS Judgment as the basis for your classification. 3.The simulator will remain in freeze during the JPM.4.There are two Time Critical elements for this JPM; Time ofEvent Classificationand completion of the Notification Form.Inform the Evaluator at the completion ofeach Time Criticalelement.5.For the purposes of this JPM, initial event time zero and the starttime forclassification will begin when the Evaluator tells you to begin.6.Reactor Trip time will be 30 minutes prior to T-0 announced by theRecord all work on the NJPA-1003 HandoutINITIAL CONDITION:1.The plant was in MODE 1.2.Weekend night shift, only routine evolutions in progress.

3.The 'C' Charging pump was tagged out for maintenance on theprevious shift. No other equipment is out of service.4.An event occurred resulting in an automatic Reactor Trip and SafetyInjection.5.The crew entered EOP-1.0, REACTOR TRIP OR SAFETY 6.The CRS has received the following reports from the crew.a."Reactor Trip."b."Turbine Trip."c."Safety Injection" d."ESFLS complete on Train A and Train B" e."The ' B' Charging pump failed to start"f."RCS pressure is less than 1418 psig with flow on FI-943,7.The crew transitioned from EOP-1.0 to "EOP-2.0, LOSS OF8.The CRS was subsequently notified that the 'A' Charging pump has9.The crew is taking action using EOP-2.0, at step 13.10.The simulator was frozen at this point.SAFETY CONSIDERATIONS:NoneOPERATOR INSTRUCTIONS:JPM BRIEFING SHEETHAND THIS PAPER BACKTO YOUR EVALUATOR WHEN YOU FEEL THAT YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

EPP-002 ATTACHMENT I PAGE 1 of 11 REVISION 36 NUCLEAR POWER PLANT EMERGENCY NOTIFICATION FORM 1.A DRILLB ACTUAL EVENT MESSAGE # _______ 2.A INITIALB FOLLOW-UP NOTIFICATION: TIME___________ DATE_____/_____/_____ AUTHENTICATION # ________ 3.SITE: V. C. SummerConfirmation Phone # (____)_________________ 4.EMERGENCY CLASSIFICATION:A UNUSUAL EVENT B ALERT C SITE AREA EMERGENCY D GENERAL EMERGENCYBASED ON EAL #______________ EAL DESCRIPTION: ______________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________________5.PROTECTIVE ACTION RECOMMENDATIONS:A NONE B EVACUATE ___________________________________________________________________________________________________ C SHELTER _____________________________________________________________________________________________________ D CONSIDER THE USE OF KI (POTASSIUM IODIDE) IN ACCORDANCE WITH STATE PLANS AND POLICY. E OTHER________________________________________________________________________________________________________ 6.EMERGENCY RELEASE:A None B Is Occurring C Has Occurred 7.RELEASE SIGNIFICANCE:A Not applicable B Within normal operatinglimits C Above normal operating limits D Under evaluation 8.EVENT PROGNOSIS:A Improving B Stable C Degrading 9.METEOROLOGICAL DATA:Wind Direction* from _______ degrees Wind Speed* _______mph (*May not be available for Initial Notifications) Precipitation* _______ Stability Class* A B C D E F G 10.A DECLARATIONB TERMINATION Time ________________ Date _____/______/_______ 11.AFFECTED UNIT(S): 2 3 All 12.UNIT STATUS: (Unaffected Unit(s) Status Not Required for Initial Notifications) U1 _____% Power Shutdown at Time _____________ Date ___/_____/____ B U2 _____% Power Shutdown at Time _____________ Date ___/_____/____ C U3 _____% Power Shutdown at Time _____________ Date ___/_____/____ 13.REMARKS: _________________________________________________________________________________________________________________________________________________________________________________________________________________________________________ ____________________________________________________________________________________________________________________________ FOLLOW-UP INFORMATION (Lines 14 through 16 Not Required for Initial Notifications) EMERGENCY RELEASE DATA. NOT REQUIRED IF LINE 6 A IS SELECTED. 14.RELEASE CHARACTERIZATION:TYPE: A Elevated B Mixed C Ground UNITS: A Ci B Ci/sec C Ci/sec MAGNITUDE: Noble Gases:__________ Iodines:___________ Particulates:__________ Other: ____________ FORM: A Airborne B Liquid Start Time __________ Date ___/_____/____Stop Time _________ Date ___/_____/____ Start Time __________ Date ___/_____/____Stop Time _________ Date ___/_____/____ 15.PROJECTION PARAMETERS:Projection period: ________Hours Estimated Release Duration ________Hours Projection performed: Time _________ Date ___/_____/____ 16.PROJECTED DOSE:DISTANCE TEDE (mrem) Adult Thyroid CDE (mrem) Site boundary 2 Miles 5 Miles 10 Miles 17.APPROVEDBY: ____________________________ Title _____________________ Time _________ Date___/_____/____ NOTIFIED BY:___________________________ RECEIVED BY: _____________________________ Time _________ Date ___/_____/____ A 1 X EPP-002 ATTACHMENT I PAGE 1 of 11 REVISION 3 NUCLEAR POWER PLANT EMERGENCY NOTIFICATION FORM 1.A DRILLB ACTUAL EVENT MESSAGE # 2. A INITIALB FOLLOW-UP NOTIFICATION: TIME DATE / / AUTHENTICATION #3.SITE: V. C. Summer Confirmation Phone # ( 4.EMERGENCY CLASSIFICATION:A UNUSUAL EVENT B ALERT C SITE AREA EMERGENCYD GENERAL EMERGENCY BASED ON EAL # EAL DESCRIPTION: 5.PROTECTIVE ACTION RECOMMENDATIONS:A NONE B EVACUATE C SHELTER D CONSIDER THE USE OF KI (POTASSIUM IODIDE) IN ACCORDANCE WITH STATE PLANS AND POLICY. E OTHER 6.EMERGENCY RELEASE:A NoneB Is Occurring C Has Occurred 7.RELEASE SIGNIFICANCE:A Not applicable B Within normal operatinglimits C Above normal operating limits D Under evaluation 8.EVENT PROGNOSIS:A Improving B Stable Degrading9.METEOROLOGICAL DATA:Wind Direction* from degreesWind Speed* mph(*May not be available for Initial Notifications) Precipitation* Stability Class* A B C D E F G10. A DECLARATIONB TERMINATION Date / / 11.AFFECTED UNIT(S):12.UNIT STATUS:1 2 3 All (Unaffected Unit(s) Status Not Required for Initial Notifications) U1  % Power Shutdown at TimeB U2 % Power Shutdown at Time C U3 % Power Shutdown at Time 13.REMARKS:FOLLOW-UP INFORMATION (Lines 14 through 16 Not Required for Initial Notifications) EMERGENCY RELEASE DATA. NOT REQUIRED IF LINE 6 A IS SELECTED. 14.RELEASE CHARACTERIZATION:TYPE: A Elevated B Mixed C Ground UNITS: A Ci B Ci/sec C µCi/sec MAGNITUDE: Noble Gases: Iodines: Particulates: Other: FORM: A Airborne Start Time Date / / Stop Time Date / / B Liquid Start Time Date / / Stop Time Date / / 15.PROJECTION PARAMETERS:Projection period: Hours Estimated Release Duration Hours Projection performed: Time Date / _/ 16.PROJECTED DOSE:DISTANCE TEDE (mrem) Adult Thyroid CDE (mrem) Site boundary 2 Miles 5 Miles 10 Miles 17.APPROVEDBY: _ Title _ Time Date / / NOTIFIED BY: RECEIVED BY: Time Date / /