ML17181A386

Exam Draft Items 2A - Delay Release

ML17181A386
Person / Time
Site:	Brunswick
Issue date:	06/30/2017
From:	NRC/RGN-II
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Download: ML17181A386 (843)

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{{#Wiki_filter:ES-401 BWR Examination Outline Form ES-401-1 Facility: Brunswick Date of Exam: December, 2016 Tier Group RO K/A Category Points SRO-Only Points K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 3 3 4 4 3 3 20 4 3 7 Emergency & N/A N/A 2 1 1 1 2 1 1 7 2 1 3 Abnormal Plant Evolutions Tier Totals 4 4 5 6 4 4 27 6 4 10 1 3 2 3 2 3 2 2 3 2 2 2 26 3 2 5 2.

Plant 2 1 1 1 1 1 2 1 1 1 1 1 12 0 2 1 3 Systems Tier Totals 4 3 4 3 4 4 3 4 3 3 3 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 3 2 2 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category.)

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

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

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

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

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

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

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

G* Generic K/As

ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO) E/APE # / Name / Safety Function K K K A A2 G* K/A Topic(s) IR # 1 2 3 1 G2.2.12; Knowledge of surveillance 295001 Partial or Complete Loss of Forced X procedures. 3.7 Core Flow Circulation / 1 & 4 AA2.05; Ability to determine and/or interpret X the following as they apply to PARTIAL OR 3.4 COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Jet pump operability. AK1.03; Knowledge of the operational 295003 Partial or Complete Loss of AC / 6 X implications of the following concepts as they 2.9 apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Under voltage/degraded voltage effects on electrical loads. AK2.01; Knowledge of the interrelations 295004 Partial or Total Loss of DC Pwr / 6 X between PARTIAL OR COMPLETE LOSS OF 3.1 D.C. POWER and the following: Battery charger. AK3.04; Knowledge of the reasons for the 295005 Main Turbine Generator Trip / 3 X following responses as they apply to MAIN 3.2 TURBINE GENERATOR TRIP: Main generator trip. AA1.06; Ability to operate and/or monitor the 295006 SCRAM / 1 X following as they apply to SCRAM: CRD 3.5 hydraulic system. AK3.03; Knowledge of the reasons for the 295016 Control Room Abandonment / 7 X following responses as they apply to 3.5 CONTROL ROOM ABANDONMENT: Disabling control room controls. AK2.02; Knowledge of the interrelations 295018 Partial or Total Loss of CCW / 8 X between PARTIAL OR COMPLETE LOSS OF 3.4 COMPONENT COOLING WATER and the following: Plant operations. AA2.02; Ability to determine and/or interpret 295019 Partial or Total Loss of Inst. Air / 8 X the following as they apply to PARTIAL OR 3.6 COMPLETE LOSS OF INSTRUMENT AIR: Status of safety-related instrument air system loads. AK2.05; Knowledge of the interrelations 295021 Loss of Shutdown Cooling / 4 X between LOSS OF SHUTDOWN COOLING 2.5 and the following: Fuel pool cooling and cleanup system. AA2.03; Ability to determine and/or interpret 3.5 the following as they apply to LOSS OF X SHUTDOWN COOLING : Reactor water level. AA1.04; Ability to operate and/or monitor the 295023 Refueling Acc / 8 X following as they apply to REFUELING 3.4 ACCIDENTS: Radiation monitoring equipment. G2.2.25; Knowledge of the bases in Technical X Specifications for limiting conditions for 4.2 operations and safety limits. EA1.05; Ability to operate and/or monitor the 295024 High Drywell Pressure / 5 X following as they apply to HIGH DRYWELL 3.9 PRESSURE: RPS. EK3.02; Knowledge of the reasons for the 295025 High Reactor Pressure / 3 X following responses as they apply to HIGH 3.9 REACTOR PRESSURE: Recirculation pump trip: Plant-Specific.

G2.4.50; Ability to verify system alarm 295026 Suppression Pool High Water X setpoints and operate controls identified in the 2.7 Temp. / 5 alarm response manual. G2.1.23; Ability to perform specific system and X integrated plant procedures during all modes of 4.4 plant operation. 295027 High Containment Temperature / 5 EA2.01; Ability to determine and/or interpret 295028 High Drywell Temperature / 5 X the following as they apply to HIGH DRYWELL 4.0 TEMPERATURE: Drywell temperature. G2.4.50; Ability to verify system alarm 295030 Low Suppression Pool Wtr Lvl / 5 X setpoints and operate controls identified in the 4.2 alarm response manual. EK1.03; Knowledge of the operational 295031 Reactor Low Water Level / 2 X implications of the following concepts as they 3.7 apply to REACTOR LOW WATER LEVEL: Water level effects on reactor power. EK1.03; Knowledge of the operational 295037 SCRAM Condition Present and X implications of the following concepts as they 4.2 Reactor Power Above APRM Downscale or apply to SCRAM CONDITION PRESENT AND Unknown / 1 REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Boron effects on reactor power (SBLC). G2.4.21; Knowledge of the parameters and logic used to assess the status of safety X 4.6 functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. EA1.01; Ability to operate and/or monitor the 295038 High Off-site Release Rate / 9 X following as they apply to HIGH OFF-SITE 3.9 RELEASE RATE: Stack-gas monitoring system: Plant-Specific. EA2.01; Ability to determine and/or interpret 4.3 the following as they apply to HIGH OFF-SITE X RELEASE RATE: Off-site. AK3.04; Knowledge of the reasons for the 600000 Plant Fire On Site / 8 X following responses as they apply to PLANT 2.8 FIRE ON SITE: Actions contained in the abnormal procedure for plant fire on site. AA2.07; Ability to determine and interpret the following as they apply to PLANT FIRE ON X SITE: Whether malfunction is due to common- 3.0 mode electrical failures. AA2.03; Ability to determine and/or interpret 700000 Generator Voltage and Electric Grid X the following as they apply to GENERATOR 3.5 Disturbances / 6 VOLTAGE AND ELECTRIC GRID DISTURBANCES: Generator current outside the capability curve. K/A Category Totals: 3 3 4 4 3/4 3/3 Group Point Total: 20/7

ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO) E/APE # / Name / Safety Function K K K A A2 G* K/A Topic(s) IR # 1 2 3 1 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 AK2.04; Knowledge of the interrelations 295009 Low Reactor Water Level / 2 X between LOW REACTOR WATER LEVEL 2.6 and the following: Reactor water cleanup. 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 295015 Incomplete SCRAM / 1 X G2.4.31; Knowledge of annunciator alarms, 4.1 indications, or response procedures. AA2.03; Ability to determine and/or interpret 295017 High Off-site Release Rate / 9 X the following as they apply to HIGH OFF- 3.1 SITE RELEASE RATE: Radiation levels: Plant-Specific. AA1.02; Ability to operate and/or monitor the 295020 Inadvertent Cont. Isolation / 5 & 7 X following as they apply to INADVERTENT 3.2 CONTAINMENT ISOLATION: Drywell ventilation/cooling system. 295022 Loss of CRD Pumps / 1 X AA2.02; Ability to determine and/or interpret 3.4 the following as they apply to LOSS OF CRD PUMPS : CRD system status EA1.01; Ability to operate and/or monitor the 295029 High Suppression Pool Wtr Lvl / 5 X following as they apply to HIGH 3.4 SUPPRESSION POOL WATER LEVEL: HPCI: Plant-Specific. EK3.01; Knowledge of the reasons for the 295032 High Secondary Containment Area X following responses as they apply to HIGH 3.5 Temperature / 5 SECONDARY CONTAINMENT AREA TEMPERATURE: Emergency/normal depressurization. 295033 High Secondary Containment Area Radiation Levels / 9 G2.4.8; Knowledge of how abnormal 295034 Secondary Containment X operating procedures are used in conjunction 3.8 Ventilation High Radiation / 9 with EOPs. EA2.01; Ability to determine and/or interpret 295035 Secondary Containment High X the following as they apply to SECONDARY 3.9 Differential Pressure / 5 CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Secondary containment pressure: Plant-Specific. EK1.02; Knowledge of the operational 295036 Secondary Containment High X implications of the following concepts as they 2.6 Sump/Area Water Level / 5 apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Electrical ground/ circuit malfunction. 500000 High CTMT Hydrogen Conc. / 5 K/A Category Point Totals: 1 1 1 2 1/2 1/1 Group Point Total: 7/3

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO) System # / Name K K K K K K A A2 A A G* K/A Topic(s) IR # 1 2 3 4 5 6 1 3 4 A4.04; Ability to manually 203000 RHR/LPCI: Injection Mode X operate and/or monitor in the 3.6 control room: Heat exchanger cooling flow. K5.03; Knowledge of the 205000 Shutdown Cooling X operational implications of the 2.8 following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): Heat removal mechanisms. K1.10; Knowledge of the physical 206000 HPCI X connections and/or cause/effect 3.4 relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the following: Condensate storage and transfer system. 207000 Isolation (Emergency) Condenser K3.03; Knowledge of the effect 209001 LPCS X that a loss or malfunction of the 2.9 LOW PRESSURE CORE SPRAY SYSTEM will have on following: Emergency generators. X G2.4.35; Knowledge of local auxiliary operator tasks during an 4.0 emergency and the resultant operational effects. 209002 HPCS K1.01; Knowledge of the physical 211000 SLC X connections and/or cause/effect 3.0 relationships between STANDBY LIQUID CONTROL SYSTEM and the following: Core spray line break detection: Plant-Specific. K2.01; Knowledge of electrical 212000 RPS X power supplies to the following: 3.2 RPS motor-generator sets. G2.2.44; Ability to interpret X control room indications to verify 4.4 the status and l operation of a system, and understand how operator actions and directives affect plant and system conditions.

A2.06; Ability to (a) predict the 215003 IRM X impacts of the following on 3.0 the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty range switch. K6.04; Knowledge of the effect X that a loss or malfunction of the 4.0 following will have on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM : Detectors. K5.03; Knowledge of the 215004 Source Range Monitor X operational implications of the 2.8 following concepts as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM: Changing detector position. K2.02; Knowledge of electrical 215005 APRM / LPRM X power supplies to the following: 2.6 APRM channels. K5.04; Knowledge of the X operational implications of the 2.9 following concepts as they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: LPRM detector location and core symmetry. A1.01; Ability to predict and/or 217000 RCIC X monitor changes in parameters 3.7 associated with operating the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) controls including: RCIC flow. K1.03; Knowledge of the physical 218000 ADS X connections and/or cause/effect 3.7 relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following: Nuclear boiler instrument system. X K3.01; Knowledge of the effect 4.4 that a loss or malfunction of the AUTOMATIC DEPRESSURIZATION SYSTEM will have on following: Restoration of reactor water level after a break that does not depressurize the reactor when required. A1.01; Ability to predict and/or 223002 PCIS/Nuclear Steam X monitor changes in parameters 3.5 Supply Shutoff associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including: System indicating lights and alarms.

K4.03; Knowledge of 239002 SRVs X RELIEF/SAFETY VALVES 3.1 design feature(s) and/or interlocks which provide for the following: Prevents siphoning of water into SRV discharge piping and limits loads on subsequent actuation of SRV's. X A2.01; Ability to (a) predict the impacts of the following on the 3.3 RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open vacuum breakers. A3.08; Ability to monitor 259002 Reactor Water Level X automatic operations of the 4.0 Control REACTOR WATER LEVEL CONTROL SYSTEM including: FWCI system initiation: FWCI. A4.02; Ability to manually 261000 SGTS X operate and/or monitor in the 3.1 control room: Suction valves. X A2.09; Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT 2.6 SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Plant air system failure. 262001 AC Electrical Distribution X G2.2.40; Ability to apply 3.4 Technical Specifications for a system. X A2.01; Ability to (a) predict the 3.6 impacts of the following on the A.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Turbine/generator trip. A3.01; Ability to monitor 262002 UPS (AC/DC) X automatic operations of the 2.8 UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including: Transfer from preferred to alternate source. G2.2.37; Ability to determine 263000 DC Electrical Distribution X operability and/or availability of 3.6 safety related equipment. K6.03; Knowledge of the effect 264000 EDGs X that a loss or malfunction of the 3.5 following will have on the EMERGENCY GENERATORS (DIESEL/JET) : Lube oil pumps.

A2.01; Ability to (a) predict the 300000 Instrument Air X impacts of the following on the 2.9 INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Air dryer and filter malfunctions. X K3.01; Knowledge of the effect 2.7 that a loss or malfunction of the Containment air system. A2.01; Ability to (a) predict the 400000 Component Cooling Water X impacts of the following on the 3.3 CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Loss of CCW pump. X K4.01; Knowledge of CCWS design feature(s) and or 3.4 interlocks which provide for the following: Automatic start of standby pump. K/A Category Point Totals: 3 2 3 2 3 2 2 3/3 2 2 2/2 Group Point Total: 26/5

ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / SRO) System # / Name K K K K K K A A2 A A G* K/A Topic(s) IR # 1 2 3 4 5 6 1 3 4 K6.02; Knowledge of the effect that a 201001 CRD Hydraulic X loss or malfunction of the following 3.0 will have on the CONTROL ROD DRIVE HYDRAULIC System: Condensate storage tanks. 201002 RMCS G2.4.49; Ability to perform without 201003 Control Rod and Drive X reference to procedures those 4.6 Mechanism actions that require immediate operation of system components and controls. 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation K6.03; Knowledge of the effect that a 202002 Recirculation Flow Control X loss or malfunction of the following 2.8 will have on the RECIRCULATION FLOW CONTROL SYSTEM: Recirculation system. 204000 RWCU 214000 RPIS 215001 Traversing In-Core Probe X G2.2.44; Ability to interpret control 4.4 room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. K1.02; Knowledge of the physical 215002 RBM X connections and/or cause/effect 3.2 relationships between ROD BLOCK MONITOR SYSTEM and the following: LPRM. A2.11; Ability to (a) predict the 216000 Nuclear Boiler Inst. X impacts of the following on the 3.2 NUCLEAR BOILER INSTRUMENTATION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Heatup or cooldown of the reactor vessel. A2.12; Ability to (a) predict the 219000 RHR/LPCI: Torus/Pool Cooling X impacts of the following on the 3.1 Mode RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve logic failure: Plant-Specific K3.09; Knowledge of the effect that a 223001 Primary CTMT and Aux. X loss or malfunction of the PRIMARY 2.8 CONTAINMENT SYSTEM AND AUXILIARIES will have on following: Nuclear boiler instrumentation.

226001 RHR/LPCI: CTMT Spray Mode X K2.02; Knowledge of the physical 2.9 connections and/or cause/effect relationships between RHR/LPCI: CONTAINMENTSPRAY SYSTEM MODE and the following: Pumps. 230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup A3.01; Ability to monitor automatic 234000 Fuel Handling Equipment X operations of the FUEL HANDLING 2.6 EQUIPMENT including: Crane/refuel bridge movement: Plant-Specific. 239001 Main and Reheat Steam 239003 MSIV Leakage Control A4.14; Ability to manually operate 241000 Reactor/Turbine Pressure X and/or monitor in the control room: 3.8 Regulator Turbine trip. K4.07; Knowledge of MAIN 245000 Main Turbine Gen. / Aux. X TURBINE GENERATOR AND 2.5 AUXILIARY SYSTEMS design feature(s) and/or interlocks which provide for the following: Generator voltage regulation. 256000 Reactor Condensate A1.04; Ability to predict and/or 259001 Reactor Feedwater X monitor changes in parameters 2.8 associated with operating the REACTOR FEEDWATER SYSTEM controls including: RFP turbine speed: Turbine-Driven-Only. 268000 Radwaste A2.10; Ability to (a) predict the 271000 Offgas X impacts of the following on the 3.3 OFFGAS SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Offgas system high flow. K5.01; Knowledge of the operational 272000 Radiation Monitoring X implications of the following concepts 3.2 as they apply to RADIATION MONITORING SYSTEM: Hydrogen injection operation's effect on process radiation indications: Plant-Specific. 286000 Fire Protection 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel Internals K/A Category Point Totals: 1 1 1 1 1 2 1 1/2 1 1 1/1 Group Point Total: 12/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Brunswick Date of Exam: December, 2016 Category K/A # Topic RO SRO-Only IR # IR # 2.1.1 Knowledge of conduct of operations requirements. 3.8 2.1.32 Ability to explain and apply system limits and precautions. 3.8 2.1.36 Knowledge of procedures and limitations involved in core 3.0

1. alterations.

Conduct of Ability to use procedures related to shift staffing, such as Operations 2.1.5 3.9 minimum crew complement, overtime limitations, etc. Ability to use procedures to determine the effects on 2.1.43 4.3 reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc. Subtotal Ability to manipulate the console controls as required to 2.2.2 4.6 operate the facility between shutdown and designated power levels. (multi-unit license) Ability to explain the variations in 2.2.4 3.6 control l board/control room layouts, systems, instrumentation, and procedural actions between units at a facility. Ability to interpret control room indications to verify the

2. 2.2.44 4.2 status and operation of a system, and understand how Equipment operator actions and directives affect plant and system Control conditions.

Ability to determine the expected plant configuration using 2.2.15 4.3 design and configuration control documentation, such as drawings, line-ups, tag-outs, etc. l 2.2.22 Knowledge of limiting conditions for operations and safety 4.7 limits. Subtotal Knowledge of radiological safety principles pertaining to 2.3.12 3.2 licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. 3. Knowledge of radiation monitoring systems, such as fixed Radiation 2.3.15 2.9 radiation monitors and alarms, portable survey Control instruments, personnel monitoring equipment, etc. 2.3.11 Ability to control radiation releases. 4.3 Subtotal Knowledge of the operational implications of EOP 2.4.20 3.8 warnings, cautions, and notes. 2.4.27 Knowledge of fire in the plant procedures. 3.4

4. Knowledge of events related to system operation/status Emergency 2.4.30 4.1 that must be reported to internal organizations or external Procedures / agencies, such as the State, the NRC, or the Plan transmission system operator.

Knowledge of local auxiliary operator tasks during an 2.4.35 4.0 emergency and the resultant operational effects. Subtotal Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 1/1 295024EA1.14 Too similar to 295020AA1.02. Reselected 295024EA1.05 as a replacement. 2/1 259002A3.05 Not representative of how plant operates. Reselected 259002A3.08 as a replacement. 2/1 262002K6.03 Too similar to 26002A3,01. Reselected 215003K6.04 1/1 295028G2.19 Cannot write at SRO level. Replaced by 295023G2.2.25. 1/1 700000AA2.08 Cannot write at SRO level. Replaced by 295021AA2.03. 1/1 295024G2.2.37 Cannot write at SRO level. Replaced by 295037G2.4.21. 2/1 217000A2.09 Cannot write operationally valid question. Replaced with 262001A2.01. 2/2 256000A2.17 Not in plant design. Reselected 219000A2.12. 2/2 268000A2.01 Unable to write a operationally valid and discriminating question. Reselected 216000 A2.11 2/2 2860000K2.03 Unable to write a discerning question. Reselected to 2260001K2.02. 1/2 295013.A2.02 Unable to write to the K/A. The plant doesnt really monitor stratification. Reselected with 295022A.2.02.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Brunswick Date of Examination: 11/28/2016 Examination Level: RO SRO Operating lest Number: Draft Administrative Topic (see Note) Type Describe activity to be performed Code* Conduct of Operations #1 R, N 2.1.25 Perform SJAE Off-Gas Radiation (RO, then SRO) Monitors Channel Check Calculation Conduct of Operations #2 R, D 2.1 .07 Determine Primary Containment Water (All) Level and Evaluate PCPL-A. Equipment Control R, N 2.2.12 Calculate Drywell Leakage Rate. (RO, then SRO) Radiation Control R, D 2.3.07 Determine Stay Time Limitations in (All) High Radiation Areas. Emergency Plan R, M 2.4.29 Classify An Emergency per PEP-02.1. (SRO Only) NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items). 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)

Conduct of Operations #1 Perform SJAE Off-Gas Radiation Monitors Channel Check Calculation RO, then SRO 2.01 .25 Ability to interpret reference materials, such as graphs, curves, tables, etc. 3.9/4.2 This is a new JPM developed for the 2016 NRC Initial Exam. The examinee will perform item 108, SJAE Off-Gas Radiation Monitors Channel Check, of 201-03.2, Reactor Operator Daily Surveillance Report, and state the status of the channel check. Then the SRO examinees will determine any required actions. Conduct of Operations #2 Determine Primary Containment Water Level and Evaluate PCPL-A All 2.01.07 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. 4.4/4.7 This is a bank JPM. The examinee will determine Primary Containment water level per EOP-01-UG, Attachment 11. Determine the current region of operation (Safe/Unsafe) on Primary Containment Pressure Limit A (PCPL-A). Equipment Control Calculate Drywell Leakage Rate RO, then SRO 2.2.12 Knowledge of surveillance procedures 3.7/4.1 This is a new JPM developed for the 2016 NRC Initial Exam. The examinee will determine the 24 hour leak rate for the equipment and floor drains, and the 24 hour total leak rate to the drywell lAW Attachment 1, Drywell Leakage Calculation, of 201-03.2, Reactor Operator Daily Surveillance Report, for Sunday Nightshift at time 2000.. Then the SRO examinees will determine if TS LCO 3.4.4 is met and if it is not met identify the latest time the Unit is required to be in MODE 3.

Radiation Control Determine Stay Time Limitations in High Radiation Areas. All 2.3.07 Ability to comply with radiation work permit requirements during normal and abnormal conditions. 3.5/3.6 This is a bank JPM. The examinee will determine the total dose accumulated for both workers and determine if any Brunswick administrative dose limitations will be exceeded. Emergency Plan Classify An Emergency per PEP-02.1 SRO Only 2.4.29 Knowledge of the Emergency Plan 3.1/4.4 This is a modified JPM that was used on the 2012 NRC Initial Exam. Changed EAL from Site Area Emergency to an Alert. The SRO examinees will determine the highest required classification and its EAL Identifier. This JPM is time critical (15 minutes).

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: Brunswick Date of Examination: 11/28/2016 Exam Level: RO SRO-l SRO-U Operating Test No.: Draft Control Room Systems: 8 for RO; 7 for SRO-l; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function

a. Reset Recirc Pump Runback, Both Recirc Pumps trip S, N, A 1

b. Mechanical Trip Valve Oil Trip Test S, N 4

c. RCIC Start wi failure to isolate 5, P, A, EN 2 U. Suppression Pool Cooing Service Water Release

                                   -                                               5, A, D, L               5

e. Vent Drywell w/ Stack Rad Mon >50% increase S, A 9

1. Shifting Caswell Beach Lube Water Pumps From The RTGB 5, N 8

g. Substitute Control Rod Position S, L 7

h. RO ONLY Test the Main Steam Isolation Valves

                 -                                                                 5, P                     3 In-Plant Systems (3 for RO); (3 for SRO-l); (3 or 2 for SRO-U)

i. LEP-Ol, Heater Drain Pumps R, D, E, L 2

j. LEP-05, SRV operation from RSDP N, R, E, L 7

k. Rack in E6 Crosstie Breaker A, D, E 6 All RD and SRO-l control room (and in-plant) systems must be different and serve different safety functions; all five 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 RD / S RD-I I S RD-U A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irectfrombank 918/4 (E)mergency or abnormal in-plant 1 I 1 / 1 (EN)gineered safety feature 1 / 1 / 1 (control room system) (L)ow-Power / Shutdown 1 I 1 I 1 (N)ew or (M)odified from bank including 1(A) 2/ 2 / 1 (P)revious 2 exams 3 I 3 / 2 (randomly selected) (R)CA 1/1I1 (S)imulator

a. Reset Recirc Pump Runback with both Recirc Pumps Tripping 202002 A2.01 Ability to predict the impacts of recirculation pump trip and based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations. 3.4/3.4 RO/ISRO/USRO This is a simulator alternate path JPM that will have the examinees reselling a Recirc Pump runback signal. When the runback is reset both recirc pumps will trip, this will require an immediate operator action to insert a manual reactor scram. This JPM is a new alternate path JPM.

b. Mechanical Trip Valve Oil Trip Test 245000 A3.01 Ability to manually operate and/or monitor in the control room:

Turbine Trip. 3.6/3.6 RO/ISRO/USRO This is a new simulator JPM that will require the examinee to perform the Mechanical Trip Valve Oil Trip Test.

c. RCIC Start Per The Hard Card Steam line break 217000 A4.08 Ability to manually operate and/or monitor RCIC system flow. 3.7/3.6 RO/ISRO/USRO This is a simulator alternate path JPM that will require the examinee to start RCIC for injection per the Hard Card and restore RPV water level. As an alternate path the steam line breaks and RCIC does not auto isolate requiring manual isolation of RCIC. RCIC is an engineered safety feature. This JPM was randomly selected from the previous exam (2015).

d. Suppression Pool Cooing Service Water Release 219000 A4.01 Ability to manually operate and/or monitor in the control room:

Pumps. 3.8/3.7 RO/ISRO This is a banked alternate path simulator JPM that will require the exam inee to place the B Loop of RHR in SPC. While placing in service the heat exchanger tubes will rupture causing a release in the service water system. The examinee will have to isolate the release path.

e. Vent Drywell wI Stack Rad Mon >50% increase 261000 A4.04 Ability to manually operate and/or monitor Primary Containment Pressure. 3.3/3.4 RO/ISRO This is a banked simulator JPM that will require the examinee to vent the Drywell via Standby Gas Treatment. This JPM is alternate path in that Main Stack Rad will rise requiring the examinee to isolate the system.

f. Shifting Caswell Beach Lube Water Pumps From The RTGB 400000 A4.01 Ability to manually operate and/or monitor in the control room: CCW indications and control. 3.1/3.0 RO/ISRO This is a new simulator JPM that will require the examinee to perform shift Caswefl Beach Lube Water Pumps From The RTGB.

g. Substitute Control Rod Position 201006 A4.06 Ability to manually operate and/or monitor in the control room:

Selected rod position indication. 3.2/3.2 RO/ISRO This is a banked ]PM that will require the examinee to substitute in the Rod Worth Minimizer the indicated rod position.

h. Test the Main Steam Isolation Valves 239001 A4.01 Ability to manually operate and/or monitor the MSIVs in the Control Room. 4.2/4.0 RO This is a new JPM that will require the examinee to perform post-maintenance testing of a MSIV. This JPM was randomly selected from the previous exam (2015).

i. LEP-Ol, Heater Drain Pumps 295031 EA1 .08 Ability to operate alternate injection system systems as they apply to Reactor Water Level Low. 3.8/3.9 RO/ISRO This is a banked in-plant JPM that will require the examinee to simulate the Auxiliary Operator actions for Alternate Coolant Injection, Heater Drain Pump Injection per OEOP-01-LEP-01. This JPM is performed in the RCA.

j. LEP-05, SRV operation from RSDP 295016 AA1 .08 Ability to operate and/or monitor Reactor Pressure as it applies to Control Room Abandonment. 4.0/4.0 RO/ISRO/USRO This is a new in-plant JPM that will require the examinee to simulate the actions associated with performing the field actions for pressure control from the RSDP (Remote Shutdown Panel)

k. Rack in E6 Crosstie Breaker 295003 AA1 .01 Ability to operate and/or monitor AC Electrical Distribution System as it applies to a partial or complete loss of A.C. power. 3.7/3.8 RO/lSRO/USRO This is a banked in-plant alternate path JPM that will require the examinee to simulate manually racking in the crosstie breaker. The charging springs on the breaker will not automatically re-charge and will have to be manually charged.

1. 201001 1 Unit One is in an outage with the condensate system under clearance.

An earthquake results in damage to the CST causing level to slowly lower. Which one of the following completes the statement below with regards to the effect on the CRD system? The CRD system will (1) when the CST level reaches approximately (2) A (1) trip (2) 3 feet B (1) trip (2) 11 feet C (1) transfer to the backup supply (2) 3 feet D (1) transfer to the backup supply (2) 11 feet Answer: B K/A: 201001 Control Rod Drive Hydraulic System K6 Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROD DRIVE HYDRAULIC System: (CFR: 41.7 / 45.7) 02 Condensate storage tanks RO/SRO Rating: 3.0/3.1 Tier 2 / Group 2 K/A Match: This meets the K/A because the student has determine the effect of the loss of the CST on the CRD system. Pedigree: New Objective: LOI-CLS-LP-008, Obj. 8 Given plant conditions, predict the effect that a loss or malfunction of the following will have on the CRDH System: b. Condensate Storage Tank

Reference:

None Cog Level: High Explanation: Under normal system operations the CRD system suction is from the condensate system. The alternate supply is from the CST, which will transfer automatically. With the condensate system under clearance these valves would be isolated. The stand pipe for the CRD suction is at 11 feet. The auto transfer for the suction for ECCS is at 3 feet.

Distractor Analysis: Choice A: Plausible because the pumps will lose NPSH and trip but the suction is at 11 feet not 3 feet. Choice B: Correct Answer, see explanation. Choice C: Plausible because an auto transfer to the CST would occur but in this case an auto transfer to the condensate system is not possible. 3 feet is the suction height for the ECCS system. Choice D: Plausible because an auto transfer to the CST would occur but in this case an auto transfer to the condensate system is not possible. The second part is correct. SRO Basis: N/A CONTROL ROD DRIVE HYDRAULIC SYSTEM 1 OP-08 OPERTING PROCEDURE Rev. 96 Page 6 of 377 3.0 PRECAUTIONS AND LIMITATIONS

1. This procedure is Reactivity Management related pet AD-OP-ALLO2O3, Reactivity Management. Those por ions of this procedure that move control rods in MODES 1 012 are considered a Direct Reactivity manipulation and Reactivity Evolution Category P2 (Reactivity Manipulation, R2) fl

2. CST level is maintained greater than I 1 feet to prevent CR0 pumps from losing suction D

2. 201003 1 Unit Two is operating at rated power when a control rod begins to drift out from position 24.

Which one of the following identifies the first action to be taken by the operator at the controls (OATC)? A Initiate a single rod scram. B Initiate a manual reactor scram. C Select and fully insert the control rod to position 00. O Select and attempt to arrest the control rod at position 24. Answer: D K/A: 201003 Control Rod and Drive Mechanism G2.4.49 Ability to perform without reference to procedures those actions that requite immediate operation of system components and controls. (CFR: 41.10 /43.2/45.6) RO/SRO Rating: 4.6/4.4 Tier 2 I Group 2 K/A Match: This meets the K/A because the question is testing the operator action required to control a drifting control rod (Chief Examiner agreed that operation of the RMCS for a rod drift would meet this K/A) Pedigree: New Objective: LOI-CLS-LP-07. Obj. lib Describe the possible cause(s) and required operator actions for the following alarms: A-S 3-2. Control Rod Drift

Reference:

None Cog Level: Fundamental Explanation: This abnormal positive reactivity addition requires response from the APP before entering the AOP. The APP requires that the operator attempt to arrest the drift at the intended position first, if it cannot be arrested but responds to RMCS to insert to 00, if it does not respond to RMCS to perform a single rod scram. If more than 1 rod drifts then a manual scram is required. Distractor Analysis: Choice A: Plausible because if the rod does not move then this is the appropriate action. Choice B: Plausible because if more than 1 rod is drifting then this would be correct Choice C: Plausible because this is the correct action if the rod is drifting in or the rod continues to drift after attempting to arrest. Choice D: Correct Answer, see explanation. SRO Basis: N/A

A2 A:; 52 54e I of I Ac: DPJF1 AUTO Afl0NS I. PJch wothiraw or unsert err:rs rossobiw retain; r:d blork of reactor p:wer us below the low mower settoint The reactor power woll restond t: the iroftin; r:d depending upon the dorert:;n of the roi droft and rod worth, and could result on a reactor E:ram of the plant os at low power oceratton. CAUSE

1. P.od in uneven position doe to:

a. leakonc Scram valve.

b. Hugh c::ling water pressure.

r. Faulure of durerti:nal control valves.

6. Slow to settle due to fuel bundle channel bow.

2 - XalfL:nrtucn in alarm curcuit. 03 SERVATIONS I. 2:d rift uniucation on the full core dosplay.

2. atc:-: error indiraticns and Rod El: Sc if reactor p:wer us below the tower setcount.

        -  A change in neutron monut:ring sUstem ne-ocr readings as a result if the drifting rod with possuble high fluz alar:os
     .     :f               roi us selected, the f:urroi group dus;lay wull indicate an odd control rod posztocn, a blank window, :r a changing control rod positu:n ir. the directuon if the druft.

S. Hugh control rod o:olin; water pressure and or flow. E If control rod drufts to the full in position, a green backiught on the full core dustlay wulI illunnate wuth no position readout on RTG2. 7 ADO CUT ELOCE alarm AOS )22) and no wuthdraw perw.ussuve licht. Greater than nornal settle tines causun; an odd :r noposition to be present when the RNCS tuw.er tunes cut ACTIONS eterw,u ne uf the affected control rcd(s is druftung or uf the rods) has scrammed using full core dusplaw, RflS, and RW&

2. Select the criftung rod and deternune doreotion of druft

a. Attenpt t: arrest the tuft and latch rcd by zerf:rning the following:

I) Atoly am;r:cruate unaert or wothdrawal sugnals to the rod usung c:s.

2) If ANN is causung rod blocks, then bypass ANN uf directed by Unit CR3.

I1E D:tinud If h. r:i ::ninu to FI Offl th prf:a oh f:11otng: CA3TIi3N

ottro1 rod colleut pistor scuc in the w hdre un1tched: c.1oI:n will allow :h :od to drift full out due to .tt5 own weiqbt when iT.erE pr5ure 1 reno-ed erher by the 1CS or by c1otng lye CLCll e T:ttfy ecror :reeo.

b. :ro: :ore t meters e +/-ne on nttor nd offgas :tivtty.

c. If rod reetonie to en RiOS nseot sunel then fully insert the rod -it o cion ci. If rod feils to letch e costo:n CI. then reatrl :nserc to drive the rod fTO1.. in I! rod fa;ls to rest:nd to IE, ti-sn initoete e songe

ntr:i rod soren.

c. efer cc Iethnl Ete:ofocec:n

       -. If rod contInues cc DJFT 21C, then trforn the foliow.nci:

e cply en insert sognel end fully insert tod cc p:sitlcT.

CONTROL ROD MALFUNCTION/MI SF0 SITION DAOP-02 .0 Rev. 28 Page 6 of 25 3.0 AUTOMATIC ACTIONS

1. Possible rod block or select block from a fa4ed reed switch or a loss of power

2. CRD pumps trip after a 3 second delay on tow suction pressure 4.0 OPERATOR ACTIONS NOTE The following should be considered for establishment as crWcal parameters during performance ot this procedure

• Reactor po.er
• Control rod position
• Thermal limits 4.1 Immediate Actions

1. Stop any power changes in progress D NOTE Detected control rod motion without a wittidraw or insert command will cause annunciator A-05 3-2. Rod Drift, to alamt IF the annundator alarms AND NO blue scram light(s) are lit on the full core display, the conservative assumption is that rod(s) are drifting [1

2. IF more than one control rod is drifting, THEN insert a manual scram AND enter 1 EQ P-01 -RSP(2EOP-01 -RSP), Reactor Scram Procedure 0

3. 202002 1 Unit One is at rated power.

Which one of the following identifies the impact of inadvertently closing the 1 A Reactor Recirculation Pump 1-B32-FO3IA, Pump A Disch Vlv? The 1A Reactor Recirculation pump speed will lower to approximately: A20% B 34% C 45.4% ID 48% Answer: B KJA: 202002 Recirculation Flow Control System K6 Knowledge of the effect that a loss or malfunction of the following will have on the RECIRCULATION FLOW CONTROL SYSTEM: (CFR: 41.7/45.7) 03 Recirculation system RO/SRO Rating: 2.8/2.8 Tier 2 / Group 2 K/A Match: This meets the K/A because the student has to determine the effect of closing the discharge valve (which causes a loss of recirc) will have on the recirc flow control system. Pedigree: new Objective: LOl-CLS-LP-002.1, Obj. 17 Explain the operation of the following VFD limiters and controls: a. Limiter #1 b. Limiter #2

Reference:

None Cog Level: Fundamental Explanation: Closing of the discharge valve will cause the pump to runback to limiter #1(34%). Distractor Analysis: Choice A: Plausible because this is the minimum speed setting. Choice B: Correct Answer, see explanation. Choice C: Plausible because this is limiter #2 setting for Unit One Choice D: Plausible because this is limiter #2 setting for Unit Two SRO Basis: N/A

4. RFCS VFD Runback #1 Logc (Figure 02-1 SD and 02-1 BE)

The lcic for VFD A Runback # is shown on Figure 02-1 8D; the logic for VFD B Runback #1 shown on Figure 02-f BE is functionally identical to that for VFD A. The initiating conditions tot Runback #1 are

• Re drculation Rump A Discharge Valve B32-F03 1 A Urnlt Switch LS-2 opens (equivalent to Discharge Valve Not Full Open);
• Total Feedwater Flow as sensed b DFCS is less than 16.4% tot 15 seconds or more.

Unit 1 Spectic VFD Parameters Parameter Value Function 1170 98.9% (166 L5 rpm) VFD Over Speed Trip (Over Speed Alarm at 93.95% or 1578.4 rpm) 2080 92.5% (1554.0 rpm) Mail mum running motor speed (based upon achieving 104.5% Core Flo. 2120 45.4% (7621 rpm) Runback #2 Active Maximum Motor Speed 4250 50.8% (653.0 rpm) Manual Runback Motor Speed Low Limit Unit 2 Specific VFO Parameters Parameter Value Function 1 170 1031% VFD Over Speed Trip (1742.2rpm) (OverSpeedAiatfliat9s.5% or 1655.1 rpm) 2080 979% Maximum running motor speed (1614.7 rpm) (based upon achieving 104.5% Core Flow) 2120 48% (606.4 rpm) Runback #2 Active Maximum Motor Speed 4250 53.6% (900.5 rpm) Manual Runback Motor Speed Low Limit VFD Parameters Common to Both Units Parameter Value Function 2090 20% (336 rpm) Minimum Running Motor Speed 2100 34% (571.2 rpm) Runback #1 Active Maximum Motor Speed

4. 203000 1 A line break has occurred in the Unit Two drywell with the following sequence of events:

1200 Drywell pressure rises above 1.7 psig 1202 RPV pressure drops below 410 psig 1203 RPV level drops to LL3 Which one of the following completes the statement below? The earliest time that the operator can throttle the 2-El l-F048A, Loop 2A RHR Heat Exchanger Bypass Valve is at: A 1205. B 1206. C 1207. D 1208. Answer: A K/A: 203000 RHR / LPCI: Injection Mode A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) 04 Heat exchanger cooling flow RO/SRO Rating: 3.6/3.6 Tier 2 / Group 1 K/A Match: This meets the K/A because the student has to determine when the HX cooling flow can be operated. Pedigree: Bank Objective: LOl-CLS-LP-01 7, Obj. 09 Given an RHR pump or valve, list the interlocks, permissives and/or automatic actions associated with the RHR pump or valve, including setpoints.

Reference:

None Cog Level: High Explanation: The heat exchanger bypass valve has a 3 minute timer that starts on a LOCA signal. Drywell pressure greater than 1 .7# and reactor pressure is less than 41 0# is the first LOCA signal. The injection valve has a 5 minute interlock initiated by the same conditions. Another LOCA signal is introduced when reactor water level less than LL3 which provides the plausibility of the distractors.

Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because this is 3 minutes from the LL3 LOCA signal. Choice C: Plausible because this is 5 minutes from the low pressure setpoint which is time limit interlock for the injection valve. Choice D: Plausible because this is 5 minutes from the LL3 LOCA signal. SRO Basis: N/A After an initiation signal is received the following actions will occur:

• all tour RHR pumps will start 10 seconds afler power is available to the E-buses.
• Recirculation pumps are tripped via LL#2
• All valves not needed for LPCI injection automatically isolate and are interlocked shut as previously described.
• Heat exchanger bypass valve FO48NB opens and cannot be throttled for 3 minutes after an initiation signal is received. This ensures a discharge path ror the RHR pumps.
• Pemiissives sent to ADS as RHR pump pressure is sensed

     >   100 psig. Both pumps in either loop are required to satisfy the ADS permissive, or one core spray loop.

• Minimum flow valve opens if injection flow in loop is < 1000 gpm decreasing after a 10 sec time delay. It automatically shuts as injection valves open and injection flow raises to> 3000 gpm increasing.
• Reactor pressure decreases through the break and/or with actuation of ADS.
• As reactor pressure decreases to 4 10 psig. the LPCI injection valves FOl bA(S) auto open. The outboard injection valve FOl 7A(B) can be throttled 5 minutes after the RPV pressure is below 410 psig.
• As pressure reaches 310 psig, recirculation pump discharge and discharge bypass valves shut and are interlocked shut in the attempt to re-flood the core.
• As pressure reaches 200 psig, the RHR system injects into both recirculation system loops by lifting the check valves and overcoming reactor pressure.

5. 205000 1 RHR Loop 2A is operating in the Shutdown Cooling mode of operation with the following parameters:

RHRSW Pump 2A Operating RHRSW Flow 4000 gpm RHR Pump 2A Operating RHR Loop A Flow 6000 gpm Which one of the following completes the statement below? The required operator action to lower the cooldown rate lAW 20P-17, Residual Heat Removal System Operating Procedure, is to throttle open: A 2-El i-FOO3A, HX 2A Outlet Vlv. B 2-E11-FO17A, Outboard Injection Vlv. C 2-El 1-F048A, HX 2A Bypass Vlv. ID 2-Eli -PDV-F068A, HX 2A SW Disch Vlv. Answer: C K/A: 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) K5 Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): (CFR: 41.5/45.3) 03 Heat removal mechanisms RO/SRO Rating: 2.8/3.1 Tier 2 / Group 1 K/A Match: This meets the K/A because the student has to know which valve would need to be operated to control the heat removal for SDC. Pedigree: New Objective: LOI-CLS-LP-017, Obj. 15 Describe how the reactor cool down rate is controlled when the RHR system is in the Shutdown Cooling mode

Reference:

None Cog Level: High Explanation: The procedure allows throttling closed the F003 or F068 or throttling open the F048. Throttling open the F048 will bypass some of the RHR flow around the heat exchanger thereby lowering cooldown rate.

Distractor Analysis: Choice A: Plausible because if the valve was throttled closed this would be correct, Choice B: Plausible because if the valve was throttled closed this would be correct, although it is not an option that is allowed in the procedure. Choice C: Correct Answer, see explanation. Choice D: Plausible because if the valve was throttled closed this would be correct. SRO Basis: N/A 4 IF a lower cooldown rate is desired. THEN PERFORM the following, as necessary, for each operating RHR loop while maintaining desired flow rate, NOT to exceed 10,000 gpm per loop: CAUTION IF El 1-fOO3A(8? is closed, THEN RHR HEATEKCHANGER 2A2B) inlet temperature, located on E41-TR-R605 Point 1(2). is NOT a valid indication of reactor coolant temperature.

a. SLOWLY THROTTLE CLOSE HX 2A(2EU OUTLET VLV. E11-F003A(B), as necessary.

b. THROTTLE CLOSED HX2A26) SWD1SCH VLV EH-PDV-F068A(5). as necessary, to reduce RHRSW flow rate.

c. SLOWLY THROTTLE OPEN HX 24(28)

BYPASS VLV. Eli-F048A (8,1, as necessary. rnamtaining RHR flovTate greater than 400 gprn

6. 206000 1 A Group 1 isolation has occurred on Unit One.

HPCI has been placed in the pressure control mode of operation lAW 1 OP-I 9, High Pressure Coolant Injection System Operating Procedure. HPCI flow controller, E41-FIC-R600, is in manual with the output at midscale. Which one of the following completes the statement below? lithe 1-E41-F008, Bypass To CST Valve, is throttled (1) too far, this may result in HPCI (2) A (1) open (2) tripping on overspeed B (1) open (2) operation below 2100 rpm C (1) closed (2) tripping on overspeed D (I) closed (2) operation below 2100 rpm Answer: B K/A: 206000 High Pressure Coolant Injection System Ki Knowledge of the physical connections and/or cause-effect relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the following: (CFR: 41.2 to 41.9 /45.7 to 45.6) 10 Condensate storage and transfer system RO/SRO Rating: 2.8/3.1 Tier 2 / Group 1 K/A Match: This meets the K/A because this is testing the cause-effect relationship between HPCI and the flowpath to the CST. Pedigree: Bank Objective: LOl-CLS-LP-019. Obj. 8 Describe the methods available for controlling RPV pressure and/or RPV cooldown when operating the HPCI System in the Pressure Control mode. (LOCT)

Reference:

None Cog Level: high Explanation: Opening F008 will increase flow, causing turbine speed control to lower turbine speed to maintain desired flow. Opening valve too far can result in RPM below 2100 (OP-19, Section 8.2). Closing F008 will cause turbine speed to increase, but the governor limits turbine speed to a maximum value (4100 RPM) below the overspeed trip.

Distractor Analysis: Choice A: Plausible because opened is correct and an overspeed condition may be thought correct if the flowpath is considered incorrectly. Choice B: Correct Answer, see explanation. Choice C: Plausible because throttling closed will increase the speed of the turbine. Choice D: Plausible because may be thought correct if the flowpath is considered incorrectly. SRO Basis: N/A From OP-19: CAUTION Throttling E41-FOO8 open may cause turbne speed reductiion to tess than 2100 rpm, if opened too tar.. From the SD: Operation of the HPCI Turbine below the minimum rated speed of 2100 rpm may result in a failure of ttie auxiliary oil pump tram repeated startup cycles. A loss of the auxiliary oil pump wilt prevent starting of ie HPCI Turbine,

7. 209001 1 Unit Two is operating at rated power.

Due to a circuit malfunction an inadvertent LOCA initiation occurs in the Div II Core Spray logic causing A-03 (2-6), CORE SPRAY SYSTEM/I ACTUATED, to alarm. Which one of the following completes both statements below? Core Spray Pump(s) (1) will start. (2) will start. A (1) 2BONLY (2) All DGs B (1) 2BONLY (2) DG2 and DG4 ONLY C (1) 2Aand2B (2) All DGs ID (1) 2Aand2B (2) DG2 and DG4 ONLY Answer: A KJA: 209001 Low Pressure Core Spray System K3 Knowledge of the effect that a loss or malfunction of the LOW PRESSURE CORE SPRAY SYSTEM will have on following: (CFR: 41.7 / 45.4) 03 Emergency generators RO/SRO Rating: 2.9/3.0 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the knowledge of a malfunction of the CS logic has on the EDG. Pedigree: New Objective: LOl-CLS-LP-018, Obj. 14 List three systems, other than the Core Spray System, which are initiated or isolated by the Core Spray System logic.

Reference:

None Cog Level: High Explanation: For CS the logic will only start that divisions pump (RHR would start the other divisions pump) for the CS logic to the DGs either divisions signal will start all DGs.

Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because the first part is correct and since it is divisional for the pump starts the student may think that it would start only the Div II DGs. There ate signals that would start divisional DGs. Choice C: Plausible because the student may think the CS logic is similar to the RHR logic for pump starts and the second part is correct. Choice D: Plausible because the student may think the CS logic is similar to the RHR logic for pump starts and since it is a Div Illogic the student may think that it would start only the Div II DGs. There are signals that would start divisional DGs. SRO Basis: N/A I r ACS ag 1 :f 3 CE EP 215 21 ACTUATE Il E b:s 2E .S s.ça2

2. I E ; 2:r 2re.v zr. 23 23 Us1 g r&z:r  : 3

2. I tfl,, :T.. FIDW 1$t 3D Ti: 2jL3, L If cI:d, Du:.rd inj:i:n 11r, EllFIC43,

3. hn r:tr t 1r d:r. 410 t.q,, I:b:.:d I:Lo YI;,

E2F:JS3, S. iThn lccp f1:: crr :han 1502 r, Mm F1..i E: 7lj, 2F2313

r::n :A, svs2El, nd  : t:I:rt tTA, A21--SE2, :1:s

5. tiz 12 I:1 ssi, d  :.  : s.Dku ID

i TA575452,

5. f:: Dryw11 C:1:s 3 C trmp

13. All nr:r at II. t:r E-m Ir I: J::.1 7.alv. S117, 22 A32C? Sr:m: W.er Inlet S102, 1:e

2. P:m:r lcw 1v1 thr 45 mn:h Hiqh dr11 (1 7 pig i ::nj.:n w.m:h 1:w r::r (4. pig

        . C.r:um

8. 2110001 Which one of the following completes the statement below concerning Core Spray Line Break Detection differential pressure instrument?

The (1) leg of this DP instrument senses (2) core plate pressure via the SLC/Core Differential Pressure penetration. A (1)variable (2) below B (1)variable (2) above C (1)reference (2) below D (1)reference (2) above Answer: D K/A: 211000 Standby Liquid Control System Ki Knowledge of the physical connections and/or cause-effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following: (CFR: 41 .2 to 41 .9 I 45.7 to 45.8) 01 Core spray line break detection RO/SRO Rating: 3.0/3.3 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the physical connection of SBLC and CS line break detection. Pedigree: Last used on 10-1 NRC Exam Objective: CLS-LP-18, Obj. 10 Explain the principle of operation of the CS Line Break Detection Instrumentation

Reference:

None Cog Level: fundamental Explanation: This system is comprised of a differential pressure detector which provides Control Room annunciation on detected high DP. The high pressure reference leg of this instrument is exposed to above core plate pressure via the SLCICore Differential Pressure penetration. The low pressure of this instrument is normally exposed to above core pressure via the Core Spray injection line. This results in the instrument normally measuring core DP (not including core plate DP).

Distractor Analysis: Choice A: Plausible because the examinee may confuse the reference and variable legs and SLC does discharge below the core plate Choice B: Plausible because the examinee may confuse the reference and variable legs Choice C: Plausible because it is the reference leg and SLC does discharge below the core plate. Choice D: Correct Answer, see explanation SRO Basis: N/A DYWL *

          /;,..  -:AM /V/

b Ioa7 OOtk

                          /

I cC PtL! I HI

                                              /CE

This system is comprised of a differential pressure detector which provides Control Room annunciation on detected high R The hugh pressure reference leg otthis instrument is exposed to above core plate pressure via the SLCCore D[ufferenhuuat Pressure penetration. The low pressure of this instrument is normally exposed to above core pressure via the Core Spray injection line. This results in the instrument normally measuring core AP (not including core plate .P:t A break in the Core Spray injection line between the reactor vessel penetration and the core shroud would expose the low pressure side ot the instrument to the lower pressure of the region outside The shroud This would be sensed as an increased differential pressure and Control Room annunciator ould alert the Operator. Although other indications would be available, this alarm would also indicate a break in the line between the E21-FOO6B(A) check valve and the reactor vessel penetration. The Gore Spray pipe break detection instruments are located on the Reactor Building 20 elevation SD-tB Rev6 Page29ot53

9. 212000 1 Which one of the following identifies the normal power supply to RPS MG Set 2B?

A 480V MCC 2CA B 480V MCC 2CB C 480V MCC 2XC D 480V MCC 2XD Answer: B K/A: 212000 Reactor Protection System K2 Knowledge of electrical power supplies to the following: (CFR: 41 .7) 01 RPS motor-generator sets RO/SRO Rating: 3.2/3.3 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the power supply to the RPS MG Set Pedigree: New Objective: CLS-LP-03, Obj 18b State the power supplies for the following: RPS MG Set B

Reference:

None Cog Level: Fundamental Explanation: Power for the Motor Generator Sets is tapped off two phases of the normal 480 VAC MCC 1 CAll CB (2CA/2CB) power supply for the motor through a stepdown transformer (480V to 1 20V) from E5/E6 (E71E8). Selectable reserve power to the Bus is provided from 120 VAC 1 E5(2E7) or 1 E6f2E8), and is normally selected to Division I. In the event that either RPS M-G Set fails to operate, the alternate power source must be manually selected. Distractor Analysis: Choice A: Plausible because 2CA supplies RPS MG Set A. Choice B: Correct Answer, see explanation Choice C: Plausible because 2E7 is normally the alternate RPS power supply. Choice D: Plausible because 2E8 is the alternate alternate RPS power supply. SRO Basis: N/A

3 0 IlPyi.\ PLO CM 0L SU 1 ET 1 Y/E D Q AflC-.E 1o-%i;-i-F1EL A. E SE ?O VOk A D 0 1SS SJCAF 00-3TEV I EDA1DI

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                              .4 E

rcDT &tAM SOt 5411 SS0 0

                                           .4-j--- r9 LIIiQ :rI                             to    -   --

H14 fit, D = 3 0 I (IVJIJtL t Ct I rn tç,j. R) _1i),

to o Cr 1535 SUO(I CV I i) 0 0 r.ih TM N DtD t4Q4 IiIr,:IPNtRAn:4 o o

10. 2150021 Which one of the following identifies the LPRM detector level that provides input to the Rod Block Monitor system for indication ONLY, and is NOT used for the purpose of generating rod blocks?

A Level A B Level B C Level C D Level D Answer: A K/A: 215002 Rod Block Monitor System Ki Knowledge of the physical connections and/or cause-effect relationships between ROD BLOCK MONITOR SYSTEM and the following: (CFR: 41.2 to 41 .9 / 45.7 to 45.8) 02 LPRM RO/SRO Rating: 3.2/3.1 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing the connection between RBM and LPRMs Pedigree: Bank Objective: LOI-CLS-LP-09.6, Obj 5a List the PRNMS system signals/conditions that will cause the following actions: APRM / RBM Rod Blocks

Reference:

None Cog Level: Fundamental Explanation: The level A inputs are sent to RBM-A for processing/output to the LPRM Display Meters on the 4-Rod Display. Level A is for indication only RBM-A Receives all four level C inputs lower left and upper right level B inputs upper left and lower right level D inputs RBM-B Receives all four level C inputs upper left and lower right level B inputs lower left and upper right level D inputs Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because LPRMs have a B level that input to the RBMs Rod Blocks. Choice C: Plausible because LPRMs have a C level that input to the RBMs Rod Blocks. Choice D: Plausible because LPRMs have a D level that input to the RBMs Rod Blocks.

SRD Basis: N/A The A level LPRM detectors are not insed for RBr1 nput processing, wtiile bott REM channels use all C level detectors. This gves an accurate representation of actual power around the control rod. The It and D detectors are distributed evenly between the two REM channel& An example ct LPRM input to a both REM channers with a four-string rod selected is tvo E level LPRMs, tour c level LPRMS. and Iwo D level LPRMs for each channel. The REM circuitry undergoes a nulling and Filtering sequence when a rod is selected and therefore a delay of at least 25 seconds must be allowed beIween selection and rod movement, A Rod lnhibt signal is SD-O9.6 Rev. 12 Page25of95

11. 2150031 Unit One is performing a startup with the reactor just declared critical.

While ranging IRM G from range 1, the IRM will not change ranges and remains on Range 1. Which one of the following completes both statements below? When IRM G indication first exceeds (1) on the 125 scale, annunciator A-05, 2-4, IRM UPSCALE, will alarm. The action requited lAW A-05, 2-4, IRM UPSCALE, is to (2) A (1) 70 (2) place the joystick on P603 for the IRM G to Bypass B (1) 70 (2) withdraw the IRM G detector to maintain reading on scale C (1) 117 (2) place the joystick on P603 for the IRM G to Bypass D (1) 117 (2) withdraw the IRM G detector to maintain reading on scale Answer: A K/A: 21 5003 Intermediate Range Monitor System A2 Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41 .5 / 45.6) 06 Faulty range switch RO/SRO Rating: 3.0/3.2 Tier 2 I Group 1 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL. This meets the K/A because it is testing what will happen with a faulty range switch and the action required. Pedigree: New Objective: LOl-CLS-LP-009.1, Obj. 3a List the SRMIIRM system signals/conditions that will cause the following actions and the conditions under which each is bypassed: Rod Blocks (LOCT) LOI-CLS-LP-009.1, Obj. 14a Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event: SRMIIRM Upscale alarm (LOCT)

Reference:

None Cog Level: High

Explanation: With the reactor critical the indication will continue to rise. The Upscale alarm will come in at 70 on the 0-125 scale. The Upscale Hi/Inop alarm comes in at 117 on the 0-125 scale. lAW with the APP the action to take is to bypass the IRM. In the case of the SRMs an action to take could be to withdraw the SRM to maintain on scale readings. Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because the first part is correct and the second part could be correct if this was an SRM. Choice C: Plausible because 117 is an alarm setpoint for the IRMs and the second part is correct. Choice D: Plausible because 117 is an alarm setpoint for the IRMs and the second part could be correct if this was an SRM. SRO Basis: N/A

tnrt 1 SP A33 0-4 Page 1 of C IPX CPM CALl Roi w:thdrawal block ibypassed whec. React or lrda E.tch in REflt C. 1PM canneD. :ndccaoes greater than :r t: on ClOS scale InlPr:;er rangon: of 1PM c:tanrels dunn: reactor startup a hu if:

3. 1PM detector faalure 4- :t:ring refuel cutages, 1PM stc.king due to r.ce peneration froSt work actovutres an dnwell,. such as weldrng E CirnuCt maCjunction C9 SER7AOICDfE C. 1PM :hanneC indicatrng :reaten than cc ezual t:- :n 0125 scale C. 1PM :nanne_ u;scaCe CISC ALARM: arLen nniicatrn; light on

3. ROD 0.T1 910:1. AOS CC; a:ac.s

4. Rod withdrawal pernissive indicatung lioh: cif ACTIONS If :n progress, stop withdrawal of ::ntro. ccds M:n:t:r 1PM indiratr:ns to detern.ne affected channels -

C&UTI DN should be retcsitroned carefully in crder tc prevent ?-::g:

3. Rep:s:tion affertei 1PM range switch to ne:-:t hrgher range -

4. If a sudden rrse ifl rndacated reactcr power :ccurred on more than one 1PM channel, verify correct rod withdrawal seguence 15 L-eing used and insert insequence c:ntrol rcds as necessary c.: turn cower cisc.

S. If :2CC detect:r faCuce :r crrcui: ralfuncticn cs suspected, cerf:rn the f:llcwtng:

a. Refer *c Ierhni:al Specrfacaticn 3.3.0.1 and OPM 3.3 f:r 1PM channel :uerab:lity reaunrenencs.

b. Notify :Lit ORE.

c. Sytass affected channel usin: PM bypass swatch.

d. Ensure a WR s prepared.

IAPP-A-05 Rev. 75 Page 24 of 94

12. 2150032 A reactor shutdown is in progress.

All IRMs on range I reading between 15 and 20. IRM B detector is failing downscale. Which one of the following completes both statements below? A-05 (1-4) IRM Downscale, alarm setpoint is < (1) on the 125 scale. When the IRM downscale alarm is received, a rod block (2) be generated. A(1)3 (2) will B (1) 3 (2) will NOT C (1) 6.5 (2) will O (1) 6.5 (2) will NOT Answer: D K/A: 215003 Intermediate Range Monitor (IRM) System K6 Knowledge of the effect that a loss or malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM : (CFR: 41.7/45.7) 04 Detectors RO/SRO Rating: 3.0/3.0 Tier 2 / Group 1 K/A Match: This meets the K/A because this is testing a failure/malfunction of a detector effect on the IRM system (whether it generates a rod block) Pedigree: New Objective: LOI-CLS-LP-009-A, Obj. 3a List the SRM/IRM system signals/conditions that will cause the following actions and the conditions under which each is bypassed: Rod Blocks

Reference:

None Cog Level: High Explanation: The downscale setpoint for the IRMs is 6.5 on the 125 scale. The rod block is bypassed under these conditions because the IRMs are all on Range 1.

Distractor Analysis: Choice A: Plausible because 3 is the downscale tech spec setpoint for SRMs and if the IRMs were not all on range 1 a rod block would be generated. Choice B: Plausible because 3 is the downscale tech spec setpoint for SRMs and the second part is correct. Choice C: Plausible because the first part is correct and if the IRMs were not all on range 1 a rod block would be generated. Choice D: Correct Answer, see explanation. SRO Basis: N/A

                                                                                 ; :.s 14

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2. O:rr.ttr rt:ut.

ASE I- :r.ii trrg 1s than :r c!ua1 EE.:r. th 1123 ca1 when t:s tanoi wttth ts not on Ranc

1. Irr.t:otr nc:n f i2! oh1 thrn; tactot tartu :r abu:d:n.

3. 1- dt:t:tn:.: f1ly st.t 4.. .M d,:a:t:r fai1i:.

3. *2.t:t na1.z::t:n.

C3SEP.7ATICNS

1.  :: hann1 .nftatinc 2s than cr ua1 to E .3 cn rh -123 a1e.

2. 2P d:wn:a1 :m:s:: jndt:at:nc 1ngh is Zn.

3. CJT ELC21: A3 22 alatn, if afferted I2 :hannel is not on

ange 1.

4. f the affe:tei 1- channEL ts n:t Zn .ance 2, the r:d w:hira;a1 oeisie tndicatin 1iht w:11 be cff.

13. 2150041 Which one of the following identifies the criteria for when SRM detectors can first begin to be withdrawn from the core lAW OGP-02, Approach To Criticality And Pressurization Of The Reactor?

A When all IRMs are above range 3. B When SRM counts reach 2 x iO counts. C When RTRCT PERMIT light is illuminated. D When SRM/IRM overlap has been established. Answer: D K/A: 21 5004 Source Range Monitor System K5 Knowledge of the operational implications of the following concepts as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM: fCFR: 41.5/45.3) 03 Changing detector position RO/SRO Rating: 2.8/2.8 Tier2/Groupl K/A Match: This meets the K/A because it is testing when SRMs can change detector positions. Pedigree: New Objective: LOI-CLS-LP-009-A, Obj 3b List the SRM/IRM system signals/conditions that will cause the following actions and the conditions under which each is bypassed: Retract Permissive (SRM) only (LOCT)

Reference:

None Cog Level: Fundamental Explanation: When SRM/IRM overlap has been established then SRM can be withdrawn to maintain an indicated SRM count rate between 100 cps and 200,000 cps. Distractor Analysis: Choice A: Plausible because this is the logic setpoint at which the SRM can be fully withdrawn Choice B: Plausible because this is the point at which the SRM must be fully withdrawn. Choice C: Plausible because this is an indication that is used during the withdrawal of the SRMs Choice D: Correct Answer, see explanation. SRO Basis: N/A

NOTE

• SRMIIRM overlap is required to be demonstrated for all operable iR1 channels prior to withdrawing SRMs from the Filly inserted position. SRMlRM overlap exists when IRM channels show an increase to at least twice their pie-startup levels and indicate at least 10% of scale (Ce.. 12.5 on the digital readout 0-125 scale) before the first SRM channet teaches 5 x cps (Technical Specifications, SR 3.3.11.6) 0
• If desired, the level of the highest readng IRM (pre-starfup) may be doubled and that value used as overlap criteria for all IRMS. This method wilt allow the operator to compare IRM channel response to a single value wtiich is at least twice the pre-startup levels of the individual IRMs 0 APPROACH TO CRITICALITY AND PRESSURIZATION OCP-02 OFTHE REACTOR Rev. 109 Page 16 of 54 6.2 Pulling Rods To Achieve Criticality (continued NOTE
• Wh IRM channels below Range 3, the SRM channels will initiate a rod withdrawal block when either of the following conditions exists:

SRM channel indicates greater than 2 x iO cps 0 O SRM channel indicates tess than 102 cps with its detector NOT full in 0

• SRM detectors are withdrawn two at a time so that the reactor flux level conditions are being monitored by channels that are NOT being affected by detector movement 0

32. WHEN SRMIIRM overlap has been confirmed.

THEN withdraw SRM detectors as required to maintain an indicated SRM count rate between 102 cps and 2 cps CAUTION Repositioning IRM range switches is performed by one operator, using one hand, on one trip system atatime t8.16} 0

33. As reactor power rises, reposition the IRM range switches to maintain IRM indication on recorders between 15 and 50 on the 0-l25scale

34. WHEN all OPERABLE IRM channels are above Range 3 A1I prior to reaching Range 7, THEN fully withdraw all SRM detectors .

14. 2150051 Which one of the following identifies the power supply to the APRM channel NUMACs?

A All APRM channels receive 120 VAC power from UPS B All APRM channels receive 120 VAC power from both RPS Bus A and RPS Bus B C APRM Channels 1 & 3 receive power from ONLYI 20 VAC RPS Bus A APRM Channels 2 & 4 receive power from ONLY1 20 VAC RPS Bus B D APRM Channels 1 & 3 receive power from Division I 24/48 VDC APRM Channels 2 & 4 receive power from Division II 24/48 VDC Answer: B KJA: 215005 Average Power Range Monitor/Local Power Range Monitor K2 Knowledge of electrical power supplies to the following: (CFR: 41 .7) 02 APRM channels RO/SRO Rating: 2.6/2.8 Tier 2 / Group 1 KJA Match: This meets the K/A because it is testing the power supply to the NUMACs. Pedigree: Modified from 2015 NRC Exam Objective: LOl-CLS-LP-09.6, Objective 7a Describe the operational relationships between the PRNMS and the following: Reactor Protection System

Reference:

None Cog Level: Fundamental Explanation: Each APRM channel NUMAC is equipped with a dual power supply arrangement with one supply from RPS Bus A and the other supply from RPS Bus B. All four APRM channels maintain power on loss of either supply as long as the other supply is available Distractor Analysis: Choice A: Plausible because UPS supplies power to the APRM ODA and recorder Choice B: Correct Answer, see explanation. Choice C: Plausible because this is the power supply arrangement for the voters. Choice D: Plausible because other ranges of nuclear instrumentation (SRM/IRM) receive their power from here. SRO Basis: N/A

2.8.8 PRNMS Power Supplies The Power Range Neutron mondonng System uses one Quadruple Voltage Power Supply (QLVPS) chassis and four Dual Low Voltage Power Supplies (DLVPSL one for each bay of the PRNMS panel, to provide redundant power to the NUMAC dnstwments. These LVPS convert 120 VAC to low voltage D& See Figure 09.6-14 Each APRM instrument receives power from two power supplies, LVPS I and LVPS 4. LVPS 1 is fed from RPS Bus A while LVPS 4 is fed from RPS Bus B. Therefore, a loss olan RPS Bus will not affect operation of the APRM NUMACS. Each RBM instrument also SD-Q6 Rev.11 Page32ot94 4.3.1 Reactor Protection System APRM channels provide signals to open contacts in the scram trip logic of the RPS System under various conditions discussed previously. The RPS System provides power to each of the four APRM instruments, which in turn provide power to all subsystems driven from the APRM instruments or NUMAC. Both RPS busses, A and B, provide power to each APRM instrument, as well as, each RBM. Therefore, a loss of one RPS bus will riot affect operation of the PRNr.IS. The reactor mode switch provides input to each APRM instairnent to determine when to enforce the tixed or flow biased scram trip and rod block settings. OPRM circuitry is enabled only when power/flow conditions are met and the mode switch in RUN. j SD-09.6 Rev. 11 Page 43 of 94 Which one of the following is the power supply to APRM Channel 4 NUMAC on P608? k 120 VAC RPS B. 120 VAC UPS C. 24148 VDC Div I D. 24148 VDC Div II

15. 2150052 Which one of the following completes the statement below?

An APRM must have at least (1) of the assigned LPRMs operable with at least (2) LPRM inputs per axial level operable. A (1) 18 (2) 2 B (1) 18 (2) 3 C (1) 17 (2) 2 D (1) 17 (2) 3 Answer: D K/A: 215005 Average Power Range Monitor/Local Power Range Monitor K5 Knowledge of the operational implications of the following concepts as they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: (CFR: 41.5/45.3) 04 LPRM detector location and core symmetry RO/SRO Rating: 2.9/3.2 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the knowledge of the LPRM inputs per axial level that ate required and the minimum number of inputs for core symmetry that are required. Pedigree: New Objective: LOI-CLS-LP-09.6, Obj. 13b Given plant conditions, predict the effect of a single or multiple LPRM failure on the following: APRM

Reference:

None Cog Level: Fundamental Explanation: An APRM channel must have a minimum of 3 LPRM inputs pet level and a total of 17 LPRM inputs to be operable Distractot Analysis: Choice A: Plausible because an OPRM requires 18 LPRMs with at least 2 LPRM inputs to each cell. Choice B: Plausible because an OPRM requires 18 LPRMs and 3 per level is correct for APRMS. Choice C: Plausible because 17 is correct for APRMs and OPRMs require at least 2 LPRM inputs to each cell. Choice D: Correct Answer, see explanation. SRO Basis: N/A

4.21 LPRM LPRM System failure, dependinig on the extent or failure type. can cause the loss of LPRM functions including the loss of indicatiion, incorrect operation of rod bllccR or scram protection. Generally, the following symptoms are exhibited for LPRM failure for the affected LPRM:

• Indicates upscale, accompanied by an upscale alarm.
• indicates downscale, accompanied by a downscate alarm.
• Indicator reads erratically.

The results ot an LPRM failure may lead to an APRM or OPRM becoming inoperable. An APRM channel must have a mininium 013 LPRM inputs per level and a total ot 17 LPRM inputs to be operable. SD-O9.6 Rev. 12 Page 45 of 9:5 An OPRM cell must have a rnin[mum ot 2 LPRM inputs to each cell and a total of 18 cells to be operable.

16. 2160001 A Unit Two plant cooldown is being performed with the following plant conditions:

Reactor water level 175 inches, steady Reactor pressure band 500 700 psig Drywell ref leg temp 175°F (REFERENCE PROVIDED) Which one of the following completes both statements below? The lowering of reactor pressure causes the NOO4NB/C (Narrow Range) reactor water level instruments indicated level error to (1) The reactor water level that would correspond to Low level 4 (LL4) is (2) A (1) increase (2) -60 inches B (1) increase (2) -65 inches C (1) decrease (2) -60 inches D (1) decrease (2) -65 inches Answer: A K/A: 216000 Nuclear Boiler Instrumentation A2 Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) 11 Heatup or cooldown of the reactor vessel RO/SRO Rating: 3.2/3.3 Tier 2 / Group 2 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL. The first part of the question deals with predicting the effect of a cooldown on indicated level error while the second part has the student has to determine based on the lowering pressure what LL4 value would be which is the value that emergency depressurization would be required. They must utilize the lower end of the pressure band to determine LL4. If LL4 cannot be maintained then ED is required. Pedigree: New Objective: LOI-CLS-LP-001 .2, Objective 5a Explain the effect that the following will have on reactor vessel level and/or pressure indications: Plant heatup/cooldown

Reference:

OEOP-0 1 -UG, Attachment 26

Cog Level: High Explanation: The indicated level error is sensitive to changes in the saturation density of the bulk water as a function of system pressure. The amount of the indicated level error is also a function of the difference in the actual water level and the variable leg instrument tap elevation. As the saturation density increases (pressure decreases) the indicated level error will increase for the narrow and wide range instruments and decrease for the fuel zone and shutdown range instruments due to calibration criteria. From 01-37.11, TAF, LL4, and LL5 values should be determined based on the reference leg area temperature and RPV pressure compensation curves, using RPV pressure at the low end of the established RPV pressure control band. Based on the low end of the band of 500 psig and < 200°F in the drywell the LL4 value would be -60 inches. Distractor Analysis: Choice A: Correct Answer, see explanation Choice B: Plausible because the first part is correct and the second part would be correct for 700 psig. Choice C: Plausible because this would be correct for the fuel zone or shutdown range instruments and the second part is correct. Choice D: Plausible because this would be correct for the fuel zone or shutdown range instruments and the second part would be correct for 700 psig. SRO Basis: N/A

USERS GUIDE OEOP-0I-UG Rev. 067 Page 94 of 155 ATTACHMENT 26 Page 1 of 1

                               <<Unit 2 RPV Level at LL 4 (Minimum Steam Cooling RPV Level)>>

0 1111 .IIJJ

                -10                                 ABOVE LL-4
                -20 C) uJ 2:

C z

                -40
           -J U
           >    -50 U
           -J C    -60                                                         pr LEG U                                                               /

4 C -70 RELEO Th&1 P C OR z -80 EQUAL TO 2C1 BELOW

                -90 LL- 4 III II II
              - 100 1,150 boo      300      500      700       900    11100 60 200         400     600       800     1,000 RPV PRESSURE (PSIG)

When RPV pressure is less than GO psig, use indicated level. LL-4 is -27.5 inches. 4.1.2 System Pressure (Heat-up and Cool-down) The indicated level error is sensitive to changes in the saturation density of the bulk water as a function of system pressure. The amount of the indicated level error is also a function of the difference in the actual water level and the variable leg instrument tap elevation. As the saturation density increases (pressure decreases) the indicated level error will increase for the narrow and wide range instruments and decrease for the fuel zone and shutdown range instruments due to calibration criteria. As actual water level decreases, the amount of error will decrease because less vessel water level is acting on the instrument. SD-01.2 Rev.10 Page36of85

. TRANSIENT MITIGATION GUIDELINES 001-37.11 Rev. 4 Page 170125 5..3 1(2)EOP-O1-RVCP. Reactor Vessel Control Procedure Level Leg

a. The CR5 directs an initial RPV level band of +166 to +206 inches. The reactor operator actually maintains a RPV levet band of +170 to

                 +200 inches to provide additional margin to the reactor scram and turbne trip set points. The CR5 ma direct a Mdened band based on plant conditions and other contrdling procedures associated with the transient.

b. If RPV level is above TAF, injection flow should be controlled so as to control the cooldown rate below 100°F/hr.

c. If RPV level is below TAP, RPV level should be rapi:dly restored to above TAP, and then injection flow reduced so as to control the cooldown rate below 100°F/hr.

d. TAP, LL4, and LLS values shou[d be deterniined based on the reference leg area temperature and RPV pressure compensation curves, using RPV pressure at the low end at the established RPV pressure control band.

17. 2170001 Following a loss of feedwater, RCIC automatically initiated and subsequently tripped on low suction pressure.

Current plant status is: Reactor water level is 150 inches RCIC flow controller in Manual set at 200 gpm Subsequently, the following actions are taken: RCIC suction transferred to Torus E51-V8, Turbine Trip and Throttle Valve is closed E51-V8 is re-opened PF push button on the RCIC flow controller is depressed Which one of the following identifies the indicated flow on the RCIC flow controller that would be observed for these conditions? A Ogpm B 200 gpm C 400 gpm D 500 gpm Answer: C K/A: 217000 Reactor Core Isolation Cooling System Al Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) controls including: (CFR: 41.5 I 45.5) 01 RCIC flow ROISRO Rating: 3.7/3.7 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the prediction of what RCIC flow will be when operating the RCIC system. Pedigree: New Objective: CLS-LP-016-A, Obj.16c Describe how the following evolutions are performed during operation of the RCIC System: Adjusting RCIC flow in the Reactor Level Control mode.

Reference:

None Cog Level: high Explanation: The RCIC Turbine is provided with a solenoid operated remote electrical tripping device, which when actuated (in this case by low suction pressure), will close the Turbine Trip and Throttle Valve, E5l-V8. Resetting of the remote electrical tripping device may be accomplished from the RTGB. The RCIC system is restarted after auto initiation and turbine trip by fully closing the V-8, and re-opening the V-8. Located on the controller face is a PF (programmable function) pushbutton which when depressed an automatic transfer from manual to automatic at a predetermined setpoint of 400 GPM will result. This button (PF) has no function if the controller is already in automatic.

Distractot Analysis: Choice A: This is plausible because this answer would be correct for these actions following a high RPV water level trip of RCIC Choice B: Plausible because this would be correct if the operator did not depress the PF pushbutton. Choice C: Correct Answer, see explanation. Choice D: Plausible because the FE push button would raise RCIC flow to rated (400 gpm) and not maximum per procedure (500 gpm). Achieving 500 gpm would require the flow control setpoint to be manually raised. SRO Basis: N/A From SD-16: Also located on the controller lace is a PE (programmable function) pushbutton. When depressed an automatic transter from MANUAL to AUToMATIC at a predetermined setpoint of 400 GPM will result. NOTE: This button (FE) has no function it the controller is already in AUTO MAIl C. For various internal processing failures, the controller is designed to hold the last output and automatically switch to MANUAL giving the operator manual control capability. Barring operator intervention, this failure could result in rising or lowering RCIC flow and would be indicated by the red FAIL lamp on the controller face. Failure display code can then be checked usng the side panel keypad. A down scale failure of the controller is possible and would result in turbine operation at well below the normal minimum speed of 2000 rpm. An upscale failure is highly unlikely but would result in turbine speed at or above the maximum running speed of 4600 rpm. Failures associated with the dynamic response are also highly unlikely but would produce either excessively sluggish responses or dynamic instability (full scale oscillations) when in the Automatic mode Programmable settings internal to the controller are maintained during a loss of 24 Vdc power supply by a lithium battery. It this battery voltage drops to a pre-determined low value, the yellow ALARM light will flash II the input signals are not within the Limits of -6.3% to 1061% or if the input or output signals are not intact, the Yellow ALARM light will come on solid. SD-16 Rev. 12 Page29of120

REACTOR CORE ISOLATION COOUNG SYSTEM OPERA11NG PROCEDURE I 2OP-1 6 Rev. 120 Page 99of 99 ATTACHMENT 9 Page 1 of 1

                   <<RCIC Instructional Aid for EOPs>>

RESTARTING RCIC AFTER AUTO INITIATON AND TURBINE TRIP (2OP-16 Section 5.!)

1. ENSURE THE [51-VS (VALVE POSIIflON) AND [51-VS MOTOR OPERATOR) ARE CLOSED C

2. PLACE RCIC FLOW CONTROL IN MANUAL IM) AND ADJUST OUTPUT TO 0% C

3. JOG OPEN [51-VS UNTIL THE TURBINE SPEED 15 CONTROLLED BY THE GOVERNOR C

4. FULLYOPENE5I-V8 C

5. SLOWLY RAISE TURBINE SPEED UNTt FLOW RATE or AT LEAST 120 GPM C

6. ENSURE [51-FOlD IS CLOSED WITH FLOW GREATER THAN 80 GPM C

1. WHEN SYSTEM CONDITIONS ARE STABLE, THEN ADJUST SETPOINT, AND TRANSFER RCIC FLOW CONTROL TO AUTO (A) C

8. SLOWLY ADJUST FLOW RATE USING RCIC FL&W CONTROL INAUTO(A C

9. ENSURE THE FOLLOWING:

BAROMETRIC CNDSR VACUUM PUMP HAS STARTED C SBGT STARTED (2OP-10) C SGT-V8 AND SGT-V9 ARE OPEN C

18. 2180001 Which one of the following completes both statements below concerning the Automatic Depressurization System (ADS) reactor water level inputs from the Nuclear Boiler System?

The (1) instruments provide LL3 inputs to ADS initiation logic. The (2) range instruments provide LLI inputs to ADS logic. A (1) FuelZone (2) Narrow B (1) FuelZone (2) Shutdown C (1) Wide range (2) Narrow D (1) Wide range (2) Shutdown Answer: C K/A: 217000 Automatic Deptessurization System Ki Knowledge of the physical connections and/or cause-effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following: (CFR: 41 .2 to 41.9 / 45.7 to 45.8) 03 Nuclear boiler instrument system RO/SRO Rating: 3.7/3.8 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the connection between ADS and level indicators. Pedigree: New Objective: LOl-CLS-LP-0001 .2, Obj 4a List the systems which receive input from the Vessel Instrumentation system for the following: Level signal

Reference:

None Cog Level: Fundamental Explanation: B21-LT-N031(Wide Range) provide LL3 initiation from NO31A and C for Logic B and from N031 B and D for Logic A. B21 -LT-N042 (Narrow Range) provide LU confirmatory from N042A for Logic B and from N042B for Logic A.

Distractor Analysis: Choice A: Plausible because the fuel zone instruments covers LL3 (45 inches) and the second part is correct. Choice B: Plausible because fuel zone instruments covers LL3 (45 inches) and the shutdown range covers the LL1 setpoint (166 inches). Choice C: Correct Answer, see explanation. Choice D: Plausible because the wide range is correct and the shutdown range covers the LL1 setpoint (166 inches). SRO Basis: N/A 4.1.2 Automatic Operation The ADS logic automatically opens the ADS valves in the event the HPCL System tails to maintain reactor level during a LOA. The seven ADS valves open automat[cally when all the following conditions are met on either of two logic channels (A or B) associated with ADS: Reactor low water level (LL3 from 621 -LTS-N03 1 A and C or B and D). Reactor conflmmtory low water level (LL1 from 621 -LTS-N042A or B). Operation of both pumps of an RHR loop or one Core Spray pump as indicated by a pump discharge pressure of 115 psig (either El 1-PS-NO 16A AND C or B AND D or El i-PS-NO2OA AND C or B AND D for RHR or either E21-PS-NOO8A AND El 1-PS-NOO9A or E2i-PS-NOO8B AND E21-PS-N009B for CS). A time delay of 83 seconds has elapsed (timer B2 1-TDPU-K5A or B). AUTO/INHIBIT switches in AUTO for either or both logic channels A and B. SD-20 Rev.3 PAGE 26 of 62

rl[ FIGURE 20-10 w ADS Logic Channel LA1 In I2 I I I I (I I

                                                -J it H

9 - Li LL U4 I j I Li

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19. 2180002 Unit One is operating at power with Core Spray Pump 1 B under clearance.

A small break LOCA occurs simultaneously with a Loss of Off-site Power to both units. DGI and DG4 fail to start and tie onto their respective E bus. The following plant conditions exist on Unit One: A-03 (5-1) Auto Depress Timers Initiated In alarm A-03 (6-9) Reactor Low Wtr Level Initiation In alarm RPV pressure 600 psig Drywell pressure 13 psig Which one of the following completes both statements below? ADS (1) auto initiate. After ADS is initiated (either automatically or manually), RPV water level will be restored with RHR Loop (2) A (1) will (2) A B (1) will (2) B C (1) wiIINOT (2)A D (1) wiIINOT (2) B Answer: C KJA: 217000 Automatic Depressurization System K3 Knowledge of the effect that a loss or malfunction of the AUTOMATIC DEPRESSURIZATION SYSTEM will have on following: (CFR: 41 .7 / 45.4) 01 Restoration of reactor water level after a break that does not depressurize the reactor when required RO/SRO Rating: 4.4/4.4 Tier 2 I Group 1 K/A Match: This meets the K/A because it is testing the knowledge of the effect of the malfunction on auto initiation of ADS and how level will be restored. Pedigree: Last used on 10-1 NRC exam Objective: CLS-LP-20 Obj. 16b Given plant conditions, predict how the following will be affected by a loss or malfunction of ADSISRVs: Reactor water level

Reference:

None

Cog Level: high Explanation: With the loss of offsite power and 1 B CS pump under clearance this would leave only one pump available in each RHR loop. Therefore ADS logic is lost. Level will continue to lower until the ADS valves are manually opened (emergency depressurization) at which time the running low pressure pumps will be able to add water. Injection would be from the A Loop of RHR as the B Loop injection valves do not have power. Distractor Analysis: Choice A: Plausible because ADS does have initiation conditions except that the logic will not have the appropriate pumps lined up for injection. Choice B: Plausible because ADS does have initiation conditions except that the logic will not have the appropriate pumps lined up for injection. B Loop of RHR does not have power to the injection valves Choice C: Correct Answer, see explanation. Choice D: Plausible because ADS will not auto initiate but the B Loop of RHR does not have power to the injection valves. SRO Basis: N/A SD-20 4.1.2 Automatic Operation The ADS logic automatically opens the ADS valves in the event the HPCI System fails to maintain reactor level during a LOCA. The seven ADS valves open automatically when all the following conditions are met on either of two logic channels (A or B) associated with ADS:

• Reactor confirmatory low water level (LL1 from B21-LTS-N042A or B).
• Operation of both pumps of an RHR loop or one Core Spray pump as indicated by a pump discharge pressure of 115 psig (either Eli -P5-NOl 6A AND C or B AND D or Eli -PS-NO2OA AND C or B AND D for RHR or either E21-PS-NOO8A AND E11-PS-NOO9A or E2i-PS-NOO8B AND E21-PS-NOO9B for CS).
• A time delay of 83 seconds has elapsed (timer B21-TDPU-K5A or B).
• AUTO/INHIBIT switches in AUTO for either or both logic channels A and B. Reactor low water level (LL3 from B2i-LTS-NO31A and C or B and D).

20. 223001 1 Which one of the following completes the statement below concerning the Fuel Zone instruments, N036 and N037, during a loss of drywell cooling?

The reference leg density will (1) causing the indicated level to read (2) than actual level. A (1) rise (2) higher B (1) rise (2) lower C (1) lower (2) higher O (1) lower (2) lower Answer: C K/A: 223001 Primary Containment System and Auxiliaries K3 Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following: fCFR: 41 .7 / 45.4) 09 Nuclear boiler instrumentation RO/SRO Rating: 2.8/3.1 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing the knowledge of a loss of DW cooling has on instrumentation. Pedigree: Bank Objective: LOl-CLS-LP-001.2, Obj. 05c Explain the effect that the following will have on reactor vessel level and/or pressure indications: High containment (primary and secondary) temperatures.

Reference:

None Cog Level: High Explanation: The reference leg length is longer than the variable leg length, therefore secondary temp increasing makes the instrument read higher than actual level.

Distractor Analysis: Choice A: Plausible because density is a function of temperature and the temperature is rising. The second part is correct. Choice B: Plausible because density is a function of temperature and the temperature is rising. The second part is plausible because if the first part is seen as right then this would be correct. Choice C: Correct Answer, see explanation. Choice D: Plausible because the first part is correct and the second part is the opposite of the answer. SRO Basis: N/A

21. 2230021 Unit One is at 75% power.

The 1A RPS MG set trips. RPS Bus A has NOT been transferred to an alternate power supply. Which one of the following identifies the Main Steam Line Isolation Valve (MSIV) logic lamp status on P601 panel? Inboard MSIV Logic Outboard MSIV Logic AOO BOO 0 COO 0 0 Answer: C K/A: 223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off Al Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including: (CFR: 41 .5 / 45.5) 01 System indicating lights and alarms RO/SRO Rating: 3.5/3.5 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the ability to predict the light status on a loss of a power supply Pedigree: New Objective: LOI-CLS-LP-012, Objective 12 Given plant conditions, determine how the following will affect PCIS:

c. Loss of RPS

Reference:

None Cog Level: High Explanation: See Notes Section. RPS A provides power to PCIS Logic A. PCIS Logic A is Inboard AC and Outboard DC indicating lights on P601.

Distractor Analysis: Choice A: Plausible because first part is correct. Outboard light is DC. Choice B: Plausible because second part is correct. Inboard light is AC. Choice C: Correct Answer, see explanation. Choice D: Plausible because the lights are just reversed. This would be true for Loss of RPS B. SRO Basis: N/A 4.3.10 AC Distribution RPS MG sets supply power to the following PCIS related components: RPS Bus A PCIS Trip System A logic PCIS Trip Channels Al and A2 logic Inboard isolation logic for valves: Inboard reactor water sample valve Main Steam Line drains Shutdown cooling suction RWCU Inboard RHR Sample valves DiyeIl floor and equipment drains CAC/CAMS/PASS far LU and Hig11 DrweII pressure Valve operating power: Inboard reactor water sample valve Inboard RHR Sample valves Drywell floor and equipment drains Inboard AC MSIV solenoids Reactor Building Vent Exh Rad Monitor NO 1 OA Main Steam Line Rad Monitors A and C (alami function only) SD-12 Rev. 11 Page65of2O8 P601 panel. These lights are arranged above the MSIV control switches as follows: TABLE 25-3, MSIV ISOLATION SIGNAL STATUS Light INBD DC INBD AC OUTBD DC OUTBD AC Solenoid 125 VDC RPS A 125 VDC RPS B Power A B PCIS Logic B A A B SD-25 Rev. 14 Page 16of79

22. 234000 1 Which one of the following identifies the affect if both Refuel Bridge hoist grapple hooks are not open five seconds after placing the Engage/Release switch to Release?

A Fuel Hoist Interlock is generated. B Engage amber light extinguishes. C Fault lockout is generated. D Grapple hooks will reclose. Answer: D K/A: 234000 Fuel Handling A3 Ability to monitor automatic operations of the FUEL HANDLING EQUIPMENT including: (CFR: 41.7 / 45.7) 01 Crane/refuel bridge movement RO/SRO Rating: 2.6/3.1 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing the ability to monitor the crane grapple hooks auto re-close feature. Pedigree: New Objective: LOl-CLS-LP-58. 1, Obj 13 Describe the operation of the grapple if the ENGAGE/RELEASE Switch is positioned to RELEASE and both grapple hooks are not open within 5 seconds when the main hoist is loaded.

Reference:

None Cog Level: Fundamental Explanation: If the grapple does not indicate released (open) within 5 seconds, the solenoid is de-energized and the grapple hooks re-close. The switch must then be taken to the ENGAGE position to reset the logic prior to making another attempt to release the grapple. Distractor Analysis: Choice A: Plausible because a Fuel Hoist Interlock is generated for a number of reasons. Choice B: Plausible because this is an indication of operation of the grapple hooks. Choice C: Plausible because a fault lockout is generated for a number of reasons. Choice D: Correct Answer, see explanation. SRO Basis: N/A

23. 239002 1 Which one of the following identifies the SRV component that will prevent siphoning of water into the SRV discharge piping?

A Vacuum breaker B Check Valve C T-Quencher O Sparger Answer: A K/A: 239002 Safety Relief Valves K4 Knowledge of RELIEF/SAFETY VALVES design feature(s) and/or interlocks which provide for the following: (CFR: 41 .7) 03 Prevents siphoning of water into SRV discharge piping and limits loads on subsequent actuation of SRVs RO/SRO Rating: 3.1/3.3 Tier 2 / Group 1 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL. This meets the K/A because it is testing the knowledge of the design feature that prevents siphoning of water. Pedigree: New Objective: LOI-CLS-LP-020, Obj. 7d State the purpose of the following: SRV tailpipe vacuum breakers

Reference:

None Cog Level: Fundamental Explanation: Following operation of the valve, a vacuum is created in the SRV tailpipe as the steam condenses. Water in the line above the suppression pool water level would cause excessive pressure at the SRVs discharge when and if the valve reopened. For this reason, a vacuum relief valve is provided on each SRV tailpipe to prevent drawing water up into the line due to this steam condensation following SRV operation. Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because this is a component that is typically used to provide an anti-siphon break. Choice C: Plausible because this is a component on the SRV that the steam discharges through and has holes in the pipe which could be thought of an anti-siphon type break. Choice D: Plausible because this is a component on the SRV that the steam discharges through and has holes throughout the pipe which could be thought of an anti-siphon type break. (The supplemental fuel pool cooling sparger has this design to prevent siphoning of water) SRO Basis: N/A

24. 241000 1 Which one of the following identifies the criteria for tripping the main turbine lAW the Unit Two Scram Immediate Actions of OEOP-01-UG, Users Guide?

A When APRMs indicate downscale trip. B When steam flow is less than 3 Mlbs/hr. C When reactor water level is 160 inches and rising. D When reactor mode switch is placed in SHUTDOWN. Answer: A K/A: 241000 Reactor/Turbine Pressure Regulating System A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) 14 Turbine trip RO/SRO Rating: 3.8/3.7 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing the ability of tripping the turbine from the control room. Pedigree: New Objective: LOl-CLS-LP-300-C, Obj. 2 List the immediate operator actions for a reactor scram.

Reference:

None cog Level: Fundamental Explanation: The main turbine is tripped after reactor power is below 2% which is indicated by APRM downscale trip lights illuminated. Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because this is a criteria for placing the mode switch to shutdown which is an immediate operator action. Choice C: Plausible because this is a criteria for a reactor feed pump which is an immediate operator action. Choice D: Plausible because this is an immediate operator action that is performed on the scram. SRO Basis: N/A

ATTACHMENT 38 Page 1 of 1 Unit 2 Scram Immediate Actions fOEOP-G1 -UG) SCRAM IMMEDtATE ACTIONS

1. Ensure SCRAM valves OPEN by manual SCRAM or ARt initiation.

2. WHEN steam flow less than 3 x f0 lb/hr.

THEN place reactor mode switch in SHUTDOWN.

3. [E reactor power below 2% (APRM downscale trip).

THEN trip main turbine,

4. Ensure master RPV level controller setpoint at +170 inches 5, IF:

• Two reactor feed pumps running AND
• RPV level above +160 inches AND
• RPV level rising, THEN trip one.

25. 245000 1 Which one of the following completes both statements below concerning the Main Generator Voltage Regulator?

The automatic voltage regulator maintains a constant generator (1) voltage. While in the automatic voltage regulation mode, the manual voltage regulator setting (2) automatically follow the automatic setpoint. A (1) field (2) does B (1) field (2) does NOT C (1) terminal (2) does ID (1) terminal (2) does NOT Answer: D K/A: 245000 Main Turbine Generator and Auxiliary Systems K4 Knowledge of MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS design feature(s) and/or interlocks which provide for the following: (CFR: 41 .7) 07 Generator voltage regulation RO/SRO Rating: 2.5/2.6 Tier 2 / Group 2 K/A Match: This meets the K/A because this is testing the design of the auto regulator as to what it controls and whether the manual regulator automatically follows the auto regulator. Pedigree: Bank Objective: LOl-CLS-LP-027.0, Obj 7c Given a simplified diagram of the Main Generator Voltage Regulator, explain how:

a. the MANUAL regulator controls Generator output voltage

b. the AUTOMATIC regulator controls Generator output voltage

c. to transfer from one Voltage Regulator to the other

Reference:

None Cog Level: Fundamental Explanation: The AVR controls terminal voltage while the manual regulator controls field voltage. The manual voltage regulator does not track the setpoint of the AVR, this must be manually adjusted in the control room.

Distractor Analysis: Choice A: Plausible because the MVR controls field voltage and the DG manual voltage regulator does track the auto regulator setpoint. Choice B: Plausible because the MVR controls field voltage and the second part is correct. Choice C: Plausible because the first part is correct and the DG manual voltage regulator does track the auto regulator setpoint. Choice D: Correct Answer, see explanation. SRD Basis: N/A 2.15 Excitation Control (Refer to Figure 27-12) The Silicon Controlled Rectifier (SCR) bridge circuit is used as a variable DC voltage source to control the exciter field current as required by the AC or DC regulator. The source of the control signal for the SCRs is determined by the Regulator Mode Selector Switch (43C5) located on Panel XU-L When Manual is selected, the DC regulator maintains a constant generator field voltage that is determined by the Manual Volts Adjust Rheostat. When the Automatic regulator is selected, the AC regulator maintains a constant generator terminal voltage. 2.17.8 Generator Voltage Regulator Differential Voltmeter This is a standard voltmeter that measures the magnitude and polarity of the difference between the DC regulator output signal and the AC regulator output signal. When shifting control from the DC voltage regulator to the AC regulator or back, it is important to ensure that the signals are the same. As an example, l the meter reads to the clockwise of zero, then the manual regulator output is less than the automatic regulator. If the meter reads counter clockwise of zero, then the manual signal is larger than the automatic signal. The meter indicates 0-10 volts in both directions. Failure to have the regulator control signals matched when shifting regulator modes may result in transients on the generator output The severity of the transient would be determined by the direction and magnitude of the mismatch. [D-27 Rev. 19 Page22of129

26. 259001 1 Unit One Reactor Feed Pump 1 B is operating in automatic DFCS control at 4500 RPM.

The DFCS control signal to Reactor Feed Pump 1 B woodward governor immediately fails downscale. Which one of the following completes the statement below? Reactor Feed Pump lB speed will: A lowerto0 rpm. B lowertol000rpm. C lower to 2450 rpm. C remain at 4500 rpm. Answer: D KIA: 259001 Reactor Feedwater System Al Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including: (CFR: 41.5 / 45.5) 04 RFP turbine speed: Turbine-Driven-Only RO/SRO Rating: 2.8/2.7 Tier 2 / Group 2 K/A Match: This meets the K/A because it is testing the ability to predict the response in parameters. Pedigree: New Objective: LOl-CLS-LP-032.2, Obj. 13d Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event: Loss of signal interface between controllers and processor.

Reference:

None Cog Level: High Explanation: If RFPT A(B) MAN/DFCS selector switch is in DFCS, and DFCS control signal subsequently drops below 2450 rpm, or increases to greater than 5450 rpm, then Woodward 5009 digital controls will automatically assume RFPT speed control and maintain current pump speed. Distractor Analysis: Choice A: Plausible if the student believes that a loss of input signal will cause the controller to use 0 as the input for the speed of the pump. (i.e. HPCI/RCIC controllers will fail to zero) Choice B: Plausible because an idled REP is maintained at 1000 rpm. Choice C: Plausible because 2450 is the low end of the controller function. Choice D: Correct Answer, see explanation. SRO Basis: N/A

NOTE: If RFPT A(B) M4WDFCS selector switch is in DFCS, and DECS control sgnal subsequently drops below 2450 rpm, or increases to greater than 5450 rpm, then Woodward 5009 digital controls will automatically assume RFPT speed control and maintain current pump speed. un this condition, the RFPT will only respond to LOWERJRAJSE speed control sitch commands until MANIDFCS selector switch is placed in MAN, DFCS GIRL RESET pushbutton s depressed, and MANIDFCS selector switch returned to DFCS. 3 13 Plant management has recommended one RFPT be idled at 1000 rpm with the discharge valve closed, during conditions with one RFPT in service.

27. 259002 1 Which one of the following completes both statements below concerning the reactor feed pump turbine (RFPT) DFCS controls?

During a RFPT startup, transfer to DFCS control is performed when RFPT speed is approximately (1) DFCS will automatically control the speed of the RFPT up to (2) A (1) 1000 rpm (2) 5450 rpm B (1) 1000 rpm (2) 6150 rpm C (1) 2550 rpm (2) 5450 rpm ID (1) 2550 rpm (2) 6150 rpm Answer: C K/A: 259002 Reactor Water Level Control System A3 Ability to monitor automatic operations of the REACTOR WATER LEVEL CONTROL SYSTEM including: (CFR: 41 .7 / 45.7) 01 Runout flow control RO/SRO Rating: 3.0/3.0 Tier 2 I Group 1 K/A Match: This meets the K/A because this is testing the upper limit of the auto (DFCS) controls which in essence prevent pump runout of the reactor feed pumps. Pedigree: new Objective: LOI-CLS-LP-032.2, Obj. 5d Describe the operation of the DFCS in the following operating modes: Master Level Control Mode (auto and manual)

Reference:

None Cog Level: Fundamental Explanation: DFCS will be placed into service with the manual output set at 2550 RPM. The DFCS system will control the RFPT speed from 2450 5450 RPMs

Distractor Analysis: Choice A: Plausible because 1000 RPM is the idle speed of the RFPT and the second part is correct. Choice B: Plausible because 1000 RPM is the idle speed of the RFPT and 6150 is the overspeed setpoint of the woodward controls. Choice C: Correct Answer, see explanation. Choice D: Plausible because the first part is correct and 6150 is the overspeed setpoint of the woodward controls. SRO Basis: N/A

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev 206 Page 50 of 408 6.1.5 Reactor Feed Pump Startup trom Idle Speed to Injection at Low Pressure Conditions (continued) NOTE When using RFPT A(S) LowerRaise speed control switch, reactor feed pump turbine speed will change at a rate of 50 rpm per second if switch is held in LOWER or RAISE for greater than 3 seconds, the rate of change will rise to 3/5 rpm per second

6. Maintain RFPT A(S) thscharqe pressure at least 100 psig greater than reactor pressure by adustftng RFPT A(S) LoweriRatse speed control switch until RFPT speed is approximately 2550 rpm END ftM. LEVEL R3 REACTIVITY EVOLUTION L Direct Radwaste Operator to monitor effluent conductivity for each in service COD -

         &     WHEN RFPTA(B) speed is approximately 2550 rpm.

THEN raise C32-SIC-R601 A(S) [RPT A(S) Sp Ctl] output to match DICS Stpt and Speed SIpt on Panel P603 to within 100 rpm - NOTE

• When RIPI A(S) Man/DFCS control switch is placed in DTCS, C32-SIC-R60 IA(S) [REPT A(S) Sn CtL] wi[L conol REPT speed U
• When RPI A(S) Man/DFCS control switch is placed in UCS, and DTCS is in control, the REPT A(S) DTCS Ctrl light will be ON U
• If REPT A(S) ManDFCS seleor switch is in DECS and DTCS controL signal subseqLiently drops to less than 2450 rpm or rises to greater than 5450 rpm, Woodward 5009 digital controls will automatically assume RIPT speed control and maintain current pump speed In this condition, the RIPT will only respond to LoweriRaise speed control switch commands unl the Man1DrCS selector switch is placed in MAN, DFCS Ctd Reset pushbutton is depressed, and the Man/DCCS selector switch returned to DFCS U

9. Confirm the following REPT A(S) speed signals on Panel P603 agree within 100 rpm:

• DTCS Stpt (speed demand from DECS)
• Speed Stpt (speed demand from 5009 control)
• Act Spd (actual REPT speed)

10. Place Man/DFCS control switch in DICS

. CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERA11NG PROCEDURE Rev 206 Paqe I o 408 3M PRECAUTIONS AND LIMITATIONS (continued) 5 Any o the fcllo.nq nditons .ill automatically trip a mactor reed pump turbine: RFPT Woodard U09 overspoed qreater Than or equal to 6150rpm..

28. 261000 1 Unit One primary containment venting is being performed lAW I OP-i 0, Standby Gas Treatment System Operating System, with the following plant status:

1-VA-i F-BFV-RB, SBGT DW Suct Damper Open i-VA-i D-BFV-RB, Reactor Building SBGT Train iA Inlet Valve Closed 1-VA-i H-BFV-RB, Reactor Building SBGT Train lB Inlet Valve Closed Which one of the following completes both statements below concerning the predicted SBGT response if drywell pressure rises to 1.9 psig? 1-VA-iF-BFV-RB (1) Both 1-VA-i D-BFV-RB and i-VA-i H-BFV-RB (2) A (1) auto closes (2) auto open B (1) auto closes (2) remain closed C (1) remains open (2) auto open ID (1) remains open (2) remain closed Answer: A K/A: 261000 Standby Gas Treatment System A4 Ability to manually operate and/or monitor in the control room: (CFR: 41 .7 / 45.5 to 45.8) 02 Suction valves RO/SRO Rating: 3.1/3.1 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the ability to monitor SBGT suction valves. Pedigree: Last used on 2014 NRC Exam Objective: LOl-CLS-LP-004.1, Obj 5 List the signals and setpoints that will cause a Secondary Containment isolation

Reference:

None Cog Level: High Explanation: The filter train fans will automatically start on High Drywell Pressure. The following actions occur: 1) SBGT Reactor Building suction dampers (1 D-BFV-RB and 1 H-BFV-RB) open, 2) SBGT DW Suct Damper (F-BFV-RB) closes. The SBGT Train NB Suction & Discharge Valves on Ui do not auto open. These valves on U2 do auto open, so there could be a misconception on these valves (inlet vs. suction dampers).

Distractor Analysis: Choice A: Correct Answer, see explanation Choice B: Plausible because 1 F does auto close and SBGT Train lA/B Suction Valves (10 & 1 E) on Unit One only do not auto open Choice C: Incorrect since SBGT will auto realign from primary containment to the Reactor Building on system initiation Choice D: Incorrect since SBGT will auto realign from primary containment to the Reactor Building on system initiation and SBGT Train lA/B Suction Valves (1C & 1E) on Unit One only do not auto open SRO Basis: N/A 2.1.6 Fan A 100% capacity, heavy-duty, industrial type Fan and motor assembly is provided in each SBGT filter train. Each Fan will produce The required 2700 3300 scfm flow through its associated filter train. Each Fan is driven by a direct-drive AC motor which is energized from a redundant and separate emergency power supply. The Unit 1 A and B Fans are powered from 480 VAC MCCs IXE and 1 XF respectively and Unit 2 A and B Fans from 2XE and 2XE The fitter train fans may be operated manually from controls located at RTGB XU-51. The filter train fans will automatically start if any ot the following Secondary Containment isolation conditions exist: (Figure 10-2) 1 Low Reactor Water Level, LL #2

2. High Drywell Pressure

3. Reactor Building Ventilation Radiation (Figure 10-3) 3.2.6 Automatic

1. Upon receipt of an automatic initiation signal both trains of SBGT will start.

Unit 1 ONLY The dampers associated with Unit I SBGT System will receive automatic open signals when an initiation signal is received EXCEPT for the train inlet and outlet dampers, fBFV5-t B,1 0,1 Eand 1 C). Should these normally open dampers be manually closed focally via their CLOSEIOPEN pushbuttons, they will NOT automatically reopen and the associated SBGT will not automatically start. SD-b Rev.7 Page 1601 38

29. 262001 1 Unit One and Unit Two are both operating at rated power.

Which one of the following completes both statements below lAW Unit One Tech Spec 3.8.1, AC Sources Operating, LCO statement? The Unit Two SAT (1) required to be OPERABLE. (2) Diesel Generators are required to be OPERABLE. A (1) is (2) Two B (1) is (2) Four C (1) isNOT (2) Two ID (1) i5NOT (2) Four Answer: B K/A: 262001 A.C. Electrical Distribution G2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 /43.2/43.5 /45.3) RO/SRO Rating: 3.4/4.7 Tier 2 / Group 1 K/A Match: This meets the K/A because this is testing the items above the line for TS 3.8.1. Pedigree: New Objective: LOl-CLS-LP-050, Obj. 16 Given plant conditions, determine whether given plant conditions meet minimum Technical Specifications requirements associated with the 230 KV Electrical Distribution system.

Reference:

None Cog Level: Fundamental Explanation: With both Units in Mode 1, both SATs and both UATs and all four DGs are required to be operable.

Distractor Analysis: Choice A: Plausible because the first part is correct and there are only two Unit One DGs but all four are required for the LCO to be met. Choice B: Correct Answer, see explanation. Choice C: Plausible because this is the Unit 2 SAT and asking if it is required for Unit 1 TS and whether only the 2 Unit One DGs are required or all of the DGs. Choice D: Plausible because this is the Unit 2 SAT and asking if it is required for Unit 1 TS and the second part is correct. SRO Basis: N/A AC SourcesOperaUng 3.8 1 3.8 ELECTRICAL. POWER SYSTEMS 3.8 1 AC SourcesOperating LCO 3.8.1 The following AC electrical por sources shall be OPERABLE a, Two Unit 1 qualitied circuits between the oftsite transmission network and the onsite Class 1E AC Electrical Power Distribution System.,

b. Four diesel generators (DGS), and c Two Unit 2 qualified circuits between the ottsite transmission network and the onsite Class JE AC Electrical Power Distribution System.

APPUCABIUTY. MODES 1, 2. and 3. AC11ONS NOTE LCO 3.O,4.b is not applicable to DGs.

30. 262002 1 Unit One is operating at rated power.

Subsequently, breaker AU9, Feed to 480V Substation E5 trips. Which one of the following completes the statement below? 120V UPS Distribution Panel 1A is: A de-energized. B energized from MCC lOB. o energized from the Standby UPS. D energized from 250V DC SWBD A. Answer: D K/A: 262002 Uninterruptable Power Supply (A.C. /D.C.) A3 Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (AC/D.C.) including: (CFR: 41.7/ 45.7) 01 Transfer from preferred to alternate source RO/SRO Rating: 2.8/3.1 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the ability to monitor the transfer to the alternate power source. Pedigree: New Objective: LOl-CLS-LP-052, Obj. 5 Given plant conditions, determine the lineup of the primary UPS, the Standby UPS, and their reserve sources.

Reference:

None Cog Level: High Explanation: The UPS system is normally aligned such the primary inverter is powering UPS loads. The standby inverter is energized but bypassed with the Manual Bypass switch in Bypass Test. The static transfer switch of the Primary inverter (and also the Standby inverter) is receiving an input from the alternate (hard) source. If the primary power source is lost (in this case the loss of E5 which powers MCC CA) the alternate power source from the 250V batteries will keep the loads energized with no need for the inverter to swap to the hard source.

Distractor Analysis: Choice A: Plausible because the normal power source is lost. Choice B: Plausible because this is the hard source for the Distribution Panel. Choice C: Plausible because this is an available power source for the Distribution Panel. Choice D: Correct Answer, see explanation. SRO Basis: N/A FIGURE 52-7 Basic Vital UPS System NORMAL SOURCE ALTERNATE SoURCE 1PriV AC Ph A ]Th kv

                                                                           .1A;    #cv 5CKVA STANDBY UPS 113                                       I RECTIFIER        INVERTER       STATICS ITCH                )N. C or or: :.BI             ELOCKINri tIOtE I                                                    MANUAL BYPASS Sw, I L                                                                         J
    \OR.L SOR:E 3D 6DHA I                                                                      ALTERNATE I                                                                      SOURCE UORVA PRIMARY        IS   (Pa) r H
                  -            -    INVERTER        STATIC SV]TEH  -
                     ;h1HTBl_

dLD t; S_C PP ULD::KIN: IIIDIJE I MANUAL BYPASS S I L J SYSTEM OUTPUT D2/?U3V AC JPh 60Hz

31. 2630001 Unit Two is operating at full power when a loss of DC Distribution Panel 4B occurs.

Which one of the following completes both statements below? RCIC is (1) for injection from the RTGB. RCIC (2) isolation logic has lost power. A (1) available (2) inboard B (1) available (2) outboard C (1) unavailable (2) inboard 0 (1) unavailable (2) outboard Answer: D K/A: 263000 D.C. Electrical Distribution G2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5/45.12) ROISRO Rating: 3.6/4.6 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the operability of RCIC/ADS. Pedigree: New Objective: LOl-CLS-LP-051, Obj. 7 Given plant conditions, determine the effect that a loss of DC power will have on the following:

d. Reactor Core Isolation Cooling.

e. Automatic Depressutization System.

Reference:

None Cog Level: High Explanation: RCIC initiation, trip logic, governor control and outboard isolation logic is powered from Division 11125 VDC panels 4B for unit 2. Without this power source RCIC cannot start either automatically or manually and Div II isol valves will not auto isolate

Distractor Analysis: Choice A: Plausible because the student may believe that the logic has a backup power supply or confuse the HPCI and RCIC power supplies. Part 2 is plausible because HPCI inboard isolation logic would be inoperable. Choice B: Plausible because the student may believe that the logic has a backup power supply or confuse the HPCI and RCIC power supplies. Part 2 is correct, see explanation. Choice C: Part 2 is correct, see explanation. Part 2 is plausible because HPCI inboard isolation logic would be inoperable. Choice D: Correct Answer, see explanation. SRO Basis: N/A LOSS OF DC POWER OAOP-39.O Rev. 042 Page 32 ot 34 ATTACHMENT 6 Page 1 ot2

                    <<Plant Effects from Loss of DC Distribution Panel 38(48)>>

RCIC: Will NOT auto initiate, outboard isolaton logic INOPERABLE (E51 -FOOB, -F029, and -F066 wiLl NOT auto close), RCIC turbine will NOT tnp except on overspeed. RCIC flow controller and EGM INOPERABLE (no flow control or indication), E51 -F045 wil NOT auto close on high water level, E51-F004, -F054, and -F026 fail closed. RCIC isolation is required in accordance with APP 1f2)-A-031-4.

32. 264000 1 Unit Two has lost off-site power.

DG3 started and tied to its respective E Bus. Sequence of events: 1200 DG3 ties to E3 1205 DG3 lube oil temperature rises above 190°F 1206 DG3 lube oil pressure drops below 27 psig Which one of the following identifies when DG3 will trip? A Immediately at 1205. B Immediately at 1206. C 45 seconds after 1205. O 45 seconds after 1206. Answer: B K/A: 264000 Emergency Generators (Diesel/Jet) K6 Knowledge of the effect that a loss or malfunction of the following will have on the EMERGENCY GENERATORS (DIESEL/JET): (CFR: 41.7 / 45.7) 03 Lube oil pumps RO/SRO Rating: 3.5/3.7 Tier 2 / Group 1 K/A Match: This meets the K/A because it is testing the effect of a loss of lube oil on the EDG. Pedigree: Bank Objective: LOI-CLS-LP-039, Obj. 4a Given plant conditions, determine if EDGs will trip: After an auto start (LOCT)

Reference:

None Cog Level: High Explanation: Hi lube oil temperature bypassed by auto start signal (LOOP and LOCA). Low lube oil pressure trip never bypassed. On a start of the DG the low lube oil trip is bypassed for 45 seconds. Distractor Analysis: Choice A: Plausible because hi lube oil temperature is a trip, but it is bypassed on the LOOP. Choice B: Correct Answer, see explanation. Choice C: Plausible because there is a 45 second time delay associated with the lube oil trip on an initial start of the EDG. Choice D: Plausible because there is a 45 second time delay associated with the lube oil trip on an initial start of the EDG. SRO Basis: N/A

SD-39 EMERGENCY DIESEL GENERATORS SYSTEM DESCRIPTION Rev. 20 Page 46 of 166 333 Automatic Stop Control (Figure 3914) Under conditions where continued Diesel Generator operation may cause damage to The Diesel itselt automatic shutdors are provided The shutdown signals will vary dependent upon whether the engine has been started manually or automatically. When operating due to receipt of an automatic start signal the following trips and k)ckout are provided:

• Low lube oil pressure 27 psig a Overspeed 575 (561 to 569) rpm When operating as a result of an initiation from a normal non-emergency start the following ffips and lockouts are enforced in add Won to those listed above:
• High lube oil temperature 19OCF O High jacket water temperature 200CF a JacketWaterLow pressure 12 psig The low lube oil pressure, and low lacket water pressure shutdowns are blocked br the first forty-five second on initiation of an engine start sequence (auto or rnanual. This permits the conditions to be established which wilt prevent these shutdowns during engine operation

33. 272000 1 Unit Two is performing a startup lAW OGP-02, Approach to Criticality and Pressurization of the Reactor.

lAW OGP-02, which one of the following identifies the radiation monitor(s) that will require the alarm setpoints raised when HWC is placed in service? A D12-RM-K603A,B,C,D, Main Steam Line Rad Monitors B ARM Channel 2-9, U-2 Turbine Bldg Breezeway C D12-RR-4599-1,2,3, Main Stack Rad Monitors D ARM Channel 2-4, Cond Filter-Demin Aisle Answer: A K/A: 272000 Radiation Monitoring System K5 Knowledge of the operational implications of the following concepts as they apply to RADIATION MONITORING SYSTEM: (CFR: 41.7/45.4) 01 Hydrogen injection operations effect on process radiation indications RO/SRO Rating: 3.2/3.5 Tier 2 / Group 2 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL. This meets the K/A because this is testing the operational implication as to which tad monitor, if asked as the operational effect on the individual rad monitor this would provide no discrimatory value. Pedigree: New Objective: LOI-CLS-LP-059, Obj. 8 Explain why Chemistry must be notified when starting and securing the HWC System.

Reference:

None Cog Level: Fundamental Explanation: The excess Hydrogen injected into the reactor coolant creates the driving force to shift the Nitrogen-16 distribution ratio, resulting in a larger fraction of the Nitrogen-16 forming volatile Ammonia and a smaller fraction forming Nitrites and Nitrates. This additional volatile Ammonia is then carried over in the reactor steam resulting in higher background radiation levels. Any increase in Hydrogen injection rates will result in a proportional increase in background radiation levels and vise-versa. OGP-02 has a step for ensuring that the tad monitors are adjusted based on this background tad level increase.

Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because when HWC is placed in service the rad levels will increase minimally and HWC H2 is injected in the reactor feed pumps. Choice C: Plausible because sufficient decay time is available for N-16 such that radiation levels wouldnt raise that much in this area. Choice D: Plausible because when HWC is placed in service the rad levels will increase minimally and this is downstream of the HWC 02 injection point. SRO Basis: N/A APPROACH TO CRFIICALITY AND PRESSURIZA11ON OGP-02 OFTHE REACTOR Rev. 110 Page 6 of 54 3.0 PRECAUTIONS AND LIMITATIONS (continued)

15. B21-t032A and B21-F032B(Feedwater Supply Line Isolation Valves), are stop-chock valves. Those valves are designed to prevent leakage from the reactor into the feedwater system. These valves are not designed to positively close against condensate system pressura. As such, with the reactor deprossurizod and the condensate system in sentce, these valves may teak by, causing reactor water level to rise

16. The Main Steam Line Radiation Monitor (MSLRM) Htqh-Hiqh Radiaon setpoint is adjusted assuming HWC is in service If HWC is removed from service for an extended period of time (greater than one week). 1(2)OP-59, Hydrogen Water Chemistry System Operating Procedure requires BESS determine if a MSLRt1 High-High Radiation sotpoint adjustment is required D

17. The HWC System will normally be placed in service immediately after establishing the following conditions:

• At least one Condensate Booster Pump feeding the reactor with minimum flow valve closed 0
• At least one SJAE operating at greater than or equal to half-load U

34. 286000 1 Which one of the following identifies the distribution system that provides the normal power supply to the Unit Two Reactor Building Fire Alarm Control Panel?

A 48VDC B I2OVAC C 125 VDC D 480 VAC Answer: B K/A: 286000 Fire Protection System K2 Knowledge of electrical power supplies to the following: (CFR: 41 .7) 03 Fire detection system ROISRO Rating: 3.6/3.8 Tier 2 / Group 2 K/A Match: This meets the K/A because this is testing the power supply to fire detection. Pedigree: New Objective: LOl-CLS-LP-042, Obj. 7 Identify the electrical distribution system which supplies power to the fire/smoke detection circuits.

Reference:

None Cog Level: Fundamental Explanation: The Detection Systems control panels are powered from 120 VAC (reference Table 42-1) and convert the AC power to DC power for the rest of the system. The Detection Systems control panels are powered from 120 VAC; except Caswell Beach which is 125 VDC. Distractor Analysis: Choice A: Plausible because the plant has a 24/48 VDC system and the panel converts the incoming power supply to a DC source. Choice B: Correct Answer, see explanation Choice C: Plausible because the Caswell Beach fire detection is supplied from 125 VDC. Choice D: Plausible because 480 VAC could easily be thought to supply the power based on DG supply in emergencies. SRO Basis: N/A

TABLE 42-i Page 1 012 Fire Detection Control Panel Power Supplies COMPONENT POWER SUPPLY Control Building flre Alarm Control Panel 120 VAC Distilbutiori Panel 2G-CB Radwaste Building Fire Alarm Control Panel 120 VAC Distflbution Panel DRWD Unit 2 Reactor Building Fire Alarm Control 120 VAC Distiibution: Panel 2C-RX Panet Unit I Reactor Building Fire Alarm Control Emergency 120 VAC Distribution Panel Panel JB-RX Unit 2 Turbine Building Fire Alarm Control 120 lAO Distilbulion Panel 2C-TB2 Panel Unit I Turbine Building Fire Alarm Control 120 VAC Distiibution Panel IC-TB2 Panel

35. 295001 1 Unit One is operating at 70% power when the OATC observes indications for a failed jet pump. Subsequently, Recirc Pump 1A trips.

Which one of the following completes both statements below lAW 1AOP-04.0, Low Core Flow? Performance of the jet pump operability surveillance for (1) Loop Operation is required. If it is determined that a jet pump has failed, the required action is to (2) A (1) Single (2) reduce reactor power below 25% rated thermal power B (1) Single (2) commence unit shutdown lAW OGP-05, Unit Shutdown o (1) Two (2) reduce reactor power below 25% rated thermal power D (1) Two (2) commence unit shutdown lAW OGP-05, Unit Shutdown Answer: B K/A: 295001 Partial or Complete Loss of Forced Core Flow Circulation G2.2.12 Knowledge of surveillance procedures. (CFR: 41.10/ 45.13) RO/SRO Rating: 3.7/4.1 Tier 1 / Group 1 K/A Match: This meets the K/A because this is testing knowledge of which surv. is required and the action if it has failed the surv. Pedigree: New Objective: LOl-CLS-LP-302-C, Obj 4 Given plant conditions and AOP-04.0, determine the required supplementary actions.

Reference:

None Cog Level: High Explanation: The indications given are for a failed jet pump which lAW the AOP require the surveillance performed for determination of a failed jet pump. Unlike the selection of the power to flow map the PT only looks at the recirc pumps for determination of single loop or two loop operation. The power to flow maps for single loop are not used until the APRM setpoint adjustments are made.

Distractor Analysis: Choice A: Plausible because single loop is correct and 25% is the requirement for when the PT is required to be performed. Choice B: Correct Answer, see explanation. Choice C: Plausible because APRM setpoint adjustments have not been made which is a determination of how to use the power to flow maps and 25% is the requirement for when the PT is required to be performed. Choice D: Plausible because APRM setpoint adjustments have not been made which is a determination of how to use the power to flow maps and the second part is correct. SRO Basis: N/A

LOW CORE FLOW 2AOP-04.O Rev. 3? Page 180125 4.2 Supplementary Actions (continued) NOTE Jet pump failure is indicated by the following E

• Reduction En generator megawatt output on GEN-WMR-760 (Net UnJt Megav.tts)
• Reduction in core thermal power
• Rise in indicated total core 110w on 621-R6 113 (Core A PressureJCore Flow) recorder
• Reduction in core plate differential pressure on 621-R613 (Core A Pressure/Core Flow) recorder
• Rise in recirculatton toop flow in the loop with a tailed let pump on 632-R6 14 (Recirculation Flow) recorder CAUTI ON Under conditions of jet pump failure, indicated oore flow on Process Computer Point U2CPWTCF and B2f-R613 (Core A Pressure/Core Flow) recorder, w[lt NOT be accurate. Accurate core flow is available from Process Computer Point U2NSSWDP (Core Plate Differential Pressure) or Attachment 1, Estimated Total Core Flow vs. Core Support Plate Delta P for B2C22. Until Step 23.b( t), the operating point on the Power-to-Flow Map will 4QI be accurate. Indicated total core flow on B21-R6 13 (Core A Pressttre/Cote Flow) recorder wilt continue to be inaccurate until the failed jet pump is repaired

23. iEet pump failure is suspected, THEN perform the following:

a. IF reactor power is greater than or equal to 25%,

THEN ensure the following:

• OPT-I 3:!, Reactor Recirculation Jet Pump Operability, is perFormed tor two loop operation l1 OR
• OPT-i 3.4, Reactor Recirculation Jet Pump Operability for Single Loop Operation, is perfomled for single loop operation

b IF any let pump is determined to be INOPERABLE. THEN perform the following:

1) Ensure the input to the Power-to-flow Map has been changed from WTCF to core plate differential pressure (2) Notify the Duty Reactor Engineer the input to the Power-to-Flow Map has been changed from WTCE to core plate differenUat pressure (3) Commence unit shutdown in accordance with OGP-05, Unit Shutdown, in compliance with Technical Specification 3.4.2

36. 295003 1 Unit One is operating at rated power.

The load dispatcher reports degraded grid conditions with the following indications: Generator frequency 59.7 hertz 230 KV Bus 1A voltage 205 KV 230 KV Bus lB voltage 205 KV El voltage 3690 volts E2 voltage 3685 volts Which one of the following completes both statements below? The (1) may be damaged with continued operation under these conditions. lAW OAOP-22.0, Grid Instability, a reactor scram and turbine trip (2) required. A (1) mainturbineblades (2) is B (1) main turbine blades (2) is NOT C (1) emergency bus loads (2) is O (1) emergency bus loads (2) is NOT Answer: D K/A: 295003 Partial or Complete Loss of A.C. Power AK1 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF AC. POWER: (CFR: 41.8 to 41.10) 03 Under voltage/degraded voltage effects on electrical loads RO/SRO Rating: 2.9/3.2 Tier 1 / Group 1 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL. This meets the K/A because this is testing the degraded voltage conditions. Pedigree: Last used on the 10-2 NRC Exam Objective: LOI-CLS-LP-302-G, Obj. 4b Given plant conditions and any of the following AOPs, determine the required supplemental actions: AOP-22.0, Grid Instability

Reference:

None Cog Level: High

Explanation: There are frequency based criteria in AOP-22.0 (Caution directly preceding step 3.2.1) for tripping the turbine to prevent resonance vibration of low pressure blades due to off frequency operation. Time limits include, 5 minute ranges and 1 minute ranges. At this current frequency, the Main Turbine can be operated indefinitely, which will not cause turbine damage. Sustained low voltage provides for higher running currents which will damage running ESF motors. Per the automatic actions section of AOP-22.0, the degraded voltage relays will actuate when emergency bus voltage has dropped below 3700 VAC for 10 seconds. This trips the Master/Slave breakers (BOP bus supply to E Buses) and the DGs start and load. Distractor Analysis: Choice A: Plausible because turbine blade damage can occur due to off frequency operation and AOP-22.0 does have a low frequency range limit to scram the reactor and trip the turbine. Choice B: Plausible because turbine blade damage can occur due to off frequency operation and the second part is correct. Choice C: Plausible because damage to E Bus loads is correct and AOP-22.0 does have a low frequency range limit to scram the reactor and trip the turbine. Choice D: Correct Answer, see explanation SRO Basis: N/A III II.I II..,.1I I il UI I..UIL ...UI tI It UI LUUft r,JVI...). 2.4 Protective Relaying Protective reayng is designed to solate any faulted component or portion of the electrical system, while mantaining continuity of power to the unfaulted portion of the system. The most corn many used protective devices inclLide:

1. Undervotage (27 Device) Relays. CJndervotage relays actuate on a low voltage condition, and usually are time delayed to account for mornentar. transient conditions, such as fault cieanng and bus transfers. The degraded grid voltage relays are provided with a substantally longer time delay to prevent actuation due to motor starting transients. Undervoltage relays provide a variety of protective functions including supply breaker tnps and closure permissives, large motor breaker trips and ciosure permissives, and automatc starting of the Emergency Dese I Generators SD-50.1 Rev 19 Page27of 131

4.2 Abnormal Operation 4.2.1 Abnormal Frequency Conditions When system frequency reaches 59.8 hertz, Annunciator, UA-0S. window 1-2. GEN BUS UNDER FREG RELAY is activated. Operators are di-ected to respond per AOP-2Z0, Generator Abnormal Frequency Conditions. This is done to stabi[ze loads on the system. One of the most pobable causes of an under frequency conthion would be the loss o another large generating unit or units, when the on-line reserie capacity is nadequate for curent system loads. Rapid response and close coordinaton with the load dispatcher are required to ensure system stabliN. Abnormal frequency operaton can develop resonant frequences that may induce vbations in the low pressure turbine blades. The vibration can cause turbine bades to fatgue and possibly fail dunng operation. The effect increases propornonal.y in relation to the magnitude of the frequency difference, and the length of time at the abnormal frequency. SD-27 Rev. 15 Page 51 of 127 GRID INSTABILITY OAOP-22.0 Rev. 27 Page 5 01 14 3.0 AUTOMATIC ACTIONS

1. ffl emergency bus voltage has lowered to less than 3700 volts (approximately equal to SOP bus voltage) ror greater than 10 seconds, THEN the masterislave breakers to the E bus open and associated diesel generator starts and loads C

GRID INSTABILITY DAOP-22.0 Rev. 27 Page 6 of 14 4.2 Supplementary Actions NOTE A sudden rise in system frequency may be observed due to additibnal generation or load shedding. Automatic load shedding (110% of system Ioad occurs at each of the following frequencies: 59.3. 59.0, and 58.5 CAUTION The maximum allowable time at a given frequency is as foows E

• Below 58.1 Hz, operation is prohibited
• Between 58.1-58.5 Hz. operation for I minute is atlowed
• Between 58.6-59.3 Hz, operation for 5 minutes s allowed a Between 59.4-60,6 Hz, operation is allowed indelinitely
• Between 60.7-61.4 Hz, operation totS minutes is allowed
• Between 61.5-61.9 Hz, operation for 1 minute is allowed
• Above 61.9 Hz, operation is prohibited CAUTION a Off-frequency operation can stimulate resonance vibration in low pressure blades 0
• A total loss of on-site power (LOOP) should be anticipated ii the turbine is tripped 0
• With gild voltage or frequency unstable or grid vulnerability identified, diesel generators should fI be paralleled with any E bus connected to the grid since severe load ings may occur and possibly overload the diesel generators 0

1. IF the maximum allowable time at a given frequency is exceeded, THEN perform the following:

a. ffl reactor power is greater than or equal to 26%,

THEN insert a manual scram 0

b. Trip the main turbine 0

c. IF the unit was scrammed.

THEN enter lEaP-C 1-RSP2EOP-01-RSP), Reactor Scram Procedure 0

GRID INSTABILITY OAOP-22.0 Rev. 27 Page 901 14 4.2 Supplementary Actions (continued)

10. IF system frequency is high, THEN:

a. Establish communication with the Load Dispatcher 0 b, Continue unit generatiion as directed by the Unit CR5 coordinate with the Load Dispatcher 0 C. IF tripping the turbine becomes imminent, THEN rapidly reduce power in an attempt to tower frequency to less than 601 Hz prior to t1pping the main turbine 0

11. IF notified by the Load Dispatcher system voltage is unable OR will be unabte to support a LOCA, OR abnormal frequency conditions persist.

THEN follow the guidelines in 0OIt0J, BNP conduct of Operations Supplement 0

12. IF any diesel generator is toaded to an E bus connected to the grid, THEN restore The diesel generator to standby in accordance with applicable procedures 0

13. IF system voltage is less than 3700 volts for greater than 10 seconds, THEN ensure:

The affected E bus master/slave breakers OPEN 0 The affected diesel generator starts and loads LI

37. 295004 1 Which one of the following completes both statements below?

lAW OAOP-39.0, Loss of DC Power, before 125 VDC battery voltage reaches (1) remove loads as directed by the Unit CRS. lAW 1 EOP-01 -SBO, Station Blackout, if either division battery chargers can NOT be restored within (2) then load strip the affected battery. A (1) lO5volts (2) 1 hour B (1) lOSvolts (2) 2 hours C (1) 129 volts (2) 1 hour D (1) 129 volts (2) 2 hours Answer: A K/A: 295004 Partial or Complete Loss of D.C. Power AK2 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following: (CFR: 41 .7 / 45.8) 01 Battery charger RO/SRO Rating: 3.1/3.1 Tier 1 / Group 1 K/A Match: This meets the K/A because this is testing knowledge of the relationship between the loss of DC power and time requirement to re-energize the battery charger. Pedigree: New Objective: LOI-CLS-LP-051, Obj. 14 Describe the consequences/problems associated with the following: a. Battery chargers remaining out of service during a loss of off-site power / station blackout.

Reference:

None Cog Level: Fundamental Explanation: AOP-39.0 directs to load strip before reaching 105 VDC to prevent cell reversal. The alarm for undervoltage comes in at 129 VDC. The station Blackout procedure states that if the battery charger is not energized in 1 hour to load strip the batteries. There is a time critical 2 hour action in the SBO procedure for opening the Reactor Building doors.

Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because the first part is correct and the second part is a time critical action time limit in the SBO procedure. Choice C: Plausible because 129 volts is the annunciator setpoint for the batteries and the second part is correct. Choice D: Plausible because 129 volts is the annunciator setpoint for the batteries and the second part is a time critical action time limit in the SBO procedure. SRO Basis: N/A LOSS OF DC PYi.ER 0AOP-39.I Rev 41 Page 7 of 36 4.2 Supplementary Actions Loss of Batteiv Chargers: a Monitor 125V and 24V DC battery voltages C

b. IF power has been removed from the battery chargers for greater than 1 hour, THEN remove selected loads trom the battery based on 001-50, 1251250 and 24/48 VDC Electrical Load List and Unit CRS direction C

c. Before 1 25V DC battery voltage reaches the tow voltage limit of 105 volts, remove loads as directed by tile Unit CRS as necessary to maintain battery voltage greater than 105 volts C U. Betore 24V battery voltage reaches the low voltage limit 012 I volts, remove loads as directed by the Unit CRS as necessary to maintain battery voltage greater than 21 volts C

e. IF battery charger AC power has been lost due to Station Blackout, THEN enter 1 EOP-0 i-SBO(2EOP-01 -SBO), Station Blackout C

(it IE1 1i$.TT tEN 3rr5 C.NNcT te THEN l.3a4 r1p ie1 t3flr ec EDP-1-SE)-it

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         \      THEN        /

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                                    .3 rirhfl reJreoDet et,i, 2hc*jra 5t9e eneiii ço Tuçe jpirt per EQI4]I EP12.

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38. 295005 1 Which one of the following identifies the reason an operator is directed to trip the main turbine as an immediate action lAW OAOP-32.0, Plant Shutdown From Outside Control Room?

A To initiate a scram on TSVITCV closure. B To prevent cold start challenges to Diesel Generators. C The turbine cannot be tripped once the Control Room is evacuated. O To bring bypass valves into operation until Remote Shutdown Panel control is established. Answer: B K/A: 295005 Main Turbine Generator Trip AK3 Knowledge of the reasons for the following responses as they apply to MAIN TURBINE GENERATOR TRIP: (CFR: 41.5/45.6) 04 Main generator trip RO/SRO Rating: 3.2/3.2 Tier 1 / Group 1 K/A Match: This question requires the operator to have knowledge of the reason for turbine/generator trip. AOP-32 was used to include plausibility of distractors. Pedigree: Bank Objective: LOI-CLS-LP-302E, Obj. 6 Given plant conditions and entry into OAOP-32.0, Plant Shutdown From Outside Control Room, explain the basis for a specific caution, note, or series of procedure steps.

Reference:

None Cog Level: Fundamental Explanation: Following a reactor scram, the turbine control valves throttle shut in an effort to control RPV pressure at the setpoint of 928 psig. Without operator action, the turbine control valves will fully close, causing the generator to motor. Reverse power on the generator will cause a generator primary lockout and auto start of the diesel generators. The main turbine is therefore manually tripped to prevent it from automatically tripping on generator reverse power. This also reduces the number of cold start demands on the diesel generators. Distractor Analysis: Choice A: Plausible because a reactor scram is inserted as a step in the AOP, but it is performed earlier. Choice B: Correct Answer, see explanation Choice C: Plausible because the procedure states to perform the step prior to exiting the control room but it could still be done at the turbine front standard. Choice D: Plausible because this would allow use of the bypass valves, but MSIVs are manually closed prior to leaving the control room. This brings SRVs into operation. If MSIVs are not closed prior to leaving the control room, RPS EPA breakers are opened prior to establishing control at Remote Shutdown panel, which would close MSIVs.

SRO Basis: N/A I REACTOR SCRAM PROCEDURE I 001-37.3 I II BASIS DOCUMENT Rev.O16l I Page7of23 5.2 Step RSP-2 I P*rforn scram iminediaw aclthnz Step RSP-2 includes the potential for multiple sensor and sensor relay fadures in the automatic RPS logic where an automatic reactor scram should have initiated but did not. If needed a manual scram is inserted to accomplish an automatic action wtich should have taken place. A manual reactor scram is also required when directed from other EOP5 and no condition exists which would have automatically initiated a reactor scram (e.g.. entry from PCCP because of high torus temperature). Step RSP-2 also addresses other Reactor Operator scram immediate actions and includes:

• ARI initiation is an additional means of inserting control rods if needed.
• Placing the reactor mode switch to shutdown. When the reactor mode switch is placed in SHUTDOWN position, a diverse and redundant reactor scram signal is generated by the RPS logic. If the mode switch is taken out of RUN prior to RPV pressure decreasing to 835 psig, the MSIV closure due to low main steam line pressure is prevented.

For Unit 2 only, if the mode switch is taken out of RUN vvtien steam flow is above 33%, the MSIVs will close. Therefore, for Unit 2 the mode switch is placed in SHUTDOWN after steam flow is below 3x1 O lb/hr.

• Following a reactor scram, the turbine control valves throttle shut in an effort to control RPV pressure at the setpoint of 928 psig. Without operator action, the turbine control valves will fully close, causing the generator to motor. Reverse power on the generator will cause a generator primary lockout and auto start of the diesel generators. The main turbine is therefore manually tripped to prevent it from automatically tripping on generator reverse power. This also reduces the number of cold start demands on the diesel generators.

39. 295006 1 Unit One has entered RSP with the following conditions:

Six control rods are at position 02, all others are fully inserted B Recirc Pump has tripped Which one of the following completes both statements below? The control rods will be inserted by (1) lAW OEOP-01-LEP-02, Alternate Control Rod Insertion. After the control rods are inserted, a CRD flow rate of approximately (2) will be established. A (1) placing the individual scram test switches to the Scram position (2) 30 gpm B (1) placing the individual scram test switches to the Scram position (2) 45 gpm C (1) driving rods using RMCS (2) 30gpm O (1) driving rods using RMCS (2) 45 gpm Answer: C K/A: 295006 Scram AM Ability to operate and/or monitor the following as they apply to SCRAM: (CFR: 41.7 / 45.6) 06 CRD hydraulic system RO/SRO Rating: 3.5/3.6 Tier 1 / Group 1 K/A Match: This meets the K/A because this is testing operation of CRD controls after a scram. Pedigree: new Objective: LOl-CLS-LP-300-C, Obj. 10 Given plant conditions and the Reactor Scram Procedure, determine the required operator actions

Reference:

None Cog Level: High Explanation: Even if the reactor will remain shutdown under all conditions without boron the LEP is used to insert the control rods using RMCS. If more control rods were out then the scram test switches would be an option. If a recirc pump is tripped then CRD flow is set to 30 gpm to minimize the stratification in the bottom head region.

Distractor Analysis: Choice A: Plausible because this is an option used to insert the control rods in the LEP. The second part is correct. Choice B: Plausible because this is an option used to insert the control rods in the LEP. The second part is the nominal setting for the CRD flowrate. Choice C: Correct Answer, see explanation. Choice D: Plausible because the first part is correct and the second part is the nominal setting for the CRD flowrate. SRO Basis: N/A

7. WHEN either:

• Afl control rods in El RO
• Qn! one control rod NOT tulflly inserted El RO
• NO mote than 10 control rods withdrawn to position 02 AND NO control rod withdrawn beyond positbn 02 U RO
• Reactor engineering has determined the reactor will remain shutdown under all conditions without boron U RO THEN perForm Section 2.2, Control Rod Insertion Verification on Page 7 U RO

10. W.Jjy control rod NOT fully inserted, THEN insert control rods:

a Record in Control Room log the control rod number and position oty rods NOT fully inserted RO ALTERNATE CONTROL ROD INSERTION OEOP-0i -LEP-02 Rev. 029 Page 120137 2.2.3 Control Rod Verification Actions (continued)

b. Bypass RWM RO c Insert control rods with Emergency Rod In Notch Override switch RO

ALTERNATE CONTROL ROD INSERTION OEOP-01-LEP-02 Rev. 029 Page 24 0137 2.6.3 Scram Individual Control Rods Actions (continued)

10. Unit 1 Only: Insert control rods with indMdual scram test swlitches:

a. Identity y control rod NOT inserted to or beyond Position 00 RO NOTE

• A Sound powered phone jack is located on the column beside Panel XU-76 and in Panels XU-12, 58, 49 and 61 D
• The preferred sound-powered phone switchboard bus for use is Bus 1 0

b. Establish communication between Panel P610 and Control Room 0 RO NOTE The individual scram test switch SCRAM position is dowr 0 C. Place individual scram test switch to SCRAM position for control rod NOT inserted to or beyond Position 00 0 RO

LWRNATE CONTROL ROD INSERTION OEOP-01-LEP-02 Rev. 029 Page iii f 37 2.23 Control Rod Verification Actions continued) (3) IF both CRD pumps mnniing. THEN stop one CRD pump RD (4) Set the setpont tape on Cl l(C12)-FCR00 (CR0 Flow Control) to 30 gpm RD NOTE The adhons in Section 2.2.3 Step 9.h(5) may be repeated as necessary C (5) Adjust cooling water differential pressure. CRD flow rate and drtve pressure:

• Cl 1(Ci2)-FC-R600 (CR0 Flow Control: to maintain cooling water difrerential pressure between 10 and 26 psid RD
• if a reactor recirculation pump is tripped, THEN establish a CRD flow rate of approximately 30 gpm RD
• IF both reactor recirculation pumps running, THEN establish a CRD flow rate beeen 30 and 60 gpm RD

40. 295009 1 A total loss of Unit One feedwater results in reactor water level lowering to 87 inches.

Drywell pressure is 2.1 psig. Reactor water level is being restored with RCIC and CRD. Which one of the following completes both statements below? RVCP (1) required to be entered. The expected response of the G31 -FOOl, Inboard RWCU Isolation Valve, and the G31-F004, Outboard RWCU Isolation Valve, is that (2) should be closed. A (1) is (2) ONLY the G31-F004 B (1) is (2) BOTH C (1) is NOT (2) ONLY the G31-F004 ID (1) is NOT (2) BOTH Answer: B KIA: 295009 Low Reactor Water Level AK2 Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following: (CFR: 41 .7 /45.8) 04 Reactor water cleanup RO/SRO Rating: 2.6/2.6 Tier 1 / Group 2 K/A Match: This meets the K/A because this is testing the LL2 relationship to Group 3 (RWCU) isolation. Pedigree: New Objective: LOI-CLS-LP-01 4, Obj 8 Given plant conditions, determine if the RWCU system should have isolated, including expected changes in RWCU System components

Reference:

None Cog Level: High Explanation: Based on conditions RVCP should be entered. By knowing the entry conditions for RVCP (2# DW pressure) this eliminates the RSP. The low level condition will isolate the FOOl and F004. There are some signals that will isolate only the F004 only.

Distractot Analysis: Choice A: Plausible because the first part is correct and some of the Group 3 signals do only close the F004. Choice B: Correct Answer, see explanation. Choice C: Plausible because the RSP would be entered, but there is an entry condition for RVCP (2# in the DW). Some of the Group 3 signals do only close the F004. Choice D: Plausible because the RSP would be entered, but there is an entry condition for RVCP (2# in the DW). The second part is correct. SRO Basis: N/A USERS GUIDE OEOP-0t-UG Rev. 067 Page 52 of 156 ATTACHMENT 1 PageS of 15

                                  <<Group Isolation Checklist>>

Group 3 Isolation Signals Signal Tech Spec Value Setpoint Value Low Level 2 +101 inches ÷105 inches High Differential Flow 73 gpm 43 gprn (after 28.5 minute time delay) Area High Temperature 150°F 140°F Area Ventilation T High 50F 47°F Non-Regen Hx Outlet N/A 135°F Temp Hi SLC Initiation NA N/A RWCU Outside Pump/HI 120°F 115°F Rms RWCU Differential Flow 30 minutes 28.5 minutes High Time Delay Group 3 Isolation Valves Control Room RTGB Panel H12-P601 Valve Number Power Supply Normal Unit 1(Unit 2) Position Fail Position Checked [Note 11 G3l-F00l 1XC(2XC)/EI(E3) NO [Note 2] FAI G31-F004 IXDB(2XDB) NO FAI [DCI Note 1: SLC Initiation and RWCU Non-Regen Hx Outlet Temperature Hi signals do NOT isolate The RWCU Inlet Inboard Isolation Valve, G31-FQ0i.

41. 2950161 CAUJ7ON There are seven keViock NORMAL/LOCAL sitches located on Diesel Generator 2 control paneL Six of these are located in a row The seenth sitcti is located in the tow above the six switches. The six switches in a row must be placed in LOCAL before placing the seventh stch in LOCAL.

Which one of the following identifies the reason the seventh switch is the last one to be placed in LOCAL while performing OASSD-02, Control Building? To prevent a loss of DG2 caused by the: A output breaker circuitry not being isolated from the fire area. B lube oil control circuitry not being isolated from the fire area. C loss of redundant power supply fuses for the output breaker circuitry. D loss of redundant power supply fuses for the engine run control circuitry. Answer: D K/A: 295016 Control Room Abandonment AK3 Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT: (CFR: 41.5/45.6) 03 Disabling control room controls RO/SRO Rating: 3.5/3.7 Tier 1 / Group 1 K/A Match: This meets the K/A because the six local switches remove control room controls and the seventh switch supplies an alternate power supply to the equipment. Pedigree: Bank Objective: LOI-CLS-LP-304, Obj. 21 Explain why the Diesel Generator NORMAL/LOCAL switches must be placed in LOCAL in a particular sequence.

Reference:

None Cog Level: Fundamental Explanation: The six switches in a row isolate DG2 engine and generator control circuitry from the control room (the fire area) since a fire induced fault in wiring in the fire area may result in loss of the DG. The seventh switch inserts redundant control power fuses to the circuitry that has been isolated in the event a fault has already resulted in blowing the normal fuses. This seventh switch must be turned last with the potentially faulted circuitry already isolated or the alternate fuses may also blow making the DG unavailable. The DG engine lockout is already tripped if the DG had been running since the operator is directed to trip the DG using emergency stop. Of the first six switches, they include: Diesel START/STOP (2 switches) Diesel Governor (2 switches) - Generator Voltage Regulation (2 switches)

Distractor Analysis: Choice A: Plausible because this is what one the first six switches are performing. Choice B: Plausible because a loss of lube oil will prevent the DG from operating. Choice C: Plausible because the output breaker does have an alternate power supply. Choice D: Correct Answer, see explanation. SRO Basis: N/A SD-39 EMERGENCY DIESEL GENERATORS Rev 20 SYSTEM DESCRIP11ON Page 39 of 166 The Governor Control At Setpoint indicator light provides a status of the DRU speed reference for the 23D1A governor. The light is an indicator that the Governor Control System is ready to operate at the Setpoint speed. During actual operation of the D(, the Governor Control At Setpoint indicator light may or may not be illuminated depending on the speed of the DC. Voltage Adjust Switches Iwo three position (RA1SENEUT-LOWER) spring return to NEUT switches are provided per engine to permit the adjustment of voltage regulators from the local panel regardless of EDG mode of operation. The auto adjust switch is normally used. ASSD Keylock Switches Brass handled two-position NORM LOCAL ASSD keylock switches on the local engine panels permit the operator to transfer control of the engine and generator to the local control panel. ASSD operations are performed When a fire exists in the pfant and components required to be operated may be damaged b the fire, These switches isolate control room controls and indications to isolate the EDG control circuitry from potential fire induced faults. There are six ASSD stches (2 for EDG runistop controls, 2 for governor controls, and 2 for voltage regulation controls) located on each local EDG panel. When in the 1ASSD mode, operation of the Diesel engine can only be accomplished by the LOCAL EMERGENCY STOP and LOCAL EMERGENCY START pushbuttons. In addition to the six ASSD switches, for EDG 2 and 4 only, there is a seventh ASSD switch located above the other six switches. This switch provides an alternate set ot control power fuses for EDG control circuitry. This may be necessary since fire induced faults may have blown normal control fuses. When operating the ASSD switches for EDGs 2 or 4, the seventh switch must be turned last after the potentially faulted circuitry has been isolated to prevent blowing the alternate fuses, making the EDG unavailable to provide power to Sate Shutdown loads.

42. 2950171 During accident conditions, the source term from the Unit One Reactor Building must be estimated. Three RB HVAC supply fans and three RB HVAC exhaust fans are running.

lAW OPEP-03.6.1, Release Estimates Based on StacWVent Readings, which one of the following is the calculated release rate? ATTACHMENT 2 Page 1 of I Source Term Calculation From #1 RX Gas (1 -CAC-AQH.12644) METER FLOW1 EFFlClENCY2 RELEAS E READING icfrn) FACTOR RATE TIME (cprn) (pCiisec) 43 203 CF?1 per

                           &0 E+3 I m[nute ago                           exhaut ar If not available use 43,200 cftn per exhaust tan times the number of fans operat[ng The eThciency factors can be obtained from OE&RC-2020 contact E&RC counting room.

Release Rate = (cpm) x (cfm) x (Etflciency Factor) A 2.2 E+3 iCi/sec. B 6.6 E+3 j.iCi/sec. C 1.3 E+4 tCi/sec. D 6.6 E+4 pCi/sec. Answer: B K/A: 295017 High Off-Site Release Rate AA2 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.10/43.5/45.13) 03 Radiation levels RO/SRO Rating: 3.1/3.9 Tier 1 / Group 2 K/A Match: This meets the K/A because it is testing the source term for a release off-site. Pedigree: Bank Objective: LOl-CLS-LP-30 1 A, Obj. 6 Determine data required for offsite dose projection in accordance with AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment, and PEP-03.6.1, Release Estimates Based Upon Stack/Vent Readings.

Reference:

None Cog Level: High Explanation: Per Attachment 2 the calculated release rate is: Meter reading (CPM) X Flow (43,200 per fan X no of discharge fans) X efficiency factor or (4 E+3) (43,200 X 3) (1.275 E-5) = 6.6 E+3 mCi/sec Distractor Analysis: Choice A: Plausible because it is the calculation without multiplying times the number of running exhaust fans. Choice B: Correct Answer, see explanation. Choice C: Plausible because it uses the total number of fans running vs. the number of exhaust fans. Choice D: Plausible because it is the correct numerical value but is off by a factor of 10. SRO Basis: N/A

43. 2950181 Unit Two is operating at 65% power when the following are observed:

0 U 100 RCC RBCCV PUMP HEAD TANK DISCH HEADER PUMP MOTOR LEVEL HVLO PRESS LOW TEb1P H UA-3 UA-3 A-6 (In Afarm) (In Ataim) (In Alarm) E-60 Ptw 2B 2C E-40 2.F [1 E-20 RB CCW DISCHARGE PRESSURE PX-P1-671-1 Which one of the following identifies the operator actions that are required lAW OAOP-1 6.0, RBCC W System Failure? A Commence a plant shutdown lAW OGP-05, Unit Shutdown. B Reduce system heat load by removing RWCU and Fuel Pool Cooling from service. C Reduce Reactor power as necessary to clear the Recirc Motor high temperature alarm. D Trip all RBCCW Pumps, insert a manual reactor scram, and trip both recirc pumps. Answer: D K/A: 295018 Partial or Complete Loss of Component Cooling Water AK2 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER and the following: (CFR: 41 .7 / 45.8) 02 Plant operations RO/SRO Rating: 3.4/3.6 Tier 1 I Group 1 K/A Match: This meets the K/A because it is testing the relationship of the loss of RBCCW and the actions required for plant operations Pedigree: Last used on the 04 NRC Exam

Objective: LOl-CLS-LP-302-H, Obj. 4a Given plant conditions, determine the required supplementary actions in accordance with the following AOPs: 0AOP-16.0, RBCCW System Failure

Reference:

None Cog Level: High Explanation: A complete loss of RBCCW is defined as discharge header pressure below 60 psig and all available RBCCW pumps running (AOP-16.0). A complete loss requires a manual scram. FPCCU is removed only for partial loss, power reduction as necessary to maintain drywell temperature is not appropriate since a scram is required. Distractor Analysis: Choice A: Plausible because the unit is requited to be shutdown but the procedure designates a scram not a normal unit shutdown. Choice B: Plausible because this is an action if it is not a complete loss of RBCCW Choice C: Plausible because this is an action if it is not a complete loss of RBCCW. Choice D: Correct Answer, see explanation. SRO Basis: N/A

RBCOW SYSTEM FAiLURE OAOP-16.O Rev. 31 Page 9 of 18 4.2 Supplementary Actions (continued)

b. IF ADHR Mode piping is NOT the source of the leakage, THEN re-align RBCCW Pumps A and D from ADHR Mode to RBCCW Mode, as necessary E NOTE A complete loss of RBCDW is defined as discharge header pressure less than 60 psig, high temperature alarms on components supplied by RBCOW, and all available (no more than three) RBCCW pumps operating on the RBCDW header C

4. IF there is a complete loss of RBCCW.

THEN C

a. Trip all RBCCW pumps (including RBCDW Drywell HVAC Cooling Pump if operating on the affected unit and pumps operating in ADHR Mode) C

b. Close the following valves:

• RCC-V28 (RBCCW To OW Isol Vlvs) C
• RCC-V52 (RBCCW TO OW Isol Vlvs) C

c. Trip RWCU pump(s) C

d. Isolate RWCU System by closing the following valves:

• 031-FOOl (RWCU Inboard Isol VIv) C
• 031-FOO4 (RWCU Outboard isol Vlv) C

e. Reduce reactor power with recirc flow in accordance with OENP-24.5, Fomi 2, Immediate Reactor Power Reduction Instructions C

f. Insert a manual scram C

g. Enter 1EOP-01-RSP(2EOP-O1-RSP), Reactor Scram Procedure, AND perform concurrently with this procedure C

h. Trip both reactor recirculation pumps by performing the following:

(1) Depress VFD A Emerg Stop C (2) Depress VFD B Emerg Stop C

44. 295019 1 Unit Two has entered OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures, due to a loss of instrument air pressure.

Which one of the following completes both statements below? The Diesel Generator starting air (1) affected. The VA-2A-BFIV-RB, RB HVAC Butterfly Isolation Valve, fails (2) A (1) is (2) open B (1) is (2) as-is C (1) is NOT (2) open ID (1) is NOT (2) as-is Answer: D KJA: 295019 Partial or Complete Loss of Instrument Air AA2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: (CFR: 41.10 / 43.5 / 45.13) 02 Status of safety-related instrument air system loads RO/SRO Rating: 3.6/3.7 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing status of equipment on a loss of air Pedigree: New Objective: LOI-CLS-LP-302K, Objective 6 Summarize the consequences associated with improper equipment operation specified in OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures

Reference:

None Cog Level: Fundamental Explanation: A loss of instrument air will not make the DGs inoperable because they have their own dedicated air system. The BFIVs fail as-is.

Distractor Analysis: Choice A: Plausible because the first part is correct and since air operated valves can be designed to fail open, closed or as-is. Choice B: Correct Answer, see explanation. Choice C: Plausible because the DGs use air for starting and since air operated valves can be designed to fail open, closed or as-is. Choice D: Plausible because the DGs use air for starting and since air operated valves can be designed to fail open, closed or as-is. SRO Basis: N/A PNEUMATIC (AIRJN [TRO 13 EN) SYSTEM FAI LU RES DAOP-2D .0 Rev. 46 Page 23 of 28 ATTACHMENT 2 Page 2 of 2 Components Required To Perform A Safety-Related Function After Loss of Normal Instrument Air CJIIII.JIENT LOSS OF AIR DESCRIPTION FAILED POSITION RNA-Y313 Ck VIv To Io N2 Backup From RNA OPEN RNA-V314 Ck Vlv To Io N2 Backup From RNA OPEN RNA-V3f 5 Ck VIv To Ioo N2 Backup From RNA CLOSED RNA-V316 Ck VIv To iso N2 Backup From RNA CLOSED VA-1(2)A-&FlV-RB Ar Operators For RB HVAC AS IS VA-I (215-BFIV-RB Air Operators For RB HVAC AS IS VA-I(2)C-BFI V-RB Air Operators For RB HVAC AS IS VA-1(2D-BFlV-RE Air Operators For RB HVAC AS IS DIESEL ENG 1 Diesel Generator OPERABLE DIESEL ENG 2 Diesel Generator OPERABLE DIESEL ENG 3 Diesel Generator OPERABLE DIESEL ENG 4 Diesel Generator OPERABLE

45. 295020 1 l&C Techs inadvertently cause a low level 3 (LL3) signal.

Unit Two plant conditions are: Reactor pressure 930 psig Drywell pressure 1.7 psig, steady Drywell temp (average) 140°F, slow rise Drywell leak calculation Normal Which one of the following completes the statement below? All Drywell Cooler Fans are: A tripped, but can be overridden on. B tripped, and cannot be overridden on. C running, but can be tripped at the RTGB. D running, and cannot be tripped at the RTGB. Answer: A K/A: 295020 Inadvertent Containment Isolation AA1 Ability to operate and/or monitor the following as they apply to INADVERTENT CONTAINMENT ISOLATION: (CFR: 41.7 / 45.6) 02 Drywell ventilation/cooling system RO/SRO Rating: 3.2/3.2 Tier 1 / Group 2 K/A Match: This meets the K/A because it is testing what the DW coolers do on an isolation signal. Pedigree: Bank Objective: LOl-CLS-LP-04, Obj. 20 Given plant conditions determine if the drywell coolers should auto start or trip

Reference:

None Cog Level: High Explanation: LOCA signal on LL3 closes Group 10 which fails dampers open, but also trips fan motors. Override for LOCA trip can be performed as long as a LOCA does not really exist which is overridden in back panels (XU-27/XU-28). The low level condition also is a scram signal which provides an auto start signal for the DW Coolers which is prioritized by the trip signal.

Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because the fans do trip and if the conditions were different they would not be able to be overridden. Choice C: Plausible because the fans do auto start on a scram signal or usually when the dampers are opened and under different conditions they would be able to be tripped from the RTGB. Choice D: Plausible because the fans do auto start on a scram signal or usually when the dampers are opened and under different conditions they would not be able to be tripped from the RTGB. SRD Basis: N/A Placing a Unit 2 Dnjwell Cooling Fan control switch in START causes the fans discharge damper to open. WHEN the discharge damper is lull open, the fan will start. The control switch should be held in the START position until the discharge damper is full open. The RBCW cooling water valve to the coils will open concurrently with a fan start. Placing the Drywelt CooLing 10 Fan control itch in START causes the lns discharge damper to open. WHEN the discharge damper is full open, the fan will start. The control switch should lie held in the START position until the fan starts. The common air inlet damper and the RBCCW cooling water valve to the coils will open concurrently with a fan start. WHEN the control switch for Drywell Cooling Fan 1 A, 1 C, or 1 D is placed in START, the associated fan starts and the discharge backdraft damper opens from the fan air flow, The discharge damper position indication does not input to the start logic for Drywell Coaling Fans IA, 1C, and ID. The RBCW coaling water valve to the coils will open concurrently with a fan start. The Dtywell Lower Vent dampers can be positioned to either MIN or MAX position by a o-position control switch on Panel XU-3. Normal plant operating position for these dampers is the MIN position. Placing these dampers to MAX position during plant operation may produce extreme temperature excursions in the upper drywell regions. Low scram air header pressure will reposition these dampers to the MAX position and automatically start any idle drywell cooling fan selected for AUTO. Drywell Cooler Overtide Switches, VA-CS-5993!5994, are provided in Panels XU-27128 to facilitate vatious modes of Drywelt cooler operation as required by the EOPs. The Pneumatic Nitrogen System or Reactor Building Non-Interruptible Instrument Air pneumatically operates the drwell cooling fans discharge dampers. These dampers will fail open on loss of pneumatics. Unit 2 and 10 dtywell cooling fans discharge dampers fail closed on loss of the associated 120 VAC distribution panel. A contactor in the associated fans 480 VAC breaker provides drywell cooler FAN ON indication on RTGB Panel XU-3. SD-04 Rev. 9 Page 17 ot 103

The drywell coolers receive a LOCA trip sgnal from the Core Spray initiaton relays

46. 295021 1 Unit One in MODE 5.

The fuel pool gates ate removed. SDC Loop B is in service. Fuel pool cooling assist is in operation. The RHR Loop B pumps tripped and can NOT be restarted. Which one of the following completes both statements below? (consider each statement separately) Fuel pool cooling assist (1) Fuel pool cooling assist (2) capable of being aligned to the SDC Loop A lAW 1 OP-i 7, Residual Heat Removal System Operating Procedure. A (1) remains in service (2) is B (1) remains in service (2) is NOT C (1) islost (2) is O (1) islost (2) is NOT Answer: D K/A: 295021 Loss of Shutdown Cooling AK2 Knowledge of the interrelations between LOSS OF SHUTDOWN COOLING and the following: (CFR: 41.7 / 45.8) 05 Fuel pool cooling and cleanup system RO/SRO Rating: 2.7/2.8 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing the relationship of using SDC and the Fuel Pool. Pedigree: New Objective: LOI-CLS-LP-01 7, Obj 5 Given a drawing of the RHR system, trace the flow path for all of the six (6) modes of operation.

Reference:

None Cog Level: High Explanation: Fuel pool cooling assist mode utilizes the B Loop of RHR so that when it is lost so too will the fuel pool cooling assist operations. If the gates were installed then the A Loop of SDC could be used with the B loop discharge flowpath, but with the gates removed this is NOT an option.

Distractor Analysis: Choice A: Plausible because the students may think that the FPC pumps provide the motive force for this mode of operation and if the gates were installed then this would be correct. Choice B: Plausible because the students may think that the FPC pumps provide the motive force for this mode of operation and the second part is correct. Choice C: Plausible because the first part is correct and if the gates were installed then this would be correct. Choice D: Correct Answer, see explanation SRO Basis: N/A Bi 1 Fuel Pool Cooling Assist Mode With Fuel Pool Gates Removed U CAUTION The folIovlng section has the potential to significantly raise area dose rates. 811.1 Initial Conditions Datefriime Started Initials 1 Reactor in Mode 5 with fuel pool gates removed. Z Fuel pool temperature can NOT be maintained tess than 125°F.

3. OPT-08.OC has been completed satisfactorily within previous 92 days.

4. Fuel Pool Cooling system in operation in accordance with lOP- 13 with available ftte[ pool cooling heat exchangers in operation.

5. RHR Loop B is operating in shutdown cooling in accordance with Section 5.7 or 5.8.

47. 295023 1 Unit Two is performing refueling operations when the refueling SRO reports that a spent fuel bundle has been dropped in the cattle chute.

The following radiation monitoring alarms are received: UA-03 (3-7) Area Rad Refuel Floor High UA-03 (4-5) Process Rx Bldg Vent Rad Hi Which one of the following identifies the Immediate Action that is required lAW OAOP-05.0, Radioactive Spills, High Radiation, and Airborne Activity? A Verify Group 6 isolation. B Evacuate all personnel from the refuel floor. C Place Control Room Emergency Ventilation System in operation. D Isolate Reactor Building Ventilation and place Standby Gas Treatment trains in operation. Answer: C K/A: 295023 Refueling Accidents AA1 Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS: (CFR: 41.7 / 45.6) 04 Radiation monitoring equipment RO/SRO Rating: 3.4/3.7 Tier 1 1 Group 1 K/A Match: This meets the K/A because it is testing the immediate operator actions for a radiation event. Pedigree: Bank Objective: LOl-CLS-LP-302J, Obj. 5 List the immediate operator actions required to be performed in accordance with OAOP-05, Radioactive Spills, High Radiation, and Airborne Activity

Reference:

None Cog Level: Fundamental Explanation: This is an Immediate Action identified in AOP-05.0. Distractor Analysis: Choice A: Plausible because this is an auto action not an immediate operator action of the AOP Choice B: Plausible because lAW OAOP-05.0 this is the first supplemental action. Choice C: Correct Answer, see explanation. Choice D: Plausible because RBHVAC isolation and SBGT start requires PROCESS RXBLDG VENT RAD HI-HI (UA-03 3-5) in alarm and these are supplementary actions in the procedure.

SRO Basis: N/A RADIOACTIVE SPILLS. HIGH RADIATION, AND OAOP-05.O AIRBORNE ACT MTY Re 32 Page 5 of 15 4.0 OPERATOR ACTIONS NOTE The tollowing should be considered for estabtishment as critical parameters during perlonnance of this procedure- Li a Area radiation levels

• Personnel habitability in the affected area 4.1 Immediate Actions

1. IF a fuel assembly was dropped or damaged, THEN ensure the Control Room Emergency Ventlation System (CREVS)isinoperation.{7.1.i).., Li

48. 295024 1 Unit Two is operating at rated power when high drywell pressure switch C72-PTM-NOO2A-1 fails high resulting in the annunciation of A-05-(5-6) Pri Ctmt Press Hi Trip.

Which one of the following completes the statement below? RPS high drywell pressure relay C72-K4A will (1) causing a (2) scram. A (1) energize (2) half B (1) energize (2) full C (1) de-energize (2) half O (1) de-energize (2) full Answer: C K/A: 295024 High Drywell Pressure EA1 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE: (CFR: 41.7 / 45.6) 05 RPS RO/SRO Rating: 3.9/4.0 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing the ability to monitor RPS (half scram condition) for a high DW pressure condition Pedigree: New Objective: LOI-CLS-LP-003, Objectives: 7.g Given plant conditions state the Normal, Initiation, and Fail position/condition of the following components: (Open/Closed Energized/De Energized) RPS Logic

9. Given any scram signal, describe the logic arrangement for the signal including what combination of signals will cause a Full Scram.

Reference:

None Cog Level: High Explanation: The RPS relays are de-energize to actuate and a single relay actuates the alarm and will cause a half scram.

Distractor Analysis: Choice A: Plausible because there are logics that are energize to actuate and the half scram is correct. Choice B: Plausible because there are logics that ate energize to actuate and there are also logics that only require one instrument to actuate (Nuclear instrumentation). Choice C: Correct Answer, see explanation. Choice D: Plausible because the first part is correct and some logics do cause a full scram (Nuclear instrumentation). SRO Basis: N/A C 2 ci *1:-:2 rcEsE sr TRP AJT2 ACTD13

2. If th rr.Iry catinnt prur h;h r1p .gn1 i rei.d nly PPS Tnt .:rn will

2. If zn:r&ry :n iant irune hgh trip :gn.l s ri:d i bcth R Tnip fys-sas, s rearn:r S:ns. w.il n.::ur DVICE/ iOINTE ris_. 072E4A2 5SL Izt:h 2T::::I2Al, 5C, CI, *:r 2 1.7 zsig

FIGURE 03-15 High Drywell Pressure Trip TRIP CHANNEL Ski TRIP CHANNEL AZ 1 C71(2)-P r C71f2H7TM NOO2C1 PANLL PANLL XU65 XU6E

   /K4A                                                      K4C NOTE PRESSURE SWITCH CON1ACTS OPEN ON HIGH DRYWELL PRESSURE CONDITION TRIP CHANNEL RI                                      TRIP CHANNEL 82 C72t2:PfM                                               C71(2)TM N0028-1                                                 NDC2D-1 4   K4B                                                   -

SD-03 Rev. 12 Page13or9O

49. 295025 1 Unit One was operating at power when a turbine trip occurred.

85 control rods fail to insert. Reactor pressure peaks at 1145 psig. Which one of the following completes both statements below? The reactor recirc pumps (1) tripped. Tripping of the reactor recirc pumps results in a rapid decrease in reactor powerdueto (2) A (1) must be manually (2) voiding of the moderator B (1) must be manually (2) a reduction in reactor water level C (1) have automatically (2) voiding of the moderator ID (1) have automatically (2) a reduction in reactor water level Answer: C K/A: 295025 High Reactor Pressure EK3 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE: (CFR: 41.5 / 45.6) 02 Recirculation pump trip RO/SRO Rating: 3.9/4.1 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing the reason the recirc pump is tripped. Pedigree: Bank Objective: LOI-CLS-LP-002, Obj. 30 Given Plant conditions determine if the ATWS-RPT protection logic should have actuated

Reference:

None Cog Level: Fundamental Explanation: The Anticipated Transient Without Scram circuit provides an alternate means of reducing reactor power in the unlikely event that the control rods fail to insert into the core following a Reactor Protection System actuation signal. Tripping of the VFD Input Circuit Breakers (ICB) will rapidly reduce recirculation flow. This results in a rapid decrease in reactor power because of the voiding of the moderator. Setpoints for ATWS trip are high reactor pressure 1137.8 psig and low reactor level LL2 105

Distractor Analysis: Choice A: Plausible because the ATWS procedure directs the pumps to be tripped and the second part is correct. Choice B: Plausible because the ATWS procedure directs the pumps to be tripped and level is reduced in the AIWS procedure which does lower power. Choice C: Correct Answer, see explanation. Choice D: Plausible because the first part is correct and level is reduced in the AIWS procedure which does lower power. SRO Basis: N/A 3.2.6 Anticipated Transient Without Scram Recirculation Pump Trip (A7WS.RPT The Mticipatecl TransFent Without Scram circuit provides an alternate means of reducing reactor pcer in the unl!iiely event that the control rods fail to insert into the core following a Reactor Protecflon System actuation signa[ Tripping of the VEt) Input Crcuit Breakers (ICS) will rapidly reduce recirculatfion flow. This results in a rapid decrease in reactor power because of the voiding of the moderato[ Two signals are used for the initiation of A1NS-RPT These Signals are LL2 reactor vessel water level and high reactor vessel pressure Each of these parameters is monitored by totir sensors. Two level or pressure instruments in one of tv.o logic trains are required to energize relays which trip both Recirculation Pumps. SD-021 Rev 0 Page 72 of 182

50. 295026 1 Unit One failed to scram following a loss of off-site power with the following plant conditions:

Reactor Power 5% RPV Water Level -55 inches (N036) RPV Pressure 850 psig Which one of the following completes both statements below? This UA-12 (5-4) alarm is expected to be received when SPIMS DIV I suppression pool water temperature first reaches (1) BULK V(TR TEMP . The RHR logic requirements to place torus cooling in service SETPINT isi under the current plant conditions will require (2) A (1) 95°F (2) placing the Think Switch to Manual first and then bypassing the 2/3rd core height interlock B (1) 95°F (2) bypassing the 2/3rd core height interlock first and then placing the Think Switch to Manual C (1) 105°F (2) placing the Think Switch to Manual first and then bypassing the 2/3rd core height interlock ID (1) 105°F (2) bypassing the 2/3rd core height interlock first and then placing the Think Switch to Manual Answer: B K/A: 295026 Suppression Pool High Water Temperature G2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 / 43.5 / 45.3) RO/SRO Rating: 4.2/4.0 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing when the torus temperature alarm setpoint and what controls need to be operated to establish cooling. Pedigree: New Objective: LOl-CLS-LP-01 7, Obj 09 Given an RHR pump or valve, list the interlocks, permissives and/or automatic actions associated with the RHR pump or valve, including setpoints.

Reference:

None Cog Level: High

Explanation: LOCA signal is sealed in due to being less than LL3 (45 inches) RPV water level is less than 2/3rd core height (-47 inches) therefore the keylock switch and then the Think switch is required (sequencing is essential). When the torus reaches 95°F this alarm will come in, 105°F is the TMax alarm. Distractor Analysis: Choice A: Plausible because the first part is correct and the second part is opposite of the requited actions. Choice B: Correct Answer, see explanation. Choice C: Plausible because this is the alarm setpoint for the SPTMS DIV I BULK WTR TEMP SETPT TMAX, and the second part is opposite of the required actions. Choice D: Plausible because this is the alarm setpoint for the SPTMS DIV I BULK WTR TEMP SETPT TMAX, and the second part is correct SRD Basis: N/A

Unit 1 arr caa: -4 Page 1 :f 2 3flMS DVI I lUll. iflE EP SE:PCEtI NOTE: tnoperability of t.hrs anncrsrat:r nay resrlt tn a TPY kerurrei

mpensatory :-easc:-e

        -     :i*;h suppresstcrs pool bdfl averare water reir.terarure -

CISER7AIIDDTS 1 r:erorder hannel I tn CA22RCEIA indreares rnrreasrno auppresaton ;::1 renrerature 132 indicator tllu:nrnated CATT44E1 - A1I2113

        -     If su;tressi:n pool temperature is a;trrarhnc tEaF and no rn trresa that ::uld add hear to me sutpressrtn po:l ther.

refer to AOPlL 2, A1:n:rmal trtnarv C:ntarrx.ent *::r.ditt:ns and 12P3L2, Safety leltef %alve larluree

2. If sutpress:cn co:l temperature ts greater than 93°F due to addrng heat to the surpressicn pool from a;;roved testing procerures, then refer to the apzrourrate teat tr:ceiure to marntarn suppression tool tenterature below L3SE

3. If auttressr:n c::1 temzerature ts greater than 93°F and n:

testing is rn progress that rruld add hear to the su-rcreasi:n pool, then enter Ec;-::rc:p, Primary :onmazent :ontrol, and ACP142, Abnormal rimary Containnent Conditions. 4 If a Dircuit or etuotment nalfuncroon is susected, ensure that a

              !fP/WO is prepared.

t17I2E. EETPCII::S s;U:s !2icr:rooess:r CA>-IU-42l 93°F

Unit 1 c:: iga 1 of C zr: i EU:z: wir E s:rr n AUTO AUTrONS CAUSE C - Eigh up si:n p::1 tti avaracia wace opratltra 03 $EcvATrONS Ra:order Channal C tn OCTA-44ClA djoa in:raainci auaaion poti nneratura, TMA indi:a::: ilLuninatai 2TT44CE1 A0TICIE If pool ta ara.tura i qaatar s i teoing i :n ora that ::t1d add haat o: tha pei:n po:1, than antar EC ClP Primary Oo a:nranr Control, and Aoi4:, C:ntairnr Con ition! if oo ai:aady If orassion pool rcara:-u:a i atroa:hn 10SF dc:e to addong haao to th trrion cool frc!rt aptrcvf tastjng rrocafuras than :afar to tha aforoDroata taa: nr:caiura to naon:ain attppresaion pool t paratura ba1o

3. If up:n pool tamparacura i graatar than 1CEF than atop all tastin; and anca: EOPC2PCCP imar Containment Con:::1 and AOP14.C, Abnormal Proma:y Contairnuano C:nd:oon.

4 If a :ir:u+/-t or tOtmanE malftnooo:n is ucacted than ensure that a WRWO os prepared. iE7ICE;EETCINIE Cicrorrooaor CAITY442El

FIGURE 17-12 CoolingSpray Permissive Logic 11 CLOSED IN & AFTEJ-{ MANUAL (SEAL IN) Si 10 - CLOSED IN MAN LIAL S1i t4 Cl OEP>2fl CORE HFICHT K698 2/SCORE (SEAL IN) HEIGHT OVED Y Runs CLOSLO ON LPUI NIl Al ION EleNa R555 KSnR (fU24JFO2S) 1 I CONTAINMLNT SPRAY R596 (w3 PERMISSIVE 1 f (FO1G.F021.F027)

51. 2950281 Unit Two is in MODE 3 following a Station Blackout.

lAW OEOP-01 -SBO-01, Plant Monitoring, the AO has reported the following temperatures from the RSDP temperature recorder 2CAC-TR-778: Point 1 290°F Point2 118°F Point 3 255° F Point 4 230°F Point 5 191°F Point6 117°F (REFERENCE PROVIDED) Which one of the following represents the correct calculated Drywell temperature? A 205°F B 249°F C 258°F D 267°F Answer: B K/A: 295028 High Drywell Temperature EA2 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: (CFR: 41.10 / 43.5 / 45.13) 01 Drywell temperature RO/SRO Rating: 4.0/4.1 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing the students ability to determine drywell temperature. Pedigree: Bank Objective: LOl-CLS-LP-303-B, Obj. 3 Given plant conditions, control room or remote shutdown panel indications, and SBO-04, calculate the following parameters: a. Drywell Temperature

Reference:

Attachment 4 of OEOP-01 -SBO-01, Plant Monitoring Cog Level: Fundamental Explanation: Attachment 4 of OEOP-01-SBO-01, Plant Monitoring, has a calculation worksheet for figuring Drywell temperature from RSDP temperature recorder readings. 290 *0.141 = 40.89 255

• 0.404 = 103.02 230
• 0.455 = 104.65 248.56

Distractor Analysis: Choice A: Plausible because this is the average of points 1 - 3 used in calculation. Choice B: Correct Answer, see explanation. Choice C: Plausible because this is the average of points 1 3, & 4. Choice D: Plausible because this is performing the calculation backwards (points 4, 3, 1) SRO Basis: N/A PLANT MONITORING OEOP-0i-SBO-0T Rev. 0 Page 16 ci 18 ATTACHMENT 4 Page 1 of 1 Drywell Temperature Calculation Using RSDP Recorder Inputs Values obtained from Recorder CAC-TR-778 Above 70 Elevation PT1 290 xO:141= 4089 T Between 28 and 45 Elevation PT3 255 xO.404= 103.02 F Between 10 and 23 Elevation PT4 230 xO.455= 104.65 °F Average Drywell Temperature 248.56 (Sum of 3 Regional Weighted Areas)

52. 295029 1 Unit Two is performing RVCP with HPCI in pressure control.

Subsequently, torus water level reaches -23 inches. Which one of the following completes both statements below? The E41-F004, CST Suction Vlv, will (1) The E41-F008, Bypass to CST Vlv, will (2) A (1) close (2) close B (1) close (2) remain open C (1) remain open (2) close ID (1) remain open (2) remain open Answer: A K/A: 295029 High Suppression Pool Water Level EA1 Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: (CFR: 41.7 / 45.6) 01 HPCI RO/SRO Rating: 3.4/3.5 Tier 1 / Group 2 K/A Match: This meets the K/A because it is testing operation of HPCI on high torus level Pedigree: New Objective: LOl-CLS-LP-01 9. Obj. 3p Given plant conditions, predict how the HPCI System will respond to the following events: High/low Suppression Pool water level

Reference:

None Cog Level: High Explanation: The torus water high level condition (>-25 inches) will cause the torus suction valves to open. When either valve is full open the F008 and F004 will close.

Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Plausible because the first part is correct. The F008 will get a close signal when the F041 or F042 is full open. Choice C: Plausible because the high level does not directly close the F008 valve. The second part is correct. Choice D: Plausible because the high level does not directly close the F004 or F008 valve. SRO Basis: N/A FIGURE 19-7 CST Suction Valve, E41 -F004. Control Logic ri :I-; wi*ri C (I:) &ii- CI II-.% HI-l ( CI )-.HI-4 CL1 rIOT çJLI JF .N N)T tUtL CE4 JLL DP [42ULLOPC J CLCOL ON CO ..VL ThO

                           -I-C_CSLI OPEN                                                  CLOSE

LII 0

      -T 0

0 0 q w U u-. 1 0 I r5 0 I) E 4-In EZ

53. 295030 1

-18 Unit One is operating at rated power when A-01 (3-7) Suppression Chamber Lvi Hi/Lo, is received.

E-23 The BOP Operator verifies the alarm using CAC-Ll-4177, Supp Pool Level, indicator on Panel XU-51. (indication provided to the left) Which one of the following identifies the action that is 3 required lAW A-01 (3-7) Suppression Chamber Lvi H,YLo? The water level in the Unit One torus must be: A lowered by using Core Spray and routed to Radwaste. B lowered using RHR and routed to Radwaste. C raised by opening the HPCI suction from the CST. ID raised by opening the Core Spray suction from the CST. Answer: D K/A: 295030 Low Suppression Pool Water Level G2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 I 43.5 I 45.3) ROISRO Rating: 4.2/4.0 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing ability to know whether the alarm is due to high or low level and knowledge of how to correct. Pedigree: New Objective: LOl-CLS-LP-302-D, Obj 2 Given plant conditions and AOP-14.0, determine the required supplementary actions.

Reference:

None Cog Level: High Explanation: The student will verify that level is low using the provided indication and then will determine that the level must be raised lAW the APP. The low level alarm comes in at -30.5 inches and the high level alarm comes in at -27.5 inches. Level can be raised using RHR or the Core Spray systems.

Distractor Analysis: Choice A: Plausible because it is a combined alarm and if it is assumed that a high water level condition exists the CS system can take a suction from the torus to correct the level condition, but is not allowed in the procedure. Choice B: Plausible because it is a combined alarm and if it is assumed that a high water level condition exists the RHR system is utilized in the procedure to lower level. Choice C: Plausible because level is low requiring it to be raised and the HPCI system could gravity drain to the torus, but is not allowed by the procedure. Choice D: Correct Answer, see explanation. SRO Basis: N/A Unit 1 APt 1- 3-tape 1 of 2 SUPPRESSION CRAD-SEP. LVI HIt LU 3t2r0 a:fl ONE NONE

1. Sucpression uccl water level hith 2 inches

2. Suppression prol water level low cs+/- inches)

3. Circuit malfunctoon 03 SERVATIONS Suppressior. o::1 water level fCAC121E212, CACL:417 CAC -LA-I .3 2 2 1 NOTE: Rapid changes in suppressior. cccl pressure due to conditior.s such as inerting cr art in1eaage can cause level flu:tuatzons in supcression pooi up to L rnch cr more.

  ;.cnous NOTE:        ECCS :ceepfrll statrons makeuc fl:w to the suctressoor. c:::1 rs apro:-:tmate1y 27 gpm

1. If the cause cf the annunoratcr os a tianned evolutocn, then refer to the acproociate :;erating rroceiuce to maintair. supcressicn poo water level.

2. If the cause cf the annuncrator :s n:t a tanned evclu.tior., ther.

determtne the rause of addit::n or l:ss of water to suppression pool and minimize evclutror.s which add or remove water to cr fr:m the suzpressicn cool.

3. If auppressron ccol water level s high or l:w, then enter

                   )AOP14.J to draor. or foll the suppression pc:l as necessarU.

4. If suporessoon pool water level is greater than 27 inches or less than -31 inches, then enter JECt02PCCF.

2. If a circuit malfur.ction os suspected, ensure a 110 is prepared.

ABNORMAI PRIMARY CONTAINMENT CONDITIONS DAOP-i4.O Rev. 30 Page 15ot36 4.2.4 Suppression Pool Level HighiLow

           .ffi. suppression pool [evel is approaching -27 inches.

THEN lower suppression pool level to Radwaste in accordance Mh IOP-17(20P-17), Residual Heat Removal System Operating Procedure 0

2. IF suppression pool level is approactting -31 inches.

THEN raise suppression pool revel in accordance with the totiowing applicable procedure 0 Unit 1 Only:

• lOP-if, Residual Heat Removal Systeni Operating Procedure 0
• lOP-I B, Core Spray System Operating Procedure 0 Unit 2 Only:
• 20 P-il, Residual Heat Removal System Operating Procedure 0
• 20P 18, Core Spray System Operating Procedure 0

54. 295031 1 Unit One is executing the ATWS procedure with the following plant conditions:

Reactor power 12% Reactor pressure 940 psig, controlled by EHC Reactor water level 170 inches, controlled by feedwater Which one of the following identifies the reason the ATWS procedure directs deliberately lowering RPV water level to 90 inches? A Reduces reactor power so that it will remain below the APRM downscale setpoint. B Provides heating of the feedwater to reduce potential for high core inlet subcooling. o Reduces challenges to primary containment if MSIVs close. D Promotes more efficient boron mixing in the core region. Answer: B K/A: 295031 Reactor Low Water Level EK1 Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.8 to 41.10) 03 Water level effects on reactor power RO/SRO Rating: 3.7/4.1 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing the knowledge of why level is lowered in a ATWS Pedigree: Bank Objective: LOl-CLS-LP-300-E, Obj 7 Explain the reason for lowering reactor water level while performing the Anticipated Transient Without Scram Procedure.

Reference:

None Cog Level: fundamental Explanation: To prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, reactor water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude. Twenty-four inches below the lowest nozzle in the feedwater sparger (i.e. 90 inches) has been selected as the upper bound of the reactor water level control band. This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment). This water level is sufficiently high that even without bypassing the low reactor water level MSIV isolation, reactor water level can be controlled with the feedwater pumps to preclude the isolation.

Distractor Analysis: Choice A: Plausible because since the operator can re-establish injection at 90 inches irrespective of power level. Power will lower as level is lowered but 90 inches will not guarantee APRMs are downscale Choice B: Correct Answer, see explanation. Choice C: Plausible because since there is no current challenge to containment from heat input. If level is lowered due to containment heat input, 90 inches is not specified as the top of the level band. This would be either TAF or the level at which downscales are received Choice D: Plausible because since lowering level will reduce natural circulation and reduce boron mixing. ATWS procedure directs raising level back to the normal band (1 70-200 inches) once hot shutdown boron weight is injected SRO Basis: N/A AW.S PROCEDURE BASIS DOCUMENT OO[-3T5 Rev. 015 Page 13 otG2 5.4 Step RCIL-2 NOtE Rcqu wi immtt] 1 2th recrcuIaion pumps nppd wtth povrs oLo.c it.it::ki, . Illi 5 lr CANNOr bt ANC RPV Ie%,I is iboe 90 inches, I[ctl St THEN t.rminte nd prevent njdion no tho RPV unIcss uscd ic nicT boron

• HPI
• CcSpny
• LPCF
• rI1-2jRcIc If reactor power is greater than 23% with both reactor recircutation pumps tripped and RPV level above 90 inches, RPV level needs to be promptly reduced below the feedwater nozzles, to avoid thermal hydraulic instabilities. This is accomplished by termination and prevention of injection systems, from identified systems, particularly feedwater, within 120 seconds.

To prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, RPV level is initially lowered sufficiently below the elevation of the feedwater sparger nozzles. This ptaces the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that ate susceptible to oscillations, initiation and growth of oscillations is pnncipally dependent upon subcooling at the core inlet, the greater the subcooling, the more likely oscillations will commence and increase in magnitude.

55. 295032 1 Which one of the following identifies the reason for performing Emergency Depressurization due to exceeding Maximum Safe Operating Temperatures lAW 00 1-37.9, Secondary Containment Control Procedure Basis Document?

A Prevent an unmonitored release. B Preserve personnel access into the reactor building. C Prevent damage to equipment required for safe shutdown. D Ensure ODCM site boundary dose limits are not exceeded. Answer: C K/A: 295032 High Secondary Containment Area Temperature EK3 Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: (CFR: 41.5/45.6) 01 Emergency/normal depressurization RO/SRO Rating: 3.5/3.8 Tier 1 / Group 2 KJA Match: This meets the K/A because it is testing the reason ED is performed for high secondary containment temperatures. Pedigree: Bank Objective: LOI-CLS-LP-300-M, Obj 13a Given plant conditions and the SCCP, determine the required actions if the following limits are exceeded: Maximum Safe operating values with a primary system discharging into secondary containment.

Reference:

None Cog Level: Fundamental Explanation: The MSOT values are the area temperatures above which equipment necessary for the safe shutdown of the plant will fail. These area temperatures are utilized in establishing the conditions which reactor depressurization is required. The criteria of more than one area specified in this step identifies the rise in reactor building parameters as a wide spread problem which may pose a direct and immediate threat to secondary containment integrity, equipment located in the RB, and continued safe operation of the plant. Distractor Analysis: Choice A: Plausible because this is a purpose of SCCP not the reason for ED on Temperature. Choice B: Plausible because this is the reason for max safe operating rad levels. Choice C: Correct Answer, see explanation. Choice D: Plausible because this is a purpose of SCCP not the reason for ED on Temperature. SRO Basis: N/A

SECONDARY CONTAINMENT CONTROL 001-37.9 PROCEDURE BASIS DOCUMENT Rev. 004 Page 8 of 33 5.1 Step SCCP1 E,:o1 , I The conditions vtuch requure entry to SCCP are symptomatic of conditions vtch, if not corrected, could degrade into an emergency. Adverse effects on the operability çf equipment located in the reactor building and conditions directly challenging secondary containment integnty or spent fuel pool cooling were specifically considered in the selection of these entry conditions. In addition, personnel accessibility to some of the areas may be required to pertom certain actions specified in the procedure This was also considered in making these detemlinations. An area temperature or area differential temperature above its maximum normal operating level is an indication that steam from a primary stem may be discharging into the reactor building. As temperatures continue to increase, the continued operability of equipment needed to carry out EOP actions may be compromised.

56. 295034 1 Which one of the following completes both statements below?

lAW OAOP-5.4, Radiological Releases, RRCP is entered when the Turbine Building Vent Rad Monitor indication exceeds an (1) EAL. lAW RRCP, before the radioactivity release rate reaches a (2) Emergency EAL, Emergency Depressurization is required. A (1) UnusualEvent (2) Site Area B (1) UnusualEvent (2) General o (1) Alert (2) Site Area o (1) Alert (2) General Answer: D K/A: 295034 Secondary Containment Ventilation High Radiation G2.4.08 Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (CFR: 41.10/43.5 / 45.13) RO/SRO Rating: 3.8/4.5 Tier 1 / Group 2 K/A Match: This question matches the KA because it tests the knowledge if the AOP and EOP are performed in conjunction with each other. Pedigree: new Objective: LOl-CLS-LP-302-J, Obj. 3c Given plant conditions, determine the required Supplementary Actions in accordance with: OAOP-05.4, Radiological Release

Reference:

None Cog Level: Fundamental Explanation: The AOP states that when an Alert EAL is entered then ENTER RRCP. Before a GE is declared ED is required to be performed. (A scram is required before a SAE is declared)

Distractor Analysis: Choice A: Plausible because an Unusual Event is the first declaration in the EAL network and a SAE is the criteria for a scram in RRCP. Choice B: Plausible because an Unusual Event is the first declaration in the EAL network and the second part is correct. Choice C: Plausible because the first part is correct and the SAE is the criteria for a scram in RRCP. Choice D: Correct Answer, see explanation SRO Basis: N/A RADIOLOGICAL RELEASE OAOP-05A Rev. 0 Page 6 of 13 3.0 AUTOMATIC ACTIONS tcontInued a SBGT starts a Group 6 isolation valves close

4. jf UA-03 2-8, Radwaste Effluent Rad Hi H, in ALART1, THEN Dl2-V27AB) (RW Liq Effluent Disch VIvs) close

5. IF UA-23 3-6, Main Steam Line Rad Hi-Hi/Inap, in ALARM, THEN:

a Mechanical vacuum pumps trip

• OG-V7 fCndsr Hogging Valve) closes 4.0 OPERATOR ACTIONS 4.1 Immediate Actions None 4.2 Supplementary Actions

1. IF AT ANY TIME elevated radiation levels are determined to be from resin injection pgy, THEN go to DAOP-26.0, High Reactor Coolant or Condensate Conductivity U

2. IF AT ANY TIME gaseous release rate exceeds an Alert level, THEN enter OEOP-04-RRCP, Radioactivity Release Control Procedure U

z 2 0 CL (-) w r UL U3L

57. 295036 1 Following an unisolable RWCU line break in the reactor building the following conditions exist:

South Core Spray Room temperature 155°F South RHR Room temperature 300°F UA-1 2 (2-3) South Core Spray Room Flood Level Hi, in alarm UA-1 2 (2-4) South RHR Room Flood Level Hi, in alarm UA-1 2 (1-4) South RHR Room Flood Level Hi-Hi, in alarm (REFERENCE PROVIDED) Which one of the following completes both statements below? lAW OEOP-01-UG, Users Guide, (1) equipment required for safe shutdown will fail. lAW SCCP, Emergency Depressurization (1) required. A (1) ONLY the South RHR room (2) is B (1) ONLY the South RHR room (2) is NOT C (1) the South RHR room AND Core Spray room (2) is o (1) the South RHR room AND Core Spray room (2) is NOT Answer: B K/A: 295036 Secondary Containment High Sump / Area Water Level EK1 Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: (CFR: 41.8 to 41.10) 02 Electrical ground! circuit malfunction RO/SRO Rating: 2.6/2.8 Tier 1! Group 2 K/A Match: This meets the K/A because this is testing the implication of high water level on equipment and whether ED is required. Pedigree: New Objective: LOl-CLS-LP-300-M, Obj, 1 3a Given plant conditions and the Secondary Containment Control Procedure, determine the required action if the following limits are exceeded: Maximum Safe operating values WITH a primary system discharging into Secondary Containment

Reference:

OEOP-01 -NL, EOP/SAMG Numerical Limits And Values, Attachment 3, Containment Parameters, Table 3-B, Secondary Containment Area Temperature Limits

Cog Level: High Explanation: Distractor Analysis: Choice A: Plausible because the first part is correct and for ED two areas in the same parameter must be at max safe conditions, while this question has two parameters in the same area. Choice B: Correct Answer, see explanation. Choice C: Plausible because both areas have a max normal condition and for ED two areas in the same parameter must be at max safe conditions, while this question has two parameters in the same area. Choice D: Plausible because both areas have a max normal condition and the second part is correct. SRO Basis: N/A

USERS GUIDE 0EOP-t-UG Rev. 067 Page Wot 156 3O DEFINITIONS toontinued) a Core Spray Loop A

• core Spray Loop B
• RHR Loop A (one or to pumps running) a RHR Loop B (one or o pumps running)

32. Maximum Normal Operating (Parameter): The highest value of the identited parameter expected to occur during normal plant operating conditions with alt directly associated support and control systems functioning properly.

33. Maximum Pressure Suppression Primary Containment Water Level: The highest primary containment water level at which the pressure suppression capability of the containment can lie maintained This corresponds to the bottom of the ring header.

34. Maximum Safe Operating Radiation Level: The radiation level above which personnel access necessary for the safe shutdown of the plant II be precluded.

If the maximum safe operating radiation level is exceeded in an area (but is within the EQ envelope as contained in DR-227, Document Reference for Environmental Qualification Service Conditions) and then later clears and is subsequently followed by another area exceeding maximum safe operating radiation level, action for one area exceeding maximum sate operating radiation level should be taken.

35. Maximum Safe Operating Temperature: The temperature above which equipment necessary for the safe shutdown of the plant may fail. This temperature is utilized in establishing the conditions under which RPV depressurization is required. Separate temperatures are provided for each Secondary Containment area. If the maximum sate operating temperature is exceeded in an area and then later clears and is stibsequently followed by another area exceeding maximum safe operating temperature, action for two areas exceeding maximum safe operating temperature should be taken.

36. Maximum Safe Operating Water Level: The water level above which equipment necessary for the safe shutdown of the plant may tail. This water level is utilized in establishing the conditions under which RPV depressurization is required. Separate water levels are provided for each Secondary Containment area. If the maximum safe operating water level is exceeded in an area and then later clears and is subsequently followed by another area exceeding maximum sate operating water level, action for two areas exceeding maximum sate operating water level should be taken.

        /,
  /7 WHEN f  Srrii :jLrhLLur Max Safe OR EQ I  envelope n more tkIn 1\\\\\%, one &eci THEN Cofltlrn.I*

Jr EMERGENCY DEPRESSURIZATION RLQUIRED. AHEACHr.IENT 3 Page 73 of 87 Containment Parameters Secondary Containment Area Tern perawre Limits Table 3-B PLANT PLANT LOCATION MAX NORM MAX SAFE AUTO GROUP AREA DESCRIPTION OPERATING OPERATNG ISOLATION VALUE (°F/ ALLIE (F) N CORE N CORE SPRAY 120 1Th N/A SPRAY ROOM S CORE S CORE SPRAY 120 175 N/A SPRAY ROOM RWCU PMPROOMA PMP ROOM 6 140 225 3 HX ROOM N RHR N RHR EQUIP ROOM ITS 295 N/A S RHR S RHR EQUIP ROOM 175 295 NIA RCIC EQUIP ROOM 165 295 5 HPCI HPCI EQUIP ROOM 16S 165 4 STEAM RCIC STM TUNNEL 190 295 5 TUNNEL HPCISTMTUNNEL 19D 295 4 20 FT 20 FT NORTH 140 200 NfA 20 FT SOUTH 140 200 N/A 50 FT 50 FT NW 140 200 N/A 50 FT SE 140 200 N1A REACTOR MULTIPLE AREAS ALARM N/A 3, 4, AND/OR 5 BLDG ANNUN. SETPOINT A-02 5-7 REACTOR MSIV PIT ANNUN. ALARM N/A 1 BLDG A-06 6-7 SETPOINT

58. 295037 1 Which one of the following identifies the effect of accomplishing injection of Cold Shutdown Boron Weight during an ATWS?

The reactor is shutdown: A and will remain shutdown under all conditions. B and may return to power if a cooldown is initiated. C and will return to power if RPV water level is raised. O and may return to power as Xenon depletes during the first 24 hours. Answer: A K/A: 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown EK1 Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: (CFR: 41.8 to 41.10) 03 Boron effects on reactor power (SBLC) RO/SRO Rating: 4.2/4.4 Tier 1 / Group 1 K/A Match: This meets the K/A because the student will have to know the effects that boron has to overcome to have the reactor remain shutdown during an ATWS. Pedigree: Bank Objective: LOI-CLS-LP-005, Obj 3 List the positive reactivity effects that must be overcome by SLC injection

Reference:

None Cog Level: Fundamental Explanation: Injection of the CSBW into the RPV will provide adequate assurance that the reactor is and will remain shutdown. It is the least weight of soluble boron which, if injected into the RPV and mixed uniformly, will maintain the reactor shutdown under all conditions. This weight is utilized to assure the reactor will remain shutdown irrespective of control rod position or RPV water temperature Distractor Analysis: Choice A: Correct Answer, see explanation Choice B: Plausible because a cooldown would add positive reactivity. Choice C: Plausible because raising level would add positive reactivity Choice D: Plausible because xenon would add positive reactivity SRO Basis: N/A

. USERS GUIDE OEOP-31-UG Rev. 057 . PageBof 156 3.0 DEFINITIONS (continued)

12. Chugging: An intermittent condensation phenomenon which occurs at the downcomer extt when the drwell is pressurized due to a small high energy (steam) leak inside the drywelL When a steam bubble collapses at the exit of the downcomers, the rush of water filling the void (some of it drai up into the downcomer pipe) induces severe stress at the jiunction of the downcomer vent header. Repeated application of this stress can cause these joints to experience fatigue failure (i.e. crack) thereby creating a pathway which bypasses the pressure suppression function of the containment.

13. Cold Shutdown Boron Weight: The least weight of soluble boron which, if injected into the RPV and mixed uniformly, will maintain the reactor shutdown under all conditions. This weight is utilized to assure the reactor will remain shutdown irrespective of control rod position or RPV water temperature.

59. 295038 1 A radioactive release has occurred in the Radwaste Building.

Which one of the following completes both statements below? The Radwaste Building HVAC (1) provide for the adsorption of noble gases. This discharge will be monitored by the (2) A (1) will (2) Main Stack Radiation Monitor B (1) will (2) Turbine Building Wide Range Gaseous Monitor (WRGM) C (1) wilINOT (2) Main Stack Radiation Monitor D (1) wilINOT (2) Turbine Building Wide Range Gaseous Monitor (WRGM) Answer: C K/A: 295038 High Off-Site Release Rate EA1 Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.7 / 45.6) 01 Stack-gas monitoring system RO/SRO Rating: 3.9/4.2 Tier 1 / Group 1 K/A Match: This meets the K/A because the student will have to determine the place that the radwaste ventilation system discharges to and the rad monitor that monitors it. (ability to monitor) Pedigree: New Objective: LOI-CLS-LP-37.2, Obj 6e State the interrelationship between the Radwaste Building Ventilation and the following: Process Radiation Monitoring

Reference:

None Cog Level: Fundamental Explanation: The radwaste vent system does not have any charcoal adsorbers in the system, only HEPA filters. The discharge is to the main stack.

Distractor Analysis: Choice A: Plausible because other systems provide for the adsorption of noble gases and the second part is correct. Choice B: Plausible because other systems provide for the adsorption of noble gases and the other systems (Turbine Building Vent) is monitored by the WRGM. Choice C: Correct Answer, see explanation. Choice D: Plausible because the first part is correct and the other systems (Turbine Building Vent) is monitored by the WRGM. SRO Basis: N/A 1.1 System Purpose The purpose of the Radwaste Building Heating, Ventilation, and Air Conditioning System is to maintain areas at a temperature which provides optimal operation of equipment, comfort and safety of personnel, and to control/prevent radiological release& A negative pressure relative to atmosphere is maintained by this system. Flow through the building is limited to prevent the spread of contamination and to mitigate the consequences of an inadvertent release of radioactive material to the building. All air exiting the Radwaste Building is monitored via the main stack radiation monitor. The Radwaste Building exhaust air volume provides dilution flow to the plant stack for condenser oft-gas, condenser air removal, and the Reactor Building standby gas trains Several items of personnel safety may be mentioned relative to the building HVAC Systent When the Radwaste Building Ventilation system is functioning in concert with the Control Building Ventilation System, niajor pressure differentials between the volumes are manageable. It the relative pressure between the two volumes is high, personnel access doors may not function as predicted. If the Radwaste Building is under a significant negative pressure and the Control Building pressure is positive in relation to the Radwaste Building, the door will open rapidly. Certain areas of the Radwaste Building have high noise levels attributed to the ventilation system. Areas close to the supply and exhaust ducts are especially noisy. The upper elevations of the Radwaste Building, especially around the supply and exhaust plenums have very high noise level& It is imperative that hearing protection is used in these areas. A list of definitions and abbreviations used in this material is in Attachment 1. SD-37.2 Rev. 4 Page 5 of 28

60. 300000 1 Unit One is operating at rated power when the following alarms are received:

UA-01 (4-4) lnstr Air Press-Low UA-01 (5-1) Air Dryer 1A Trouble The AO reports that the cause of the alarms is due to filter blockage. Which one of the following completes both statements below? The Service Air Dryer malfunction will cause SA-PV-5067, Service Air Dryer Bypass Valve, to open when pressure first lowers to (1) lAW OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures, the required action isto (2) A. (1) 105 psig (2) place the I B Service Air Dryer in service B. (1) 105 psig (2) set the service air dryer maximum sweep value to zero C. (1) 98psig (2) place the 1 B Service Air Dryer in service D. (1) 98psig (2) set the service air dryer maximum sweep value to zero Answer: C K/A: 300000 Instrument Air System A2 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: (CFR: 41 .5 / 45.6) 01 Air dryer and filter malfunctions RO/SRO Rating: 2.9/2.8 Tier 2 / Group 1 K/A Match: This meets the KA because it is predicting the response on the system and then using procedure (AOP-20) determine the action required. Pedigree: New Objective: LOl-CLS-LP-046, Obj. 6 Given plant conditions, determine if the following automatic actions should occur:

a. Service Air Isolation g. Air Dryer bypass.

Reference:

None Cog Level: High

Explanation: 98 psig is when the bypass valve auto opens, the 105 psig is the isolation setpoint for Service Air. The AOP will direct placing the standby Air Dryer in service. Distractor Analysis: Choice A: Plausible because 105 is the isolation setpoint for the service air system and the second part is correct. Choice B: Plausible because 105 is the isolation setpoint for the service air system and this is an action in the AOP but would not be performed for this failure. If there is a high demand then this is performed to limit the amount of air that is used for the blowdown of the air dryer filter when cycling filters. Choice C: Correct Answer, see explanation. Choice D: Plausible because the first part is correct and this is an action in the AOP but would not be performed for this failure. If there is a high demand then this is performed to limit the amount of air that is used for the blowdown of the air dryer filter when cycling filters. SRO Basis: N/A

PNEUMATIC (AlRNlTR0 GEN) SYSTEM FM LURES 0AOP-20 .0 Rev. 46 Page 13of28 4.2 Supplementary Actions (continued) NOTE

• Service Air System pre-filter or after-filter differential pressure should NOT exceed V5psid fl
• In service air compressor tgh discharge pressure (Unit 1: greater than or equal to 125 psig, Unit 2: greater than or equal to 130 psig) or relief valves lifting could be an indication of air dryer high differential pressure potentially caused by power failures resulting in valves in the flow path fading dosed U
• l(2)SA-PV-5067 [Serv Air Dryer 1(2 iA Bypass Pressure Control Valve)], is located in the Turbine Building air compressor area U

i. IF UA-0 1 5-3, Air Dryer I(ZIA Trouble, is in alarm, THEN perform the following:

(1) Unit 1 Only: Confirm I -SA-PV-5067 (Serv Air Dryer IA Bypass Pressure Control Valve), is OPEN U CAUTION The service air dryer provides a low dew point pneumatic source to downstream components. A low dew point is necessary to insure long term reliability of these components. The time the dryer is bypassed should be minimized (Li fl U (2) Unit I Only:1F i-SA-PV-5067 is NOT open, THEN open l-SA-V5089 (Serv Air Dryer Manual Bypass Valve) U (3) Unit 2 Only: Confirm 2-SA-PV-5067 (Serv Air Dryer 2A Bypass Pressure Control Valve), is OPEN U (4) Unit 2 Only: IF 2-SA-PV-5067 is NOT open, THEN open 2-SA-V5089 (Serv Air Dryer Manual Bypass Valve) U (5) IF available, THEN place lB Service Air Dryer in service AND shutdown i(2)A Service Air Dryer in accordance with OOP-46, Instrument and Service Air System Operating Procedure U

PNEUMATIC (AIRITROGEN) SYSTEM FAILURES UAOP-2D.O Rev 46 Page 9 of 28 4.2 Supplementary Actions (continued) c IF air s NOT cross-tied, J4p cross-tie operation will .fQJ cause a loss ot instrument air on the unafFected unit THEN perform the following: (1) Obtain permission from the non-affected unit C (2) Ensure i-SA-PV-5071 (Cross-Tie Valve), Located on Unit 1, Panel XU-2, is OPEN C (3) Ensure 2-SA-PV-5071 (Cross-Tie Valve), located on Unit 2, Panel XU-2, is OPEN C (4) IF opening the cross-tie valve degrades the non-affected unit, THEN return to Step 1 .L(4) C

d. IF the in service air dryer is in sweep node, THEN consider securing sweep mode in accordance with Attachment 1, Setting Service Air Dryer(S) Maximum Sweep VakeToZero

61. 3000002 Unit One is in MODE 3 following a seismic event and reactor scram with the following plant conditions:

Reactor level 55 inches Reactor pressure 500 psig Drywell pressure 9 psig Division I PNS header pressure 93 psig Division II PNS header pressure 98 psig Which one of the following completes both statements below? Div I Backup N2 Rack Isol Vlv, RNA-SV-5482 is (1) Div II Backup N2 Rack Isol Vlv, RNA-SV-5481 is (2) A. (1) open (2) open B. (1) open (2) closed C. (1) closed (2) open D. (1) closed (2) closed Answer: B K/A: 300000 Instrument Air System K3 Knowledge of the effect that a loss or malfunction of the (INSTRUMENT AIR SYSTEM) will have on the following: (CFR: 41.7 / 45.6) 01 Containment air system RO/SRO Rating: 2.7/2.9 Tier 2 / Group 1 K/A Match: This meets the KA because it is testing the effect of the low pressure (loss or malfunction) of the air system on containment air (N2 backup). Pedigree: Last used on 2007 NRC Exam Objective: LOl-CLS-LP-046-A, Obj. 8 Given plant conditions, determine the effects that the following conditions will have on the Pneumatic System: (LOCT) b. Low Instrument Air/Pneumatic Nitrogen (IAN/RNNPNS) Header Pressure

Reference:

None Cog Level: High

Explanation: No LOCA signal is present so the Backup N2 valves will not be open on a Core Spray initiation signal. The Backup N2 valves open at 95 psig or lower in the PNS header. This would result in Division I Backup N2 valve (5482) being open and Division 11(5481) being closed. Distractor Analysis: Choice A: Plausible if the student believes that either division will open both valves. Choice B: Correct Answer, see explanation. Choice C: Plausible if the student uses the valves for the division separation. Choice D: Plausible if the student only checks the LOCA signal and not the low pressure signal. SRO Basis: N/A Unit I APP U.A-O1 -1 Page 1 o2 RB INSTR AIR RECE[VER 1A PRESS LOW AUTO ACTIONS

1. RNA-SJ-54a2. Hgh Pressure Bottle Rack Isolato.fl Vle, opens. supIng SRVs antt CAC-t6 with a pneumatic source.

DEVICEISETPOINTS RNA-PSL-3596 9E. pscg decreasing Unit I APP UA-O1 1-2 Page 1 012 RB INSTR AIR RECEIVER lB PRESS LOW AUTO ACTIONS

1. RNA-SV-5481, High Pressure Bottle Rack Isolatlon VaIe, opens, supplying SRJs and CAC-17 with a pneumatic source.

DEVICE1SETPOINTS RNA-PSL-3597 95 pslg decreasing

62. 400000 1 Unit One is operating at rated power with the following conditions:

CSW Pump IA trips Conventional header pressure lowers to 35 psig Which one of the following completes both statements below? If CSW header pressure remains at this pressure for (1) seconds, the SW-V3, SW To TBCCW HX5 Otbd lsol VIv, and SW-V4, SW To TBCCW HXs lnbd Isol Vlv, will close to a throttled position. lAW OAOP-1 9, Conventional Service Water System Failure, the SW-V3 and SW-V4 are reopened (2) A. (1) 30 (2) ONLY after a reactor Scram is inserted B. (1) 30 (2) if system pressure is restored by starting the standby CSW pump C. (1) 70 (2) ONLY after a reactor Scram is inserted D. (1) 70 (2) if system pressure is restored by starting the standby 05W pump Answer: D K/A: 400000 Component Cooling Water System (CCWS) A2 Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: (CFR: 41 .5 / 45.6) 01 LossofCCWpump RO/SRO Rating: 3.3/3.4 Tier 2 / Group 1 K/A Match: This meets the KA because it is testing the auto start signal/logic for a cooling water system. Pedigree: Last used on the 2010 NRC exam Objective: CLS-LP-302-H, Obj. 4 Given plant conditions and any of the following AOPs, determine the required supplementary actions: d. OAOP-19.0, Conventional Service Water System Failure

Reference:

None Cog Level: High

Explanation: IF conventional service water header pressure remains below 40 psig for 70 seconds, THEN:

             - SW TO TBCCW HXS OTBD ISOL, SW-V3 closes to a throttled position
             - SW TO TBCCW HXS INBD ISOL, SW-V4 closes to a throttled position The Standby CSW pump should start and restore CSW header pressure to normal prior to the SW valves throttling closed. If the standby CSW pump fails to auto start, manually starting the pump will restore CSW header pressure. AOP-19 provides guidance to re-open the SW valves only after header pressure has been restored and the cause of low pressure is known (pump trip).

Distractor Analysis: Choice A: Plausible because 30 seconds is when the DG cooling water valves close and a Scram is inserted only after the SW valves have closed to the throttled position AND CSW header pressure cannot be immediately restored above 40 psig under this condition all CSW pumps would be shutdown. Choice B: Plausible because 30 seconds is when the DG cooling water valves close and system pressure restored by the STBY pump start is correct. Choice C: Plausible because 70 seconds is correct and a Scram is inserted only after the SW valves have closed to the throttled position AND CSW header pressure cannot be immediately restored above 40 psig under this condition all CSW pumps would be shutdown. Choice D: Correct Answer, see explanation SRO Basis: N/A CONVENTIONAL SERVICE WATER SYSTEM OAOP-1 9.0 FAILURE Rev. 26 PageSorll 3.0 AUTOMATIC ACTIONS Standby pump selected to the conventional service water header starts at 40 psig U

2. IF all conventional service water pumps are tripped, THEN:

• SW-V36 (SW To CW Pumps Inbd VIv). closes U
• SW-V37 (SW To CW Pumps Otbd Vlv), closes U
• CWIPs trip on low bearing lubricating water flow (5 6 gpm, time-delayed 15 minutes). resulting in loss oV condenser vactturn U

3. IF conventional service water header pressure remains less than 40 psigfor7o seconds, THEN:

• SW-V3 (SW To TBCCW HXs Otbd Isol), closes to a throttled position LI
• SW-V4 (SW To TBCCW HX5 Inbd Isol), closes to a throttled position C

. CONVENTIONAL SERVICE WATER SYSTEM OAOP-iq.o FAILURE Rev. 26 Page 9 of ii 4.2 Supplementary Actions (continued)

d. Attempt to isolate any source of leakage U

e. Ensure discharge valves are CLOSED on shutdown pump(s) U

1. Check service water traveling screens for excessive build-up AND wash it excessive buildup is occurring U

g. Check service water trash racks for excessive build-up AND notify Maintenance to clean [f excessive buildup is occurring U h Locally monitor each pump discharge strainer differential pressure U Check Annunciator Panel UA-0i for lit annunciators U

10. Reter to Technical Specification 3.7.2, Service Water (SW) System and Ultimate Heat Sink (UHS) for operability requirements U ii. WHEN conventional service water header pressure is restored to normal AND the cause of low header pressure has been corrected, THEN:

a. Open SW-V3 (SW To ThCCW HXs Otbd 1501) U

b. Open SW-V4 (SW To TBCCW HX5 lnbd Isol) U

63. 400000 2 Unit Two Nuclear Service Water (NSW) pumps are aligned as follows in preparation for equipment realignment:

 -      DISCHARGE             NUCLEAR SER\/ICE f        DISCHAPGE JLV SWV20 NtCLEAR 5tRJCt VLV SWV 1 9                                                           iJATERPUUP 2R VATEP PJNP 2A Ej                     2P6 STOP   j     ZTAR
                                                /

Subsequently, Off-site power is lost. Which one of the following identifies how NSW Pumps 2A and 2B will respond when the DG5 re-energize their E Buses? A. Both NSW Pumps auto start immediately when their associated E Bus is energized. B. Both NSW Pumps auto start five seconds after their associated E Bus is energized. C. NSW Pump 2B auto starts immediately after the associated E Bus is energized. NSW pump 2A does NOT auto start. D. NSW Pump 2B auto starts five seconds after the associated E Bus is energized. NSW pump 2A does NOT auto start. Answer: A K/A: 400000 Component Cooling Water System (CCWS) K4 Knowledge of CCWS design feature(s) and or interlocks which provide for the following: (CFR: 41.7) 01 Automatic start of standby pump ROISRO Rating: 3.4/3.9 Tier 2 / Group 1 K/A Match: This meets the KA because it is testing the auto start signal/logic for a cooling water system. Pedigree: Bank

Objective: LOl-CLS-043, Objective 8a State the power supply (bus and voltage) for the following Service Water System components: Nuclear Service Water Pumps.

Reference:

N/A Cog Level: High Explanation: NSW pumps auto start immediately after LOOP signal regardless of mode selector switch or discharge valve position. 5 second timer applies only on a LOCA. Distractor Analysis: Choice A: Correct Answer, see explanation Choice B: Plausible because this would be the case with a LOCA signal present. Choice C: Plausible because exam inee must know the power supply scheme. Choice D: Plausible because examinee must know the power supply scheme. SRO Basis: N/A Each pump is powered by a 4160 VAC motor supplied from the emergency bus power supplies: Component Power Supv lA CSW pump E4 lB CSW pump El 1C CSW pump E2 lANSWpump El lB NSW pump E2 2ACSWpump E3 2B CSW pump E4 2CCSWpump El 2ANSWpump E3 2B NSW pump E4 SD-43 Rev. 25 Page 10 of 87

In addition to the low header pressure auto start, the NSW pumps will start five seconds after receipt of a LOCA signal, regardless of mode selector switch or discharge valve position. For example, a Division I LODA signal from either Unit 1 or Unit 2 will auto stan the iA and 2A NSW pumps; the Division II LOA logic will auto start 10 and 2B NWS pumps. The NSW pumps. powered through the 4160 VAC emergency buses, will also automatically start immediately after the start of the diesel generators and reenergization of the emergency buses on loss of off-site power (LOOP), regardless of mode selector swttch or discharge valve posWon For example, a Division I LOOP sDgnal from either Unit I or Unit 2 will auto start the 1A and 2A NSW pumps; the Division II LOOP logic will auto start 10 and 2B NSW pumps. If a LOCA signal exists on the division sensing the LOOP, auto start will occur after five seconds, provided that a LOOP signal is not present on the opposite unit On a dual unit LOOP the NSW pump(s) of the LOCA (and non LOA) unit start immediately after the emergency buses are reenergized by their respective diesel generators without the five second delay.

64. 600000 1 Which one of the following identifies the potential consequence of failing to place backup nitrogen in service by placing RNA keylock switches in LOCAL lAW OASSD-02, Control Building?

RNA keylock switch noun names: 2-RNA-CS-001, Override Switch For Valve RNASV-5482 2-RNA-CS-002, Override Switch For Valve RNA-SV-5253 A. Misoperation of RCIC. B. Loss of drywell cooling. C. Inability to operate SRVs. D. Spurious operation of MSIV5. Answer: C KIA: 600000 Plant Fire On Site AK3 Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE: (CFR: 41.5 / 45.6) 04 Actions contained in the abnormal procedure for plant fire on site RO/SRO Rating: 2.8/3.4 Tier 1 / Group 1 K/A Match: This matches the KA because it tests the reason a step in the ASSD procedure is performed. The ASSD procedures are the plant fire procedures. Pedigree: Bank Objective: LOl-CLS-LP-304, Obj. 25k Given ASSD procedures and plant conditions, predict the consequences of FAILURE to perform the following actions: Deenergize RNA-SV-5482 and RNA-SV-5253 via keylock switches RNA-CS-001 and RNA-CS-002.

Reference:

None Cog Level: fundamental Explanation: The Reactor Building MCC Operator places the key lock switches to the LOCAL position to ensure Nitrogen System is lined up to provide reliable operation of the SRVs. Distractor Analysis: Choice A: Plausible because actions for the operation of RCIC are contained in the ASSD procedures. Choice B: Plausible because a loss of pneumatics would cause the DW cooler dampers to close. Choice C: Correct Answer, see explanation. Choice D: Plausible because actions to prevent the spurious operation of the MSIV is contained in the ASSD procedure. SRO Basis: N/A

SECTION Bi UNIT 2 RX BLDG MCC OPERATOR ACTiONS Initial Actions and RCIC Operations 1 £5 WHEN directed to start RCIC. THEN PERFORM the following at NICC 2XDB:

1. OPEN RCIC TURB TP & fliP VLV, E51-V8, at Compt 637 D (Row C1)

2. OPEN RCIC TUPA STM SPLY VLV E51-F045, at D Compt 644 (Row F2).

3 INFORM Unit 2 CR8 RCIC should be running. D 11 WHEN directed, THEN PERFORM the following at Unit 2 Reactor Building 50 foot elevation: 1.71 PLACE keylock switch 2-PNA-CS-OO I in LOCAL for valve D 2-RNA-SV-5482. 1.72 PLACE keylock switch 2-RNA-CS-002 in LOCAL for valve D 2-PNA-SV-5253. 1.7.3 INFORM the Unit 2 CR8 that backup nitrogen has been D made available for SRV operation. 1-8 IF directed, THEN TRANSFER RCIC suction from CST to suppression pool at MCC ZXDB as follows: 1.8:1 OPEN RCIC SUPP POOL SUCT VLy, E5i-F031, at D Compt 645 (Row Gi). 1.8.2 OPEN PCIC SUPP POOL SUCT VLV, E51-FO2 at D Compt 646 (Row G2). 1.8.3 CLOSE RCIC CST SUCT VLV, E5i-FOfl at Compt 638 D (Row C2). OASSD-02 Rev. 57 Page 31 of 154

65. 700000 1 A grid disturbance occurs with the following Unit One plant parameters:

Generator Load 980 MWe Generator Reactive Load 160 MVARs, out Generator Gas Pressure 50 psig (REFERENCE PROVIDED) Which one of the following identifies both available options that will place the Unit within the Estimated Capability Curve? A. Raise gas pressure to 58 psig or lower power to 940 MWe. B. Raise gas pressure to 58 psig or raise reactive load to 240 MVARs. C. Raise gas pressure to 58 psig or lower reactive load to 70 MVARs. D. Lower power to 940 MWe or raise reactive load to 240 MVARs. Answer: A K/A: 700000 Generator Voltage and Electric Grid Disturbances AA2 Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: (CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8) 03 Generator current outside the capability curve RO/SRO Rating: 3.5/3.6 Tier 1 / Group 1 K/A Match: This meets the K/A because the tests the ability to determine action needed to remain within capability curve. Pedigree: Last used on 2014 NRC exam Objective: CLS-LP-27, Obj. 9 Given the Generator estimated capability curves, hydrogen pressure and either MVARS, MW, or power factor, determine the limit for MW and MVARS.

Reference:

1OP-27 Attachment 2, Estimated Capability Curves Cog Level: High Explanation: Based on the conditions the student should plot the current location on the graph. Plot MWe along the bottom and MVARs up the side. Where these two points intersect, based on 50 psig gas pressure line is outside of the safe area. (Must be inside the curve to be safe) Lowering MWe or raising gas pressure are the only options. For this case lowering or raising MVARs would still be outside the curve.

Distractor Analysis: Choice A: Correct Answer, see explanation Choice B: Plausible because raising pressure will move the plant within the limits of the curve. Raising MVARS will not move the plant within the limits of the curve. Choice C: Plausible because raising pressure will move the plant within the limits of the curve. Lowering MVARS will not move the plant within the limits of the curve. Choice D: Plausible because raising MWe will move the plant within the limits of the curve. Raising MVARS will not move the plant within the limits of the curve. SRD Basis: N/A 600 48PSIG  :  : 9.85 psIc-z  : - 30 Psic - - 400  :  : 15 PSIG 200 -  :  :  :

                                                                                                      .1 o           200                                       -

kOO 71r: 600 o Li - - - - -

            -800       -                      -   -          -        -         -     -

200 400 600 800 1000 KILOWATTS

66. G2.1.01 1 Which one of the following completes both statements below lAW AD-OP-ALL-i 000, Conduct of Operations?

With the Unit operating at rated, steady state power, a key parameter that the OATC must monitor to assure a constant awareness of its value and trend is (1) An end to end control panel walk down shall be performed every (2) and documented in the Narrative Logbook. A. (1) jet pump flow (2) one hour B. (1) jetpumpflow (2) two hours C. (1) steam flow / feed flow (2) one hour D. (1) steam flow / feed flow (2) two hours Answer: D K/A: G2.1 .01 Knowledge of conduct of operations requirements. (CFR: 41.10 / 45.13) ROISRO Rating: 3.8/4.2 Tier 3 K/A Match: This meets the K/A because it is testing knowledge of the Conduct of Operations Manual Pedigree: New Objective: LOl-CLS-LP-201-D, Obj. lj Explain/describe the following lAW AD-OP-ALL-i 000, Conduct of Operations, 001-01.01, BNP Conduct of Operations Supplement and OPS-NGGC-1314, Communications: Control Board walkdown and monitoring requirements

Reference:

None Cog Level: Fund Explanation: lAW the conduct of operations document board walk downs must be completed every two hours and section 5.5.6 lists the key parameters to watch. Distractor Analysis: Choice A: Plausible because jet pump flow has a daily surveillance requirement and if a watchstander is relieved for greater than one hour it must be entered in narrative logbook. Choice B: Plausible because jet pump flow has a daily surveillance requirement and part two is correct. Choice C: Plausible because part one is correct and if a watchstander is relieved for greater than one hour it must be entered in narrative logbook. Choice D: Correct Answer, see explanation.

SRO Basis: N/A CONDUCT OF OPERAflONS AID-OP-ALL-I000 Rev. 5 Page 2i of $5 556 Control Board Monitoring (continued) k Unless involved in activies where Reactor Operator involvement is required by the Conduct of Operations (for example reactivity manipuIatins, peer checks or detailed panel reviews, the operator shall monitor the fottowing key parametors at a frequency to assure a constant awareness oHher value and trend:

• Rx Power
• RPV level (BWR
• Steam generator pressure (PWR
• RCS temperature
• Steam generator level (PWR
• RCS pressure
• Steam flow / feed flow Prossunzer level (PWRI CONDUCT OF OPERATIONS AD-OP-ALL-i 000 Rev. 5 Page 28 of 85

   .6   Control Board Monitoring (continued)

3. The CR5 ensures that a licensed operator performs an end to end control panel walk down event two hours. The walk down shalt be documented in the Narrative Loqbook.

CONDUCT OF OPERATIONS AD-OP-ALL-i 000 Rev. 5 Page 40 of 85

4. Whenever a watch station is relieved for greater than one hour, this information shall be entered in a Narrative Loj Program. a formal turnover and shift turnover sheet will be completed, including the logs signed over.

67. G2.1.321 Which one of the following completes the statement below?

1 OP-i 0, Standby Gas Treatment System Operating Procedure, prohibits venting the drywell and the suppression pool chamber simultaneously with the reactor at power because this would cause the: A. unnecessary cycling of reactor building to torus vacuum breakers. B. unnecessary cycling of torus to drywell vacuum breaker. C. SBGT Train water seal to blow out of the trough. D. pressure suppression function to be bypassed. Answer: D K/A: G2.1 .32 Ability to explain and apply system limits and precautions. (CFR: 41.10 /43.2 / 45.12) RO/SRO Rating: 3.8/4.0 Tier 3 K/A Match: This meets the K/A because it is testing the ability to explain the system precaution. Pedigree: Last used on 2012 NRC exam Objective:

Reference:

None Cog Level: Fundamental Explanation: Per OP-b, torus and drywell cannot be vented at the same time in Modes 1, 2 or 3. per the LER reference, this could result in bypassing pressure suppression function. Distractor Analysis: Choice A: Plausible because these vacuum breakers prevent drawing a negative pressure in the suppression pool. Cross connecting the drywell and the suppression pool free air space will not cause a negative pressure in the suppression pool. Choice B: Plausible because this lineup equalizes pressure between the drywell and the suppression pool free air space since the vacuum breakers operate on a d/p between the spaces this would bypass them, not open them. Choice C: Plausible because venting containment through large valves with an elevated pressure may blow out the SBGT water seal. Choice D: Correct Answer, see explanation. SRO Basis: N/A

STANDBY GAS TREATMENT SYSTEM OPERATHN G OP-iD PROCEDURE Rev 66 Page 4 ot49 1.0 PURPOSE

1. fills procedure provides irtstnuctional çuidance for operation of tile Standby Gas Treatment System and its associated deluge system.

2.0 SCOPE

1. This procedure provides the prerequisites, precautions, limitations, and instructional guidance for startup, normal operation, shutdown, and infrequent operation of the Standby Gas Treatment System and its associated deluge system.

3.0 PRECAUTIONS AND LIMITATIONS 1 The Standby Gas Treatment System will NOT automaticatly start if the control switch is in STBY D

2. Venting the dryweit and suppression pool simultaneously is !I performed when the plant is in MODE 2, or3. (8.1.1)

STAN DBY GAS TREATMENT SYSTEM OPERATING I OP-iD PROCEDURE Rev 66 Page 31 ot 49

8.0 REFERENCES

8.1 Commitments

1. LER 1-97-011, Dryweti and Torus Inerting/Deinerting Lineup Results in Unanalyzed Suppressioti Pool Bypass Path

68. G2.1.361 A core reload is in progress during a refueling outage. The initial loading of fuel bundles around each SRM centered 4-bundle cell was completed with all four SRMs fully inserted and reading 50 cps.

It is now approximately half way through the core loading sequence and SRMs read 80 cps. Which one of the following completes the statement below lAW OFH-1 1, Refueling? Fuel movement must be suspended when any SRM reading first rises to upon insertion of the next fuel bundle. A. 100 cps B. 160 cps C. 250 cps D. 400 cps Answer: B KJA: G2.1 .36 Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 /43.6 / 45.7) RO/SRO Rating: 3.0/4.1 Tier 3 K/A Match: This meets the K/A because it is testing the fuel movement requirements that an RO would monitor. Pedigree: New Objective: LOI-CLS-LP-305, Objectives 18 Given the conditions during a refueling outage state the operator actions required for rising SRM count rates and/or inadvertent criticality.

Reference:

None Cog Level: High Explanation: An increase in counts by a factor of two during a single bundle insertion is reason to suspend fuel movements. An increase by a factor of five from the baseline is also a reason.

Distractor Analysis: Choice A: Plausible because this is a doubling of the baseline counts which is used for a different criteria for suspension of fuel movements. Choice B: Correct Answer, see explanation Choice C: Plausible because this is an increase of the baseline counts by a factor of five which is a reason to suspend fuel movements. Choice D: Plausible because this is an increase of the counts by a factor of five which is a reason to suspend fuel movements. SRO Basis: N/A FH-11:

24. Suspension of fuel movement and notification of the Reactor Engineer is required. if either of the following occur An SRM reading rise by a factor of two upon insertion of any single bundle. During a spiral reload, this restriction applies only after the initial loading of fuel bundles around each SRM is complete. During a Core Shuffle, this restriction does NOT apply to the SRM that is having an adjacent fuel bundie inserted orremoved C

• An SRM rise by a factor of five relative to the SRM baseline count rate recorded on Attachment 6, Documentation for SRM Baseline C

25. SRM count rate may drop to less than 3 cps during either of the following conditions:

• With less than or equal to four fuel assemblies adjacent to the SRM and NO other fuel assemblies in the associated core quadrant C
• During a core spiral offload C

69. G2.2.02 I Unit Two is conducting a routine power reduction for rod pattern improvement.

The Reactivity Management Plan contains actions for the RO to insert a group of four rods from position 24to position 12. Which one of the following completes the statement below lAW AD-OP-ALL-0203, Reactivity Management? The movement of these rods should be: A. single notched for the entire movement. B. continuously inserted to the final intended position. C. continuously inserted and stopped four notches prior to reaching the intended position and then single notched into the final intended position. D. continuously inserted and stopped one notch prior to reaching the intended position and then single notched into the final intended position. Answer: D K/A: G2.2.02 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. (CFR: 41 .6 I 41.7 / 45.2) RO/SRO Rating: 4.6/4.1 Tier 3 K/A Match: This question matches the KA because it tests the genetic requirements of control rod movement during any power level. Pedigree: Bank Objective: LOl-CLS-LP-201-D, Obj. 22f Explain the following regarding AD-OP-ALL-0203, Reactivity Management: The procedural requirements for positioning intermediate control rods

Reference:

None Cog Level: High Explanation: If a rod is to be moved between 46 and 02 three notches or less, it must be single notched the entire move. When moving a control rod four notches or more, the control rod should be stopped one notch prior to reaching the intended position and then single notched into the final intended position. Distractor Analysis: Choice A: Plausible because this would apply if the movement was <four notches. Choice B: Plausible because this would apply under emergency conditions. Choice C: Plausible because the rod does have to be single notched into its final position but the rod can be continuously move if greater than four notches not for four notches. Choice D: Correct Answer, see explanation.

SRO Basis: N/A REACTIV 1W MANAGEMENT AD-OP-ALL-0203 Rev. 2 Page 47 of 90 5.2.8 [BWRJ Single Recirculation Loop Operation (7.1.5)

1. Standards a Single-Loop operation for extended periods of time is discouraged.

2. Expectations

a. Plant procedures that address Single Recirculation Loop Operation will identity applicable limitations and trip criteria,

b. For operations not covered by an approved procedure the Operational Decision Making process will be used to evaluate continued operation in Single-Loop.

c. The risk associated with single recirculation loop operations shall be carefully considered and appropriate contingencies will be developed.

d. Operator Jlfl shall be conducted for planned Single-Loop operations.

5.2.9 Control Rod Manipulations Standards

a. Ensure all control rod movements are made in a deliberate, carefully controlled manner while constantly monitoring nuclear instrumentation and redundant indications of reactor power and neutron flux.

2. Expectations

a. [BQVR] To minimize the possibility of mispositioning a control rod when inserting or withdrawing to an intermediate position (notch positions 02 through 46), the following practices shall be followed:

(1) When moving a control rod four notches or more, the control rod should be stopped one notch prior to reaching the intended position and then single notched into the final intended position. This guidance does not supersede any other requirement to single notch control rods. (2) When moving a control rod three notches or less, the control rod should be single notched for the entire move.

70. G2.2.04 1 Which one of the following identifies the Unit Two Scram Immediate Operator Action that utilizes a different criteria for performance than on Unit One?

A. Tripping of the main turbine. B. Tripping one of the running feed pumps. C. Master level controller setpoint setdown. D. Placing the reactor mode switch to Shutdown. Answer: D K/A: G2.2.04 Ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility. (CFR: 41.6 / 41.7 / 41.10 / 45.1 / 45.13) RO/SRO Rating: 3.6/3.6 Tier 3 K/A Match: This meets the K/A because it is testing the differences between the Units Pedigree: Bank Objective: LOI-CLS-LP-300-C, Obj. 2 List the immediate operator actions for a Reactor Scram. (LOCT)

Reference:

None Cog Level: fund Explanation: On Unit Two the mode switch cannot be placed to shutdown until MSL flow is less than 3 Mlbms. This restriction does not exist on Unit One. Distractor Analysis: Choice A: Plausible because this is an immediate operator action. Choice B: Plausible because this is an immediate operator action. Choice C: Plausible because this is an immediate operator action. Choice D: Correct Answer, see explanation. SRO Basis: N/A

Unit 2 Scram Immediate Actions (OEOP-O1-UG) SCRAM IMMEDIATE ACTIONS Ensure SCRAM valves OPEN by manual SCRAM or API initiatiow 2 WHEN steam flow less than 3 x io lb/hi, THEN place reactor mode switch in SHUTDOWN & IF reactor power below 2% (APRM downscale trip), THEN trip main turbine 4 Ensure master RPV leI controller setpoint at + 170 inches 5 IF:

• Two reactor feed pumps running AND
• RPV level above + 160 inches AND
• RPV level rising, THEN trip one Unit 1 Scram Immediate Actions (DEOPO1-UG)

SCRAM IMMEDIATE ACTIONS

1. Ensure SCRAM valves OPEN by manual SCRAM or API initiationS 2 Place reactor mode swach in SHUTDOWN.

3. IF reactor power below 2% (APRM downscale trip),

THEN trip main turbina

4. Ensure master RPV level controller setpoint at + 170 inches.

5 IF:

• Two reactor teed pumps running AND
• RPV level above + 160 inches AND
• RPV level rising, THEN trip ona

71. G2.2.44 I

-       TC                          indications after initiating SLC during an ATWS.

SLC tiB I z15 SiUB VALVE tONflNUIt( 15-12 P

                                                           ]

P1 1 2I SI 1X: PUMP ZA EY C CCi) A G x x E-6

                               *1         [

0 a 0 S L C FUMP A&3 L41 *-S 1) RE.DIOR PR[ IRL C4SLRR1.,E 1-RC1A PRESSURE CA 1Pl< Which one of the following completes both statements below? Squib valve (1) has failed to fire. lAW 20P-05, Standby Liquid System Operating Procedure, the OATC is required to (2) A.(1) A (2) place the CS-Si, SLC Pump A & B, in the PUMP A RUN position B. (1) A (2) leave the CS-Si, SLC Pump A & B, in the PUMP A/B RUN position C. (1) B (2) place the CS-Si, SLC Pump A & B, in the PUMP A RUN position D. (1) B (2) leave the CS-Si, SLC Pump A & B, in the PUMP A/B RUN position Answer: C

KJA: G2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5/45.12) ROISRO Rating: 4.2/4.4 Tier 3 K/A Match: This meets the K/A because it is testing knowledge of the indications and what action is required based on the system lineup. Pedigree: Previously used on the 2014 NRC exam Objective: LOl-CLS-LP-005, Obj 13-Predict the effect of the following on the Standby Liquid Control System, and based on those predictions use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: a. Failure of one or both squib valves to fire.

Reference:

None Cog Level: Hi Explanation: The SLC squib valve continuity lights are normally lit and go out when fired on SLC initiation. Per OP-05, if one squib valve fails to fire, two pump SLC operation may still continue provided reactor pressure is below 1184 psig, which it is not. Distractor Analysis: Choice A: Plausible because the student may think that the light is illuminated when the squib valve fires and securing 1 pump is correct. Choice B: Plausible because the student may think that the light is illuminated when the squib valve fires and if reactor pressure was lower this would be correct. Choice C: Correct Answer, see explanation Choice D: Plausible because the B squib did not fire and if reactor pressure was lower this would be correct. SRO Basis: N/A NOTE: The SLC pump discharge relief valve should NOT actuate with two pumps operating and only one squib valve open unless reactot pressure exceeds 1 184 psig, wtich is possible during an AWJS even with 10 SRV5 open

2. IF SLCA SQUIB VALVE CONTINUtTYOR SLCB SQUIB VALVE CONTINUITY indicating light on Panel P603 remains on AND reactor pressure is greater than or equal to 1184 psig, THEN PERFORM tile following:

a PLACE SLC PUMP A & B Control Sitch. C4fCSS1r to the SLCPUMPA OR SLC PUMP B position

b. ENSURE tile selected SLC pump red indicating light on.

72. G2.3.12 1 Two operators are required to enter a room that is posted as a Locked High Radiation Area (LHRA) to hang a clearance for scheduled work.

Which one of the following completes both statements below? The radiation level at which a LHRA posting is required is (1) in one hour at 30 centimeters from the radiation source. The LHRA key is obtained from (2) A. (1) >l00mrem (2) the Shift Manager B. (1) >l00mrem (2) a RP Technician C. (1) >l000mrem (2) the Shift Manager D. (1) >1000 mrem (2) a RP Technician Answer: D K1A: G2.3.12 Knowledge 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. (CFR: 41.12 / 45.9 / 45.10) RO/SRO Rating: 3.2/3.7 Tier 3 K/A Match: This question matches the KA because it is testing the rad requirements for entering a LHRA. Pedigree: Bank (from Farley) Objective: LOl-CLS-LP-201-F, Obj. 10 Explain the requirement regarding control of High Radiation Areas per E&RC-0040.

Reference:

None Cog Level: Fundamental Explanation: Locked High Radiation Area (LHRA) criteria is an area, accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 1 .0 rem (1000 mrem) (10 mSv) in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates or an area accessible to individuals with dose rates in excess of 1 .0 rem per hour at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates but less than 500 rads in one hour at one meter from the radiation source or from any surface penetrated by the radiation. The Shift Manager has a LHRA key for emergency use.

Distractor Analysis: Choice A: Plausible because this is the limit for a high radiation area nota LHRA. The Shift manager has a key for LHRA but it is for emergency use, not scheduled work. Choice B: Plausible because this is the limit for a high radiation area not a LHRA. The second part is correct. Choice C: Plausible because the first part is correct and the Shift manager has a key for LHRA but it is for emergency use, not scheduled work. Choice D: Correct Answer, see explanation. SRO Basis: N/A

14. High Radiation Area (HRA: An area, accessible to individuals in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 0.1 rem (100 mrem)flrnsv) in one hour at 30 cm from the radiation source or 30 cm from any surface the radiation penetrates

15. Hot Spot fHS): An accessible, focalized source of radiation with a contact dose tate of greater than 100 mrem per hour and greater than five times the general area dose rate at 30 cm.

16. Licensed Material: Source material, special nuclear material, or byproduct material received, possessed, used, transferred or disposed of under a general or specific license.

17. Lacked High Radiation Area (LHRA): An area, accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 1.0 rem (1000 mrem) (10 mSv) in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates or an area accessible to individuals with dose rates in excess of 1.0 rem per hour at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates but less than 500 rads in one hour at one meter from the radiation source or from any surface penetrated by the radiation.

ACCESS CONTROLS FOR HIGH, LOCKED HIGH, AND AD-RP-ALL-201 t VERY HIGH RADIATION AREAS Rev. 2 Page 9 of 29 5.1 General Instructions (continued)

12. Entry into HRAs, LHRAs, or VHRAs require a briefing pcrAD-RP-ALL-2011, Radiaon Protection Briefings. {i.1.2}

13. HRA, LI-IRA less than 10 Rhr, and LHRA greater than or equal to 10 RJhr master keys may be under the confrol of the Operations Shift Manager for emergency use.

73. G2.3.15 1 Which one of the following identifies the DW radiation value indicated above?

A. 1OR/hr B. 20 RIhr C. 100 R/hr D. 200 R/hr Answer: D K/A: G2.3.1 5 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.12 / 43.4 / 45.9) RO/SRO Rating: 2.9/3.1 Tier 3 K/A Match: This question matches the KA as it requires knowledge of the DW rad monitoring system to answer question. Pedigree: Bank Objective: CLS-LP-11.1, Obj. 03a Describe the function/operation of the following: Drywell High Range Radiation Monitors

Reference:

None Cog Level: Fundamental

Explanation: Drywell high range area monitors provide indications of gross fuel failure and are used to determine emergency plan emergency action level associated with abnormal core conditions. With the function switch in the ALL position the upper (red) scale is used, meter readings are taken from the upper scale between 1 1,000,000 RJh. Current indication of 200 RIh Distractor Analysis: Choice A: Plausible because this is the reading on the bottom scale. Choice B: Plausible because if function switch is not taken into account the answer could be 20 RJh. Choice C: Plausible because if the reading on the bottom scale is adjusted by a factor of 10. Choice D: Correct Answer, see explanation. SRO Basis: N/A

74. G2.4.20 1 A transient has occurred on Unit Two with the following plant conditions:

RPV pressure 1000 psig Drywell ref leg area temp 197°F Rx Bldg 50 temp 135°F Wide Range Level 170 inches (NO26NB) Shutdown Range Level 160 inches (NO27NB) (REFERENCE PROVIDED) Which one of the following completes both statements below concerning the level instruments that can be used to determine reactor water level lAW EOP Caution 1? Wide Range Level instruments N026A/B (1) be used. Shutdown Range Level instruments N0027A/B (2) be used. A. (1) can (2) can B. (1) can (2) can NOT C. (1) can NOT (2) can D. (1) can NOT (2) can NOT Answer: B K/A: G2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10/43.5/45.13) RO/SRO Rating: 3.8/4.3 Tier 3 K/A Match: The question meets the KA because it is testing the knowledge of EOP Caution 1 which deals with the water level instruments availability to determine level. Pedigree: New Objective: LOl-CLS-LP-300-B, Objective 16 Given Plant conditions, determine if the RPV water level instrument is providing valid trending information lAW Caution 1.

Reference:

Caution 1 (EOP-O1-UG, Att. 19, Att. 22 & Att. 31 pages 1 and 2) Cog Level: High Explanation: N026s can be used since reading >20 and RB 50 temp is <140 degrees and N027s cannot be used since in unsafe region for minimum indicated level

Distractor Analysis: Choice A: Plausible because the first part is correct and if Attachment 19 is only looked at then this is plausible. Choice B: Correct Answer, see explanation. Choice C: Plausible because if the temperature was a little higher on the RB 50 foot this would be correct and if Attachment 19 only is looked at for the second part this would be correct. Choice D: Plausible because if the temperature was a little higher on the RB 50 foot this would be correct and the second part is correct. SRO Basis: N/A USERS GUIDE OEOP-0l-UG Rev. 067 Page 90 of 156 ATTACHMENT 22 Page of I

                                       <<Shutdown Range Level Instrument fNO27A, 8) Caution>>

30G SAFE -

               -J 250 LU U
               -J 0

fi it LU 0 200 0 z

                            .-                           -UNSAFE 150  -                              I ifiJ1L1I11 I iIIItIILIiI 150           250           350          450 100            200           300          400 REFERENCE LEG AREA DRYWELL TEMP (°F

75. G2.4.27 1 A fire has been reported and confirmed in the turbine building breezeway.

A fire hose being used to control/suppress the fire. Which one of the following completes both statements below lAW OPFP-O1 3, General Fire P/an? The RD is required to sound the fire alarm and announce the location of thefire (1) A call for offsite assistance to the Brunswick County 911 Center (2) required. A. (1) ONLYonce (2) is B. (1) ONLY once (2) is NOT C. (1) three times (2) is D. (1) three times (2) is NOT Answer: C K/A: G2.4.27 Knowledge of fire in the plant procedures. (CFR: 41.10 / 43.5/45.13) RO/SRO Rating: 3.4/3.9 Tier 3 K/A Match: This meets the K/A because it is testing knowledge of the actions contained in the plant fire procedure Pedigree: New Objective: FPT-CLS-LP-205 Lesson plan discusses the actions for the control room but no objective is listed.

Reference:

None Cog Level: Fundamental Explanation: The operator aid (from the General Fire Plan, PFP-013) for the control room operators states to announce the fire location 3 times. The procedure also states to request off site assistance if a fire hose is used for extinguishing the fire.

Distractor Analysis: Choice A: Plausible because EP announcements are performed once and the second part is correct. Choice B: Plausible because EP announcements are performed once and the second part because the stem says that the fire is under control. Choice C: Correct Answer, see explanation Choice D: Plausible because EP announcements are performed once and the second part because the stem says that the fire is under control. SRO Basis: N/A GENERAL FIRE PLAN QPFP-013 Rev. 48 Page 27 of 35 ATTACHMENT 2 Page of 2

                   <<(Information Use) Control RoomQperator Fire Actions>>

Sound fire alarm, announce location of the fire 3 times, then. .. .......... E

• Announce Fire brigade muster at the fire house.9
• IF fire is outside the Protected Area, THEN announce:

All personnel NOT involved in fire fighting or direct support activities are to evacciate the iiwolved area immediately.

• IF fire is inside the Protected Area, THEN announce:

All personnel in the affected area are to evacuate the involved area immediately and report to your normal work location. If your normal work location is inaccessible, report to the O&M lunch room or TAC auditorium as conditions dictate 0

• Announce:

Use of the PA and radio is restricted to emergency fire communications, except as directed by the Unit CRS for operational safety concerns. 0

2. Announce the fire over Unit 1 and Unit 2 radio channels 0

c. IF the investigating operator confirms a fire AND any of the following conditions exist, THEN immediately request off site assistance by calling 911: C

• Extreme force is necessary to gain entry into fire area C

76. S209001 I During a LOCA and LOOP on Unit One, the following plant conditions exist:

An Emergency Depressurization has been performed due to RPV water level The Reactor Building -17 foot and 20 foot elevations are NOT accessible due to radiation levels. ALL ECCS pumps are unavailable. Which one of the following completes the statement below? The CRS will direct demin water injection to the RPV, lAW OEOP-0l-LEP-01, Alternate Coolant Injection, Section: A. 2.4.3.3a, Demineralized Water Actions, Inject demineralized water through Core Spray Loop A B. 2.4.3.3c, Demineralized Water Actions, Inject demineralized water through RHR Loop A C. 2.4.3.3d, Demineralized WaterActions, Inject demineralized water through HPCI D. 2.4.3.3e, Demineralized Water Actions, Inject demineralized water through RCIC Answer: A KJA: 209001 Low Pressure Core Spray System G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: 41.10/43.5/45.13) RO/SRO Rating: 3.8/4.0 Tier 2 / Group 1 KJA match: This question meets the K/A because an emergency condition exists in the stem (ED followed by no HP injection/high area rad levels). In addition, the SRO is required to have knowledge of the local emergency procedure and that it contains actions the AD must take in the field in order to inject (AD must locally open demin keepfill bypass valves for various injection sources). Contrasting the given conditions, the procedural knowledge of AD field actions, and system knowledge of valve locations will have the operational effect of determining which injection source is viable. Pedigree: New Objective: LOl-CLS-LP-300-J Obj 4a Given plant conditions, determine which system should be utilized to restore RPV water level and/or pressure when executing the following:

a. Alternate Cooling Injection Procedure with EOP-01-LEP-01.

Reference:

None Cog Level: High

Explanation: Demineralized water injection requires knowledge from the LEP that the system Keepfill Bypass Valves are to be opened. The only keepfill bypass valve that is not inaccessible due to radiation is the Core Spray Loop A Keepfill Bypass valve. Therefore, Core Spray Loop A is the section to use in order to inject demin water to the RPV. Distractor Analysis: Choice A: Correct answer, see explanation. Choice B: Plausible because it is a section in 2.4.3, it is incorrect because the keepfill bypass valve for RHR loop A is inaccessible due to radiation Choice C: Plausible because it is a section in 2.4.3, it is incorrect because the keepfill bypass valve for HPCI is inaccessible due to radiation Choice D: Plausible because it is a section in 2.4.3, it is incorrect because the keepfill bypass valve for RCIC is inaccessible due to radiation SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J. Requires the SRO to evaluate the emergency conditions in the stem, and contrast those conditions with the given local emergency procedure sections and then select the appropriate procedure section. 2.43 Demtneralized Water Actions NOTE OE&PC-05J412 slates lrat me Shift Marae ma grant immediate access to a loceO hs3fl rIiafoo area tO ma fltaiO tre tealifl d safce iari1 persccnel On me genefat pobi( er IOCFREO 53is] using the ejs mamlaned in the RIatmi Protection ke1 locken near the Unit 2 OG panel I t) Al ANY TIME a locke,j Lgfl taoaton area is entered, e1tr,m.1 RIati Prclecton SLDpc%l 1)-lEN nottly ftP RO 2 Monitor id control MLD tank leret greater than 13 reel AO NOPE Demmeralied water transtec pomp capacity is 400 gpnt 3 Perform for sstems NOT eraLng ANL1 available to proede injection to RPV D RO 3 ttijecl demtrrerali:ed water tencrigh Cole Spray Loq A

1) Ettsure P21 FOC5A 1)01)03rd Injection N) OPEN C NO (2) Ensure P21 FEC-tA (0U)board Injection \ Iv) OPEN C RO ALTERNATE COOLANT iNJECTION CEOP-01.LmP.O1 Rca 34 Page 20ot47 23i DeirriireaIized Water Actroits (continued)

NOTE P21-F023A is located on Reactor Suhling C

13) Open £21 F02A Core Spa-, Loop A Ceeplil Stal:on B ,viss .ahe;

( lneCt rtt:e4aW lr4i R,IR .op A Ii FnsuteEI1-F1AiIntordnclitVi1OPfN U RU Throttle E11-FJ17A iOutLo lnje.ion  ;. U RU rE11-Fa2A EJ1FU5 in El1.FCTA NOTE are ixa:ei ItSe l4P. me:snre U Open El l-FCS2 IRHR Li ç A eeØJ Stitcii U AU

                    $        Open El 1-FCTE RHR Sstem DenineraIe Wster FII ViI                                                  C AU 0      IflleCl ernjrierah:e1 waler lhforfl HPCI Ensure E4l4T12 iHPC     Miii Fiw  Ecasu To roc V1 ClOSED                                               U NO 21     A l&C DA Row RI. Comct B24 rHPC lirFl SV To Supp Ceambor Val.e E31 FSI21 price LrCale OFF                                              U AD NOTE E3litCsliiunNRlR                                                                       U 131      Open E4 I-V lOG 1HPCI cee p1l 5ta onE ,pass  al         U AU 131     Ensure E4l-F7 1PUIVP Disch3re Viel OPEN                  U NO C     Inject   dernneraloeU wEer I csrr ROIC l      Ensure EEl-FOSS IRCIC Mn FlOe Bpase To Tons Viul CLOSED                                              U RU
          ,SLTERUATE COGC3 I14JECTO1l                                          EEOP-ai-LE?-Th Re34 Pe22ot4 24J    Deminerlized Water Actions Icontinnied) 21     At MCC XDB Row Ni ComEt B4? tRCIC Wit Flo BEci5S To Supp Pool Vii E5I-F5191, place breaker OFF                                                      U AU NOTE E51-V70 in IateO  it SRHR                                                              U 131     Open ES 1-V?O IRCIC KeplT Slaron Eip3S VOiie)            U AU ii     Ensure E51-F0I2 IPump Discharge iv) OPEN                 U NO

77. S212000 1 Unit One is at rated power performing OPT-Ol .1.6, Reactor Protection System Manual Scram Test.

The Reactor Scram System A pushbutton has been depressed. RPS Trip System A Scram Groups light for groups one, two, three, and four are illuminated (REFERENCE PROVIDED) Which one of the following completes both statements below? The scram pilot valve solenoids associated with these lights are (1) Tech Spec 3.3.1 .1, Reactor Protection System Instrumentation, Condition B (2) required to be entered. A. (1) energized (2) is B. (1) energized (2) is NOT C. (1) de-energized (2) is D. (1) de-energized (2) is NOT Answer: B K/A: 212000 Reactor Protection System G2.2.44Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41 .5 / 43.5 / 45.12) RO/SRO Rating: 4.2/4.4 Tier 2 / Group 1 K/A match: The applicant must interpret the indications (scram group lights) in comparison with the expected conditions for the given action (Sys A pushbutton depressed). The applicant must then use that knowledge to determine whether the system meets the given limited condition for operability as stated in the TS. Pedigree: New Objective: LOI-CLS-LP-003 Rev 3 Obj 27 Given plant conditions, determine whether given plant conditions meet minimum Technical Specification requirements associated with the Reactor Protection System.

Reference:

TS 3.3.1.1 Cog Level: High

Explanation: Part I When group lights are OFF that is indication that the solenoid valves are de-energized. Groups 2 and 4 remained lit therefore, they are energized. Part 2 ONLY Condition A and C are required to be entered, due to a failure of the A3 scram channel. Only one requited channel is inoperable, but a loss of manual scram function has occurred resulting in RPS trip capability not maintained. Distractor Analysis: Choice A: Part I is the correct Answer, see explanation. Part 2 is plausible because Groups 1 through 4 are also in trip System B (although for the SV-1 18s), a novice applicant may assume that failure of these solenoids in Trip System A would mean they would not function in Trip System B. In addition, one channel per trip system is required, and since there are only lights for groups 1 through 4 in both trip systems, a novice applicant may assume that all required channels are mop. Choice B: Correct Answer, see explanation. Choice C: Part I is plausible because ARI system uses energize to function valves. Part 2 is plausible because Groups 1 through 4 are also in trip System B (although for the SV-1 18s), a novice applicant may assume that failure of these solenoids in Trip System A would mean they would not function in Trip System B. In addition, one channel per trip system is required, and since there are only lights for groups 1 through 4 in both trip systems, a novice applicant may assume that all required channels are mop. Choice D: Part I is plausible because ARI system uses energize to function valves. Part 2 is the correct Answer, see explanation. SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)J. Requires the SRO to evaluate the failure of PT for RPS, and select the appropriate> 1 hour IS condition.

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.&     One or more requIred        A.I         ° 3    charnel III hip       t2 hoIrs ch3lfleis lr3perab e a

A.2 NOTE 2 hours Not apptcab e lot Fund ns 2.3 2.b2c,23 or2l. P 3d ass2dtate hip tyslen, IC hip canIIrueCI

6P0 IrstruTn1llot 3.3.11 ._:t)N crue DONDIll D1 REQUIRED ACTh)t, ccqPLETh DiN TtP.fE S 5.1 P3cnar9eItt oleLflp flOL! Not ppIcat.e for FL,ct DflB senr It nip. D.2.b,ZD,2 or2t.

                                -a Ole OF fore Fun000rs wIth      5.2     P ac orS irS sysfum n     floL!5 one or more reqJIre1                   tfl.

Cflslne S Ir]pernt. S l Scm trip 55iT5. C. One or note Functhors wIhtJ C. I estore RPZ t1p c3pbIIfy 1 flotr lr spbHhty r Ins vuIccJ. C. euIrEI Actir snd 0 I Enter the CanSton IrnnhI5ie this ocJsi C-inipIeton Time refeeriet Ii OtC]flthItn A, 5, ftC not TaOS 3..l.1-1 flrthle nsf. cllSflnEi. S Au reuIre1 tiy qtreth 5.1 eSucau ThERMAL 1 floLrs ADODI 0.1 reTerenceth I P0ER to 26% RTP Tat331.1-l. DEFINmON OF INSTRUMENT CHANNELS AND TRf.P SYSTEMS FOR SELECTED INSTRUMENTS R Pace 58 ATTACH ME Paie CZIA-S3A, B; C72A-53A. B INSTRUMENT NUMBER: 071A-S3A, B; CZ2A-53A, B INSTRUMENT NAME: Manual Sctarn IS

REFERENCE:

3.3.11; TRM Table 3.3 1.1-1.11 TRIP CHANNEL: A3 S3A B3 S3B TRIPSYSTEM: A3-S3A B3 -S3B TRIP LOGIC: A3 and B3 = Reactor scram PLace channel in TRIPPED condition by: Pull fuse

The Reactor Manual Scram relays deenenize me Scram pilot ate solenoids for the RPS Thp System. The SV1i7 ales are in RPS Trip System A. The SV-f1 Qalies are in RS Trip System B Shorting links are normally installed around the auxiliar trip relay contacts in the Manual Scram circuits allowing this Scram signal to be bypassed. These shorting links are located in the back panels and are co[or coded red for identification. NOTE: The shorting links are removed prior to and during the time any control rod is withdrawn (except for control rods removed per Technical Specifications) during operation in reeling mode or during shutdown margin demonstration. There are o shorting links per RPS Trip System, a total of four for the entire Reactor Protection System. SD-03 Rev 12 Page 27 ot 9B

ttEACQR POTCr ON svzrte P.t?u.sL OPT-Ol. t .6 OCP.AM TEST Res 18

                                                                                 °aze 4 of 1 1 1.0    PURPOSE This test is performed to determine the OPERAS tLlfl of the Reactor Prccecticn System Manual Scram unction 2.0    SCOPE This procedure perfcmms the following:

• Aquartedy Channel Fur,ct:onal Test perTS SR 33.1 1orTabIe 3.3 1.1-1 Function 11, Manual Scram.
• Satsifes aporion of the 24 ntnthiS SR3 3.1.1 15 cc.gtSysben FunconatTestfcrTabe3.S1.1_l Function II. ManuatSoram.

3.0 PRECAUTiONS AND LIMIFAflONS 1 A haV-scram sgnat nmll e:;ist until RESE ....... 0 1.0 GENERAL INFORMATION The ollcwing annunciators will alarm durng. the performance of this test:

• A-05, I-S. Reactor Man ual Scram Sys A
• A-OS, 2-8 Reactor Man ual Scram Sys B 5.0 ACCEPTANCE CRflERtA

1. This test nay be consered satsfactorj tthen alc the fcllceing criteria are met:

a. A thp is indicated on P35 A and alarmed on RTGB Panel R12-P603 vahen CT1iC72-S3A Manual Reactor Scram System Ai push button is depressed

b. Amp is indicated on RPS B and slanned on RTGB Panel R12-P603 when Crli:C72:f.S36 iManual Reactor Scram System B:. push button is depressed.

c. Tne Scram ,alsw solenoids are DE-EfcSRGCD ihen the associated RPS is topped

ftECTOR PRo1EC ON SYSTEM MiU 8PTfl1.1 .6 ScftM TEST Rei. 10 P3ge 6 of 11 iS INSTRUCTIONS

7. I Test Preparation 1 Obtain lint CR5 permsson to perform dms test 2 Ensure all prereQuisite hszed in Section 65, Prerequ:sites are rer NOTE The length of lime a haff-scram is sealed-in is to be minimced I lFdur;r the performance cfthis procedure. the expected test resulte from a hat-scram n*riabcr are NOT obserued THEN immedately reset the half-scram and notify the Unit CR5 7.2 Marwal Scram A Test Depress CT1(CT21-53A lManual Peactor Scram System push button and observe the foio,nt actors occur

a. Plant Process Computer Event Log dsplays Man ial Scram CharnelAmp (Computer 4cint Deii:

b. IF the proper Plant Process Computer Event Log (Computer Point D533 was NOT receised in Step l.a THEN generate aiaC c Manual React r Scram System A push button light comes ON

d. A-Ott i-a, Reactor Manual Scram 5ys A, ALARMS

e. RPS Thp System A Scram Group lights 1, 2,3, and 4 located on Panel Hl2-PE33 are ClEF, ndiea:ing So-am valve solenoids are CE-ENERGIZED RPS Thp System A Scram Group lights 1, 23, and 4 located on RPS A Panel H12-Pe-D are OFF. ndcanrg Scram vale solenoids are CE-ENERGIZED

78. S215001 I Unit Two is at rated power. A TIP trace is in progress.

TIP D Valve Control Monitor indications are as follows: Squib Monitor Light ON Shear Valve Monitor Light OFF TIP Ball Valve OPEN Light ON TIP Ball Valve Closed Light OFF TIP D Drive Control Unit indications are as follows: MODE Switch MANUAL IN CORE Light ON (REFERENCE PROVIDED) Which one of the following completes both statements below? Tip Valves are Group (1) PCIVs. Tech Spec 3.6.1.3, Primary Containment isolation Valves (PCIVs), Condition (2) is required to be entered. A. (1) 2 (2)A B. (1) 2 (2) B C. (1) 6 (2)A D.(1) 6 (2) B Answer: A K/A: 215001 Traversing In-Core Probe G2.2.44Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5/43.5 / 45.12) ROISRO Rating: 4.2/4.4 Tier 2 / Group 2 K/A match: The applicant must interpret the given TIP system indications and controls to verify the status of the system. The applicant must then compare those indications and controls with the required conditions for the given mode as stated in the TS to determine the appropriate condition. Pedigree: New

Objective: LOI-CLS-LP-009.5 Obj 9 Determine whether given plant conditions meet minimum Technical Specification requirements, including the Bases, associated with the Traversing lncore Probe System.

Reference:

TS 3.6.1.3 Cog Level: High Explanation: Part 1: Tip Valves are Group 2 PCIVs. Part 2: Squib Valve Monitor Light On indicates that the squib valve continuity is lost. Therefore, The Shear Valve in the Ball valve and shear valve assembly is inoperable. The ball valve remains operable. Therefore, Condition A is entered for one PCIV inoperable. Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because squib Valve Monitor Light On indicates that the squib valve continuity is lost. Therefore, The Shear Valve in the Ball valve and shear valve assembly is inoperable. In addition, a novice applicant may assume that with the MODE switch in MANUAL, the ball valve would be inoperable as well. (However, the only position of the MODE switch that would make the TIP inoperable is OFF.) With Two PCIVS inoperable, condition B would be entered, Choice C: Part 1 is plausible because group six PCIVs also isolate on LL1 and High DW pressure. Part 2 is plausible because it is correct, see explanation. Choice D: Part 1 is plausible because group six PCIVs also isolate on LU and High DW pressure. Part 2 is plausible because squib Valve Monitor Light On indicates that the squib valve continuity is lost. Therefore, The Shear Valve in the Ball valve and shear valve assembly is inoperable. In addition, a novice applicant may assume that with the MODE switch in MANUAL, the ball valve would be inoperable as well (However, the only position of the MODE switch that would make the TIP inoperable is OFF. ) With Two PCIVS inoperable, condition B would be entered, SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. The SRO applicant is required to select the appropriate> 1 hour TS condition based on the status of the TIP system indications.

Table 09.5-2 Valve Contrc4 Moultor Indications hdi cation Comment Squib Monitor Light ON indicates that the W Shear Valve squib circuit connuity has been ba Shear VaI Monitor Light ON indicates that the squib charge in the TP Shear Valve has been detonated iF Ball Vahe OPEN Light ON indicates that TIP Ball Valve is OPEN. liP Ball Valve CLOS ED ON nd icates that TIP Ball Valve is C LOSED Light lime Delay Light ON indicates that the liP Ball Valve was not OPEN within 6 seconds from when the detector Iefl e in-shield position and the TIP Drive Motor should have stopped. Purge Light hdicates that the solenoid for the IldexerPu:rge System shoUld be energized OPEN. Fuse F5 Continuity Light hdicates poerto PCIS 6ro 2 Bus in drawer is available (Fuse F5 is not blown). 50-09.5 Rev. 7 Page 20 of 58

Table 09.5-3 TIP Drive Control Unit Indications Lqdication Comment DETECTOR POSITION Dinaric digital display of detector positon.

ilhmiinated dints) 000t reference point about one foot behind the Indexer: i750 In Shield position)

CORE LBIIT Static digital display of pre-prograninsed core top or (illuminated digits) bcrkm limits of sele: ted channel READY Light Indicates that Indexer is properly aligned to selected channel. CORE TOP Light Detector is at top of core. PT COPI Liaht Detector is above core botiPom limit PT SHIELD Light Detector is in Shield Chamber. SCAN Light Axial Flux Profile is being recorded. LOW Speed Light Detector is being driven at 3 inches per second. (15 feet per minute) REV (Reverse) Light Detector moving away from top of core. flUD (Forward) Light Detector moving towards top of cove. VALVE Light ON if TIP Ball Valve is CLOSED. Table 09.5-4 P601 TIP Indications Indication Comment TIP Valve Status Green Light

                    -                             Green Light ON indicates that each TIP Ball Valve is FULL CLOSED.

TIP Valve Status Red Light

                    -                             Red Light ON indicates that a TIP Bali Valve is NOT FULL CLOSED.

SD-OP 5 Ret 7 Page 21 of 58

Low Speed OFF MaRes lowspeed dhve a function of detector positon and i:ndependent of operator control. ON Initiates continuous low-speed detector drtve Core Limit TOP Permits digital display of selected channel pre-prog rammed top-core [imit which corresponds to tap of active fuel BOTTOM As above, except pre-programmed core bottom limit is displayed. (The core top and bottom numbers are different for each TIP channel because of different lengths of guide tube run). Mode (Switch 5-7) OFF Deenergizes power supplies in Drive Contrn[ Unit MANUAL Positions detector in conjunction with the FWD and REV position of manual switch S-i AUTO Pernifis automatic mode of operation when Auto start 3-2 is pressed. Manual Valve Control (Spring Return to Closed Position CLOSED Permits TIP Ball Valve to open automatically when mechanism is operated OPEN Opens TIP Ball Valve without energizing the Drive in the TIP Drive Mechanism. X-Y Recorder (Figure 09.5-14) Alternate plotting capability via AFORA sotware Controls on the X-Y recorder drawer are as follows: SD-09.5 Rev 7 Page 24 of 58

4.0 SYSTEM OPERATION 4.1 Normal Operational Relationships TIP traces are required to be performed periodically Technica[ Specifications require calibration of LPRM detectors at least once every effective full power month (EFPM). Traces are done to obtain new Gain Adjustment Factors (OAF) as calculated by the process computer for each LPRM. These OAFs can be used to adjust the current applied to the LPRM detectors. Current adjustment is required due to loss of detector sensitivity which results from exposure to the fission process in the core. The TIP Detector is run through the dry tube within the IPRM assembly containing the LPRM to be calibrated. A comparison is made between the TIP output and the existing LPRM reading and a OAF is calculated. The current applied to the LPRM detector is adjusted, if necessary. TIP Detector calibration is also periodically required. Capability for calibration of TIP probes is provided by a common channel which can align each TIP to the center LPRM assembly (28-29) (Figure 095-15) and TIP Dry Tube. Each TIP is run through the center LPRM assembLy Readings from the TIPs are compared and the gains are adjusted to meet an average value of the four TIP readings and so that the gains for the TiP channels tall within a specified band. The TIP System can he operated in an Automatic mode or a Manuai mode. OP-09.1 in conjunction with OENP-24, 15 covers precautions. initial conditions, and specific instructions related to the operation of the TIP System Regardless of the niode of operation there are some precautions that must be observed. The TIP Machine should never be turned off with the detector inserted past the TIP Ball Valve. This conditions prevents the isolation logic from retracting the detector and closing the TIP Ball Valve should an isolation signal he received. Also, the TIP Machines should not be left unattended if the detectors are in motion. 4.1.1 Automatic Operation Sequence The mode switch on the selected Drive Control Unit is placed in AUTO and the manual switch and low speed switch remain in OFF. This gives a permissive to the TIP logic to be run automatically. The auto start pushbutton (Drive Control Unit) is then pressed. The detector will automatically move from Shield Chamber to the entrance of the Indexer at low speed (V,1sec or IS if/ruin) and will then stop. SD-09.5 Rev. 7 Page 27 of 58

3.6 CONTAINMENT SYSTEMS 3.6.1.3 Phmars Containment Isolation Vaes (PClsi LCD 3.6.1.3 Each PCIV, e:wcept reactor buIdng-t-suppression chamber vacuum breakers shall be OPERAELE. APPLDCABIILITf: MODES 1, 2, and 3. When assocdated instrumentaton Us required to be OPERABLE per LCD 3.3.6.1, PrinianiContaUnmentllsolation FInstnmentation. ACTIONS NOTES - . 1 Penekaton flow paths may be unisolated UntermUttently under admUnUsfratve controls.

2. Separate Condtion entry is allowed for each penetraton flow path.

3. Enter applicable Conditions and Requred Achons for systems n,ade inoperabe by PCPt/s 4, Enter applicable Condibens and Requdred Acttons of LCD 3.6.1.1, Prtmwy Contahnment when PCIV leakage results in exceeding overalU containment leakage rate accept-ante citteria.

CONDITION REQUIRED ACTION COMPLETION TIME A. NOTE-- Al Isolate the affected 3 hours Only applicable to penetrabon flow path by penetration flow paths with use of at least one cosed two PCIVs. and de-activated automatc

                     --                    valve, closed manual va1ve.

blind flarge, or check valve One or more penetration with flow thrngh the calve flow paths wth one PCIV secured. inoperable except for MSIV leakage not within limit. AND (mnin, ,4& PCIVs 3.6.1.3 ACTIONS (conbnued) CONDITION REQUIRED ACTION COMPLETION TIME B. NOTE 6.1 Isolate the affected 2 hours Only applicable to penetration flow path by penetration flow paths with use of at least one closed two PCIVs. and dc-activated automatc valve, closed manual valve, or blind flange. One or more penetration flow paths with two PC P/s in operable except for MSIV leakage not within hmit. C RflTC C 4 .1.,+.-, H,.-. ,.fl.......,A

TABLE 12-2 Prinhary Containment is.olatiocs System Group led*ion Insfrumentation Setpoints ISOLATiON ISOLATiON TR:PSETPONT NOTES GROUP SI LNAL Tech Scec ActaJ Allowable INrte 1) GroupS HighSteanPtow c275% 220% NoteS Low Steam Pressure S3peg 71] peg High Turd 6th Pressure 6 psig 5 psi; Seam LireireaHlernp .17EF Note4 SeamLireturrelHgh c2EI]F 165R14]F Note4 Ant Tarp SleamLineTunreldTHigh EDF 4?F No1e4 Eqip Area High Tenp - 1TS F 1i15F EpipAreaigh ;SEiF 47t C-roup6 Lw,Leeel*1 :157 16W HighDvlIPressure l8psg lipsi; Pa Edg Ethaust Hi Pad 16 mRtr 4 mRhr Pa EdgEahaustHiTenr 14/A 135F NoteS High Main Stack Pad 00CM DOOM Note 2 C-coup? LcaStearnPressureM.1C l4ps; 115 peg High Crwefi Pressure s 1.8 psq 17 psi; Citup8 LowLeuel#1  :- 157 16W Hind Steam Corre Ftessure 137 psip 130.8 psi C-coup 9 Low Steam PressureAND a 53 psi; ?Ups High Cr1v,eII Pressure s 1.8 peg 17 psi; 0mw 10 Low Lecel #3 11/A 4.57 Hgh Ci)w&l Pressure AND 17 psi; LowPeactrFressure 410pe Note 1: AJI AcuaF vai.es from TEM Note 2: Stack radarn h gh level is Safled in acowdrcewrth the Dffshe Dose CaiIaton Manusi Note 3: After a 23.5 ninide tine deaw Note 4: After 27 minite tare delay NoteS: After aS eea:nd time delay NoteS: Specific Actial velues from SOP Users Gu:* Attadirrent 1 SC12 Pee, 11 Pae84gh2O8 TABLE 12-2 Phnury Containment Isolation System Grrnç hdation Instrumentation Selpoints ISOLATiON ISOLATiON TRiP SEFflOINT NOTES GROUP SICiIAL Tech Spec Arcuel Alcwable iNote lj VaLe Group 1 Low Lev&#3  :-i-13 45 Main Steam Lire High c 197 F 19F Tetrp TurbtieSldgAreaHiTefrp Pith 1CF NoteS Main Steam Line High Arm a 138% 137% - Notin RUN 33 3O3 Lhit2 only Low Condenser Vaouum a 7.5 Hj tO Hg Low Steam Pressure 825 crs 836 peg Grcup2 LowLeu&#1 ..IEO itS High DryvII Pressure 1.Spsg 1.Tpsig Group3 LrmLev&#2 alor 105 HghCiff9cw a73gpn 43m Note3 AJea High Temp a lEF 141] F AreaVentdTHigh aSOF 47F HELS Isciaton a 121] F I 16 F NRHX Cutlet Tarp High Pith 13&F NoteS SLClnitatcn N/A Pith Group 4 High Steam Row a 275% 223% NoteS Low Steam Pressure a 1 psig 115 psig High Turd 6th Pressure a 9 psig 7 psig Steam LirekeaHiTemp a22]F 165 F Steam LineTunrel Hih c2Ce] F 165 Ff193 F Ant Tarp Steam Lire Tunnel dT High a SO F 47 F Epjip Area Hign Temp a 175 F 16SF

79. S219000 1 Unit Two is operating at rated power with RHR Loop A operating in suppression pool cooling mode.

A-O1 (2-8) RHR Relay Logic Pwr Failure, is in alarm due to a blown fuse affecting RHR Logic A ONLY. (REFERENCE PROVIDED) Which one of the following completes both statements below? If a LOCA signal were to occur, 2-El 1 -FOl 5A, Inboard Injection Vlv, (1) open automatically on low reactor pressure. lAW Tech Spec 3.3.5.1, ECCS Instrumentation, the required channels (2) required to be placed in trip within 24 hours. A. (1) will (2) are B. (1) will (2) are NOT C. (1) wilINOT (2) are D. (1) wilINOT (2) are NOT Answer: A K/A: 219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode A2 Ability to (a) predict the impacts of the following on the RHR!LPCI: TORUS/SUPPRESSION POOL COOLING MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41 .5 / 45.6) 12 Valve logic failure RO/SRO Rating: 3.0/3.1 Tier 2 I Group 2 K/A Match: This meets the K/A because it is testing whether RHR in suppression pool cooling mode will transfer to LPCI injection mode with a failure of divisional valve logic, and whether the LCO action statement for condition B should be applied. Pedigree: New Objective: LOl-CLS-LP-01 7 Obj 7. Given plant conditions, determine if the RHR system should automatically initiate in the LPCI mode. (LOCT)

Reference:

T.S. 3.3.5.1 Cog Level: High

Explanation: Part 1: With the plant at rated power and torus cooling in service on loop A, RHR loop A is not in its normal standby lineup. The loss of Div I RHR relay logic power, will result in the failure of the A loops suppression cooling valves (f024/28) to automatically close on a LPCI initiation. However, the FO15A will auto open from the div 2 logic. Part 2: The loss of power to the Div I RHR logic will result in the loss of the functions applicable to condition C. Distractor Analysis: Choice A: This is the Correct answer, see explanation. Choice B: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because a novice candidate might believe the note that required action B.2. is only applicable to functions 3a and 3b also applies to condition B.3. - Choice C: Part 1 is plausible because a candidate might believe a loss of Div I logic would prevent the auto function of the FOl 5a, since the F024 and F028 will not automatically close under these conditions. Part 2 is plausible because it is correct, see explanation. Choice D: Part us plausible because a candidate might believe a loss of Div I logic would prevent the auto function of the FOl 5a, since the F024 and F028 will not automatically close under these conditions. Part 2 is plausible because a novice candidate might believe the note that required action B.2. is only applicable to functions 3a and 3b also applies to condition B.3. SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)J. The SRO applicant is required to select whether an action statement is applicable given a loss of Div I RHR logic.

ECCS Instrumentation 13.51 ACTION S CO NDITIO N REQUIRED ACTION COMPLETION TIME B. (continued) 82 ----NOTh-- Only applicable for runctions 3.a and lb. Declare Hiah Pressure 1 hour from Coolant Injedion (HPCI1 discoverg of loss of System inoperable. HPCI initiation capability AND 8.3 Place channel in trip. 24 hours C. As required by Required Cl ---NOTES----- Action A.1 and referenced in 1. Only applicable in Table 3.3.5l1. MODES 1. 2. and 3.

2. Only applicable for runctions 1.c. 1.d. 2,c, 2.d. and 2.f.

Declare supported 1 hour from feature(s) inoperable when disco:eri of loss of its redundant feature ECCS initiation capability initiation capability is for feature(s) in both inoperable. dM si ons AND C.2 Restore channel to 24 hours OPERABLE status. (continued)

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I Page29of34 AUACHMENT 4 Paoe I of 2 cc Plant Effects from Loss of DC Distiibution Panel 34(4A)>> RCIC: Will NOT shutdown on reactor high water level, inboard isolation logic INOPERABLE (ES1-F007. -FOIl, and -P062 will NOT auto close). Valves E5t-F005 and -F025 fail closed. ADS: ADS Logic B is INOPERABLE. ADS will initiate from ADS Logic A ifcoreSpray Pump Borboth RHR Loop B pumpsareoperating. HPCI: Will NOT auto inWate, outboard isolation logic INOPERABLE (E41-F003. -F041, and -P075 will NOT auto close). HRCI how controller and EGM INOPERABLE (no flow control or indication), HPCI trip logc INOPERABLE, valves E41-F053, -P054, and -F026 fail closed. HPCI isolation is required in accordance with APP 112)-A-01 5-5. CS Loop A: Will NOT auto initiate (manual operation possible but miPimum flow valve will NOT auto open, and injection valves can NOT be opened simultaneously). RPS Logic A: Will NOT have 10 second time delay pr[or to reset of full scram, power lost to backup scram valves. RHR Loop A: Will auto initiate from RHR Logic B, however the following effects exist: Pumps can NOT be restarted it stopped by control switch, pumps will NOT trip on No Suction Path Interlock. LOCA interlocks NOT functional, mm flow valve will NOT auto open, Loop A Containment Spray can NOT be initiated. If a loss of DC Distribution Panel 3B(4B) has also occirned, RHR Loop A will NOT auto initiate (manual operation possible)

80. S239002 1 Unit One was operating at power when a Group 1 isolation and reactor scram occurred.

Reactor pressure is 950 psig and being manually controlled by SRVs. An SRV is stuck open with a stuck open SRV tailpipe vacuum breaker. Torus and Drywell sprays have been initiated lAW PCCP (REFERENCE PROVIDED) Which one of the following completes both statements below? The SRV is discharging through the open vacuum breaker directly into the (1) The highest EAL classification for this event is a(n) (2) A. (1) drywell (2) Alert B. (1) drywell (2) Site Area Emergency C. (1) suppression chamber air space (2) Alert D. (1) suppression chamber air space (2) Site Area Emergency Answer: A K/A: 239002 Safety Relief Valves A2 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41 .5 / 45.6) 01 Stuck open vacuum breakers RO/SRO Rating: 3.0/3.3 Tier 2 / Group 1 K/A match: THIS QUESTION WAS PRE-SuBMITTED FOR APPROVAL. The applicant must determine that the effect of the stuck open vacuum breaker on the SRV is that it is now bypassing the pressure suppression function and discharging directly into the DW. The applicant must also determine that the ultimate effect of the given conditions is that the RCS barrier has been lost, and that the consequences of this conditions and its potential effects on the health and safety of the public are mitigated by declaring an ALERT lAW with OPEP-2.1. Pedigree: New Objective: LOl-CLS-LP-020 Obj 1 5e Given plant conditions, predict how ADS/SRVs will be affected by the following: Failure of the SRV tailpipe vacuum breakers.

Reference:

OPEP-02.1

Cog Level: High Explanation: Part 1: SRV tailpipe vacuum breakers relieve to the drywell, therefore a stuck open tailpipe breaker with a stuck open SRV would discharge steam directly into the drywell. Part 2: With a stuck open relief valve and stuck open vacuum breaker, a LOCA is occurring. With Torus pressure sufficient to warrant drywell sprays (11 .5 psig) drywell pressure has more than exceeded 1 .7 psig. Therefore the conditions for a loss of the RCS barrier ( Primary Containment Pressure >1 .7psig due to RCS leakage) are met and an ALERT should be declared based on FA1 .1 loss of RCS barrier. Distractor Analysis: Choice A: Correct Answer, see explanation. Choice B: Part I is plausible because it is correct, see explanation. Part 2 is plausible because a group 1 isolation has occurred which is a primary containment isolation signal, an unisolable LOCA is occurring, and a novice candidate might assume that this meets the conditions for a loss of the containment barrier (Unisolable direct downstream pathway to the environment exists after primary containment isolation signal), this loss coupled with the loss in the explanation would meet the criteria for a loss of two barriers, and a declaration of an SAE lAW FS1 .1 They also might think that this condition would not be consistent with a LOCA. Choice C: Part 1 is Plausible because with a failed open Suppression Chamber to Drywell Vacuum Breaker would allow DW steam to go directly to the suppression Chamber Air Space. in addition, a failed open SRV with a broken tailpipe and broken downcomer would directly pressurize the torus air space. Part 2 is plausible because it is correct, see explanation. Choice D: Part 1 is Plausible because with a failed open Suppression Chamber to Drywell Vacuum Breaker would allow DW steam to go directly to the suppression Chamber Air Space. in addition, a failed open SRV with a broken tailpipe and broken downcomer would directly pressurize the torus air space. Part 2 is plausible because a group 1 isolation has occurred which is a primary containment isolation signal, an unisolable LOCA is occurring, and a novice candidate might assume that this meets the conditions for a loss of the containment barrier (Unisolable direct downstream pathway to the environment exists after primary containment isolation signal), this loss coupled with the loss in the explanation would meet the criteria for a loss of two barriers, and a declaration of an SAE lAW FS1 .1. They also might think that this condition would not be consistent with a LOCA. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J The SRO applicant is required to select the appropriate ALERT emergency classification based on the given conditions which equates to a loss of the RCS barrier.

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                                                             ._3 lIE lI2 Pl[,PT STE c

1: i LINE t -u CL, 0 C) permissives. 4.2.2 Failure of the SRV Tailpipe Vacuum Breakers In the event an SRV tailpipe vacuum breaker fails in the open position. the result is a direct path of steam from the reactor to the drjwetl (i.e., LOCA). In the event an SRV vacuum breaker fails in the closed position, the result is the possible creation of a vacLium in the tailpipe upon closure otan SRV, resulting in the draNing of water into the SRV tailpipe. SD-20 Rev 3 PAGE 27 of 62

FSti[112131 I I FAIIIII2I3I I I I Loss or pcenbal Icss c two bartiers TabIe F-i) Any Ices or ny potential Ices c iher FLeI Clad or RCS

                                                                    *:Table F-1i table F-I Fission Product Barrier Threshold Matrix Reactor Coolant System Barrier                                                                        Containmeni 55                          Loss                              Potential Loss                                         Loss

1. RPV leiel cannot be resred and led nnbjad>TAFcrcaosnotbe de[errrir,ed 1 UNISOLA9E break in cry of the I UNlSCLGLE primary system 1. tJfUSC1LA2&E pririary cern leakage that eakae at results in eaceeding results in e edin one or mere Secondary

                  - Main 515am re                 i     EITHER of the oIlcwing:                    Containnent area temperature Metimum
                  - HPCI steam iine                        One or more Secondary                   Safe Ciperahng Lirrs
                 -  RCIC steam line                        Ccn1ainmen area ralic1icn                JEOP-O3-SCCP Table SC-I
                  - RWCU                          I        Maximum Niotmnal Oper.atng Limils
                  - Feedwatet                     I        (OEOP-C-SCCP Table SC-3)

One or more Secenary 2 Emergency Depreceurizalion is i Cantainmnent area temperature required Mciimum NoITnal Operating Lintits (UEOP.D3-SCCP Table SC-i) I PhmaryCancanmnenlpressure 1. LLNNEDrapiddropin Primary

1.7 psg due to RCS leakage ontainrrent pressure following Primary Contcir,rtent pressure rise Ncne

2. Primary Containment pressure response not consistent with LOCA conditions

1. Cyssell radiation > 27 R(hr with reactcr shibitcn Nos Note

1. UNISOLAELE d:reot downstream pathway to the eruireriment exists alter Primary None Containment isoladon sign&

iI Sun[XIufls FL[t itri ii LU,4 9I1 4.2.7 Containment Response with Vacuum Breaker Failitre Suppression Chamber to Drywen Vacuum Breakers Failed Opeli Steam flows from drwe[I to suppression chamber tttrouh the open vacuum breakers equalizing pressure beeen the to immediately. The steam is not forced through the watetr of the suppression pool: therefore it wi[l operate only as a surface condenser. As a result. the drjweli pressure will probably exceed the des[gn pressure. To prevent this occurrence, light indication is proded for each vacuum breaker. It indicates if the valve is off its seat and i:s displayed in the Contro[ Room.

2. Suppression Chamber to DrwelI Vacuum Breakers Failed Closed The steam in the drywell wifl condense and drwe[I pressure will decrease. With vacuum breakers fated shut, pressure cannot equalize beteen the suppression pool and the dr,vell. The pressure in the drjwei[ may decrease such that the suppression chamber to drjwelldifferential pressure limit 1U psid) is eceeded.

Vent pipe buckling can occur if suppression chamber pressure is 10 psia greater than drjwel[ pressure. To prevent this situation, the vacuum breakers are operationa[ly checked periodically and 133% capacity is provided with ten vacuum breakers SD-04 Rev. £ Page 48 of 103

81. S2610001 Unit Two is operating at rated power.

Subsequently, a Div I pneumatic leak occurs causing drywell pressure to rise to 1 .9 psig. Which one of the following completes both statements below? The SBGT trains (1) running. The Div I pneumatics are required be isolated lAW (2) A. (1) are NOT (2) OEOP-0i -SEP-i 6, Drywell Systems Isolation B. (1) are NOT (2) OAO P-20 .0, Pneumatic (Air/Nitrogen) System Failures C. (1) are (2) 0 EOP-0 1-SEP-16, Drywell Systems Isolation D. (1) are (2) OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures Answer: C K/A: 261000 Standby Gas Treatment System A2 Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41 .5 I 45.6) 09 Plant air system failure RO/SRO Rating: 2.4/2.6 Tier 2 I Group 1 K/A match: A failure of plant air in the drywell would require that the drywell be vented due to increasing DW pressure. In this case, since the impact of the failure is DW pressure increasing> 1.7 psig, the specific impact on SBGT is that it can not be used to vent containment. The procedure to mitigate the consequences of the failure is the leak is required to be isolated using SEP-16. (NOTE: the conditions are NOT entry conditions for the procedures) Pedigree: New Objective: LOI-CLS-LP-046A Obj 13. Predict the effect that a loss or malfunction of the Pneumatic System would have on plant operation.

Reference:

None Cog Level: High Explanation: With drywell pressure >1.7 psig, 20P-10 cannot be used to vent the drywell. SBGT has automatically started at 1 .7 psig. With drywell pressure> 1.7 psig, PCCP directs the use of SEP-16 to isolate containment leaks, and SEP 16 contains the steps to isolate DIV I pneumatics.

Disttactor Analysis: Choice A: Part 1 is plausible because PCCP directs 20P-10 to control DW pressure <1.7 psig. Part 2 is plausible because it is correct, see explanation. Choice B: Part 1 is plausible because PCCP directs 20-10 to control DW pressure <1 .7 psig. Part 2 is plausible because AOP-20 directs the isolation of various pneumatic sources, but while executing the EOPs the EOPs take precedence. Choice C: Correct Answer, see explanation. Choice D: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because AOP-20 directs the isolation of various pneumatic sources, but while executing the EOPs the EOPs take precedence. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J. The applicant is required to select the appropriate procedure for venting the drywell and isolating pneumatics based on the given conditions. D RWyE LL SiSTE MS ISOLATION 0EOP-0I-SEP-f 6 Rev. PageGoI7 22 Drywell Pneumatics Isolation 22.1 Manpower Requireti

• 1 Reactor Operator 22.2 Special Eqtiipment None 22.3 Operator Actions NOTE If both divisions are isolated MSIV and SRV operation wil[ be limited to accumulators Q t IF Division I pneumatic leakage suspected, THEN:

a. Notify CRS RD

b. Close RNA-SV-5262 (Div I Non-lntrpt RNA) D RD

c. Close RNA-SV-5253 (Div I Ru N2 Supp To DW Isd Vlv) D RD 7 IF flhjicunn II nnIinvir Ikri c!icnctirI

SiNDBY GAS TREATMENT SYSTEM OPERATNG 20P-lO PROCEDURE Rev. 81 Pane 19 of 49 6.32 VentinO Containment Via SBGI DatefTEime Started Confirm the tollowing Initial Conditions are met:

• DryeIl pressure has risen to greater than 0.15 psig
• SBGT System is in STANDBY in accordance cith Section 6.1.1
• One of the tolIotng:

O Plant stack radiation monitor is in service and CAC-CS-5519 ICAC Purge Vent Isol Ovrdi is En OFF O E&C has sampled the drweil atmosphere and has determined that it is suitable for release

• Unit CR3 approval is obtained prior to venting

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tpig pre.s ort1rmert AIgn elecUcal pmer p. P-J1-OO-I4

                                       . 755 KW required drcç to 2.5 peig                  Terminate torus uçcays Is drcrm to 2.S peig                 Temiktote &i sçrs 4

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82. S262001 1 Unit One is operating at rated power.

Unit Two is in MODE 5 with UAT backfeed established. A main generator backup lockout occurs on Unit One. (REFERENCE PROVIDED) Which one of the following completes both statements below? All four diesels (1) automatically start. lAW Unit One Tech Spec 3.8.1, AC sources Operating, Condition E (2) required to be entered. A. (1) will (2) is B. (1) will (2) is NOT C. (1) wiIINOT (2) is D. (1) wilINOT (2) is NOT Answer: D K/A: 262001 AC. Electrical Distribution A2 Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41 .5 / 45.6) 09 Turbine/generator trip RO/SRO Rating: 2.9/3.0 Tier 2 I Group 1 K/A Match: This meets the K/A because the candidate is required to predict the status of the EDGs and determine whether the appropriate TS condition is entered. Pedigree: New Objective: LOl-CLS-LP-050 Obj 17 Given plant conditions, determine the required action(s) to be taken in accordance with Technical Specifications associated with the 230 KV Electrical Distribution System. (LOCT) (SRO Only)

Reference:

IS. 3.8.1 (blank out the LCO statement and Applicability) Cog Level: High Explanation: Part 1: DGs do not auto start on a generator backup lockout. Part 2: Condition E is not entered because a loss of two offsite circuits has not occurred. Only one offsite circuit is lost, the Unit One UAT.

Distractor Analysis: Choice A: Part 1 is plausible because all four DGs start on a generator primary lockout. Part 2 is plausible because if UAT was not in backfeed on Unit Two this would be the case. Choice B: Part 1 is plausible because all four DGs start on a generator primary lockout. Part 2 is plausible because it is correct, see explanation. Choice C: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because if UAT was not in backfeed on Unit Two this would be the case. Choice D: Correct Answer, see explanation. SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. Part two requires knowledge of the TS 3.8.1 bases to determine if Unit Two on UAT backfeed can qualify as an offsite source. In addition it requires the appropriate determination of the TS condition application. 3.2.6 Main Transformer, UAT, and Main Generator Protection

1. The Main Generator, MPT, and UAT are all: three protected by the Generator/Transformer Primary (8GGP) and Backup (86GB) Lockout Relays.

The Main Generator is provided additional protection by the Main Generator Differential Lockout Relays (86G).

a. Generator/Transformer Primary, Backup, and Main Generator Differential Relay trip actions are as follows:

(1)Main Generator Output breakers trip and lock out f2)Main turbine trips. f3)Main generator exciter field breaker trips and locks out. f4)UAT 4160 supply breakers to B, C and D buses trip and lock out f5)SAT teeders to C and D buses auto close. (6)Four diesel generators auto start for the Main Generator Differential Lockout or the Generator/Transformer Primary Lockout. They do not auto start for a Generator/Transformer Backup Lockout. SD-50 Rev. 23 Page 46 of 140

Offsite power a sAupplied to the 23(1 ki swutchyards from the transmission netork by eight bansmnssion lines. Fmm the 230 1W switchyards tao qualied e!ecthca[y and phystalty separated circuits provide AC power, through either a startup auxitiarg transformer (SATi or badifeethng via a unit auiIiarytransfomier(UATi, to 4.161W SOP buses. A single circuit path (master/slave breakers and i;nMerconnectng cabees:l ni eath SOP bus provides offtite power to its associated downstream 4.16 1W emergency bus. A detalied description of the ofisite power nelpwoirk and circuits to the onsite Class IE ernergeocy buses is found in the UPSAR, Sections 8.2 and 8.3 (Ref. 2J. A qualified offsite circuit consists of all breakers, transfomleffs, switches, interrupting devices, cabling and contids required to transmit power from either 230 1W bus (bus A or B) to the onsite Class 1 E emergency buses. The Unit I main generator provides the nomial source of power to 4.16 1W emergency buses El and E2 via its respective UAL The Unit 2 main generator provides the normal soi.wce of power to 4.16 kV emergency buses E3 and E4 via its respective UAT. In the event of a (continuedi Brunswick Unit I B 3.6.1-i Revision No. 31 I AC SourcesOperating B 3.6.1 BASES BACKGROUND unit trip, an automatic transfer from the normal circuit (main generator (continued) output via the UAT) to the respective unit SAT occurs resulting in the SAT supplying power to o 4.16kv emergency buses. As such, the Unit 1 SAT provides the preferred source of power to emergency buses El and E2 and the Unit 1 UAT (backfeed mode) is the alternate source of power to emergency buses El and E2. The Unit 2 SAT provides the preferred source of power to emergency buses E3 and E4 and the Unit 2 UAT (backfeed mode) is the alternate source of power to emergency buses E3 and E4. Each UAT can only be consdered a qualified oifsite source if it is capable of being powered from the 230 kV switchyard (Ref. 3).

AC Sources Operating 3.6.1 3.6 ELECTRICAL POWER SYSTEMS 3.6.1 AC Sourtes OperatinQ LCO 3.6.1 The foH.ing AC eIerical poer sources shall be OPERABLE

a. Two Unit I qualified circuits between the oftsite transmission network and the onsite Class IE AC Elestrical Power Distribution System:

b. Four diesel generators IDG5): and

c. Two Unit 2 qualified circuits between the offsite transmission network and the onsite Class IE AC Eledrical Power Distribution System.

APPLICABILITY: MODES 1, 2, and 3. ACTiO N S LCD 3.Q.4.b is not applicable to DOs. E. Two or more offsite circuits E.1 Declare required feature(s) t2 hours tram inoperable for rca sons other inoperable when the discol cry of than Condition B. redundant required Cor,dition E feature(s) are inoperable. cencunent with in operability of redundant required feature(s) AND E.2 Restore all hut one offsite 24 hours circuit to OPERABLE status, (centinued) Brunswick Unit I 3.6-5 Amendn,ent No. 264

83. S271000 I Unit Two is operating at rated power.

UA-48 (5-4) AOG System Bypass, has been alarming for 1 minute due to High-High off gas flow (REFERENCE PROVIDED) Which one of the following completes both statements below? AOG-XCV-142, Guard Bed Isolation Valve, (1) automatically close. ODCM 7.3.10, Gaseous Radwaste Treatment System, Condition A entry (2) required. A. (1) will (2) is B. (1) will (2) is NOT C. (1) will NOT (2) is D. (1) willNOT (2) is NOT Answer: C K/A: 271000 Offgas System A2 Ability to (a) predict the impacts of the following on the OFFGAS SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (C FR: 41 .5 / 45.6) 10 Offgas system high flow RO/SRO Rating: 3.1/3.3 Tier 2 I Group 2 K/A match: The applicant is required to predict the status of the AOG system (XCV-1 42) based on high-high offgas flow, and the required procedural actions (i.e ODCM). Pedigree: New Objective: LOl-CLS-LP-030 Obj 9 Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM and COLR, determine the requited action(s) to be taken in accordance with Technical Specifications, the TRM or ODCM associated with the Condenser Air Removal/Augmented Offgas System.

Reference:

ODCM 7.3.10 Cog Level: High

Explanation: Part 1: With XCV-1 42 remaining open the probable cause for UA-48 (5-4) is off gas flow high, all other conditions (besides circuit failure) for this alarm would cause a closure of the XCV-142. Part 2: with UA-48 (5-4) in alarm, HCV-102 is open. This would bypass the AOG portion of the Gaseous Radwaste Treatment System, leading to reduced hold up times and increased main stack rad levels. ODCM 7.3.10 requires AOG in operation so the comp measure is required. Distractor Analysis: Choice A: Part 1 is plausible because high-high cooler condenser condensate level would also cause UA-48 (5-4) to alarm, but it would also close the XCV-142. Part 2 is plausible because it is correct, see explanation. Choice B: Part 1 is plausible because high-high cooler condenser condensate level would also cause UA-48 (5-4) to alarm, but it would also close the XCV-142. Part 2 is plausible because with the opening of the HCV-102, an additional flowpath is provided around the charcoal adsorbers and the normal flowpath remains in service. Choice C: This is the Correct answer, see explanation. Choice D: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because with the opening of the HCV-102, an additional flowpath is provided around the charcoal adsorbers and the normal flowpath remains in service. SRO Basis: Conditions and limitations in the facility license. [10 CFR 55.43fb)(1). Requires the SRO applicant to have basis knowledge to determine whether ODCM compensatory actions are required based on plant status of the AOG system. Unit2 2AP P-UA-48 5-4 Page 1 ot 2 AOG SYSTEM BYPASS AUTO ACTIONS

1. AOG SYSTEM BYPASS VALVE, AOG-HCV-f02, opens CAUSES I High hydrogen Train A

2. High hydrogen Train B

3. High-high cooler condenser condensate level

4. High-high off-gas flow 5 Circuit failure Unit2 2APP-UA-48 1-4 Page 1 of 2 COOLER CNDSR DRN LEVEL HI AOG SYS BYP AUTO ACTIONS

1. GUARD BED ISOLATION VALVE. AOG-XC V-142, closes

2. AOG SYSTEM BYPASS VALVE AOG-HCV-102. opens

Ut2Lt 2 APP Wi 4 22 Page 3 oc 2 D:SCKAPSE 112 CONC t:ci AUtO ACT IONS

3. sotaticn to AOC System. lcoscs xcv-iIu. 1(7, 1(2, 2(0, and Sc after a 30 second ttme delayt

2. Cpeo ACG-HCV-102.

GASEOUS RADWASTE ThEAThENT SYSTEM B 7.110 B 7.3.10 GASEOUS RADWASTE TREATMENT SVSTEM BASES This requirement protdes reasonable assurance i at the releases of radioactije matedas in gaseous effluents will be kept as low as reasonab[9 achtevable. This specification iniplements the requirements of 10 CFR Pad SD.36a. General Design Criterion 60 off Appendix A to 10 CFR Pad 50, and The design obØctives giien in Section ll.D off Appendix Ito 10 CFR Part

50. The GASEOUS RADWASTE TREATMENT SYSTEM refers to the 30-minide offgas holdup line, stack filter house flifration, and the Augmented Off-Gas-Treatment System.

GASEOUS RAS%VASTE TREATMENT SYSTEM 7.3.10 7.3.10 GASEOUS RAOASTE TREATMENT SYSTEM 00CM S 73.10 TIe GASEOUS RASWASTE TREATMENT SYSTEM shat be ir, OPerabon APPL1CASIUTY Wfle.neier the Main Corterser Air Ejector IeVacuat3n) System is n operabcn COMPENSATORY MEASURES coNorntN REQUiRED COMPENSATORY COMPLETION MEASURE TiME A. GASEOUS RASaVASTE A.I Place GASEOUS 7 Eas TREATMENT SYSTEM not RAOWASTE in ooent3n. TREATM ENT SYSTEM In operation

5. 5.1 SubmIt a Speoai Report to 3] bays NRC that isenbfles tte NOTE requireti inoaenbie

Require-I Compensatory Measuree.1 stiai te etpmeni and the reasons rorthe competed If the Condibo.n iroperabi W ociresbee

                                   .         actors taken to restore Require-I Compensatory                tIe required inoperabie measure and associsie                 equipment to OPERAELE status, art a surrmary Comr4eboI Time not met.

descilatan of the onrremMe actions taken to preient recunence.

W4TER CfrIEMIST9Y GUIDEUNES Page AUACHM Pa Condenser Air Inleakage CONDENSER AIR INLEAKAGE: MODE I OIAtNOSTIC PARAMETER SAMPLE OLAGNOSTIC REQUIRED ACTIONS IF I FREQUENCY 1IUt REMARKS EXCEEDED Liir t pphes t.rni paet. At 1i] scfri the AOG 3yten YP5 Dcelop ni niplement n in-Ic Air In-leakage, ic AOi3HCV-D2liiflI open CDn1pcnsatry unes are reqL:ired byODCM 73 10

84. S295001 I Unit One is operating at 88% power with the following conditions:

Jet Pump Flow Loop A (B21-R61 JA) 29 Mlbs/hr Jet Pump Flow Loop B (B21-R611B) 33 Mlbs/hr Total Core Flow (UJCPWTCF) 62 Mlbs/hr Which one of the following completes both statements below lAW Tech Spec 3.4.1, Recirculation Loops Operating, and Bases? (consider each statement separately) The current Jet Pump Flow mismatch (1) If Jet Pump Flows are not matched within limits, then the loop with the (2) must be considered not in operation. A. (1) is within limits (2) lower flow B. (1) iswithin limits (2) higher flow C. (1) is not within limits (2) lower flow D. (1) is not within limits (2) higher flow Answer: C K/A: 295001 Partial or Complete Loss of Forced Core Flow Circulation AA2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : (CFR: 41.10 / 43.5 / 45.13) 05 Jet pump operability RO/SRO Rating: 3.1/3.4 Tier 1 / Group 1 K/A match: This question has the SRO candidate determine jet pump operability based on given core flow conditions. Pedigree: Bank NRC 10-1 Objective: CLSLP002*34 Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR determine the required action(s) to be taken in accordance with Technical Specifications associated with the Reactor Recirculation System. (SRO/STA only)

Reference:

None Cog Level: High

Explanation: Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Jet pump loop flow mismatch should be maintained within the following limits:

             -jet pump loop flows within 10% (maximum indicated difference 7.5 x106 lbs/hr) with total core flow less than 58 x106 lbs/hr
             - jet pump loop flows within 5% (maximum indicated difference 3.5 x106 lbs/hr) with total core flow greater than or equal to 58 x106 lbs/hr Distractor Analysis:

Choice A: Plausible because flow mismatch is within limits for lower reactor power level. Choice B: Plausible because flow mismatch is within limits for lower reactor power level and because the belief that the higher flow loop will experience excessive vibration could cause them to select the higher flow response Choice C: Correct Answer, see explanation. Choice D: Plausible because the belief that the higher flow loop will experience excessive vibration could cause them to select the higher flow response SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. The candidate is required to determine whether jet pump flow is within the limits and then use TS bases information to determine which loop is considered not in operation. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.41.1 NOTE Not required to be performed until 21 hours after both recirculation loops are in operation. VeAfy recirculation loop jet pump flow mismatch with 24 hours both recirculation loops in operaton:

a. 10% of rated core flow when operating at

                              <75% of rated core flow and

b. 5% of rated core flow when operating at 75% of rated core flow.

Redrculation Loops Operating B 3.41 EASES (cendnuedj SURVEILLANCE SR 3.41.1 REQUIREMENTS This SR ensures the rec.irc.ulation loops are within the allowable limits for mismatch. At low core flow Ci.e., c 75% of rated cere flow, the MCPR requirements provide larger margins to the fuel cladding integritc Safety Limit such that the potential adverse effed of early boiling transition during a LOCA is reduced. A. larger flow mismatch cart therefore, be all owed when core flow is c 75% of rated core flow. The resirculation loop jet pump flow. as used inthis Surveillance., isthe sumnnation of the flows from all of the jet pumps assodated with a single reditulation loop. The mismatch is measured in terms &the percent of rated core flow If the flow mismatch exceeds the specified limits, the loop with the lower flow is censidered not in operation. The SR is not required when both loops are not in operation sincethe mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours afterboth loops are in operation. The 24 hour rrequency is consistent with the Surveillance rrequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to deted off normal jet pump loop flows in a timely manner. REFERENCES 1. UrSAR. Section 5.4.1.3.

2. UrSAR, Chapter 15.

3. io CR 50.36(c2)(iij.

85. S2950131 Unit Two is performing 2EOP-01 -SBO, Station Blackout, as the blacked out unit.

The control room instrumentation for torus temperature is unavailable. Which one of the following completes both statements below? The CRS will direct torus temperature monitoring locally using (1) lAW 001-37.8, Primary Containment Control Procedure Basis Document, RCIC can operate without equipment damage with a suction temperature up to (2) A. (1) OEOP-01-FSG-08, Flex Instrumentation (2) 145°F B. (1) OEOP-0J-FSG-08, Flex Instrumentation (2) 190°F C. (1) OEOP-01 -SBO-01, Plant Monitoring (2) 145°F D. (1) OEOP-01 -SBO-01, Plant Monitoring (2) 190°F Answer: D K/A: 295013 HIGH SUPPRESSION POOL TEMPERATURE AA2 Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: (CFR: 41.10 /43.5/45.13) 02 Localized heating/stratification RO/SRO Rating: 3.2/3.5 Tier 1 / Group 2 K/A match: This question requires the applicant to determine that with the unavailability of average suppression pool temperature in the MCR during a SBO, individual temperature detectors can be read at the RSDP, by using TR-778 points 6 and 7 and then direct the appropriate procedure or attachment to accomplish this task. In addition, it requires the candidate to know the consequences of localized heating at the RCIC suction. Pedigree: New Objective: LOI-CLS-LP-004-A, Obj 19- Given plant conditions determine the required action(s) to be taken in accordance with Technical Specifications, associated with the Primary Containment System. (SRO/STA only) (LOCT)

Reference:

None Cog Level: Fundamental Explanation: Part 1: EOP-01-SBO-01 states, IF Control Room containment parameters NOT available, THEN: Monitor containment parameters at the RSDP, Attachment 3, Instruments Available With Loss of All AC Power and UPS Deenergized. Part 2: 001-37.8 states, RCIC can operate without damage up to 190°F based on EC 96336.

Distractor Analysis: Choice A: Part 1 is plausible because the use of FSG-08 is used if local monitoring is required and no indication available on the RSDP. Part 2 is plausible because while performing AOP-32, RCIC suction is swapped to the torus at 145°F. Choice B: Part 1 is plausible because the use of FSG-08 is used if local monitoring is required and no indication available on the RSDP. Part 2 is plausible because it is correct, see explanation. Choice C: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because while performing AOP-32, RCIC suction is swapped to the torus at 145°F Choice D: Correct answer, see explanation. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J. This question requires the SRO candidate to select the appropriate procedure to use for suppression pool temperature monitoring when instrumentation is lost in the MCR during a SBO.

PLANT MON ITORING OEOP-Llt-SBO-O1 Rev. 0 Page 5 of 1 2.3 Monitoring Actions C continued NOTE [B21-u-Ro4B and C32-PR-RO9 will NOT be available on the RTGB C I IF RPV level will be mon[tored on B2i-Ll-R6O4BX (N026B, at RS DP, THEN: C RO Place B21-CS-3345 (Reactor Water Level Ncrmal!Locai Swch) to LOCAL C AO

3. IF AT ANY TIME local nonitorini requited.

THEN obtain as needed per EOP-0t-FSG-O C RO

• RPV Level C AO
• RPV Pressure C AD
• Drywell Pressure C AO
• Torus Pressure and Level C AO
• Containmentlemperature C AO

4. IF Control Room containment parameters NOT available, THEN:

a. Monitor containment parameters at the RSDP, Attachment 3, Instruments Available With Loss of All AC Power and UPS Deenergized C AO

b. Periodically notify Control Room of values and trend C AO

P RI NtARY CONTAINMENT CONTROL DCI -37. a PROCEDURE BASIS DOCUMENT Rev DOS Page 16 of 57 5.9 Steps Tif-iS and VT-16 4 I BEFORE 1 orus water I Ensure RCIC aligned to GST fT16 The nornial aflgnment and preferred suction source for RCIC is normally tile CST. This alignment may be altered during SBO conditions, Extended Loss of AC Power (ELAP) conditions or for operational considerations. RC[C is the preferred injection system tor ELP conditions and may he preferred during SBC conditions if lower iniection flow and longer system operation time is desired to limit battery depletion. in addition, RCiC may be in service for RPV pressure control. As stated in Caution 4. operation of RCIC with suction temperature above 190P may result in equipment damage since the lube oil and control oil is cooled by the water being pumped Very high lube oil temperatures can result in loss of lubricating qualities in the oil and cause damage to the bearings. RCLC can operate without damage up to 19GF based on EC 9633& However, if SRVs are being used for RPV pressure control, tows water temperature may eventually become incompatible with long term RCIC operation. Therefore RCIC suction is aligned to the CST before the tows water temperature reaches 19OF.

fliL 4. APP LIA-21 2-5 Page 1 of 2 DRYWELUSUPP POOL HIGH TEMP NOTE: This procedure is only to be used in conjunction with 6AOP-32. Plant Shutdown From Outside Control Room. AUTO ACTIONS 1 Low RBCCW pressure starts idle RBCON pump. CAU SE 1 Insufficient number of drywell coolers in service.

2. Dr,ell purge exhaust fans not operating properly.

3 Improper operation of RBCCW System.

4. Drywell equipment drain heat exchangers not working properly.

5 Excessive drgweil equipment leakage. 6 Circuit malFunction. 0 BSERVAT IONS 1 Verify proper operation of drelI coolers. I Verify proper operation of drpell purge exhaust fans. 3 Verify proper operation of RBCCW System. 4 Verify proper operation of dnpell equipment drain heat exchangers

5. Dryvell pressure.

ACTLO NS 1 Place RHR Loop B in suppression pool cooling per OAOP-32.

2. II suppression pool average water temperature exceeds 145:F, then line up the RCIC suction to the CST.

3. Identify and isolate source of drweII leakage.

4. If desired, UA-29 2-5 may be cleared by performing 001-63, Section 6.2, Step 6, SV100 Alarm Acknowledge Function, at 2-CAC-TR-778.

5. If a circuit malfunction is suspected ensure that a WR is prepared.

DEVICEJSETPOINTS CAC-TE-778-1, 3, 4 (Drywell) 200:F CAC-TE-778-5, 6, 7 (Suppression Pool) 1 4StF 8F 2APP-UA-29 Rev.6 Page l3of 16

86. S2950151 Unit One is performing the ATWS Procedure with the following conditions:

A-05 (2-6) Reactor Vess Lo Level Trip, is illuminated A-06 (1-6) Reactor Vess Lo Lo Water Level Sys A, is NOT illuminated A-06 (2-6) Reactor Vess Lo Lo Water Level Sys B, is NOT illuminated MSIV5 are closed Reactor pressure peaked at 1141 psig and is now being controlled 800-1 000 psig. Torus water temperature is 105°F and rising Reactor power is 25% lAW 001-37.5, ATWS Procedure Basis Document, which one of the following identifies the action that will have the highest priority? A. SLC initiation. B. Inhibiting ADS. C. Trip both Reactor Recirc Pumps D. Termination and prevention RPV injection. Answer: D K/A: 295015 Incomplete SCRAM G2.4.31 Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41 .10 I 45.3) ROISRO Rating: 4.2/4.1 Tier 1 / Group 2 K/A match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL. This question requires knowledge of the annunciator response procedure status (where level is at) to determine the appropriate course of action while executing the EOP. Also for prioritization of the action to direct. Pedigree: New Objective: 300E-1 7e Given plant conditions and the Anticipated Transient Without Scram Procedure, determine the following: Priority of execution given to each leg of the procedure.

Reference:

None Cog Level: Hi Explanation: With Reactor vessel to lo water level not illuminated, RPV water level is >90. Reactor power is also >23% with the RR pumps tripped. lAW RC/Q-8 and RC/L-2, terminate and prevent is immediately required in order to prevent THI.

Distractor Analysis: Choice A: Plausible because SLC initiation is required with >2% power and rising torus temperatures in order to not exceed the HCTL. However, it is not the first step required because power is >23%. Choice B: Plausible because Inhibiting ADS is required, However, it is not the first step required because power is >23%. Choice C: Plausible because tripping the recirc pumps would be done prior to terminate and prevent but with RPV pressure peaking at> 1138 ARI would have tripped the pumps, Level is not at LL2 yet according to the alarms. Choice D: Correct Answer, see explanation. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J. This SRO applicant is required to direct the appropriate actions out of the ATWS EOP based on plant conditions. The action is contained in a note in the EOP and not a general strategy of the EOP. jTCDTt r RIy :ec] rcm

          !3S     a                   E1   ted fret,. :LLTi25a11; Lee1    riit                     r.t tZi-1 O3EL LANr EZCT

1. pb iii a T:h pc

a. .Z.:4

2. cac. , Lc 22 1APP-A-06 Rev 69 Page 11 o185

KTWO PROCEDURE RC UVCLJMENT 03l37.5 Re.. 14 Page 13 of 5.4 Step RCJL-2 o , 1 kOtfflhIp.I*% I;.! n

                       £&ltb.T ,    ii RI,             5rnctns UlrN            l4         rrrhi r I       kji, Ii.. iC ii. .- 4,W, I,.i I rwrl

• 1 .Olfr,,_1frFs *.,1 I *-J I .

11 If reactor power is greater than 234 wth both reactor red rtulacon plrtps tripped and RPV level above go inches, RPV level needs to be prumItiy reduced blova the feedwater nozzles, to toid thermal hytirauhc instabilities. TI-is is accomplished by terminallcn and prevetlicn of injection systems. frc*i, identified systems, parttcu[aiyfeedwater, within 120 seconds. To prevent or mtiigate the consequences of any la-ge reglarneubcnfiue oscillations induced by neuonicthermaI-hydraulc instabiIites, RPt esel is initially lcwered su-tficiently below the elevaton of the feedwater spaiger nozzles. This pisces ti-is feethvaler spagers in the steam spats prcnditg effective hsatng of the relatively cold feedwater and cumin alinGihe pOtential for hgh core tAct subcccling For ccndticns that are susceptble to oscillations, initiation and growth of oscillations is principally dependent upon sbcccling at the core inlet the greater the subcooling, the more likely oscillations will commence and increase in mniIcjde. If reactor power is at or below the APRM downscc zip setpoint 2%, is highly unlikely that the core bulk b&ng bcundy would be below that which pruvtdes suitable stability margin For operaton at high powers and low flows. A minimum bdling bouldary ci- 4 ft above the bottom of active fuel has been shcwn to be effective as a stabitty control because a relatively long two-phase celumn is required to develop a coupled neutronicl thermal-hydraulic instablity. Rrthermore, floeddensity vaiations would be limted with reactor power this [ov since the core has a relatively ow average uaid content.

ATV PROCEDURE fiAc[ DOCUMENT GCl-3T5 Peg. 14 Page 68 of 55 5.33 Step RCSQ-6 throuah RCIQ-8 BE.PORE f (c - ¶rn N075 r J LmvN p.cfcn J torus If reactor puer is below fl, the operator is directed to inject boron beftie water temperature reaches IDtF. This allows suffiu:snt time For Rot Shutdown Boron Weight HSBWl cf hucun to be injected As lung as the core rernahs sub merged ithe preferred method of cure cooling:I. fuel ntegnty and PPS inte;rftyn nut directly chalenged even under fallure-to-scram condiic4ls. A scram faflurecoupled with an MSS Latcn hcwever, resulte in rapid heatup of the tows due to the steam dischrged from the PPS rita SRVs. The ohallenjte to ocntainmwrt thus becomes the lofting factor which defines the reui iremwrt fcc boron injection. f tunis temperature and PPS pressure cannot be maintan cit b.eluwi the Heat Capacity Temperature Limt lHCTL. rapid epressuh:at un of the PPS will be required. To avoid depressuhoing the PPS isith the reactor at power, it is desirable to shut down the reactor pdor to reaching HCTL. thi. mninting the quantity of heat rejected to the tows. The Bcrcn lnection initiation Temperature IS lIT) is defined so as to achiese tlis when practeable

PCE U?E OCJ&1ENT Q3 -37. Re. 14 Page 22 5 RCIL-1O

                 *0

• I

__j___ - TIbU .b Thllnft, law P101101 IlS1ltja0Ofl [Yl 10 tfl* fl ,.* src4s fl1%l VOt I B.ed on raact pewer bn -oe 2%, Step RCL-2 initiay Iowred RPV teeI, to the feedwater sparger. by terminathg and preawrtimg njectcn from identtftd stEcns. fall of the condif one in TateQ-2 are met, Step RCL4 will lower RPV Iewl further ma supprees reactor poww. When any ccnt-micn in Table Q-2 is no anger met the operator is directe1 to confnue to subsequent steps which will establish a nev RPd Iseel band.

kV.vc PDCEDURE S&DID cooJr,IENT C2I-35 Rev. 14 PageD3 of 55 SW Step RCt-lt through RC.t-1O 4continue-d Termn.xing and preventing reoison frcrn:

• Condensate and Fe&t4ater is addressed in 1OP-32 i2OP-321, Condensate And Feerwiazer System Coeratin Procedure. and coves terrninatng ard prevenong heccion by either tripping both ReacicrFeed PuripsiRFPsi orby rIn;cneRFP
• Core Spray is acooniplished by tripping the associxed Core Spray loops operabrg pump.
• HPCI is addressed in IOP-tP 120P-tPi, High Pressure Coolant Injection Systeni Operating Proc.ed Lire, and covers terrrinsing and preventing irjeorion when HPCI is either operating or not operating
• RHR is accomplished by tipping the associated RHR loops operating pump(si.

The Boron Injection Initiation Temperature (SIlT is a function of reactor power and is the torus temperature before whoh boron ir eotion must be initiated if a reactor depressunoabon. due to esceeding the Heat Capacity Terperature Lint (HCTLj, is to be precluded. This temperature is hOP. The combination of high reactor power :aboee the APRkt downsoa4e thpj, high torus temperature jabase SI lT, and an open SR V or high drywell pressure (above the scram setpoirt, are symptomatic of heat berg rejected to the torus at a raze in caress of that ehioh can be removed by the torus oootng system. Unless mitigated, these oonditiora ultimately result in loss of NPSR ía ECCS purnips taking suction on the tows, oortainment overpressuization, and (ultirnatelyl loss of Primary Contain cent integrity, which in tim could lead to a loss of adequ ate core oooling arid uncontrolled release of radhoactiufty to the environment. The conditions listed in Table 0-2, corrtned with the in ability to shut bonn the reactor through corerd rod insertion, dictate a requirement to prorrptiy further reduce reactor poner in order to preserve Primary Containment integrity since, as long as these conditore exist tows heatup cli continue. Since RPV level is only alowed to drop to TAF before injection is restarted. if RPV level is already bebon TAF. then the &ective of the step has been accomplished. Further lowering of RPV level is not necessary, and the steps which deliberately lover RPV level re bypassed

87. S295021 I Unit One is in MODE 4, when a loss of SDC occurs due to RCS leakage.

Cul UNPLANNED os of RPVi ertciyfr 15 mhnutes orcnger I I I 4 I 5 I cu1.1 UNPLANNED loss of reactor cootant resuts in RPV water level less than a required lower limit for 15 mEn. (Note 1) Which one of the following completes both of the statements below? The minimum required RPV water level to support natural circulation is (1) lAW OPEP-02.2.1, Emergency Action Level Technical Bases, the Unusual Event required lower limit is defined as RPV water level less than (2) A. (1) 200 inches (2) 105 inches B. (1) 200 inches (2) 166 inches C. (1) 254 inches (2) 105 inches D. (1) 254 inches (2) 166 inches Answer: B K/A: 295021 Loss of Shutdown Cooling AA2 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING : (CFR: 41.10 / 43.5 I 45.13) 03 Reactor water level RO/SRO Rating: 3.5/3.5 Tier 1 / Group 1 K/A Match: This meets the K/A because it is testing the ability of the applicant to determine the water level designated for a UE with a loss of SDC. Pedigree: Modified from 10-1 Objective: LOl-CLS-LP-301-B Obj. 6 Define the relationship between fission product barrier loss/potential loss and each of the four emergency classifications.

Reference:

None Cog Level: Fundamental Explanation: Part 1: lAW OAOP-15.0, the minimum required level for natural circulation is 200 inches. Part 2: lAW OPEP-02.2.1, the RPV water level lower limit is the low end of any established band.

Distractor Analysis: Choice A: Parti is plausible because this is correct, see explanation. Part 2 is plausible because this would be the lower limit if no band was established for SDC in MODE 4. Choice B: Correct Answer, see explanation Choice C: Part 1 is plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC. Choice D: Correct Answer, see explanation. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]. The applicant is required to have bases knowledge of the EAL procedure to select the appropriate method of implementing the UE criteria for a lowering RPV level in MODE 4. EAL determination is an SRO only task. Question from NRC 10.1 exam: While in Mode 4 a loss of Shutdown Cooling (SDC) occurs. Which one 01 the following completes both statements? The minimum required Reactor Water Level to support Natural Circulation is (1) inches. An Alert declaration is first required after an unplanned RPV pressure increase greater than (2) psig due to a loss of RCS coo[ing. A. (1) 200 (2) 135 B (1) 200 (2) 10 C. (1) 254 (2) 135 D. (1) 254 (2) 10

LOSS OF SHUTDOWN COOLING OAOP-1&0 Rev.31 Page 7 of 25 4.2 Supplementary Actions (continued

2. IF forced circulation has been lost.

AND natural circuiatthn has NOT been established. THEN ensure reactor vessel water level is being maintained between 200 inches and 220 inches as read on B21-Ll-R6OSA(B) (RPV Water Level), OR as directed by the Unit CRS based on plant conditions until forced circulation is restored C ATTACHMENT 1 Page 45 of 204 EAL Bases jh is IC addresses the inability to restore and maintain water leei to arequed minimum v& -. (orthe lower limit ofa level band). or a loss ofthe abilityto monitor RPV level concurrentwith indications of coolant leakage. Either ofthese conditions is considered to be a potential degradation ofthelevel of safety of the plant Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An U N PLANNED event that results in water level decreasing below a procedurally required limitwarrants the declaration of an Unusual Event due to the reduced water inventorythat is availableto keep the core covered. This EAL recognizes thatthe minimum required RPV level can change several times during the course ofa refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified forthe current plant conditions. cannot be maintained for 15 minutes or longer The minimum level istypicallyspecified in the applicable operating procedure but may be specified in another controlling document. The 15-minute threshold duration allis sufflcienttimefor prompt operator actionsto restore and maintain the expected water level This criterion excludes transient conditions causing a brief lowering of water level. Continued loss ofRCS inventory may resultin escalation to theAlertemergencyclassification level via either IC CAl or CA3. BNP Basis Reference(s):

1. OEOP-O1-NL EOP-SAMG NUMERICAL LIMITS AND VALUES. Table IE

2. SD-O1 .2 Reactor Vessel Instrumentation Figure 01.2-1 Reactor Water Level Instrument Ranges

3. l(2)APP A7 2-2 (Reactor Water Level HiiLow)

4. OGP-O5 Cold Shutdown to Refueling (Head Unbolted) step 5.1.14

5. NEI 99-01 CU1

88. 5295023 1 Which one of the following completes both statements below?

lAW Tech Spec 3.9.6, Reactor Pressure Vessel (RPV) Water Level, the minimum water level over the top of irradiated fuel assemblies seated within the RPV during movement of irradiated fuel assemblies in the RPV is (1) The Tech Spec bases for the minimum water level is to provide for (2) during a fuel handling accident. A. (1) l9feetll inches (2) iodine retention B. (1) l9feetll inches (2) shielding of radioactive decay particles C. (1) 23feet (2) iodine retention D. (1) 23feet (2) shielding of radioactive decay particles Answer: C K/A: 295023 Refueling Accidents G2.2.25Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7 / 43.2) RO/SRO Rating: 3.2/4.2 Tier 1 / Group 1 K/A Match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL. This meets the K/A because it is testing the TS bases for the RPV water level during movement of irradiated fuel assemblies in the RPV. Pedigree: New Objective: LOI-CLS-LP-200-B Obj 12. Identify conditions and limitations in the facility license. (SRO/STA only)

Reference:

None Cog Level: Fundamental Explanation: Part 1: LCO 3.9.6 states, RPV water level shall be = 23 ft above the top of irradiated fuel assemblies seated within the RPV. Part 2: The minimum water level of 23 ft allows a decontamination factor of 200 to be used in the accident analysis for iodine. This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the damaged fuel assembly rods is retained by the water.

Distractor Analysis: Choice A: Part 1 is plausible because lAW TS 3.7.7 the spent fuel storage pool water level shall be = 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. Part 2 is plausible because it is correct, see explanation. Choice B: Part 1 is plausible because lAW TS 3.7.7 the spent fuel storage pool water level shall be = 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. Part 2 is plausible because the water does provide shielding from radioactive decay particles, but that is not the TS bases for the minimum water level. Choice C: Correct Answer, see explanation Choice D: Part 2 is plausible because it is correct, see explanation. Part 2 is plausible because the water does provide shielding from radioactive decay particles, but that is not the TS bases for the minimum water level. SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)J. Requires the SRO applicant to know the IS bases for RPV water level. RPV Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure Vessel (RPV) Water Level LCD 3.9.6 RPV water level shall be 23 ft above the top of irradiated fuel assemblies seated within the RPV. APP LICABILITY: During movement of irradiated fuel assemblies within the RPV. During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPv. Spent Fuel Storage Pool Water Level 3.7.7 3.7 PLANT SYSTEMS 3,7.7 Spent Fuel Storage Pool Water Level LCO 3.7.7 The spent fuel storage pool water level shall be 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool.

B 39 REFUELING OPERAI1ONS B 39.6 Reactor Pressure vessel (RPV Water Level BACKGROUND The mcvenientoffu&l assemblies orihardling of control to swithin the RPvwequiires a minimum water level of 23ft alcie the top ofrradiated fuel assemblies seated witfun the RPV. During refueling, this niaintains a sufcient water level in the reactor esseL Sufficient water is necessarj to retain iodine ssian product activity in the ater in the event of a fuel handling accident (Refo. I and 2i. Siuftcilent iodine activity would be retained to limit offsfte doses from the accident to well beici the ID CFR 5G.67 exposure guidelines (Ref. 3. APPUCABLE During movement of fuel assenthhies or handling of control rods, the SAFETY ANALYSES water level in the RPV is an initial condibon design parameter in the ana[sis of a fuel handling accident in containment postulated by Regulatory Guide 1.183 (Ref. Il. A rnnimurin water level of 23 ft abows a decontamination factor of 2DD to be used in the accident analysis for iodine (Ref. 1). This relates to the assumption that 95% of the total iodine released from the pellet to cladding gap of all the damaged fuel assemily rods is retained by the water. Analysis of the fuel handling accident inside containment is deschibed in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours prior to fuel handling, the ana[sis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequatel captured by the water and that offsite doses are ntaintained welt below the allowable limits of Reference 3. RPV water level satisiles Criterion 2 of 1 CFR EO.3Ec(2(ii:i (Ref. 41. (coritiri ued

89. S295026 I An event on Unit One has resulted in the following plant conditions:

Reactor pressure: 1000 psig Reactor Water Level 120 inches Control Rod position Unknown APRMs Downscale Drywell pressure: 3 psig Torus pressure: 2 psig Torus water temp: 152°F Torus water level: -36 inches (REFERENCE PROVIDED) Which one of the following identifies the required actions for reactor pressure control? A. Exit the RC/P flowpath of ATWS, and go to OEOP-01 -EDP, Emergency Depressurization. B. Exit the RC/P flowpath of RVCP, and go to OEOP-01-EDP, Emergency Depressurization. C. Remain in the RC/P flowpath of ATWS, and exceed 100°F/hr cooldown rate if necessary. D. Remain in the RC/P flowpath of RVCP, and exceed 100°F/hr cooldown rate if necessary. Answer: C K/A: 295026 Suppression Pool High Water Temperature G2.1 .23Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10/43.5/45.2/45.6) RO/SRO Rating: 4.3/4.4 Tier 1 / Group 1 K/A match: The applicant is required to determine which procedure to implement based on HCTL Pedigree: New Objective: CLS-LP-300L, Obj. 5a Given the Primary Containment Control Procedure, determine the appropriate operator actions if any of the following limits are approached or exceeded: Heat Capacity Temperature Limit

Reference:

OEOP-01-UG, Attachment 7, Heat Capacity Temperature Limit Cog Level: Hi Explanation: With rods at an unknown position, an ATWS has occurred. Since HCTL is close to the unsafe region (but not violating it), exceeding the cooldown rate in the RC/P flowpath of ATWS is warranted.

Distractor Analysis: Choice A: Plausible because a novice applicant may misinterpret the graph and believe HCTL is in the unsafe region, and an ED is warranted. Choice B: Plausible because a novice applicant may believe with APRMs downscale an ATWS has not occurred, and may misinterpret the graph believing an ED is warranted. Choice C: Correct answer, see explanation. Choice D: Plausible because a novice applicant may believe with APRMs downscale an ATWS has not occurred. Exceeding the cooldown rate is correct because of the operating point on the HCTL graph. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5) SRO is required to select the appropriate EOP action based on HCTL. UE GUIOE OEOP-G1-UG Rev. 35 Pago12 Pe 1 of 1 Heat Capacity TeoperaLi;r Limit NTh ivrr. U-2l0 HHH+ UNSAFE ABOVE

                                              .H SELECTED LINE 200                Li             I D    190 ou 1JZ 1 Lu   1704.               :                                   --I02SFT Li   160,---                                                     (-Il 25FT I                          -
                                                                            %(-12 5OFT irn ..zr              :

r (-1325FT 140 - 44% SAFE BELOV 130 SELECTED LII 120 I 5S0FT 110 -t 100 s

10) 300 500 700 900 1,100 a 200 400 600 000 1,000 RPV PRESSURE (PSIG)

( START) While in this procedure: IF ThEN RPV Iee1 CANNOT be deierrhed Exit ftCLAtl. ft.P flcaptls .3rd go to ECP-C1-RXFP Emecgency deeeaItzaton IsQ has been reqt3red Proceed to Ext CP 1lpII and go to Rextor Is OlD Wtrdut boro1 wderj cc4,dItIora (te Q-r I Terminate bcra9 I9jecb3rli reqIlIred oy cdh E0Ps

2. ExlttNn IIwrht 3rd toEC-Dl-RVCP

C

F THEN I nir pi Cntru rcirn 1rur: 1 cerze a ctpr HflLC4T e Ir bitj! Mhita Pprur ea iC1L ftCf, 1I 1f rJ n O5 t MJy1 CLDC qwIIz& prsalird Ms DP-:

• De1e.t r I r,mu I I&i1rtl If ECP-21-EEP-1I r.o, U.

U1r cnI 33 IZrR IJ! IrdcJflcr 2t a !rajr II9 tr3 A rrui EUX . 1E1]t1Izlrg curE cco wt1 I CLDE

                                                . Eiperrg FJ.

Ill rr1rIn1z I.

90. S295035 1 Unit Two is operating at rated power. PCCP has been entered due to high torus water temperature with the following plant conditions:

UA-12 (3-3) Rx Bldg Duff Press High/Low, is in alarm UA-05 (6-10) Rx Bldg Isolated, is in alarm. Reactor Building Pressure (indication on the left) Which one of the following completes both statements below? Reactor Building pressure is (1) The CRS will direct Reactor Building HVAC restarted lAW (2) A. (1) positive (2) 20P-37.1, Reactor Building Heating and Ventilation System Operating Procedure B. (1) positive (2) OEOP-01 -SEP-04, Reactor Building HVAC Restart Procedure C. (1) negative (2) 20P-37.1, Reactor Building Heating and Ventilation System Operating Procedure D. (1) negative (2) OEOP-01 -SEP-04, Reactor Building HVAC Restart Procedure Answer: C K/A: 295035 Secondary Containment High Differential Pressure EA2 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: (CFR: 41.8 to 41 .10) 01 Secondary containment pressure RO/SRO Rating: 3.8/3.9 Tier 1 / Group 2 K/A match: The applicant is required to interpret secondary containment pressure and select the appropriate procedure based on this interpretation. Pedigree: New Objective: LOI-CLS-LP-300-M Obj 11 Given plant conditions involving Reactor Building HVAC system isolation and the Secondary Containment Control Procedure, determine if the Reactor Building HVAC system should be restarted.

Reference:

None Cog Level: Hi

Explanation: Part 1: With indications at upscale > +0.5 inches of h20, reactor building pressure is negative. Part 2: lAW with UA-5 (6-10) and UA-1 2 (3-3) RBHVAC is restarted 20P37.1. Distractor Analysis: Choice A: Part 1 is plausible because the reading is +.05 inches of h20 and upscale high, a novice candidate could mistake this for a positive pressure indication. Part 2: is plausible because it is correct, see explanation. Choice B: Part 1 is plausible because the reading is +0.5 inches of h20 and upscale high, a novice candidate could mistake this for a positive pressure indication. Part 2: is plausible because RBHVAC is restarted using SEP-04, when in SCCP when LL2 and high drywell pressure needs to be defeated. In addition the title is rbhvac restart procedure, and it is a SEP. Choice C: Correct Answer, see explanation. Choice D: Part 1 is plausible because it is correct, see explanation. Part 2: is plausible because RBHVAC is restarted using SEP-04, when in SCCP when LL2 and high drywell pressure needs to be defeated. In addition the title is rbhvac restart procedure, and it is a SEP. SRQ Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [JO CFR 55.43(b)(5)] The SRO applicant is required to interpret secondary containment pressure and select the appropriate procedure based on this interpretation. At h-ID 35

                                                                                ;g 1 ci 1.

Th 3L0 ti: .Z5S LOW EP.eactcr 3uilding 2frnti. iIL:w AUTO ACTEONS I - ct:r Build+/-.nq uot1y .r..i hic CAUSE iqh cc lcw di! rnc..1 pstr btwr. ch Reacc:r 1dn ni accsherc t:esslir. Circuit mifun:ti:n. O33ERVATDQNS I. Rectcc Buildinq Static Pressure Indicatcr 2VAP1297, cm RT9

us.

AC II CITS Ii sec:ndary :cnt xtmet lntegrity is r u2.red and dii+/-erenctal pressure zs lcw, enter CEOPD3-SCCP Secondary Ccnraiment Conrr:l, execute :cncurrencly w:th thIs procedure. C. Inform E&R Chemistry Reactcr Building Ventilation is not n servIce.

3. Vertfy that the valve lineup is correct per 20P571, React:r Building Heattng and Ventilation System.

4. Start up the svsten per Secti:n 31 ci S. Ii a circuit malfunction :s susected, ensure that a WP.WC Is submtted.

1Ifiil 1 dP (I.I ii 1 ri II P .1 i l&

      ,1I)           PH I   U            .1 0            rj   uJ i:         *1) pr N         H t,r                P      ii Ii.)ii        PP                     H 0 Wi         II I

M Dl II I 4) Ii ni ,IIl; P I ii p = 4 1111.11 4, LI li) Ci p 1-14) 11)11 H n P U In p.a ii 4)41 4I u 4*14 4).) -ii tp II

      .11           .ir          )

1 .1 41 p 4) Ci C1 ill p p p -c

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      .iI            pP             iJ                         
      .1-1    4      ii  P P ii 11                  1.

1141 II 4) l.)14 14 LIP. 4,li-IP - 13 14. P til Ii t IL 2 I ii Pd 41

           .1    P   4i  II i-I  .i  ii UI       fl.C    11PII 1  p   p    ii  p p         p PP           *iP*1                  *U1    4)2 410P444,41fl4)

II,II

91. 52950371 Unit Two is in an ATWS executing RXFP, with the following plant conditions:

Injection to the RPV has been terminated and prevented The Minimum Number of SRVs Required for Emergency Depressurization are open. Table P4 Minimum Steam Cooling Pressure Open SRVs Pressure fpsig) 7or more 120 6 145 5 175 4 220 3 300 2 455 I 915 lAW RXFP, which one of the following completes the statement below? The CRS should direct injection to the RPV when EITHER: (1) SRV remains open OR when reactor pressure lowers below the Minimum Steam Cooling Pressure of (2) A. (1) NO (2) 175 psig B. (1) NO (2) 455 psig C. (1) ONLYone (2) 175 psig D. (1) ONLYone (2) 455 psig Answer: A K/A: 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown G2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7/43.5 / 45.12) RO/SRO Rating: 4.0/4.6 Tier 1 / Group 1

K/A Match: This meets the K/A because it is testing the required steam cooling pressure for adequate core cooling. Pedigree: New Objective: LOl-CLS-LP-300-F Obj.3 Given the Reactor Flooding Procedure, which steps have been completed and plant parameter values, determine the required operator actions.

Reference:

None Cog Level: High Explanation: Part 1: lAW RXFP-9/1O, injection is reestablished when either no SRVs are open or Part 2: Reactor is below the MSCP, which in this case for 5 SRVs (MNSRED) is 175 psig. Distractor Analysis: Choice A: Correct answer, see explanation. Choice B: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because this is the minimum steam cooling pressure for having the Minimum Number of SRVs Required for Decay Heat Removal open. Choice C: Part 1 is plausible because once injection is re-established, RXFP-1O directs injection to continue until at least one SRV is open. In addition, with no SRVs open or below the MSCP adequate core cooling could possibly not exist, so a candidate might assume we would never wait for that condition to reestablish injection. Part 2 is plausible, because it is correct, see explanation. Choice D: Part 1 is plausible because once injection is re-established, RXFP-1O directs injection to continue until at least one SRV is open. In addition, with no SRVs open or below the MSCP adequate core cooling could possibly not exist, so a candidate might assume we would never wait for that condition to reestablish injection. Part 2 is plausible because this is the minimum steam cooling pressure for having the Minimum Number of SRVs Required for Decay Heat Removal open. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]. This requires the SRO to have knowledge of more than the overall sequence of events in the EOPs or mitigating strategy. Since it requires the SRO to know the MNSRED and how to implement Table-P3.

USERS CJIIDE GEOP-01-UG Rev 067 Page ii of 156 3.0 DEFINITIONS (continued)

37. Maximum Subcritical Banked Withdrawal Position: The lowest control rod position to which all controls rods may be withdrawn in bank and the reactor wit nonetheless remain shutdown under all conditions. This position is utilized to assure the reactor will remain shutdown irrespective of reactor water temperature

38. Minimum Core Steam Flow: The Lowest core steam flow which is sufficient to preclude any clad temperature from exceeding 1500T even if the reactor core is not completely covered.

39 Minimum Debris Retention Injection Rate: The Lowest RPV injection rate at which it is expected that core debris will be retained in the RPV when RPV leveL cannot be determined to be above the bottom of active mel. (Attachment 17)

40. Minimum Indicated Level: The highest RPV level instrument indication which results from off-calibration instrument run temperature conditions when RPV level is actually at the elevation of the instrument variable leg tap.

41. Minimum Number of SRVS Required for Decay Heat Removal:

The least number of SRVs (2) which will renrnve all decay heat from the core at a pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow.

42. Minimum Number of SRVS Required for Emergency Depressurization: The number of SRVs (5) which correspond to a minimum steam cooling pressure sufficientty low that the ECCS with the lowest head will be capable of making up the SRV steam flow.

I

                     /

(Either WHEN I 3s0P

• RPV pressure is below I

L MSCP (Table R3i

                     \         THEN continue
                                            /PJFP-9 IL commence and slowly raise RPV injection usmg Table L-3 systems until:                                               -  5

• At least one SRV is OPEN AND
• RPV pressure above MSCP (Table P-3) but as low as practicable

92. S295038 I Unit Two is executing RRCP with the following plant conditions:

Main Stack Rad Monitor, D12-RM-23S, is reading 2.3E+08 .jCl/sec Turbine Building Vent Rad Monitor, D12-RM-23, is reading 2.5E+07 pCI/sec Real-time dose assessment using actual meteorology indicates 0.92 Rem TEDE and 5.1 Rem thyroid CDE at the site boundary (REFERENCE PROVIDED) Which one of the following completes both statements below? lAW RRCP, Unit Two turbine ventilation is required to be in the (1) ventilation lineup. The highest EAL classification for this event is (2) A. (1) recirculation (2) Site Area Emergency B. (1) recirculation (2) General Emergency C. (1) once through (2) Site Area Emergency D. (1) once through (2) General Emergency Answer: B K/A: 295038 High Off-Site Release Rate EA2 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE : (CFR: 41.10 / 43.5 / 45.13) 01 Off-site RO/SRO Rating: 3.3/4.3 Tier 1 / Group 1 K/A match: The candidate is required to compare the given radiation release values (including site boundary), and compare those to the EALs for rad effluent. Based on this comparison the candidate must make the correct EAL designation. Pedigree: NEW Objective: CLS-LP-301-B Obj 9: Given a hypothetical abnormal event and plant operating mode, use OPEP-02.1 to properly classify or re-classify the event

Reference:

OPEP-02.1 Cog Level: Hi

Explanation: Part 1: lAW the first override in RRCP (RRCP-2) TB ventilation is required to be placed in the recirculation mode of operation. Part 2: Due to Site Area Boundary dose >5000 mrem thyroid CDE, a GE is the highest classification. Distractor Analysis: Choice A: Patti is plausible because it is correct, see explanation. Part 2 is plausible because the main stack and turbine building tad monitors are reading > the SAE setpoint. However, the dose assessments results are above a GE classification. Choice B: Correct Answer, see explanation. Choice C: Patti is plausible because unit two has a once through mode of operation, and a novice applicant might assume that would be used to minimize turbine building dose rates. Part 2 is plausible because the main stack and turbine building tad monitors are reading > the SAE setpoint. However, the dose assessments results are above a GE classification. Choice D: Parti is plausible because unit two has a once through mode of operation, and a novice applicant might assume that would be used to minimize turbine building dose rates. Part 2 is plausible because it is correct, see explanation SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [i 0 CFR 55.43(b)(5)] The SRO candidate is required to compare the given radiation release values (including site boundary), and compare those to the EALs for rad effluent. Based on this comparison the candidate must make the correct EAL designation.

RAD IOACT13J ITY RELEA SE C ONTRDL 001-3710 PROCEDURE BASIS DOCUMENT Rev 01 1 Page 7 of i 5.2 Step RRCP-2 White in thl procedure: IF ThEt Ei1hr, Per1cm (P 31.3 113 enIkihcin iHl iTfl:)WN

• Pc TB FIW.J ii, f:sjSi9,iifl ?.IIik,
• TB venta1ion is in Once 1 t,rouh la1e Ensere tcln TI Ar Fitier EeIst Fens Hit ,iJ Fuel ture ind;cjed b Ensure Contuzil Bu:ltnt Ernerqency r;pi;tiitJ t fl
• Miiit Seain Une Rti t-II tU4 23. 2.f
• Pro ssO11LasR.1t-Ii)LiAC3 5-2:,
• Prncesc oo t P Rid H f-4i P,ictctr hiiiW n broderi #qut FRfltiij.t* tu rtt n itinq prt er [jMr, 11L12 Step RRCP-2 is a procedure ovethde which applies the entire time RRCP is being executed. Each of the three components specify applicable conditions and direct performance of actions as discussed below 5.2.1 Step RRCP-2 First Override Continued personnel access to the turbine building may be essential for responding to emergencies or transients which niay degrade into emergencies The turbine building is not an air tight structure, and radioactivity release inside the turbine building would not only limit personnel access but would eventually lead to an unmonitored ground level release, or release via the turbine building ventilation if operating in the once-through lineup Operation of the turbine building ventilation in the recirculation lineup helps to improve turbine building accessibility. In addition, since both units share a common turbine building airspace, if the building is intact, removing turbine building ventilation from once through lineup will terminate a large unfiltered votume discharge flow path for a leak on either unit Due to normal operational requirements when in once through lineup, at least one Air Filter Exhaust Fan and WRGM will be in seniice providing a monitored and filtered discharge flowpath.

,:U;:.z:., . :z. ....rr.r..-2r.fl.t.-_u Table R.1 ElTluent Monitor CLassrfica4on Threshokis

   *rn-ut-:t-rrnr-,L-trz ==t_.=-                  -   - --     z.r..              .-.rnn Releise Point                Monitor                GE                    SAl                  Alert                    liE M3tnctRa                              D12-M-2s         Z13ECt5ec              13E+tt.,.seo         atJ7jctjsec                 2+O.CLssc ectodgntNotteGas                   CAC-,i2H-12t4-3                                                     ,                6.I4E+C1pn Turtuns  u icrg Vent Rad               Ot2-t-2         I 07E+36 iCtsec      tL7E+C7 se             1 7E+]        Csez      I.13E+t    jCtsec 5er,1 Vte1 Et7luert Rad              Ci2-RM-KtC                                                                            2 Xli aarni a Raste Eluert Rad                      Dl2-M-KtC4                                                                          2 K li-N 3iarnl V.       t::r4vr:z trttc                     :r:.r              rzztt4:-a

GENERAL EMERGENCY SITE AREA EMERGENCY NQ1 Nesaseoflaseoasrsdictactt&ity rehuwrgm oniwaoepraater Thai 1,DBOrrenTEOEori,D1Omremb1)md DDE tsi tietease CS QSSCOUE tawioaccsr resucag in nne sose greater Wa: too attest TEDE or 500 strain Thyrod CSE I I I 2 3 ) 4 I 5 JOEFI I 1 2 I a j 4 S IDEFI RGI.1 RSII in the absence of reat-time dose assessment, read irg on any In the absenoe of real-done dose assessment reading on any 15 Table R-t effluent radiation monitor> ootumn SAE for 15 Tabe R-1 effuent radiation monitor> column GE for min. Wctes 1,23. 4i mm. iNotes 1,2. 3, 4 RGI.2 RSI2 Dose assessment using actial meteorology indicates doses Dose assessment using actual meteorology indicates doses > 1000 m rem TEDE or 5000 mrem thyroid CDE at or beyond > 100 mt-em TEDE or 500 mrwn thyroid CDE at or beyond the SITE EGUNDARY iNote 4) the SiTE BOUNDARY qNote 4) RGIi RS13 Field survey resuts indicate EITHER of the following at or Reid survey results indicate EITHER of the following at or beyond the SITE SQUNDARY: beyond the STE BOUNDARY

-  Closed window dose rates> 1000 mRhr espeoted to          -  Closed aindcw dose rates> CO mRhr expected to continue for 60 mm.                                           continue for 50mm
-  An a!ses of feld su r.wy samples in doate thyroid CDE      -  Analyses of Steid surey samples indicate thy-oid CDE 5000 mt-em for 60 mm. of inhalation.                       >500 mrerst for 60 mm of inhalation.

iNotes 1,2) Nates 1.2)

93. S600000 1 Unit One and Unit Two are executing OASSD-O1, Alternative Safe Shutdown Procedure Index, due to a fire in Main Control Room back panels requiring Main Control Room evacuation. Current plant conditions are:

Unit One and Two have scrammed All MSIVs are shut Which one of the following completes both statements below? The CRS will enter OASSD-02, Control Building, and (1) OASSD-O1. The CRS will direct actions to achieve a safe shutdown using (2) A. (1) exit (2) HPCI B. (1) exit (2) RCIC C. (1) concurrently perform (2) HPCI D. (1) concurrently perform (2) RCIC Answer: B K/A: 600000 Plant Fire On Site AA2 Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: (CFR: 41.10 /43.5 /45.13) 07 Whether malfunction is due to common-mode electrical failures RO/SRO Rating: 2.6/3.0 Tier 1 / Group 1 K/A match: THIS QUESTION WAS PRE-SUBM1TTED FOR APPROVAL. The applicant is required to determine that based on a fire in the MCR requiring evacuation, common mode failures of electrical equipment could occur and interpret the procedure steps that require exiting ASSD-01 and entering the standalone ASSD-02. In addition, the applicant will determine which train of ASSD equipment (HPCI/RCIC) will remain unaffected by the potential common mode electrical failure. Pedigree: Modified 14 NRC Objective: LOI-CLS-LP-301 Obj 20 Given a fire in an ASSD area, describe the potential impact that the fire may have on Safe Shutdown Equipment

Reference:

None Cog Level: Fundamental

Explanation: Part 1: OASSD-02, Control Building, is an outside Control Room shutdown procedure for both units. This procedure is a stand-alone post fire shutdown procedure for a Control Room evacuation and requires the reactors to be in hot shutdown/manually scrammed prior to leaving the Control Room. There are parts of the control building that this procedure is not used for, i.e. battery rooms. Part 2: The safe shutdown strategy for control room evacuation requires the B train of ASSD equipment (RCIC). Distractor Analysis: Choice A: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because the A train of Safe shutdown equipment (HPCI) is used in other ASSDs. Choice B: Correct Answer, see explanation Choice C: Part 1 is plausible because for every other ASSD fire OASSD-01 is performed concurrently with specific ASSD sub procedures. Part 2 is plausible because the A train (HPCI) of Safe shutdown equipment is used in other ASSDs. Choice D: Part 1 is plausible because for every other ASSD fire OASSD-01 is performed concurrently with specific ASSD sub procedures. Part 2 is plausible because it is correct, see explanation. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]. requires the SRO applicant to decide the appropriate transition to an event specific subprocedure based on fire in the control building. 2014 Question: A fire in the control building fire area requires ent into OASSD-Oi, Alternative Safe Shutdown Procedure Index. The CRS has determined that alternate safe shutdown actions are required Both Unit One and Unit Two have been manually scrammed. Which one of the following completes the statements below lAW DASSD-Oi? The next action that is required is to (I) Following this action both units will (2) A. (1) place MSIV control switches in close (2) perform OASSD-Oi, Alternative Safe Shutdown Procedure Index concurrently with OASSD-02, Control Building.

6. (1) trip both Reactor Recirc pumps (2) perform OASSD-O1, Alternative Safe Shutdown Procedure Index concurrently with OASSD-02, Control Building.

O (1) place MSIV control switches in close (2) exit OASSD-01, Pjternative Safe Shutdown Procedure Index and enter OASSD-02, Control Building D. (1) trip both Reactor Recirc pumps (2) exit OASSD-01, Alternative Safe Shutdown Procedure Index and enter QASSD-02, Control Building

15.2 IF the fire is in the Control BuiIinç fire area. AND control room evacuation is required. ThEN PERFORM the follosin

a. MANUALLY SCRAM Unit 1 reactor.

Ii PLACE Unit 1 MSIV control switches in CLOSE.

c. MANUALLY SCRAM Unit 2 reactor.

d. PLACE IJnit 2 MSIV control switches in CLOSE.

e. Both units EXFI this procedure AND ENTER ASSD-2, Control Building.

OASSD-O1 Rev 41 Page 501 16

3.0 OPERATOR ACTIONS 3.5.3 IF the fire is NOT in the Control Buildinu, THEN ENTER the applicable ASSD procedure AND EXECUTE concurrently with this procedure. NOTE: A [oss of drveIi tooting can be deterruined using CAC-TR-442-1A. CAC-TR4426-2A. in (he Control Room or CAC-TR77B, points I, 3., and 4, at the Remote Shutdown Panel. The time the loss of dre[I tooting occurred can be determined from the recorder disp[ay information 3.5A IF RCIC ot HPCI is injecting AND dr9wel[ cooling has been lost, THEN START reactor vesse[ coo[down at 1OOFthr or greater within 6 minutes of the toss ot dreiI cooling. 35.5 IF drjweli temperature control is lost, THEN PERFORM the tol[owing to preseRe containment overpressure lot RHR pump net positive suction head:

1. STOP A, B. C and D RBCCW pumps for the afleeted unit.

2 CLOSE the following valves for the affected unit

                -        DW EQUIP DRAIN D ISOL VL G i6-FOf
                -        OW EQUIP DRAIN OTBO 1SOL VLV, 6
                -        OW FLOOR DRAiN IND ISOL VLV. 616-F003
                -        OW FLOOR DRAIN OTBO 1SOL VLV 616-FOOt OASSD-O1                                 Rev 41                          Page 6 of 16

2.1 This procedure is entered from Alternative Safe Shutdo.vn Procedure Index, DASSD-DI, AND 2.2 Unit CR8 has determined both reactors are to be broghtto safe and stab[e conditions tram outside the Control Rocn usnj ASSD Train B. 3.0 OPERATOR ACTIONS 3.1 CONTINUE [mplementation of this procedure by perfom,ing the steps in Section A. 3.2 IF the tire is extinguished whiie executtn this procedure AND the Unit CR8 detemines no action within this procedure is required. THEN EXIT this procedure. 4.0 RESTORATION 4.1 RETURN plant to general operating condition as directed by plant management. OASSD-02 Rev. 57 Page 3 of 154

94. SG2.1.05 1 Which one of the following completes both statements below?

(Consider each statement separately.) lAW Tech Spec 5.2.2, Facility Staff, the shift crew composition may be less than the minimum requirement for a period of time not to exceed (1) for an unexpected absence of on-duty shift crew members. lAW 001-01.01, BNP Conduct of Operations Supplement, the minimum required number of Auxiliary Operators for manning a shift at BNP is (2) A. (1) one hour (2) three B. (1) one hour (2) nine C. (1) two hours (2) three D. (1) two hours (2) nine Answer: D K/A: G2.1 .05 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 / 43.5/ 45.12) ROISRO Rating: 2.9/3.9 Tier 3 K/A match: knowledge of required tech spec and conduct of ops shift manning requirements. Pedigree: 2012 BNP NRC Objective: LOI-CLS-LP-200-B Obj.12.-ldentify conditions and limitations in the facility license.

Reference:

None Cog Level: Low Explanation: lAW the procedure, 9 AO makeup the minimum shift staffing and two hours is the time to find a replacement. One hour is the time on stepping out limitation of the control room personnel. The tech Specs 5.2 only address the number of AOs for the Units which is 3, this does not take into account ASSD and Fire Brigade. Distractor Analysis: Choice A: Plausible because TS 5.2.2 requires 3 AOs for both Units which does not take into account ASSD and Fire Brigade requirements. One hour is the stepping out time limit for control room personnel. Choice B: Plausible because nine is correct but one hour is the stepping out time limit for control room personnel.

Choice C: Plausible because TS 5.2.2 requires 3 AOs for both Units which does not take into account ASSD and Fire Brigade requirements. Choice D: Correct Answer, see explanation SRO Basis: Conditions and limitations in the facility license. (10 CFR 55.43(b)(1 )). Requires the SRO applicant to know the limitations for shift staffing in the license. BNP CONDUCT OF OPER41IONS SUPPLEMENT OOI-Dl .01 Rev. 73 Page 15 01191 5.5 Operations Shift Staffing 5.5.1 General

1. In addition to the requirements ot AD-OP-ALL-100D. the following requirements apply:

a The following table outlines the administrative guideline for the nomial Operations shift complement. Any deviation from the normal shift complement must remain in accordance with Section 522 ol Technical Specifications, applicable sections of OASSD-G0. Users Guide. OFPP-031 Fire Brigade Staffing Roster and Equipment Requirements, and OERP, Radiological Emergency Response Plan (ERP) (Attachment 13, Operations Staffing Roster contains a listing of required ERO watch Stations and qua[iflcations for each and ASSD positions.) BNP Watchstations BNP Shift Complement License Shift Manager (SMI I Shift Manager SRO Control Room Superiisor (CRS) 2 CRSs (1 for each unit) SRO Reactor Operator (RO) 4 Reactor Operators (typically, 2 RO!SRO for each unit) Auxiliary Operator AO) 9 (includes 2 in Radwaste) NA Operations Center SRO 1 Operations Center SRO SRO STA [Note 1] 1 STA 5Th Qualified Notes: Organ izaticin 5.2 5.2 Organization 5.2.2 Facibty Staff (continued)

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, when either unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the confrol room. Wth one unit in MODE 1, 2, or 3 and the other unit defueled, the minimum shift crew shall include a total of two SROs and two ROs.

c. Shift crew composition may be less than the minimum requirement of 10 CFR 5D.54(m)(2Xi) and Spectficatons 5.2.2.a and 5.2.2.g for a period of Ume not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members proded immediate action is taken to restore the shift crew compositon to within the minimum requirements.

d. An individual qualified in radiation protection prccedures shall be on &te when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Deleted.

95. SG2.1.43 I Following the bypass of Unit Two feedwater heaters 4A and 5A, the following plant conditions exist:

Reactor Power is 60% Feedwater Temperature is 330° F Final Feedwater Temperature vs Power Nominal FWTemp 110.3T Reduced RI PVVR Nominal F% Temp Reduced iUT FFWT 65% 394.4 381.4 296.4 61% 3911 383.1 295.5 63% 391.! 381 .1 294.6 62% 3904 380.1 293.t 61% 389.0 379M 292.8 60% 387.6 377.6 291.9 lAW 001-01.01, BNP Conduct of Operations Supplement, which one of the following completes both statements below? (consider each statement separately) The CRS (1) required to implement the thermal limit penalties for FHOOS (feedwater heater out of service). Entry into Tech Spec 3.0.3 (2) required if final feedwater temperature is less than the 110.3°F reduced final feedwater temperature value. A. (1) is (2) is B. (1) is (2) is NOT C. (1) is NOT (2) is D. (1) is NOT (2) is NOT Answer: B K/A: G2.1 .43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc. (CFR: 41 .10 / 43.6 / 45.6) ROISRO Rating: 4.1/4.3 Tier 3 K/A match: The applicant is required to use the final feedwater temperature reduction attachment to determine if the effect of the feedwater reduction is severe enough on reactivity to requite implementation of thermal limit penalties.

Pedigree: New Objective: CLS-LP-032 obj 27 Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR, determine whether given plant conditions meet minimum Technical Specifications requirements associated with the Condensate and Feedwater System.

Reference:

NONE Cog Level: High Explanation:  : Parti: A final fw temp of 385°F is less than the nominal FW temp for 60% power, but >10°F reduced from nominal, therefore the thermal limit penalties for FHOOS do not need to be implemented. Part 2:There are NO core operating limits specified in the COLR for operation beyond 110.3°F Final Feedwater Temperature. Thermal limits CANNOT be verified to be within the limits specified in the COLR, which requires entry into the Actions of LCD 3.2.1, 3.2.2, and 3.2.3. These LCOs require thermal limits be restored within 4 hours, LCD 3.0.3 is not entered, Distractor Analysis: Choice A: Part 1 is correct, see explanation. Part2 is plausible because a candidate may believe since there are no thermal limits specified in the COLR for this condition, LCD 3.0.3 would be applicable. Choice B: Correct Answer, see explanation Choice C: Part 1 is plausible because a final 1w temp of 330° F is less than the nominal FW temp reduced by 10°F for 60% power, but greater than the 110.3°F reduced FFWT, a novice applicant may believe thermal limit penalties are only applied at the 110.3°F value. Part2 is plausible because a candidate may believe since there are no thermal limits specified in the COLR for this condition, LCO 3.0.3 would be applicable. Choice D: Part 1 is plausible because a final fw temp of 330°F is less than the nominal FW temp reduced by 10°F for 60% power, but greater than the 110.3°F reduced FFWT, a novice applicant may believe thermal limit penalties are only applied at the 110.3°F value. Part 2 is correct, see explanation. SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)J This question requires that the applicant determines whether the TS thermal limits should incur a penalty. In addition, it also requires that the candidate determines whether LCD 3.0.3 applies for a given condition.

LCD 3.0.3 When an LCD is not met arid the associated ACTtDNS are not met, an assocted ACTLON is not provided. or if directed bythe associated ACTIONS, the unit shall be placed in a MDGE or other specified condition in which the LCD is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable. in:

a. MODE 2 within 7 hours:

b. MODE 3 within 13 hours: and

c. MODE 4 within 37 hours.

Exceptions to this Specification are stated in the indidual Specifications. Where corredi,e measures are carnpleted that permit operation in aordance with the LCD or AOTIONS. completion of the actions requited by LCD 3.0.3 is not requited. LCD 3.0.3 is only applicable in MODES 1,2. arid 3. 8NP CONDUCT OF OPERATIONS SUPPLEMENT 0O[-91 .01 ReQ. 76 Page 131 of 191 ATTACHMENT f9 Pace 3 o 4

                       << Equipmeiit Out Of Service Contingencies>>

EQOS Required Action Note 1.2] Power Condition

                              . Reduce rCactor power to 5C%

qn

                              . Impment OGP-14.

SLO Any

• Implement applicable SIC power to flow map.

                              . IF 23% RIP. THEN implement theral lirit pena ty.
                              . ReFerence 15341.

• Implement thermal limit penafty.

IBVOOS 23% RTF

• Reference 13 3. f.
• IF the aIue in the 1 1 O.3F Reduced R Temp column of 1 1210P-32, 23%RTP Att, FHCOS THEN enter [CC 3.2.1. 3.2.2 and 3.23.

(FWTR)

• IF> JOT below nominal FW mperarure, (FFTRi 23% RTP tn,piement appl:caoe FWTR power to flow map.

0 tnmp[ement thermal lin,tt penalty. a Ininlantant nnIirtkIa fl,1T ra..nar fly... ni .n CONDENSATE AND FEED TER SYSTEM 20P-32 OPERATING PROCEDURE Rev 206 Piqe400 of 408 ATTACHMENT 6 Pdqe 2 of 2 Final Feedwater Temperature vs Power

                                .                   Nominal RVTemp                    110.3F Reduced RX PWR        Nominal RV Temp              Reduced 10F                          FPuVT 65%                  394.4                     3844                            296.4 64%                  393.1                     383.1                           295.5 63%                 391.1                     381.7                           2946 62%                  3904                      380.4                           293.7 61%                  389.0                     379.0                           292.8 60%                 37.6                      3Th6                            291.9 59%                 386.2                     3762                            290.9

96. SG2.2.15 1 Unit One is operating at rated power.

A-03 (2-2) Auto Depress Control Pwr Failure, is in alarm due to Fuse F5 being blown. (REFERENCE PROVIDED) Which one of the following completes both statements below? Fuse F5 is located on ADS Logic (1) ADS (2) operable. A.(1) A (2) is B. (1) A (2) is NOT C. (1) B (2) is D. (1) B (2) is NOT Answer: C K/A: G2.2.1 5 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc. (CFR: 41.10/43.3/45.13) RO/SRO Rating: 3.9/4.3 Tier 3 K/A match: THIS QUESTION WAS PRE-SUBMITTED FOR APPROVAL. This question requires the candidate to use a drawing to determine operability of ADS. Pedigree: New Objective: LOI-CLS-LP-020 OBJ 1 5d. Given plant conditions, predict how ADS/SRVs will be affected by the following: Loss of DC power

Reference:

1-FP-05887 (Block out references to which logic string is logic A and B) Cog Level: High Explanation: Part 1: Fuse F5 is located on the alternate power source, only logic B has an alternate power source. Therefore, the fuse is on logic B, Part2: Since the drawing is shown in the de-energized state, fuse 5 being blown will have no impact on ADS instrumentation. ADS remains on its normal power source.

Distractor Analysis: Choice A: Part 1 is plausible the fuse is located on 125V DC 3A power or the operator might forget which train of logic has two power supplies. Part 2 is plausible because it is correct, see explanation. Choice B: Part us plausible the fuse is located on 125V DC 3A power or the operator might forget which train of logic has two power supplies. Part 2 is plausible because if the drawing was shown in the energized state, this would be correct. Choice C: Correct Answer, see explanation. Choice D: Part 1 is plausible because it is correct, see explanation. Part 2 is plausible because if the drawing was shown in the energized state, this would be correct. SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)J Requires the SRO candidate to use a drawing to determine the status of ADS power and to know whether that loss effects ADS operability. CtOE3 r ,, LE flP5

97. SG2.2.22 1 Unit Two is operating at rated power.

While performing OPT-07.2.4A, Core Spray Loop A Operability, Core Spray Room Cooler A fails to start when Core Spray Pump A is started. The reactor building AD reports that the room cooler tripped on thermal overload. lAW AD-OP-ALL-i 000, Conduct of Operations, which one of the following completes both statements below? (consider each statement separately) CoreSprayLoopA is (1) A one time reset of the thermal overload (2) allowed before a Maintenance and Engineering evaluation. A. (1) OPERABLE (2) is B. (i) OPERABLE (2) is NOT C. (1) INOPERABLE (2) is D. (i) INOPERABLE (2) is NOT Answer: D K/A: G2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41 .5 I 43.2 I 45.2) ROISRO Rating: 4.0/4.7 Tier 3 K/A match: Requires knowledge of conduct of ops procedure to determine whether Core Spray A meets the conditions for operability in the tech specs based on cooler operation. Pedigree: Bank NRC 08 Objective: CLS-LP-1 8, Obj. 18. Given plant conditions and TS, including bases, TRM, ODCM, and COLR, determine the required actions to be taken in accordance the TS associated with the Core Spray System.

Reference:

None Cog Level: Hi Explanation: Part 1: Per 001-01 .01, When any ECCS Room Cooler is determined to be INOPERABLE, then the ECCS equipment associated with that room cooler is to be declared INOPERABLE per the applicable Technical Specifications. Part 2: Per AD-OP-ALL-i 000, the breaker should only be reset once the condition is identified and corrected, and plant conditions dictate the reset before maint and eng personnel are available.

Distractor Analysis: Choice A: Part us plausible, because the room cooler is not part of the Core Spray system listed in the tech spec bases. Part 2 is plausible because during transient conditions the breaker could be reset, however, the plant is in a stable condition. Choice B: Part 1 is plausible, because the room cooler is not part of the Core Spray system listed in the tech spec bases. Part 2 is correct, see explanation. Choice C: Part 1 is correct, see explanation. Part 2 is plausible because during transient conditions the breaker could be reset, however, the plant is in a stable condition. Choice D: Correct Answer, see explanation. SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)J Requires the SRO candidate to have knowledge of conduct of ops procedure to determine whether Core Spray A meets the conditions for operability in the tech specs based on cooler operation. Co NO UCT 0 F OPERATI ONS AD-C P-ALL-i 000 Rev. 4 Page 80 of 1133 5.19 Resetting Protective Devices {71.4} 5.19.1 Standards

1. Protective devices should not be reset without a clear understandin of the reason for the protective device trip.

2. The overriding priority for the operating crew upo the trip of any protective device is to stabilize the plant and restore the systems to the safest possible condition 5.1 9.2 Expectations

1. Protection devices which have actuated (breakers, fuses, bistables. MDV thermal overloads, lockouts, etc,) shou[d only be restored with shift superision approval, under the following conditions. The fol:lowing conditions do not app[y to 120 volt breakers that only supply lighting or receptacles.

a. The cause of the actuation has been identifted and corrected t) Restoring the protective device is not recommended unless plant conditions dictate that the component repositioning must be complted before Maintenance and Engineering personnel are available Remote operation of the component with no personnel in the immediate area after resetting the protective device is recommended if repositioning is required prior to completion o the evaluation by Maintenance and Engineering.

2. The SM may approve additional protective device resetting after consultation with Engineering

BNP CONDUCT OF OPERATiONS SUPPLEMENT OOL-aDl .01 Rev. 73 Page 36 of 191 5.16.2 Degraded Equipment Controls System C otuponent Related Guidance (continued (1) Reference IRM Appendix F, Safety Function Determination Program (SFDP;, Atachments 1 and 2 to assist with determination of Technical Specification 3.8.1 and 3.9.7 requirements and to assess the possible impact on supported systems. (2) If an evaluation of the SFDP is performed, then document the evaluation and the results in the the narrative log or on Attachment 26, if the narrative log IS not available.

4. ECCS Room Coolers (Ti ,3}

NOTE

• The following step is not required to be performed it the EOCS Room Cooler is INOPERABLE due to the loss of a 4160V or 490V E-Bus. E-Bus [NOPERABILIfl impacts the OPERABILITY of ECCS subsystems. Technical Specifications and the SFDP will provide Required Actions to be taken for the loss of the E-Bus.
• in Mode 4 and Mode 5, ECCS Room Coolers are not required to be OPERABLE to support OPERABILITY of the associated ECCS Systems.

a. When any ECCS Room Cooler is determined to be INOPERABLE, then the ECCS equipment associated with that room cooler is to be declared INOPERABLE per the applicable Technical Specifications.

EXAMPLE The RHR Room Coolers are to be considered redundant components required to support the operation of RHR. Therefore. should a room cooler be found or made INOPERABLE, a 7 day Active LCO is required to be established on the RHR system. Likewise, should both room coolers be found INOPERABLE, the action required is the same as if both RHR loops and HPCI were INOPERABLE. Should it be identified that one RHR Room Cooler is INOPERABLE and one RHR Loop is also INOPERABLE (specific combinations do not matter), the action is as if only one RHR Loop is INOPERABLE (7 days:. flnntrnl Riiildinn HVAfl Air flnrnnrnccnrc

98. SG2.3.11 1 Following a small steam line break in the drywell plant conditions are as follows:

Drywell pressure: 25 psig and rising Drywell hydrogen: 1.3% Suppression Chamber hydrogen: 1 .2% Torus level: 42 inches Which one of the following completes both statements below? The CRS is required to direct venting containment lAW OEOP-O1-SEP-O1, Primary Containment Venting, using (1) Venting of the (2) will be directed first. A. (1) Section 2.1, Containment Pressure Control (2) drywell B. (1) Section 2.1, Containment Pressure Control (2) torus C. (1) Section 2.2, Containment Hydrogen Control (2) drywell D. (1) Section 2.2, Containment Hydrogen Control (2) torus Answer: D K/A: G2.3.1 1 Ability to control radiation releases. (CFR: 41.11 / 43.4 /45.10) ROISRO Rating: 3.8/4.3 Tier 3 K/A match: Requires the ability to determine the procedure section for venting, and the correct sequence of termination of venting. Pedigree: New Objective: CLSLP300L*08d Given the Primary Containment Control Procedure and plant conditions, determine if the following actions are required: Venting the primary containment IRRESPECTIVE of radioactivity release rate limits

Reference:

None Cog Level: High Explanation: Part 1: Following the H2 leg of the PCCP with the given conditions will drive you to step PC/G-9 which directs you to Vent Containment per EOP-01 -SEP-01, since H2 is the driving condition for venting, then section 2.2 is the appropriate section to implement. Part 2: lAW SEP-01, the torus is vented first as long as the torus water level is less than 6 feet.

Distractor Analysis: Choice A: Part 1 is plausible because H2 concentration is less than the entry limit into POOP (entry at 1 .5%), and Containment pressure is >11 .5 psig (pressure for DW sprays), therefore a novice applicant might believe that the appropriate procedure section required is for containment pressure control. Part 2 is plausible since venting of the drywell is performed first if torus water level is >6 feet. Choice B: Part 1 is plausible because H2 concentration is less than the entry limit into PCCP, and Containment pressure is >11 .5 psig (pressure for DW sprays), therefore a novice applicant might believe that the appropriate procedure section required is for containment pressure control. Part 2 is correct. Choice C: Part 1 is plausible since it is correct, see explanation. Part 2 is plausible since venting of the drywell is performed first if torus water level is >6 feet. Choice D: Correct Answer, see explanation.. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J Requires knowledge of diagnostic step in EOP, and selection of appropriate emergency contingency procedure.

                       /=\

7 WHEN Oyweli2Rtoc.s H2 re.sches 1% I THEN

                      \continue/
                                    /CG-5 Notify E&C to sample cotanenl fr vening.

rnp Hydroe.9 Water Cemis1y. PG-7 r WHEN j wilOerrred ODCM MQI be exceeied L\ THEN

                     \\ COfltiflUe,,/ PCG4 Vent cojtainmeit per EOP-O1-SEP-Oi

PRt MARY CONTAINMENT VENI[NG DEOP-Di-SEP-Ol Re D2 Page 12 of 21 2.2.3 Continrnent Hydrogen Control ctions (continued) e Close VA-D-BFV-RB :S6Gi A [sal Daniperi D IRO

t. Close VA-H-B FY-RE :SBGT B [so I Damperi D IRO ci. IF venting the torus, THEN open:

(1) CAC-V7 (Tows Purge Exh VIv) C IRO (2) CAC-Va (Tows Purge Exhi VIv) U RD

h. IF venting the drjwelt, THEN open:

(1) CAC-V9 (DryeII Purge Exh VIv) C RD (2) CAC-ViG (DrweII Purge Exh VIv) C RD

i. Open VA-F-BFV-RB (SBGT E Suct Damper) C RD i3 IF c[irected to terminate torusentitig.

THEN: a Enstire pnmarj containment purging term[nated per EOP-O1-SEP-05 C RD

99. SG2 4.30 1 Unit Two is operating at rated power with LPCI A inoperable and the following sequence of events occurs:

0000 7 day completion time for LCD 3.5.1, ECCS Operating, Condition A expires and Condition C is entered requiring that the Unit be placed in MODE 3 in 12 hours. 0030 Plant shutdown is commenced per LCD 3.5.1, Condition C. 0050 LPCI A is repaired and declared operable; LCD 3.5.1 Conditions A and C are exited. 0100 Management decides to continue the plant shutdown as planned to complete other maintenance items. 0230 Unit Two in MODE 3 (REFERENCE PROVIDED) Which one of the following identifies the reportability requirements, if any, for this event? A report to the NRC: A. is not required. B. would be submitted no later than 0400. C. would be submitted no later than 0430. D. would be submitted no later than 0630. Answer: C K/A: G2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11) RO/SRO Rating: 2.7/4.1 Tier 3 K/A match: Applicant required to determine report status for the given condition. Pedigree: Bank Objective: LOl-CLS-LP-201-D, Obj 11 Explain the following regarding NRC Reporting requirements per AD-LS-ALL-0006, Notification/Reportability Evaluation: d. Determination of clock start time for reportable events (LOCT)

Reference:

001-0 1 .07 Attachment 1 Cog Level: High Explanation: A TS required shutdown requires a 4 hour NRC report. The time starts when the shutdown is started, Completion of the shutdown required by TS is an LER.

Distractor Analysis: Choice A: Plausible because the plant was not shutdown due to the TS as LPCI was repaired and the TS exited. Choice B: Plausible because this would be 4 hours from when the TS shutdown condition was entered. Choice C: Correct Answer, see explanation. Choice D: Plausible because an LER is required after completing a TS required shutdown, but in this case the shutdown was not IS requited. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]. Requires SRO administration procedure knowledge of reportability requirements based on plant conditions. NOTIFICATIONS 001-01.07 Rev 35 Page 25 of 43 ATTACHMENT 1 Page 2 ol 7 Reportabihty Evaluation Checklist NOTE

• If the answer to any of the following questions is YES, the event is reportable within 4 hours.
• It all answers to the following questions are NO, the event is not reportabLe vlthin 4 hours.

4 HOUR REPORTABIUTY ITEM YES NO DESCRIPTIVE QUESTION NOTE Includes any Safety Limit ioDaon Thch Spec 22L 1 Is plant shutdown required by tecinical specifications being initiated? [10 CER S072{bii2 Plant Shutdown Required by Technical Specifications (See Section 3.2.1 of this report)

§ 5012(b)(2)f 1) The initiation of any nuclear          § 50.73(a)(2)fi)(A) The completion of any plant shutdown required by the plants Technical        nuclear plant shutdoi required by the plants Specifications.                                        TechnicaL Specifications

Discussion The 10 CFR 50.12 reporting requirement is intended to capture those events for which TS require the initiation of reactor shutdown to prode the NRC with early warning of safetysignificant conditions serious enough to warrant that the plant be shut down. For 10 CFR 50.72 reporting purposes, the phrase initiation of any nuclear plant shutdown includes action to start reducing reactor power; i.e., adding neqative reactivity to achieve a nuclear plant shutdown required by TS This includes initiation of any shutdown due to expected inability to restore equipment prior to exceeding the [CO action time. As a practical matter, in order to meet the time lim its for reporting under 10 CFR 50.12, the reporting decision should sometimes be based on such expectations. (See Example 4.) The initiation of any nuclear plant shutdown does not include mode changes required by TS if they are initiated after the plant is already in a shutdown condition. A reduction in power for some other purpose, not constituting initiation of a shutdown required by TS, is not reportable under this criterion. For 10 CFR 50.73 reporting purposes, the phrase completion of any nuclear plant shutdown is defined as the point in time during a TS-required shutdown then the plant enters the first shutdown condition required by an LCO (e.g., hot standby (Mode 3) for RVRs with the Standard Technical Specifications (STS)). For example. if at 0200 hours a plant enters an [CO action statement that states, testore the inoperable channel to operable status within 12 hours or be in at least Hot Standby within the next 6 hours, the plant must be shut down (i.e., at least in hot standby) by 2000 hours. An [ER is required if the inoperable channel is not returned to operable status by 2000 hours and the plant enters hot standby. An LER is not required if a failure was or could have been corrected before a plant has completed shutdown (as discussed above) and no other criteria in 10 CER 50.73 apply. NOTIFICATIONIREPORTABILIfl EVALUATION AD-LS-ALL-0006 Rev. 0 Page 11 of 17 5.4 Making Emergency Notification System and LER Reports (continued) Table 1. Emergency Notification System Reporting Overview Event or ENS notification ENS notification ENS notification 60-day LER Job Aid condition within 1 hour within $ hours within 8 hours section Plant shutdown Initiation of SD corn pleton of a 4.3, 44 (5/0) required by required by Tech 3/0 required by Tech specs Specs [50.72 Tech Specs [50.73 (b): 2)/i] (a)(2i( i)IA)

100. SG2.4.35 I Unit One and Unit Two have entered SBO procedures at time 1300 due to a loss of all onsite and offsite power. Which one of the following completes both statements below? lAW 1 EOP-01-SBO, Station Blackout, opening the reactor building roof hatch is required to be performed before (1) lAW 001-37.14, Station Blackout Procedure Basis Document, the reactor building doors and roof hatch are opened to ensure (2) A. (1) 1330 (2) equipment availability B. (1) 1330 (2) habitability C. (1) 1500 (2) equipment availability D. (1) 1500 (2) habitability Answer: D K/A: G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: 41 .10/43.5/45.13) ROISRO Rating: 3.8/4.0 Tier 3 K/A match: Requires the applicant to have knowledge of when to implement the AD task in a EOP sub procedure (opening RB roof hatch during SBO), and the operational effects if not completed (jeopardized habitability). Pedigree: New Objective: LOl-CLS-LP-303-B Obj 2 Given plant conditions, EOP-01-SBO Flowchart, and SBO Support Procedures, determine the required operator actions. Temperature analysis states that access would be prohibited in the RB building due to 117 elevation ceiling temperature if the hatch was not opened.

Reference:

None Cog Level: Fundamental Explanation: This is a time sensitive action from the SBO procedure that directs the RB roof hatch to be opened within 2 hours of the start of the SBO.

Distractor Analysis: Choice A: Part 1 is plausible because this is the time critical action time limit in SBO procedure for opening control panel doors. Part 2 is plausible because high temperatures could be thought to jeopardize equipment availability, however the hatch and doors are opened to ensure habitability. Choice B: Part 1 is plausible because this is the time critical action time limit in SBO procedure for opening control panel doors. Part 2 is correct, see explanation. Choice C: Part 1 is correct, see explanation. Part 2 is plausible because high temperatures could be thought to jeopardize equipment availability, however the hatch and doors are opened to ensure habitability. Choice D: Correct, see explanation. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)f 5)] Knowledge of when to implement attachments and the basis for the step. STATION BLACKOUT 001-37.14 PROCEDURE BASIS DOCUMENT Rev 00 I Page 40 ot42 525 Steps 580-27 and 580-28 4 Open r icrhiiiIcFtig ckinrs pew EOP-O 1 -SO-4 F---.. RB mnf Time Sensitive hafrii req uired npeii ihtn 2 hours

         $g. alternate fuel poo niakeupspray                   Time iquipmwn per FOP 10 fFP t2 It RAP conditions exist, reactor buitding temperatures will use rapidly due to the loss of building ventilation. The refuel floor roof hatch and 20 elevation personnel access doors are blocked opened to provide alternate ventilation. The retuel floor roof hatch should be opened as soon as resources are available and is required to be open within 2 hours of the SBO start time recorded at Step SBO-1. The 20 elevation personnel access doors shoLild be opened as soon as resources are available and are required to be open within 6 hours of the SBO start time recorded at Step SBO-i by the text procedure. The reactor building temperature analysis (BNP-MECHFLEX-Q001) shows 117 elevation ceiling temperature will reach 114°F at time 2 hours, which is approaching the temperature that access uld be prohibited. Alternate ventilation should be established as eafly as possible based on pliorities and available resources.

Appendix D Scenario Outline Form ES-D-1 Facility: Brunswick Scenario No.: 1 Op-Test No.: 2016 Draft Examiners: Operators: Initial Conditions: Unit Two is operating at rated power. 1A NSW Pump is under clearance for planned maintenance. 2C TCC pump is aligned to Unit One. APRM 2 failed downscale and bypassed Turnover Start the 2C Condensate Booster Pump and Secure the 2A Condensate Booster Pump. 20P-32, Section 6.3.6 is completed up to step 6.3.6.11. Event Maif. Event Event No. No. Type* Description 1 N - BOP Swap Condensate Booster Pumps C -ATC Inadvertent HPCI Initiation w/ failure to trip 2 ESO14F C CRS

                              -                                                           fTS)(AOP-03.0) 3        ZUA343                    Off Gas Filter Hi d/P RC055       C ATC
                              -         Recirc Pump Trip D         C CRS
                              -                                                           fTS)(AOP-04.0) 5                     R ATC
                              -         Reduce reactor power for Single Loop Operation C  - BOP     Heater Drain Deaerator Controller Failure 6        CFO39F C  - CRS                                                            (AOP-23.0)

Fuel Failure 7 NBOO5F M (RSP)(AOP-05.4) K2501A Manual Scram FailureAlternate Rod Insertion 8 C K2503A Close Group I Valves Torus cooling Valve Failure 9 RHO2OF C (PCCP) (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Event Event Description Summary The Unit will be operating at rated power and after taking the shift turnover will swap 1 condensate booster pumps lAW 20P-32. (Start 2C, secure 2A) After the condensate booster pumps have been swapped, an inadvertent HPCI 2 initiation will occur. The HPCI system will not trip but can be manually isolated. AOP 03.0 and Tech Specs will be entered. After Tech Specs are addressed the Off Gas Filter Hi d/P alarm will be received. The standby Off Gas filter will be placed in service. VFD coolant leakage will occur that will cause a trip of the 2B Recirc pump. The supply 4 breaker does not trip and must be opened manually. OAOP-04.0 will be entered. Technical Specifications will be addressed. The plant will be greater than the allowable for single loop operation. Recirc flow must be lowered and/or control rods must be inserted to reduce power. With only one recirc pump in operation reactor power must be less than 50% and core flow must be greater than 30.8 Mlbs but less than 45 MIbs. After the power reduction the Heater Drain Deaerator controller will fail. AOP-23.0 6 will be entered. Fuel failure will begin requiring entry into ADP-05.4. When Process Off gas Hi Hi alarm with the Main Stack rising a scram will be inserted and Group I valves closed. The crew will be required to (1) Scram, (2) Close Group I valves. (Each of which are critical 8 tasks) The manual scram pushbuttons will not work, Alternate Rod Insertion (ARI) will be initiated to scram the rods. 9 Torus cooling valve will trip on thermals (will reset if attempted).

Appendix D Scenario Outline Form ES-D-1 Facility: Brunswick Scenario No.: 2 Op-Test No.: 2016 Draft Examiners: Operators: Initial Conditions: Unit Two is at 95% power. 1A NSW Pump is under clearance for planned maintenance. APRM 2 has failed downscale and is bypassed. 2C TCC Pump is aligned to Unit One. Turnover The OATC will reduce power to 850 MWe Gross (reactivity plan is to use recirc flow) The BOP operator will then Isolate 230 kV Delco West (Line 30) lAW the marked up of 2OP-50, Section 6.2.6. Event Maif. Event Event No. No. Type* Description 1 R ATC

                               -         Lower power to 850 MWe to remove 230 kV Line 30 2                      N BOP
                               -         Remove 230 kV Line 30 from service RDOOJ M       C ATC
                               -         Rod Drift (26-1 1)     C CRS
                               -                                                                      (TS)

C BOP

                               -         ADHR primary pump trip 4        K4526A C-CRS                                                                   (AOP)

C ATC

                               -         Recirc Loop B Flow transmitter Failure 5        NIO63F (TS)

C BOP

                               -         Heater Drain Deaerator Pump Trip 6       CFO89F C-CRS                                                                   (AOP)

Small steam leak in DW results in an ATWS requiring terminate 7 CAOO8F M and prevent actions (RSP)(ATWS) 8 K2119A C SLC Mode Switch Failure 9 K2624A C Alternate Rod Insertion reset failure (N)ormal, (R)eactivity, (I)nstrument, fC)omponent, (M)ajor

Event Event Description Summary 1 After taking the watch the CRS will direct power reduced to 850 MWe. 2 The BOP will isolate 230 kV Line 30. Control Rod 26-1 1 will start to drift in. The crew will enter OA0P-02.0 and take action lAW 2APP-A-05 (3-2). When the high temperature alarm is received Engineering will 3 report that scram times cannot be assured based on past history of the control rod. Determine TS 3.1.3 condition Cl to insert the control rod in 3 hours and C2 to disarm the control rod within 4 hours. After Tech Specs are addressed the Alternate Decay Heat Removal (ADHR) primary pump will trip. A0P-38.0 will be entered The Recirc Loop B flow transmitter to APRM Channel 4 will fail downscale resulting in a rod block and a trip input to each voter. The crew will respond per APP5 and bypass 5 APRM 4. The APRM will be declared Inoperable per TS 3.3.1.1, Condition A and placed in trip within 12 hours. APRM TS Actions to be taken requires the APRM mode selector switch to be place in INOP lAW 001-18 A motor overload will occur on Heater Drain Pump 2A. The crew will reference APP UA-06 1-7, Bus 2D 4KV Motor OvId and determine which pump has the overload 6 condition. The crew should start HDP 2C and secure HDP 2A. The crew may reference AOP-23.0. A small steam leak in the DW results in rising Drywell pressure requiring a reactor 7 scram. An ATWS will occur, conditions will require terminate and prevent actions to be performed. 8 When SLC is initiated, the mode switch will fail and the pumps will not start. LEP-03 will be executed to inject the boron into the core. Alternate Rod Insertion (ARI) will not reset, the crew will perform LEP-02 to drive 9 control rods into the core. When level is stabilized after terminating and preventing ARI will be repaired to allow the rods to be manually scrammed.

Appendix D Scenario Outline Form ES-D-1 Facility: Brunswick Scenario No.: 3 Op-Test No.: 2016 Draft Examiners: Operators: Initial Conditions: Unit Two is at 100% power. 1A NSW Pump is under clearance for planned maintenance. 2C ICC Pump is aligned to Unit One. Turnover The BOP will perform PT-40.2.1 1, Main Generator Voltage Regulator Manual And Automatic Operational Check. Event No. [ Maif. No. Event Type* Event Description Perform PT-40.2. 1 1 Main Generator Voltage Regulator 1 N-BOP Manual And Automatic Operational Check 2 ZA41 1 DWEDT Pump failure C-ATC VFD Cell Failure 3 RCO53F C-CRS (TS)(AOP) 4 R-ATC Power maneuver C-BOP NSW Pump 2B Trip (failure of standby to start) 5 CWOI9F C-CRS (TS)(AOP) C-BOP CWIP Trip 6 CWO39F C-CRS (AOP) M RWCU leak I Scram 7 RWO13F C SBGT Fails to start (AOP)(RSP)(SCCP) M ED 8 K1507A C Failure of 2 ADS valves to open (EDP) (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Event Event Description Summary Perform PT-40.2.1 1, Main Generator Voltage Regulator Manual And Automatic 1 Operational Check. Annunciator A-04 1-1, Drywell Equip Drain Sump Lvl Hi, will annunciate and the 2 sumps will not auto start. One of the sump pumps will need to be manually started A power cell in VFD A will fail. Recirc Pump 2A speed will lower and a speed hold will initiate. Loop flows will be outside mismatch limits. 4 The crew will reset the speed hold and match loop flows. NSW Pump B will trip and the crew will start NSW Pump A. Since 1A NSW Pump is 5 out of service, Tech Specs will apply. Crew will enter 0AOP-1 8.0, Nuclear Service Water System failure, and carry out appropriate actions. Circulating Water Pump 2A will trip on motor winding fault, and the standby 6 Circulating Water Intake Pump will be started. OAOP-37.0 will be entered due to lowering vacuum. A large un-isolable RWCU leak will occur. Crew will enter AOP-5.0 and SCCP. The 7 CRS should direct a SCRAM. SBGT will fail to auto start and should be manually started. Secondary containment conditions will worsen, forcing the CRS to direct an Emergency Depressurization (or Anticipation of Emergency Depressurization) due to 8 high water levels. If Anticipation is performed, the second area high water level will annunciate requiring the emergency depressurization. Two ADS SRVs will fail to manually open. The CRS should direct opening two additional SRVs.

Appendix D Scenario Outline Form ES-D-1 Facility: Brunswick Scenario No.: 4 Op-Test No.: 2016 Draft Examiners: Operators: Initial Conditions: Unit Two is at 100% power. 1A NSW Pump is under clearance for planned maintenance. 2C ICC Pump is aligned to Unit One. Turnover The BOP will start CREV in the area high radiation mode for inspection testing lAW OOP-37, Section 6.1.3. Event MaIf. No. Event Event No. Type* Description 1 N-BOP Manual start of CREV in area high radiation mode. C-ATC C32-NOO4A Fails High 2 NBOO7F (TS) EEO3OM- C-BOP MCC 2TD trip / Standby Stator Water Cooling Pump fails to auto 2TD C-CRS start C-ATC RCIC steam leak 4 ESO25F C-CRS (AOP)(TS) C-BOP TCC Pump Failure 5 K4516A C-CRS (AOP) 6 R-ATC Power Reduction M Loss of Off-Site Power / Scram 7 EEOO9F C DG3 failure to auto start / DG4 Duff 0/C (RSP)(PCCP)(AOP) C SRV Failure I Tailpipe Break / DW Spray Logic Failure ESOO4F 8 M ED on PSP CAO2OF (AOP)(EDP) (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Event Event Description Summary 1 The BOP will start CREV in the area high radiation mode lAW 20P-37. After CREV is started, C32-LT-NOO4A will fail high. The crew will reference Tech 2 Spec 3.3.2.2 and determine a 7 day LCO exists to place the failed channel in the tripped condition. The crew should select level B per OP-32. MCC 2TD will trip and the standby stator cooling water pump will fail to auto start. The standby stator cooling water pump can be manually started. The 2D air compressor will also be lost and OAOP-20.0 may be entered. Unit One may be contacted to place the 1 D Air Compressor in lead. A break in the RCIC steam line in the south RHR room will occur. The break can be 4 isolated by closing either the E51-F007 or the E51-F008. The crew will respond to the steam leak lAW AOP-05.0. TBCCW Pump 2B will trip and TBCCW low header pressure will alarm. The crew will 5 respond per OAOP-1 7.0. TBCCW pressure will recover and actions for partial loss of TBCCW will be performed. 6 A power reduction will be required lAW AOP-17.0. A Loss of Offsite Power will occur. The crew will respond per OAOP-36. 1. DG3 will 7 fail to auto start and DG4 will trip on Duff 0/C. The BOP operator will start DG3 to energize bus E3 and perform OAOP-36.1. SRV F will fail open. AOP-30 will be entered. The SRV will not reset using the control switch. Pulling fuses lAW AOP-30 results in loss of indication but the SRV 8 remains open. SRV F tailpipe will rupture, pressurizing containment. The DW Spray logic (think switch) will fail causing an inability to spray the torus or drywell. Emergency Depressurization is required when PSP is violated.

Appendix D Scenario Outline Form ES-D-1 Facility: Brunswick Scenario No.: 5 Op-Test No.: 2016 Draft Examiners: Operators: Initial Conditions: A2X sequence at step 166. Core Spray Loop B under clearance, remaining ECCS LP systems protected. Permission for continuous withdrawal has been granted for the rods going from 12-48. IRM A was bypassed due to spiking and the paperwork is being evaluated by the WCC SRO for its return to service. Turnover BOP OP complete Step 6.3.46 of OGP-02, Approach to Criticality and Pressurization of the Reactor. lAW the reactivity plan the ATC operator is to raise power to 6-10% using rods. Event Mall. Event Event No. No. Type* Description 1 N-BOP Place 2A RFPT level control in automatic 2 R-ATC Raise reactor power using control rods C-ATC Difficult to move control rod 3 RDO32M C-CRS (AOP) C-BOP 4 K45100 Steam Packing Exhauster Trip CCRS C-ATC IRM Failure 5 NIO18F C-CRS (TS) ED lAD C-BOP DG3 I E3 I E7 control Power loss 6 CJ6 C-CRS (AOP)fTS) M Lowering Torus Level I 7 CAOO2F C RHR F028A mech trip I RHR F024B thermal trip I CS FO2OA broke (PCCP) Scram I Emergency Depressurization 8 RPOO8F M (RSP)(ATWS)(EDP) (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Event Event Description Summary OGP-02, Approach to Criticality and Pressurizations of the Reactor will be completed 1 starting at Step 6.3.46 to place the REP master controller into Auto. The crew will raise power by pulling control rods in preparation for placing the Mode 2 switch to RUN. Rod pulls will commence at Step 161 (42-39 @ 12) of the A2X sequence. Control rods will continue to be withdrawn raising power. When control rod 42-23 is selected for withdrawal, it will be stuck at position 12. AOP-02 may be entered and 20P-07, Section 8.2 actions are required to withdraw a difficult intermediate control rod. SPE 2A will trip causing a loss of gland sealing header pressure. SPE 2B will be placed in service While withdrawing control rods, IRM C will fail upscale causing a rod block and half scram. SRO will address IRM A and C inoperability lAW TS 3.3.1.1. Once 5 addressed, l&C will report IRM A is ready to be returned to service following proper channel check. The crew will take the actions of the APP and bypass IRM C and reset the half scram. DC Panel 2A will trip resulting in loss of control power to DG 3, Bus E3 and Bus E7. The crew will respond per AOP-39.O and transfer the control power to alternate. DG 6 3, Bus E3 and Bus E7 are inoperable until transferred to alternate supply. Once control power is transferred, a 7 day action is required to restore to the normal source. The BOP operator will return DG 3 to AUTO lAW AOP-39.O. Torus level will begin to lower due to an unisolable leak on RHR suction. If attempted to raise torus water level, on RHR A loop the E11-F028A (Torus Discharge Isol Vlv) 7 will trip when opened, on RHR B loop the El l-F024B (Torus Cooling Isol Vlv) will thermal trip when opened, and on Core Spray the E21-FOO2A (Core Spray Pump A Suction Valve From The Condensate Storage Tank) handwheel will be broke. Before level reaches -5.5 feet in the torus a reactor scram is required. When torus 8 water level reaches -5.5 feet emergency depressurization is required. The crew can anticipate emergency depressurization.

(_~ DUKE ENERGY. BRUNSWICK TRAINING SECTION OPERATIONS TRAINING INITIAL LICENSED OPERATOR SIMULATOR EVALUATION GUIDE 2016 NRC SCENARIO 1 SWAP CBP, HPCI INITIATION, OFFGAS FILTER DP, RR PP TRIP, HDD CONT FAILURE, FUEL FAILURE, SCRAM FAILURE, SPC VLV FAILURE REVISION 0 Developer: ~od ~otue Date: 07/07/2016 Technical Review: Va1e ?I~ Date: 9/12/2016 Validators: 'D~ 1flo£6. Date: 09/06/16 s~~

              ~~

Facility Representative: c::k- o/,11er q- Date: O'i/ zi/ro1~ '---*---*---------------------------~

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 2 of 48 REVISION

SUMMARY

.____o_J Scenario developed for 2016 NRC Exam.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 3 of 48 TABLE OF CONTENTS 1.0 SCENARIO OUTLINE .................................................................................................................. 4 2.0 SCENARIO DESCRIPTION

SUMMARY

................................................................................... 5 3.0  CREW CRITICAL TASKS ............................................................................................................ 6 4.0  TERMINATION CRITERIA .......................................................................................................... 6 5.0  IMPLEMENTING REFERENCES .............................................................................................. 7 6.0  SETUP INSTRUCTIONS ............................................................................................................. 8 7.0  INTERVENTIONS ....................................................................................................................... 10 8.0  OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES .................................... 12 ATTACHMENT 1 - Scenario Quantitative Attribute Assessment ..................................................... 43 ATTACHMENT 2 - Shift Turnover ........................................................................................................ 48

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 4 of 48 1.0 SCENARIO OUTLINE Event Malf. No. Type* Event Description 1 N-BOP Swap Condensate Booster Pumps C-ATC Inadvertent HPCI Initiation w/ failure to trip 2 ES014F C-CRS (TS}(AOP-03.0) C-BOP 3 ZUA343 Off Gas Filter Hi d/P C-CRS C-ATC Recirc Pump Trip 4 RC055D C-CRS (TS}(AOP-04.0) 5 R-ATC Reduce reactor power for Single Loop Operation C-BOP Heater Drain Deaerator Controller Failure 6 CF039F C-CRS (AOP-23.0) Fuel Failure 7 NB005F M (RSP)(AOP-05.4) K2501A Manual Scram Failure - Alternate Rod Insertion 8 K2503A c Close Group I Valves Torus cooling Valve Failure 9 RH020F c (PCCP)

                *(N)ormal, (R)eactivity, (C)omponent or Instrument,       (M)ajor

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 5 of 48

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2.0 SCENARIO DESCRIPTION

SUMMARY

Event Description The Unit will be operating at rated power and after taking the shift turnover will swap 1 condensate booster pumps IAW 20P-32. (Start 2C, secure 2A) After the condensate booster pumps have been swapped, an inadvertent HPCI 2 initiation will occur. The HPCI system will not trip but can be manually isolated . AOP-03.0 and Tech Specs will be entered . After Tech Specs are addressed the Off Gas Filter Hi d/P alarm will be received . The 3 standby Off Gas filter will be placed in service. VFD coolant leakage will occur that will cause a trip of the 28 Recirc pump. The 4 supply breaker does not trip and must be opened manually. OAOP-04.0 will be entered . Technical Specifications will be addressed .. The plant will be greater than the allowable for single loop operation. Recirc flow must be lowered and/or control rods must be inserted to reduce power. With only one recirc 5 pump in operation reactor power must be less than 50% and core flow must be greater than 30.8 Mlbs but less than 45 Mlbs. After the power reduction the Heater Drain Deaerator controller will fail. AOP-23 .0 6 will be entered. Fuel failure will begin requiring entry into AOP-05.4. When Process Off gas Hi Hi 7 alarm with the Main Stack rising a scram will be inserted and Group I valves closed . The crew will be required to (1) Scram, (2) Close Group I valves. (Each of which are 8 critical tasks) The manual scram pushbuttons will not work, Alternate Rod Insertion (ARI) will be initiated to scram the rods. 9 Torus cooling valve will trip on thermals (will reset if attempted).

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 6 of 48 3.0 CREW CRITICAL TASKS Critical Task #1 Insert control rods with ARI IAW the Scram hard Card or LPC when Process Off gas Hi Hi annunciator is received and Main Stack Rad is rising. Critical Task #2 Close Group 1 isolation valves when Process Off gas Hi Hi annunciator is received and Main Stack Rad is rising . 4.0 TERMINATION CRITERIA When all control rods are inserted and suppression pool cooling is being placed in service (F024A valve failure recognized) the scenario may be terminated.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. O Page 7 of 48 5.0 IMPLEMENTING REFERENCES NOTE: Refer to the most current revision of each Implementing Reference. Number Title A-01 (1-4) HPCI LOW FLOW A-01 (3-5) HPCI ISOLATION TRIP SIG A INITIATED f-- -- A-01 (4-4) HPCI SYS PRESS LO 2APP-UA-03, 5-3 OFFGAS STBY FILTER DI FF-HIGH 2APP-A-07, 2-3 RECIRC VFD B ALARM UNACK 2APP-A-07, 3-3 RECIRC VFD B ALARM 2APP-A-07, 3-4 RECIRC VFD B COOLING SYS TROUBLE 2APP-A-07, 4-3 RECIRC VFD B TRIP WARNING 2APP-A-07, 4-6 RECIRC LOOP B ONLY OUT OF SERVICE 2APP-A-07, 5-3 RECIRC VFD B TRIPPED 2APP-A-07, 5-5 PUMP B SEAL STAGING FLOW HI/LO 2APP-A-05, 4-8 OPRM TRIP ENABLED 2APP-UA-23, 4-4 E REHEATER FIRST STAGE LEVEL HI-LO 2APP-UA-23, 4-5 W REHEATER FIRST STAGE LEVEL HI-LO 20P-30 Condenser Air Removal And Off-Gas Recombiner System I 2AOP-04 Low Core Flow 2APP-UA-04, 2-10 HD DEAERATOR LEVEL HIGH-LOW. 2APP-UA-03, 4-6 AREA RAD RADWASTE BLDG HIGH 2APP-UA-03, 5-2 PROCESS OFF-GAS RAD HIGH 2APP-UA-03, 5-7 AREA RAD TURBINE BLDG HIGH J

                                                                                       ~

2APP-UA-23, 2-6 MAIN STEAM LINE RAD HI 2APP-UA-23, 3-6 I MAIN STEAM LINE RAD Hl-Hl/INOP OAOP-05.4 I Radiological Release 2EOP-01-RSP I Reactor Scram Procedure I 2EOP-01-RVCP Reactor Vessel Control Procedure l i OEOP-02-PCCP Primary Containment Control Procedure I OEOP-04-RRCP Radioactivity Release Control Procedure I

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 8 of 48 6.0 SETUP INSTRUCTIONS

1. PERFORM TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 5, Checklist for Simulator Exam Security.

2. RESET the Simulator to IC-11.

3. ENSURE the RWM is set up as required for the selected IC.

4. ENSURE appropriate keys have blanks in switches.

5. RESET alarms on SJAE, MSL, and RWM NUMACs.

6. ENSURE no rods are bypassed in the RWM.

7. PLACE all SPDS displays to the Critical Plant Variable display (#100).

8. ENSURE hard cards and flow charts are cleaned up

9. TAKE the SIMULATOR OUT OF FREEZE

10. LOAD Scenario File.

11. ALIGN the plant as follows:

Manipulation

1. Ensure 2C TCC Pump is in service on Unit One.

2. Bypass APRM 2

12. IF desired, take a SNAPSHOT and save into an available IC for later use.

13. PLACE a clearance on the following equipment.

I Component Position APRM2 Blue Tag

14. INSTALL Protected Equipment signage and UPDATE RTGB placard as follows:

Protected Equipment

1. 2A and 2B NSW pumps

2. 2A FPC Pump/Hx, 2A RCC Pump, and 2C Demin Transfer Pump.

15. VERIFY OENP 24.5 Form 2 (Immediate Power Reduction Form) for IC-11 is in place.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 9 of 48

16. ENSURE each Implementing References listed in Section 7 is intact and free of marks.

17. ENSURE all materials in the table below are in place and marked-up to the step identified.

Required Materials 20P-32, Section 6.3.6 has been completed up to step 6.3.6.11

18. ADVANCE the recorders to prevent examinees from seeing relevant scenario details.

19. PROVIDE Shift Briefing sheet for the CRS.

20. VERIFY all actions contained in TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 4, Simulator Training Instructor Checklist, are complete.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. O Page 10 of 48 7.0 INTERVENTIONS TRIGGERS Trig Type ID 1 Remote Function EE_UTSHED4 - [UNIT TRIP LOAD SHED SEL SW, 2A COND BOOSTER PUMP] 1 Remote Function EE_LSHED4 - [LOCA LOAD SHED SEL SW, 2A CONDENSATE BOOSTER PUMP] 2 Remote Function EE_LSHED6 - [LOCA LOAD SHED SEL SW, 2C CONDENSATE BOOSTER PUMP] 2 Remote Function EE_UTSHED6 - [UNIT TRIP LOAD SHED SEL SW, 2C COND BOOSTER PUMP] 3 Malfunction ES014F - [INADVERTANT HPCI SYS INITIATION] 4 Annunciator ZUA343 - [OFF GAS FILTER DIFF-HIGH] 5 Malfunction RCOSSD - [VFD COOLING SYSTEM LEAKAGE] 6 Trigger Command DOD:Q2735RRH 7 Trigger Command DOD:Q2735LGH 8 Malfunction CF039F - [HTR ORN DEAER LVL CNTRLR FAILURE] 9 Malfunction NB005F - [FUEL FAILURE] 10 Malfunction NBOlOF - [GROSS FUEL FAILURE] 11 Remote Function Ml_IACBLRMl - [UNIT 1 CABLE SPREAD ROOM VENT FANS] 12 Remote Function Ml_ZVACS918_1- [UNIT 1 CB MECHANICAL EQUIP ROOM VENT FANS CS) 13 Remote Function CF_ZVCF5719 - [HWC-SV-5719 - H2 INJECTION OPEN VLV] 14 Remote Function CF_ZXCF5717 - [HWC-SV-5717 - H2 INJECTION CLOSE VLV] 15 Trigger Command and:zua343 Trig# Trigger Text 6 !K2735NWD - Not [2B RECIRC STOP] delete Q273SRRH 7 K2735BPH - [2B RECIRC STOP] delete Q2735LGH 15 QK214RRZ - [AOG SYS ISOL HCV-101 RED]

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 11of48 MALFUNCTIONS Malf Mult Current Target Rmp Description Actime Dactime Trig ID ID Value Value time ES014F INADVERTANT HPCI SYS INITIATION False True 3 RCOSSD VFD B VFD COOLING SYSTEM LEAKAGE 0.00 100.00 00:01:00 5 CF039F HTR ORN DEAER LVL CNTRLR FAILURE False True 8 NBOOSF FUEL FAILURE 0.00 100.00 00:10:00 9 NBOlOF GROSS FUEL FAILURE 0.00 100.00 00:10:00 10 RH020F Ell-F024B FULL FLOW

• VLV Ell-F024B True True RP006F MANUAL SCRAM DEFEAT True True N1032F APRM2 APRM FAILS LO True True REMOTES Current Target Rmp Remfld Multld Description Actime Trig Value Value time CC_IACW4518 2C TBCCW PUMP UNIT ALIGNMENT 1 1 LOCA LOAD SHED SEL SW, 2A CONDENSATE EE_LSHED4 DISABLE ENABLE 1 BOOSTER PUMP LOCA LOAD SHED SEL SW, 2C CONDENSATE EE_LSHED6 ENABLE DISABLE 2 BOOSTER PUMP UNIT TRIP LOAD SHED SEL SW, 2A COND EE_UTSHED4 DISABLE ENABLE 1 BOOSTER PUMP UNIT TRIP LOAD SHED SEL SW, 2C COND EE_UTSHED6 ENABLE DISABLE 2 BOOSTER PUMP Ml_IACBLRMl UNIT 1 CABLE SPREAD ROOM VENT FANS AUTO OFF 11 UNIT 1 CB MECHANICAL EQUIP ROOM VENT Ml_ZVACS918_1 NEUT STOP 12 FANS CS CF_ZVCF5719 HWC-SV-5719 - H2 INJECTION OPEN VLV NORMAL OPEN 13 CF_ZXCF5717 HWC-SV-5717 - H2 INJECTION CLOSE VLV NORMAL CLOSE 14 PANEL OVERRIDES Actu Position I Override Rmp Tag ID Description al Actlme Dactime Trig Target Value time Value KlCOBA REMOTE TURBINE TURB_TRIP OFF OFF Q2735RRH 2B RECIRC RED ON/OFF ON ON Q2735LGH 2B RECIRC GREEN ON/OFF OFF OFF ANNUNCIATORS Override Window Description Tagname OVal AVal Actime Dactime Trig Type 4-3 OFF GAS FILTER DIFF-HIGH ZUA343 ON ON OFF 4

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 12 of 48 8.0 OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES EVENT 1: SWAP CBPs Simulator Operator Actions Ensure Monitored Parameters is open and Scenario Based Testing Variables are loaded. Initiate Trigger 1 to ENABLE 2A CBP Unit Trip and LOCA Load Shed Selector Switches. Initiate Trigger 2 to DISABLE 2C CBP Unit Trip and LOCA Load Shed Selector Switches Initiate Trigger 13 to place HWCH-SV-5719 to open. Initiate Trigger 14 to place HWCH-SV-5717 to close. Simulator Operator Role Play Acknowledge request to enable 2A CBP Unit Trip and LOCA Load Shed selector switches, after Sim Operator activates the trigger report that the action is complete. Acknowledge request to disable 2C CBP Unit Trip and LOCA Load Shed selector switches, after Sim Operator activates the trigger report that the action is complete. Acknowledge request to place HWCH-SV-5719 (Condensate Booster Pump C H2 Injection Isolation Valve) in AUTO, have SIM OP initiate trigger 13 and report valve is open. Acknowledge request to place HWCH-SV-5717 (Condensate Booster Pump B H2 Injection Isolation Valve) in CLOSE and then report that the action is complete Acknowledge any requests for Radwaste actions. Acknowledge request to perform 20P-32, Attachment 10. Report that you will co-ordinate the performance of the attachment with the wee. Evaluator Notes Plant Response: 2C CBP is started and 2A CBP is secured . Objectives: SRO - Directs BOP to swap Condensate Booster Pumps BOP - Swap Condensate Booster Pumps RO - Monitors the plant Success Path: Condensate Booster Pumps are swapped Event Termination: When directed by the Lead Evaluator, go to Event 2.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 13 of 48 EVENT 1: SWAP CBPs Time Pos EXPECTED Operator Response NOTES SRO Conduct shift turnover shift briefing . Direct CBPs to be swapped. (20P-32, Section 6.3.6) May conduct a brief (see Enclosure 1 on page 43 for format) RO Monitors the plant Swap Condensate Booster Pumps IAW 20P-32, Section 6.3.6: Make a PA announcement for starting 2C Condensate Booster Pump and Securing 2A Condensate Booster Pump. Should also check that Bus C is clear of personnel, may state that they would use plant cameras. Performs 20P-32, Section 6.3.6. Per the turnover, Steps 1-10 are complete. BOP Directs an AO to perform steps 12a, 12g, and 12h. Directs Radwaste to perform steps 21 and 22. Either directs an AO or the WCC SRO to perform Attachment 10 of 20P-32.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 14 of 48 _ CONDENSATE AND FEEDWAT ER SYSTEM 20P-32 OPERATING PROCEDURE Rev. 206 Page 133 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

9. !E 2B Condensate Booster Pump is to be started.

THEN perform th e fo llowing:................................................................. _ WA

a. Open COD-V5016 (Condensate Booster Pump B Inboard Mechanical Seal Vent Valve)........................ .................... .... ...... NlA _

b. WHEN a continuous solid stream of water has been observed for at least 2 minutes.

THEN close COD-V50 16 (Condensate Booster Pump B Inboard Mechanical Seal Vent Valve) ..... .. ....... ....... ........ ... ......... _ WA__

c. Open COO-V5017 (Condensate Booster Pump B Outboard Mechanical Seal Vent Valve) ................................... .... ....... ........ _ NIA__

d. WHEN a continuous sohd stream of water has been observed for at least 2 minutes, THEN close COD-V50 17 (Condensate Booster Pump B Outboard Mechanical Seal Vent Valve) ........ .......... ........ ..... ....... _ _NfA__

10. !E 2C Condensate Booster Pump is to be started .

THEN perform the following: ..................... .. ... .... ...................... ........... .. _ _RO__

a. Open COD-V5018 (Condensate Booster Pump C Inboard Mechanical Seal Vent Valve) .. ..... ........................ .. .... ... .............. __AO _

b. WHEN a continuous solid stream of water has been observed for at least 2 minutes.

THEN close COO-V50 18 (Condensate Booster Pump C Inboard Mechanical Seal Vent Valve) . ..................... .......... ......... _ AO _

c. Open COD-V5019 (Condensate Booster Pump C Outboard Mechanical Seal Vent Valve) . .... ...... ........................... ......... ....... _ _p.o _

d. WHEN a continuous sohd stream of water has been observed fo r at least 2 minutes, THEN close COD-V50 19 (Condensate Booster Pump C Outboard Mechanical Seal Vent Valve) ............... ........ ............... __ AO _

11 . Note status of the following alarms:

• UA-13 . 1-1, Gen-XfmrPn maryLIO Unit Trip ................... ... ... ...... _
• UA-13. 1-2. Generator 0 1ff LIO Unit Trip ...... .......... ....... ..... ........ _
• UA-13 . 1-3. Gen-Xfmr Backup LIO Unit Trip ................ ..... ....... .. _

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 15 of 48 CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev. 206 Page 134 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

12. IF all the alarm s m Step 11 are CLEAR.

THEN perform the follow1ng:[8.7.5] ** - --********************************* ********* ****** -

a. Place the fo llowing switches in ENABLED for the off going condensate booster pump:

• Unit Tnp Load Shed Selector Switch .......... ..................... _
• LOCA Load Shed Selector Switch ................................... .

b. Ensure the onoommg oondensate booster pump mode selector switch in MAN .................................. .. ..... .................... ..

c. Confinn oncormng con densate pump discharge valve closes . rf OPEN :

• COD-V4 (Condensate Booster Pump A Disch Viv) .. ........ _
• COD-VS (Condensate B ooster Pump B Disch Viv) .......... _
• COD-V6 (Condensate Booster Pum p C Disch Viv) .... ...... _

NOTE The start/stop con trol switch for a Condensate Booster Pump must be held in START until the recircu lati on valve is full OP EN. The recirculation valve does NOT stroke OPEN until all other condensate booster pump perm issives are met. While opening th e recirc valve . condensate booster pump suction pressure will lower as add itiona l flow up to 2000 gpm is rou ted to the con denser. .......... ........................... .... ... ... .. D

d. St art the oncoming condensate booster pum p..... ...................... _

e Confinn oncoming condensate booster pump discharge valve opens :

• COD-V4 (Condensate Booster Pum p A Disch Viv) .... .... .
• COD-VS (Condensate Booster Pump B Disch Viv) ......... .
• COD-V6 (Condensate Booster Pump C Disch Viv} ......... .

2016 NRG SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 16 of 48 CONDENSATE AND FEEDWATER SYSTE M 20P-32 OPE RATI NG PROCEDURE Rev. 206 Page 135 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

f. WHEN condensate booster pump discharge pressure stabilizes .

THEN perform the following: ..... ................................................ . (1) Stop off going condensate booster pump ........................ _ (2) IF B2 1-F032A (Feedwa ter lsol Viv) AND B2 1-F032B (Fecdwater lsol Viv) am OPEN . THEN plac e the off going condensate booster pump mode switch in AUTO ............... ........................................

g. Place the following switches in DISABLED for oncoming condensate booster pump :

• Unit Tnp Load Shed Selector Switch ... ............................ _
• LOCA Load Shed Selector Switch ................ ................ ... .

h. IF HWC 1s in service.

THEN perform the fol lowing : ...................................... ................ _ (1) Place HWC condensate H2 injection isolation solenoid valve . at HWC local panel H2J. for oncoming condensate booster pump. 1n AUTO:

• HWCH-SV-57 17 (Condensate Booster Pump A H2 Injection Isolation Valve) ....... . .............. ... . .. .
• HWCH-SV-57 18 (Condensate Booster Pump B H2 Injection Isolation Valve) ....... ............................ _
• HWCH-SV-57 19 (Condensate Booster Pump C H2 Injection Isolation Valve) ............................... _

(2 ) Confirm HWC condensate H2 1niect1on 1solat1on solenoid va lve . at HWC local panel H2J . for oncom ing condensate booster pum p. 1s OPEN.

• HWCH-SV-57 17 (Condensate Booster Pump A H2 ln1ect1on Isolation Valve) .... ....... .... .... ......... .
• HWCH-SV-57 18 (Condensate Booster Pump B H2 Injection Isolation Valve) ................. . .... .. .... _
• HWCH-SV-57 19 (Condensate Booster Pump C H2 ln1cct1on Isolation Valve) ......................... .

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 17 of 48 CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev. 206 Page 136 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued) (3 ) Place the HWC condensate H2 injection isolation solenoid va lve . at HWC local panel H2J. for off going condensate booster pump in CLOSED.

• HWCH-SV-57 17 (Condensate Booster Pump A H2 Injection Isolation Valve } ................................ ... _
• HWCH-SV-5718 (Condensate Booster Pump B H2 Injection Isolation Valve) ..................................._
• HWCH-SV-57 19 (Condensate Booster Pump C H2 Injection Isolation Valve) ****************- *** **************- ___ _

i. Go to Step 20 . .............................................. .............................. _ --* _

13. Place the fo llowin g switches in DISABLED for the oncoming Condensate Booster Pump :

• Unit Trip Load Shed Selector Switch ... ................... ...... ......... ...... _ _NIA_ _
• LOCA Load Shed Selector Switch ..... ...... ... ......................... .... ... _ N/A _

14. Place the oncom ing condensate booster pump mode selector svvitch in MAN . .... .............. ... ............. ................ ....... ....... ..................... .. _ .NIA

15. Confirm oncoming condensate booster pump discharge valve doses.

• COD-V4 (Condensa te Booster Pump A Disch Viv) ....... .............. _ NIA
• COD-V5 (Condensate Booster Pump B Disch Viv) ..... ...... ... ...... . _ NIB,__
• COD- V6 (Condensate Booster Pump C Disch Viv) ............. .... ... .

NOTE The start/stop con trol switch for a condensate booster pump must be held in START until the recirculation va lve is full OP EN. The recirculation valve does NOT stroke OPEN until all other condensate booster pump perm1ssives are met. While opening the recirc valve . condensate booster pump suction pressure will lower as add itional flow up to 2000 gpm is ro uted to the condenser. ....... .... ................... ... ... ...... ...... . ISl'A 16 . Start oncoming condensa te booster pump ..... ..... .................... .. .. ....... ... .NIA__

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 18 of 48 - ---~ -- - - CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev. 206 Page 137 of 408 6.3.6 Transferring to Standby Condensat e Booster Pump (continued)

17. Confirm oncoming condensate lx>oster pump discharge valve opens:

• COD-V4 (Condensate Booster Pump A Disch Viv) NIA
• COD-VS (Condensate Booster Pump B Disch Viv). .............. ... ... _NlA
• COD-VG (Condensate Booster Pump C Disch Viv) ........ ..... ....... _ NIA

18. WHEN condensate booster pum p discharge pressure stab1llzcs ,

THEN perform the fo llowing* .............................. ........ ................ .

a. Stop off going condensate lx>oster pump .. ................................. _ NIA

b. IF B21-F032A (Feedwa ter lsol Viv) ANO B2 1-F032B

( ecdwater lsol Viv) are OPEN. THEN place the off going conden sate lx>oster pump mode switch in AUTO ..... ... ............. .............................................. _ NIA

c. Place the following sW1tches i n ENABLED for the off going condensate lx>oster pump:

• Unit Tnp Load Shed Selector Switch .. .............................. .NLA
• LOCA Load Shed Selector Switch ................................... .NJA _

t 9. !E HWC 1s in service . THEN perform the following:.......................... .............................. .... ... _NIA_

a. Place oncornrng condensa te booster pu mp. HWC condensate H2 in1ecbon 1solat1on solenoid valve . at HWC loca l panel H2J, in AUTO:

• HWCH-SV-5717 (Conden sate Booster Pump A H2 ln1ect1on Isolation Valve) ................... .......... .................... _ NIA _
• HWCH-SV-57 18 (Condensa te Booster Pump B H2 ln1ect1on Isolation Va lve)... ... .. .......................... ............... NIA
• HWCH-SV-57 19 (Conden sa te Booster Pump C H2 ln1cct1on Isolation Valve)..... .. ... ... ... ......... .. ... .. ............... NLA

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 19 of 48 CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATI NG PROCEDURE Rev. 206 Page 138 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

b. Confirm oncormng condensate booster pump. HWC condensate H2 injection isolation solenoid valve . at HWC local panel H2J. OPE N:

• HWCH-SV-5717 (Condensa te Booster Pum p A H2 Injection Isolation Valve)................................................. .. _ NfA
• HWCH-SV-5 718 (Condensate Booster Pum p B H2 Injection Isolation Valve)................................................... _ NJA _
• HWCH-SV-5719 (Condensate Booster Pump C H2 Injection Isolation Valve) ................................................... _ .NfA __

c. Place off going condensate booster pump. HWC condensate H2 injection isolation solenoid valve . at HWC local panel H2J.

in CLOSED:

• HWCH-SV-5717 (Condensate Booster Pum p A H2 Injection Isolation Valve) ..................... ....... .................... ... _ Joj/_A _
• HWCH-SV-57 18 (Condensate Booster Pum p B H2 Injection Isolation Va lve) ........... .................. ................ ...... _ .N/A _
• HWCH-SV-5 719 (Condensa te Booster Pum p C H2 Injection Isolation Va lve) ..... ............................. .... ............. _ N/A __

20. !E three condensate pumps are in service per Step 5.b(2 ) .

THEN perform Section 6.3.23. Securing from Three Condensate Pum p Operati on .. ........................... ............ ............................. .. ............. _ 2 1. Direct Radwaste Operator to remove the additiona l CFO or COD placed in service in Step 5.a ...................... .................................... ........ _ 22 . Direct Radwaste Operator to monitor COD effluent conductivity for each dem ineralizer in service .. ........ ............. ......................... ............ ... . 23 . !E any condensate pump Unit Trip Load Shed Selector Switch AND LOCA Load Shed Selector SwHch was manipulated . THEN complete Attachment 9, Condensate Pump Unit Trip Load Shed/LOCA Load Shed Selector Switch Alignment Documentation ... ... _ 24 . Complete Attachment 1O. Condensa te Booster Pum p Unit Trip Load Shed/LOCA Load Shed Selector Switch Alignment Documentation . ... .... ................ .. ...... ... ...... ........... .............. .... ....... ......... .

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 20 of 48 CONDENSATE AND FEEDWATER SYSTEM 20 P-32 OPERATING PROCEDURE Rev. 206 Page 406 of 408 ATTACHMENT 10 Page 1 of 1 Condensate Booster Pump Unit Trip Load Shed/LOCA Load Shed Selector Switch Alignment Documentation NOTE

• When performing this attachment. pum p selector switches. for pump(s) which have NOT been rea ligned may be marked NA. ............................................................. D
• The rondcnsate booster pump selected for standby operation (Mode Selector Switch in AUTO) i s require d to have both the LOCA l oad Shed Selector Switch and Unit Trip l oad Shed Selector Switch in ENABLED ............... .. .. .............................. D

1. Circle the required selector swi tch position.

Position/ Number Descnption Indication Checked Verified fNote 11 Turb ine Buildino - 4 160V SwitchQear Bus 2D - Elev 20 ft 2B Condensate Booster Pump Unit Trip l oad ENABLED/ AD9 Shed Se lector Switch DISABLED 2B Condensate Booster Pump LOCA Load Shed ENABLED/ AD9 Selector Switch DISABLED Turbine Building - 4 160V SwitchQear Bus 2C - Elev 20 ft 2A Condensate Booster Pump Unit Tri p Load ENABLED/ AC 1 Shed Selector Switch DISABL ED 2A Condensate Booster Pump LOCA Load Shed ENABLED/ AC 1 Se lector Switch DISABL ED 2C Condensate Booster Pump Unit Trip Load ENABLED/ AC2 Shed Selector Switch DIS ABLED 2C Condensate Booster Pump LOCA Load Shed ENABLED/ AC2 Selector Switch DISABL ED

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 21 of 48 EVENT 2: HPCI INITIATION Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 3 to initiate HPCI. After isolated, delete HPCI Inadvertent Initiation. Simulator Operator Role Play If contacted as Reactor Engineer to look at thermal limits due to HPCI injection, report that he will evaluate and monitor If contacted as l&C to assist with troubleshooting, after HPCI is isolated and Tech Spec are addressed, remove the HPCI initiation and trip failures and then report that l/C found a relay failure that caused the initiation signal. It will take about 4 hours to replace the relay but the initiation signal is now clear. The trip pushbutton had a loose wire, which has been re-attached . If contacted as chemistry acknowledge request for coolant samples for indications of fuel failure. Evaluator Notes Plant Response: HPCI will inadvertently initiate. The crew will verify level and then secure HPCI. HPCI manual isolation pushbutton will fail. If injection occurs, the crew will enter AOP-03. Technical Specifications will be addressed. Objectives: SRO - Direct actions in response to an inadvertent HPCI initiation and potential positive reactivity addition Determine actions required for LCO per Technical Specifications RO- Respond to an inadvertent HPCI initiation and potential positive reactivity addition Success Path: Verify HPCI initiation signal not present and isolate HPCI Event Termination: Go to Event 3 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 22 of 48 EVENT 2: HPCI INITIATION Time Pos EXPECTED Operator Response Comments Direct crew to trip HPCI following verification of SRO false initiation Direct crew to isolate HPCI on failure of trip SRO pushbutton (May use isolation pushbutton or direct steam supply valves to be closed). Direct crew to enter and execute OAOP-3.0 SRO Positive Reactivity Addition , if injection has occurred. Contact maintenance to look at the HPCI Initiation signal. SRO May also contact Reactor Engineer to look at thermal limits. Evaluate Tech Spec 3.5.1 ECCS - Operating SRO Condition 01 - Verify RCIC is OPERABLE Condition 02 - Restore HPCI in 14 days Direct HPCI shutdown IAW 20P-19 after l/C SRO confirms signal has cleared. SRO May direct RCIC and ADS to be protected. May conduct a brief SRO (see Enclosure 1 on page 43 for format) Monitor reactor plant parameters during BOP evolution

2016 NRG SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 23 of 48 EVENT 2: HPCI INITIATION Time Pos EXPECTED Operator Response Comments Verify false HPCI initiation signal (No LL2 signal ATC present or high drywell pressure) Trip HPCI by pushing the HPCI trip pushbutton, ATC recognize failure of trip May depress Manual Isolation System A pushbutton to isolate HPCI. OR ATC May isolate the steam supply valves (E41-F002 and F003) to HPCI to isolate system. Enter and execute AOP-3.0 Positive Reactivity ATC Addition Respond to the following A-01 alarms: 1-4, HPCI LOW FLOW ATC 3-5, HPCI ISOLATION TRIP SIG A INITIATED 4-4, HPCI SYS PRESS LO ATC Perform 20P-19 Section 6.2 to shutdown HPCI.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 24 of 48 HIGH PRESSURE COOLANT INJECTION SYSTEM 20P-19 OPERATING PROCEDURE Rev. 140 Page 27 of 120 6.2 Shutdown 6.2.1 HPCI System Shutdown NOTE HPCI runl'lrg in A uto or Manual is a n R2 Reactivity Management Ei.ulution ........................ D

1. Confirm either Step 1 a or Step 1.b of the following initial conditions are met

a. !E the HPCI System has automatically initiated, THEN confirm one of the fo llowing eXists: ..................................... _ __

• llle system is NO longer reqlired to maintain reactor water level.. ...................................................................... ..... _ __
• The automati c initiatmn sigrol is HQ! valid ........................ _ __

b. IF the HPCI System was manually started, THEN confirm both of the follo'M ng eXists: .................................... _ __

• The system is NO longer reqlired to maintain reactor water level. .............................. ........................................... _ __
• The system is .t!Q longer reqlired to maintain reactor pressure ........................ ............ ..... ...... .... ....... .. ........ _ __

2. !E HPCI automatically initiated, Il:!fili perform the foUoW'ing : .................................................................... _ __

a. Confirm rrom at least two independent indications at least one of the following conditions exist: .............................................. _ __

• Adequate core cooling is ensured ........................................ _ __
• The initiation signal was NOT valid .......... ........... . .... .
• HPCI is NOT fUnctiorvrg property in the A UTOMATIC mode ..................... ......................... .. .. . .....

b. Place E4 1-S21 (Vacuum Pump ) control switch in START. ...... _ __

c Depress E4 1-S17 (Initiation Signal/Reset) pustbutton ........... _ __

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 25 of 48

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 26 of 48 HIGH PRESSVRE COOLANT INJECTION SYSTEM 20P-19 OPERATING PROCEDURE Rev. 140 Page 28 of 120 6.2.1 HPCI System Shutdown (continued) NOTE Operation of the HPCI S'f.)tem lllder miniml.lll now conditions is minimized . Section 6.2.1 Step 3 through Section 6.2.1 Step 11 are performed as expeditiously as possible. {8.12} ............................................................................................................................ D

3. Ensure E41-F006 (HPCI Injection Viv) is CLOSED .................................. _ __

4. Ensure E41-F008 (Bypass To CST Vlv) is CLOSED . ***************************- - -

5. IF OPEN AliQ. NOT in use for RCIC operation.

THEN close E4 1-F01 1 (Redundant lsol To CST Viv) ............................ _ __

6. Ensure E41-F012 (Min Flow Bypass To Torus Viv) opens .................... _ __

NOTE f Turbine Trip pushbutton is released before E4 1-FOO1 (Tl.lbine Steam Supply Viv) is fu Oy closed . the tl.11.>i ne wiD attempt to restart ...... .... ................ .......................................... D

7. Close E41-FOO 1 (Turbine Steam Supply VIV) and immediately depress and hold Turbine Trip pushbutton until E4 1-FOO1 is fully CLOSED ............................................................................................... _ __

EN D R.M. LEVEL R2 Reactivity Evolution NOTE E4 *1-V8 (Turbine Stop Valve) closes while the Tl.11.>ine Trip pushbutton is depressed. However. wnen the Turbine Trip pustt>utton is released, E41-V8 will reopen until the HPCI oil pressure source is removed vmich will allOw spring pressure to close the va Ive .............................................................................................................. D

8. Ensure E4 1-V8 (Turbine Stop Valve ) closes ............................................ _ __

9 Ensure E4 1-S20 (Auxiliary Oil Pl.lllp) auto starts as the turbine speed !Owers ......................... .............................. ..................................... _ __

10. Close E4 1-F059 (C oo li~ Water S~plyVlv) ........................................... _ __

11 Ensure E4 1-F012 (Min FIOw Bypass To Torus Viv) closes .................... _ __ 12 Ensure HPCI E41-F IC-R600 (Flow Control) is in AUTO (A) ............... _ __

13. Adjust HPCI E4 1-FIC-R600 (Flow Control) setpoint to 4300 gpm

2016 NRG SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 27 of 48 HIGH PRESSl.RE COOLANT INJECTION SYSTEM 20P-19 OPERATING PROCEDURE Rev. 140 Page 29 of 120 6.2.1 HPCI System Shutdown (continued) NOTE

• Technical Specification 3.6.1.6.1 (MODE 1, 2, or 3) requires completion of OPT-02.3.1B. Suppression Pool to Drywell Vacwm Breaker Position Check, within 6 oours after arry discharge of steam to the ~pressi on chamber from arry source and within 6 hours following an operation that causes arry of the vacuum breakers to open......................................................................................................... D
• Section 62 .1 Step 14 ensures compliance with Technical Specifications . ...................... o

14. IF AT AN Y TIME in MODE *1. 2 , or 3, aner arry discharge of steam to the suppression chamber from arry source I!::!fil! ensure OPT-02.3.1B, S~pression Pool to Dry.veil Vacuun Breaker Position hdication Check. is completed within 6 hours.

[8.2 .1) ................................................................................................................._ __

15. WHEN *15 mirutes have elapsed after tripping the HPCI turbine.

THEN stop E4 1-S2 1 (Vacuum Pump )......................................................... _ __

16. Pl3ce E41-S21 (Vacuum Pump ) control switch in AUTO......................... _ __

17. WHEN differential temperature across the HPCI turbine bearings reduces to approximately 0°F, as indicated on E41-TR-R605, THEN stop E41-S20 (Auxiliary Oil Pump). *************************************************- - -

18. Pl3ce E4 1-S20 (Auxiliary Oil Pump) in AUTO ............................................_ __

19. Pl3ce E41 -S22 (Barometric Cndsr Condensate Pump) control switch in AUTO . ****************************************************** *****************************************- - -

20. Ensure E4 1-F025 (Coro Pump Disch otbd Drai n Viv) is OPEN ............ _ __

21 . Shut down Standby Gas Treatment System per 20P-10, Standby Gas Treatment System Operating Procedure. Section for Control Room Manua l Shutdown , aoo return to Section 6.2.1 Step 22 ............... _ __

22. Complete Section6 .1.1 to return the system to STANDBY. ................... _ __

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. O Page 28 of 48 EVENT 3: OFF GAS FILTER HI D/P Simulator Operator Actions At the direction of the lead evaluator, Initiate Trigger 4 to bring in Off-Gas Filter Diff-Hi annunciator Simulator Operator Role Play IF contacted as Outside AO to verify Off-gas filter Diff pressure, report local DP indication is reading 13 inches water. IF contacted as Unit One report steps 6.3.3.2 and 3 are complete. (1-0G-FV-244-4, 1-0G-FV-244-5, and 1-AOG-HCV-101 are closed) IF contacted as AO report 1-0G-CD-V7 is CLOSED (Step 4) IF contacted as AO report 2-0G-CD-V7 is OPEN (Step 5) IF contacted as Outside AO to verify Off-gas filter Diff pressure after filter swap, report local DP indication is reading 3 inches water. When contacted as AO report 2AOG-HCV-101 is in OPEN position. Evaluator Notes Plant Response: Off-Gas Filter Diff-Hi alarm annunciates. Objectives: SRO -Direct actions in response to a Off-gas Filter Diff-Hi alarm. BOP - Respond to clogged off-gas filter IAW APP and 20P-30. Success Path: Swap Off-gas filters per OP-30. Event Termination: Go to Event 4 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 29 of 48 EVENT 3: OFF GAS FILTER HI D/P I Time Pos EXPECTED Operator Response Comments Direct crew to perform the actions of UA-3, 4-3, SRO OFF GAS FILTER DI FF-HIGH alarm Direct crew to swap Off-gas filters per OP-30 Section 6.3.3. SRO Should identify that a PRR needs to be written for the APP (or verified that one has already been written) to reference the correct procedure section to swap off-gas filters (Section 6.3.3) May conduct a brief SRO (see Enclosure 1 on page 43 for format) ATC Monitors the plant. Respond to UA-3, 4-3, OFF GAS FILTER DIFF-HIGH alarm BOP Dispatch Outside AO to verify Off-gas filter Diff Hi locally Place Off-gas Standby filter in service per OP-BOP 30 Section 6.3.3

2016 NRG SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 30 of 48 CONDENSER AIR REMOVAL AND OFF-GAS 20P-30 RECOMBINER SYSTEM Rev. 98 Page 68 of 167 6.3.3 Placing Off Gas Standby Filter In Service

1. Ensure the following Initial Conditions are met:

a. Off Gas filter is in service. ********-** ******* ************-************* ******* ********- - -

b. Off Gas standby filter is available for operation ...... ............... ...... _ __

2. Ensure the following Unit 1 valves are CLOSED:

• 1-0G-FV-244-4 (OG stby Filt Inlet lsol Viv) ................................ _ _ _
• 1-0G-FV-244-5 (OG stby Filt Inlet ISOI Viv) *-********* **-* *****************- - -

3. Ensure 1-AOG-HCV-101 (Standby Off Gas Filter Outletlsolation Valve) is CLOSED. ***************--*****-*******-*** *-****-**************-****-******-************- --

4. Ensure 1-0G-CD-V7 (standby Filter Drain Valve) is CLOSED to prevent a cross-connect of Unit 1 and Unit 2 Off Gas Systems ............._ __

5. Open 2-0G-CD-V7 (Standby Filter Drain Valve). ...... .................. .... ......_ __

6. Open 2-0G-FV-244-4 and 2-0G-FV-244-5 (OG Stby Filt lnl Vlvs). *******- --

NOTE If 2-AOG-CS-316"1 (AOG Sys Viv Cont Sel SW) on Panel XU-80 is NOT in CENT, 2-AOG-HCV-10'1 (standby Off Gas Filter Outlet Isolation Valve) CANNOT be operated from the control Room ...***-***-**--*****-******-******** ....................... ............... ................ D

7. Open 2-AOG-HCV-101 (standby Off Gas Filter Outlet Isolation Valve).......... ... .. .. .............. ................. ................... .. ........ ... ........ ............. _ _ _

8. .!E. Off Gas filter is to be taken out of service.

THEN close 2-0G-FV-244-1, 2-0G-FY-244-2 and 2-0G-FV-244-3 (OG Filter lsol Vlvs) ............................................................. ..... ........ .... .. _ __

9. Ensure Off Gas standby filter pressure differential is less than 1o inches of water .................... ....... ... ................................. ...... ..... ........ _ __

10. At Local Control Panel H2E, place 2-AOG-HCV-101 (AOG System Isolation Valve) control switch to OPEN ........ ...................... ..... ....... ....... _ __

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev.a ~ Page 31 of 48 EVENT 4/5: 28 REACTOR RECIRC PUMP TRIP I EXIT SCRAM AVOIDANCE REGION Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 5, VFD B Coolant Leakage. Note: Short time delay before first alarm (1 .5 min), followed by pump trip. Malfunctions required: VFD B Coolant Leakage/ 4 kV breaker override Simulator Operator Role Play If contacted as the AO to investigate VFD Alarms, wait until Recirc pump has tripped and report that the coolant pumps are tripped due to coolant leakage. If contacted as reactor engineer, report you will monitor thermal limits and use OENP-24.5 rods as needed to get below MELLL line. If contacted as NIT to backup OPRM data acknowledge the request. If requested as RBAO to reduce Recirc purge flow, wait 3 minutes and report actions complete. If contacted as l/C, acknowledge report of VFD coolant leak. If contacted as chemistry, acknowledge request for coolant sample for indications of fuel failure. If asked the results of the previous sample are still be processed. Evaluator Notes Plant Response: A coolant leak develops on VFD B that will cause the Recirc pump to trip. The supply breaker does not trip and must be opened manually. 2AOP-04.0 will be entered. Technical Specifications will be addressed. The plant will be in the Scram Avoidance Region of the power to flow map. Recirc flow must be increased or control rods must be inserted to exit this region . With only one Recirc pump in operation reactor power must be less than 50% and core flow must be greater than 30.8 Mlbs but less than 45 Mlbs. Objectives: SRO - Direct Shift Response To Recirculation Pump Trip Per 2AOP-04.0 ATC: Respond to a Recirc Pump trip IAW 2AOP-04.0, Reduce reactor power. Success Path: Identifies that the Recirc supply breaker did not trip and manually trips the breaker. Manipulates reactor power to exit the scram avoidance region using rods and I or Recirc flow Event Termination: Go to Event 6 at the direction of the Lead Evaluator.

2016 NRG SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 32 of 48 EVENT 4/5: 28 REACTOR RECIRC PUMP TRIP I EXIT SCRAM AVOIDANCE REGION Time Pos EXPECTED Operator Response Comments SRO Direct response to Annunciators. SRO Direct entry into 2AOP-04.0, Low Core Flow Determine region of operation on power/flow SRO map (computer display 806 may be used for reference) Direct actions to reduce power to .:s,50% and Insert rods if greater than MELLL line. SRO (Flow must be maintained >30.8 but <45 Mlbs/hr and power must be <50% for single recirc pup operation) TS 3.4.1 Recirculation Loops Operating SRO Determine Condition A applies Required Action A.1: APLHGR limits and APRM setpoints must be adjusted within 6 hours May conduct a brief SRO (see Enclosure 1 on page 43 for format) BOP Plant Monitoring May determine region of operation on BOP power/flow map (computer display 806 may be used for reference) May monitor for THI BOP Check the VFD HMI screens, report loss of cooling.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 33 of 48 EVENT 4/5: 28 REACTOR RECIRC PUMP TRIP I EXIT SCRAM AVOIDANCE REGION Time Pos EXPECTED Operator Response Comments Respond to the following alarms during this event: A-07: 2-3, RECIRC VFD B ALARM UNACK 3-3, RECIRC VFD B ALARM 3-4, RECIRC VFD B COOLING SYS TROUBLE 4-3, RECIRC VFD B TRIP WARNING 4-6, RECIRC LOOP B ONLY OUT OF ATC SERVICE 5-3, RECIRC VFD B TRIPPED 5-5, PUMP B SEAL STAGING FLOW HI/LO A-05: 4-8 OPRM TRIP ENABLED UA-23: 4-4, E REHEATER FIRST STAGE LEVEL HI-LO 4-5, W REHEATER FIRST STAGE LEVEL HI-LO Observes indications on the VFD B HMI on Panel XU-4 Dispatch AO to investigate Diagnose and report supply breaker not tripped. IAW APP A-7, 5-3, Recirc VFD B Tripped and confirmation on XU-4, 2-B32-YFD-VDT-002B, Recirc VFD 2B identifies that both VFD coolant pumps are tripped, this indicates that the supply breaker should have tripped . Opens Recirc Pump B - 4 kV Supply Breaker Determine if valid core flow indication exists on process computer {WTCF).

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 ~ Page 34 of 48 EVENT 4/5: 28 REACTOR RECIRC PUMP TRIP I EXIT SCRAM AVOIDANCE REGION Time Pos EXPECTED Operator Response Comments Determines region of operation on power/flow map (computer display 806 may be used for reference) May do one or both of the following to lower power to less than 50%:

1. Insert control rods Turns select power on Selects a control rod from the ENP-24.5, Immediate Power Reduction sheet. (see page 35)

Drive the selected rod as specified for each group of rods as needed ..

2. Lower core flow using the running Recirc Pump Reduce CRD flow to 30 gpm per 20P-08.

If charging pressure high alarms, may request CRD Pump A discharge valve throttled closed per the APP Maintain core flow >30.8 E6 lb/hr to prevent excessive cooldown of idle loop. or If Core Flow is <30.8 E6 lb/hr log bottom head and loop temperature every 15 minutes Monitor for THI Notify Reactor Engineer Notify chemistry

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 _ £lage 35 of 48 FOR SIMULATOR USE ONLY IC-11 Revised 3-22-2015 FORM2 Page2of3 Immediate Reactor Power Reduction Instructions Sheet 1 of 2 (Control Rod Insertions ONL ~ CRS Control Correct Rod Control Rod Licensed Second Projected Rod Selected and Position Operator Licensed Power Verified* Operator Reduction (% CTP) 18-19 08 To 00 18-35 08 To 00 34-35 08 To 00 34-19 08 To 00 2% ... 26-11 36 To 00 10-27 36 To 00 26-43 36 To 00 42-27 36 To 00 18% ... 26-27 48 To 00 2% 'Y 18-11 48 To 00 18-43 48 To 00 34-43 48 To 00 34-11 48 To 00 13% ... 10-"19 48 To 00 10-35 48 To 00 42-35 48 To 00 42-19 48 To 00 8% 'Y

      *concurrent Verification of rod selection requi red prior to rod movement

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 36 of 48 EVENT 6: HEATER DRAIN DEAERATOR CONTROLLER FAILURE Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 8 to fail Heater Drain Controller. If directed to place controller in Manual or to swap master controllers, Delete CF039F. Simulator Operator Role Play If contacted as TBAO to investigate, report LC-91 is in master and is sending a full open signal. If asked by l&C to investigate controller failure, acknowledge the request. When HOD level is stabilized and if directed to place controller in Manual or to swap master controllers, have Sim Operator delete CF039F and report controller in manual maintaining level Evaluator Notes Plant Response: Heater Drain Tank Lowers Low level alarm at 32" Both Heater Drain pumps trip at 24" Condensate Booster Pump C auto start if power is not sufficiently reduced Objectives: Enter OAOP-23.0 May reduce power Control level using HD-57 Success Path: Manually control level in Heater Drain Tank using the HD-57 Event Termination: Go to Event 7 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 37 of 48 EVENT 6: HEATER DRAIN DEAERATOR CONTROLLER FAILURE Time Pos EXPECTED Operator Response Comments Direct entry in OAOP-23.0, SRO Condensate/Feedwater System Failure May direct a power reduction to stabilize SRO Condensate/Feedwater Directs manual control with HD-57 to stabilize SRO HD Tank level SRO Directs l&C to investigate SRO May contact Shift Manager May conduct a brief (See Enclosure 1, page 43 SRO for format of the brief. ATC Monitor plant Announce entry into OAOP-23.0, ATC Condensate/Feedwater System Failure May reduce Reactor power IAW OENP-24.5 as ATC directed by CRS.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 38 of 48 EVENT 6: HEATER DRAIN DEAERATOR CONTROLLER FAILURE Time Pos EXPECTED Operator Response Comments Acknowledge and report alarm: UA-4 2-10 HD DEAERATOR LEVEL HIGH-LOW. BOP Alarm at 30 inches and lowering. Pump trip at 24 inches and lowering . BOP Diagnose HD Pump discharge valves full open Enter and announce OAOP-23.0, BOP Condensate/Feedwater System Failure Trips one of the operating Heater Drain pump BOP Maintains heater drain deaerator level less than Move to the next event when level is being 60 inches indicated on HEATER DRAIN controlled with the HD-V57. DEAERATOR LEVEL, HD-Ll-97 BOP If level reaches 60 inches UA-4, 3-10 may alarm and the HOD Moisture removal valves will open. After level is stabilized the APP has direction for re-closing the moisture removal valves. May dispatch TBAO to check HD Pump Air-Operated Discharge Level Control Valves, HD-BOP LV-91-1, 2, & 3. May direct TBAO to place HOD level control in Manual IAW 20P-35 Section 6.3.8. or swap BOP controller IAW 20P-35, Section 6.3.8 Monitors main condenser vacuum and BOP condensate parameters May have to secure a CSP if one auto started BOP during the evolution.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 ~ Page 39 of 48 EVENT 7/8/9: FUEL FAILURE I SCRAM I SPC F0248 VALVE FAILURE Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 9 to activate the Fuel Failure Monitor MSL Rad and as power is reduced , adjust Fuel Failure and Gross Fuel Failure to ensure MSL Hi-HI is received . (If NB005F reaches 100, Manually initiate Trigger 10). When requested as Unit 1 to stop Unit 1 Cable Spread Vent Fans, Initiate Trigger 11 . When requested as Unit 1 to stop CB Mech Equipment Room Vent Fan, Initiate Trigger 12. When requested delete malfunction to reset thermal overload on torus cooling water valve. When directed by the Lead Evaluator, place the simulator in FREEZE Do not reset the simulator until receipt of concurrence to do so from the lead examiner Simulator Operator Role Play If PEP-3.4.7 is requested, acknowledge the request. If asked as E&RC, acknowledge the request to obtain actual/projected off site dose rates. If asked as E&RC, acknowledge the request to sample off gas and reactor coolant for potential fuel damage. If asked as E&RC, acknowledge the request to obtain noble gas dose rates. If requested to investigate breaker for the Torus cooling valve (F024B) report that it has tripped on thermals. If asked to reset have Sim Op delete malfunction and report breaker is reset. Evaluator Notes Plant Response: Fuel failure will occur. When the Main Stack Rad indication is rising then a reactor scram and Group I isolation is inserted. Reactor Scram Pushbuttons are failed , other methods are available {i.e. ARI). Torus coolinQ valve will thermal out, can be reset. Objectives: SRO- Directs actions for a Reactor Scram, and RRCP. ROs - Insert a scram IAW RRCP and close Group I valves. Success Path: Rods inserted with ARI or mode switch. Scenario Termination: When rods are inserted and SPC is being placed in service the scenario may be terminated . Remind students not to erase any charts and not to discuss the scenario until told to do so by the evaluator/instructor.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 ~ Page 40 of 48 EVENT 7/8/9: FUEL FAILURE I SCRAM I SPC FAILURE Time Pos EXPECTED Operator Response Comments Direct entry in to AOP-5.4, SRO Radiological Releases. Directs reactor power to be reduced to clear UA-SRO 03 5-2 Asks Unit One to calculate site boundary dose SRO per PEP-3.4.7. Asks HPs for field surveys SRO Ensure AOG in service Notify E&RC to sample off- gas and reactor SRO coolant for potential fuel damage. When UA-03 4-2 is received : Verifies Main Stack Rad levels rising . Ensures AOG bypass valve is closed . May direct a power reduction When UA-23 3-6 is received, Directs the Critical Task #1 SRO insertion of a manual scram. (May direct sooner as a conservative decision) Critical Task #2 Directs the closure of Group I valves (MSIVs, Steam Line Drains (B21-F016 and F019) and Recirc Sample Valves (832-F019 and F020) When Torus temperature is greater than 95° F, directs placing Torus Cooling in service. SRO Gives permission for a reset of the torus cooling water valve thermals.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 41 of 48 EVENT 7/8/9: FUEL FAILURE I SCRAM I SPC FAILURE Inserts a manual reactor scram and recognizes failure of manual scram channel B. ATC See Enclosure 2 (page 44) for Scram Hard Card ATC Initiates ARI per the Scram Hard Card. Critical Task #1 Perform immediate actions for Reactor scram After steam flow is less than 3 x 106 lb/hr, place the reactor mode switch to shutdown When APRM downscale trip, then Trip the main turbine ATC Ensure the master reactor level controller setpoint is +170 inches When RPV water level is above 160 inches and rising then trips one reactor feed pump. - Close Group / Isolation valves. Critical Task #2 Closes MSIVs Closes Main Steam Line Drain valves 821 - ATC F016 and F019. Closes Reactor Recirc Sample Valves B32-F019 and F020 May operate SRVs as directed to stabilize pressure and control pressure as directed by the ATC CRS.

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 42 of 48 EVENT 7/8/9: FUEL FAILURE I SCRAM I SPC FAILURE Report Annunciators when received UA-03 (5-7) Area Rad Turbine Bldg High BOP UA-03 (5-2) Process Off-Gas Rad High UA-23 (2-6) Main Steam Line Rad Hi UA-23 (3-6) Main Steam Line Rad Hi-Hi/lnop Announces and enters OAOP-5.4, Radiological Releases May operate SRVs as directed to stabilize pressure and control pressure as directed by the CRS. When Torus temperature is greater than 95° F, places Torus Cooling in service. (Enclosure 3, page 45) Recognize failure of torus cooling water valve and dispatch an AO to investigate breaker. Maintains level as directed by the CRS

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 43 of 48 ENCLOSURE 1 Page 1 of 1 CONDUCT OF OPERATIONS AD-OP-ALL-1000 Rev. 6 Page 90 of 90 ATTACHMENT 8 Page 1 of 1

                        << Crew Brief Template >>

0 Announce "Crew Brief* Begin Brief 0 All crew members acknowledge announcement (Az Required) 0 Update the crew as needed: 0 Describe what happened and major actions taken 0 Procedures in-progress D Notifications: Rec:ip D Maintenance O Engineering O Others (Dispatcher, Station Management, etc.) 0 Future Direction and priorities 0 Discuss any contingency plans (Az Required) 0 Solicit questions/concerns from each crew member: 0 ROs Input 0 CRS 0 STA D Are there any alarms unexpected for the plant conditions? D What is the status of Critical Parameters? ( Az Required) EAL 0 Provide EAL and potential escalation criteria 0 Restore normal alarm announcement? (Yes/No) Flnl:h Brief 0 Announce "End of Brief"

2016 NRG SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 44 of 48 ENCLOSURE 2 Page 1of1 5.4 SCRAM Card Enter applicable leg: ........................ .. ................... ......... ............................ 0 Scram ATWS All Control Rods FULL-IN ................ 0 Indications of Hydraulic/Electrical ATWS ...................... ............. ........... 0 RPV Water Level.. .................. ....... .. D Ensure ARI initiated ..................... .. . 0

      - - - - - - - inches                                         Reactor Power ................................. D RPV Pressure ...... ............................ O

_ _ _ _ _ _ _ psig ------- % Communicate ATWS report Communicate scram report to CRS .............. ....... .............. .......... 0 to CRS ........... ............. ..................... 0 IF enabled, Place SULCV in service ........... ....... O THEN initiate a recirc pump manual runback ......... .... ....... .. ...................... D Insert Nuclear Instrumentation ........ O IF reactor power above 2% OR Ensure Turbine Oil System CANNOT be determined, Operating .... .................................... . O THEN trip both recirc pumps ........... D Ensure Reactor Recirculation Report reactor power to CRS ....... ... D Pump speed at 34% ........................ D Exit scram card and perform Ensure Heater Drain Pumps EOP-0 1-LEP-02 ..... .................. ........ 0 tripped .. ........................................... D Exit scram card ............................... D

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 45 of 48 ENCLOSURE 3 Page 1of2 ATTACHMENT SA Page 1 of1 Emergency Suppression Pool Cooling Using Loop A (20P-17) NOTE : This attachment is NOT to be used for normal system operations. START RHR SW A LOOP (CONV} START RHR SW A LOOP (NUC} OPEN SW-V101 D OPEN SW-V105 D CLOSE SW-V1 43 D OPEN SW-V102 D START CSW PUMPS AS NEEDED D CLOSE SW-V143 D IF LOCA SIGNAL IS PRESENT THEN D START PUMPS ON NSW HOR AS NEEDED D PLAC E RHR SW BOOSTER PUMPS IF LOCA SIGNAL IS PRESENT THEN PLACE D A & C LOCA OVERRI DE SW ITCH RHR SW BOOSTER PUM PS A & C LOCA TO MANUAL OVERRIDE OVERRIDE SWITCH TO MANUAL OVERRIDE START RH R SW PMP D ST ART RHR SW PMP D ADJ UST E11 -PDV-F068A D ADJU ST E11 -PDV-F068A D ESTABLISH CLG WTR TO VITAL HOR D ESTABLISH CLG WTR TO VITAL HOR D ST ART ADDITIONAL RHR SW PUMP D ST ART ADDITIONAL RHR SW PUMP D AND ADJ UST FLOW AS NEEDED AND ADJ UST FLOW AS NEEDED START RHR LOOP A IF LOCA SIGNAL IS PRESENT, THEN D VERIFY COOLING LOGIC IS MADE UP IF E1 1-F015A IS OPEN, TH EN D CLOSE E11 -F017A START LOOP A RHR PMP D OPEN E11 -F028A D THROTTLE E11 -F024A D THROTTLE E 11-F048A D START ADD ITIONAL LOOP A RHR PMP 0 AND ADJ UST FLOW AS NEEDED 2 2/'106 '1 2 S/'1 062

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 46 of 48 ENCLOSURE 3 Page 2 of 2 ATTACHMENT 88 Page 1 of1 Emergency Suppression Pool Cooling Using Loop B (20P-17) NOTE: This attachment is NOT to be used for normal system operations. START RHR SW B LOOP (NUC} START RHR SW B LOOP (CONV} OPEN SW-V1 05 0 OPEN SW-V101 0 CLOSE SW- V143 0 OPEN SW-V102 0 START PMPS ON NSW HDR AS NEEDED 0 CLOSE SW-V143 0 IF LOCA SIGNAL IS PRESENT THEN 0 ST ART CSW PUMPS AS NEEDED 0 PLACE RHR SW BOOSTER PUMPS IF LOCA SIGNAL IS PRESENT THEN PLACE 0 B & D LOCA OVERRIDE SW ITCH RH R SW BOOSTER PUMPS B & D LOCA TO MAN UAL OVERRIDE OVERRIDE SWITCH TO MANUAL OVERRIDE ST ART RHR SW PMP 0 ST ART RHR SW PMP 0 ADJ UST E 11-PO V-F068B 0 ADJ UST E1 1-PDV-F068B 0 ESTABLISH CLG WTR TO VITAL HOR 0 EST ABUSH CLG WTR TO V ITAL HOR 0 ST ART ADD ITIONAL RHR SW PUMP 0 ST ART ADDITIONAL RHR SW PUMP 0 AND ADJ UST FLOW AS NEEDED AN D ADJUST FLOW AS NEEDED START RHR LOOP B IF LOCA SIGNAL IS PRESENT. THEN 0 VERIFY COOLING LOGIC IS MADE UP IF E 11 -F015B IS OPEN, THEN D CLOSE E11 -F017B ST AR T LOOP B RHR PMP D OPEN E11 -F028B D TH ROTTLE E 11-F024B D TH ROTTLE E11 -F048B D START ADDITIONAL LOOP B RHR PMP 0 AN D ADJU ST FLOW AS NEEDED 2 2/'1 063 2 S/'1064

2016 NRC SCENARIO 1 LOI SIMULATOR EVALUATION GUIDE Rev. O Page 47 of 48 ATTACHMENT 1 - Scenario Quantitative Attribute Assessment NUREG 1021 Category Scenario Content Rev. 2 Supp. 1 Req. Total Malfunctions 5-8 8 Malfunctions after EOP 1-2 2 Entry Abnormal Events 2-4 4 Major Transients 1-2 1 EOPs Used 1-2 2 EOP Contingency 0-2 1 Run Time 60-90 min 90 Crew Critical Tasks 2-3 2 Tech Specs 2 2 Instrument I Component 2-0ATC Failures before Major 4 2-BOP Instrument I Component 2 2 Failures after Major Normal Operations 1 1 Reactivity manipulation 1 1

LOI SIMULATOR EVALUATION GUIDE Page 48 of 48 ATTACHMENT 2-Shift Turnover Brunswick Unit 2 Plant Status Station Duty Workweek E. Neal B. Craig Manager: Manager: Mode: 1 J Rx Power: 100% J Gross*/Net MWe*: 977 I 951 Plant Risk: Green Current EOOS Risk Assessment is: SFPTimeto 49.7 hrs Days Online: 80 days 200 Deg F: Turnover: 20P-32, Section 6.3.6 has been completed up to step 6.3.6.11. Protected 2A FPC Pump/Hx, 2A RCC Pump, and 2C Demin Transfer Pump for Equipment: Fuel Pool Decay Heat Removal and inventory makeup. 2A/B NSW Pumps due to 1A NSW pump maintenance 1A NSW Pump is under clearance for planned maintenance. 2C TCC Pump is in service on Unit One. Comments: APRM 2 has failed downscale and is bypassed. The BOP operator will continue procedure to swap Condensate Booster Pumps (place CBP 2C in service and remove CBP 2A from service for maintenance).

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev.206 Page 131 of 408 6.3.6 Transferring to Standby Condensate Booster Pump

1. Confirm at least one condensate booster pump in operation ................ _BO_ _

NOTE

• Lowering CFO and COD system differential pressure will raise Condensate Booster Pump suction pressure. This may also minimize the potential of an override or bypass condition of the COD and CF D's ..................................................... IXI
• If HWC is in service, from operating experience, UA-23, 2-6 Main Steam Line Rad Hi may alarm when starting a condensate booster pump. [8.7.17] ........................ W

2. Notify Chemistry performing this section............................................... RO J.Johnson Person Notified

3. Designate the oncoming and off going condensate booster pumps below:.................................................................................................... RO Oncoming Condensate Booster Pump c

Off going Condensate Booster Pump A

4. Confirm Condensate Booster Pump suction pressure is greater than 85 psig on either COD-Pl-30-1 or PPC point U2CODL071. ........... RO

5. IF Condensate Booster Pump suction pressure is less than or equal to 85 psig, THEN perform the following:................................................................. NIA

a. Direct Radwaste operator to perform the following to lower CFO and COD system differential pressure:

(1) IF available, THEN place an additional CFO in service ........................ _hUA_ (2) IF available, THEN place an additional COD in service ....................... N/A

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev.206 Page 132 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

b. IF after performing Step 5.a, Condensate Booster Pump suction pressure is still less than or equal to 85 psig, THEN perform the following: ...................................................... __w.A__

(1) Record the following in the Operator's Log:

• Values from both suction pressure indicators. ....... NIA
• Reactor power level. .............................................. __hliA___

(2) Perform Section 6.3.22, Starting a Third Condensate Pump for Operation During Abnormal Plant Conditions to place the third Condensate Pump in service. ............... N/A (3) Record third Condensate Pump start to support Condensate Booster Pump start in Operator log. ............. NIA

6. Direct Radwaste operator to monitor for proper operation of hotwell level control. ........................................................................................... RO

7. Ensure proper oil level in oncoming condensate pump oil reservoir..... RO CAUTION Proper venting of the condensate booster pump mechanical seal chamber, prior to starting the pump is critical to seal performance and longevity............................................ ~

8. IF 2A Condensate Booster Pump is to be started, THEN perform the following: .................................................................__hliA___

a. Open COD-V5014 (Condensate Booster Pump A Inboard Mechanical Seal Vent Valve) . ..................................................... ~

b. WHEN a continuous solid stream of water has been observed for at least 2 minutes, THEN close COD-V5014 (Condensate Booster Pump A Inboard Mechanical Seal Vent Valve)...... ................................... N/A

c. Open COD-V5015 (Condensate Booster Pump A Outboard Mechanical Seal Vent Valve).................... .. ................................ N/A

d. WHEN a continuous solid stream of water has been observed for at least 2 minutes, THEN close COD-V5015 (Condensate Booster Pump A Outboard Mechanical Seal Vent Valve) . ..................................... __hliA___

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev.206 Page 133 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

9. IF 2B Condensate Booster Pump is to be started, THEN perform the following: ................................................................. ___NLA_

a. Open COD-V5016 (Condensate Booster Pump B Inboard Mechanical Seal Vent Valve). ..................................................... NIA

b. WHEN a continuous solid stream of water has been observed for at least 2 minutes, THEN close COD-V5016 (Condensate Booster Pump B Inboard Mechanical Seal Vent Valve)........ ................................. N/A

c. Open COD-V5017 (Condensate Booster Pump B Outboard Mechanical Seal Vent Valve)...................................................... NIA

d. WHEN a continuous solid stream of water has been observed for at least 2 minutes, THEN close COD-V5017 (Condensate Booster Pump B Outboard Mechanical Seal Vent Valve) ...................................... _m~

10. IF 2C Condensate Booster Pump is to be started, THEN perform the following:................................ ................................. RO

a. Open COD-V5018 (Condensate Booster Pump C Inboard Mechanical Seal Vent Valve)...................................................... AO

b. WHEN a continuous solid stream of water has been observed for at least 2 minutes, THEN close COD-V5018 (Condensate Booster Pump C Inboard Mechanical Seal Vent Valve)......................................... AO

c. Open COD-V5019 (Condensate Booster Pump C Outboard Mechanical Seal Vent Valve)...................................................... AO

d. WHEN a continuous solid stream of water has been observed for at least 2 minutes, THEN close COD-V5019 (Condensate Booster Pump C Outboard Mechanical Seal Vent Valve). ..................................... AO

11. Note status of the following alarms:

• UA-13, 1-1, Gen-Xfmr Primary LIO Unit Trip ............................... _ __
• UA-13, 1-2, Generator Diff LIO Unit Trip .................................... _ __
• UA-13, 1-3, Gen-Xfmr Backup LIO Unit Trip .............................. _ _,

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev. 206 Page 134 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

12. IF all the alarms in Step 11 are CLEAR, THEN perform the following:[B. 7 .S] ...................................................... _ __

a. Place the following switches in ENABLED for the off going condensate booster pump:

• Unit Trip Load Shed Selector Switch ............................... _ _ __
• LOCA Load Shed Selector Switch .................................... _ _ __

b. Ensure the oncoming condensate booster pump mode selector switch in MAN ................................................................ _ _ __

c. Confirm oncoming condensate pump discharge valve closes, if OPEN:

• COD-V4 (Condensate Booster Pump A Disch Viv) .......... _ __
• COD-VS (Condensate Booster Pump B Disch Viv) .......... _ __
• COD-V6 (Condensate Booster Pump C Disch Viv) .......... _ __

NOTE The start/stop control switch for a Condensate Booster Pump must be held in START until the recirculation valve is full OPEN. The recirculation valve does NOT stroke OPEN until all other condensate booster pump permissives are met. While opening the recirc valve, condensate booster pump suction pressure will lower as additional flow up to 2000 gpm is routed to the condenser .................................................. D

d. Start the oncoming condensate booster pump ........................... _ _ __

e. Confirm oncoming condensate booster pump discharge valve opens:

• COD-V4 (Condensate Booster Pump A Disch Viv) .......... _ __
• COD-VS (Condensate Booster Pump B Disch Viv) .......... _ __
• COD-V6 (Condensate Booster Pump C Disch Viv) .......... _ __

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev.206 Page 135 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

f. WHEN condensate booster pump discharge pressure stabilizes, THEN perform the following: ...................................................... _ __

(1) Stop off going condensate booster pump ........................ _ __ (2) IF B21-F032A (Feedwater lsol Viv) AND B21-F032B (Feedwater lsol Viv) are OPEN, THEN place the off going condensate booster pump mode switch in AUTO ....................................................... _ __

g. Place the following switches in DISABLED for oncoming condensate booster pump:

• Unit Trip Load Shed Selector Switch ............................... _ __
• LOCA Load Shed Selector Switch .................................... _ __

h. IF HWC is in service, THEN perform the following: ...................................................... _ _

(1) Place HWC condensate H2 injection isolation solenoid valve, at HWC local panel H2J, for oncoming condensate booster pump, in AUTO:

• HWCH-SV-5717 (Condensate Booster Pump A H2 Injection Isolation Valve) ................................... _ __
• HWCH-SV-5718 (Condensate Booster Pump B H2 Injection Isolation Valve) ................................... _ __
• HWCH-SV-5719 (Condensate Booster Pump C H2 Injection Isolation Valve) ................................... _ __

(2) Confirm HWC condensate H2 injection isolation solenoid valve, at HWC local panel H2J, for oncoming condensate booster pump, is OPEN:

• HWCH-SV-5717 (Condensate Booster Pump A H2 Injection Isolation Valve) ................................... _ __
• HWCH-SV-5718 (Condensate Booster Pump B H2 Injection Isolation Valve) ................................... _ __
• HWCH-SV-5719 (Condensate Booster Pump C H2 Injection Isolation Valve) ................................... _ __

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev.206 Page 136 of 408 6.3.6 Transferring to Standby Condensate Booster Pump {continued) (3) Place the HWC condensate H2 injection isolation solenoid valve, at HWC local panel H2J, for off going condensate booster pump in CLOSED.

• HWCH-SV-5717 (Condensate Booster Pump A H2 Injection Isolation Valve) ................................... _ __
• HWCH-SV-5718 (Condensate Booster Pump B H2 Injection Isolation Valve) ................................... _ __
• HWCH-SV-5719 (Condensate Booster Pump C H2 Injection Isolation Valve) ................................... _ __

i. Go to Step 20 ............................................................................. _ __

13. Place the following switches in DISABLED for the oncoming Condensate Booster Pump:

• Unit Trip Load Shed Selector Switch ........................................... _ __
• LOCA Load Shed Selector Switch .............................................. _ _ _

14. Place the oncoming condensate booster pump mode selector switch in MAN ........................................................................................ _ __

15. Confirm oncoming condensate booster pump discharge valve closes.

• COD-V4 (Condensate Booster Pump A Disch Viv) ..................... _ __
• COD-V5 (Condensate Booster Pump B Disch Viv) ..................... _ __
• COD-V6 (Condensate Booster Pump C Disch Viv) ..................... _ __

NOTE The start/stop control switch for a condensate booster pump must be held in START until the recirculation valve is full OPEN. The recirculation valve does NOT stroke OPEN until all other condensate booster pump permissives are met. While opening the recirc valve, condensate booster pump suction pressure will lower as additional flow up to 2000 gpm is routed to the condenser .................................................. D

16. Start oncoming condensate booster pump ............................................ _ __

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev.206 Page 137 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

17. Confirm oncoming condensate booster pump discharge valve opens:

• COD-V4 (Condensate Booster Pump A Disch Viv) ..................... _ __
• COD-VS (Condensate Booster Pump B Disch Viv) ..................... _ __
• COD-V6 (Condensate Booster Pump C Disch Viv) ..................... _ __

18. WHEN condensate booster pump discharge pressure stabilizes, THEN perform the following: ................................................................. _ __

a. Stop off going condensate booster pump ................................... _ __

b. IF B21-F032A (Feedwater lsol Viv) AND B21-F032B (Feedwater lsol Viv) are OPEN ,

THEN place the off going condensate booster pump mode switch in AUTO. ......................................................................... _ __

c. Place the following switches in ENABLED for the off going condensate booster pump:

• Unit Trip Load Shed Selector Switch ................................ _ __
• LOCA Load Shed Selector Switch .................................... _ __

19. IF HWC is in service, THEN perform the following: ................................................................. _ __

a. Place oncoming condensate booster pump, HWC condensate H2 injection isolation solenoid valve, at HWC local panel H2J, in AUTO:

• HWCH-SV-5717 (Condensate Booster Pump A H2 Injection Isolation Valve) ................................................... _ __
• HWCH-SV-5718 (Condensate Booster Pump B H2 Injection Isolation Valve) ................................................... _ _
• HWCH-SV-5719 (Condensate Booster Pump C H2 Injection Isolation Valve) ................................................... _ _ _.

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev. 206 Page 138 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued)

b. Confirm oncoming condensate booster pump, HWC condensate H2 injection isolation solenoid valve, at HWC local panel H2J, OPEN:

• HWCH-SV-5717 (Condensate Booster Pump A H2 Injection Isolation Valve) ................................................... _ __
• HWCH-SV-5718 (Condensate Booster Pump B H2 Injection Isolation Valve) ................................................... _ __
• HWCH-SV-5719 (Condensate Booster Pump C H2 Injection Isolation Valve) ................................................... _ __

c. Place off going condensate booster pump, HWC condensate H2 injection isolation solenoid valve, at HWC local panel H2J, in CLOSED:

• HWCH-SV-5717 (Condensate Booster Pump A H2 Injection Isolation Valve) ................................................... _ __
• HWCH-SV-5718 (Condensate Booster Pump B H2 Injection Isolation Valve) ................................................... _ __
• HWCH-SV-5719 (Condensate Booster Pump C H2 Injection Isolation Valve) ................................................... _ _

20. IF three condensate pumps are in service per Step 5.b(2) ,

THEN perform Section 6.3.23, Securing from Three Condensate Pump Operation ....................................................................................._ __

21. Direct Radwaste Operator to remove the additional CFD or CDD placed in service in Step 5.a .................................................................. _ __

22. Direct Radwaste Operator to monitor CDD effluent conductivity for each demineralizer in service ............................................................... .- - -

23. IF any condensate pump Unit Trip Load Shed Selector Switch AND LOCA Load Shed Selector Switch was manipulated, THEN complete Attachment 9, Condensate Pump Unit Trip Load Shed/LOCA Load Shed Selector Switch Alignment Documentation ...... _ __

24. Complete Attachment 10, Condensate Booster Pump Unit Trip Load Shed/LOCA Load Shed Selector Switch Alignment Documentation ......................................................................................._ __

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev.206 Page 139 of 408 6.3.6 Transferring to Standby Condensate Booster Pump (continued) Date/Time Completed Performed By (Print) Initials Reviewed By Unit CRS/SRO

CONDENSATE AND FEEDWATER SYSTEM 20P-32 OPERATING PROCEDURE Rev.206 Page 406 of 408 ATTACHMENT 10 Page 1 of 1 Condensate Booster Pump Unit Trip Load Shed/LOCA Load Shed Selector Switch Alignment Documentation NOTE

• When performing this attachment, pump selector switches, for pump(s) which have NOT been realigned may be marked NA .............................................................. D
• The condensate booster pump selected for standby operation (Mode Selector Switch in AUTO) is required to have both the LOCA Load Shed Selector Switch and Unit Trip Load Shed Selector Switch in ENABLED ................................................. D

1. Circle the required selector switch position.

Position/ Number Description Indication Checked Verified rNote 11 Turbine Building - 4160V Switchgear Bus 2D - Elev 20 ft 2B Condensate Booster Pump Unit Trip Load ENABLED/ AD9 Shed Selector Switch DISABLED 2B Condensate Booster Pump LOCA Load Shed ENABLED/ AD9 Selector Switch DISABLED Turbine Building -4160V Switchgear Bus 2C - Elev 20 ft 2A Condensate Booster Pump Unit Trip Load ENABLED/ AC1 Shed Selector Switch DISABLED 2A Condensate Booster Pump LOCA Load Shed ENABLED/ AC1 Selector Switch DISABLED 2C Condensate Booster Pump Unit Trip Load ENABLED/ AC2 Shed Selector Switch DISABLED 2C Condensate Booster Pump LOCA Load Shed ENABLED/ AC2 Selector Switch DISABLED

faIb DUKE ENERGY. BRUNSWICK TRAINING SECTION OPERATIONS TRAINING INITIAL LICENSED OPERATOR SIMULATOR EVALUATION GUIDE 2016 NRC SCENARIO 2 LOWER POWER, REMOVE 230KV LINE FROM SERVICE, ROD DRIFT, ADHR PP TRIP, RECIRC LOOP FLOW FAILURE, HDD PP TRIP, ATWS, SLC MODE SWITCH FAILURE, ARI FAIL TO RESET REVISION 0 Developer: d Vo& Date: C7IOZIO/6 Technical Review: Va%4 Date: %Il2/2C6 Validators: e eoe Date: 09/06//6 Facility Representative: Date:

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 2 of 72 REVISION

SUMMARY

Scenario developed for 2016 NRC Exam.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 3 of 72 TABLE OF CONTENTS 1.0 SCENARIO OUTLINE 4 2.0 SCENARIO DESCRIPTION

SUMMARY

5 3.0 CREW CRITICAL TASKS 6 4.0 TERMINATION CRITERIA 6 5.0 IMPLEMENTING REFERENCES 7 6.0 SETUP INSTRUCTIONS 8 7.0 INTERVENTIONS 10 8.0 OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES 12 ATTACHMENT 1 - Scenario Quantitative Attribute Assessment 49 ATTACHMENT 2 Shift Turnover 72

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 4 of 72 1.0 SCENARIO OUTLINE Event MaIf. No. [ Type* Event Description 1 R ATC

                           -        Lower power to 850 MWe to remove 230 kV Line 30 2                   N - BOP    Remove 230 kV Line 30 from service RDOO1 M      C ATC
                           -        Rod Drift 3       (26-11 C-CRS                                                             (TS)

K4526A C BOP

                           -        ADHR Secondary pump trip 4

C-CRS (AOP) N1063F C ATC

                           -        Recirc Loop B Flow transmitter Failure 5

C-CRS (TS) CFO89F C BOP

                           -        Heater Drain Deaerator Pump Trip 6

C-CRS (AOP) Small steam leak in DW results in an ATWS requiring 7 CAOO8F M terminate and prevent actions (RSP)(ATWS) 8 K2119A C SLC Mode Switch Failure 9 K2624A C Alternate Rod Insertion reset failure

                *(N)ormal    (R)eactivity,  (C)omponent or Instrument,    (M)ajor

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 5 of 72 2.0 SCENARIO DESCRIPTION

SUMMARY

Event Description 1 After taking the watch the CRS will direct power reduced to 850 MWe. 2 The BOP will isolate 230 kV Line 30. Control Rod 26-1 1 will start to drift in. The crew will enter OAOP-02.0 and take action lAW 2APP-A-05 (3-2). When the high temperature alarm is received Engineering will 3 report that scram times cannot be assured based on past history of the control rod. Determine TS 3.1.3 condition Cl to insert the control rod in 3 hours and 02 to disarm the control rod within 4 hours. After Tech Specs are addressed the Alternate Decay Heat Removal (ADHR) Seconday pump will trip. AOP-38.0 will be entered The Recirc Loop B flow transmitter to APRM Channel 4 will fail downscale resulting in a rod block and a trip input to each voter. The crew will respond per APPs and bypass 5 APRM 4. The APRM will be declared Inoperable per TS 3.3.1.1, Condition A and placed in trip within 12 hours. APRM TS Actions to be taken requires the APRM mode selector switch to be place in INOP lAW 001-18 A motor overload will occur on Heater Drain Pump 2A. The crew will reference APP UA-06 1-7, Bus 2D 4KV Motor OvId and determine which pump has the overload 6 condition. The crew should start HDP 20 and secure HDP 2A. The crew may reference AOP-23.0. A small steam leak in the DW results in rising Drywell pressure requiring a reactor 7 scram. An ATWS will occur, conditions will require terminate and prevent actions to be performed. When SLC is initiated, the mode switch will fail and the pumps will not start. LEP-03 8 will be executed to inject the boron into the core. Alternate Rod Insertion (ARI) will not reset, the crew will perform LEP-02 to drive 9 control rods into the core. When level is stabilized after terminating and preventing ARI will be repaired to allow the rods to be manually scrammed.

2016 NRC SCENARIO 2 LOt SIMULATOR EVALUATION GUIDE Rev.0 Page6of72 3.0 CREW CRITICAL TASKS Critical Task #1 Insert control rods lAW LEP-02 Critical Task #2 Direct Alternate Boron Injection lAW LEP-03 Critical Task #3 Terminate and prevent injection from HPCI/Condensate and Feedwater/LP ECCS 4.0 TERMINATION CRITERIA When all rods are inserted and level is being controlled above TAF the scenario may be terminated.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 7 of 72 5.0 IMPLEMENTING REFERENCES NOTE: Refer to the most current revision of each Implementing Reference. Number Title A-05, 3-2 ROD DRIFT OAOP-02.O CONTROL ROD MALFUNCTIQN/M ISPOSITI ON UA-18, 6-1 BUS E4 4KV MOTOR OVLD. UA-O1, 2-3 ADHR PRIMARY LOOP TROUBLE UA-01, 3-3 ADHR SECONDARY LOOP TROUBLE OAOP-38.O LOSS OF FUEL POOL COOLING A-06, 2-8 APRM UPSCALE A-06, 3-8 APRM UPSCALE TRIP/INOP A-06, 5-7 FLOW REF OFF NORMAL A-05, 2-2 ROD OUT BLOCK A-05, 4-8 OPRM TRIP ENABLED UA-5, 3-5 SBGT SYS B FAILURE UA-5, 4-6 SBGT SYS A FAILURE

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 I Page 8of72 6.0 SETUP INSTRUCTIONS

1. PERFORM TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 5, Checklist for Simulator Exam Security.

2. RESET the Simulator to IC-i 1.

3. ENSURE the RWM is set up as required for the selected IC.

4. ENSURE appropriate keys have blanks in switches.

5. RESET alarms on SJAE, MSL, and RWM NUMACs.

6. ENSURE no rods are bypassed in the RWM.

7. PLACE all SPDS displays to the Critical Plant Variable display (#100).

8. ENSURE hard cards and flow charts are cleaned up

9. TAKE the SIMULATOR OUT OF FREEZE

10. LOAD Scenario File.

11. ALIGN the plant as follows:

Manipulation Ensure 2C TCC pump is in service on Unit One. Bypass APRM 2 RCC Pump D in service for ADHR RCC Pump A in service for RBCCW

12. IF desired, take a SNAPSHOT and save into an available IC for later use.

13. PLACE a clearance on the following equipment.

Component Position APRM 2 Blue tag

14. INSTALL Protected Equipment signage and UPDATE RTGB placard as follows:

Protected Equipment

1. 2A and 2B NSW pumps

2. 2A FPC Pump/Hx, 2A RCC Pump, and 2C Demin Transfer Pump.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page9of72

15. VERIFY OENP 24.5 Form 2 (Immediate Power Reduction Form) for IC-i i is in place.

16. ENSURE each Implementing References listed in Section 7 is intact and free of marks.

17. ENSURE all materials in the table below are in place and marked-up to the step identified.

Required Materials Marked up of 20P-50, Section 6.2.6

18. ADVANCE the recorders to prevent examinees from seeing relevant scenario details.

19. PROVIDE Shift Briefing sheet for the CRS.

20. VERIFY all actions contained in TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 4, Simulator Training Instructor Checklist, are complete.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 10 of 72 7.0 INTERVENTIONS TRIGGERS Trig Type ID 1 Malfunction RDOO1M [CONTROL ROD SLOW INSERTION DRIFT] 2 Annunciator ZA512 [CRD HYD TEMP HIGH] 3 Trigger Command MFD:RDOO1M,26-11 4 Remote Function RD_RDELDIS [ELECTRICAL DISARM OF ROD] 5 Dl Override K4526A [RBCCW PMP D AUTO] 5 Dl Override - K4526A - [RBCCW PMP D AUTO] 5 DI Override K4526A - [RBCCW PMP D AUTO] 5 DO Override - Q4526AMW [RBCCW PMP D ADHR MODE] 5 DO Override Q4526LG4 [RBCCW PMP D OFF G] 5 Malfunction RPO11F [ATWS 4] 6 Remote Function CC_MODE [RBCCW/ADHR VALVE LINEUPS] 6 Remote Function CC_MSS [RBCCW/ADHR PUMP MODE SELECTOR SWITCH] 7 Remote Function CC_PDV [RBCCW PUMP DISCHARGE VALVE] 8 Malfunction N1063F [RECIRC LOOP B XMIUER FAILURE] 9 Malfunction -_____ CFO89F [HEATER DRAIN PUMP MOTOR WINDING FAULT] 10 Malfunction NBO06F [MSL BRK BEFORE FLOW RESTRICTOR] 11 Remote Function EPIAEOPJP1 [BYPASS LL-3 GROUP I ISOL (SEP-10)] Trig # frrigger Text 3 KM118EDN - [SCRAM TEST SWITCH 26-11] true deletes RDOO1M ANNUNCIATORS Window Description Tagname Override OVal AVaI Type Actime Dactime Trig 1-2 CRD HYD TEMP HIGH ZA512 ON ON OFF 2

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 11 of 72 MALFUNCTIONS Maif Mult . . Description Current I Target Rmp Actime Dactime Trig ID ID Value Value time CONTROL ROD SLOW INSERTION RD001M 26-11 False True 1 NIO63F APRM 4 RECIRC LOOP B XMIUER FAILURE 0.00 125.00 8 NBOO6F A MSL BRK BEFORE FLOW RESTRICTOR 0.00 1.Oe-1 0:03:00 10 H EATE R D RAIN PUMP M OTOR CFO89F A False True 9 WINDING FAULT RPO11F ATWS 4 False True 5 RPOO5F AUTO SCRAM DEFEAT True True NIO32F APRM 2 APRM FAILS LO True True REMOTES Remf Id Mult Id Description Current Thret Actime Trig RD_RDELDIS 26-11 ELECTRICAL DISARM OF ROD ARM DISARM 4 CC_MODE PUMP-A RBCCW/ADHR VALVE LINEUPS RBCCW ADHR 6 RBCCW/ADHR PUMP MODE CC MSS A RBCCW ADHR 6 SELECTOR SWITCH CC_PDV A_V38_V5114 RBCCW PUMP DISCHARGE VALVE 1.0000 1.Oe-01 7 CC_IACW458 2CTBCCW PUMP UNITALIGNMENT 1 1 EP_IAEOPJP1 BYPASS LL-3 GROUP I ISOL (SEP-10) OFF ON 11 PANEL OVERRIDES Tag ID P05 I Actual Override Description Actime Dactime Trig K4526A RBCCW PUMP D OFF OFF/RESEST OFF ON 5 K4526A RBCCW PMP D AUTO AUTO OFF OFF 5 K4526A RBCCW PMP D ON ON ON__ OFF 5 Q4526LG4 RBCCW PMP D OFF 6 ON/OFF OFF OFF 5 Q4526AMW RBCCWPMPDADHRMODE ON/OFF ON OFF 5 K2119A S/B LIQ PUMP A & B PUMP_A OFF OFF K2119A S/B LIQ PUMP A & B PUMP_A&B OFF OFF K2119A S/B LIQ PUMP A & B PUMP_B OFF OFF K2624A CS-5562 ARI RESET OFF OFF K2625A CS-5560 ARI INOP OFF OFF

2016 NRC SCENARIO 1 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 12 of 49 8.0 OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES EVENT 1: SWAP CBPs Simulator Operator Actions Ensure Monitored Parameters is open and Scenario Based Testing Variables are loaded. Initiate Trigger 1 to ENABLE 2A CBP Unit Trip and LOCA Load Shed Selector Switches. Initiate Trigger 2 to DISABLE 20 CBP Unit Trip and LOCA Load Shed Selector Switches Initiate Trigger 13 to place HWCH-SV-5719 to open. Initiate Trigger 14 to place HWCH-SV-5717 to close. Simulator Operator Role Play Acknowledge request to enable 2A CBP Unit Trip and LOCA Load Shed selector switches, after Sim Operator activates the trigger report that the action is complete. Acknowledge request to disable 2C CBP Unit Trip and LOCA Load Shed selector switches, after Sim Operator activates the trigger report that the action is complete. Acknowledge request to place HWCH-SV-571 9 (Condensate Booster Pump C H2 Injection Isolation Valve) in AUTO, have SIM OP initiate trigger 13 and report valve is open. Acknowledge request to place HWCH-SV-5717 (Condensate Booster Pump B H2 Injection Isolation Valve) in CLOSE and then report that the action is complete Acknowledge any requests for Radwaste actions. Acknowledge request to perform 20P-32, Attachment 10. Report that you will co-ordinate the performance of the attachment with the WOO. Bus C Area is reported as clear if requested or if camera is checked. Evaluator Notes Plant Response: 20 CBP is started and 2A CBP is secured. Objectives: SRO Directs BOP to swap Condensate Booster Pumps BOP Swap Condensate Booster Pumps RO Monitors the plant Success Path: Condensate Booster Pumps are swapped Event Termination: When directed by the Lead Evaluator, go to Event 2.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 13 of 72 EVENT 1: LOWER POWER TO 850 MWE Time Pos EXPECTED Operator Response NOTES SRO Conduct shift turnover shift briefing. Direct power to be reduced using recirc flow to 850 MWe. (20P-02, Section 6.2.1) Contacts chemistry for samples due to 15% power change. May contact Load dispatcher to inform of power decrease. May conduct a brief (See Enclosure 1, page 62 for format of the brief. Reduces reactor power using recirc lAW 20P-RO 02 Section 6.2.1 May null the DVM meter. BOP Monitors the plant

2016 NRC SCENARIO 2 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 14of72 t REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev. 168 Page 45 of 250 6.2 Shutdown 6.2.1 Lowering SpeedtPower Using Individual Recirculation Pump Control Or Recirc Master Control

1. Confirm reactor recirculation pump in operation in accordance with Section 6 1.2 NOTE

• Recirculation Pump speed changes are performed when directed by OGP-05, Unit Shutdown, and OGP-i2, Power Changes Other operating procedures are used simultaneously wth this procedure as directed by OGP-05, Unit Shutdown, and OGP-i2, Power Changes C
• Speed changes are accomplished by depressing Lower Slow, Lower Mednim, or Lower Fast pushbuttons. The Lower Slow pushbutton changes Recirc pump speed at 0.06%/decrement at 1 rpm/second The Lower Medium pushbutton changes Recirc ptimp speed at 0.28%/decrement at 5 rpm/second. The Lower Fast pushbutton changes Recirc pump speed at 2.8%Idecrement at 100 rpm/second C

2. IF AT ANY TIME any of the following conditions exist.

THEN enter 1AOP-04.0, Low Core Flow {8 1.9)

• Entry into Region A of Power to Flow Map
• OPRM INOPERABLE AND any of the following 0 Entry into Region B of Power to Flow Map 0 Entry into 5% Buffer Region of Power to Flow Map 0 Entry into OPRM Enabled Region and indications of THI (Themial Hydraulic Instability) exist

2016 NRC SCENARIO 2 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 15 of 72 REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev 168 Page 46 of 250 6.2.1 Lowering SpeedlPower Using Individual Recirculation Pump Control Or Recirc Master Control (continued) CAUTION

• The OPRM System monitors LPRMs for indication of themial hydraulic instability çtHl). When greater than or equal to 25% power and less than or equal to 60% recirculation flow, alamis and automatic trips are initiated upon detection of THI. Pump operations are governed by the limits of the applicable Power Flow Map, as specified in the COLR, {8. 1 9) C
• Entry into the 5% Buffer Region warrants increased monitoring of reactor instrumentation for signs of Themial Hydraulic Instability Time in the 5% Buffer Region presents additional risk and is rninimized.{8. 1 .9} C
• With core flow less than 57.5 x .108 lbshr, jet pump loop flows are required within 10% (maximum indicated difference 6.0 x 108 lbs/br). Wh core flow greater than or equal to 57.5 x 108 lbsfhr, jet pump loop flows are required within 5% (maximum indicated difference 3.0 x 108 lbs/hr)
• When Recirc Pump speeds ate less than 40%, decreasing speed using a Lower Fast pushbutton can result in a Speed Hold condition due to exceeding the tegen torque limit C BEGIN R.M. LEVEL R21R3 REACTIVITY EVOLUTION

3. IF desired to lower the speed of both redrculation pumps simultaneously, THEN depress Redrc Master Control Lower (Slow Medium Fast) pushbutton 4 IF desired to lower the speed of an individual recirculation pump, THEN depress the Recirc VFD A(B) Lower (Slow Medium Fast) pushbutton

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 16 of 72 REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev. 168 Page 47 of 250 6.2.1 Lowering SpeedlPower Using Individual Recirculation Pump Control Or Recirc Master Control (continued)

5. Confirm the folloving, as applicable Recirc Pump Af B) Speed Demand, Calculated Speed, and Actual Speed have lowered

• Reactor power lowers
• B32-R6 1 7(R6l 3) [Recirc Pump Af B) Discharge Flow] lowers
• B32-VFD-IDS-003A(B) [Recirc VFD 2A(B) Output Waflmeterj lowers
• B32-VFD-IDS-00 1A(B)JRecirc VFD 2A(B) Output Frequency Meter] lowers END R.M. LEVEL R2/R3 REACTIVITY EVOLUTION Dateiflme Completed Perfomed By (Print) Initials Reviewed By:

Unit CRS/SRO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 17 of 72 EVENT 2: ISOLATE 230 KV DELCO WEST LINE 30 Simulator Operator Actions Simulator Operator Role Play Evaluator Notes Plant Response: 230 kV Delco West line is isolated Objectives: SRO - Direct 230kV Delco West Line isolated ATC Plant monitoring BOP Performs 2OP-50 Section 6.2.6 for isolating ONLY the Delco West Line Success Path: 230 kV Delco West (Line 30) isolated Event Termination: Go to Event 3 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 18 of 72 EVENT 2: ISOLATE 230 KV DELCO WEST LINE 30 Time Pos EXPECTED Operator Response Comments Directs 230kV Delco West Line isolated lAW SRO marked up version of 2OP-50, Section 6.2.6. BOP Performs 20P-50, Section 6.2.6 RO Monitors the plant.

2016 NRC SCENARIO 2 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 19 °? I PLANT ELECTRIC SYSTEM OPERATING 20P-50 PROCEDURE Rev 147 Page 58 of 281 6.2.6 De.energlzlng The 230 kV Switchyard 1 Ensure the Unit 2 230 kV switchyard Is ENERGIZED AD

2. Ensure the 4kV Auxihary Electrical Systems are DE-ENERGIZED in accordance with Section 6 2,3 . .. .... ....... ................ N-I SRO 3 Ensure the SAT is DE-ENERGIZED in accordance with Section 6.2.4 N-i SRO 4 Ensure Caswell Beach Pumping Station is DE-ENERGIZED in accordance with Section 6.2.5 N-i SRO

5. Ensure required LCOs for Technical Specification Sections 3.81.

3.8.2, 38.7 and 3.8.8 are initiated SRI)

6. Obtain Load Dispatchers permission to dc-energize the 230 Ky switchyard AD A Powers The Delco West Une ONLY Person Contacted

7. Place Auto Reclose switches for the toIIo1ng PCBs in MAN:

• 315 (Bus 25 Whiteville 230KV Breaker) N-i SRO
• 31 A (Bus 2A Whiteville 230 kV Breaker) N-I SRO
• 305 (Bus 2B Delco West Line 230 kV Breaker)
• 30A (Bus 2A Delco West Line 230 kV Breaker)
• 28B (Bus 25 Wallace 230 kV Breaker) N- I SRO
• 28A (Bus 2A Wallace 230 kV Breaker) N-i SRI)
• 27B (Bus 2B Town Creek 230 kV Breaker) N-I SRI)
• 27A (Bus 2A Town Creek 230 kV Breaker) N-i sro

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 20 of 72 PLANT ELECTRIC SYSTEM OPERATING 20P-50 PROCEDURE Rev. 137 Page 59 of 281 6.2.6 De-energizing The 230 kV Switchyard (continued) CAUTION PCB Supervisory switch must be in LOCAL before the associated PCB is operated from Panel XU-5 D

8. Place Supervisory switches for the following PCBs in LOCAL:

• 31 B (Bus 26 Whitevifle 230 kV Breaker) N-I SRO
• 306 (Bus 26 Delco West Line 230 kV Breaker)
• 286 (Bus 26 Wallace 230 kV Breaker) N-i SRO
• 27B (Bus 26 Town Creek 230 kV Breaker) N-i SRO
• 31A (Bus 2A Whiteville 230 kV Breaker) N-i SRO
• 30A (Bus 2A Delco West Line 230 kV Breaker)
• 28A (Bus 2A Wallace 230 kV Breaker) N-i SRO
• 27A (Bus 2A Town Creek 230 kV Breaker) N-i SRO

9. Open 316 (Bus 2B Whiteville 230 kV PCB) N-I SRO JO. Confirm 316 (Bus 2B Whiteville 230 kV PCB) is OPEN by observing the indicating lights N- I SRO

11. Open3lA(Bus2AWhiteville23OkVPCB) N-i SRO

12. Confirm 3IA (Bus 2A Whaeville 230 kV PCB) is OPEN by observing the indicating lights N- I SRO

13. Open 30B (Bus 26 Delco West Line 230 kV PCB) 14 Confirm 306 (Bus 26 Delco West Line 230 kV PCB) is OPEN by observing the indicating lights

15. Open 30A (Bus 2A Delco West Line 230 kV PCB)

16. Confirm 30A (Bus 2A Delco West Line 230 kV PCB) is OPEN by observing the indicating lights

17. Open 286 (Bus 26 Wallace 230 kV PCB) N-i SRO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 21 of72 PLANT ELECTRIC SYSTEM OPERATING 20P-50 PROCEDURE Rev 147 Page 60 of 281 6.2.6 De-energizing The 230 kV Switchyard (continued) 18, Confirm 28B (Bus 2B Wallace 230 kV PCB) is OPEN by observing the indicating lights N-i SRO

19. open 28A (Bus 2A Wallace 230 kV PCB) N-I SRO

20. Confirm 28A (Gus 2A Wallace 230 kV PCB) is OPEN by observing the indicating lights N-i SRO

21. open 27B (Bus 2B Town Creek 230 kV PCB) N- I SRO

22. Confirm 27B (Bus 2B Town Creek 230 RV PCB) is OPEN by observing the indicating lights N-i SRO

23. Open 27A (Bus 2A Town Creek 230 kV PCB) .................,................ N-i SRO

24. Confirm 27A (Bus 2A Town Creek 230 kV PCB) is OPEN by observing the indicating lights N-i SRO NOTE It work is to be performed on a 230 kV bus, the manual disconnects are to be opened 0

25. Place Supervisory switches for the following PCB5 in REMOTE:

• 31 B (Bus 2B Whiteville 230 kV Breaker) , ,,,.,.,, ,.,,,.....,,,,,,,

                                                                                              ,,,, N-i SRO

• 3DB (Bus 2B Delco West Line 230 kV Breaker)
• 28B (Bus 2B Wallace 230 kV Breaker) N-i SRO
• 27B (Bus 2B Town Creek 230 kV Breaker , N-i SRO
• 31A (Bus 2A Whiteville 230 kV Breaker) N-i SRO
• 30A (Bus 2A Delco West Line 230 kV Breaker)
• 2$A (Bus 2A Wallace 230 kV Breaker) N-i SRO
• 27A (Bus 2A Town Creek 230 kV Breaker) N-i SRO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 J Page 22 of 72 PLANT ELECTRIC SYSTEM OPERATING 20P-50 PROCEDURE Rev 147 Page 61 01281 6.2.6 De-energizing The 230 kV Switchyard (continued) Date/lime Completed Pei1omed By (Print) Initials Reviewed By Unit CRS/SRO N-i, Partial usage to isolate only the Delco West 230 kV Line (Line 30)

2016 NRC SCENARIO 2 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 23 of 72 EVENT 3: ROD DRIFT Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger I to drift CR 26-1 1 into the core. When CR 26-1 1 is inserted to 00, Initiate Trigger 2 to activate CRD High Temperature alarm. If control rod is scrammed, verify the rod drift malfunction deletes. Two minutes after control rod is disarmed or scrammed, delete CRD HYD TEMP HIGH alarm. If asked to disarm CRD 26-11 Initiate Trigger 4. Simulator Operator Role Play If contacted as the RE to address thermal limits, acknowledge the request. When contacted for scramming control rod 26-1 1, report that Thermal Limits will NOT be exceeded by this single rod scram. If asked as the RBAO to investigate HCU for control 26-11, report that the HCU scram outlet riser is hot to the touch. When contacted as the RBAO and after high temperature alarm has been actuated, report that the CRD temperature is 390°F and slowly rising. When contacted as the System Engineer report that based on past history of this rod (26-11) scram times cannot be guaranteed. If asked as the RBAO to disarm control rod, coordinate with Sim Operator after 5 minutes. If requested, close/reopen the 113 valve (Charging Header Isolation Valve) as necessary As RBAO, Report Accumulator pressure 980# after rod has been scrammed. Evaluator Notes Plant Response: Control Rod 26-1 1 will drift full in. Crew should enter AOP-02.0 and take action lAW 2APP-A-05 (3-2). When the high temperature alarm is received, Engineering will report that scram times cannot be assured based on past history of the control rod. Determine TS 3.1.3 condition Olin 3 hours and 02 within 4 hours. Objectives: SRO - Direct actions in response to a drifting control rod and evaluate Tech Specs. RO Respond to a drifting control rod. Success Path: The drifting control rod is fully inserted, determined that the control rod must be placed under clearance and electrically disarmed. Event Termination: Go to Event 4 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 24 of 72 EVENT 3: ROD DRIFT Time Pos EXPECTED Operator Response Comments SRO Direct actions of 2APP-A-05 (3-2) ROD DRIFT Direct entry into OAOP-02.0, Control Rod SRO Malfunction/Misposition. After System Engineer reports that the scram times cannot be guaranteed, according to Note 2 in TS Table 3.1.4-1 the rod must be declared inoperable. Tech Spec 3.1.3 Control Rod Operability SRO Condition C. One or more control rods inoperable for reasons other than Condition A or B Required Action C.1 Fully insert inoperable control rod (3 hrs) C.2 Disarm the associated CRD_(4_hrs) Contact System Engineer on high temperature condition of control rod. SRO Contact RE to inform of rod drift and to evaluate thermal limits May direct the control rod to be scrammed to attempt to reseat the leaking outlet valve SRO lAW A-05 (3-2) ROD DRIFT May conduct a brief (See Enclosure 1, page 62 for format of the brief. Monitor reactor plant parameters during evolution. May read APP actions for the OATC to perform

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 25 of 72 EVENT3: ROD DRIFT Time Pos EXPECTED Operator Response Comments Acknowledge alarms: A-05 (2-2) ROD OUT BLOCK ATC A-05 (3-2) Rod Drift Announce and enter OAOP-02.0, Control Rod Malfunction/Misposition. Perform the actions of APP-A-05 (3-2) ROD DRIFT as follows: LI Determine which control rod is drifting. LI Select the drifting control rod and determine direction of drift. LI Attempt to arrest the drift by giving a withdraw signal. LI If rod continues to drift in, apply an RMCS insert signal and fully insert to position 00. ATC LI Attempt to locate and correct the cause of the rod malfunction as follows: LI Check and adjust cooling water header pressure if required. LI Direct AC to check for leaking scram valve. LI May direct an AC to check HCU temperature on R018 temperature recorder (in the Rx Bldg.) Monitor core parameters, main steam line ATC radiation and off-gas activity. Perform 2OP-07 Section 6.3.17, Single Rod The examiner will prompt the Scram from RPS Test Panel. performer that the Green light is ON ATC and the control rod is fully inserted CRS will NA appropriate steps. when step 6.3.17.11 is performed.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 26 of 72 REACTOR MANUAL CONTROL SYSTEM OPERA11rK3 20P-O? PROCEDURE Rev. 1O Page N of 162 6.3.17 Single Rod Scram From RPS Test Panel Confirm the following initial conditions are mets All applicable prerequisites in Section 5.0 are met

• Attachment 1 has been reviewed
• Communications are established between RPS Test Panel and the Control Room
• Reactor Engineer recommends performance of this section and has determined Techniil Specification Thermal Limits will NOT be exceeded by this single rod scram Reactor Engineer

2. IF AT ANY TIME it becomes necessary to scram a single control rod for operability concerns THEN perform OPT-14.2.1, Single Rod Scram Insertion Times Test for that control rod

3. Obtain permission from the Unit CRS to perform this section -

CRS 4, Document applicable control rod to be scrammed in the space provided Control Rod 5, IF recommended by Reactor Engineering to support diagnostic data, THEN record the following

• Reactor pressure psig Applicable accumulator pressure:

psig

2016 NRC SCENARIO 2 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 27 of 721 REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-07 PROCEDURE Rev 10 Page 95 of 162 6.3.17 Single Rod Scram From RPS Test Panel (continued) BEGIN R.M. LEVEL R21R3 REACTIVITY EVOLUTION

6. Select applicable control rod at P603 ............................................

CV

7. Close C 12-1 13 (Charging Water Riser Isolation Valve) for the applicable control rod......

8. IF RWM scram time recording is recommended by Reactor En gin a or in g.

THEN perform the followtng

a. Have Reactor Engineering connect temporary scram time test cable to single rod scram interface box ilocated on terminal strip GM in P616-RMCS cabinet and route cable up to RPS Test Panel P610 in accordance wtth Attachment 12, (Reference Use) Test Cable Arrangement tor RWM Scram Recording Reactor Engineer (1) Insert black lead into NEUTRAL socket on the P610 test panel IV (2) Insert red lead into socket corresponding to control rod tobe testedat P610 IV

9. Monitor control rod position

10. IF AT ANY TIME the control rod does NOT fully scram after loring the scram test switch.

THEN immediately notify the Unit CRS to determine operability of the rod (Technical Specificatjn 3.1.3)

11. Using a currently licensed ROSRO, perform the following

a. Scram the applicable control rod by Iowenng the scram test switch on RPS Test Panel P610 to the scram (down) position CV

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 28 of 72 REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-0/ PROCEDURE Rev. 105 Page 96 of 162 6.3.17 Single Rod Scram From RPS Test Panel (continued)

b. WHEN the scrammed control rod is fully inserted OR 10 seconds have elapsed (whichever occurs first).

THEN return applicable scram test switch to the normal (up) IV

12. Confimi rod posdion display indicates °00 for scrammed rod and the GREEN Full In light is ON ..

13. IF control rod did NOT fully insert, THEN reference Technical Specifications for OPERABILITY.

CRS NOTE Holding Emergency Rod In Notch Override switch in EMERGENCY ROD IN position for a penod of time will flush any ingested crud from the dnve to help prevent double notching D

14. Hold the Emergency Rod In Notch Override switch in EMERGENCY ROD IN position for at least 15 seconds and record insert stall flow.

stall flow stali flow stali flow 1: 2 3:

15. Repeat Section 6.3.17 Step 14 two additional times END RM. LEVEL R21R3 REACTIVITY EVOLUTION

16. Slowly open applicable C 12-113 (Charging Water Riser Isolation Valve) I IV

11. Confimi assoaated accumulator pressure is greater than 955 psig

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 29 of 72 REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-01 PROCEDURE Rev. 105 Page 91 of 162 6.3.17 Single Rod Scram From RPS Test Panel (continued)

18. IF RWM scram time was recorded, THEN perform the following

a. Contact Reactor Lngineering to upload data Reactor Engineer

b. Remove temporary scram timing cables from P616 and P610 IV

c. Perform the following to delete RWM scram data buffers:

(1) Select SCRAM DATA screen on RWM Operator Display in the Control Room (2) Press DLLLTL softkey to delete scram data (3) Confirm SCRAM DATA screen displays

• ROD SCRAM TIMING FUNCTION: READY
• ROD SCRAM TIMING DATA NOT TRANSFERRED

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 30 of 72 EVENT 4: ADHR SECONDARY PUMP TRIP Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 5 to trip the running ADHR Pump. When informed to align 2A RCC pump to ADHR mode Initiate Trigger 6 If asked to throttle closed the RCC-V51 14, Initiate Trigger 7. When asked to re-open the RCC-V51 14, then adjust the remote to 1.0. Simulator Operator Role Play If directed to investigate the trip of RCC Pump D, report the pump is tripped on overcurrent. When directed to align RBCCW Pump 2A to ADHR mode lAW 20P-21 Section 6.3.16 (steps 2b through 2i) have Sim Op align pump to ADHR mode and inform BOP Op that the steps are complete. When contacted as RBAO report radiation monitor is aligned per 20P-21 Section 6.3.18 step 4. RCC-V5154 (Rad Monitor Bypass Standby Isolation Valve) is CLOSED RCC-V51 16 (Rad Monitor Bypass ADHR Isolation Valve) is OPEN RCC-V5115 (Rad Monitor Bypass Common Mode Isolation Valve) is OPEN When contacted report RCC-V51 14 (RBCCW Pump 2A ADHR Mode Discharge Valve) is throttled 90% closed. (20P-21 Section 6.3.18 Step 5a) When contacted report RCC-V51 14 (RBCCW Pump 2A ADHR Mode Discharge Valve) is full open. (20P-21 Section 6.3.18 Step 5c) Evaluator Notes Plant Response: The running ADHR Secondary Loop Pump (RCC Pump D) will trip. The crew will have to start RCC Pump C. Shutdown RCC Pump A. Re-align RCC Pump A for ADHR mode and then start the pump for ADHR. (AOP-38.0 will be entered). Objectives: SRO Direct swapping of RCC pumps and then direct starting of RCC Pump in ADHR Mode. RO - Swap RCC pumps, Place RCC Pump in ADHR Mode. Success Path: Standby ADHR Pump placed in service. Event Termination: Go to Event 5 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 31 of72 EVENT 4: ADHR SECONDARY PUMP TRIP Time Pos EXPECTED Operator Response Comments Direct entry into OAOP-38.0, Loss of Fuel Pool SRD Cooling Direct swapping of RBCCW pumps Start RBCCW Pump C, secure A. Direct alignment of RBCCW Pump A to ADHR Mode. Direct starting RBCCW Pump 2A in ADHR Mode. Direct I/C to investigate trip of RBCCW Pump 2D. May conduct a brief (see Enclosure 1 on page 62 for format)

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 32 of 72 EVENT 4: ADHR SECONDARY PUMP TRIP Time Pos EXPECTED Operator Response Comments RD Plant Monitoring Report trip of RBCCW Pump 2D (running in ADHR Mode) BOP UA-01 3-3, ADHR SECONDARY LOOP TROUBLE Announce and enter AOP-38.0, Loss of Fuel Pool Cooling Perform 20P-21, Section 6.3.10 (page 33)to swap RBCCW pumps. (Start C and secure A) Plant announcement for the start of 20 RCC Pump and securing of 2A RCC Pump. Perform 2OP-21, Section 6.3.16 (page 34)to align RBCCW Pump A into ADHR Mode. Direct RB AD to perform steps 2b through 2i. Step 3 is N/A Perform 2OP-21, Section 6.3.18 (page 37)to start RBCCW Pump A in ADHR Mode. Notifies E&C, starting ADHR pump Step 2 is N/A Step 3 is N/A Direct the RB AD to perform step 4 and 5a. Announce starting of RCC Pump 2A. Direct the RB AD to perform step 5c. May direct AD to ensure primary loop is operating lAW 2OP-13.1.

I 2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 33 of 72 REACTOR BUILDING CLOSED COOLING WATER 20-21 SYSTEM OPERATING PROCEDURE Rev. 93 Page 47 of 149 6.3.10 Transferring to the Standby RBCCW Pump RBCCW Mode-Ensure the following initial conditions ate met:

• Applicable prerequisites listed in Section 5,0, Prerequisites ate met
• RBCCW System in operation with two pumps aligned for RBCCW Mode in service 2, Start the standby RBCCW pump by placing the associated pump control switch in ON:
• RBCCWPUMP2A
• RBCCW PUMP 2B
• RBCCW PUMP 2C
• RBCCWPUMP2D

3. Secure the desired RBCCW pump by placing the associated pump control switch in OFF:

• RBCCW PUMP 2A
• RBCCW PUMP 2B
• RBCCW PUMP 2C
• RBCCWPUMP2D 4 IF a third RBCCW pump is aligned to RBCCW Mode, AN. RBCCW discharge header pressure has stabilized, THEN place the pump control switch in AUTO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 j Page 34 of 72 REACTOR BUILDING CLOSED COOLING WATER 2OP-2 I SYSTEM OPERATING PROCEDURE Rev, 93 Page 68 of 149 6.316 Alignment of RBCCW Pump from RBCCW Mode to ADHR Mode Ensure the following initial condition is met

• One RBCCW Heat Exchanger is aligned to ADHR Mode per Section 6314
• Key for RBCCW/ADHR Mode Selector Switch has been obtained from one of the following:

G Control Rm Key Locket key 98 O WCC Key Locker-key 167 or 168 ... NOTE RBCCW Pump 2A and RBCCW Pump 2D can support either RBCCW Mode or ADHR Mode. A Mode Selector Switch is located on the pump breaker and a wiiite ADHR Mode indicating light is on the RTGB. This switch determines which of the two header pressures (RBCCW or ADHR) will be monitored for the pump auto start on low header pressure when the pump control switch is placed in AUTO. When the Mode Selector Switch is placed in the ADHR Mode position, the white light is ON0ntheRTGB 0

2. IF aligning RBCCW Pump 2A to ADHR Mode, THEN perform the following

a. Ensure RBCCW Pump 2A control switch is in OFF b Close RCC-V32 (RBCCW Pump 2A RBCCW Suction)

c. Close RCC-V38 (RBCCW Pump 2A RBCCW Mode Discharge Valve)

d. Open RCC-V5105 (RBCCW Pump 2A ADHR Mode Suction Valve) e Open RCC-V51 14 (RBCCW Pump 24 ADHR Mode Discharge Valve) f Open RCC-V303 (RBCCW Pump 2A Casing Vent Valve)

g. WHEN a steady stream of water is present, THEN close RCC-V303 (RBCCW Pump 2A Casing Vent Valve) h Ensure 2-RCC-SS-7667 (Pump 2A RBCCWIADHR Mode Selector Switch) located at MCC 2XE. in ADHR

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 35 of 72 REACTOR BUILDING CLOSED COOLING WATER 20P-2i SYSTEM OPERATING PROCEDURE Rev. 93 Page 69 of 149 6.3.16 Alignment of RBCCW Pump from RBCCW Mode to ADHR Mode (continued) Remove key from 2-RCC-SS-7667 (Pump 2A RBCCW1ADHR Mode Selector Switch)

j. Confirm the ADHR white indicating light on the RTGB for RBCCW Pump 2A is ON

3. IF aligning RBCCW Pump 2D to ADHR Mode.

THEN perform the following a Ensure RBCCW Pump 2D control switch is in OFF b Close RCC-V5 107 (RBCCW Pump 2D RBCCW Mode Suction Valve)

c. Close RCC-V51 11 (RBCCW Pump 2D RBCCW Mode Discharge Valve)

d. Open RCC-V5104 (RBCCW Pump 2D ADHR Mode Suction Valve)

e. Open RCC-V5 113 (RBCCW Pump 2D ADHR Mode Discharge Valve)

t. Open RCC-V5 139 (RBCCW Pump 2D Casing Vent Valve)

g. WHEN a steady stream of water is present, THEN close RCC-V5 139 (RBCCW Pump 2D Casing Vent Valve)

h. Ensure 2-RCC-SS-7668 (Pump 20 RBCCW/ADHR Mode Selector Switch) located at MCC 2XD, in ADHR

i. Remove key from 2-RCC-SS-7668 (Pump 2D RBCCW/ADHR Mode Selector Switch)

j. Confirm the ADHR white indicating light on the RTGB for RBCCW Pump 2D is ON

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 36 0172 REACTOR BUILDING CLOSED COOLING WATER 20P-2 I SYSTEM OPERATING PROCEDURE Rev. 93 Page 70 of 149 6.3.16 Alignment of RBCCW Pump from RBCCW Mode to ADHR Mode (continued) NOTE ADHR Mode piping is placed either in Standby Mode or in service to ensure RBCCW circulation and proper chemistry control when NOT undergoing maintenance 0

4. Place ADHR in service per Section 6.3.18. Starting an RBCCW Pump ADHR Mode DateiTime Completed Performed By (Print) Initials Reviewed By Unit CRS/SRO

[. 2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 37 of 72 REACTOR BUILDING CLOSED COOLING WATER 20P-2 1 SYSTEM OPERATING PROCEDURE Rev. 93 Page 74 of 149 6.3.18 Starling an RBCCW Pump ADHR Mode

1. Ensure the following initial conditions are met:

• Designated RBCCW Pump is aligned to ADHR Mode per Section 63.16
• IF RBCCW pump is operating in ADHR Mode, THEN E&C notified commencing startup of ADHR Mode Person Notified NOTE
• RBCCW Pump 2A and RBCCW Pump 2D can support either RBCCW Mode or ADHR Mode. A Mode Selector Switch is located on the pump breaker and a white ADHR Mode indicating light is on the RTGB This switch determines which of the two header pressures (RBCCW or ADHR) will be monitored for the pump auto start on low header pressure when the pump control switch is placed in AUTO. When the Mode Selector Switch is placed in the ADHR Mode position, the white light is ON on the RTGB 0
• RBCCW Pump 20 will NOT auto re-start when power returns after a LOOP or bus under voltage condition with the control switch in ON or AUTO. The control switch must be placed in OFF/RESET prior to restarting the pump 0 CAUTION Two pump operation in ADHR Mode subjects RCC-V37 (RGCCW Pump lA Discharge Check Valve) and RCC-V51 10 (RBCCW Pump 10 Discharge Check Valve) to accelerated wear. This lineup is expected to be utilized only when maximum ADHR capacity is required [8.7.2] 0

2. IF desired to start a second pump aligned to ADHR Mode, THEN perform the following a Obtain concurrence from Engineering to start a second pump in the ADHR Mode Person Contacted b GotoStep5.b

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 38 of 72 REACTOR BUILDING CLOSED COOLING WATER 20P-2 I SYSTEM OPERATING PROCEDURE Rev 93 Page 75 of 149 6.3.18 Starting an RBCCW Pump ADHR Mode (continued)

3. IF the ADHR Mode has been shutdown tot greater than 72 hours OR maintenance has been pertomied.

THEN fill and vent the ADHR piping per Section 6.3.13

4. Ensure the follong valve alignment for system radiation monitoring:

• RCC-V51 54 (Rad Monitor Bypass Standby Isolation Valve) is CLOSED
• RCC-V5 116 (Rad Monitor Bypass ADHR Isolation Valve) is OPEN
• RCC-V51 15 (Rad Monitor Bypass Common Mode Isolation Valve) is OPEN

5. For the RBCCW pump aligned to ADHR Mode to be started, perform the following a Throttle 80% to 95% closed the associated pump discharge valve:

• RCC-V51 14 (RBCCW Pump 2A ADHR Mode Discharge Valve)
• RCC-V51 13 fRBCCW Pump 2D ADHR Mode Discharge Valve)

b. Start an RBCCW pump aligned to ADHR Mode by placing the associated pump control switch in ON:

• RBCCWPUMP2A
• RBCCWPUMP2D c lFthrotDed in Step5a THEN slov1y open the associated pump discharge valve
• RCC-V51 14 (RBCCW Pump 2A ADHR Mode Discharge Valve)
• RCC-V5 1 13 (RBCCW Pump 20 ADHR Mode Discharge Valve)

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 39 of 72 REACTOR BUILDING CLOSED COOLING WATER 20P-21 SYSTEM OPERATING PROCEDURE Rev. 93 Page 76 of 149 6.31$ Starting an RBCCW Pump ADHR Mode (continued) d Ensure a log entry is made stating RBCCW pumps are in service in the ADHR Mode and Engineering has been notified NOTE The normal parameters for Supplemental Spent Fuel Pool Cooling are provided in , Normal System Operation Parameters, Equipment manipulations to maintain these parameters are perfomied per 20P-. 13.1, Alternate Decay Heat Removal System Primary Loop Operating Procedure D

6. IF a Primary Loop pump is operating, THEN maintain Primary Loop flow per 2OP-1 3.1, Alternate Decay Heat Removal System Primary Loop Operating Procedure

7. Ensure Plant Process Computer setup as follows per OOP-55, Plant Process and ERFIS Computer Systems Operating Procedure:

• PPC U2RCCA1 11 point ENABLED
• PPC U2RCCAO95 Value Monitoring setup with the nominal flow values per Attachment 1 Section 2.5 for the number of RBCCW Pumps in ADHR Mode to provide audible alarms for ADHR secondary flow changes Date/Time Completed Performed By (Print) Initials Reviewed By Unit CRS/SRO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 40 of 72 EVENT 5: RECIRC LOOP B FLOW TRANSMITTER FAILURE Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 8 to activate Recirc Loop B Flow failure. Simulator Operator Role Play If contacted as l&C to investigate, acknowledge the request. After LCO entries have been determined and SRO is waiting for I&C, call as WCCSRO and request APRM 4 be placed in tripped condition to support I&C trouble shooting. The WOO will hang the status control tag paperwork. If asked to pull fuses (for TRM 3.3 actions, 2-012A-F1 Labeled ROD OUT BLOCK RELAYS 012A in P616 panel) acknowledge the request Evaluator Notes Plant Response: Flow reference off normal alarm, rod block and scram signal to all 4 voters Flow transmitter signals are displayed on PC display 845, and on individual NUMACs by selecting Input Status. Objectives: SRO - Determine LCO for APRM 4 inoperability and direct placing channel in trip. RO Respond To A Flow Unit/Transmitter Failure Per APP A-06 5-7. Success Path: ARPM 4 TS 3.3.1.1 declaration and placed in trip condition lAW 001-1 8. Event Termination: Go to Event 6 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 2 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 41 of 72 EVENT 5: RECIRC LOOP B FLOW TRANSMITTER FAILURE Time Pos EXPECTED Operator Response Comments Direct actions of APPs SRO Direct I&C to investigate Evaluate Tech Spec 33.1.1 Reactor Protection System Instrumentation TS 3.3.1.1, Function 2b, Required Action Al. With one or more required channels inoperable, place in trip condition in 12 hours Evaluate TRM 3.3 Control Rod Block Instrumentation TRM 3.3, Function la, Required Condition Al. With one of the required channels not operable - 24 hours to restore to operable. Refers to 001-18 for actions to place APRM 4 in a tripped condition. Direct APRM 4 mode selector switch placed in INOP to allow I&C troubleshooting. May conduct a brief (see Enclosure 1 on racie 62 for format)

2016 NRC SCENARIO 2 LOt SIMULATOR EVALUATION GUIDE Rev. 0 Page 42 of 72 EVENT 5: RECIRC LOOP B FLOW TRANSMITTER FAILURE Time Pos EXPECTED Operator Response Comments BOP Monitors the plant. May check back panel APRM indications. Acknowledges, refers to & reports annunciators A-6 2-8 APRM UPSCALE 3-8 APRM UPSCALE TRIP/INOP ATC 5-7 FLOWREFOFFNORMAL A-5 2-2 ROD OUT BLOCK 4-8 OPRM TRIP ENABLED Diagnose and report failure of APRM 4 Flow Transmitter Obtains key number 114 from the SRO key locker to place APRM 4 in trip. Places APRM mode selector switch in INOP lAW 001-18.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 43 of 72 EVENT 6: HEATER DRAIN DEAERATOR PUMP TRIP Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 9 to trip a Heater Drain Pump. Simulator Operator Role Play If contacted as AD to investigate, wait until pump is tripped and report over-current flags for all phases of 2A HDP 4KV breaker on Bus 2D If contacted as RE for reduced FW Temp, acknowledge any requests. If asked as l&C to investigate, acknowledge the request Evaluator Notes Plant Response: Heater Drain Pump 2A shaft seizes and trips on overcurrent. Heater Drain tank level will rise and the crew will throttle HD-V57 to stabilize HD Tank level. If the standby HDP is not started, REP suction pressure will lower during the transient requiring power reduction to stabilize Condensate/feedwater. Objectives: SRO Directs OAOP-23, Condensate/Eeedwater System Eailures, and possible OAOP-03.0, Positive Reactivity Addition, entry. RD Diagnose HDP pump trip and start the standby HDP. Success Path: 2C HDP started with HDD level recovered in the normal band. Event Termination: Go to Event 7 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 44 of 72 EVENT 6: HEATER DRAIN DEAERATOR PUMP TRIP Time Pos EXPECTED Operator Response Comments Direct annunciator response for UA-04: 4-10 HD PUMPATRIP 2-10 HD DEAERATOR LEVEL HIGH-LOW 3-10 HD DEAERATOR LEVEL HIGH TRIP SRO Direct entry into OAOP-23, Condensate/Feedwater System Failures Direct starting standby HDP. May direct 2AOP-3.0, Positive Reactivity Addition, entry if power rises due to the HDD Ext Trip. May direct monitoring of final feedwater temperature. May direct maintenance to investigate trip May conduct a brief (see Enclosure 1 on page 62 for format) RO Plant Monitoring May reduce power lAW OAOP-23 to stabilize reactor water level.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 45 of 72 EVENT 6: HEATER DRAIN DEAERATOR PUMP TRIP BOP Recognize and report annunciators: UA-04 4-10 HD PUMP A TRIP 2-10 HD DEAERATOR LEVEL HIGH-LOW 3-10 HD DEAERATOR LEVEL HIGH TRIP UA-06 1-7 BUS 2D 4 KVMOTOR OVLD Manually starts 2C HDP lAW APP or AOP. Enter and announce OAOP-23, Condensate/Feedwater System Failures. Monitors final feedwater temperature (FFWT) lAW 20 1-03.2 May open the HD-V57 to assist in HDD level recovery. Directs an AC to 4.16 KV Switchgear Bus 2D to investigate 2A HDP trip Verifies auto actions for HD DEAERATOR LEVEL HIGH TRIP, if it occurs.

1. Non-return valves (EX-Vi 1 and EX-V1 2) to deaerator close. (Only close if turbine load is below 500 MWe)

2. HDD Extraction Line B moisture removal valves (MVD-LV-266 and MVD-LV-267) open.

May reference 2OP-35 to recover MRVs following HDD level restoration.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page46ofZ2 EVENTS 7/819: STEAM LEAK IN DW ATWS I SLC SWITCH FAILURE I ARI RESET FAILURE Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger JO to activate small steam leak in DW. If requested to defeat Group I LL3, wait 2 minutes, and install jumpers (Trigger 11) If requested to install LEP-02, Section 2.3 jumpers, wait 5 minutes, and inform the SRO that the jumpers are installed (RPOO5F already active). Simulator Operator Role Play Acknowledge request as I&C to investigate failure of SLC switch. If requested as l&C to investigate the failure of the ARI reset failure, acknowledge the request. Evaluator Notes Plant Response: Most control rods will fail to insert on the scram. The crew will respond to the ATWS per EOP-01-ATWS. When SLC initiation is attempted, the switch positions will not work. The crew will enter LEP-03 and align for alternate boron injection using CRD. The scram cannot be reset due to failure of the ARI to reset. Objectives: SRO - Direct actions to control reactor power per EOP-01-ATWS.. RD Perform actions for an ATWS per EOP-01-ATWS. Success Path: Lower level to control power, inject SLC, insert control rods.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 47 of 72 EVENTS 71819: STEAM LEAK IN DW ATWS I SLC SWITCH FAILURE I ARI RESET FAILURE Time Pos EXPECTED Operator Response Comments Enter RSP and transition to ATWS. Direct mode switch to shutdown when steam flow < 3 Mlbs/hr. Direct ARI initiation. Direct Recirc Pumps Tripped. SRO Direct SLC initiation, then LEP-03, CRITICAL TASK #2 Alternate Boron Injection. Direct ADS inhibited. Direct RWCU isolation verification. Direct LEP-02, Alternate Rod Insertion CRITICAL TASK #1 Direct Group 10 switches to override reset. Direct terminate and prevent HPCI/Feedwater (CS/RHR when LOCA signal received) to CRITICAL TASK #3 lower level to 90 inches. When level reaches 90 inches, evaluate Table Q-2: If not met, establishes a level band of LL4 to +90 inches. If met, directs RPV injection to remain terminated. When Torus temperature is greater than 95° F, enters PCCP and directs Torus Cooling. (See Enclosure 5, page 68) Directs Drywell cooling restored per SEP-10. Direct injection established to maintain RPV level LL4 to TAF (or the level at which APRMs indicate downscale)

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 48 of 72 EVENTS 7/8/9: STEAM LEAK IN DW ATWS I SLC SWITCH FAILURE I ARI RESET FAILURE Time Pos EXPECTED Operator Response Comments Place mode switch to shutdown when steam RO flow < 3x1 6 lb/hr. Initiates ARI. Trips Recirc Pumps. Initiates SLC. CRITICAL TASK #2 Determines SLC switch failure. Directs LEP-03, Alternate Boron Injection Recognizes failure of SLC switch and reports to CRS. Monitor APRMs for downscale. Performs LEP-02, Alternate Rod Insertion.

                    .          ..     .                       CRITICAL TASK #1 Section 2.1, Initial Actions (see page 48)

Section 2.3, Reset RPS and Initiate a Manual Scram (see page 51) Section 2.4, Reactor Manual Control System (RMCS) (see page 54) May also perform Section 2.5, Increasing Cooling Water Header Pressure (see page 56). Recognizes failure of ARI to reset, informs CRS

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 49 of 72 EVENTS 7/8/9: STEAM LEAK IN DW ATWS / SLC SWITCH FAILURE I ARI RESET FAILURE Time Pos EXPECTED Operator Response Comments BOP Places ADS in inhibit. Places Group 10 switches to override! reset Terminate and prevent injection to RPV. CRITICAL TASK #3 Terminates and prevents HPCI lAW Hard Card. (Enclosure 2, page 63) Terminates and Prevents Feedwater lAW Hard Card._(Enclosure_3,_page_64) May place HPCI in service for level control during ATWS when directed by the SRO. (Enclosure 6, page 70) Restart REP to maintain level as directed by SRO. (Enclosure 4, page 65) When Torus temperature is greater than 95° F, places Torus Cooling in service. (Enclosure 5, page 68)

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 50 of 72 ALTERNATE CONTROL ROD INSERTION OEOP-O 1 -LEP-02 Rev. 029 Page 4 of 37 1.0 ENTRY CONDITIONS

• As cFrected by Emergency Operatng Procedures fEOPs)
• As directed by Severe Accident Management Guideline (SAMGs) 2.0 OPERATOR ACTIONS 2.1 Initial Actions 2.1.1 Manpower Required
• 1 Reactor Operator
• 1 Auxiliary Operator 2.1.2 Special Equipment None NOTE
• Two-handed operation is allowed during implementation of this procedure in order to minimize delay in control rod insertion 0
• Section 2.1 .3 Step 1 through Step 6 may be performed concurrently with the rest of this procedure 0
• The system designation Cl 1 is for Unit 1 and C12 for Unit 2 0 2.1.3 Operator Actions

1. Monitor reactor power on APRMs until IRM recorders on scale 0 RO

2. Insert IRMs and monitor reactor power on IRM recorders 0 RO 3 Downrange IRMs to bring them on scale 0 RO

4. WHEN lRMs on Range 30R below.

THEN insert SRMs 0 RO

5. Monitor reactor period 0 RO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 51 of72 ALTERNATE CONTROL ROD INSERTION OEOP-0 1 -LEP-02 Rev. 029 Page 5 of 37 2.1.3 Operator Actions (continued)

6. Monitor control rod position using

• Process computer ... . U RO
• SPDS U RO
• RWM U RO
• Four rod U RO
• Full core dspIay U RO

7. WHEN either:

• All control rods in U RO
• Qjjy one control rod .!I4QI fully inserted U RO
• NO more than 10 control rods withdrawn to position 02 AND NO control rod withdrawn beyond position 02 U RO
• Reactor engineenng has determined the reactor will remain shutdown under all conditions without boron U RO THEN perform Section 2.2. Control Rod Insertion Verification on Page 7 U RO

2016 NRC SCENARIO 2 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 52 of 72 ALTERNATE CONTROL ROD I NSERTION OEOP-0 1 -LEP-02 Rev. 029 Page 6 of 37 2.1.3 Operator Actions (continued)

8. Insert control rods by one or more methods.

• Section 2.3, Reset RPS and Initiate a Manual Scram on Page 15 0 RO
• Section 24, Reactor Manual Control System (RMCS) on Page 18 0 RO
• Section 2.5, Increasing Cooling Water Header Pressure on Page 20 0 RO
• Section 26, Scram lndvidual Control Rods on Page 22 0 RO
• Section 2.7, De-energize Scram Solenoids and Vent Scram Air Header on Page 26 0 RO
• Section 2.8, Venting Over Piston Area on Page 32 0 RO

H 2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 53 of 72 ALTERNATE CONTROL ROD INSERTION OEOP-01=LEP-027 Rev. 029 Page 15 of 37 2.3 Reset RPS and Initiate a Manual Scram 2.3.1 Manpower Required

• I Reactor Operator 2.3.2 Special Equipment
• RO Desk Locked Drawer 0 4jumpers(15, 16, 17 and 18) 0 2.3.3 Manual Scram Actions NOTE Section 2 3.3 Step 1 and Step 2 may be perfomled concurrently 0

1. IF an automatic scram signal present AND power available to RPS bus.

THEN install jumpers to bypass reactor scram:

• Jumper 15 in Panel H 12-P609, Terminal Board DD, from right side of Rise Cf 1A(C72A)-F WA to Terminal 4 of Relay C71A(C72A)-K12E 0 RO
• Jumper 16 in Panel Hi 2-P609, Terminal Board BB, from left side of Fuse C7IAfC72A)-F14C to Terminal 4 of Relay C71A(C72A)-K12G ., 0 RO
• Jumper 17 in Panel H12-P611, Terminal Board DDjrorn right side of Fuse C7iA(C72A)F 146 to Terminal 4 of Relay CZ1A(C72A)-K12F 0 RO
• Jumper 18 in Panel H12-P61 1, Terminal Board BB, from left side of Fuse C7 1A(C72A)-F14D to Terminal 4 of Relay C7 lA(C72A)-K12H 0 RO

2. Inhibit ARI:

a. Place Ci i(C12)-CS-5560 (ARI Auto/Manual Initiation Switch) toINOP 0 RO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 54 of 72 ALTERNATE CONTROL ROD INSERTION OEOP-O I -LEP-02 Rev. 029 Page 16 of 37 2.3.3 Manual Scram Actions (continued) Li Place and hold Cl 1fC12)-CS-5562 (ARI Reset) switch in RESET C RO C. WHEN 5 seconds have elapsed, THEN retease C RO d Confirm red TRIP light located above C I I(C12)-CS-5561 (ARI Initiation) OFF C RO

3. Ensure Disch Vol Vent & Drain Test switch in ISOLATE C RO

4. Confirm CLOSED:

• Cf 1fC12)-V139 (Disch Vol Vent VIv) C RO
• Cl 1fC12)-CV-F0i0 (Disch Vol Vent Vlv) C RO
• Cl lfCi2)-V 140 fDisch Vol Drain VIv) C RO
• Ci l(C12)-CV-F0li (Disch Vol Drain Vlv) C RO

5. Reset RPS C RO

6. IF either RPS A OR B can be RESET, THEN go to Section 2.3.3 Step 8 C RO

7. IF RPS CANNOT be RESET, THEN return to Section 2 1.3 Step 7 C RO

8. Place Disch Vol Vent & Drain Test switch to NORMAL C RO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 55 of 72 ALTERNATE CONTROL ROD INSERTION OEOP-0 1 -LEP-02 Rev. 029 Page 170137 2.3.3 Manual Scram Actions (continued)

9. Confirm OPEN

• Cl 1fC 12)-V139 (Disch Vol Vent Vlv) C RO
• Cl 1(C12)-CV-F0l0 fDisch Vol Vent Vlv) C RO
• Ci lfCi2)-V 140 (Disch Vol Drain Vlv) C RO
• C11fC12)-CV-F0il fDisch Vol Drain Vlv) C RO

10. WHEN the scram discharge volume has drained for approximately 2 minutes OR A-05 1-6, SDV Hi-Hi Level RPS Trip clears, THEN continue C RO 1 1. venting control rod over piston area per Section 2.8, THEN notify AO to secure venting prior to inserting a manual scram C RO

12. Manually scram the reactor C RO

13. IF control rods moved inward AND control rods NOT inserted to

            .Q  beyond   Position 00, THEN return to Section 2.3 3 Step 3                                        C RO 14   lFjjj control rods inserted to OR beyond Position 00 OR control rods did NOT move inward, THEN return to Section 2.1.3 Step 7                                        C RO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 56 of 72 ALTERNATE CONTROL ROD I NSERT1ON OEOP-0 1 -LEP-02 Rev. 029 Page 18 of 37 2.4 Reactor Manual Control System fRMCS) 2.4.1 Manpower Required 1 Reactor Operator 2.4.2 Special Equipment RO Desk Locked Drawer 0 Unit 1 Only: 1 5450 key for RWM U 0 Unit 2 Only: 1 5451 key for RWM C 2.4.3 RMCS Actions I IF a reactor scram sealed in, THEN ensure available CRD pumps operating C RO

2. Ensure Cl l(C12)-FC-R600 (CRD Flow Control) in MAN C RO

3. IF a CRD pump NOT operating, THEN:

a Close The in-service Cl lfCl2)-FOO2A(FOO2B) (Flow Control Vlv) C RO

b. Start one CRD pump C RO c Adjust Ci 1(Ci2)-FC-R600 (CR0 Flow Control) to greater than or equal to 30 gpm C RO

d. IF available, THEN start The second CR0 pump C RO

4. IFNO CRD pump can be started, THEN return to Section 21.3 Step 7. C RO

2016 NRC SCENARIO 2 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 57 of 72 ALTERNATE CONTROL ROD INSERTION OEOP-0l-LEP-02 Rev. 029 Page 19of37 2.4.3 RMCS Actions (continued)

5. Insert control rods with RMCS:

a. Throttle open Ci 1(C12)-FOO2AfEQO2B)(Flow Control VIv) until drive water differential pressure greater than or equal to 260 psid D RO

b. IF dnve water differential pressure less than 260 psid, THEN throttle closed C I ifCl2)-PCV-F003 (Drive Pressure Vlv) until drive water differential pressure greater than or equal to 260 psid 0 RO

c. Bypass RWM 0 RO

d. Insert control rods with Emergency Rod In Notch C)vernde switch 0 RO

6. WHEN jfl control rods inserted to beyond Position 00 CANNOT be inserted th RMCS, THEN return to Section 2.1.3 Step 7 0 RO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0

                                                                                   -   Page 58 of 72 ALTERNATE CONTROL ROD I NSERTION                                    OEOP-0 1 -LEP-02 Rev. 029 Page 20 of 37 2.5     Increasing Cooling Water Header Pressure 2.5.1   Manpower Required

• 1 Reactor Operator 2.5.2 Special Equipment None 2.5.3 Cooling Water Header Actions

1. IF a reactor scram sealed in, THEN ensure available CRD pumps operating C RO 2 if a CRD pump operating, THEN:

a Ensure Cl 1(C12)-FC-R600 (CRD How Control) in MAN C RO

b. Close The in-service C) i(C12)-EOO2A(F0026)(FIow Control VIv) C RO C. Start one CRD pump C RO d Adjust Ci 1fC12)-FC-R600 (CRD Flow Control) to greater than or equal to 30 gpm C RO

e. IF available, THEN start the second CRD pump C RO

3. IFNO CRD pump can be started, THEN return to Section 2.1.3 Step 7 C RO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 59 of 72 ALTERNATE CONTRC)L ROD INSERTION OEOP-0 1 -LEP-02 Rev. 029 Page 21 0137 2.5.3 cooling Water Header Actions (continued)

4. IF a reactor scram NOT sealed in, THEN maximize cooling water header pressure:

a. Ensure C1i(C12)-FC-R600(CRD Flow Control) in MAN and fully open the in service Cl 1(C12)-FOO2AfFOO2B)fFIow Control Vlv) D RO b Fully open Cl i(C12)-PCV-F003 (Drive Pressure Vlv) D RO

5. WHEN aN control rods inserted to OR beyond Position 00 control rods NOT moving inward, THEN return to Section 2.1.3 Step 7 U RO

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 60 of 72 TERM I NATI ON Simulator Operator Actions When directed by the Lead Evaluator, delete the following commands: Malfunction K2624A, ARI Reset Malfunction K2625A, ARI INOP Malfunction RPO1 1 F, ATWS 4 (Make sure RPS is reset and scram air header pressurized before deleting) When directed by the Lead Evaluator, place the simulator in FREEZE DO NOT RESET THE SIMULATOR PRIOR TO RECEIPT OF CONCURRENCE TO DO SO FROM THE LEAD EXAMINER Simulator Operator Role Play After Sim Operator has deleted SDV malfunction, Inform the CRS that a loose wire was found on ARI switch and it has been repaired. Evaluator Notes Plant Response: When actions are taken to control reactor water level during the ATWS after terminating and preventing, ARI will be repaired and rods can be inserted. Objectives: SRO - Directs actions for an ATWS. RO Insert control rods lAW LEP-02. Success Path: Rods inserted with LEP-02, Alternate Rod Insertion. Scenario Termination: When all rods are inserted and level is being controlled above TAF with injection established, the scenario may be terminated. Remind students not to erase any charts and not to discuss the scenario until told to do so by the evaluatorlinstructor.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev. 0 L Page 61 0172 TERMINATION Time Pos EXPECTED Operator Response Comments Exit ATWS and enter RVCP when all rods are SRO in. Direct level restored to 170 200 inches after rods are all in. RO Confirms ARI reset when reported fixed. Inserts a scram after discharge volume has CRITICAL TASK #1 drained for 2 minutes. Reports all rods in. Maintains reactor pressure as determined by the SOP CRS. Maintains level as directed by the SCO. Restores level to 170 200 inches after all rod inserted. (Enclosure 4, page 65, contains actions for restart of Condensate and Feedwater)

L 2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 62 of 72 ENCLOSURE 1 Page 1 of 1 CONDUCT OF OPERATIONS AD-OP-ALL-i 000 Rev. 6 Page 90 of 90 ATTACHMENT 8 Page 1 of 1

                        <   Crew   Brief Template>>

El Announce Crew Brief Begin Briet El All crew members acknowledge announcement ( A: Required) El Update the crew as needed. El Describe what happened and major actions taken El Procedures In-progress El Notifications: Recap El Maintenance El Engineering El Others (Dispatcher, Station Management, etc.) El Future Direction and priorities El Discuss any contingency plans (A: Required) El Solicit questionsconcerns from each crew member: El ROs Input [I CR5 El STA El Are there any alarms unexpected for the plant conditions? El What is the status of Critical Parameters? (A: Required) EAL El Provide EAL and potential escalation criteria El Restore normal alarm announcement2 (Yes/No) Fini:h Brief El Announce End of Brief

I 2016 NRC SCENARIO 21 LOl SIMULATOR EVALUATION GUIDE I Rev.0I

                                   - -. .      L____.           *_ag*63of72 ENCLOSURE 2                                Page 1 of 1 SECURING HPCI INJECTION 1.0   INITIAL CONDITIONS WHEN DIRECTED BY ZOP4 1-LPC TO TEIa1INATL AND PREVENT HPCI INJECTION. OR     . -                      -  ..... 0

2. WHEN DIRECTED BY OEOP-O1-RXrP TO UTERMINATE AND PREVENT HPCI INJECTION. OR - ...... ..... 0

3. WHEN PERMISSION GIVEN BY THE UNIT CR5 TO SECURE HPCI INJECTION WITH A HPCI ALW0 START SIGNAL PRE5EJ4 - - - -....- -... a 10 PROCEDURAL STEPS

1. IF HPCI IS NOT OPERATING, PERFORM THE rOLL.OwING:

a PLACE HPCI AUXILLkRY OIL PUMP CONTROL SWITCH IN PULL-TO-LOCK - ..... 0 2.IFHPCI IS OPERATING. PERFORM THE rOLL0wING:

b. DEPRESS AND HOLD THE HPCI TURBINE TRIP PUSHBUTTON - 0 c WHEN HPCI TURBINE SPEED IS 0 RPM. AND HPCI TURBINE CONTROL VALVE, E41-V9 B CLOSED, THEN PLACE HPCI AUXILIARY OIL PUMP CONTROL SWITCH IN PULL-TO-LOCK. -... 0

d. WHEN HPCI TURB BRG OIL PRESS LO, 40142,15 SEALED IN. THEN RELEASE THE HPCI TURBINE TRIP PUSHBUTTON. - - ..... 0

a. ENSURE HPCI TURBINE STOP VALVE, E41-V8, AND HPCI TURBINE CONTROL VALVE, E41-V9.

REMAN CLOSED, AND HPCI DOES NOT RESTART. ..... - .. - ..... 0 1 211368 811369

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page64of72 ENCLOSURE 3 Page 1 of I Terminating and Preventing Injection From Condensate and Feedwater During EOPs (20P-32) IF desired TRIP all operatmg REPs.

2. IF one or more REPS are in service IDLE one RFP as follows:

a. IF two REPS are operating THEN TRIP one.

b. PERFORM either of the following for the operating REP 1 PLACE MAN/DECS control switch to tiAN.

2. RAPIDLY REDUCE speed to approximately 1000 rpm with the LOWERJRAISE speed control sitch.

OR I PLACE REPT Speed Control in M (MANUAL)

2. SELECT DEM and RAPIDLY REDUCE speed to approximately 2550 rpm.

3. CLOSE the following valves:

           -      FW HTR 5A OUTLET VLVS, FW-V6
           -      EW HTR 5B OUTLET VLVS, FW-V8 OR FW HTR 4A INLET VLV, FW-V1 18
          -       FWHTR4BINLETVLV,EW-V119

4. ENSURE the SULCV is closed by perfomiing the toIIowing

a. PLACE SULCV, in M (Manual) E

b. SELECT DEM and DECREASE signal until VALVE DEM indicates 0%

5. ENSURE FW-V 120, is closed.

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 65 of 72 ENCLOSURE 4 Page 1 of 3 Aligning Condensate and Feedwater After Terminating and Preventing Ensure FW-EV-177 (Feedwater Recirc to Condenser Vlv) CLOSED.,....... 0

2. Ensure EW Control Mode Select in 1 ELEM 0 3 Ensure at least one valve OPEN 0

• B21-F032A (Eeedwater lsol Vlv) 0
• B21-F032B (Feedwater Isol Vlv) 0

4. IFNO REP operating, THEN: 0

a. Ensure REPT A(B) Sp Ctl:

(1) lnM(rnanual) 0 (2) Prnp A(B) Dern at 0.0 PCT 0

b. Place FW-FV-46(47) [REP (NB) Recirc VIv] En OPEN 0

c. Ensure: 0

• RV-V3fV4) [REP (NB) Disch Vlv) OPEN 0
• REP A(B) Manual/DECS control switch in MANUAL 0 d Depress 0 (1) Reactor Water Level High Reset A 0 (2) Reactor Water Level High Reset B 0 (3) Reactor Water Level High Reset C 0 (4) REP A(B) Reset 0

e. Confirm OPEN 0

• REPA(B)LPStopVlvs 0
• REP A(B) HP Stop Vlvs 0 f Depress REP A(B) REPT Start 0

g. WHEN at rpm, THEN raise REP A(S) to at least 2550 rpm 0

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 66 of 72 ENCLOSURE 4 Page2of3 Aligning Condensate and Feedwater After Terminating and Preventing (continued)

5. IE desired to transfer REP At B) to DECS.

THEN a Ensure speed at least 2550 rpm 0

b. Depress DECS Ctrl Reset 0

c. Place Manua[IDECS control switch in DECS 0

6. Raise REP A( B) speed until discharge pressure approximately 100 psig above RPV pressure band 0 0/1550 5/1372

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 67 of 72 ENCLOSURE 4 Page 3013 Injection After Terminating and Preventing Condensate and Feedwater WHEN RPV injection directed. THEN as needed 0

• Adjust SULCV Valve Dern 0
• Throttle FW-V120 (FW Htrs 4&5 Byp Vlv) 0

2. WHEN automatic control desired, THEN 0

a. Confirm RPV level greater than +170 inches 0 b Ensure FW-V120 (EW Htrs 4&5 Byp Vlv) CLOSED 0 c Open RY-Vi0 (FW Recirc To Cond Isol VIv) 0 d Adjust SULCV to between 25 PCT and 55 PCI using 0

• SULCV Valve Dem 0
• FW-FV-177 (Eeedwater Recirc To Condenser Vlv) 0

e. Ensure Mstr REPT SpfRx Lvi Ctl 0

1) In M (manual) 0 (2) Level Setpoint at current RPV level 0

f. Place SULCV in A (automatic) 0

g. Adjust as needed to control RPV level: 0

• Mstr REPT SpfRx Lvi Ctl Level Setpoint 0
• FW-FV- 177 fFeedwater Recirc To Condenser Vlv) 0 011551 Sf1552

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 68 of 72 ENCLOSURE 5 Page 1 of 2 ATTACHMENT 8A Page 1 of I Emergency Suppression Pool Cooling Using Loop A f2OP-17) NOTE: This attachment is NOT to be used for normal system operations START RHR SW A LOOP (CONV) START RHR SW A LOOP (NUC) OPEN SW-VIOl OPEN SW-V105 CLOSE SW-V 743 OPEN SW-V102 START CSW PUMPS AS NEEDED E CLOSE SW-Vf43 IF LOCA SIGNAL IS PRESENT THEN START PUMPS ON NSW HDR AS NEEDED PLACE RHR SW BOOSTER PUMPS IF LOCA SIGNAL IS PRESENT THEN PLACE E A & C LOCA OVERRIDE SWITCH RHR SW BOOSTER PUMPS A & C LOCA 10 MANUAL OVERRIDE OVERRIDE SWITCH TO MANUAL OVERRIDE START RHR SW PMP START RHR SW PMP ADJUST El 1-PD V-F068A E ADJUST Eli-PD V-FO6BA ESTABLISH CLG WTR TO VITAL HDR ESTABLISH CLG WTR TO VITAL HDR START ADDITIONAL RHR SW PUMP START ADDITIONAL RHR SW PUMP AND ADJUST FLOW AS NEEDED AND ADJUST FLOW AS NEEDED START RHR LOOP A IF LOCA SIGNAL IS PRESENT, THEN VERIFY COOLING LOGIC IS MADE UP IF El l-FOJ5A IS OPEN, THEN E CLOSE E1l-FO17A START LOOP A RHR PMP OPEN E1l-F028A THROTTLE E71-F024A El THROTTLE E1i-F046A El START ADDITIONAL LOOP A RHR PMP E AND ADJUST FLOW AS NEEDED 2 211061 2 Si 1062

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0l Page 69 of 721 ENCLOSURE 5 Page 2of2 ATTACHMENT 8B Page 1 of 1 Emergency Suppression Pool Cooling Using Loop B (20P-17) NOTE: This attachment is NOT to be used for nornial system operations. START RHR SW B LOOP (NUC) START RHR SW B LOOP fCONV) OPEN SW-V105 OPEN SW-ViOl CLOSE $W-V143 OPEN SW-V102 START PMPS ON NSW HDR AS NEEDED CLOSE SW-V143 IF LOCA SIGNAL IS PRESENT THEN START CSW PUMPS AS NEEDED PLACE RHR SW BOOSTER PUMPS IF LOCA SIGNAL IS PRESENT THEN PLACE B & D LOCA oVERRIDE SWITCH RHR SW BOOSTER PUMPS B & D LOCA TO MANUAL oVERRIDE OVERRIDE SWITCH TO MANUAL OVERRIDE START RHR SW PMP START RHR SW PMP ADJUST Eli-PD V-F0688 ADJUST Eli-PD V-F0688 ESTABLISH CLG WTR TO VITAL HDR ESTABLISH CLG WTR TO VITAL HDR START ADDITIONAL RHR SW PUMP START ADDITIONAL RHR SW PUMP AND ADJUST FLOW AS NEEDED AND ADJUST FLOW AS NEEDED START RHR LOOP B IF LOCA SIGNAL IS PRESENT. THEN VERIFY COOLING LOGIC IS MADE UP IF El 7-E0158 IS OPEN THEN E CLOSE Eii-F0178 START LOOP B RHR PMP OPEN El l-F0288 THROTTLE El l-F024B THROTTLE Ell-F0488 START ADDITIONAL LOOP B RHR PMP E AND ADJUST FLOW AS NEEDED 2 2/1063 2 Sf1064

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 70 of 72 ENCLOSURE 6 Page 1 of 1 HPCI INJECTION IN EOPS

1. IF HPCI IS TRIPPED ON HIGH WATER LEVEL, DEPRESS HIGH WATER LEVEL SIGNAL RESET, E41-S25, PUSH BUTTON, AND ENSURE THE INDICATING LIGHT IS OFF.

2. ENSURE AUXILIARY OIL PUMP IS NOT RUNNING 3 ENSURE E41-V9 AND E41-V8 ARE CLOSED 4 OPEN E41-F059

5. PLACE HPCI FLOW CONTROL, E4 1-FIC-R600, IN MANUAL fM).

AND ADJUST OUTPUT DEMAND TO APPROXIMATELY MIDSCALE, USING THE MANUAL LEVER.

6. START VACUUM PUMP AND LEAVE IN START

7. OPEN E41-F001

8. START AUXILIARY OIL PUMP AND LEAVE IN START

9. OPEN E41-F006, IMMEDIATELY AFTER E41-V8 HAS DUAL INDICATION 10 ENSURE E4I-V9 AND E41-V8 ARE OPEN

11. WHEN SPEED STOPS INCREASING, THEN ADJUST SPEED TO APPROXIMATELY 2100 RPM

12. ADJUST HPCI FLOW CONTROL, E41-FIC-R600, TO OBTAIN DESIRED FLOW RATE

13. ENSURE E41-F012 IS CLOSED WHEN FLOW IS GREATER THAN 1400 GPM 14 ADJUST HPCI FLOW CONTROL, E4I-FIC-R600, SETPOINT TO MATCH SYSTEM FLOW, AND THEN PLACE E4 1-FIC-R600 IN AUTO (A)

15. ENSURE E41-F025 AND E41-F026 ARE CLOSED

16. START SBGT (OP-JO)

17. ENSURE BAROMETRIC CNDSR CONDENSATE PUMP IS OPERATING

2016 NRC SCENARIO 2 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 71 of72 ATTACHMENT I - Scenario Quantitative Attribute Assessment NUREG 1021 Category Scenario Content Rev. 2 Supp. I Req. Total Malfunctions 5-8 7 Malfunctions after EOP 1-2 2 Entry Abnormal Events 2-4 2 Major Transients 1-2 1 EOPs Used 1-2 2 EOP Contingency 0-2 2 Run Time 60-90 mm 90 Crew Critical Tasks 2-3 3 Tech Specs 2 2 Instrument I Component 2 OATC Failures before Major 2 BOP Instrument I Component 2 2 Failures after Major Normal Operations 1 1 Reactivity manipulation 1 1

LOl SIMULATOR EVALUATION GUIDE Page 72 of 72 ATTACHMENT 2 Shift Turnover Brunswick Unit 2 Plant Status Station Duty Workweek E. Neal B. Craig Manager: Manager: Mode: I Rx Power: 100% Gross*/Net MWe*: 977/ 951 Plant Risk: Green Current EOOS Risk Assessment is: SFP Time to 49.7 hrs Days Online: 82 days Turnover: Protected 2A FPC Pump/Hx, 2A RCC Pump, and 2C Demin Transfer Pump for Equipment: Fuel Pool Decay Heat Removal and inventory makeup. 2A/B NSW Pumps due to 1A NSW pump maintenance. IA NSW Pump is under clearance for planned maintenance. APRM 2 has failed downscale and is bypassed. 2C TCC Pump is in service on Unit One. Comments: The OATC reduce power to 850 MWe Gross (reactivity plan is to use recirc flow) The BOP operator will then Isolate 230 kV Delco West (Line 30) lAW the marked up of 20P-50, Section 6.2.6.

PLANT ELECTRIC SYSTEM OPERATING I 20P-50 I PROCEDURE I Rev. 147 I Page 58 of 281 6.2.6 De-energizing The 230 kV Switchyard

1. Ensure the Unit 2 230 kV switchyard is ENERGIZED AD

2. Ensure the 4kV Auxiliary Electrical Systems are DE-ENERGIZED in accordance with Section 6.2.3 N-I SRO

3. Ensure the SAT is DE-ENERGIZED in accordance with Section 6.2.4 N-I SRO

4. Ensure Caswell Beach Pumping Station is DE-ENERGIZED in accordance with Section 6.2.5 N-I SRO

5. Ensure required LCOs for Technical Specification Sections 3.8.1, 3.8.2, 3.8.7 and 3.8.8 are initiated SRO

6. Obtain Load Dispatchers permission to de-energize the 230 kV switchyard AD A Powers the Delco West Line ONLY Person Contacted

7. Place Auto Reclose switches for the following PCBs in MAN:

• 31 B (Bus 2B Whiteville 230 kV Breaker) N-I SRO
• 31A (Bus 2A Whiteville 230 kV Breaker) N-i SRO
• 30B (Bus 2B Delco West Line 230 kV Breaker)
• 30A (Bus 2A Delco West Line 230 kV Breaker)
• 28B (Bus 2B Wallace 230 kV Breaker) N-i SRO
• 28A (Bus 2A Wallace 230 kV Breaker) N-I SRO
• 27B (Bus 2B Town Creek 230 kV Breaker) N-I SRO
• 27A (Bus 2A Town Creek 230 kV Breaker) N-I SRO

PLANT ELECTRIC SYSTEM OPERATING 2OP-50 PROCEDURE Rev. 147 Page 59 of 281 6.2.6 De-energizing The 230 kV Switchyard (continued) CAUTION PCB Supervisory switch must be in LOCAL before the associated PCB is operated from Panel XU-5 D

8. Place Supervisory switches for the following PCBs in LOCAL:

• 3iB (Bus 2B Whiteville 230 kV Breaker) N-i SRO
• 30B (Bus 2B Delco West Line 230 kV Breaker)
• 28B (Bus 2B Wallace 230 kV Breaker) N-i SRO
• 27B (Bus 2B Town Creek 230 kV Breaker) N-i SRO
• 31A (Bus 2A Whiteville 230 kV Breaker) N-i SRO
• 30A (Bus 2A Delco West Line 230 kV Breaker)
• 28A (Bus 2A Wallace 230 kV Breaker) N-i SRO
• 27A (Bus 2A Town Creek 230 kV Breaker) N-i SRO

9. Open 31 B (Bus 2B Whiteville 230 kV PCB) N-i SRO i 0. Confirm 3i B (Bus 2B Whiteville 230 kV PCB) is OPEN by observing the indicating lights N-i SRO Ii. Open 3iA (Bus 2A Whiteville 230 kV PCB) N-i SRO

12. Confirm 31A (Bus 2A Whiteville 230 kV PCB) is OPEN by observing the indicating lights N-i SRO

13. Open 30B (Bus 2B Delco West Line 230 kV PCB)

14. Confirm 30B (Bus 2B Delco West Line 230 kV PCB) is OPEN by observing the indicating lights

15. Open 30A (Bus 2A Delco West Line 230 kV PCB) i6. Confirm 30A (Bus 2A Delco West Line 230 kV PCB) is OPEN by observing the indicating lights

17. Open 28B (Bus 2B Wallace 230 kV PCB) N-i SRO

PLANT ELECTRIC SYSTEM OPERATING 20P-50 PROCEDURE Rev. 147 Page 60 of 281 6.2.6 De-energizing The 230 kV Switchyard (continued)

18. Confirm 28B (Bus 2B Wallace 230 kV PCB) is OPEN by observing the indicating lights N-i SRO

19. Open 28A (Bus 2A Wallace 230 kV PCB) N-i SRO

20. Confirm 28A (Bus 2A Wallace 230 kV PCB) is OPEN by observing the indicating lights N-I SRO

21. Open 27B (Bus 2B Town Creek 230 kV PCB) N-i SRO

22. Confirm 27B (Bus 2B Town Creek 230 kV PCB) is OPEN by observing the indicating lights N-i SRO

23. Open 27A (Bus 2A Town Creek 230 kV PCB) N-I SRO

24. Confirm 27A (Bus 2A Town Creek 230 kV PCB) is OPEN by observing the indicating lights N-i SRO NOTE If work is to be performed on a 230 kV bus, the manual disconnects are to be opened D

25. Place Supervisory switches for the following PCBs in REMOTE:

• 31 B (Bus 2B Whiteville 230 kV Breaker) N-i SRO
• 30B (Bus 2B Delco West Line 230 kV Breaker)
• 28B (Bus 2B Wallace 230 kV Breaker) N-I SRO
• 27B (Bus 2B Town Creek 230 kV Breaker N-i SRO
• 3iA (Bus 2A Whiteville 230 kV Breaker) N-i SRO
• 30A (Bus 2A Delco West Line 230 kV Breaker)
• 28A (Bus 2A Wallace 230 kV Breaker) N-i SRO 27A (Bus 2A Town Creek 230 kV Breaker) N-I SRO

PLANT ELECTRIC SYSTEM OPERATING 20P-50

                . PROCEDURE Rev. 147 Page 61 of 281 6.2.6    De-energizing The 230 kV Switchyard (continued)

Date/Time Completed Performed By (Print) Initials Reviewed By Unit CRSISRO N-i, Partial usage to isolate only the Delco West 230 kV Line (Line 30)

DUKE ENERGY. BRUNSWICK TRAINING SECTION OPERATIONS TRAINING INITIAL LICENSED OPERATOR SIMULATOR EVALUATION GUIDE 2016 NRC SCENARIO 3 PT-40.2.1 1, DWEDT FAILURE, VFD CELL BYPASS, NSW PUMP TRIP, CWIP PUMP TRIP, RWCU LEAK, SBGT START FAILURE, ED, ADS VLV FAILURE REVISION 0 Developer: dVo&t Date: C?/C7/20t6 Technical Review: Vatc4c Date: 9/12/2076 Validators: Vae/d Date: 09/07/16 ade Facility Representative: Date:

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 2 of 47 REVISION

SUMMARY

0 Scenario developed for 2016 NRC Exam.

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 30147 TABLE OF CONTENTS 1.0 SCENARIO OUTLINE 4 2.0 SCENARIO DESCRIPTION

SUMMARY

5 3.0 CREW CRITICAL TASKS 6 4.0 TERMINATION CRITERIA 6 5.0 IMPLEMENTING REFERENCES 7 6.0 SETUP INSTRUCTIONS 7.0 INTERVENTIONS 10 8.0 OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES 12 ATTACHMENT 1 - Scenario Quantitative Attribute Assessment 41 ATTACHMENT 2 Shift Turnover 47

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 4 of 47 1.0 SCENARIO OUTLINE Event MaIf. No. Type* Event Description 1 Perform PT-40.2.11, Main Generator Voltage Regulator N-BOP Manual And Automatic Operational Check 2 ZA41 1 DWEDT Pump failure 3 C-ATC VFD Cell Failure RCO53F C-CRS (TS)(AOP) 4 R-ATC Power maneuver 5 C-BOP NSW Pump 2B Trip (failure of standby to start) CWO19F C-CRS (TS)(AOP) 6 C-BOP CWIP Trip CWO39F C-CRS (AOP) M RWCU leak / Scram 7 RWO13F C SBGT Fails to start (AOP)(RSP)(SCCP) M ED 8 Ki 507A C Failure of 2 ADS valves to open (EDP)

                *(N)ormal    (R)eactivity,  (C)omponent or Instrument,   (M)ajor

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 5 of 47 2.0 SCENARIO DESCRIPTION

SUMMARY

Event Description Perform PT-40.2. 1 1, Main Generator Voltage Regulator Manual And Automatic 1 Operational Check. Annunciator A-04 1-1, Drywell Equip Drain Sump Lvl Hi, will annunciate and the 2 sumps will not auto start. One of the sump pumps will need to be manually started A power cell in VFD A will fail. Recirc Pump 2A speed will lower and a speed hold will initiate. Loop flows will be outside mismatch limits. 4 The crew will reset the speed hold and match loop flows. NSW Pump B will trip and the crew will start NSW Pump A. Since 1A NSW Pump is 5 out of service, Tech Specs will apply. Crew will enter OAOP-1 8.0, Nuclear Service Water System failure, and carry out appropriate actions. Circulating Water Pump 2A will trip on motor winding fault, and the standby 6 Circulating Water Intake Pump will be started. OAOP-37.0 will be entered due to lowering vacuum. A large un-isolable RWCU leak will occur. Crew will enter AOP-5.0 and SCCP. The 7 CRS should direct a SCRAM. SBGT train A will fail to auto start and should be manually started. Secondary containment conditions will worsen, forcing the CRS to direct an Emergency Depressurization (or Anticipation of Emergency Depressurization) due to 8 high water levels. If Anticipation is performed, the second area high water level will annunciate requiring the emergency depressurization. Two ADS SRVs will fail to manually open. The CRS should direct opening two additional SRVs.

L 2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page6of47 3.0 CREW CRITICAL TASKS Critical Task #1 Insert a reactor scram prior to any area reaching its Max Safe Operating Value Critical Task #2 Perform Emergency Depressurization when two plant areas exceed max safe operating water level. 4.0 TERMINATION CRITERIA When emergency depressurization has been performed and the reactor has been depressurized to <100 psig the scenario may be terminated.

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page7of47 5.0 IMPLEMENTING REFERENCES NOTE: Refer to the most current revision of each Implementing Reference. Number Title A-04, 1-1 DRYWELL EQUIP DRAIN SUMP LVL HI A-06, 3-1 RECIRC VFD A ALARM UNACK A-06, 4-5 RECIRC LOOP A ONLY OUT OF SERV OAOP-04.0 Low Core Flow 2OP-02, 6.1.3 Reactor Recirculation System Operating Procedure 6.2.1 6.3.4 UA-01, NUCLEAR HEADER SERV WTR PRESS-LOW UA-01,4-i0 NUCLEAR HDR SW PUMP B TRIP UA-18, 6-1 BUS E4 4KV MOTOR OVLD. OAOP-18.0

• NUCLEAR SERVICE WATER SYSTEM FAILURES UA-01, 1-7 CIRC WATER PUMP A TRIP OAOP-37.0 Low Condenser Vacuum UA-03, 2-7 AREA RAD RX BLDG HIGH OAOP-05.0 Radioactive Spills, High Radiation, And Airborne Activity UA-5, 4-6 SBGT SYS A FAILURE

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 8of47 6.0 SETUP INSTRUCTIONS

1. PERFORM TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 5, Checklist for Simulator Exam Security.

2. RESET the Simulator to IC-i 1.

3. ENSURE the RWM is set up as required for the selected IC.

4. ENSURE appropriate keys have blanks in switches.

5. RESET alarms on SJAE, MSL, and RWM NUMACs.

6. ENSURE no rods are bypassed in the RWM.

7. PLACE all SPDS displays to the Critical Plant Variable display (#100).

8. ENSURE hard cards and flow charts are cleaned up

9. TAKE the SIMULATOR OUT OF FREEZE

10. LOAD Scenario File.

11. ALIGN the plant as follows:

Manipulation Ensure 2C TCC pump is in service on Unit One. Loaded in Scenario File Ensure 2B NSW pump is running, 2A in standby

12. IF desired, take a SNAPSHOT and save into an available IC for later use.

13. PLACE a clearance on the following equipment.

Component Position

14. INSTALL Protected Equipment signage and UPDATE RTGB placard as follows:

Protected Equipment

1. 2A and 2B NSW pumps

2. 2A FPC Pump/Hx, 2A RCC Pump, and 2C Demin Transfer Pump.

15. VERIFY OENP 24.5 Form 2 (Immediate Power Reduction Form) for IC-il is in place.

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page9of4z

16. ENSURE each Implementing References listed in Section 7 is intact and free of marks.

17. ENSURE all materials in the table below are in place and marked-up to the step identified.

Required Materials OPT-40.2.1 1

18. ADVANCE the recorders to prevent examinees from seeing relevant scenario details.

19. PROVIDE Shift Briefing sheet for the CRS.

20. VERIFY all actions contained in TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 4, Simulator Training Instructor Checklist, are complete.

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev. 0 PagelOof47 7.0 INTERVENTIONS TRIGGERS Trig Type ID 1 Annunciator ZA411 [DRYWELL EQUIP DRAIN SUMP LVL HI] 2 Trigger Command did:k2115a 1TriggerCommand did:k2116a 4 DI Override K2721K [VFD A LOWER FAST] 4 Malfunction RCO53F [VFD A POWER CELL COMMUNICATION FAILURE] 5 Malfunction CWO19F [NUC SERVICE WATER PUMP MOTOR WINDING FAULT] 6 Malfunction CWO39F [CIRC WATER INTAKE PUMP MOTOR WINDING FAULT] 7 Malfunction RWO13F [RWCU BRK IN TRIANGLE ROOM 77] 9 Remote Function RW_ZVRWOO4M [G31-F004 OUTBOARD ISOLATION VALVE] 10 Annunciator ZUA1214 [SOUTH RHR RM FLOOD LEVEL HI-HI] Trig # trigger Text 2 K2115JBU [DRYWELL EQUIP DR PUMP A] 3 K2116JBU [DRYWELL EQUIP DR PUMP B] 9 K141OJCK [RWCU VLV G31-F004J 11 K61O1WOV [SBGT SYS A]

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 PageJJof47 MALFUNCTIONS MaIf Mult Current Target Rmp Description Actime_fDactime Trig ID ID Value ValueJtime RWO15F G31-F001 FAILURE TO AUTO CLOSE True True RWO16F G31-F004 FAILURE TO AUTO CLOSE True True VFD A POWER CELL COMMUNICATION RCO53F CELL Bi False True 00:00:01 4 FAILURE NUC SERVICE WATER PUMP MOTOR CWO19F B False True 5 WINDING FAULT CIRC WATER INTAKE PUMP MOTOR CWO39F A False True 6 WINDING FAULT RWO13F I RWCU BRK IN TRIANGLE ROOM 77 0.00 100.00 00:10:00 j 7 RWO17F G31-F001

• REAOR WTR CLEANUP VLV G31-F001 True True REMOTES RemfMuftiescption CC_1ACW4518 2CTBCCWPUMPUNITACIGNMENT 1 1 RW_ZVRWOO4M G31-F004 OUTBOARD ISOLATION VALVE ON OFF 9 PANEL OVERRIDES Tag ID Pos I Override Description Actime Dactime Trig K61O1B SBGT SYS A PREF ON OFF Q61O1ARV SBGT SYS A CONT PREF R 4 ON/OFF OFF ON K2115A DRYWELL EQUIP DR PUMP A OUT OFF ON K2115A DRYWELL EQUIP DR PUMP A NORM ON OFF K2116A DRYWELL EQUIP DR PUMP B OUT OFF ON K2116A DRYWELL EQUIP DR PUMP B NORM ON OFF
• K1505A AUTO DEPRESS VLV B21-FO13D OPEN OFF OFF K1511A AUTO DEPRESS VLV B21-FO3A OPEN OFF OFF K4B2OA NUC HDR SW PMP A DISCH VLV AUTO - ON - OFF K2721K VFD A LOWER FAST LOWER FAST OFF ON 00:00:01 4 Q2721LWF jVFDA LOWER FAST ON/OFF OFF JOFF ANNUNCIATORS Window Description Tagname AVal Actime Dactime Trig 3-1 DRYWELL EQUIP DRAIN SUMP LEAK HI ZA431 OFF OFF OFF 1-1 DRYWELL EQUIP DRAIN SUMP LVL HI z11 ON ON OFF 1 1-4 SOUTH RHR RM FLOOD LEVEL HI-HI ZUA1214 ON ON OFF 10

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 12 of 47 8.0 OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES EVENT 1: PT-40.2.11 Simulator Operator Actions Ensure Monitored Parameters is open and Scenario Based Testing Variables are loaded. Simulator Operator Role Play Acknowledge any requests for the Load Dispatcher. When asked the voltage regulator operation was smooth and in the same direction of of the rheostat. Evaluator Notes Plant Response: Objectives: SRO Directs BOP to perform PT-40.2. 11 BOP Performs PT-40.2.1 1 RO Monitor Balance of Plant Success Path: PT-40.2.11 is completed. Event Termination: When directed by the Lead Evaluator, go to Event 2.

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 13of47 EVENT 1: PT-40.2.11 Time Pos EXPECTED Operator Response NOTES SRO Conduct shift turnover shift briefing. Direct performance of PT-40.2. 1 1 May conduct a brief (see Enclosure 1 on page 45 for format) RD Monitors the plant Performs PT-40.2.1 1 BDP See attached procedure.

2016 NRC SCENARIO 3 LOt SIMULATOR EVALUATION GUIDE Rev.0 Page 14 of 47 MAIN GENERATOR VOL.TAGE REGULATOR MANUAL OPT-40.2 .11 AND AUTOMATIC OPERATIONAL CHECK I Rev.6 L Page4ofl3 1.0 PURPOSE The purpose of this test is to demonstrate the OPERABILITY of the voltage regulator transfer circuitry and exercise the regulator potentiometers. 2.0 SCOPE

1. This test is performed once every 92 days and demonstrates OPERABILITY of voltage regulator transfer circuitry and exercises the regulator potentiometers

2. This test may also be used to demonstrate proper operation of the voltage regulator potentiometer and transfer circuitry, after completion of maintenance.

3.0 PRECAUTIONS AND LIMITATIONS

1. Main generator Ioadng is within the limits of the Generator Reactive Capability Curve shown on Attachment 1, Estimated Capability Curve, and with a minimum of 20 MVAR (positive) C

2. This test is NOT performed if erratic operation of the voltage regulator is noted immediately prior to the performance of this test C

3. The Load Dispatcher is to be informed when the main generator automatic voltage regulator is .QI in service. Log entries are made documenting the notification. {9. 11) C 4.0 GENERAL INFORMATION None 5.0 ACCEPTANCE CRITERIA 1 This test may be considered satisfactory when the following criteria are met.

a DC regulator output variation is smooth and in the same direction as the rheostat movement. b AC regulator output variation is smooth and in the same direction as the rheostat movement

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page J5of 47 MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2. 1 1 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 5of 13 6.0 PREREQUISITES

1. Confirm Generator and Exciter System in operation in accordance with 1 (2)OP-27, Generator and Exciter System Operating Procedure

2. Confirm Plant Electrical System in operation in accordance with 1 f2)OP-50, Plant Electric System Operating Procedure

3. Confirm DC Electrical System in operation in accordance with 1 (2)OP-5 1, DC Electrical System Operating Procedure

4. Confirm 120 Volt AC UPS, Emergency, and Conventional Electrical Systems in operation in accordance with 1 (2)OP-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure

5. Confirm NO system load changes are anticipated

2016 NRC SCENARIO 3 LOt SIMULATOR EVALUATION GUIDE Rev.0 Page 16 of 47 MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.i 1 AND AUTOMATIC OPERATIONAL CHECK Rev 6 Page 6 of 13 7.0 INSTRUCTIONS 7.1 General 1 Obtain permission from Unit CRS to perform this test

2. Ensure all Prerequisites listed in Section 5 0 are met 7.2 Operate ZOCS (Gen Manual Volt Adj Rheo)

1. Ensure 43CS (Regulator Mode Selector) in AUTO

2. Station an operator at the Excitation Regulator and Control cubicle in the Turbine Building on the 70 ft elevation west to monitor regulator output during the following steps NOTE

• Section 7.2 Step 3 and Section 7.2 Step 4 are repeated as necessary to ensure proper operation/indication of the manual rheostat 0
• DC regulator output is locally monitored using D 1VM (D.C. Reg Output) 0

3. Raise 7OCS (Gen Manual Volt Adj Rheo) until the Upper Limit light comes ON NOTE The Intemied light will come ON during lowering of 7OCS (Gen Manual Volt Adj Rheo) and will remain ON after the Low Limit light is ON 0

4. Lower 7OCS (Gen Manual Volt Adj Rheo) until the Low Limit light comes ON

5. Using 7OCS (Gen Manual Volt Adj Rheo) on the RTGB, null Gen Volt Reg Duff Volt meter

6. IF D1VM (D.C. Reg. Output) variation was NOT smooth AND in the same direction as rheostat movement, THEN go to Section 7.3 Step 7

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 17 of47J MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-30.2 11 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 7 of 13 7.2 Operate ZOCS (Gen Manual Volt Adj Rheo) (continued)

7. IF DIVM (D.C. Reg. Output) variation was smooth in the same direction as rheostat movement, THEN perform the following {9. 1 .1)

a. Notify the Load Dispatcher the main generator voltage regulator is being placed in MANUAL Person Notified

b. Document the Load Dispatcher notification in the log

c. Place 43CS (Regulator Mode Selector) in MAN

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 18 of 47 MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2. ii AND AUTOtvIATIC OPERATIONAL CHECK Rev. 6 Page 8 of 13 7.3 Operate 9OCS (Gen Auto Volt Adi Rheo) NOTE

• Section 73 Step 1 and Section 7.3 Step 2 may be repeated as necessary to ensure proper operationlindication of the automatic rheostat 0 AC regulator output may be locally monitored using Al VM (kC Reg. Output) 0

1. Raise 9OCS (Gen Auto Volt Ad] Rheo) until the Upper Limit tight comes ON

2. Lower 9OCS (Gen Auto Volt Adj Rheo) until the Low Limit light comes ON

3. Null Gen Volt Reg 01ff Volt meter on the RTGB using 9OCS (Gen Auto Volt Adj Rheo)

4. IF Al VM (AC. Reg Output) variation was NOT smooth AND in the same direction as rheostat movement, THEN go to Section 7.3 Step 6

5. IF A1VM fA.C. Reg Output) variation was smooth AND in the same direction as rheostat movement, THEN perform the following {9. 1.1 }

a Place 43CS (Regulator Mode Selector) in AUTO

b. Notify the Load Dispatcher the main generator voltage regulator is in AUTOMATIC Petson Notified

c. Document Load Dispatcher notification in the log

6. IF extended manual voltage regulator operation becomes necessary, THEN coordinate with the Load Dispatcher to maintain minimum generator MVAR load and generator voltage in accordance with the System Operation section of 1 f2)OP-27, Generator and Exciter System Operating Procedtire 7 iF. either regulator output variation was !QI smooth ANP in the same direction as the rheostat, THEN prepare a W/R for the regulator

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0

                                                                                    -    Page19of47 MAIN GENERATOR VOLTAGE REGULATOR MANUAL                                          OPT-40.2 ii AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 9of 13 7.4     Restoration 1     Perform review of completed procedure sections to verify Section 5.0, Acceptance Criteria, for tests performed, have been met IV

2. if Acceptance Cntena is NOT met, THEN perform following:

a. Report any equipment found INOPERABLE or NOT meeting Acceptance criteria to Supervisor

b. Ensure CR has been initiated

3. Ensure required information has been recorded on Attachment 2, Certification and Review Form

4. Notify Unit CRS when this procedure is complete or found to be unsatisfactory

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 20 of 47 EVENT 2: DWEDT PUMP FAILURE Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger I to activate the DWED Sump Lvl Hi Annunciator. If the simulator is left in run the DWED Sump Lvi Hi Alarm will annunciate on its own after NOTE approximately 50 minutes. (The malfunctions will still work if it is allowed to annunciate) When either sump pump has been running for 30 seconds delete malfunction for the DWED Sump Lvl Hi Annunciator. Simulator Operator Role Play Acknowledge requests as I&C for troubleshooting DWED Sump Pump auto start failure. If asked, the last time the sumps were pump was 4 hours ago. Evaluator Notes Plant Response: Annunciator A-04 (1-1), Drywell Equip Drain Sump Lvi Hi. Objectives: RO Pump the DWEDT Success Path: Pumps the DWEDT. Event Termination: Go to Event 3 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 21 of47 EVENT 2: DWEDT PUMP FAILURE Time Pos EXPECTED Operator Response Comments Direct actions of APPs Direct RO to start DWEDS Pump, if asked. SRO Contact I/C for troubleshooting the failure of the DWEDS to auto start. Refer to APP: RO A-04 (1-I), Drywell Equip Drain Sump Lvl Hi Diagnose failure of DWEDS Pump Start a DWEDS Pump (may use OOP-47 Section 5.3.5) Verifies pump shuts off after a period of time. BOP Monitors the plant

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 220147 FLOOR AND EQUIPMENT DRAIN SYSTEM OOP-4 1 OPERATING PROCEDURE Rev, 28 Page 21 o156 5.3.5 Manually Pumping Drywell Floor Or Equipment Drain Sumps Ensure the following:

a. Drywell floor or Equipment Drain sump needs to be manually pumped to determine in-leakage rates OR

b. Drywell Floor or Equipment Drain sump needs to be manually pumped as detemined by the Unit CRS............,..

2. On Panel P603, place control switches for the applicable sump pump(s) in START AND then in AUTO:

• G16-COO1A (Diywell Floor Drain Pump 1(2)A
• G16-COO1B (Drywell Floor Drain Pump 1(2)B)
• G16-COO6A (Drywell Equip Drain Pump 6A)
• G16-0006B (Drywell Equip Drain Pump 6Ba DateTirne Completed Performed By (Print) Initials Reviewed By Unit CR$/SRO

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 23 of 47 EVENT 3/4: VFD A CELL FAILURE I MANEUVER POWER Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 4 to activate VFD A Cell failure. Simulator Operator Role Play If contacted as I&C to investigate, acknowledge the request. If asked as Reactor Engineer for guidance on restoring Loop flow limits, ask the CRS for their recommendations, then concur with that recommendation. Evaluator Notes Plant Response: A power cell in VFD A will fail. Recirc Pump 2B speed will lower and a speed hold will initiate. Loop flows will be outside mismatch limits. The crew will respond per AOP 04.0, reset the speed hold and match loop flows or lower the speed of 2B to get within Tech Spec limits. Objectives: SRO - Direct Shift Response To A Recirculation Flow Control Failure Causing A Decreasing Flow Per AOP-04.0 RO Respond To A Recirc Flow Control Failure Decreasing Per AOP-04.0 Success Path: Reset the speed hold condition and match recirc loop flows. Event Termination: Go to Event 5 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 24 of 47 EVENT 314: VFD A CELL FAILURE I MANEUVER POWER Time Pos EXPECTED Operator Response Comments SRO Direct entry into AOP-04.O With recirculation loops flows mismatched, enter LCO 3.4.1 Condition A. NOTE: May balance loops and not enter Tech. Specs. Question examinee about Tech Spec actions if not entered. TS 3.4.1 Condition A.1. Satisfy the requirements of the LCO within 6 hours by restoring matched flows or impose limits specified by the LCO. NOTE: Declare the loop with lower flow not in operation. Direct speed hold reset on VFD A Direct loop flow mismatch restored to within limit Direct l&C to investigate cell failure May conduct a brief (see Enclosure 1 on page 45 for format) BOP Monitors the plant. Determine cause to be cell failure at HMI

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 25 of 47 EVENT 314: VFD A CELL FAILURE I MANEUVER POWER Time Pos EXPECTED Operator Response Comments Reference applicable APPs: RO A-06, 3-1, Recirc VFD A Alarm Unack A-06, 4-5, Recirc Loop A Only Out Of Serv Recognize/report lowering Recirc A speed/speed hold Enter/announce 2AOP-04.0, Low Core Flow Determine Loop flow outside mismatch limits Core flow >57.5 MIbs, Jet Pump flows must be within 3 MIbs. Reset speed hold on VFD A lAW 20P-02 Section 6.3.4. (see page 26) Restore loop flows to within limits as directed by CRS. Lower the B Recirc Pump Speed lAW 20P-02 Section 6.2.1. (see page 27) Raise the A Recirc Pump Speed lAW 20P-02 Section 6.1.3. (see page 28)

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 26 of 47 REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev. 168 Page 66 of 250 6.3.4 Recovery From Recirc VFD Speed Hold Condition 1 Confirm Recfrc VED A(B) Speed Hold yellow light ON at Panel P603

2. Ensure the cause of the Speed Hold condition has been identified

3. Ensure Plant conditions have stabilized

4. Check the following parameters are approximately the same:

• Recirc Pump A( B) Speed Demand
• Recirc Pump A(B) Actual Speed
• Recitc Pump Af B) Calculated Speed

5. Depress Recirc VED A(B) SP Hold Reset to reset the speed hold condition

6. Confirm Recitc VED AtE) Speed Hold yellow status light is OFF

7. Check flow conditions stable END R.M. LEVEL R2/R3 REACTIVITY EVOLUTION

8. Adjust Recirc VFD speed and Recirc flow as directed by the Unit CRS END R.M. LEVEL R21R3 REACTIVITY EVOLUTION Datemme Completed Performed By (Pnnt) Initials Reviewed By:

Unit CRS/SRO

2016 NRC SCENARIO 3] LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 27 of 47 REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev. 168 Page 4501 250 6.2 Shutdown 6.2.1 Lowering Speed/Power Using lndwidual Recirculation Pump Control Or Recirc Master Control

1. Confirm reactor recirculation pump in operation in accordance with Section 6 1.2 NOTE

• Recirculation Pump speed changes are performed wten directed by OGP-05, Unit Shutdown, and 0GP- 12, Power Changes Other operating procedures are used simultaneously with this procedure as directed by OGP-05, Unit Shutdown, and OGP-12, Power Changes U
• Speed changes are accomplished by depressing Lower Stow. Lower Medium, or Lower Fast pushbuttons. The Lower Slow pushbutton changes Recirc pump speed at 0.06%/decrement at 1 rpm/second The Lower Medium pushbutton changes Recirc pump speed at 0.28%/decrement at 5 rpm/second The Lower Fast pushbutton changes Recirc pump speed at 2 8%/decrement at 100rpm/second U

2. IF AT ANY TIME any of the following conditions exist, THEN enter 1AOP-040, Low Core Flow.{8.1 .9)

• Entry into Region A of Power to Flow Map
• OPRM INOPERABLE any of the following 0 Entry into Region B of Power to Flow Map 0 Entry into 5% Buffer Region of Power to flow Map 0 Entry into OPRM Enabled Region and incFcations of THI (Themial Hydraulic Instability) exist

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 28 of 47 REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev 168 Page 46 of 250 6.2.1 Lowering SpeedlPower Using Individual Recirculation Pump Control Or Recirc Master Control (continued) CAUTION

• The OPRM System monitors LPRMs for indication of thermal hydraulic instability (THI). When greater than or equal to 25% power and less than or equal to 60% recirculation flow, alarms and automatic trips are initiated upon detection of THI. Pump operations are governed by the limits of the applicable Power Flow Map, as specified in the COLR. {8. 1 .9) C
• Entry into the 5% Buffer Region warrants increased monitoring of reactor instrumentation for signs of Thermal Hydraulic Instability. Time in the 5% Buffer Region presents additional risk and is minirnized.{8,1.9} C
• Wh core flow less than 57.5 x 10 lbs/hr, jet pump loop flows are required within 10% (maximum indicated difference 6.0 x 10e lbslhr) With core flow greater than or equal to 57.5 x 10e lbslhr, jet pump loop flows are required within 5% (maximum indicated difference 3.0 x 1O Ibsihr) C
• When Recirc Pump speeds are less than 40%, decreasing speed using a Lower Fast pushbutton can result in a Speed Hold condition due to exceeding the regen torque limit C BEGIN R.M. LEVEL R21R3 REACTIVITY EVOLUTION

3. iF. desired to lower the speed of both recirculation pumps simultaneously, THEN depress Recirc Master Control Lower (Slow Medium Fast) pushbutton

4. iF. desired to lower the speed of an individual recirculation pump, THEN depress the Recirc VFD A(B) Lower (Slow Medium Fast) pushbutton

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 29 of 47 REACTOR RECIRCULATION SYSTEM OPERTlNG 20P-02 PROCEDURE Rev 168 Page 47 of 250 6.2.1 Lowering SpeedlPower Using Individual Recirculation Pump Control Or Recirc Master Control (continued)

5. Confirm the following, as applicable Recirc Pump A(B) Speed Demand. Calculated Speed, and Actual Speed have lowered

• Reactor power lowers
• B32-R617(R613) [Recirc Pump A(B) Discharge Flow] lowers
• B32-VFD-lDS003A(B) [Recirc VFD 2A( B) Output Wattrneterj lowers
• B32-VFD-IDS-OO1A( B) JRecirc VED 2Af B) Output Frequency Meter] lowers END R.M. LEVEL R21R3 REACTIVITY EVOLUTION Date/Time Completed Performed By (Print) Initials Reviewed By:

Unit CRS/SRO

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 30 of 47 REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev. 168 Page 38of 250 61.3 Raising SpeedlPower Using Individual Recirculation Pump Control or Recirc Master Control 1 Ensure the following Initial Conditions are met: a Reactor Recirculation Pumps in operation in accordance with Section 6.1 .2

b. Recirculation Pump flow limits are CLEAR NOTE

• Recirculation Pump speed changes are performed wtien directed by OGP-04, Increasing Turbine Load to Rated Power, and OGP12, Power Changes. Other operating procedures are used simultaneously with this procedure as directed by OGP-04, Increasing Turbine Load to Rated Power, OGP-12, Power Changes, or the Unit CRS 0
• Speed changes are accomplished by depressing Raise Slow or Raise Medium pushbuttons. The Raise Slow pushbutton changes Recirc pump speed at 006%/increment at 1 rpm/second. The Raise Medium pushbutton changes Recirc pump speed at 0.28%/increment at 5 rpm/second 0 CAUTION The OPRM System monitors LPRMs for indication of thermal hydraulic instability (THI). When greater than or equal to 25% power and less than or equal to 60%

recirculation flow, alarms and automatic trips are initiated upon detection of THI Pump operations are governed by the limits of the applicable Power Flow Map, as specified in the COLR. {8. 1.9) 0

2. IF AT ANY TIME any of the following conditions exist, THEN enter 2AOP-04.0, Low Core Flow {8 19)

• Entry into Region A of Power to Flow Map
• OPRM INOPERABLE AND any of the following 0 Entry into Region B of Power to Flow Map O Entry into 5% Buffer Region of Power to Flow Map O Entry into OPRM Enabled Region and indications of THI (Thermal Hydraulic Instability) exist

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 31 of 47 REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev 168 Page 39 of 250 6.1.3 Raising SpeedlPower Using Individual Recirculation Pump Control or Recirc Master Control (continued) CAUTION

• The OPRM System monitors LPRMs for indication of thermal hydraulic instability THl). When greater than or equal to 25% power and less than or equal to 60% recirculation flow, alarms and automatic trips ate initiated upon detection of THI. Pump operations is be within the limits of the applicable Power-Flow Map, as specified in the COLR. The Scram Avoidance Region is avoided. {8.1.9} El
• With core flow less than 57.5 x .10e Ibshr, jet pump loop flows are requited within 10% (maximum indicated difference 6.0 x l& Ibs/hr). With core flow greater than or equal to 57.5 x 10e lbs/hr. jet pump loop flows are required within 5% (maximum indicated difference 3.0 x 10 lbs/br) El
• If total reactor feedwater flow lowers to less than 16.4% of rated flow. Speed Limiter Number 1 v.iIl cause the Recirculation Pumps to mn back to 34%

speed This signal must be manually reset in accordance with Section 6.3.3 El

• When total core flow is greater than 43 rnlbThr, Speed Limiter Number 2 will cause a wnback to approximately 48% speed if reactor water level is less than 182 inches and either reactor feed pump A or B suction flow is less than 14 9%

of individual REP rated suction flow. This signal must be manually reset using Section 6.3.3 El BEGIN R.M. LEVEL R2/R3 REACTIVITY EVOLUTION

3. IF desired to raise the speed of both Recirc Pumps simultaneously, as directed by the Unit CR5, THEN depress Recirc Master Control Raise Slow or Raise Medium pushbutton

4. IF desired to raise the speed of an individual Recirc Pump, as directed by the Unit CRS, THEN depress the VFD A(B) Raise Slow or Raise Medium pushbutton for the Recirc Pump

5. Confirm the following, as applicable

• A rise in Recirc Pump Af B) Speed Demand. Calculated Speed, and a rise in Actual Speed
• A rise in Reactor power
• A rise in B32-R617(R613) [Recirc Pump A(B) Discharge Flow]

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page32of47 REACTOR RECIRCULkIION SYSTEM OPERATING 20P-02 PROCEDURE Rev 168 Page 40 of 250 6.1.3 RaIsing Speed/Power Using Individual Recirculation Pump Control or Recirc Master Control (continued)

• A nse in B32-VFD-IDS-OO3Af B) [Recirc VFD 2A(B) Output Wattmeterj .. .
• A nse in B32-VFD-IDS-O0 lAf B) [Recirc VFD 2Af B) Output Frequency Meter]

END R.M. LEVEL R21R3 REACTIVITY EVOLUTION Date/rime Completed Performed By (Print) Initials Reviewed By Unit CRS1SRO

2016 NRC SCENARIO 3 LOt SIMULATOR EVALUATION GUIDE Rev. 0 Page 33 of 47 EVENT 5: NSW PUMP B TRIP (FAILURE OF STANDBY TO START) Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 5 to trip the 2B NSW Pump. Simulator Operator Role Play If contacted as OAO to investigate NSW pump and breaker, After the pump has tripped report 51 devices on all three phases ate tripped at the breaker on E4 If contacted as l&C to investigate, acknowledge the request. Evaluator Notes Plant Response: The running NSW pump will TRIP on motor overload. The STBY NSW pump will fail to AUTO start. The BOP operator should recognize the failure and manually start the STBY NSW pump. With a Ui NSW pump under clearance will requite entry into TS. Objectives: SRO - Direct actions for loss of NSW Determine actions required for LCO per Technical Specifications RO Respond to the failure of an automatic start of the A NSW pump Success Path: Determine TS required actions and Start 2A NSW Pump. Event Termination: Go to Event 6 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 34 of 47 EVENT 5: NSW PUMP B TRIP (FAILURE OF STANDBY TO START) Time Pos EXPECTED Operator Response Comments Direct entry into OAOP-18.0, NSW System SRO Failure. Contact maintenance to investigate trip of 2B NSW Pump. May also report to I/C that 2A NSW Pump did not auto start. Evaluate Tech Spec 3.7.2 Service Water System and Ultimate Heat Sink.

               . Determine 2B NSW pump inoperable
               . Determine 1A NSW Pump inoperable due to clearance.
               . Pet the Bases, 3 NSW pumps required site wide.
               . 3.7.2 Condition B. One required NSW pump inoperable for reasons other than condition A. Required Action B.J Restore required NSW pump to Operable status in 7 days May direct 2C CSW pump to be placed on the NSW header.

May conduct a brief (see Enclosure 1 on page 45 for format)

2016 NRC SCENARIO 3 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 35 of 47 EVENT 5: NSW PUMP B TRIP (FAILURE OF STANDBY TO START) Time Pos EXPECTED Operator Response Comments Monitor reactor plant parameters during ATC evolution. Acknowledge / reference BOP UA-1 8 (6-1) BUS E4 4KV MOTOR OVLD Recognize trip of 2B NSW pump and lowering NSW system pressure. Announce and execute OAOP-18.0, NSW System Failure. Recognize the failure of the STBY NSW pump to start and starts standby pump.

               . Places 2A NSW pump in Manual.
               . Starts 2A NSW Pump.

Refer to alarms.

               . UA-01 (1-10) NUCLEAR HEADER SERV WTR PRESS-LOW
               . UA-01 (4-10) NUCLEAR HDR SW PUMP B TRIP
               . UA-05 (1-9) FAN CLG UNIT CS PUMP RM A INL PRESS LO
               . UA-05 (2-9) FAN CLG UNIT CS PUMP RM B INL PRESS LO May align the 2C CSW pump to the NSW header.

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 36 of 47 EVENT6: CWIPTRIP Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 6 to activate CW Pump A trip Simulator Operator Role Play If asked as Outside AO, acknowledge request to check pump. After 2-3 minutes, call back and report that shear pin on the traveling screens for CW Pump A broke. If asked as TBAO, identify that breaker AB8 on 4160 V Switchgear 2C is tripped on overcurrent. No other abnormalities. If asked as I&C to investigate, acknowledge the request If asked for prestart checks for the 2C CWIP, report prestart checks are SAT. If asked to verify no personnel are around the 2C Bus, report all clear. Evaluator Notes Plant Response: Circ Water Pump A will trip and annunciator UA-01, 1-7, CIRC WATER PUMP A TRIP, will alarm. After investigating the cause of the alarm, another Circ Water Pump should be started lAW the APP. Objectives: SRO Direct actions of APP-UA-01, 1-7, CIRC WATER PUMP A TRIP Direct Emergency Depressurization BOP Perform action of APP UA-01, 1-7, CIRC WATER PUMP A TRIP RO Monitor plant parameters Success Path: Another Circ Water pump is be started. Event Termination: Go to Event 7 at the direction of the Lead Evaluator

2016 NRC SCENARIO 3 LOt SIMULATOR EVALUATION GUIDE Rev. 0 Page 37 of 47 EVENT 6: CWIP TRIP Time Pos EXPECTED Operator Response Comments Direct actions of APP-UA-01, 1-7, CIRC WATER SRO PUMP A TRIP. May direct entry into enter OAOP-37.0, Loss Of Condenser Vacuum May direct power lowered to 90% May conduct a brief on when Reactor Scram is required (see Enclosure 1 on page 45 for format) ATC Plant Monitoring May lower power as directed by the CRS. (See page 27) BOP Take actions lAW APP-UA-01, 1-7, CIRC WATER PUMP A TRIP (see page 38) NOTE: CW SQL VALVES MODE SELECTOR SWITCH will need to be placed into position D to start C CWIP May announce and enter OAOP-37.0, Loss Of Condenser Vacuum Direct AOs to investigate pump and pump breaker to determine cause of pump trip.

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 38 of 47 Unit 2 APP UA-O1 1-7 Page 1 01 3 CWPUMPATRIP AUTO ACTIONS

1. CW Pump A trips CAUSE

1. Instantaneous overcurrent

2. Time overcurrent

3. Phase overcurrent

4. Differential overcurrent or phase angle (lockout relay)

5. Condenser pit flood level hi-hi

6. Low lube water flow

7. High traveling screen A dP (48 in. water) AND screen A stopped

8. High traveling screen A UP (48 in. water) AND high screen B, C or D dP (18 in. water)

9. LOCA Load Shed

10. Unit Trip Load Shed

11. Circuit malfunction OBSERVATIONS

1. Condenser vacuum decreasing (process computer points T000, TOO;,

Recorder OG-PR-23 on XU-2, and l-OG-Pl-23-iA, -2A on XU-80)

2. Generator output decreasing

3. Local relay indication at the breaker compartment

4. Circulating water discharge temperature increasing (BOP typer)

5. CW PUMP LUBE WATER FLOW-LOW (UA-Ot 5-7) alarm 6, TURB BLDG NW CNDSR PIT FLOOD LVL HI fUA-28 6-6) alarm

7. TURB BLDG E CNDSR PIT FLOOD LVL HI (UA-28 6-5) alarm

8. TURB BLDG SW CNDSR PIT FLOOD LVL HI (UA-28 6-7) alarm

9. CW SCREEN DIFF HI HI (UA-Oi 1-4) alarm

10. CW SCREEN A DIFF HIGH OR STOPPED (UA-O1 1-5) alarm

11. CW SCREEN B DIFF HIGH OR STOPPED fUA-O 1 2-5) alarm

12. CW SCREEN C DIFF HIGH OR STOPPED (UA-Ol 3-5) alarm

13. CW SCREEN D DIFF HIGH OR STOPPED fUA-0l 4-5) alarm ACTIONS

1. II a radioactive liquid release is in progress, terminate the release 2 II reactor power is less than 90% OR a CWIP pump can be started within 5 minutes, THEN START an available CWIP.

3. II reactor power is greater than 90% AND an available CWIP pump was NOT started within 5 minutes, then power must be reduced to 90 to 92%

prior to starting a CWIP 2APP-UA-O; Rev. 83 Page 15 of 109 I

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 39 of 47 EVENT 7: RWCU LEAK I SBGT FAILS TO START Simulator Operator Actions At the discretion of the lead evaluator, Initiate Trigger 7 to activate the RWCU Leak. Simulator Operator Role Play If contacted as engineering, acknowledge request for EQ envelopes for the U2 Reactor Building. If HPs contacted to perform field surveys acknowledge the request. If directed to reset breakers for the RWCU isolation valves, wait 2 minutes and report HP has restricted access to the reactor building. If directed to co-ordinate entry with the HPs, wait 15 minutes and report the breakers will not reset. Evaluator Notes Plant Response: A large un-isolable RWCU leak will occur. Crew will enter AOP-5.0 and SCCP. SRO should direct a SCRAM. Objectives: SRO - Direct response to un-isolable primary system breach in secondary containment. RO Respond to un-isolable primary system breach in secondary containment. Perform SCRAM actions. Success Path: Reactor scram is inserted before max norm operating value is exceeded. Event Termination: When a reactor scram is inserted and SCCP entered.

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 40 of 47 EVE NT 7: RWCU LEAK I SBGT FAILS TO START Time Pos EXPECTED Operator Response Comments Direct entry into OAOP-5.0, Radioactive Spills, SRO High Radiation, And Airborne Activity Direct RO to trip and isolate RWCU. Announce and enter SCCP procedure Direct a reactor manual scram prior to any area reaching its Max Safe Operating Value Critical Task #1 May direct a cool down at normal cool down rates (<100°F/hr). Request EQ envelopes for the U2 Rx Bldg Enter and execute RVCP. D Direct RO/BOP to stabilize reactor pressure below 1050 psig. D Verify Instrument operability per Caution 1. D Direct crew to not use NO26NB due to 50 temperatures after 50 alarm reported. D Direct verification of group isolations, ECCS initiations and DG starts as appropriate. D Direct RO/BOP to restore and maintain reactor water level 1 70-200 Recognize when alarm A-2 6-8, RB 20/50 FT ELEV TEMP HI, is reported that if 50 elevation is greater than 140°F that the Wide Range (N026) level indicators are inaccurate. Contact I/C for assistance with RWCU isolation valve failures

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 41 of 47 EVENT 7: RWCU LEAK I SBGT FAILS TO START Time Pos EXPECTED Operator Response Comments Insert Reactor scram as directed by CRS Critical Task #1 Depresses both of the manual scram pushbuttons. Place mode switch to shutdown when steam flow < 3x106 lb/hr. IF reactor power is below 2% (APRM downscale ATC trip), THEN TRIP the main turbine. ENSURE the master reactor level controller setpoint is +170. IF two reactor feed pumps are running, AND reactor vessel level is above 160 AND rising, THEN TRIP one. ATc/ Maintain reactor pressure as directed by CRS. BOP ATc/ Maintain reactor water level as directed by SRO. BOP

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 42 0147 EVENT 7: RWCU LEAK I SBGT FAILS TO START Time Pos EXPECTED Operator Response Comments Respond to UA-03 2-7, AREA RAD RX BLDG HI. Enter and execute OAOP-5.0, Radioactive Spills, High Radiation, And Airborne Activity. D Evacuate Unit 2 Reactor Bldg. D Direct AO to close PIV-33 RB Sprinkler BOP Shutoff Valve. D Direct E&RC to take applicable OAOP-5.0 actions. D Check area radiation readings at back panels. D Diagnose source of radiation as RWCU leak. Recognize and report to CRS alarm A-2 6-8, RB 20/50 FT ELEV TEMP HI. Responds to UA-5, 4-6, SBGT SYSA Failure Recognize failure of SBGT to start, places SBGT train A switches to start ATc/ Maintain reactor pressure as directed by CRS. BOP ATc/ Maintain reactor water level as directed by SRO. BOP

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 43 of 47 EVENT 8: EMERG DEPRESS I ADS VALVE FAILURE I TERMINATION Simulator Operator Actions 2 minutes after receiving Annunciator UA-12 (2-4) SOUTH RHR RM FLOOD HI, or when anticipation of emergency depressurization is performed, Initiate TRIGGER 10 (South RHR RM Flood HI-HI) When directed by the lead evaluator, place the simulator in FREEZE DO NOT RESET THE SIMULATOR PRIOR TO RECEIPT OF CONCURRENCE TO DO SO FROM THE LEAD EXAMINER Simulator Operator Role Play Evaluator Notes Plant Response: Secondary containment conditions will worsen, forcing the SRO to direct an Emergency Depressurization due to high water levels. Two ADS SRVs will fail to manually open. SRO should direct opening two additional SRVs. Scenario will end when reactor pressure reaches 1 00#. Objectives: SRO Evaluate plant conditions and direct an Emergency Depressurization. RO Performs actions for Emergency Depressurization. Success Path: ED has been performed. Scenario Termination: When emergency depressurization has been performed and the reactor has been depressurized to <100 psig the scenario may be terminated. Remind students not to erase any charts and not to discuss the scenario until told to do so by the evaluator/instructor.

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 44 of 47 EVENT 8: EM ERG DEPRESS I ADS VALVE FAILURE I TERMINATION Time Pos EXPECTED Operator Response Comments SRO Continue reactor cooldown per SCCP direction. Direct Emergency Depressurization when CRITICAL TASK #2 RHR RM FLOOD LEVEL HI-HI alarm (Two plant areas with water levels above Max Safe South CS and RHR) Direct RO/BOP to open 7 ADS valves. If informed by RO/BOP that 2 SRVs failed to open, direct opening additional SRVs until 7 SRVs are open. Enter PCCP when torus temperature exceeds 95°F. Directs all available loops to be placed in suppression pool cooling. Recognize and report South CS and South RHR A1C/ Room Flood Hi-Hi alarms. BOP Open seven ADS valves as directed by SRO. CRITICAL TASK #2 Recognize failure of 2 ADS valves to OPEN and report to SRO. Open 2 additional SRVs as directed by SRO. Maintain reactor water level as directed by SRO. Place available loops in suppression Pool Cooling lAW hard card. (see page 40)

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page45of47 ENCLOSURE 1 Page 1 of 1 CONDUCT OF OPERATIONS AD-OP-ALL-i 000 Rev. 6 Page 90 of 90 ATTACHMENT 8 Page 1 of 1

                         << CrQw Briof TompIatQ>>

D Announce Crew Brief B.gln BrIQI C AU ciew members acknowledge announcement (A RaquirGU) C Update the crew as needed C Descnbe what happened and major actions taken C Procedures In-progress C Notifications R.c3p C Maintenance C Engineering C Others (Dispatcher, Station Management. etc.) C Future Direction and pnonties C Discuss any contingency plans (A RQqutr.d) C Solicit questions/concerns from each crew member: C ROs Input C CR5 C STA C Are there any alarms unexpected for the plant conditions? C What is the status of Critical Parameters? (A R.qulred) EAL C Provide EAL and potential escalation criteria C Restore normal alarm announcement? (Yes/No) FlnIh Brief C Announce End of Bnef

2016 NRC SCENARIO 3 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 46 of 47 ATTACHMENT I - Scenario Quantitative Attribute Assessment Category Scenario Content Rev.2Supp.iReq. Total Malfunctions 5-8 7 Malfunctions after EOP Entry 1-2 2 Abnormal Events 2-4 4 Major Transients 1-2 2 EOPs Used 1-2 2 EOP Contingency 0-2 1 Run Time 60-90 mm Crew Critical Tasks 2-3 2 Tech Specs 2 2 Instrument I Component 2 OATC Failures before Major 2 BOP Instrument I Component Failures after Major 2 2 Normal Operations 1 1 Reactivity manipulation 1 1

LOl SIMULATOR EVALUATION GUIDE Page 47 of 47 1 ATTACHMENT 2 Shift Turnover Brunswick Unit 2 Plant Status Station Duty Workweek E Neal B Craig Manager: Manager: Mode: 1 Rx Power: 100% Mode: 1 Plant Risk: Green Current EOOS Risk Assessment is: SFP Time to 49.7 hrs Days Online: 80 days Turnover: Protected 2A FPC Pump/Hx, 2A RCC Pump, and 2C Demin Transfer Pump for Equipment: Fuel Pool Decay Heat Removal and inventory makeup. 2A/B NSW Pumps due to 1A NSW pump maintenance 1A NSW Pump is under clearance for planned maintenance. 2C TCC Pump is in service on Unit One. Comments: The BOP will perform PT-40.2.1 1, Main Generator Voltage Regulator Manual And Automatic Operational Check.

DUKE ENERGY Continuous Use BRUNSWICK UNIT 0 SURVEILLANCE TEST PROCEDURE OPT4O.2.1 I MAIN GENERATOR VOLTAGE REGULATOR MANUAL AND AUTOMATIC OPERATIONAL CHECK REVISION 6

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.1 1 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 2 of 13 REVISION

SUMMARY

PRR 00652903 DESCRIPTION This procedure revision upgraded the procedure to the new PAS template and AD-DC-ALL-0202, Writers Manual for Controlled Procedures Manual. Statements for new procedure Scope, General Information and Records sections were added in accordance with the Writers guide format for test procedures. Added Estimated Capability Curve graphic from 1 (2)OP-27 for operator efficiency. Addressed PRR 00686591 to reorder steps Section 7.3 Step 5 to place the voltage regulator in automatic then perform notifications. No technical changes were made to the procedure. Revised by ER. Sessoms.

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.11 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 3 of 13 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE 4 2.0 SCOPE 4 3.0 PRECAUTIONS AND LIMITATIONS 4 4.0 GENERAL INFORMATION 4 5.0 ACCEPTANCE CRITERIA 4 6.0 PREREQUISITES 5 7.0 INSTRUCTIONS 6 7.1 General 6 7.2 Operate 7OCS (Gen Manual Volt Adj Rheo) 6 7.3 Operate 90C5 (Gen Auto Volt Adj Rheo) 8 7.4 Restoration 9 8.0 RECORDS 10

9.0 REFERENCES

10 ATTACHMENT 1 Estimated Capability Curve 12 2 Certification and Review Form 13

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.11 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 4 of 13 1.0 PURPOSE The purpose of this test is to demonstrate the OPERABILITY of the voltage regulator transfer circuitry and exercise the regulator potentiometers. 2.0 SCOPE

1. This test is performed once every 92 days and demonstrates OPERABILITY of voltage regulator transfer circuitry and exercises the regulator potentiometers.

2. This test may also be used to demonstrate proper operation of the voltage regulator potentiometer and transfer circuitry, after completion of maintenance.

3.0 PRECAUTIONS AND LIMITATIONS

1. Main generator loading is within the limits of the Generator Reactive Capability Curve shown on Attachment 1, Estimated Capability Curve, and with a minimum of 20 MVAR (positive) D

2. This test is NOT performed if erratic operation of the voltage regulator is noted immediately prior to the performance of this test

3. The Load Dispatcher is to be informed when the main generator automatic voltage regulator is NOT in service. Log entries are made documenting the notification. {9.1.1} D 4.0 GENERAL INFORMATION None 5.0 ACCEPTANCE CRITERIA

1. This test may be considered satisfactory when the following criteria are met:

a. DC regulator output variation is smooth and in the same direction as the rheostat movement.

b. AC regulator output variation is smooth and in the same direction as the rheostat movement.

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.1 I AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 5 of 13 6.0 PREREQUISITES

1. Confirm Generator and Exciter System in operation in accordance with 1(2)OP-27, Generator and Lxciter System Operating Procedure

2. Confirm Plant Electrical System in operation in accordance with 1 (2)OP-50, Plant Electric System Operating Procedure

3. Confirm DC Electrical System in operation in accordance with 1 (2)OP-51, DC Electrical System Operating Procedure

4. Confirm 120 Volt AC UPS, Emergency, and Conventional Electrical Systems in operation in accordance with I (2)OP-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure

5. Confirm NO system load changes are anticipated

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.11 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 6 of 13 7.0 INSTRUCTIONS 7.1 General

1. Obtain permission from Unit CRS to perform this test

2. Ensure all Prerequisites listed in Section 5.0 are met 7.2 Operate 7OCS (Gen Manual Volt Adj Rheo)

1. Ensure 43CS (Regulator Mode Selector) in AUTO

2. Station an operator at the Excitation Regulator and Control cubicle in the Turbine Building on the 70 ft elevation west to monitor regulator output during the following steps NOTE

• Section 7.2 Step 3 and Section 7.2 Step 4 are repeated as necessary to ensure proper operation/indication of the manual rheostat C
• DC regulator output is locally monitored using D1VM (D.C. Reg. Output) C

3. Raise 7OCS (Gen Manual Volt Adj Rheo) until the Upper Limit light comes ON NOTE The Intermed light will come ON during lowering of 7OCS (Gen Manual Volt Adj Rheo) and will remain ON after the Low Limit light is ON C

4. Lower 7OCS (Gen Manual Volt Adj Rheo) until the Low Limit light comes ON

5. Using 7OCS (Gen Manual Volt Adj Rheo) on the RTGB, null Gen Volt Reg Duff Volt meter

6. IF D1VM (D.C. Reg. Output) variation was NOT smooth AND in the same direction as rheostat movement, THEN go to Section 7.3 Step 7

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.11 AND AUTOMATIC OPERATIONAL CHECK Rev 6 Page 7 of 13 7.2 Operate ZOCS (Gen Manual Volt Adj Rheo) (continued)

7. IF DIVM (D.C. Reg. Output) variation was smooth AND in the same direction as rheostat movement, THEN perform the following: {9.1.1}

a. Notify the Load Dispatcher the main generator voltage regulator is being placed in MANUAL Person Notified

b. Document the Load Dispatcher notification in the log

c. Place 43CS (Regulator Mode Selector) in MAN

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.11 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 6 of 13 7.3 Operate 9OCS (Gen Auto Volt Adj Rheo) NOTE . Section 7.3 Step 1 and Section 7.3 Step 2 may be repeated as necessary to ensure proper operation/indication of the automatic rheostat D . AC regulator output may be locally monitored using A1VM (A.C. Reg. Output) D

1. Raise 9OCS (Gen Auto Volt Adj Rheo) until the Upper Limit light comes ON

2. Lower 9OCS (Gen Auto Volt Adj Rheo) until the Low Limit light comes ON

3. Null Gen Volt Reg Duff Volt meter on the RTGB using 9OCS (Gen Auto Volt Adj Rheo)

4. IF Al VM (AC. Reg. Output) variation was NOT smooth AND in the same direction as rheostat movement, THEN go to Section 7.3 Step 6

5. IF Al VM (A.C. Reg. Output) variation was smooth AND in the same direction as rheostat movement, THEN perform the following: {9.1.1}

a. Place 43C5 (Regulator Mode Selector) in AUTO

b. Notify the Load Dispatcher the main generator voltage regulator is in AUTOMATIC Person Notified

c. Document Load Dispatcher notification in the log

6. IF extended manual voltage regulator operation becomes necessary, THEN coordinate with the Load Dispatcher to maintain minimum generator MVAR load and generator voltage in accordance with the System Operation section of 1 (2)OP-27, Generator and Exciter System Operating Procedure

7. IF either regulator output variation was NOT smooth AND in the same direction as the rheostat, THEN prepare a W/R for the regulator

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.11 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 9 of 13 7.4 Restoration

1. Perform review of completed procedure sections to verify Section 5.0, Acceptance Criteria, for tests performed, have been met IV

2. IF Acceptance Criteria is NOT met, THEN perform following:

a. Report any equipment found INOPERABLE or NOT meeting Acceptance Criteria to Supervisor

b. Ensure CR has been initiated

3. Ensure required information has been recorded on Attachment 2, Certification and Review Form

4. Notify Unit CRS when this procedure is complete or found to be u nsatisfactory

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.11 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 10 of 13 8.0 RECORDS Completed portions of this procedure are transmitted to QA records for retention per Quality Assurance Program requirements.

9.0 REFERENCES

9.1 Commitments

1. VAR-002-1.lb, Voltage and Reactive (VAR) Reliability Standard, Federal Energy Regulatory Commission 9.2 Technical Specifications None 9.3 Updated Final Safety Analysis Report None 9.4 Drawings None 9.5 Procedures

1. 1 OP-27, Generator and Exciter System Operating Procedure

2. IOP-50, Plant Electric System Operating Procedure

3. 1 OP-51, DC Electrical System Operating Procedure

4. JOP-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure

5. 2OP-27, Generator and Exciter System Operating Procedure

6. 2OP-50, Plant Electric System Operating Procedure

7. 2OP-51, DC Electrical System Operating Procedure

8. 20P-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure

MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.1 1 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 11 of 13 9.6 VendorlTechnical Manuals I FP-85650, Power System Stabilizer Equipment

2. FP-20183, Steam Turbine-Generator, Generator Section fGEK-1 4870) 9.7 Miscellaneous Documents I. SD-27, Main Generator and Exciter System

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MAIN GENERATOR VOLTAGE REGULATOR MANUAL OPT-40.2.11 AND AUTOMATIC OPERATIONAL CHECK Rev. 6 Page 13 of 13 ATTACHMENT 2 Page 1 of 1 Certification and Review Form Date Completed Reason For Test time Completed Routine Surveillance Unit %Pwr DWO# GMWE 0 Other (explain) General Comments and Recommendations: Initials Name (Print) Test procedure performed by: Exceptions to satisfactory performance: Corrective action required: Test Procedure has been completed SAT or UNSAT (circle as appropriate): Unit CRS/SRO Signature Date Test procedure has been reviewed by: Shift Manager Signature Date

DUKE ENERGY. BRUNSWICK TRAINING SECTION OPERATIONS TRAINING INITIAL LICENSED OPERATOR SIMULATOR EVALUATION GUIDE 2016 NRC SCENARIO 4 START CREV, NOO4A FAILURE, STATOR COOLING TRIP, RCIC STEAM LEAK, TCC FAILURE, LOOP, DG3 FAILURE, SRV TAILPIPE, ED REVISION 0 Developer: d oc Date: ZIfI2C/6 Technical Review: Vatc% Date: 9///2O6 Validators: Ved Date: O9/O//6 Facility Representative: ,-y Date: 24w

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 2 of 51 REVISION

SUMMARY

0 Scenario developed for 2016 NRC Exam.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 3of51 TABLE OF CONTENTS 1.0 SCENARIO OUTLINE 4 2.0 SCENARIO DESCRIPTION

SUMMARY

5 3.0 CREW CRITICAL TASKS 6 4.0 TERMINATION CRITERIA 6 5.0 IMPLEMENTING REFERENCES 7 6.0 SETUP INSTRUCTIONS 8 7.0 INTERVENTIONS 10 8.0 OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES 12 ATTACHMENT 1 - Scenario Quantitative Attribute Assessment 39 ATTACHMENT 2Shift Turnover 51

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 4of51 1.0 SCENARIO OUTLINE Event

         ]  Maif. No.
                         ] Type*                            Event Description 1                     N-BOP      Manual start of CREV in area high radiation mode.

C-ATC C32-LT-NOO4A Fails high 2 NBOO7F C-CRS (TS) EEO3OM- C-BOP MCC 2TD trip / Standby Stator Water Cooling Pump fails to 3 2TD C-CRS auto start C-ATC RCIC steam leak 4 ESO25F C-CRS (AOP)(TS) C-BOP TCC Pump Failure 5 K4516A C-CRS (AOP) 6 R-ATC Power Reduction M Loss of Off-Site Power / Scram 7 EEOO9F C DG3 Duff 0/C I DG4 failure of output breaker to close (RSP)(PCCP)(AOP) C SRV Failure / Tailpipe Break I DW Spray Logic Failure ESOO4F 8 M EDonPSP CAO2OF (AOP)(EDP)

                *( N )ormal,   (R)eactivity,  (C)omponent or Instrument,    ( M)ajor

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.Q] Page5of5l 2.0 SCENARIO DESCRIPTION

SUMMARY

Event Description 1 The BOP will start CREV in the area high radiation mode lAW OOP-37, Section 6.1.3. After CREV is started, C32-LT-NOO4A will fail high. The crew will reference Tech 2 Spec 3.3.2.2 and determine a 7 day LCO exists to place the failed channel in the tripped condition. The crew should select level B per OP-32. MCC 2TD will trip and the standby stator cooling water pump will fail to auto start. The standby stator cooling water pump can be manually started. The 2D air compressor will also be lost and OAOP-20.0 may be entered. Unit One may be contacted to place the 1 D Air Compressor in lead. A break in the RCIC steam line in the south RHR room will occur. The break can be 4 isolated by closing either the E51-F007 or the E51-F008. The crew will respond to the steam leak lAW AOP-05.0. TBCCW Pump 2B will trip and TBCCW low header pressure will alarm. The crew will 5 respond per OAOP-1 7.0. TBCCW pressure will recover and actions for partial loss of TBCCW will be performed. 6 A power reduction will be required lAW AOP-17.0. A Loss of Offsite Power will occur. The crew will respond per OAOP-36.1. DG3 will trip on Duff 0/C and DG4 output breaker will fail to close, can be closed manually. SRV F will fail open. AOP-30 will be entered. The SRV will not reset using the control switch. Pulling fuses lAW AOP-30 results in loss of indication but the SRV 8 remains open. SRV F tailpipe will rupture, pressurizing containment. The DW Spray logic (think switch) will fail causing an inability to spray the torus or drywell. Emergency Depressurization is required when PSP is violated.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page6of5l 3.0 CREW CRITICAL TASKS Critical Task #1 Close DG4 output breaker Critical Task #2 Emergency Depressurize when violating PSP 4.0 TERMINATION CRITERIA When all rods are inserted and level is being controlled above TAF the scenario may be terminated.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 7of51 5.0 IMPLEMENTING REFERENCES NOTE: Refer to the most current revision of each Implementing Reference. Number Title UA-14 (4-2) CE MACH ROOM VENT FAN TRIP A-07 (4-2) FW CTL SYS TROUBLE - OAO P-23 CONDENSATE/FEEDWATER SYSTEM FAILURE UA-06 (2-5) SUB 2F 480V FEEDER BKR TRIP UA-13 (6-6) RFP B CONTROL TROUBLE UA-02 (i- STAT COOLANT INLET FLOW-LOW UA-02 (1-9) LOSS OF STAT COOLANT TRIP CKT ENER UA-02 (2-8) STAT COOLANT PRESS-LOW UA-02 (6-9) EXCITER COOLANT FLOW-LOW UA-03 (2-4) TBCCW PUMP DISCH HEADER PRESS LOW OAOP-17.0 TURBINE BUILDING CLOSED COOLING WATER SYSTEM FAILURE A-03 (4-8) OPRM TRIP ENABLED OAOP-36. 1 LOSS OF ANY 4160V BUSES OR 480V E-BUSES

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page8of5l 6.0 SETUP INSTRUCTIONS

1. PERFORM TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 5, Checklist for Simulator Exam Security.

2. RESET the Simulator to IC-i i.

3. ENSURE the RWM is set up as required for the selected IC.

4. ENSURE appropriate keys have blanks in switches.

5. RESET alarms on SJAE, MSL, and RWM NUMACs.

6. ENSURE no rods are bypassed in the RWM.

7. PLACE all SPDS displays to the Critical Plant Variable display (#100).

8. ENSURE hard cards and flow charts are cleaned up

9. TAKE the SIMULATOR OUT OF FREEZE

10. LOAD Scenario File.

11. ALIGN the plant as follows:

Manipulation Ensure 2C TCC pump is in service on Unit One. Loaded in Scenario File Ensure 2B Stator Cooling Pump running and 2A in standby

12. IF desired, take a SNAPSHOT and save into an available IC for later use.

13. PLACE a clearance on the following equipment.

Component Position

14. INSTALL Protected Equipment signage and UPDATE RTGB placard as follows:

Protected Equipment

i. 2A and 2B NSW pumps

2. 2A FPC Pump/Hx, 2A RCC Pump, and 2C Demin Transfer Pump.

15. VERIFY OENP 24.5 Form 2 (Immediate Power Reduction Form) for IC-il is in place.

2016 NRC SCENARIO 4 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page9of5l

16. ENSURE each Implementing References listed in Section 7 is intact and free of marks.

17. ENSURE all materials in the table below are in place and marked-up to the step identified.

Required Materials

18. ADVANCE the recorders to prevent examinees from seeing relevant scenario details.

19. PROVIDE Shift Briefing sheet for the CRS.

20. VERIFY all actions contained in TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 4, Simulator Training Instructor Checklist, are complete.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page lOof 51 7.0 INTERVENTIONS TRIGGERS Trig Type ID 1 Malfunction NBOO7F [RX LVL TRANSMITTER C32-NOO4A FAILS] 2 Malfunction EEO3OM [INDIVIDUAL BUS FAILURE] 4 Malfunction ESO25F [RCIC STM BRK SOUTH RHR] 5 Annunciator ZUA324 [TBCCW PUMP DISCH HEADER PRESS LOW] 5 AO Override G4H11G14 [TBCCW DISCHARGE PRESS TOC-PI-556] 5 DI Override K4517A [TB CCW PMP B ON] 5 DI Override K4517A [TB CCW PMP B ON] 5 DO Override Q4517LG4 [TB CCW PMP B OFF G] 6 Malfunction EEOO9F [LOSS OF OFF-SITE POWER] 7 Remote Function SW_IAVSW193 [SW-V193 MAN ISOL NSW TO RBCCW] 7 Remote Function SWVHSW146L [CONV SW TO RBCCW HXS V146] 8 Remote Function RP_IAEPAMGA [RPS M-G SET A EPA BKRSJ 8 Remote Function RP_IARPSA [RESTART RPS MG SET A] 9 Remote Function RP_IAEPAMGB [RPS M-G SET B EPA BKRS] 9 Remote Function RP_IARPSB [RESTART RPS MG SET B] 10 Remote Function ED_ZIEDH11 [PNL 2AB-RX PWR (E7=NORM/E8=ALT)] 10 Remote Function ED_ZIEDHXO [PNL 32AB PWR fE7=NORM/E8=ALT)] 10 - Remote Function ED_ZIEDHO8 [PNL 2AB PWR (E7=NORM/E8zALT)J 11 Malfunction ESOO4F [ADS VALVE F FAILS OPEN] 12 DO Override Q15O8LGJ [SRV VLV B21-F013F GREEN] 12 DO Override Q15O8RRJ [SRV VLV B21-FO3F RED] 12 DO Override Q1520SA9 [AMBER LED +5V] 12 Malfunction CAO2OF [SRV F TAIL PIPE RUPTURE] 13 Remote Function MI_ZVACS918_1 [UNIT 1 CB MECHANICAL EQUIP ROOM VENT FANS CS] 14 Remote Function MI_IACBLRM1 [UNIT 1 CABLE SPREAD ROOM VENT FANS] Trig # rigger Text

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Pagellof5l MALFUNCTIONS MaIf Mult . Current Target Rmp Description Actime Dactime Trig ID ID Value Value time DG OUTPUT BREAKER FAIL TO AUTO DGO06F DG False True CLOSE DGO26F DG3 DIFFERENTIAL FAULT False True NBOO7F RX LVL TRANSMITTER C32-N004A FAILS 0.00 100.0 00:02:00 1 EEO3OM 2TD INDIVIDUAL BUS FAILURE False True 2 ESO25F RCIC STM BRK SOUTH RHR

                                      -                       0.00    5.0       00:10:00                         4 EEOO9F                  LOSS OF OFF-SITE POWER                 False   True                                       6 ES0O4F                  ADS VALVE F FAILS OPEN                 False   True                                       11 CAO2OF                  SRV F TAIL PIPE RUPTURE                False   True                  00:01:00             12 REMOTES Remf Id        Mult Id   Description                                 Current      Target Actime      Trig CC_IACW4S18                  2CTBCCW PUMP UNITALIGNMENT                 1            1 SW_VHSW146L                  CONVSWTORBCCWHXSV146                       SHUT         OPEN                        7 SW1AVSW93                    SW-V193 MAN ISOL NSW TO RBCCW              OPEN         CLOSE                        7 RPIARPSA                     RESTART RPS MG SET A                       NORMAL       RESET                        $

RPIAEPAMGA RPS M-G SET A EPA BKRS SET SET 00:00:05 8 RP_IARPSB RESTART RPS MG SET B NORMAL RESET 9 RP_IAEPAMGB RPS M-G SET B EPA BKRS SET SET 00:00:05 9 EDZIEDHO8 PNL 2AB PWR fE7=NORM/E8zALT) NORMAL ALT 00:00:30 10 ED_ZIEDH PNL 2AB-RX PWR (E7=NORM/E8=ALT) NORMAL ALT 00:02:30 10 ED_ZIEDHXO PNL 32AB PWR (E7=NORM/E8=ALT) NORMAL ALT 00:04:30 10 UNIT 1 CB MECHANICAL EQUIP ROOM VENT MI ZVACS918 1 NEUT STOP 13 FANSCS MI_IACBLRM1 UNIT 1 CABLE SPREAD ROOM VENT FANS AUTO OFF 14 EDIARKAX5 X-TIE BKR E7-E8 (AX5) RACK STATUS OUT IN 00:05:00 15 ED_IARKAIO X-TIE BKR E8-E7 (AlO) RACK STATUS OUT IN 00:02:30 15

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Pagel2of5l PANEL OVERRIDES Pos Won I Actual Override Tag ID Description Actime Dactime Trig K4517A TB CCW PMP B ON OFF OFF ON -________ _____ 5 K4517A TB CCW PMP B ON ON ON OFF 5 Q4517LG4 TB CCW PMP B OFF G ON/OFF OFF OFF 5 Q15O8LGJ SRV VCV B21-F013F GREEN ON/OFF ON OFF 12 Q15O8RRJ SRV VLV B21-FO3F RED ON/OFF OFF OFF 12 Q1520SA9 AMBER LED +5V ON/OFF OFF OFF 12 K5412A STAT COOLANT PMPA AUTO OFF OFF TBCCW DISCHARGE PRESS TOC-G4H11G14 39 80.2739 39 5 K1727A CONTSPRAYVLVCONTROL NORMAL ON OFF K1727A CONT SPRAY VLV CONTROL MANUAL OFF OFF K1727A CONT SPRAY VLV CONTROL RESET OFF OFF K1227A CONTSPRAYVLVCONTROL NORMAL ON OFF K1227A CONT SPRAY VLV CONTROL MANUAL OFF OFF K1227A CONTSPRAYVLVCONTROL RESET OFF OFF ANNUNCIATORS Override Window Description Tagname OVal AVaI Actime Dactime Trig TBCCW PUMP DISCH HEADER PRESS 2-4 ZUA324 ON ON OFF 5

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Pagel3of5l 8.0 OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES EVENT 1: Manual Start of CREV Simulator Operator Actions Ensure Monitored Parameters is open and Scenario Based Testing Variables are loaded. When contacted to secure the Ui CB Mech Equipment Room Vent Fans Initiate Trigger 13 When contacted to stop the Cable Spread Room 1 Vent Fans Initiate Trigger 14 Simulator Operator Role Play Evaluator Notes Plant Response: Objectives: SRO Directs BOP to manually start CREV BOP Manual Start of CREV RD Monitors the plant Success Path: CREV manually started Event Termination: When directed by the Lead Evaluator, go to Event 2.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 14of51 EVENT 1: Manual Start of CREV Time Pos EXPECTED Operator Response NOTES SRO Conduct shift turnover shift briefing. Direct CREV to be started in the area high radiation mode lAW OP-37. May conduct a brief (see Enclosure 1 on page 44forformat) RO Monitors the plant Manually starts CREV in the area high radiation BOP mode lAW OOP-37, Section 6.1.3.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page l5of 51 CONTROL BUILDING VENTILATION SYSTEM OQP-31 OPERATING PROCEDURE Rev. b2 Page 16of82 6.1.3 Manual Startup of the Control Building Emergency Recirculation System

1. Confirm the following initial cenditions are met:

• All applicable prerequisites as listed tn Section 5,0 are met. .....
• The Control Building Emergency Recirculation System has failed to start after an iniation signal or
• A manual start in accordance with UAOP-05.O, Radioactive Spills, High Radiation, and Airborne Activity, is required (8.11}

or

• Surveillance or inspection tests are required....

NOTE

• Indicabons for the Control Building Ventilation System are located on Panel XU-3 on both units D
• Controls for the Mechanical Equipment Room Ventilation Fans and the Control Building Wash Room Exhaust Fan are on XU-3 on Units 1 and 2 0
• Controls for the Cable Spread Room ventilation fans are on Panel XU-3 for the respective unit 0

2. Perform the following to place the Control Building Emergency Recirculation System in the area high radiation mode (includes Secondary Containment Isolation):

NOTE

• Placing one of the 2A(B)-ERF-CB (CD Emerg Recirc Fans) in ON will INOP the automatic start function of the non-operating fan El
• Controls for the Control Building Emergency Recirculation Fans are on Panel XU-3 on Unit 2 El

a. Place one of the 2A(B)-ERF-CB (CB Ernerg Recirc Fans) in ON

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 16 of 51 CONTROL BUILDING VENTILATION SYSTEM OOP-3/ OPERATING PROCEDURE Rev 62 Paqe 11ot82 61.3 Manual Startup of the Control Building Emergency Recirculation System (continued) CAUTION Detection of heat in the charcoal bed, detectors 2-FP-CB-4-20 and 2-F P-CB-4-2 I for A or detectors 24 P-CB-1-14 and 2-TP-CB-4-15 for B. wilt trtp the associated Emergency Recireulation t an ....................................................... 0

b. Confirm 2L-D-CB (Ctl RM Norm Mu Air Drnpr) closes

c. Confirm VA-2J-D-CB (GB Emorg Recirc Damper) opens.

d. Stop 2D-[T-CB (GB Washroom Exhaust Tan) and confirm associated damper closes NOTE The Conol Building Mechanical Equipment Room Vent Tans can be stopped only by simultaneously placing both Units control switches in OTT 0

e. Simultaneously place both Units control switches in OTT. for 2T-ST-CB and 2L-CF-CB (GB Mechanical Equip Room Vent Fans) to stop the fans and confirm associated supply and exhaust dampers close

f. Stop 2A-SF-CB and 2A-EF-CB (Cable Spread Room 2 Vent Tans) and confirm associated supply and exhaust dampers close

g. Stop lA-SF-GB and 1A-EF-CB (Cable Spread Room 1 Vent Tans) and confirm associated supply and exhaust dampers close

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page l7of 51 CONTROL BUlL DING VENTILATION SYSTEM OOP -31 OPERATING PROCEDURE Rev. 62 Page 18of82 6.13 Manual Startup of the Control Building Emergency Recirculation System (continued) NOTE The Control Building Emergency Recirculation System is now in operation for high radiation conditions U

3. Perform the following to place the Control Building Emergency Rocirculation System in the fire modo NOTE Placing one of the 2A(B)-[RF-CB (CU Emerg Recirc Fans) in ON will INOP the automatic start function of the non-operating Ian U a Place one of the 2NB)-ERF-CB (CU Lmcrg Rccirc Fans) in ON CAUTION Detection of heat in the charcoal bed, detectors 2-FP-CB-1-2O and 2-FP-CB-4-21 for A or detectors 2-FP-CB-4-14 and 2-FP-CB-1-15 for B, will trip the associated Emergency Recirculation Fan U

b. Confirm 2L-D-CB (Ctl RM Norm Mu Air Dmpr) closes

c. Confirm VA-2J-D-CB (CU [merg Recirc Damper) opens -

U Stop 2D-L F-CD (CU Washroom Exhaust Fan) and confirm associated damper closes NOTE The Control Building Emergency Recirculation System is now in operation for fire conditions U

4. WHEN the initiating oonditions have cleared, THEN place Control Building Ventilation System in operation in accordance with Secon 6.1.4

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 18 of 51 EVENT 2: C32-LT-NOO4A FAILS HIGH Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger I to fail C32-LT-NOO4A upscale. Simulator Operator Role Play If contacted as TB AO to check UPS Panel yb, Ckt #3, acknowledge request, then report no tripped breakers on UPS Panel yb. If asked, the inverters on the trip cabinets are energized. If contacted as maintenance or I&C to investigate trip, acknowledge request Evaluator Notes Plant Response: C32-LT-NOO4A will fail high. The crew will reference Tech Spec 3.3.2.2 and determine a 7 day LCO exists to place the failed channel in the tripped condition. The crew should select level B per OP-32. Objectives: SRO Determine TS LCO for C32-LT-NOO4A failing high RO Transfer DFCS to control to B Success Path: TS LCO 3.3.2.2, Condition A One feedwater and main turbine high water level trip channel inoperable. Required Action A.1 Place channel in trip within 7 days. DFCS Feedwater Level Select transferred to B Event Termination: Go to Event 3 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Pagel9of5l EVENT 2: C32-LT-NOO4A FAILS HIGH Time Pos EXPECTED Operator Response Comments Acknowledges annunciator report SRO A-07 4-2 FWCTL SYS TROUBLE Contacts I&C to investigate. May direct FWCS Level control to be selected to Level B. Determines TS 3.3.2.2 Condition A A.1 Place channel in trip in 7 days. May conduct a brief (see Enclosure 1 on page 44 for format) Acknowledges and reports annunciator report of RO A-07 4-2 FW CTL SYS TROUBLE Diagnose failure of the C32-NOO4A If directed by the CRS, shifts LEVEL NB select switch to Position B. BOP Monitors the plant

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page200f5l EVENT 3: MCC 2TD LOSS I STATOR COOLING STANDBY PUMP FAILURE Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 2 to trip the feeder breaker to MCC 2TD. If requested to place the 1 D Air Compressor in lead Activate Remote Al_2DLEAD, DELTA SA-CS-7892 (LEAD/LAG SWITCH) to 1 D LEAD, 2D LAG Simulator Operator Role Play When asked as the TB AD to investigate the 2F feeder breaker trip, report a trip of the feeder breaker to MCC 2TD, (ATO) on 480V Substation 2F is tripped with the white overcurrent indicating flag protruding from the breaker. If asked as I&C to investigate, acknowledge any requests for MCC trip / Auto start failure. If asked do not recommend re-energizing 2TD until an investigation can be completed. If asked to investigate/acknowledge the 2B RFP alarm, acknowledge the local panel alarm and report that the alarm on the local panel is HPU Pump 2 Running in Stby. If asked the standby pump is operating with no problems noted. If dispatched to verify proper operation of the standby Stator Water Cooling Water Pump or the 2B air compressor, report no problems with the operation of the pump/compressor are noted. If contacted as Ui, report that the 1 D air compressor is running as lag compressor. If asked to place the 1D Air Compressor in lead, after SIM OP activates remote, report 1D air compressor has been placed in lead. Evaluator Notes Plant Response: The crew will respond to a trip of MCC 2TD with the standby stator cooling water pump failure to auto start. The standby stator cooling water pump can be manually started. The 2D air compressor will also be lost (loss of controls) and OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures, may be entered. Objectives: SRO - Direct the standby Stator Cooling Water pump to be started. RD Start the standby Stator Water Cooling pump identify 2D air compressor failure. Success Path: Standby Stator Cooling Water Pump started and actions of OAOP-20.0 Pneumatic (Air/Nitrogen) System Failures, addressed. Event Termination: Go to Event 4 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 21 of5l EVENT 3: MCC 2TD LOSS I STATOR COOLING STANDBY PUMP FAILURE Time Pos EXPECTED Operator Response Comments Acknowledges report of alarms received/cleared SRO for the BOP/RO. Directs BOP operator to start the standby stator water cooling pump. May ask for I&C to investigate

1) The trip of the feeder breaker to 2TD

2) The failure of the standby Stator Water Cooling pump to auto-start.

May direct entry into OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures,. May review the load list for MCC 2TD (001-50.11). May conduct a brief (see Enclosure 1 on page 44 for format) AIC Monitors the plant.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page22of5l EVENT 3: MCC 2TD LOSS I STATOR COOLING STANDBY PUMP FAILURE Time Pos EXPECTED Operator Response Comments Report alarms to the CRS. UA-6, 2-5 Sub 2F 480V Feeder Bkr Trip UA-13, 6-6 RFP B Control Trouble BOP UA-2, 1-8 Stat coolant Inlet Flow-Low UA-2, 1-9 Loss of Stat Coolant Trip Ckt Ener UA-2, 2-8 Stat Coolant Press-Low UA-2, 6-9 Exciter Coolant Flow-Low Start the standby Stator Water Cooling Pump. UA-2, 4-9 Stator Cool Reserve Pump Running will annunciate on starting of the standby pump and then will clear when the 2B pump is placed in off. Dispatch an AO to investigate the Sub 2F Feeder Breaker Trip. May dispatch an AO to verify proper operation of the Stator Water Cooling pump that was started. May Dispatch an AO to investigate the alarm on the 2B RFP. May enter and announce OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures, for the trip of 2D Air Compressor. May ask Unit One to place the 1 D Air Compressor in the lead position. May place the 2D NC in Stop.

2016 NRC SCENARIO 4 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 23 of 51 EVENT 4: RCIC STEAM LEAK Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 4 to initiate a RCIC steam leak. (Increase as necessary to have room temperatures slowly rising until system is isolated) Simulator Operator Role Play After the initial alarms for the steam leak are received, report as RB AC steam is blowing out of the -17 foot in South RHR room and you are leaving the building. If contacted as I&C to investigate, acknowledge the request. If contacted to close 2-FP-P1V33, Unit 2 Reactor Building Sprinkler Shutoff Valve, wait three minutes and report that the valve is closed. Evaluator Notes Plant Response: A break in the RCIC steam line in the south RHR room will occur. The break can be isolated. If the system is delayed from being isolated, observe temperatures in the Reactor Building (specifically South RHR Room temperature), before any area exceeds MSOTL, a Reactor Manual Scram should be inserted. The RCIC system should be declared inoperable and Tech Specs entered. Objectives: SRO - Determine RCIC should be isolated and actions required for LCO per Technical Specifications RO Respond to an isolable RCIC steam line break. Success Path: Evaluate Tech Specs to determine required actions as outlined in SRO actions below. Event Termination: Go to Event 5 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 4 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page 24 of 51 EVENT 4: RCIC STEAM LEAK Time Pos EXPECTED Operator Response Comments SRO Diagnose RCIC leak and direct RCIC isolation May direct entry into OAOP-05.0 May direct reactor building evacuated Contact maintenance about the RCIC steam leak Refers to Tech Spec 3.5.3 RCIC System and determines: CONDITION A. RCIC System inoperable. REQUIRED ACTION: A.1 Verify by administrative means HPCI System is OPERABLE. Immediately AND A.2. Restore RCIC System to OPERABLE status. 14 days May conduct a brief (see Enclosure 1 on page 44 for format)

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page25of5l EVENT 4: RCIC STEAM LEAK Time Pos EXPECTED Operator Response Comments Respond to alarms: UA-03 3-5, PROCESS RX BLDG VENT RAD RD HI-HI UA-03 2-7, AREA PAD RX BLDG HIGH UA-05 6-10, RX BLDG ISOLATED Diagnose RCIC steam line leak Isolate RCIC by closing either isolation valve: E51-F007 (Steam Supply Inboard Isol Vlv) and/or E51-F008 (Steam Supply Outboard Isol Vlv) May reference procedure 20P-16, Section 6.3.4. (see page 25) BOP Monitors the plant. May announce and enter AOP-05 May direct AC to close 2-FP-PIV33

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 26 of 51 REACTOR CORE ISOLA11ON COOLING SYSTEM 2OP-16 OPERATING PROCEDURE Rev. 120 Page 3?of 99 6.3.4 Isolating the RCIC System Steam Supply

1. Confirm all applicable prerequisites listed in Section 5.0 are meL.......,

2. IF rapid isolation of RCIC steam line is desired, THEN perform the following;

a. Close E5l-100f (Steam Supply Inboard lsol Vlv).....,

b. Close L51-F008 (Steam Supply Outboard Isol Vlv)............

CAUTION Opening the L51-t045 (Turbine Steam Suppty Vlv) to dc-pressurize the RCIC steam line will roll the RCIC turbine D

3. iF. rapid isolation is N.Qi desired.

THEN perform the following to isolate and dc-pressurize the RCIC steam supply line

a. Close [51-FOOl (Steam Supply Inboard Isol Vlv)

b. Open MVD-V5002 (HPCI/RCIC Cond Dm Line Back Press Onfice Bypass Valve)

c. Open L51-F045(furbine Steam Supply Vlv) and monitor turbine response U. Close E51-t025 (Supply Drain Pot lnbd Drain Vlv)

e. Close E51-F026 (Supply Drain Pot Othd Drain Vlv) f WHEN RCIC steam line has boon dc-pressurized for approximately 2 minutes, THEN close E51-FO0$ (Steam Supply Outboard Isol Vlv)

g. Close [51-F045 (Turbine Steam Supply Vlv)

h. Close MVD-V5002 (HPCI/RCIC Cond Drn Line Back Press Orifice Bypass Valve)

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 27 of 51 REACTOR CORE ISOLATION COOLING SYSTEM 20-16 OPERATING PROCEDURE Rev. 120 Page 38 of 99 6.3.4 Isolating the RCIC System Steam Supply (continued) NOTE

• Technical Specification 3.6.1 .6 1 (MODES 1, 2, or 3) requires completion of OPT-02.31B. Suppression Pool to DrelI Vacuum Breaker Position Check, within 6 hours after any discharge of steam to the suppression chamber from any source and within 6 hours following an operation that causes any of the vacuum breakers to open D
• Section 6.3.4 Step ii ensures compliance with Technical Specifications and may be completed as required during the performance of the procedure 0 IF inMODLS 1, 2, or3, THEN ensure OPT-023,1 B, Suppression Pool to Drywell Vacuum Breaker Position Chock, is completed within 6 hours after any discharge of steam to the suppression chamber from any source. {8.1.1}

Dateifime Completed Performed By (Print) Initials Reviewed By Unit CRS/SRO

2016 NRC SCENARIO 4 LOI SIMULATOR EVALUATION GUIDE Rev. 0 Page28of5l EVENT 5/6: TCC PUMP B TRIP I POWER REDUCTION Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 5 to trip the 2B TCC Pump. When power is reduced change TCC pressure override to current value and activate over one minute then delete annunciator override, when pressure is at current value delete override. Simulator Operator Role Play If contacted as the TB AC, wait one minute and report that 2B TCC pump is hot to the touch and the breaker is tripped (magnetic). If contacted as Unit One CRS, report Unit One is using the 2C TCC Pump and cannot be released to Unit Two operation If contacted as I&C to investigate 2B TCC Pump, acknowledge request. If contacted as Unit One to start the 1 D air compressor, report 1 D Air Compressor is running. If contacted as RE for power reduction or Reactivity Plan, ask the CRS what their recommendation is, then concur with that recommendation. If contacted as chemistry acknowledge request for sample due to a 15% power change. Evaluator Notes Plant Response: TBCCW Pump 2B will trip and TBCCW low header pressure will alarm. The crew will respond per OAOP-1 7.0. With 2C TBCCW Pump supplying Unit 1, a power reduction will be required. TBCCW pressure will recover and actions for partial loss of TBCCW will be performed. Objectives: SRO Direct entry into OAOP-17.0 RO Power reduction with Recirc flow and restoration of TCC pressure Perform actions for a partial loss of TCC. Success Path: TCC pressures restored to normal with reactor power reduced to the tecirc flow limit. Event Termination: Go to Event 7 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 [ Page29of5l EVENT 516: TCC PUMP B TRIP I POWER REDUCTION Time Pos EXPECTED Operator Response Comments Acknowledge report of annunciator UA-03 2-4 TBCCW PUMP DISCH HEADER SRO PRESS LOW Direct entry into OAOP-17.0, Turbine Building Closed Cooling Water System Failure. Direct power reduction lAW OENP-24.5 to the Recirc flow limit. Directs I&C to investigate loss of 2B TOO pump. Acknowledge report of annunciator A-03 4-8 OPRM TRIP ENABLED Contact chemistry to sample coolant because of the power reduction (>15%) Briefs crew on reactor scram if TCC pressure is not restored above 42 psig within 4 minutes of reaching 47Mlbm/hr. May conduct a brief (see Enclosure 1 on page 44 for format)

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page300f5l EVENT 5/6: TCC PUMP B TRIP I POWER REDUCTION Time Pos EXPECTED Operator Response Comments RD Plant Monitoring. Reduces reactor power with Recirc Flow to ENP-24.5 flow limit of 47 Mlbm/hr. May use the Manual Runback for flow (see page 34) Acknowledge and report annunciator A-03 4-8 OPRM TRIP ENABLED Acknowledge and Report annunciator BOP UA-03 2-4 TBCC W PUMP DISCH HEADER PRESS LOW Diagnose loss of 2B TCC Pump. Announce and enter OAOP-1 7.0, Turbine Building Closed Cooling Water System Failure. (see page 30) Performs step 4.2.3 (page 30) Performs Step 4.2.6 (page 31) Report annunciator UA-03 2-4 TBCC W PUMP DISCH HEADER PRESS LOW clear.

2016 NRC SCENARIO 4 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 31 of5l TURBINE BUILDING CLOSED COOLING WATER OAOP-l 7.0 SYSTEM FAILURE Rev. 033 Page 60116 4.2 Supplementary Actions (continued) NOTE TBCCW pump power supplies are as follows: 0 a TBCCW Pump 1A, MCC 1 TJ

• TBCCW Pump 1B, MCC 1TM
• TBCOW Pump 2A, MOO 2TJ
• TBCOW Pump 2B, MOO 2TM
• TBCOW Pump 20, MCC 2TH, with an automatic transfer switch to select MCC 1 TH as the power supply on loss of power to 2TH NOTE In accordance with OAP-013, Plant Equipment Control, tripped breakers (thermally or magnetically) should ii4QI be reset except in an emergency situation until an evaluation of the circuit condition has been performed. Breakers that have tripped thermally may be reset as deemed necessary by the Unit ORS for continued reliable operation of the plant 0 b IF TBCCW pump breakers local thermal or magnetic trips have activated, THEN perform the following.

(1) Initiate a WO for evaluation of the affected circuit 0 (2) WHEN directed by the Unit CR5, THEN reset tripped breakers 0

3. IF only one TBCCW pump is in service AND TBCCW pressure is less than 42 psig, THEN perform the following:

a. Reduce reactor power with recirc flow in accordance with OENP-24.5, Form 2, Immediate Reactor Power Reduction Instructions 0

b. IF TBCCW pressure is greater than 42 psig within 4 minutes, THEN perform Section 4.2 Step 6, on page 7 0 c IF TBCCW pressure is NOT greater than 42 psig within 4 minutes, THEN perform Section 4 2 Step 7, on page 9 0

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 32 of 51 TURBINE BUILDING CLOSED COOLING WATER OAOP-i 7.0 SYSTEM FAILURE Rev 033 Page 7o1 16 4.2 Supplementary Actions (continued)

4. IF a IBCCW system leak is suspected, THEN:

a. Monitor TBCCW Head Tank level 0 b Maintain TBCCW Head Tank level in accordance with 2OP-44, Turbine Building Closed Cooling Water System Operating Procedure 0

c. Check system piping to locate leakage 0 d Isolate any leakage found 0

e. Monitor temperatures on equipment cooled by TBCCW 0

5. IE TBCCW heat exchanger outlet temperature is greater than 1 i0F OR component temperatures are nsing, THEN reduce reactor power as necessary to reduce TBCCW temperature 0

[ NOTE A partial loss of TBCCW or service water is defined as reduced cooling available with the expectation that nom,aI cooling can be quickly re-established 0

6. IF there is a partial loss of TBCCW or service water, THEN perform the following a Ensure all available TBCCW pumps are operating 0 NOTE High temperature indications on equipment cooled by TBCCW in conjunction with CSW header pressure approaching 90 psig are indications of a Conventional Service Water System failure. (7.1 1} 0 b IF a failure of CSW is indicated.

THEN enter OAOP-19.0. Conventional Service Water System Failure, perform concurrently with this procedure 0

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 33 of 51 TURBINE BUILDING CLC)SED COOLING WATER OAOP- 17.0 SYSTEM FAILURE Rev. 033 Page 80116 4.2 Supplementary Actions (continued)

c. Reduce system heat load by removing the following loads as plant conditions permit.

(1) Out-of-service equipment 0 (2) Sample coolers 0 (3) Bus-duct cooling 0 NOTE

• If only the main or standby compressor is operating on the unaffected unit and the idle compressor is available, there should be sufficient compressed air capacity to support air demand C
• Service Air Compressors 1 B and 2B are fQI designed to individually carry the full system demand of both units when the cross tie valves ate open C d IF air pressure can NOT be maintained.

THEN enter OAOP-20.0. Pneumatic (Air/Nitrogen) System Failures AND perform concurrently with this procedure e IF Unit 1 and Unit 2 Service Air Systems are cross-tied, THEN: (1) Ensure the unaffected units air compressors have sufficient capacit to support air demand 0 (2) Ensure the unaffected units Service Ar Compressor D is operating 0 IF Unit 1 and Unit 2 Service Air Systems are JQI cross-tied AND it is possible to cross-tie, THEN: NOTE

• If only the main or standby compressor is operating on the unaffected unit and the idle compressor is available, there should be sufficient compressed air capacity to support air demand 0
• Service Air Compressors I B and 2B are 4QI designed to individually carry the full system demand of both units when the cross tie valves are open C (1) Obtain permission from the unaffected units CRS to cross-tie the Service Air Systems 0

L 2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 34 of 51 TURBINE BUILDING CLOSED COOLING WATER I OAOP-17.0 I SYSTEM FAILURE I Rev. 033 I Page 9 of 16 4.2 Supplementary Actions (continued) (2) Ensure 2-SA-PV-5071 (Cross-Tie Valve) on Unit 2 Panel XU-2, is OPEN 0 (3) Ensure l-SA-PV-5071 (Cross-Tie Valve) on Unit 1 Panel XU-2. is OPEN 0 (4) if the uninvolved units air systems are adversely affected, THEN perform the following at the direction of the Unit CR5 or Reactor Operator;

• if Service Air Dryer 16 is in standby, ORin service on Unit 1, THEN close 2-SA-PV-507 1 (Cross-Tie Valve),

using the control switch located on Unit 2 PanelXU-2 0 OR

• IF Service Air Dryer 16 is in service on Unit 2, THEN close 1-SA-PV-507l (Cross Tie Valve).

using the control switch located on Unit 1 Panel XU-2 0 g Trip the affected uniVs air compressors 0

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 35 of 51 REACT)R RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev. 168 Page 94 of 250 6.3.16 Initiation Of A Manual Runback NOTE The Man Runback feature is enabled only wflen both Recirc Pumps are operating at greaterthan 54.8% speed 0 BEGIN R.M. LEVEL R21R3 REACTIVITY EVOLUTION 1 Confirm the following Initial Conditions are met.

• Manual Runback Enabled wNte light on Panel P603 is ON
• lmrnedate power reduction is requwed, wtiich reduces total core flow to 47 mlbmr, or
• The Unit CRS directs inEtiaton of a Manual Runback NOTE A Manual Runback lowers both Recirc Pump speeds at 100 rpm/second to 53.6%

speed, which results in approximately 47mlbfhr core flow. The Manual Runback can be reset by depressing the Man Runback pushbutton a second time 0

2. Depress the Man Runback pushbutton 3 Confirm the following:

a Both Recirc Pump speeds are lowenng b The Manual Runback Enabled light is flashing C 2-A-06, 3-2 (2-A-07, 2-4), Recirc Flow A(B) Lirnt, annunciator isON d Resultant Core Flow is approximately 47 mlb/hr, unless manually RESET END R.M. LEVEL R21R3 REACTIVITY EVOLUTION

4. Go to 2AOP-04.0, Low Core Flow

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 36 of 51 EVENT 7: LOOP I SCRAM I DG FAILURES Simulator Operator Actions At the discretion of the lead evaluator, Initiate Trigger 6 to active the LOOP. Acknowledge and silence Fireworks alarms Acknowledge Unit 1 alarms as needed. If directed to align RBCCW to CSW cooling, wait 4 minutes and Initiate Trigger 7. If directed to restart RPS MG sets, wait 3 minutes and insert the following as requested: For RPS A Initiate Trigger 8 and/or for RPS B Initiate Trigger 9. If directed to swap AB panels Initiate Trigger 10 and inform Sim Role Player when timed out. If directed to tack in 480V cross-tie breakers Initiate Trigger 15 Simulator Operator Role Play If requested to monitor DGs, acknowledge alarms on DG local Alarm Panel (Instructor Aids/Panels) and report alarms if requested If directed to align RBCCW to CSW cooling, wait 4 minutes and inform Sim Operator to align RBCCW to CSW cooling then report valve open. If directed to restart RPS MG sets, wait 3 minutes and inform Sim Operator to restart RPS then report actions complete. If directed as RBAO to ensure BFIV latching mechanisms are disengaged, wait two minutes, then report latches are disengaged. If requested to transfer 2AB, 32AB, 2AB-RX, acknowledge request, inform Sim operator and when the remotes are timed out inform the control room the action is complete. If directed to cross-tie 480V after remote timers time out report breakers racked in. Evaluator Notes Plant Response: The crew will respond to a Loss of Offsite Power. The reactor will scram on MSIV closure on the LOOP. All Diesel Generators will start on the LOOP signal. DG3 will trip on Duff 0/C. DG 4 output breaker will fail to auto close. The BOP operator will close DG 4 output breaker to energize Bus E4. Objectives: SRO Direct actions of AOP-36.1 RD Close DG4 output breaker. Perform scram immediate operator actions. Success Path: Scram immediate operator actions are complete and DG4 output breaker is closed.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page37of5l EVENT 7: LOOP I SCRAM I DG FAILURES Time Pos EXPECTED Operator Response Comments SRO Direct AOP-36.J entry. Direct DG4 output breaker closed. CRITICAL TASK #1 Contacts Maintenance for failure of DG3 and DG4 output breaker. Enters and directs actions of RVCP: D Direct control of reactor pressure using SRVs (establishes pressure band 800 1000 psig) D Direct water level band of 170 200 inches Enters and directs actions of POOP: D Monitor and control Suppression Pool temperature below 95 deg F. D Direct starting available RHR Loops in Suppression pool Cooling as necessary to maintain temp below 95 F. D Monitor HOTL D Direct operation of available drywell coolers D Verify RCC operation and alignment to the d rywel I

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page38of5l EVENT 7: LOOP I SCRAM I DO FAILURES Time Pos EXPECTED Operator Response Comments Unit 2 SCRAM Immediate Actions

1. Ensure SCRAM valves OPEN by manual SCRAM or ARI initiation.

2. WHEN steam flow less than 3.0 Mlb/hr, THEN place reactor mode switch in SHUTDOWN.

3. IF reactor power below 2% (APRM downscale trip), THEN trip main turbine.

ATC

4. Ensure master RPV level controller setpoint at +170 inches.

5. IF Two reactor feed pumps running AND

• RPV level above +160 inches AND
• RPV level rising, THEN_trip_one.

Communicate scram report to CRS Place SULCV in service (See Enclosure 4 page 48) Insert Nuclear Instrumentation Ensure Turbine Oil System Operating Ensure Reactor Recirculation Pumps at 34% Ensure Heater Drain Pumps tripped Maintain reactor water level between 170 200 inches Place RHR Loops in Suppression pool Cooling as necessary (see Enclosure 3 page 46) Control reactor pressure 800 1000 psig

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page39of5l EVENT 7: LOOP I SCRAM I DG FAILURES Time Pos EXPECTED Operator Response Comments Diagnose failure of DG4 output breaker BOP Manually close DG4 Output Breaker CRITICAL TASK #1 Diagnose and report to the SRO DG3 tripped and Locked out. Perform the following OAOP-36.1 actions: Dispatch AC to monitor DGs Momentarily place DIV I NON-INTRPT RNA, SV-5262 control switch to OVERRIDE/RESET, then to OPEN, and ensure DIV I NON-INTRPT RNA, SV-5262 opens. May start the CRD system in accordance with OP-08, Section 8.17, or it may be started lAW SEP-09. Ensure the associated NSW and CSW pumps are operating. Direct an AC to swap the AB panels to their alternate source. Ensure 125V and 24V DC battery chargers return to service for each energized 480V E Bus. Perform the following to transfer RBCCW HXs from the NSW header to the CSW header: D Confirm CSW system available. Ensure at least one of the following is closed: D RBCCW HX SERVICE WATER INLET VALVE,SW-V1 03 D RBCCW HX SERVICE WATER INLET VALVE, SW-VJ 06 D Direct an AC to open CONVENTIONAL HEADER TO RBCCW HEAT EXCHANGERS SUPPLY VALVE,SW Vi 46.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 40 of 51 EVENT 7: LOOP I SCRAM I DG FAILURES Time Pos EXPECTED Operator Response Comments p Continue OAOP-36.1 actions: Ensure Control Building Ventilation started on the affected unit: Perform the following to restore drywell cooling: If three RBCCW pumps are running, then STOP one RBCCW pump, and place its control switch in AUTO. D If only one RBCCW pump is running, then START a second pump, if available. D If no RBCCW pump is running, then place all RBCCW pump control switches in OFF, and perform one of the following: IF any local drywell temperature is currently greater than the starting temperature limit OR has exceed the starting temperature limit since the initiation of the event, then perform 20P-21, Section 8.6. IF all local drywell temperatures have remained less than the starting temperature limit since the initiation of the event, then perform 20P-21, Section 5.2. ENSURE all available drywell coolers on the affected unit are operating. IF HPCI is running with suction from the CST AND CST level indication is NOT available in the Control Room or Radwaste, then monitor CST level locally and report level every hour. Start RPS MG Sets A(B) in accordance with OP 03, Section 5.2 May direct for 480V busses to be cross-tied.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 41 o151 EVENT 7: LOOP I SCRAM I DG FAILURES Time Pos EXPECTED Operator Response Comments BOP Continue OAOP-36.1 actions: Perform the following to start the Reactor Building HVAC: D If PROCESS OG VENT PIPE RAD HI-HI (UA-03, 5-4) is in alarm, and is NOT the result of a valid high radiation signal, then place CAC PURGE VENT ISOL OVRD, CAC-CS-5519, in OVERRIDE Reset the following Reactor Building Ventilation Radiation Monitors on Panel H12-P606: D PROCESS REACTOR BLDG VENTILATION RADIATION MONITOR A, D12-RM-K609A D PROCESS REACTOR BLDG VENTILATION RADIATION MONITOR B, D12-RM-K609B. Depress the following Isolation Reset Groups push buttons: D ISOLATION RESET GROUPS 1, 2, 3, 6, 8, A71-S32 D ISOLATION RESET GROUPS 1, 2, 3, 6, 8, A71-S33. D Ensure Instrument Air header pressure is greater than 95 psig. D Ensure BFIV latching mechanisms are disengaged. (Local). D Open RB VENT INBD ISOL VALVES, A BFI V-RB and C-BFI V-RB. D Open RB VENT OTBD ISOL VALVES, B BFI V-RB and D-BFI V-RB. Start three sets of Reactor Building Ventilation Fans in accordance with OP-37.1, Section 8.8 to maintain Reactor Building static pressure negative.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 42 of 51 EVENT 8: SRV FAILURE /TAILPIPE BREAK I ED PSP I TERMINATION Simulator Operator Actions Initiate Trigger 11, to fail the SRV. When contacted to pull SRV F fuses Initiate Trigger 12. (this also fails the downcomer) When directed by the Lead Evaluator, place the simulator in FREEZE DO NOT RESET THE SIMULATOR PRIOR TO RECEIPT OF CONCURRENCE TO DO SO FROM THE LEAD EXAMINER Simulator Operator Role Play If contacted to pull fuses for SRV F lAW AOP-30.O, wait 2 minutes have SIM OP Initiate Trigger 12 and report that the Fuses for SRV F have been pulled. Evaluator Notes Plant Response: An SRV will fail open and then the tailpipe will break causing a violation of PSP requiring the plant to be emergency depressurized. Objectives: SRO - Directs actions for Emergency Depressurization. RO - Perform Emergency depressurization. Success Path: Emergency depressurization performed. Scenario Termination: When emergency depressurization has been performed and RPV pressure is less than 100 p51g. the scenario may be terminated. Remind students not to erase any charts and not to discuss the scenario until told to do so by the evaluator/instructor.

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 43 of 51 EVENT 8: SRV FAILURE /TAILPIPE BREAK I ED - PSP I TERMINATION Time Pos EXPECTED Operator Response Comments SRO Directs announcement of OAOP-30.0. Direct level maintained 170 200 inches. Directs Emergency Depressurization when CRITICAL TASK #2 PSP is violated. (see Enclosure 2 on page 45) ATC Maintains level as directed by the CRS. Maintains reactor pressure as determined by the CRS. Informs CRS of SRV F failure to close. Announces and enters OAOP-30.0. Attempts to cycle control switch for stuck open SRV. Directs WCCSRO to pull fuses for SRV F Performs Emergency Depressurization when CRITICAL TASK #2 directed by the CRS. Continues OAOP-36.1 actions. BOP (see actions in event 7)

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 44 of 51 ENCLOSURE 1 Page 1 of I CONDUCT OF OPERA1IONS AD-OP-ALL-i 000 Rev. 6 Page 90 of 90 ATTACHMENT 8 Page 1 of 1

                         <<  Crow Briof Template>>

0 Announce Crew Brief Begin Brief 0 All crew members adcnowledge announcement (A Required) 0 Update the crew as needed 0 Describe what happened and major actions taken 0 Procedures in-progress 0 Notifications: Rec3p 0 Maintenance 0 Engineering 0 Others (Dispatctler, Station Management, etc.) 0 Future Direction arrd pnorities 0 Discuss any contingency plans fA Required) 0 Solicit questionsconcerns from each crew member OROs input 0 CRS 0 STA 0 Are there any alarms unexpected for the plant conditions? 0 What is the status of Critical Parameters? (A Required) EAL 0 Provide EAL and potential escalation criteria 0 Restore normal alarm announcement? (Yes/No) FInl:h Brief 0 Announce End of Brief

2016 NRC SCENARIO 4 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 45 of 51 ENCLOSURE 2 Page 1 of 1

                  <<Pressure Suppression Pressure>>
            +2
            +1 0

1 L1 C

        -J  -2 w

w -3 Cl)

            -4 UNSAFE 0   -s F-
            -6
            -7
            -8 0          10         20         30    40 TORUS PRESSURE (PSIG)

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 46 of 51 ENCLOSURE 3 Page 1 of 2

          <<Emergency Suppression Pool Cooling Using Loop A (20P-f 7)>>

NOTE This attachment is not to be used for normal system operations C Start RHR SW A LOOP fCONV) Start RHR SW A LOOP (NUC) Open SW-Viol D Open SW-V 105 C Close SW-V 143 D Open SW-V 102 C Start CSW PUMPS AS NEEDED D Close SW-V143 C IF LOCA SIGNAL IS PRESENT, Start PUMPS ON NSW HDR AS NEEDED C THEN place RHR SW BOOSTER IF LOCA SIGNAL IS PRESENT, PUMPS A & C LOCA OVERRIDE THEN place RHR SW BOOSTER PUMPS SWITCH TO MANUAL OVERRIDE A & C LOCA OVERRIDE SWITCH TO MANUAL OVERRIDE C Start RHR SW PMP C Start RHR SW PMP C Adjust El l-PDV-F068A C Adjust El i-PDV-F068A C Establish CLG WTR TO \ITAL HDR C Establish CLG WTR TO VITAL HDR C Start ADDITIONAL RHR SW PUMP Start ADDITIONAL RHR SW PUMP and adjust FLOW AS NEEDED C and adjust FLOW AS NEEDED C Start RHR LOOP A IF LOCA SIGNAL IS PRESENT, THEN C Verify COOLING LOGIC IS MADE UP IF El i-F015A IS OPEN, THEN close Eli -FO17A C Start LOOP A RHR PMP C Open El i-F028A C Throttle El i-F02$A C Throttle El i-F048A C Start ADDITIONAL LOC)P A RHR PMP and adjust FLOW AS NEEDED C 2 2/1061 2 S/I 062

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 47 of 51 ENCLOSURE 3 Page 2 of 2

           <<Emergency Suppression Pool Cooling Using Loop B (20P-f 7)>>

NOTE This attachment is not to be used for normal system operations 0 Start RHR SW B LOOP (NUC) Start RHR SW B LOOP (CONy) Open SW-Vl05 O Open SW-Viol 0 Close SW-V 143 O Open SW-Vi02 0 Start 05W PUMPS AS NEEDED O Close SW-Vi13 0 LOCA SIGNAL IS PRESENT, Start PUMPS ON NSW HDR AS NEEDED 0 THEN place RHR SW BOOSTER IF LOCA SIGNAL IS PRESENT, PUMPS B & D LOCA OVERRIDE THEN place RHR SW BOOSTER PUMPS SWITCH TO MANUAL OVERRIDE O B & D LOCA OVERRIDE SWITCH TO MANUAL OVERRIDE 0 Start RHR SW PMP 0 Start RHR SW PMP 0 Adjust El 1-PDV-F068B 0 Adjust El l-PDV-F06$B 0 Establish CLG WTR TO VITAL HDR C Establish CLG WTR TO VITAL HDR 0 Start ADDITIONAL RHR SW PUMP Start ADDITIONAL RHR SW PUMP and adjust FLOW AS NEEDED O and adjust FLOW AS NEEDED C Start RHR LOOP B if LOCA SIGNAL IS PRESENT, THEN C Verify COOLING LOGIC IS MADE UP if E1l-FO15B IS OPEN, THEN close ElI-FO17B C Start LOOP B RHR PMP C Open El 1 -F028B C Throttle El l-F024B C Throttle El l-F048B C Start ADDITIONAL LOOP A RHR PMP and adjust FLOW AS NEEDED C 2 2/1063 2 S/1064

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 48 of 51 ENCLOSURE 4 Page 1 of 2 Feedwater Level Control Following a Reactor Scram (EOP) NOTE: This attachment is not to be used for routine system operation ENSURE the following:

• 1W- V6 AND 1W -V8 OR FW-V118 AND IW-V119 closed U
• IW-FV-1 ii closed U
• IW-V120 closed U
• 1W control MODL $LL[CI in 1 ELEM U
• SULCV in M (MANUAL) closed U
• B21-FO32AANDIORS2-FO32Bopen U

2. PLACE the MSTR RIPT SPRX LVL CTL in M (MANUAL), THEN U

• ADJUST to 18?

3. IF any RtP is running, THEN:

a. PLACE RIP A(S) Recirc Vlv, conol switch to open U

b. PLACE RIPT A(S) SP CTL in M (MANUAL) U

4. IF no RI P is running, THEN:

a. PLACE RIP A(S) RECIRC VLV. control switch to open .. U

b. ENSURE the following:

• FW-V3cV4) [RIP A(S) Disch Vlv open U
• RIPT A(S) $P CTL in M (MANUAL) at lower limit U
• RIPT A(S) Man/DICS control switch in MAN U
• Reactor water level is less than 4206 inches AND RFPT A&B HIGH LEVEL TRIP reset U

c. DEPRESS RIPI A(S) RESET U

2016 NRC SCENARIO 41 LOl SIMULATOR EVALUA11ON GUIDE I Rev.0I Page 49 of 511 ENCLOSURE 4 Page2of2 Feedwtar Level Control Following a Reader Scram (EOP)

d. ENSURE RrPT 048) LP MD HP STOP VLVS open 0 a ROLL RFPT 048) to 1000 m by depressing RI? NB)

START.... .. ........ - - - 0

1. RAISE lflT A(B) to at toast 2550 m using tho LOWCR/RMSC contrd switch - - ... 0

g. DEPRESS lPT NB) DFCS CTRL RESET - 0

5. ENSURE MANIDFCS contol switch in DFCS - .._.... - .........._. 0

6. RAISE RFPT NB) SP aL weed until dschage pmsswe is oater then or equ to 100 psig above rea2orpmssse .... ...... ..... 0

1. ADJUST SULCV to ostthhsh desired injection _. - a... 0

6. F desired, THEN PLACE SULCV in A (AUTO) .. .... ..... 0

9. F needed. THEN ThROTTLE FW-V120.. - ..... 0

10. F needed, THEN GO TO 2OP-32, Condensae And Feedwater System Oporaing Procedure, for level oontml ........ .. .... ..... 0 3 211204 9 S11205

2016 NRC SCENARIO 4 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 5Oof 51 ATTACHMENT I - Scenario Quantitative Attribute Assessment NUREG 1021 Category Scenano Content Rev. 2 Supp. I Req. Total Malfunctions 5-8 8 Malfunctions after EOP 1-2 2 Entry Abnormal Events 2-4 4 Major Transients 1-2 2 EOPs Used 1-2 2 EOP Contingency 0-2 1 Run Time 60-90 mm 90 Crew Critical Tasks 2-3 2 Tech Specs 2 2 Instrument I Component 2 OATC Failures before Major 2 BOP Instrument I Component 2 2 Failures after Major Normal Operations 1 1 Reactivity manipulation 1 1

LOl SIMULATOR EVALUATION GUIDE Page 51 of 511 ATTACHMENT 2 Shift Turnover Brunswick Unit 2 Plant Status Station Duty Workweek E. Neal B. Craig Manager: Manager: Mode: 1 Rx Power: 100% Mode: 1 Plant Risk: G teen Current EOOS Risk Assessment is: SFP Time to 80 days 49.7 hrs Days Online: 200 Deg F: Turnover: Feedwater Temperature Reduction will be implemented this weekend Protected 2A FPC Pump/Hx, 2A RCC Pump, and 2C Demin Transfer Pump for Equipment: Fuel Pool Decay Heat Removal and inventory makeup. 2NB NSW Pumps due to IA NSW pump maintenance 1A NSW Pump is under clearance for planned maintenance. 2C TCC Pump is in service on Unit One. Comments: The BOP is to start CREV in the area high radiation mode for inspection testing lAW OOP-37, Section 6.1.3. (The inspection is scheduled to take three hours)

CONTROL BUILDING VENTILATION SYSTEM OOP-37 OPERATING PROCEDURE Rev. 62 Page 16 of 82 6.1.3 Manual Startup of the Control Building Emergency Recirculation System

1. Confirm the following initial conditions are met:

               .       All applicable prerequisites as listed in Section 5.0 are met
               .       The Control Building Emergency Recirculation System has failed to start after an initiation signal or
                .      A manual start in accordance with OAOP-05.0, Radioactive Spills, High Radiation, and Airborne Activity, is required

{8.1.1} or

                .      Surveillance or inspection tests are required NOTE Indications for the Control Building Ventilation System are located on Panel XU-3 on both units Controls for the Mechanical Equipment Room Ventilation Fans and the Control Building Wash Room Exhaust Fan are on XU-3 on Units 1 and 2                             D Controls for the Cable Spread Room ventilation fans are on Panel XU-3 for the respective unit                                                                         D

2. Perform the following to place the Control Building Emergency Recirculation System in the area high radiation mode (includes Secondary Containment Isolation):

NOTE Placing one of the 2A(B)-ERF-CB (CB Emerg Recirc Fans) in ON will INOP the automatic start function of the non-operating fan D Controls for the Control Building Emergency Recirculation Fans are on Panel XU-3 on Unit 2

a. Place one of the 2A(B)-ERF-CB (CB Emerg Recirc Fans) in ON

CONTROL BUILDING VENTILATION SYSTEM OOP-37 OPERATING PROCEDURE Rev 62 Page 17of82 6.1.3 Manual Startup of the Control Building Emergency Recirculation System (continued) CAUTION Detection of heat in the charcoal bed, detectors 2-FP-CB-4-20 and 2-FP-CB-4-21 for A or detectors 2-FP-CB-4-1 4 and 2-FP-CB-4-1 5 for B, will trip the associated Emergency Recirculation Fan C

b. Confirm 2L-D-CB (Ctl RM Norm Mu Air Dmpr) closes

c. Confirm VA-2J-D-CB (CB Emerg Recirc Damper) opens U. Stop 2D-EF-CB (05 Washroom Exhaust Fan) and confirm associated damper closes NOTE The Control Building Mechanical Equipment Room Vent Fans can be stopped only by simultaneously placing both Units control switches in OFF C

e. Simultaneously place both Units control switches in OFF, for 2F-SF-CB and 2E-EF-CB (05 Mechanical Equip Room Vent Fans) to stop the fans and confirm associated supply and exhaust dampers close ___

1. Stop 2A-SF-CB and 2A-EF-CB (Cable Spread Room 2 Vent Fans) and confirm associated supply and exhaust dampers close

g. Stop lA-SF-CE and IA-EF-CB (Cable Spread Room 1 Vent Fans) and confirm associated supply and exhaust dampers close

CONTROL BUILDING VENTILATION SYSTEM OOP-37 OPERATING PROCEDURE Rev. 62 Page 18of82 61.3 Manual Startup of the Control Building Emergency Recirculation System (continued) NOTE The Control Building Emergency Recirculation System is now in operation for high radiation conditions D

3. Perform the following to place the Control Building Emergency Recirculation System in the fire mode:

NOTE Placing one of the 2A(B)-ERF-CB (CE Emerg Recirc Fans) in ON will INOP the automatic start function of the non-operating fan D

a. Place one of the 2A(B)-ERF-CB (CE Emerg Recirc Fans) in ON CAUTION Detection of heat in the charcoal bed, detectors 2-FP-CE-4-20 and 2-FP-CB-4-21 for A or detectors 2-FP-CB-4-1 4 and 2-FP-CB-4-1 5 for B, will trip the associated Emergency Recirculation Fan D

b. Confirm 2L-D-CB (CtI RM Norm Mu Air Dmpr) closes

c. Confirm VA-2J-D-CB (CE Emerg Recirc Damper) opens

d. Stop 2D-EF-CB (CB Washroom Exhaust Fan) and confirm associated damper closes NOTE The Control Building Emergency Recirculation System is now in operation for fire conditions 0

4. WHEN the initiating conditions have cleared, THEN place Control Building Ventilation System in operation in accordance with Section 6.1.4

OOP-37 CONTROL BUILDING VENTILATION SYSTEM OPERATING PROCEDURE Rev. 62 Page 19 of 82 6.1.3 Manual Startup of the Control Building Emergency Recirculation System (continued) Date/Time Completed Performed By (Print) Initials Reviewed By Unit CRS/SRO

DUKE ENERGY BRUNSWICK TRAINING SECTION OPERATIONS TRAINING INITIAL LICENSED OPERATOR SIMULATOR EVALUATION GUIDE 2016 NRC SCENARIO 5 PLACE REP IN AUTO, DIFF TO MOVE ROD, SPE TRIP, IRM FAILURE, DG3IE3/E7 CP LOSS, LOWERING TORUS LEVEL, RHR/CS FAILURES, ED (TORUS LVL) REVISION 0 Developer: o o& Date: O7/7//6 Technical Review: Vac%t Date: 9///2Ol6 Validators: Vwaeo% Date: O/C//6 Facility Representative: Date:

2016 NRC SCENARIO 5 LOt SIMULATOR EVALUATION GUIDE Rev. 0 Page2of6l REVISION

SUMMARY

0 Scenario developed for 2016 NRC Exam.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 3of61 TABLE OF CONTENTS 1.0 SCENARIO OUTLINE 4 2.0 SCENARIO DESCRIPTION

SUMMARY

5 3.0 CREW CRITICAL TASKS 6 4.0 TERMINATION CRITERIA 6 5.0 IMPLEMENTING REFERENCES 7 6.0 SETUP INSTRUCTIONS 8 7.0 INTERVENTIONS 10 8.0 OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES 12 ATTACHMENT 1 - Scenario Quantitative Attribute Assessment 50 ATTACHMENT 2 Shift Turnover 60

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 4of61 1.0 SCENARIO OUTLINE Event MaIf. No. Type* Event Description 1 N-BOP Place 2A RFPT level control in automatic 2 R-ATC Raise reactor power using control rods C-ATC Difficult to move control rod 3 RDO32M C-CRS (AOP) C-BOP 4 K45100 Steam Packing Exhauster Trip C-CRS C-ATC IRM Failure 5 NIO18F C-CRS (TS) 6 ED_IADCGJ6 C-BOP DG3 I E3 I E7 Control Power loss C-CRS (AOP)(TS) Lowering Torus Level I M 7 CAOO2F RHR F028A mech trip I RHR F024B thermal trip I CS FO2OA Handwheel broke (PCCP) 8 RPOO8F Scram I Emergency Depressurization M (RSP)(ATWS)(EDP)

                *(N)ormal   (R)eactivity,   (C)omponent or Instrument,    (M)ajor

2016 NRC SCENARIO 5 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 5of61 2.0 SCENARIO DESCRIPTION

SUMMARY

Event Description Step 6.3.46 of OGP-02, Approach to Criticality and Pressurizations of the Reactor will 1 be completed starting at Step 6.3.46. The crew will raise power by pulling control rods in preparation for placing the Mode 2 switch to RUN. Rod pulls will commence at Step 161 (42-39 @ 12) of the A2X sequence. Control rods will continue to be withdrawn raising power. When control rod 42-23 is selected for withdrawal, it will be stuck at position 12. AOP-02 may be entered and 20P-07, Section 8.2 actions are required to withdraw a difficult intermediate control rod. SPE 2A will trip causing a loss of gland sealing header pressure. SPE 2B will be placed in service While withdrawing control rods, IRM C will fail upscale causing a rod block and half scram. SRO will address IRM A and C inoperability lAW TS 3.3.1.1. Once 5 addressed, I&C will report IRM A is ready to be returned to service following proper channel check. The crew will take the actions of the APP and bypass IRM C and reset the half scram. DC Panel 2A will trip resulting in loss of control power to DG3, Bus E3 and Bus E7. The crew will respond per OAOP-39.O and transfer the control power to alternate. 6 DG3, Bus E3 and Bus E7 are inoperable until transferred to alternate supply. Once control power is transferred, a 7 day action is required to restore to the normal source. The BOP operator will return DG3 to AUTO lAW AOP-39.O. Torus level will begin to lower due to an unisolable leak on RHR suction. If attempted to raise torus water level, on RHR A loop the E11-F028A (Torus Discharge Isol Vlv) 7 will trip when opened, on RHR B loop the El 1-F024B (Torus Cooling Isol VIv) will thermal trip when opened, and on Core Spray the E21-FOO2A (Core Spray Pump A Suction Valve From The Condensate Storage Tank) handwheel will be broke. Before level reaches -5.5 feet in the torus a reactor scram is required. When torus 8 water level reaches -5.5 feet emergency depressurization is required. The crew can anticipate emergency depressurization.

2016 NRC SCENARIOS LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 6of61 3.0 CREW CRITICAL TASKS Critical Task #1 Scram the reactor before torus water level drops below -5.5 feet. Critical Task #2 Emergency Depressutize the reactor when torus water level reaches -5.5 feet. OR Anticipate Emergency Depressurization of the reactor when all attempts to fill the torus have failed. 4.0 TERMINATION CRITERIA When all rods are inserted and the reactor has been depressurized to less than 100 psig the scenario may be terminated.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.O Page 7of61 5.0 IMPLEMENTING REFERENCES NOTE: Refer to the most current revision of each Implementing Reference. Number Title UA-02, 4-5 GLAND SEAL VACUUM LOSS 2OP-26.1, Section 8.1 SHIFTING STEAM PACKING EXHAUSTERS A-05, 2-4 IRM UPSCALE A-OS, 3-4 IRM A UPSCALE/INOP A-OS, 1-7 REACTOR AUTO SCRAM SYS A A-05, 4-7 NEUT MON SYS TRIP A-05, 2-2 ROD OUT BLOCK UA-17, 2-3 DG-31E3 ESS LOSS OF NORM POWER UA-19, 6-3 DG-1 CTL PWR SUPPLY LOST UA-21, 6-2 DG-3 LO START AIR PRESS UA-2 1, 6-3 DG-3 CTL POWER SUPPLY LOST OAOP-39.O LOSS OF DC POWER A-O1, 3-7 SUPPRESSION CHAMBER LVL HI/LO A-05, 5-5 PRI CMT HI/LO PRESS OEOP-O1 -SEP-18 FILLING THE TORUS OEOP-O1-SEP-15 ANTICIPATE EMERGENCY DEPRESSURIZATION

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page8of6l 6.0 SETUP INSTRUCTIONS

1. PERFORM TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 5, Checklist for Simulator Exam Security.

2. RESET the Simulator to IC-06.

3. ENSURE the RWM is set up as requited for the selected IC.

4. ENSURE appropriate keys have blanks in switches.

5. RESET alarms on SJAE, MSL, and RWM NUMACs.

6. ENSURE no rods are bypassed in the RWM.

7. PLACE all SPDS displays to the Critical Plant Variable display (#100).

8. ENSURE hard cards and flow charts are cleaned up

9. TAKE the SIMULATOR OUT OF FREEZE,

10. CLOSE the CS B Loop valves

11. LOAD Scenario File.

12. ALIGN the plant as follows:

Manipulation Insert control rods up to Step 160 of GP-10, Sequence A2X is completed. Raise pressure set to 900 psig Verify level is stable Verify drive water pressure is at 260 psid

13. IF desired, take a SNAPSHOT and save into an available IC for later use.

14. PLACE a clearance on the following equipment.

Component Position IRM A (Blue Tag) Bypassed Core Spray Loop B Red Tag

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page9of6l

15. INSTALL Protected Equipment signage and UPDATE RTGB placard as follows:

a. ADHR I FPCI Demin Transfer Pump

b. All remaining LP ECCS systems

16. ENSURE each Implementing References listed in Section 7 is intact and free of marks.

17. ENSURE all materials in the table below are in place and marked-up to the step identified.

Required Materials OGP-02 up to Step 6.3.46 OGP-10 up to step 161

18. ADVANCE the recorders to prevent examinees from seeing relevant scenario details.

19. PROVIDE Shift Briefing sheet for the CRS.

20. VERIFY all actions contained in TAP-409, Miscellaneous Simulator Training Guidelines, Attachment 4, Simulator Training Instructor Checklist, are complete.

2016 NRC SCENARIO 5 LOI SIMULATOR EVALUATION GUIDE Rev.0 I Page 10 of 61 7.0 INTERVENTIONS TRIGGERS Trig Type ID 4 Dl Override K4510C [STM PACING EXHAUSTER A CLOSE DI] 4 Dl Override - K4510C [STM PACKING EXHAUSTER A CLOSE Dl] 4 DI Override K4S1OC [STM PACKING EXHAUSTER A CLOSE DI] 5 Malfunction NIO18F [IRM C FAILS HI] 6 Remote Function ED_IADCGJ6 [LOAD BKR GJ6 SBD 2A TO 125V P 2A (DG)] 7 Remote Function ED_IADCAPD3 [DG-3 DC BKR CTL PWR ON/OFF] 7 Remote Function EG_0003 [DG-3 LOCKOUT RESET] 7 Remote Function EDIADCADG3- [DG-3 DC BKR CTL PWR (NML=2A ALThU1)] 8 Remote Function ED_IADCABE3 [SWGR E3 DC BKR CTL PWR (NML=2A ALT=U1)] 10 Malfunction CAOO2F [TORUS WATER LEAK] 11 Trigger Command DOD:Q1217LGN 12 Trigger Command - DOD:Q17O7LGN Trig # Trigger Text 11 K1217ENN - [TORUS ISO VLV E11-F028A] 12 K1707]NN [FULL FLOW VLV E11-F024B]

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 11 of 61 MALFUNCTIONS Maif Mult Cun-ent Thet Description Actime Dactime Trig CONTROL ROD WITHDRAWAL RDO32M 42-23 True True SLUGGISH NIO18F IRM C FAILS HI False True 5 CAOO2F TORUS WATER LEAK False True 10 RDO12M 42-23 STUCK CONTROL ROD True True REMOTES

                                   . .                                    Current   Target    Rmp Remf Id        Mult Id  Description                                                              Actime      Trig Value    Value     time BKR CTL DC FUSES CORE SPRAY PUMP 2B          OUT       OUT CS_ZVCS3BT                  E21-FO31B MIN FLOW                           OFF       OFF CS_ZVCS15BT                 E21-FO5B FULL FLOW TEST                      OFF       OFF CS_ZVCSO5BT                 E21-FOO5B INBD INJ VLV                       OFF       OFF CS_ZVCSO4BM                 E21-FOO4B OTBD INJ VLV                       OFF       OFF CS_ZVCSO1BT                 E21-FOO1B TORUS SUCTION                      OFF       OFF CS_VHCSOB                   E21-FO1OB OPEN/CLOSE                         CLOSE     CLOSE ED_IADCGJ6                  LOADBKRGJ6SBD2ATO125VP2A(DG)                 CLOSE     OPEN                         6 ED_IADCADG3                 DG-3DCBKRCTLPWR(NML=2AALT=U1)                NORMAL    ALT                          7 ED_IADCABE3                 SWGR E3 DC BKR CTL PWR (NML=2A ALT=U1)       NORMAL    ALT                          8 EG0003             DG-3     DG-3 LOCKOUT RESET                           NORMAL    RESET    1       00:00:02    7 RH_ZVRH24BT RHZVRH28AM E11-F024B FULL FLOW TEST E11-F028A TORUS ISOLATION J OFF OFF       OFF OFF ED_IADCAPD3                 DG-3 DC BKR CTL PWR ON/OFF                   ON        ON               00:00:01     7
                                                                                                              ]

PANEL OVERRIDES Position! Actual Oven-We Tag ID Description Actime Dactime Trig K4510C STM PACKING EXHAUSTER A CLOSE DI NORMAL ON OFF 4 K4510C STM PACKING EXHAUSTER A CLOSE DI START OFF OFF 4 K4510C STM PACKING EXHAUSTER A CLOSE DI STOP OFF ON 4 Q1217LGN TORUS ISO VLV E11-F028A GREEN ON/OFF OFF ON Q17O7LGN FULL FLOW E11-FO24B GREEN ON/OFF OFF ON

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Pagel2of6l OPERATOR RESPONSE AND INSTRUCTIONAL STRATEGIES EVENT 1: PLACING 2A RFPT CONTROLLER IN AUTOMATIC Simulator Operator Actions Ensure Monitored Parameters is open and Scenario Based Testing Variables are loaded. Simulator Operator Role Play Evaluator Notes Plant Response: Place RFPT Master Controller in Automatic lAW OGP-02, Step 6.3.46 Objectives: SRO Direct BOP to perform Step 6.3.46 of OGP-02 BOP Place RFPT Level Controller is placed in Automatic ATC Monitors plant Success Path: RFPT Master Level Controller is in Automatic and Reactor water level is controlled in band. Event Termination: Go to Event 2 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 13of61 EVENT 1: PLACING 2A RFPT CONTROLLER IN AUTOMATIC Time Pos EXPECTED Operator Response Comments SRO Direct BOP to perform Step 6.3.46 of OGP-02 RO Monitors the plant Place RFPT Master Controller in Automatic lAW BOP OGP-02, Step 6.3.46.

2016 NRC SCENARIO 5 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page l4of 61 APPROACH TO CRITICALITY AND PRESSURIZATION OGP-02 I Rev. 1101I OF THE REACTOR Page 30 of1 6.3 Heating And Pressurization Of The Reactor (continued)

e. B21-1019 (Main Steam Line Drain OtbU Isol Vtv) ............

f, B21-r016 (Main Steam Line Drain lnbd Isd Vlv) ....f iv

46. WHEN reactor feed pump discharge pressure is greater than 900 psig, THEN place C32-SIC-R600 (Mstr RIPT SpiRx Lvi Ctl) in A (automatic) as foliows:....,..............

a. Ensure C32-SIC-R600 (Mstr RIPT SpiRx Lvi Ctl), in M (manual)

b. Ensure Feedwater Control Mode Select in 1 ELEM

c. Depress SEL pushbutton on C32-SIC-R6OIA(B) [RIPT A(B)

Sp cDtlj until A(B) BIAS is indicated and ensure bias is set to 0% U. Depress SEL pushbutton on C32-SIC-R6O1A(B) [RIPI A(B) Sp Ct[) until PMP At B) OEM is displayed

e. Depress SEL pushbutton on C32-SIC-R600 (Mstr RIPT SpfRx Lvi CtI). until MASTR OEM is displayed

f. Using the raise and lower pushbuttons on C32-SIC-R600 Mstr RIPT SpIRx Lvi Ctl). set MASTR OEM to equal the PMP A(B) OEM value displayed on C32-SIC-RGO1A(B)

[RIPI A(B) Sp 0th

g. Depress NM pushbutton on C32-SIC-R60 1 A(B) [RI PT A(B)

Sp Ctl} and confirm the following:

• Indicator on control station changes to A (automatic)
• PMP OEM signal remains unchanged

h. Depress SEL pushbutton on the out-of-seriice C32-SIC-RGO1A(B)[RFPT A(B) Sp Cthl until LVL ERROR is indicated and confirm LVL ERROR is approximately 0 inches Depress NM pushbutton on C32-SIC-R600 (Mstr RIPT Sp/Rx Lvi 0th) and confirm the indicator on the control station changes to A (automatic)

2016 NRC SCENARIOS LOl SIMULATOR EVALUATION GUIDE Rev. ol Page l5of 61 APPROACH TO CRCALlTY AND PRESSURiZATION OGP-02 OFTHEREACTOR Rev 110 Page 31 of54 6.3 Heating And Pressurization Of The Reactor (continued)

j. Confirm signals for PMP AtB) DLM on C32-SIC-R601AE)

[RtPT A(B) S Ctl] and VALVL DLM on TW-LIC-3269 (SULCV Ctl) remain unchanged... k Depress NM pushbutton on rw-Llc.32bg (SULCV Ctl) and confirm the inditor on the control station changes to M(manual) CAUTION Momentarily depressing the raise or lower pushbuttons on rW-LiC-3269 SULCV CU)will cause valve demand to change in increments of 0 1%, Continually depressing the raise or lower pushbuttons will cause valve demand to change at an exponential rate D Using raise pushbutton on rw-LlC-3269 (SULCV Ctl), siowiy open the SULCV until VALV[ DEM is 100% rn. Confirm reactor water level is being maintained between 182 and 192 inches

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Pagel6of6l EVENTS 2/3: RAISE REACTOR POWER I DIFF TO MOVE ROD Simulator Operator Actions Simulator Operator Role Play If asked as the RE, continuous rod withdrawal is allowed. Evaluator Notes Plant Response: Control rods will continue to be withdrawn until control rod 42-23 which is difficult to move, requires DP-07 actions to move. Objectives: SRO - Directs and monitor reactor power ascension with control rods Direct actions for a difficult to move control rod. RD Withdraw control rods to raise reactor power Perform 20P-07 actions for difficult to move control rod Success Path: Control rod 42-23 withdrawn to position 48 by use of increase drive water DP. Event Termination: Go to Event 4 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page l7of 61 EVENTS 2/3: RAISE REACTOR POWER I DIFF TO MOVE ROD Time Pos EXPECTED Operator Response Comments Directs RD to continue to raise reactor power by SRO withdrawing control rods. (Continuous withdrawal allowed). Directs RD to perform 2OP-07. May direct AOP-02 (Control Rod malfunction) Provides notifying RE and Using 20P-07 to move rod. May conduct a brief (see Enclosure 1 on page 56 for format) BOP Monitor reactor plant parameters during evolution. Continues rod withdrawal per GP-10 (see page ATC

18) lAW guidance of 20P-07 (see page 20).

Report A-6 2-7 APRM DOWNSCALE annunciator clears. Recognizes control rod 42-23 will not move. Notifies SRO control rod 42-23 will not move. Identifies 20P-07, Reactor Manual Control System Operating Procedure, Section 6.3.2 (Control Rod Difficult to Withdraw, Control Rod NOT at Position 00) is required. (page 23) Continues rod withdrawal per GP-10 lAW guidance of 20P-07.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page l8of 61 AHACHMENT 3A Page 2 0126 Rod Sequence A2X Withdraw Check Off Sheet (Expanded Group A2) NOTES: Concurrent verification of rod selection s required PRIOR to rod movement.

2. The initials in 0. T. column verify and document the following control rod coupling integrity checks have been performed.

WHEN a control rod is withdrar to the FULL OUT position, a continuous withdraw signal has been maintained for at least 3 to 5 seconds, OR a separate notch out signal has been applied, AND

             -      ROD OVER TRAVEL (A-05, 4-2) annunciator does NOT alarm
             -      ROD DRIFT (A-05 3-2) annunciator does NOT alarm
             -      The Full Out light indication for the selected control rod is not lost
             -      The four-rod display indicates 48 for the selected control rod

3. Initials in the Rod P. I. column confirm that the rod position indications for those positions coveted by that item of the check oft sheet are operable. The RWt1 Inferred Rod Position capability may be used as an alternate method to determine Rod Position.

4. During manipulation of control rods a second Licensed Operator shalt monitor control rod selection and movement. This individual shall ensure correct placement of control rods, and document these verifications by initialing the Rod Sequence Check Off Sheet, IF inoperable Rod Position indication necessitates inserting the rod in question one notch further than its insert/withdraw limit and bypassing the rod on the RWM, THEN the second Licensed Operators initials also documents verification of this action.

I R8 I 5. Any deviation from the original rod move sequence should be reviewed by the Reactor Engineer, authorized by the Unit CRS, and documented on the proper rod sequence check off sheet. For changes in direction or control rod double notches, the affected page(s) of the sequence pull sheet must be copied, rod move documented, then the documentation must be included with the original rod sequence attachment OGP-lO Rev.43 Page lO4ot3l4

2016 NRC SCENAPJO5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 19of61 AflACHMNT 3A Page 100126 Rod Sequence A2X Withdraw Check Off Sheet (Expanded Group A2) Correct RcO Rod Selected And Postio Act OierTtael Rod P lrtals Lta5 Cornmen Number Vered Frorr.To Positton [Ncte j [Nete 3] [Note 4] [Note_1] STEP 9 BPWS4) 137 50-31 I 03 to 12 N/A 138 42-07 1 03 to 12 N/A 139 10-07 1 03 to 12 WA 140 02-3 1 I 03 to 12 N/A 141 10-39 / O3to 12 WA 142 13-47 I 08 to 12 N/A 143 34-47 I 08to12 N/A 144 42-39 1 08 to 12 N/A 145 42-23 1 08 to 12 N/A 146 34-15 I 08 to 12 N/A 147 26-07 1 08 to 12 NIA 148 18-15 I 08to12 N/A 149 10-23 08 to 12 WA 150 18-31 t 08 to 12 N/A 151 26-39 1 08 to 12 N/A 152 34-31 I 08 to 12 N/A 153 26-23 I 08 to 12 N/A STEP 10 (BPWS 4 154 50-31 1 12to48 155 42-07 I 72to48 156 10-07 / 12to48 157 02-31 / 12to48 156 10-39 I 12to48 759 18-47 1 12to46 160 34-47 / 12to48 161 42-39 / 12 to 48 162 42-23 1 12 to 48 163 34-15 I 12 to 48 164 26-07 1 12to48 165 18-15 t 12to48 166 10-23 1 12to48 167 18-31 / 12to48 168 26-39 I 12to48 169 34-31 1 12to48 170 26-23 I 12 to 48 Notes (for further detads see Page 2 o this attachmenty Concurrent Venf cation of rod select on is required PROR to rod movement

      -3 Inti.s in the Over Traver column slgn-y c iptetion of control rod ccuol rig itegfty ctecks for utly w.thdrawn ccntrol rods 3     Intals in the Rod P     - column confrm th3t the rod position r.dcatjons for those positions coverec by that tern c the check o she.t are operable 4    Column used oya Second Licensed Operator to document monitoring o ccntrot rod selecton and movement to ensure correct placement of ccntrol rods OGP-10                                                        Rev. 43                                        Page 11201 314

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 20 of 61 REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-Ol PROCEDURE Rov 1O Page 153 of 162 ATTACHMENT 15 Page 1 of 3

        <<(Reference Use)     - Section 6.1.1 Continuous Control Rod Withdrawal>>

NOTE The purpose of this attachment is to provide the Reactor Operator with guidance for control rod movement and use OENP-23.5, Reactivity Control Planning. and General Operating Procedure pull sheets as the place keeping tool for execution of steps D BEGIN R.M. LEVEL R2(R3 REACTIVITY EVOLUTION

1. Select control rod by depressing its Control Rod Select pushbutton.

2. Confirm the following:

• The backlighted Control Rod Select pushbutton is brightly ILLUMINATED.
• The white indicating light on the full core display is ON
• Rod Withdrawal Permissive indication is ON

3. Continuously withdraw control rod to position designated on General Operating Procedure or OLNP-21 .5. Reactivity Control Planning, pull sheets by holding Lmergency Rod in Notch Overnde switch to OVERRIDE, while simultaneously holding Rod Movement switch to NO CH OU1.{8J 2) 4, Monitor control rod position and nuclear instrumentation while withdrawing the control rod.

S. IF control rod fails to withdraw, THEN go to Section 6.3.1. Section 6.3 2, Section 6.3.1, or Secbon 6.3.8 to free the control rod and return to Attachment 15 Step 6.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page2lof6l REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-01 PROCEDURE Rev, 105 Page 154 of 162 ATTACHMENT 15 Page 2 of 3

         <<(Reference Use)        - Section 6.1.1 Continuotis Control Rod Withdrawal>>

6. IF the control rod is being withdrawn to an intermediate position THEN perform the following:

a Before control rod reaches the position designated on General Operating Procedure or 0[NP-23 5. Reactivity Control Planning, pull sheets, release Rod Movement and Emergency Rod In Notch Ovemde control switches.{8. 1 .2}.

b. Ensure control rod settles into desired position.

c. Confirm rod settle light is OFF.

7. IF the control rod is being fully withdrawn to position 48 THEN perform the following:

NOTE A continuous withdraw signal of approximately 3 to 5 seconds is sufficient time to ensure the control rod remains coupled. Longer continuous withdraw signals may be utilized if a control rod flush is desired [1

a. WHEN control rod reaches position 48, THEN perform either of the following:

• Maintain a continuous withdraw signal for the desired time
• Apply a separate notch withdraw signal,

b. Confirm control rod does Qj. retract beyond position 48 (Technical Specification SR 3.1.34).

c. Release Rod Movement and Emergency Rod In Notch Ovemde switches, if used, d Ensure control rod settles at position 48,

e. Confirm rod settle light is OFF.

f. Confirm control rod reed switch position indicators agree with FULL OUT indication on full core display.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 22 of 61 REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-0! PROCEDURE Rev. 105 Page 155 of 162 ATTACHMENT 15 Page 3 of 3

      <<(Reference Use)   - Section 6.1.1 Continuous Control Rod Withdrawal>>

8. Repeat Attachment 15 Step 1 through Attachment 15 Step LI, of this Attachment, for the remainder of the control rods requiring movement, using General Operating Procedure or OENP-24.5.

Reactivity Control Planning, pull sheets.{8.1.2}. END RM. LEVEL R21R3 REACTIVITY EVOLUTION

9. WHEN control rod movement is NO longer required THEN go to Section 6.1.1 Stop L

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page23of6l REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-O? PROCEDURE Rev. 105 Page 38 of 162 6.3.2 Control Rod Difficult To Withdraw And Control Rod NOT At Position 00 Control Rod

1. Record Control Rod Number above

2. Confirm the fotloving initial conditions are met:

• All applicable prerequisites in Section 5.0 are met
• Control rod will NOT withdraw in accordance with Section 61.1
• Control rod is NOT at position 00 a Unit CRS has consulted Technical Spccifitions 3.1 .3, Control Operability, and 3,3,2.1 Control Rod Block Instrumentation for the required actions pnor to the performance of Section 6.3.2, Control Rod Difficult to Withdraw And Control Rod NOT At Position 00 CR5

3. Ensure failure of the control rod to withdraw is NOT the result of a rod block from the RWMorRBM

4. Notify the Reactor Engineer Reactor Engineer CAUTION If reactor pressure is less than or equal to $00 psig jj higher than normal CRD drive water pressure is used to withdraw a control rod, then the latching function of the CRD may be lost D BEGIN R.M. LEVEL R2/R3 REACTIVITY EVOLUTION

5. Attempt to withdraw the control rod using 300 psid drive header differential pressure as follows:

a. Raise CRD dnve differential pressure to 300 psid b Attempt to withdraw control rod

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page24of6l REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-01 PROCEDURE Rev. 10 Page 39 of 162 6.3.2 Control Rod Ditficult To Withdraw And Control Rod NOT At Position 00 (continued)

c. IF control rod moves, THEN immediately restore drive pressure to 260 to 215 psid and attempt to withdraw rod in accordance with Section ( 1.1

• IF control rod will NOT continue to withdraw at normal drive pressure, THEN return dnve differential pressure to 300 psid and withdraw rod tn accordance with Section 6.1.1

d. IF control rod withdraws, THEN go to Section 6.3.2 Step 16

e. Repeat Section 6.3.2 Step 5.b and Section 63.2 Step 5.c. as necessary

6. Attempt to withdraw control rod using 350 psid dnve header differential pressure as follows:

a. Raise CRD Unve differential pressure to 350 psid . . -

b. Attempt to withdraw control rod

c. IF control rod moves, THEN immediately restore drive pressure to 260 to 2/5 psid and attempt to withdraw rod in accordance with Section 6.1.1

• IF control rod will NOT continue to withdraw at normal dnve pressure, THEN return Unve differential pressure to 350 psid and withdraw the rod in accordance with Section6l.1

d. IF control rod withdraws.

THEN go to Section 6.3.2 Step 16

16. Lower control rod drive differential pressure to betv&en 260 and 215 psid

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 25 0161 EVENT 4: STEAM PACKING EXHAUSTER (SPE) TRIP Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 4 to trip the A SPE. Simulator Operator Role Play If contacted as l&C to investigate, acknowledge the request. If asked to investigate MCC 2TA for the SPE, report that compartment CA6, OG-SPEM-A (Steam Seal SPE Motor 2A) is tripped. If asked as AC to Open 2-MVD-V52 float trap outlet valve for 2B SPE report that the valve is Open. If asked as AC to Close 2-MVD-V51 float trap outlet valve for 2A SPE report that the valve is Closed. Evaluator Notes Plant Response: The SPE trips and the exhauster valves close. APP UA-2 4-5 Gland Seal Vacuum Loss annunciates. The BOP will start the B SPE and place in service to maintain vacuum. Objectives: SRO - Direct B SPE started. RO - Diagnose A SPE failure and Starts B SPE. Success Path: SPE B is started and vacuum returned to normal. Event Termination: Go to Event 5 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 5 LOt SIMULATOR EVALUATION GUIDE Rev. 0 Page 26 of 61 EVENT 4: STEAM PACKING EXHAUSTER (SPE) TRIP Time Pos EXPECTED Operator Response Comments CR5 Direct I&C to investigate May direct entry into OAOP-37.0, Loss of Condenser Vacuum. Direct the B SPE to be started May conduct a brief (see Enclosure 1 on page 56 for format) ATC Monitors the plant. Acknowledges, refers to & reports annunciator UA-2 4-5 GLAND SEAL VACUUM LOSS May announce and enter OAOP-37.0, Loss of Condenser Vacuum. BOP Performs actions of APP (page 27) Starts standby SPE lAW 2OP-26.1 Section 6.3.1 (See page 28) Closes OG-MOV-D1 (Steam Seal SPE 2A MO Disch Vlv)

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 27 of 61 2 A;UA:: 45

;LAl2 SEAL  AUV LOSE AVID AICN3 I. Eteaa ta:T,g        hauster n:t :tera:.ng.

1. Steam parc.mg e:thaus:er surt:nd :hare valve thr:::led :l:sef, sr ;land seal vatuum :ann:t te mntaine.

3. lube leaks .n gland ezhuster ::ndenser.

4. Steam Seal Feed Valve :r the Steam a:kinr lfr.g lves are

n:r:.ng steam seal header pressure :crre::ly.

S. Iuffi:er.t ::ndemsate fl: thr:u;h the cam pa:cn e:thauster. E. .r:ut raIfun:t:.n.

1. lar.d seal va*:uum :n CIE9SE bel:w S n:hes :f

2. Om ::Ter varuur. de:reasr.r.

3. iressd :ffgas fl:w.

4. Steam seal header resstre AIIICUS Start str.dbv ezhuster and ad: ust tt s d:s :hare ae t:

           *_ been 1             and 1 hes             r 2O1.l.

1. If gland seal re;ulat:r is n:t :pertin pr:erly. refci 2A-Th12 3-5, SlEA SEAL PELA2D ESS-LCW.

3. Drain the I: seals :n the ser;ire steam pa:kng eauster.

4. If steam seal header ressurc is ab:v 7 sig, :he:k the Steam Seal Feet! alve M335F7 peratin; r:rrectly as frlI:ws:

a. Thrittle cl:sed, Mn Steam t: Seals VI: MVS1 steam seal header pressure between 1.5 and 4.0 psi;.

b. If thr:ttling the MVD3i was sutreasful in sting steam seal header p saute, then byass the Steam Seal Feed Valve zer 2Or1.l Ee:tirr. 3,3.

If :hr:ttling the MVD3l was nt: surresaful then re:pen and bypass the Steam Seal Unl:adng Valve by thr::tling :en Steam Seal Syas Unl:ad VI:, If n:s:ure s suspe::ed n Steam a:k:ng E:thaua:rir.en: lines, drain per 2D1E.l Ee:ti:n 3.2. Switrh S nches f water 2APP-UA-02 Rev. 45 Page 56 0194

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page28of6l GLAND SEALING STEAM SYSTEM OPERATING 20-26.1 PROCEDURE Rev. 46 Page 140132 6.3 Infrequent Operation 6.3.1 Shifting Steam Packing Exhausters Confirm the following initial conditions ate met: a Gland Seahng Steam System is in operation per Section 6.1

b. Condensate System is in service and is aligned to supply adequate flow to the SPE per 20P-30 section for Swapping Off-Gas Trains Dunng Normal Conditons

2. IE Steam Packing Exhauster SPE 2A is operating, THEN perform the following a Open 2-MVD-V52 (Float Trap Outlet Valve) for SEP 26

b. Start Steam Packing Exhauster SPE 26

c. Ensure OG-MOV-E2 (Steam Seal SPE 26 MO Inlet Vlv) is OPEN

d. Throttle closed OG-MOV-D 1 (Steam Seal SPE 2A MO Disch Vlv) and throttle open OG-MOV-D2 (Steam Seal SPE 26 MO Disch Vlv) while maintaining OG-PI-EPT-9 (Steam Packing Exhauster Vacuum) located on Panel XU-2. between 10 and 20 inches water vacuum

e. Ensure OG-MOV-D1 (Steam Seal SPE 2A MO Disch Vlv) is CLOSED

f. Stop Steam Packing Exhauster SPE 2A

g. Close 2-MVD-V51 (Float Trap Outlet Valve) for SPE 2A

h. Ensure OG-MOV-E1 (Steam Seal SPE 2A MO Inlet Vlv) is CLOSED

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 29 of 61 EVENT5: IRMCFAILURE Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 5, to fail IRM C upscale. Simulator Operator Role Play If contacted as the RE for IRM C inoperability, acknowledge request. When IRM C inoperability has been addressed and by Lead Examiners direction, contact the control room as WCC SRO and report IRM A can be declared Operable following a satisfactory channel check. Once declared operable the off normal tag can be removed and the WCC will follow up with the paperwork. Evaluator Notes Plant Response: The crew will continue raising power by pulling control rods in preparation for placing the Mode switch to RUN. IRM C will fail upscale causing a rod block and half scram. Objectives: SRO Determine Technical Specification application. RD - Perform actions for IRM C failure Success Path: Declare IRM A operable by channel check and bypass IRM C. Event Termination: Go to Event 5 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 30 of 61 EVENT 5: IRM C FAILURE Time Pos EXPECTED Operator Response Comments SRO Directs APP reference. Contacts I&C for IRM C failure. May contact Shift Manager also. References TS 3.3.1.1 and determines with IRM5 A & C inoperable: Condition A is applicable for Function la Required Action A.1 Place channel in trip is required within 12 hours or A.2 Place associated trip system in trip is required in 12 hours. May enter TRM 3.3 (Control Rod Block Instrumentation) Function 3 Condition A, Tracking LCO May conduct a brief (see Enclosure 1 on page 56 for format) Evaluates IRM A operability following satisfactory channel check. 20P-09, Attachment 4, 2.3.4 (Operability Guidance). NOTE: WCC provides cue that IRM A can be declared operable after channel check is SAT. Channel Check definition in the RO DSR. Channel Checks are a sufficient WO PMT for SRMs and IRMs at power unless a component failure is suspected in which case an IN curve and TDR trace is desirable Directs IRM A channel check be performed. BOP Plant Monitoring:

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 31 of6l EVENT 5: IRM C FAILURE Time Pos EXPECTED Operator Response Comments ATC Determines IRM C failed upscale. Responds and reports applicable alarms for IRM C failing upscale. A-05 7-7 REACTOR AUTO SCRAM SYS A 4-7 NEUTMONSYS TRIP 2-4 IRM UPSCALE 2-2 ROD OUT BLOCK 3-4 IRM A UPS CALE/INOP A-5 IRM A UPSCALE/INOP actions: May Reposition range switch for IRM C to bring indicated power to between 15 and 50 on the 0-125 scale. May verify IRM C Drawer Selector switch (Control Panel H12-P606) is in OPERATE. May notify CRS of Tech Spec applicability May inform CRS IRM C cannot be bypassed and half scram cannot be reset due to IRM A being bypassed. Performs channel check of IRM A for operability. RD DSR Item # 9 (IRM channel check) 201-03.2, Definition 5.1. Removes IRM A from Bypass Bypasses IRM C per APP guidance. Resets half scram per APP guidance.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 32 of 61 Ag AIS 24 a;e 1. of 2 IP.M A UPSCAiEfITQ AITC AIflONS I. P:d w hirawal bi:: byassed when r:r:r mofe 7;t:h is n PV.

2. P.ea:t:r halfStrain bypassed when r art:r .ode swi:rh 2.S in yy, CAJSE

C ei.s A, C, 2, cr Iriratlng greater than or eua :r ll on the l2

2. lt nnels A, C, 2, or S inve sirna.s:

a. I2 drawer lert:r swotrh n:t n :pera:e

b. IP1 drawer r.odule unpiurged.

2i Uetert:r hgh v:lta;e rwer supply 12w voltage.

3. P! A, C, 2, or S detertor failure.

4. propr rangIng if 1P2! A, C, I, :r S range a :t:hes during rea:t:r startup or shutdown.

irruit lfnrt:n. C5.SEPVATIONS I. 2P Channel A, C, 2, 5 drating greater than or egual to lL on the l2 soale.

2. P2ACICP AVID SA A ACS l-7 alarni.

3. POD CUT ELCC: AJS 22 alarm.

4. NEVI ON S TPP AOS 47) alarm.

P. VISALE )AS 2-4 alarm. E. IP2 Channel A, C, 2, or S uparale trip or mop VS C T OP INDI rod Inii:atIr.g liqh: s on.

7. The rod withdrawal oemni.ssive diratlng light w:ll be off.

1. X:nt:r P Channels A, C, 2, and S to deermlne afferted rhannels

2. If a sudden rise n ind:ra:ed rearror power orrurred In more than one rnel, insert in seruenre control rods as neoessary to turn the tower ncrease and verify that the correct rod wIthdrawal sequence os being used.

3. P.ep:si:ion range swotch f:r iP_I A, C, 2, :r S to bring :ndrcated power to between IS and SC on the 3125 scale.

4. Verify that IPCC A, C, I, and S Drawer Selert:r swt:hes (Control ganel 12IE are :n CEPAIE.

2APP-A-05 Rev. 68 Page 43 of 96

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.O Page33of6l

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a. fr tt T:hnrai :f.:ar::n end fir

terab.tv rrrnrs.

b. :r:f rh.

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C53I2.E LAfl EFFECTS I. P.a:tor Srrazr .f AS Tr:t .s an 2P. channe.. . asad cr :n:prabi., :hn.nai f.ca:.:n ICC :n A C:na::r ,asur nay naau.t. EFEENES Te:hnicai 5p.c;f.:ation 3.3.1.1, TAM 3.3 A A-CE 1-, AEACTCA AC SAA! 3Y5 A A A-CE 2-2, AGO CUT EIOC A ACE 24, ACC SAIE A ACE 47, NEU MON 2APP.A-05 Rev. 68 Page 44 of 96 1

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 34 of 61 EVENT 6: DG3 I E3 I E7 LOSS OF CONTROL POWER Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 6 to trip 125 VDC Panel 2A. When requested to align alternate control power: Initiate Trigger 7, to align alternate control power to the DG3 and to reset DG3 local DG engine control panel lockout. Initiate Trigger 8, to align alternate control power to E3/E7. Simulator Operator Role Play Acknowledge/reset Unit One alarms, as necessary If contacted as TBAO, report Switchboard 2A load breaker GJ6, Feed to Panel 2A, is tripped. If contacted as l&C, report problem is a due to GJ6 breaker failure, not a fault on the system. If contacted as I&C to verify alternate power to ESS cabinet, report ESS cabinet has transferred to alternate power. Evaluator Notes Plant Response: DC Panel 2A will trip resulting in loss of control power to DG 3, Bus E3 and Bus E7. The crew will respond per OAOP-39.O and transfer the control power to alternate. DG3, Bus E3 and Bus E7 are inoperable until transferred to alternate supply. Once control power is transferred, a 7 day action is required to restore to the normal source. The BOP operator will return DG 3 to AUTO lAW AOP-39.0. Objectives: SRO - Directs AOP-39 and APP actions Evaluate TS 3.8.1 and 3.8.7. RO Perform AOP-39 actions. Success Path: Restore DG3 control power and then return DG3 to Auto. Event Termination: Go to Event 7 at the direction of the Lead Evaluator.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page35of6l EVENT 6: DG3 I E3 I E7 LOSS OF CONTROL POWER Time Pos EXPECTED Operator Response Comments Direct actions of APPs: UA-21 6-2, DG-3 LO START AIR PRESS UA-21 6-3, DG-3 CTL POWER SUPPLY LOST SR UA-196-3, DG-1 CTLPWRSUPPLYLOST UA-17, 2-3, DG-3/E3 ESS LOSS OF NORM POWER Direct actions of OAOP-39.0, Loss Of DC Power Contact l&C to verify ESS cabinets have transferred to alternate power. Direct transfer of control power to alternate source Direct returning DG3 to Auto Determine Tech Specs 3.8.1 AC Sources Operating, Condition D applies. (until alt power established) D.1 Perform SR 3.8.1.1 within 2 hours and once per 12 hours AND D.3.1 Determine OPERABLE DGs not inoperable due to common cause failure 24 hours OR D.3.2 Perform SR 3.8.1.2 for OPERABLE DGs 24 hours AND Restore DG to OPERABLE status 7 days 3.8.7 Distribution Systems Operating, Condition C applies. 0.1 Declare required features supported by the inoperable DC electrical power distribution system inoperable Immediately. AND C.2 Initiate action to transfer DC electrical power distribution subsystem to its alternate DC source Immediately AND C.3 Declare required features supported by the inoperable DC electrical power distribution subsystem OPERABLE Upon completion of transfer of the required features DC electrical power distribution subsystem to its OPERABLE DC source. AND 0.4 Restore DC electrical power distribution subsystem to OPERABLE status 7 Days. May conduct a brief (see Enclosure 1 on page 56 for format)

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 36 of 61 EVENT 6: DG3 I E3 I E7 LOSS OF CONTROL POWER Time Pos EXPECTED Operator Response Comments ATC Monitors the plant. Report annunciators and review APPs: UA-21 6-2, DG-3 LO START AIR PRESS UA-21 6-3, DG-3CTLPOWERSUPPLYLOST BO UA-19 6-3, DG-1 CTL PWR SUPPLY LOST UA-17, 2-3, DG-3/E3 ESS LOSS OF NORM POWER Announce and enter OAOP-39.0, Loss of DC Power. (see page 37)

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 37 of 61 LOSS OF DC PNER OAOP-39.O Rev. 042 Page 7 of 34 4.2 Supplementary Actions Loss of Battery Chargers: a Monitor 125V and 24V DC battery voltages.... 0

b. IF por has been removed from me battery chargers for greater than 1 hour, THEN remove selected loads from the battery based on 001-50, 125/250 and 24/48 VDC Electncal Load List and Unit CRS direction 0

c. Before i25V DC battery voltage reaches the low voltage limit of 105 volts, remove loads as directed by the Unit CR5 as necessary to maintain battery voltage greater than 105 volts 0

d. Before 24V battery voltage teaches the low voltage limit of 21 volts, remove loads as directed by the Unit CRS as necessary to maintain battery voltage greater than 21 volts 0

e. IF battery charger AC power has been lost due to Station Blackout, THEN enter 1EOP-01-SBOf2EOP-01-SBO), Station Blackout 0

2. Loss of Any DC Panel.

a Determine which panel has been lost using Attachment 3, Annunciators Associated with Losses of Various DC Panels, if necessary 0

b. Dispatch an operator to investigate the cause of the loss of DC power 0 c Contact Duty l&C to determine actual electrical system ground conditions prior to transferring any panel to alternate source or reenergizing from the normal source 0 U IF l&C determines a panel is faulted, THEN DO NOT reenergize the panel until the fault is isolated 0

e. Refer to 001-50, 125/250 and 24/48 VDC Electrical Load List, for specific load information 0

f. IF switchyard control power OR 4KV bus control power is lost, THEN request the Load Dispatcher to minimize grid operations 0

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 38 of 61 LOSS OF DC POWER OAOP-39.0 Rev 042 Page 8 ot 34 4.2 Supplementary Actions (continued)

g. Using the lollowing table, determine the appropriate section based on the DC panel lost I]

(1) Go to the appropriate section for additional actions D Unit Div. DC Panel Lost Normal Power Procedure Section Supply to Panel 3A, 5A, 1 lA IA-i Section 4.2 Step 3 on page 8 1A, 7A 1A-2 Section 4.2 Step 4 on page 10 1B7B,3AB lB-i Section4.2Step5on page 14 3B, 11 B, 9A 1 B-2 Section 4.2 Step 6 on page 20 4A. 6A, 1 2A, 17 2A-i Section 4.2 Step 3 on page 8 2A. 8A 2A-2 Section 4.2 Step 4 on page 10 2 2B, 8B, 4AB, 13, MWT 2B-1 Section 4.2 Step 5 on page 14 48, 12B, 1OA 26-2 Section 4.2 Step 6 on page 20

2016 NRC SCENARIO 5 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 39 of 61 LOSS OF DC POWER OAOP-39.0 Rev. 042 Page ii ot 34 4.2 Supplementary Actions (continued) (3) WHEN directed, THEN perform the followwlg at Sub E5 C (a) Open Sub E5 1-E5-FM9-72-NORM (Normal Control Power C1rcuit Breaker), inside Cornpt. FM9 C (b) Close Sub E5 i-E5-FM9-72-ALT (Alternate Control Power Circuit Breaker), inside Cornpt. FM9 C

b. IF loss of DC Distribution Panel 2A has occurred, THEN dispatch an operator to the Diesel Generator Building C (1) WHEN directed, THEN perform the following for DG3 C (a) Open DG3 Normal Feed 8, nomal diesel generator control power breaker, in the rear upper right inside of the Excitation Control Panel C (b) Close DG3 Alternate Feed 8A. alternate diesel generator control power breaker, in the rear upper right inside of the Excitation Control Panel C (C) Confirm the Governor Control At Setpoint light is LIT within 10 seconds after control power has been restored C fd) IE the Governor Control At Setpoint light does NOT light THEN initiate a WOIWR C (e) Depress Lockout Reset pushbutton, on the local diesel generator engine control panel C (0 Confirm diesel generator Avail light on Panel XU-2isON C

() Depress DG3 Auto Switch push button on RTGB Panel XU-2 C

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 40 of 61 LOSS OF DC PYNER OAOP-39.0 Rev. 042 Page 120134 4.2 Supplementary Actions (continued) (2) WHEN directed, THEN perform the following at Bus E3 D a) Open Bus E3 125 Volt E3 Bus Normal Control Power breaker inside Compt. A14 0 fb) Close Bus E3 125 Volt E3 Bus Alternate Control Power breaker inside Compt Al4 0 (3) WHEN directed, THEN perform the following at Sub EL 0 (a) Open Sub E7 2-E7-FN 1-72-NORM fSwgr 125V DC Normal Control Power Circuit Breaker), inside Compt. FN1 0 fb) Close Sub E7 2-E7-FN 1-72-ALT (Swgr i25V DC Alternate Control Power Circuit Breaker), inside Cornpt. ENi 0

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 41 of6l LOSS OF DC PVER OAOP-39.0 Rev. 042 Page 13 of 34 4.2 Supplementary Actions (continued) d IF loss of DC Distribution Panel 2A has occurred, THEN confirm ESS Panel 1-160 is OPERABLE by performing the following: C ( 1) Alternate source from Battery Bus 1 A-I, Panel 3k is OPERABLE C NOTE . Loadside is the right side of the terminal strip C Drawing F-09118-i is the interconnection wiring diagram for ESS Panel H60 C (2) Request l&C to determine power is available indicated by measurement of 125 VDC system voltage between the following points in ESS Panel H60.

• Loadside of FU-2 to loadside of EU-4 C
• Loadside of EU-S to loadside of EU-8 C
• Loadside of FU-lO to loadside of FU-12 C
• Loadside of EU- 14 to loadside of EU-i 6 C
• Loadside of EU-is to loadside of FU-20 C

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 42 of 61 EVENT 7: LOWERING TORUS WATER LEVEL I ATTEMPT TO FILL TORUS Simulator Operator Actions At the direction of the Lead Evaluator, Initiate Trigger 10 to start Torus Water Leak NOTE: It will take 24 minutes to reach -5.5 feet in the torus. Simulator Operator Role Play If contacted to look for leaks in the RB -17 elevation, after 5 minutes report none found. When directed to open E21-FOO2A (Core Spray Pump A Suction Valve From The Condensate Storage Tank), wait 3 minutes and report that the handwheel is broken, the valve cannot be opened. When directed to align RHR Loop A, wait 3 minutes and report SEP-18 Section 2.2.3 Steps 5a-c are complete. If directed to investigate F028A breaker, report overcurrent trip, will not reset if asked. When directed to align RHR Loop B, wait 3 minutes and report SEP-18 Section 2.2.3 Steps 6a-c are complete. If directed to investigate F024B breaker, report thermal trip, will not reset if asked. Evaluator Notes Plant Response: Torus level will begin to lower due to an unisolable leak on RHR suction. If attempted to raise torus water level, on RHR A loop the Eli -F028A (Torus Discharge Isol Vlv) will trip when opened, on RHR B loop the El l-F024B (Torus Cooling tsol Vlv) will wilt trip when opened, and on Core Spray the E21-FOO2A (Core Spray Pump A Suction Valve From The Condensate Storage Tank) handwheel will be broke. Objectives: SRO -Direct actions for a lowering torus water level lAW PCCP RO Respond to a lowering torus water level lAW PCCP. Success Path: Attempts to add water to torus through RHR and Core Spray systems. Event Termination: When torus fill through RHR / CS has been attempted or a reactor scram inserted.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page43of6l EVENT 7: LOWERING TORUS WATER LEVEL I ATTEMPT TO FILL TORUS Time Pos EXPECTED Operator Response Comments SRO Direct actions of PCCP. Direct torus fill lAW OEOP-O1 -SEP-18 May conduct a brief on when Reactor Scram is required (see Enclosure 1 on page 56 for format) Report annunciator ATC A-0i 3-7, Suppression Chamber Lvl Hi/Lo Diagnose lowering torus water level. When directed by the CRS, perform OEOP-0i-SEP-i 8, Filling the Torus. (page 45) If RHR Loop A is selected, report unable to fill due to El 1 -F028A (Torus Discharge Isol Vlv) breaker magnetic trip. If RHR Loop B is selected, report unable to fill due to Eii-F024B (Torus Cooling Isol Vlv) breaker thermal trip. If CS Loop A is selected, report unable to fill due to E2i-FOO2A (Core Spray Pump A Suction Valve From The Condensate Storage Tank) handwheel broken.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 44 of 61 EVENT 7: LOWERING TORUS WATER LEVEL I ATTEMPT TO FILL TORUS BOP Monitors the plant Report A-05 5-5, Pri Cmt Hi/Lo Press Dispatch AD to look for the leak.

LOl SIMULATOR EVALUATION GUIDE 45 of 1.0 ENTRY CONDITION

• As directed by Emergency Operating Procedures (EOPs)
• As directed by Severe Accident Management Guidelines (SAMGs) 2.0 INSTRUCTIONS 2.7 Core Spray Torus Fill 2.1.1 Manpower Required
• 1 Reactor Operator
• 2 Auxiliary Operators 2.1.2 Special Equipment None 2.1.3 Core Spray Torus Fill Actions I. Select Core Spray loop to be used RO A B 2 Confirm Core Spray loop to be used

              *      !I in operation RO

• Suction aligned to torus D RO

3. Monitor and control csi level greater than 11 feet D AC

4. Monitor torus level RO

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 46 of 61 FILLING THE TORUS OEOP-0l -SEP-18 Rev. 000 Page 501 12 21.3 Core Spray Torus Fill Actions (continued) NOTE Normal tows level is -27 to -31 inches 0

5. IF Core Spray Loop A selected, THEN:

a. Unlock and slowly throttle open E2 1-FOO2A (Core Spray Pump A Suction Valve From The Condensate Storage Tank) ...... D AO b WHEN at desired tows level.

THEN 0 RO Close E21-FOO2A (Core Spray Pump A Suction Valve From The Condensate Storage Tank) 0 AO

2016 NRC SCENARIOS LOl SIMULATOR EVALUATION GUIDE Rev. oJ Page47of6l FILLING THE TORUS OEOP-Oi-SEP- 18 Rev 000 Page 6 of 12 2.2 RHR Torus Fill 2.2.1 Manpower Required

• 1 Reactor Operator
• 2 Auxdiary Operators 2.2.2 Special Equipment None 22.3 RHR Torus Fill Actions

1. Select RHR loop to be used 0 RO A B 2 Confirm RHR loop to be used NOT in operation 0 RO

3. Monitor and control MUD tank level greater than 14 feet 0 AO

4. Monitor tows level 0 RO NOTE Normal tows level is -27 to -31 inches 0

5. iF. RHR Loop A selected, THEN:

NOTE Valves located on HPCI mezzanine 0 a Close Eli -Vi 95 (RHR Keeptill Station Outlet Isolation Valve) 0 AO b Close El1-V194 fRHR Keepfill Station Inlet Isolation Valve) 0 AO c Open El 1-F082A (RHR Loop A Keeptill Station Bypass Valve) 0 AO

2016 NRC SCENARIO 5 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 48 of 61 FILLING THE TORUS QEOP-O 1-SEP-18 Rev. 000 Page 7 of 12 2.2.3 RHR Torus Fill Actions (continued)

d. Place El 1-CS-S 18A (2/3 Core Height LPCI Initiation Override Switch) to MANUAL OVERRD 0 RO

e. Momentarily place El 1-CS-S 17A (Containment Spray Valve Control Switch) to MANUAL a RO

f. Open El 1-F028A (Tows Discharge Isol Vlv) a RO g Sloviy throttle open El l-F024A (Torus CooIng Isol Vlv) a RO h WHEN at desired tows level.

THEN close El 1 -F024A (Tows Cooling 1501 Vlv) a RO

i. Close El l-FO28A (Tows Discharge lsol VIv) a RO

j. Close El 1-F082A (RHR Loop A Keeptill Station Bypass Valve) a AC

6. J.E RHR Loop B selected, THEN:

NOTE Valves located on Reactor Building 50 west a a Close E11-F098 (RHR Keeptill Station Outlet Isolation Valve) a AC ii Close El 1-F099 (RHR Keepfill Station Inlet Isolation Valve) a AC

c. Open El t-F088 fRHR Loop B Keepfill Station Bypass Valve) a AC d Place El 1-CS-Si86 (2/3 Core Height LPCI Initiation Override Switch) to MANUAL OVERRD a RC

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 49 of 61 FILLING THE TORUS OEOP-0 1-SEP-18 Rev. 000 Page sof 12 2.2.3 RHR Torus Fill Actions (continued)

e. Momentanly place El 1-CS-Si 7B (Contatnment Spray Valve Control Switch) to MANUAL C RO

f. Open El i-F028B (Tows Discharge Isol Vlv) C RO

g. Slowly throttle open El 1-F024B (Tows Cooling lsol \Iv) C RC) h WHEN at desired tows level, THEN close El l-F024B (Torus Cooling 1501 Vlv) C RO Close El 1 -F028B (Tows Discharge lsol VIv) C RO

j. Close Ei1-F088(RHR Loop B Keepflu Station Bypass Valve) C AC

7. Exit this section and go to Section 2.4 C RO

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 50 of 61 EVENT 8: SCRAM I EMERGENCY DEPRESSURIZATION Simulator Operator Actions When directed by the Lead Evaluator, place the simulator in FREEZE DO NOT RESET THE SIMULATOR PRIOR TO RECEIPT OF CONCURRENCE TO DO SO FROM THE LEAD EXAMINER Simulator Operator Role Play Evaluator Notes Plant Response: Before level reaches -5.5 feet in the torus a reactor scram is required. When torus water level teaches -5.5 feet emergency depressurization is required. The crew can anticipate emergency depressurization. Objectives: SRO - Direct ED or Anticipate ED based on torus water level. RO Perform ED or Anticipate ED. Success Path: Reactor depressurized. Scenario Termination: When all rods are inserted and RPV pressure is less than 100 psig, the scenario may be terminated. Remind students not to erase any charts and not to discuss the scenario until told to do so by the evaluatorlinstructor.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev. 0 Page 51 ofOl EVENT 8: SCRAM 1 EMERGENCY DEPRESSURIZATION lime Pos EXPECTED Operator Response Comments Direct a reactor scram before torus level CRS reaches -5.5 feet. CRITICAL TASK #1 Direct OEOP-O 1-SEP-15, Anticipate Emergency Depressurization. CRITICAL TASK #2 OR Direct Emergency Depressurization When directed to scram performs scram immediate actions (see page 52) CRITICAL TASK #1 ATC Performs Scram Hard Card (see page 53) Reports all rods in. Maintains reactor pressure as determined by the BOP CRS. Maintains level as directed by the CRS. May align condensate and feedwater lAW hard card. (See Enclosure 2 page 57) If directed performs OEOP-O1-SEP-15, Anticipate Emergency Depressurization. (see CRITICAL TASK #2 page 56) If directed opens 7 ADS valves.

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 I Page 52 of 61 Unit 2 SCRAM Immediate Actions 1 Ensure SCRAM valves OPEN by manual SCRAM or ARI initiation.

2. WHEN steam flow less than 3.0 Mlbht, THEN place reactor mode switch in SHUTDOWN.

3. iF. reactor power below 2% fAPRM doswiscale trip).

THEN trip main turbine.

4. Ensure master RPV level controller setpoint at ÷170 inches

5. IF:

• Two reactor feed pumps running AND
• RPV level above +160 inches AND
• RPV level rising.

THEN trip one.

2016 NRC SCENARIO 5 LOt SIMULATOR EVALUATION GUIDE Rev.0 Page 53of 61 SCRAM Card Enter applicable leg. 0 Scram ATWS All Control Rods FULL-IN 0 Indications of HydrauliciElectncal ATWS 0 RPV Water Level 0 Ensure ARI initiated C inches Reactor Power C RPV Pressure 0 psig Communicate AWJS report Communicate scram report t0CRS 0 toCRS 0 jf enabled, Place SULCV in service 0 THEN initiate a recirc pump manual wnback 0 Insert Nuclear Instrumentation 0 IF reactor power above 2% OR Ensure Turbine Oil System CANNOT be detemiined, Operating 0 THEN trip both tecirc pumps 0 Ensure Reactor Recirculation Report reactor power to CR5 0 Pump speed at 34% 0 Exit scram card and perform Ensure Heater Drain Pumps EOP-01-LEP-02 0 tripped 0 Exit scram card 0 8 11805 21906 51807

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 54 of 61 ANTICIPATE EMERGENCY DEPRESSURIZATION OEOP-O I-SEP-iS Rev. 0 Page 4 of 5 1.0 ENTRY CONDITION

• As directed by Emergency Operating Procedures fEOP5) 2.0 INSTRUCTIONS 2.1 Reactor Vessel Depressurization 2.1.1 Manpower Requited
• 1 Reactor C)perator 2.1.2 Special Equipment None 2.1.3 Operator Actions

1. Ensure:

• How path available from RPV to condenser 0 RO
• EHC System in operation 0 RO
• Circulating water in operation 0 RO
• Vacuum System in operation 0 RO
• Turbine Shaft Sealing System in operation 0 RO 2 IF AT ANY TIME Main Steam Line Break indicated by
• A-O6 3-6, Stm Tunnel Hi Temp Sys A 0 RO
• A-06 4-6, Stm Tunnel Hi Temp Sys B 0 RO

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 55 of6l ANTICIPATE EMERGENCY DEPRESSURIZATION OEOP-O I -SEP Rev. 0 Page 5015 2.1.3 Operator Actions (continued)

• AM6 6-7, MSIV PitTBTB Tunnel H Temp CI RO
• A-06 5-6, Mn Strn Line Hi Flow Sys A CI RO
• A-06 6-6, Mn Stm Line Hi Flow Sys B CI RO THEN terminate RPV depressurization CI RO

3. IF AT ANY TIME fuel taiture indicated by

• UA-23 2-6, Main Steam Line Rad H CI RC)
• UA-03 5-2, Process Ott-Gas Rad Hi CI RO
• UA-03 6-4, Process OG Vent Pipe Rad Hi CI RO THEN terminate RPV depressurization CI RO 4 IF AT ANY TIME RPV pressure reduction will result in loss of injection required tot adequate core cooling, THEN terminate RPV depressurization CI RO

5. IF MSIVs CLOSED, THEN equalize pressure and open MSIVs fOP-25) CI RO

6. Unit 2 only: Maintain main steam line flow less than 3xI0 Ibm/hr while performing Step 7 CI RO

7. Rapidly depressurize RPV with Main Turbine Bypass valves irrespective of cooldown tate CI RO 8, Exit this procedure and continue in procedure(s) in effect CI RO

2016 NRC SCENARIO 5 LOI SIMULATOR EVALUATION GUIDE Rev.0 Page 56 of 61 ENCLOSURE 1 Page 1 of 1 CONDUCT OF OPERATIONS AD-OP-ALL-i 000 Rev. 6 Page 90 of 90 ATTACHMENT 8 Page 1 of 1

                         <<  Crow Briof Tomplato>>

C] Announce Crew Brief Begin Brief C] All crew members acknowledge announcement (A Required) C] Update the crew as needed: C] Describe what happened and major actions taken C] Procedures in-progress C] Notifications: Reczip C] Maintenance C] Engineering C] Others (Dispatcher, Station Management, etc.) C] Future Direction arid priorities C] Discuss any contingency plans (A Required) C] Solicit questionsconcerns from each crew member C] ROs Input C] CR5 C] STA C] Are there any alarms unexpected for the plant conditions? C] What is the status of Critical Parameters? (A RequIred) EAL C] Provide EAL and potential escalation criteria C] Restore normal alarm announcement (Yes/No) Plnih Brief C] Announce End of Brief

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 57 of 61 ENCLOSURE 2 Page 1 of 2 Feedwater Level Control Following a Reactor Scram NOTE This attachment is NOT to be used for routine system operation. ENSURE the following:

• FW-V6 AND RV-V8 OR FW-V 18 AND FW-V 1 19 closed
• FW-FV177 closed E
• FW-V 120 dosed fl
• FW control MODE SELECT in 1 ELEM
• SULCV in M (MANUAL) closed E
• 821-F032A AND1OR B21-E032B open El

2. PLACE the t1STR REPT SP1RX LVL CTL in M tMANUAL), THEN:

• ADJUST to 187

3. IF any RFP is running, THEN

a. PLACE REP AfB) RECIRC VLV, control switch to open El

b. PLACE RFPT A(B) SP CTL in M (MANUAL) El

4. IF no REP is running, THEN a PLACE REP A(B) RECIRC VLV, control switch to open

b. ENSURE the following

• RFP Af B) DISCH VLV, EW-V3fV4) open El
• REPT Af B) SP CTL in M (MANUAL) at lower limit El
• REPT Af B) MAN/DECS control switch in MAN El
• Reactor water level is less than +206 inches AND REPT El A&B HIGH LEVEL TRIP reset c DEPRESS REPT A(B) RESET

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 58 of 61 ENCLOSURE 2 Page2of2 Feedwater Level Control Following a Reactor Scram ci. ENSURE REPT A(B) LP AND HP STOP VLVS open

e. ROLL RFPT Af B) to 1000 rpm by depressing REP At B) E START f RAISE REPT A(B) to approximately 2550 rpm using the LOW ERRAISE control switch

g. DEPRESS REPT AfB) DFCS CTRL RESET

5. ENSURE MAN/DECS control switch in DECS 6 RAISE REPT A(B) SP CTL speed until discharge pressure is greater than or equal to 100 psig above reactor pressure

7. ADJUST SULCV to establish desired injection

8. IF desired, THEN PLACE SULCV in A (AUTO)

9. IF needed. THEN THROTTLE RV-V 120 10 IF needed, THEN GO TO 20P-32 Section 8.17 for level control

2016 NRC SCENARIO 5 LOl SIMULATOR EVALUATION GUIDE Rev.0 Page 59 of 61 ATTACHMENT I - Scenario Quantitative Attribute Assessment Category Scenario Content Rev Total Malfunctions 5-8 7 Malfunctions after EOP Entry 1 2 2 Abnormal Events 2-4 2 Major Transients 1-2 2 EOPs Used 1-2 2 EOP Contingency 0-2 1 Run Time 60-90 mm Crew Critical Tasks 2-3 2 Tech Specs 2 2 Instrument / Component 2 OATC Failures before Major 2 BOP Instrument / Component 2 2 Failures after Major Normal Operations 1 1 Reactivity manipulation 1 1

NEUTRON MONITORING SYSTEM OPERAT1NG 20P-U9 PROCEDURE Rev 31 Page 43 of 33 ATTACHMENT 1 Page 1 nfl

                          <<Neutron Monitoring Spiking Troubleshooting Form>>

1. Initiators name Unit Two SRO

2. Check all instruments that ate spiking and the associated Unit:

flUniti fSRMA tRMA flIRME jUnit2 f$Rt.t B ZlRt.B LJRMF E S RM C RM C G S RM 0 I RM 0 1 RM H

3. Thie and date of event Today. previous stiitt

4. What is the duration of the spiking (duration of indMdual spikel? Add additional information below to characterize spiking event.

fl Seconds Minutes fl Hours

1. Ensureall rutjirud tbs vdtsjI to sup[ErtuptILAhty are a pnt!iy dtriientcwi

5. Has a WO or AR been initiated?

If yes. list number(s)._00345765 X Yes No

6. Has a leg entry been made? y

7. Is there any welding occurring in the plant? y E No

8. Are there any personnel under vessel? Yes No

9. Are there any plant evolutions in progress? Yes fl No

10. Is there any electrical switching occurring? Yes No

11. Are any control rods being moved or selected? i Yes No

12. Has there been a recent diango in the mode switch? Yes 3 No

13. Is there any major equipment being started? Yes No

14. Has there been any observed relay chatter? 1 No

15. Is there any refuel bridge movement? Yes No

16. Are the rod interlocks being affected?

X Yes No

17. Completed copy of this aftad-iment sent to engineer - Yes No Note below any additional information that may aid troubleshooting (such as 2 instruments spiking but NOT in the same manner):

Multiple upscale and downscale alarms during startup over a 15 minute period All other IRMs responded normally

LOl SIMULATOR EVALUATION GUIDE Page 61 of 61 ATTACHMENT 2 Shift Turnover Brunswick Unit 2 Plant Status Station Duty Workweek E. Neal B. Craig Manager: Manager: Mode: 2 Rx Power: 2% Gross*/Net MWe*: N/A Plant Risk: Green Current EOOS Risk Assessment is: SFPTimeto 45.7 hrs Days Online: 0 days 200 Deg F: lAW the reactivity plan the OATC is to raise power to 6-10%. A2X sequence at step 161. Turnover: . Permission for continuous withdrawal has been granted for the rods going from 12-48. Protected ADHR / FPC Loop A / Demin Transfer Pump Equipment: All remaining ECCS LP systems IRM A was bypassed due to spiking and the paperwork is being evaluated by the WCC SRO for its return to service. Comments: Core Spray Loop B under clearance, expected return in 4 hours. The BOP operator is to complete Step 6.3.46 of OGP-02, Approach to Criticality and Pressurization of the Reactor.

DUKE ENERGY BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE SIM JPM A 2016 NRC INITIAL EXAM

           -                      - RO/ISRO/USRO LESSON TITLE:      Reset Recirc Pump Runback     Both Recirc Pumps Trip LESSON NUMBER: LOT-SIM-JP-002-A09 REVISION NO:       0 Dan Hulqin                                    8/18/16 PREPARER I DATE Bob Bolin                                     9/06/16 TECHNICAL REVIEWER I DATE Grant Newton Shawn Zander                                  9/06/16 VALIDATOR I DATE 2A %r LINE SUPERVISOR I ATE 9        -

TRAflJING SUPERVISION APPROVAL I DATE LOT-SIM-JP-002-A09 Page 1 of 10 Rev. 0

Reset Recirc Pump Runback Both Recirc Pumps Trip RELATED TASKS: 202201 B401 Recover from a Reactor Recirculation Pump Runback Per QP-02 K/A REFERENCE AND IMPORTANCE RATING: 202002 A2.01 3.4/3.4 Ability to predict the impacts of recirculation pump trip and based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations.

REFERENCES:

20P-02, Section 6.3.3, Recovery From Reactor Recirculation Pump Runback 20P-02, Section 6.1.3, Raising Speed Using Individual Recirculation Pump Control 2AOP-04.0, Low Core Flow TOOLS AND EQUIPMENT: None SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): 1 - Reactivity Control (Recirculation Flow Control System) LOT-SIM-JP-002-A09 Page 2 of 10 Rev. 0

Reset Recirc Pump Runback Both Recirc Pumps Trip SETUP INSTRUCTIONS SIMULATOR SETUP Initial Conditions

1. Recommended Initial Conditions IC-b

2. Required Plant Conditions Recirculation Pump A at approximately 70% flow, Recirculation Pump B at Limiter #1.

OPRM Trip Enabled annunciator in ALARM. Malfunctions: Insert the following malfunctions: MaIf ID Current Target Mult ID Description Rmptime Actime Deactime Trig Value Value VFD B RUNBACK RCO24F VFD B False True

                           #1 ACTUATES EEO26F               Loss of 41 60V Bus B       False     True Set Trigger 1, Q2722RSM, VFD B Raise Medium pushbutton to TRUE.

Special Instructions Initiate VFD B runback and allow plant conditions to stabilize. Once runback is complete, delete malfunction RCO24F. If necessay, insert control rods toget below the MELLL Line on the Power-Flow map. Verify VFD B RUNBACK #1 ACTUATES is clear. Reset Speed Hold, if amber light is illuminated. LOT-SIM-]p-002-A09 Page 3 of 10 Rev. 0

Reset Recirc Pump Runback Both Recirc Pumps Trip SAFETY CONSIDERATIONS:

1. None EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM will be performed in the simulator on Unit Two.

Read the following to the ]PM performer. TASK CONDITIONS:

1. Reactor Recirculation Pump operation was previously in accordance with 20P-02, Section 6.1.2.

2. Recirculation Pump 2B has run back to limiter number 1, and the cause has been corrected.

3. A reactivity management briefing is complete, and your reactivity management team is available in the Control Room

4. Another operator is monitoring Nuclear Instrumentation.

INITIATING CUE: You are directed by the Unit CRS to reset the Recirculation Pump runback signal and raise flow of Reactor Recirculation Pump 2B to match flow of Recirculation Pump 2A. LOT-SIM-JP-002-A09 Page 4 of 10 Rev. 0

Reset Recirc Pump Runback Both Recirc Pumps Trip PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. Step 0 - May perform take a minute at job site prior to beginning task. Examinee may cover the following questions, as deemed necessary. What are the hazards in the area? What PPE is required? Am I using appropriate gloves? Tools/PPE inspected prior to use? Energy sources secured/isolated? Is Clearance/Tag Out sufficient? Whats the worst that can happen? Any ALARA concerns? Will I affect plant status? HU Tools needed? Step 1 - Obtain current revision of 20P-02 Sections 6.3.3 may also get Section 6.1.3. Provide current revision of 20P-02 Section 6.3.3, and if asked Section 6.7.3. SAT/U NSAT TIME START: Step 2 Verify the conditions that caused the runback have cleared, or Recirc Pump speed has been lowered below the runback setpoint. Verifies Recirc Pump speed below the runback setpoint, and condition is clear as part of Task Conditions. SAT/UNSAT Step 3 ENSURE RECIRC PUMP B SPEED DEMAND signal is approximately the same as the following:

• RECIRC PUMP B CALCULATED SPEED
• RECIRC PUMP B ACTUAL SPEED Ensures Calculated Speed and Actual Speed approximately the same.

SAT/U NSAT Step 4 - RESET the Recirc Pump runback for Reactor Recirculation Pump B as follows:

a. DEPRESS Recirc VFD B RUNBA CK RESET push button.

Runback Rest push button is depressed.

                                                              **CRITICAL STEP**            SAT/UNSAT LOT-SIM-JP-002-A09                          Page 5 of 10                                           Rev. 0

Reset Recirc Pump Runback Both Recirc Pumps Trip

b. CONFIRM yellow AUTOMATIC RUNBACK light extinguished.

Yellow Automatic Runback light is confirmed extinguished SAT/UNSAT

c. CONFIRM annunciator RECIRC FLOW B LIMIT (A-07 2-4) is clear.

Annunciator confirmed clear. SAT/UNSAT PROMPT: If asked which VFD RAISE pushbutton to use, or if the VFD RAISE SLOW pushbutton is depressed, direct Examinee as the CRS to raise speed of the Recirculation Pump using the VFD RAISE MEDIUM pushbutton. Step 5 ADJUST flow as directed by the Unit CRS. VFD RAISE MEDIUM push button is depressed in accordance with 20P-02, Section 6. 7.3, pages 38-40. CRITICAL STEP ** SAT/UNSAT INOTE: 4160 V Bus B will trip when the VFD RAISE MEDIUM push button is depressed. I PROMPT: If asked, as CR5, direct Examinee to enter and announce the appropriate AOP. I Step 6 Determines that no Recirc Pumps are running. Diagnosis failure of Bus B which causes both Recirc Pumps to lose power. SAT/U NSAT Step 7: May enter and announce entry into 2AOP-04.0, Low Core Flow 2A OP-04. 0 announced and entered. SAT/UNSAT LOT-SIM-JP-002-A09 Page 6 of 10 Rev. 0

Reset Recirc Pump Runback Both Recirc Pumps Trip Step 8 Inserts a manual reactor scram lAW 2AOP-04.0, Immediate Operator Action, depresses both RPS Channel Manual Pushbuttons. Performs the following scram immediate actions:

1. Ensure SCRAM valves OPEN by manual SCRAM orARl initiation.

2. WHEN steam flow less than 3.0 Mlb/hr, THEN place reactor mode switch in SHUTDOWN.

3. IF reactor power below 2% (A PAM downscale trip),

THEN trip main turbine.

4. Ensure master RPV level controller setpoint at + 770 inches.

5. IF two reactor feed pumps running and RPV level is greater than 760 inches and rising, THEN trip one reactor feed pump.

CRITICAL STEP ** SAT/UNSAT Step 9 Informs CRS that all rods are in, RPV water level and RPV pressure. Acknowledge scram report as the CRS. SAT/UNSAT TERMINATING CUE: Once a manual reactor scram is inserted and scram immediate actions are complete, the JPM can be terminated. TIME COMPLETED: COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. Step Critical I Not Critical Reason 1 Not Critical Administrative 2 Not Critical Given as initial conditions 3 Not Critical Observation of indications 4a Critical Action required to reset the runback 4b Not Critical Observation of indications 4c Not Critical Observation of indications 5 Critical Action required to meet Initiating Cue 6 Not Critical Observation of indications 7 Not Critical Observation of indications 8 Critical Immediate Operator action from 2AOP-04.0 9 Not Critical Communication LOT-SIM-JP-002-A09 Page 7 of 10 Rev. 0

Reset Recirc Pump Runback Both Recirc Pumps Trip REVISION

SUMMARY

0 New JPM written for 2016 Initial NRC exam. LOT-SIM-JP-002-A09 Page 8 of 10 Rev. 0

Reset Recirc Pump Runback Both Recirc Pumps Trip Validation Time: 20 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: 2 Setting: In-Plant Simulator X Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes X No EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: U Comments reviewed with Performer Evaluator Signature: Date: LOT-SIM-JP-002-A09 Page 9 of 10 Rev. 0

TASK CONDITIONS:

1. Reactor Recirculation Pump operation was previously in accordance with 20P-02, Section 6.1.2.

2. Recirculation Pump 2B has run back to limiter number 1, and the cause has been corrected.

3. A reactivity management briefing is complete, and your reactivity management team is available in the Control Room

4. Another operator is monitoring Nuclear Instrumentation.

INITIATING CUE: You are directed by the Unit CRS to reset the Recirculation Pump runback signal and raise flow of Reactor Recirculation Pump 2B to match flow of Recirculation Pump 2A. Page 10 of 10

REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev. 168 Page 38 of 250 6.1.3 Raising SpeedlPower Using Individual Recirculation Pump Control or Recirc Master Control

1. Ensure the following Initial Conditions are met:

a. Reactor Recirculation Pumps in operation in accordance with Section 6.1 .2

b. Recirculation Pump flow limits are CLEAR NOTE

. Recirculation Pump speed changes are performed when directed by 0GP-04, Increasing Turbine Load to Rated Power, and 0GP-12, Power Changes. Other operating procedures are used simultaneously with this procedure as directed by 0GP-04, Increasing Turbine Load to Rated Power, 0GP-12, Power Changes, or the Unit CRS . Speed changes are accomplished by depressing Raise Slow or Raise Medium pushbuttons. The Raise Slow pushbutton changes Recirc pump speed at 0.06%/increment at 1 rpm/second. The Raise Medium pushbutton changes Recirc pump speed at 0.28%/increment at 5 rpm/second CAUTION The OPRM System monitors LPRMs for indication of thermal hydraulic instability (THI). When greater than or equal to 25% power and less than or equal to 60% recirculation flow, alarms and automatic trips are initiated upon detection of THI. Pump operations are governed by the limits of the applicable Power Flow Map, as specified in the COLR. {8.1 .9) D

2. IF AT ANY TIME any of the following conditions exist, THEN enter 2AOP-04.0, Low Core Flow. {8.1 .9)

• Entry into Region A of Power to Flow Map
• OPRM INOPERABLE AND any of the following O Entry into Region B of Power to Flow Map o Entry into 5% Buffer Region of Power to Flow Map o Entry into OPRM Enabled Region and indications of THI (Thermal Hydraulic Instability) exist

REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev. 168 Page 39 of 250 6.1.3 Raising SpeedlPower Using Individual Recirculation Pump Control or Red rc Master Control (continued) CAUTION The OPRM System monitors LPRMs for indication of thermal hydraulic instability (THI). When greater than or equal to 25% power and less than or equal to 60% recirculation flow, alarms and automatic trips are initiated upon detection of THI. Pump operations is be within the limits of the applicable Power-Flow Map, as specified in the COLR. The Scram Avoidance Region is avoided. {8.1.9} With core flow less than 57.5 x 106 lbs/hr, jet pump loop flows are required within 10% (maximum indicated difference 6.0 x 106 lbs/hr). With core flow greater than or equal to 57.5 x 106 lbs/hr, jet pump loop flows are required within 5% (maximum indicated difference 3.0 x 106 lbs/hr) D If total reactor feedwater flow lowers to less than 16.4% of rated flow, Speed Limiter Number I will cause the Recirculation Pumps to run back to 34% speed. This signal must be manually reset in accordance with Section 6.3.3 E1 When total core flow is greater than 43 mlb/hr, Speed Limiter Number 2 will cause a runback to approximately 48% speed if reactor water level is less than 182 inches and either reactor feed pump Aor B suction flow is less than 14.9% of individual RFP rated suction flow. This signal must be manually reset using Section 6.3.3 t1 BEGIN R.M. LEVEL R21R3 REACTIVITY EVOLUTION

3. IF desired to raise the speed of both Recirc Pumps simultaneously, as directed by the Unit CRS, THEN depress Recirc Master Control Raise Slow or Raise Medium pushbutton

4. IF desired to raise the speed of an individual Recirc Pump, as directed by the Unit CRS, THEN depress the VFD A(B) Raise Slow or Raise Medium pushbutton for the Recirc Pump

5. Confirm the following, as applicable:

• A rise in Recirc Pump A(B) Speed Demand, Calculated Speed, and a rise in Actual Speed
• A rise in Reactor power
• A rise in B32-R61 7(R61 3) [Recirc Pump A(B) Discharge Flow]

REACTOR RECIRCULATION SYSTEM OPERATING 2OP-02 PROCEDURE Rev. 168 Page 40 of 250 6.1.3 Raising SpeedlPower Using Individual Recirculation Pump Control or Recirc Master Control (continued)

• A rise in B32-VFD-IDS-003A(B) [Recirc VFD 2A(B) Output Wattmeter]
• A rise in B32-VFD-IDS-OO1A(B) [Recirc VFD 2A(B) Output Frequency Meter]

END R.M. LEVEL R21R3 REACTIVITY EVOLUTION Date/Time Completed Performed By (Print) Initials Reviewed By: Unit CRS/SRO

6.3.3 Recovery From Reactor Recirculation Pump Runback

1. Ensure Reactor Recirculation Pump operation was previously in accordance with Section 6.1.2 NOTE Recirculation pump runback to approximately 48% speed occurs when reactor water level is less than or equal to 182 inches and reactor feed pump A or B suction flow is less than or equal to 14.9% of individual REP rated suction flow. A recirculation pump speed runback to 34% will occur when the recirculation pump discharge valve is NOT fully OPEN or total feedwater flow is less than 16.4% of the rated flow. Both of these conditions will require a manual reset of the runback

2. Ensure the conditions that caused the runback have cleared, or recirc pump speed has been lowered below the runback setpoint

3. Ensure the system operation has stabilized CAUTION The OPRM System monitors LPRMs for indication of thermal hydraulic instability (THI). When greater than or equal to 25% power and less than or equal to 60% recirculation flow, alarms and automatic trips are initiated upon detection of THI. Pump operations are within the limits of the applicable Power-Flow Map, as specified in the COLR. The Scram Avoidance Region is avoided. {8.1.9} U

4. IF AT ANY TIME any of the following conditions exist, THEN enter 2AOP-04.0, Low Core EIow.{8.1 .91

• Entry into Region A of Power to Flow Map
• OPRM INOPERABLE AND any of the following o Entry into Region B of Power to Flow Map O Entry into 5% Buffer Region of Power to Flow Map O Entry into OPRM Enabled Region and indications of THI (Thermal Hydraulic Instability) exist

5. Ensure Recirc Pump A(B) Speed Demand signal is approximately the same as the following:

• Recirc Pump A(B) Calculated Speed
• Recirc Pump A(B) Actual Speed

REACTOR RECIRCULATION SYSTEM OPERATING 20P-02 PROCEDURE Rev. 168 Page 65 of 250 6.3.3 Recovery From Reactor Recirculation Pump Runback (continued)

6. Reset the Recirc Pump runback for Reactor Recirculation Pump A(S) as follows:

a. Depress Recirc VFD A(B) Runback Reset pushbutton

b. Confirm yellow Automatic Runback light is OFF

c. Confirm 2-A-06, 3-2 (2-A-07, 2-4), Recirc Flow A(S) Limit, is CLEAR BEGIN R.M. LEVEL R21R3 REACTIVITY EVOLUTION

7. Adjust flow as directed by the Unit CRS END R.M. LEVEL R2!R3 REACTIVITY EVOLUTION Date/Time Completed Performed By (Print) Initials Reviewed By:

Unit CRS/SRO

DUKE ENERGY BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE SIM JPM B 2016 NRC INITIAL EXAM

           -                    - RO/ISRO/USRO LESSON TITLE:       MECHANICAL TRIP VALVE OIL TRIP TEST LESSON NUMBER: LOT-SIM-JP-026.2-01 REVISION NO:        0 Dan HuIqin                                 8/18/16 PREPARER I DATE Bob Bolin                                  9/06/16 TECHNICAL REVIEWER I DATE Kyle Cooper Shawn Zander                                9/06/16 VALIDATOR I DATE LINE S PERVISOR I   ATE TRAINING SUPERVISION APPROVAL I DATE LOT-SIM-JP-026.2-01                Page 1 of 9          Rev. 0

Mechanical Trip Valve Oil Trip Test RELATED TASKS: 245202B1 01 Perform Mechanical Trip Valve Oil Trip Test Per OP-26 K/A REFERENCE AND IMPORTANCE RATING: 245000 A3.01 Ability to manually operate and/or monitor in the control room: Turbine Trip

REFERENCES:

20P-26, Section 6.3.8, Mechanical Trip Valve Oil Trip Test TOOLS AND EQUIPMENT: None SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): 4 - Heat Removal (main Turbine Generator and Auxiliary Systems) LOT-SIM-JP-026.2-01 Page 2 of 9 Rev. 0

Mechanical Trip Valve Oil Trip Test SETUP INSTRUCTIONS SIMULATOR SETUP Initial Conditions

1. Recommended Initial Conditions IC-li

2. Required Plant Conditions Turbine is at 1800 rpm.

Malfunctions: None Special Instructions None. LOT-SIM-JP-026.2-01 Page 3 of 9 Rev. 0

Mechanical Trip Valve Oil Trip Test SAFETY CONSIDERATIONS:

1. None EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM will be performed in the simulator on Unit Two.

Read the following to the JPM performer. TASK CONDITIONS:

1. All applicable prerequisites listed in Section 5 of 20P-26 are met.

2. Last performance of 20P-26 Section 6.3.15 was successful.

INITIATING CUE: You are directed by the Unit CRS to perform 20P-26, Section 6.3.8, Mechanical Trip Valve Oil Trip Test. LOT-51M-Jp-026.2..01 Page 4 of 9 Rev. 0

Mechanical Trip Valve Oil Trip Test PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. Step 0 May perform take a minute at job site prior to beginning task. Examinee may cover the following questions, as deemed necessary. What are the hazards in the area? What PPE is required? Am I using appropriate gloves? Tools/PPE inspected prior to use? Energy sources secured/isolated? Is Clearance/Tag Out sufficient? What the worst that can happen? Any ALARA concerns? Will I affect plant status? HU Tools needed? Step 1 - Obtain current revision of 20P-26 Sections 6.3.8. Provide current revision of 20P-26 Section 6.3.8. SAT/UNSAT TIME START: Step 2 - Depress the Locked out pushbutton. Depresses the Locked Out pushbutton on the XU- 7 panel.

                                                               **CRITICAL STEP**           SAT/UNSAT Step 3    Confirms Locked Out light illuminates and UA-23, 3-3, Overspeed Trip Locked annunciator is On Confirms Locked Out white light illuminates and UA-23, 3-3, Overspeed Trip Locked annunciator is acknowledged and reported to Unit CRS.

SAT/U NSAT Step 4 Depress and hold the Oil Trip Pushbutton until Tripped light comes On, then release the Oil Trip pushbutton. Oil Trip pushbutton is depressed and held until the Tripped light is illuminated.

                                                              **CRITICAL STEP**            SAT/UNSAT LOT-SIM-JP-026.2-01                          Page 5 of 9                                           Rev. 0

Mechanical Trip Valve Oil Trip Test Step 5 - Depress and hold the Push to Reset pushbutton and confirm the Resetting light comes on. Push to Reset pushbutton is depressed and the Resetting light illuminates. CRITICAL STEP ** SAT/UNSAT Step 6 When 5 seconds have elapsed confirm the Reset light comes On and UA-23, 4-3 Turbine Overspeed Trip Reset, annunciator is received. Confirms the Reset light illuminates and Turbine Overspeed Trip Reset annunciator is acknowledged and reported to Unit CRS. SAT/U NSAT Step 7 When the Reset light comes On then release the Push to Reset pushbutton. The Push to Reset pushbutton is released and annunciator UA-23, 4-3, Turbine Overspeed Trip Reset annunciator is reset and reported as cleared to the Unit CRS. CRITICAL STEP ** SAT/UNSAT Step 8- When 10 seconds have elapsed then depress the Normal pushbutton. The Normal pushbutton is depressed and the Normal light is confirmed On and the Locked Out Light extinguishes. Annunciator UA-23, 3-3, Overspeed Trip Locked annunciator is reset and reported as cleared to the Unit CRS. CRITICAL STEP SAT/UNSAT Step 9 Informs CRS Mechanical Trip Valve Oil Trip is complete. Acknowledge report as the CRS. SAT/UNSAT 1 TERMINATING CUE: All actions in 20P-26, Section 6.3.8 have been completed. I TIME COMPLETED: COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. LOT-SIM-JP-026.2-01 Page 6 of 9 Rev. 0

Mechanical Trip Valve Oil Trip Test Step Critical I Not Critical Reason 1 Not Critical Administrative 2 Critical Action required to complete the procedure without tripping the main turbine. 3 Not Critical Observation of indications as a result of the previous step 4 Critical Action required to complete the procedure 5 Critical Action required to complete the procedure 6 Not Critical Observation of indications as a result of the previous step 7 Critical If released prior to the Reset light illuminating the trip will trip. 8 Critical Required to restore the system to normal alignment and remove the overspeed trip locked out. 9 Not Critical Communication REVISION

SUMMARY

0 New JPM written for 2016 Initial NRC exam. LOT-SIM-JP-026.2-01 Page 7 of 9 Rev. 0

Mechanical Trip Valve Oil Trip Test Validation Time: 10 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: 2 Setting: In-Plant Simulator X Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: D Comments reviewed with Performer Evaluator Signature: Date: LOT-SIM-JP-026.2-01 Page 8 of 9 Rev. 0

TASK CONDITIONS:

1. All applicable prerequisites listed in Section 5 of 20P-26 are met.

2. Last performance of 20P-26 Section 6.3.15 was successful.

INITIATING CUE: You are directed by the Unit CRS to perform 20P-26, Section 6.3.8, Mechanical Trip Valve Oil Trip Test. Page 9 of 9

TURBINE SYSTEM OPERATING PROCEDURE 20P-26 Rev. 130 Page 93 of 168 6.3.8 Mechanical Trip Valve Oil Trip Test

1. Confirm the following initial conditions are met:

              . All applicable prerequisites listed in Section 5.0 are met
             .      Turbine is at 1800 rpm, on-line or off-line
             .       IF last performance of Section 6.3.15 was unsuccessful, THEN Unit CRS permission is obtained to perform this test NOTE During performance of this test, annunciators UA-23, 3-3, Overspeed Trip Locked, and UA-23, 4-3, Turbine Overspeed Trip Reset, are expected alarms                       D

2. IF AT ANY TIME during the performance of this test the expected indications are NOT observed, THEN perform the following

a. Immediately notify the System Engineer Person Notified

b. Reference EC-293249 for expected position of linkages

3. Depress the Locked Out pushbutton

4. Confirm the following:

a. The Locked Out light comes ON

b. UA-23, 3-3, Overspeed Trip Locked, annunciator is ON

5. Depress and hold the Oil Trip pushbutton untit the Tripped light comes ON, then release the Oil Trip pushbutton

TURBINE SYSTEM OPERATING PROCEDURE 2OP-26 Rev. 130 Page 94 of 168 6.3.8 Mechanical Trip Valve Oil Trip Test (continued) NOTE

• Steps that require holding a pushbutton IN and then confirming required actions occur may be performed and then signed off after completion of the confirmation steps EC-293249 (NCR-741915, WO-1351 1606) document the Resetting Light and UA-23, 4-3, Turbine Overspeed Trip Reset, may NOT respond as indicated in Step 6.a and Step 6.c. These are indications only and do NOT impact the ability to test and reset the turbine trip logic. Therefore these steps may be marked NA D

6. Depress and hold the Push To Reset pushbutton and confirm the followingS

a. The Resetting light comes ON

b. WHEN approximately 5 seconds have elapsed, THEN the Reset light comes ON

c. UA-23, 4-3, Turbine Overspeed Trip Reset, annunciator is ON

7. WHEN the Reset light comes ON, THEN release the Push To Reset pushbutton and confirm the UA-23, 4-3, Turbine Overspeed Trip Reset, annunciator is CLEAR

8. WHEN at least 10 seconds have elapsed, THEN depress the Normal pushbutton and confirm the following

a. The Normal light comes ON and the Locked Out light goes OFF

b. The UA-23, 3-3, Overspeed Trip Locked, annunciator CLEARS

TURBINE SYSTEM OPERATING PROCEDURE 20P-26 Rev. 130 Page 95 of 168 6.3.8 Mechanical Trip Valve Oil Trip Test (continued) Date/Time Completed Performed By (Print) Initials Reviewed By Unit CRS/SRO

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE SIM JPM C -2076 NRC INITIAL EXAM - RO/ISRO/USRO LESSON TITLE: RCIC Start Steam Line Ruptures and RCIC Fails to Isolate LESSON NUMBER: LOT-SIM-JP-01 6-A05 REVISION NO: 04

           %oc So1ez                             9/tO/2015 PREPARER/DATE
                 &9%2 TECHNICAL REVIEWER I DATE Vewá Pcc%ez                           9/tO/Ot5 VALIDATOR I DATE 9e dezce LINE SUPERVISOR I DATE Q

TRAINING SUPERVISION APPROVAL I DATE LOT-SI M-JP-01 6-A05 Page 1 of 9 Rev 5

RELATED TASKS: 21 7003B1 01, Manually Startup The RCIC System Per OP-i 6 K/A REFERENCE AND IMPORTANCE RATING: 217000A4.08 3.7 3.6 Ability to manually operate and/or monitor RCIC system flow

REFERENCES:

S/969 (RCIC Hard Card) OP-16, Section 6.1.3 TOOLS AND EQUIPMENT: None. SAFETY FUNCTION (from NUREG 1123, Rev 2.): 2 - Inventory Control LOT-SIM-JP-016-A05 Page 2019 Rev 5

SIMULATOR SETUP Recommended Initial Conditions Any 100% IC Required Plant Conditions:

• RPV level <166 inches
• Inhibit ADS
• Place HPCI in PTL
• Trip REPs Triggers:

Auto: Q1619RRM, E51-F013 Red Light Equal to TRUE. Malfunctions: Event System Tag Title Value! Activate Deactivate Ramp Time (sec) Time (sec) Rate N/A ES ESO55F E51 -F007, Failure to Auto Close N/A N/A N/A N/A ES ESO56F E51-F008, Failure to Auto Close N/A N/A N/A 1 ES ESO25F RCIC Stm Brk S RHR Room 20%! 0 40 sec Trigger 1 sec. N/A ES ESO41 F RCIC Failure to Auto Start N/A N/A N/A Overrides: None Remotes None LOT-SIM-JP-016-A05 Page 3 of 9 Rev 5

SAFETY CONSIDERATIONS: None EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM may be performed on Unit 2.

4. Critical Step Basis a) Prevents Task Completion b) May Result in Equipment Damage c) Affects Public Health and Safety d) Could Result in Personal Injury Read the following to the performer.

TASK CONDITIONS:

1. Both Reactor Feed Pumps have tripped and are not available.

2. Reactor level is below 166 inches.

3. HPCI is not available.

INITIATING CUE: You are directed by the Unit CRS to place RCIC in service per the Hard Card and restore Reactor level to 166 to 206 inches. Notify the Unit CRS when all required actions are complete. LOT-SIM-JP-016-A05 Page 4 of 9 Rev 5

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. Step 1 - Perform take a minute at job site prior to beginning task. Examinee should cover the following questions, as deemed necessary. What are the hazards in the area? What PPE is required? Tools/PPE inspected prior to use? Energy sources secured/isolated? Is Clearance/rag Out sufficient? Whats the worst that can happen? Any ALARA concerns? Will I affect plant status? HU Tools needed? SAT/UNSAT TIME START: Step 2 - Ensure the following valves are open: Turbine Trip & Throttle Valve, E51-V8, and Turbine Trip & Throttle Valve Actuator, E51 -V8, and Turbine Governor Valve, E51 -V9. E57-V8 (valve position) E51-V8 (actuator position) and E57-V9 are open. SAT/UNSAT Step 3 Open Cooling Water Supply Valve, E51 -F046. E57-F046 is full open. CRITICAL STEP ** SAT/UNSAT Step 4 Start Vacuum Pump and leave switch in START. Vacuum Pump running with switch in Start. SAT/U NSAT Step 5 - Open Turbine Steam Supply Valve, E51-F045. E57-F045 is full open. CRITICAL STEP ** SAT/UNSAT Step 6 Open RCIC Injection Valve, E51 -FOl 3. E57-F073 is full open. CRITICAL STEP ** SAT/UNSAT LOT-SIM-JP-016-A05 Page 5 of 9 Rev 5

Step 7 Ensure that the RCIC turbine starts and comes up to speed as directed by RCIC FLOW CONTROL. RCIC Turbine speed observed to come up to speed. SAT/U NSAT Step 7a Raises flow to 500 gpm. Raises RCIC FLOW CONTROLLER to500 gpm. SAT/UNSAT NOTE: When Reactor water level has started to rise and when directed by the evaluator activate Trigger 1 to initiate steam line break. Step 8 Recognize the RCIC isolation and trip signal. RCIC isolation and trip is recognized. SAT/UNSAT Step 9Recognize the failure of the RCIC Steam Supply Valves, E51-F007 and E51-F008, to close. Failure of E57-F007 and E51-F008 to close is recognized. Operator refers to 2APP-A-03 (5-2 and 6-2). SAT/UNSAT Step 10 Manually close RCIC Steam Supply Valve, E51 -F007 OR RCIC Steam Supply Valve, E51-F008 OR both. E57-F007 or E57-F008 or both are closed. CRITICAL STEP ** SAT/UNSAT Step 11 Notify the Unit CRS that the RCIC Steam Pipe has ruptured and that E51 -F007 and/or E51 -F008 were manually closed to isolate the leak. Unit CRS is notified SAT/U NSAT TERMINATING CUE: When the RCIC Steam Line rupture is isolated and the Unit CRS is notified, this JPM is complete. Time Completed: NOTE: Comments required for any step evaluated as UNSAT. LOT-SIM-JP-016-A05 Page 6 of 9 Rev 5

COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. Step Critical I Not Critical Reason 1 Not Critical Administrative 2 Not Critical Not required to complete task. 3 Critical Pump will be damaged without cooling water. 4 Not Critical Not required to complete task. 5-6 Critical Required to complete task. 7-9 Not Critical Ensure and Recognize steps. 10 Critical Actions required to complete task. 11 Not Critical Informing CRS of results. REVISION

SUMMARY

5 ALL NON-TECHNICAL CHANGES:

                   . Corrected revision numbers in revision summary.

• Corrected OP-16 section in references from 5.3 to 6.1.3.
• Corrected year of signature from 205 to 2015 4 New JPM format.

Added Critical/Non Critical step explanation. 3 NewJPM format. LOT-SIM-]P-016-A05 Page 7 of 9 Rev 5

Validation Time: 15 Minutes (approximate). Time Taken: Minutes APPUCABLE METHOD OF TESTING Performance: Simulate X Actual X Unit: 2 Setting: In-Plant Simulator X Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes X No EVALUATION Pe rio rm e r: JPM: Pass Fail Remedial Training Required: Yes No Comments: O Comments reviewed with Performer Evaluator Signature: Date: LOT-SIM-JP-016-A05 Page 8019 Rev 5

Read the following to the performer. TASK CONDITIONS:

1. Both Reactor Feed Pumps have tripped and are not available.

2. Reactor level is below 166 inches.

3. HPCI is not available.

INITIATING CUE: You are directed by the Unit CRS to place RCIC in service per the Hard Card and restore Reactor level to 166 to 206 inches. Notify the Unit CRS when all required actions are complete. Page 9 of 9

REACTOR CORE ISOLATION COOLING SYSTEM 20P-16 OPERATING PROCEDURE Rev. 120 Page 97 of 99 ATTACHMENT 7 Page 1 of 1

                           <<RCIC Instructional Aid for EOPs>>

MANUAL RCIC INJECTION (20P-16 Section 5.3)

1. ENSURE THE FOLLOWING VALVES ARE OPEN: E51-V8 (VALVE POSITION), E51-V8 (ACTUATOR POSITION), AND E51-V9. C

2. OPEN E51-F046 C

3. START VACUUM PUMP AND LEAVE SWITCH IN START. C

4. OPEN E51-F045 C

5. OPEN E51-F013 C

6. ENSURE RCIC TURBINE STARTS AND COMES UP TO SPEED AS DIRECTED BY RCIC FLOW CONTROL C

7. ADJUST RCIC FLOW CONTROLLER TO OBTAIN DESIRED FLOW RATE. C

8. ENSURE E51-F019 IS CLOSED WITH FLOW GREATER THAN 80 GPM. C

9. ENSURE THE FOLLOWING VALVES ARE CLOSED: E51-F025, E51-F026, E51 F004, AND E51-F005 C

10. START SBGT (lOP-i 0)

11. ENSURE BAROMETRIC CNDSR CONDENSATE PUMP OPERATES RCIC PRESSURE CONTROL (2OP-16 SECTION 82)

1. ENSURE THE FOLLOWING VALVES ARE OPEN: E51-V8 (VALVE POSITION), E51-V8 (ACTUATOR POSITION), AND E51-V9. C

2. OPEN E51-F046 C

3. START VACUUM PUMP AND LEAVE SWITCH IN START. C

4. ENSURE E51-F013 IS CLOSED C

5. ENSURE E41-F011 IS OPEN C

6. THROTTLE OPEN E51-F022 UNTIL DUAL INDICATION IS OBTAINED C

7. OPEN E51-F045 C

8. THROTTLE OPEN E51-F022 OR ADJUST RCIC FLOW CONTROL, E51 FIC R600, TO OBTAIN DESIRED SYSTEM PARAMETERS AND REACTOR PRESSURE.

9. ENSURE E51-F019 IS CLOSED WITH FLOW GREATER THAN 80 GPM.

10. ENSURE THE FOLLOWING VALVES ARE CLOSED: E51-F025, E51-F026, E51 F004, AND E51-F005. C

11. START SBGT (20-10) C

12. ENSURE BAROMETRIC CNDSR CONDENSATE PUMP OPERATES C FOR SHUTDOWN: REFER TO 20P-16 FOR TRANSFER BETWEEN PRESSURE AND LEVEL CONTROL: REFER TO 2OP-16 2 21968 S1969

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE SIM JPM D 2016 NRC INITIAL EXAM RO!ISRO LESSON TITLE: SUPPRESSION POOL COOLING PER HARD CARD SW RELEASE LESSON NUMBER: LOT-SIM-JP-017-A12 REVISION NO: 2 Dan HuIqin 8/18/16 PREPARER I DATE Bob Bolin 9/06/16 TECHNICAL REVIEWER I DATE Shawn Zander 9/06/16 VALIDATOR I DATE (<%91ç (%/%- LINE SLJPERVISOR IATE TRAINING SUPERVISION APPROVAL I DATE LOT-SIM-JP-017-A12 Page 1 of 10 Rev.2

RELATED TASKS: 20501 4B1 01 Start Up RHR In Suppression Pool Cooling Mode Per OP-17 K/A REFERENCE AND IMPORTANCE RATING: 219000 A4.01 Ability to manually operate and/or monitor in the control room: Pumps

REFERENCES:

Hard Card- Emergency Suppression Pool Cooling Using Loop B (20P-17) S/i 064 TOOLS AND EQUIPMENT: None SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): 5 Containment Integrity SAFETY CONSIDERATIONS None LOT-SIM-JP-017-Ai2 Page 2 of 10 Rev.2

SETUP INSTRUCTIONS Recommended Initial Conditions IC-li, 100% Power, BOO Required Plant Conditions

1. Activate the fuel failure and Stuck SRV.

2. When MSL Rad Hi-Hi is in alarm and Suppression Pool Temperature is >950 F, scram the reactor, perform scram immediate actions, close the MSIVs, and allow plant conditions to stabilize.

3. Place RHR Loop A in Suppression Pool Cooling, open CSW to vital header.

4. Start both NSW pumps.

5. Place a red cap on RHR and RHR SW Pump 2D.

Triggers Trigger 1 - K17O8JCN, Eii-FO48Bto close. Malfunctions CWO1 3F, RHR B HX Tube Leak to 100% on trigger 1. ELIALSRB2B, Mechanical Trip RHRSW Booster Pump 2D to Trip on trigger 1 Overrides ZUA355 Service WTR Effluent Rad High to ON with 10 sec TD on trigger 1 Remotes RS_IARHBYPB, El i-F068B Auto Close Bypass Switch, to BYPASS. ED_IABKCJ16, Bkr Ctl DC Fuses RHR SW 2D OUT ED_IABKCFO3, Bkr OtI DC Fuses RHR 2D OUT Special Instructions None LOT-SIM-JP-0i7-A12 Page 3 of 10 Rev.2

SAFETY CONSIDERATIONS:

1. None.

EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG 1021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM will be performed in the simulator on Unit Two.

Read the following to the ]PM performer. TASK CONDITIONS:

1. A Resin Injection has occurred, currently raising power to burn off the resin.

2. RWCU has been isolated.

3. SRV C was not properly seated, subsequently it has been closed.

4. Torus temperature is >95 deg. F.

5. The A Loop of RHR has been started in Suppression Pool Cooling mode using CSW.

6. 2D RHR and 2D RHR SW Pumps are under clearance.

INITIATING CUE: You are directed to place the B Loop of RHR in Suppression Pool Cooling in accordance with the hard card using NSW and inform the Unit CRS when the required actions are complete. LOT-SIM-JP-017-A12 Page 4 of 10 Rev.2

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. Step 1 - Obtain Hard Card for Emergency Suppression Pool Cooling Using Loop B. Hard Card obtained. SAT/U NSAT TIME START: Step 2Open SW-V105, Nuc Sw Supply Vlv. Places control switch for the SW-V705 to OPEN and verifies green light out/red light on. SW- Vi 05 is open * (critical*).

                                                                 **CRITICAL STEP**SAT/UNSAT Step 3  Close SW-V143, Well Water Supply Vlv.

Verifies control switch for the SW-V143 to CLOSE and verifies green light on/red light off SW-V743 is closed. SAT/U NSAT NOTE: The performer should recognize that a LOCA signal does not exist and N/A the step for using the manual override switch. RHR SW Pump 2D is under clearance and is not available. Step4StartRHRSWPMP. Places control switch for RHR SW Booster Pump 2B to START. Verifies green light off/red light on, and verifies RHR SW Booster Pump 2B discharge pressure is rising on SW-PI- 1155-i. RHR SW Booster Pump 2B is running* (critical*)

                                                                 **CRITICAL STEP**SAT/UNSAT Step 5  Adjust El 1 -F068B, HX 2B Sw Discharge VIv.

Throttles control switch for El 7-F068B to obtain a RHR SW flow of 2000-4000 gpm on El 1-Fl-R602B.

                                                                 **CRITICAL STEP**SAT/UNSAT LOT-SIM-JP-017-A12                          Page 5 of 10                                           Rev.2

NOTE: SW-V1 17, NSW to Vital Header, should not be opened since CSW is already supplying the vital header. Opening SW-V1 17 would cross-tie the CSW and NSW headers. Although an analysis has been performed for this alignment, it is not preferred. Step 6 Establish CLG WTR TO VITAL HDR. Verifies that SW-Vl 17, Conventional SW to Vital Header Vlv, is supplying the vital header on A Loop (Red light on/Green Light off). SAT/UNSAT NOTE: The performer should recognize that a LOCA signal does not exist and that the FOl 5B is closed and N/A these two steps on the hard card. Step 7Start LOOP B RHR PMP. Places control switch for RHR Pump 28 to START. Verifies red light on/green light off Verifies RHR system pressure rises. Verifies RHR Pump pressure rises. 28 RHR Pump running*. (critical*)

                                                            **CRITICAL STEP**SATIUNSAT Step 8Open El 1-F028B, Torus Discharge Isol Vlv.

Places control switch for El l-F0288 to OPEN. Verifies red light on/green light off.

                                                            **CRITICAL STEP**SAT/UNSAT Step 9Throttle E11-F024B, Torus Cooling lsol Vlv.

Throttles control switch for El 7-F024B to obtain 6,000 to 70,000 gpm for 7 RHR pump in operation.

                                                            **CRITICAL STEP**SAT/UNSAT LOT-SIM-JP-017-A12                       Page 6 of 10                                        Rev.2

NOTE: ALTERNATE PATH BEGINS AT STEP 10 and 11: When the performer begins to throttle the F048B the RHR HX tubes will rupture resulting initially in a RHR Hi Conductivity alarm followed by a trip of the RHR SW Pump and the F068B will fail to close resulting in high service water radiation alarm. There are several ways to terminate the release, closing the F068B, stopping the running RHR pumps, or closing F047B and F003B. The critical action is to terminate the release. F068B should be closed as this is an automatic action that tailed to occur. Step 10 Throttle Eli -F048B, HX 2B Byp Vlv. Throttles control switch in CLOSE for El l-F048B.

                                                                 **CRITICAL STEP**SAT/UNSAT Step 11  Terminates the Service Water radiation release.

Closes the F068B and/or stops the running RHR pump(s) and/or closes F047B and FOO3B.

                                                                 **CRITICAL STEP**SAT/UNSAT PROMPT: If informed as Unit SRO that B Loop cannot be placed in SPC, inform the performer that another operator will be assigned to complete isolation of RHR Loop B.

Step 12 Informs Unit CR5 RHR B Loop cannot be placed in SPC. CRS informed that RHA Loop B cannot be placed in SPC.

                                                                                        *SAT/U NSAT TERMINATING CUE: When the service water radiation release has been terminated and the Unit CRS is informed this JPM is complete.

TIME COMPLETED: COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. LOT-SIM-JP-017-A12 Page 7 of 10 Rev.2

Step Critical I Not Critical Reason 1 Not Critical Administrative 2 Critical Necessary for RHR B loop SPC with Nuc SW. 3 Not Critical Unnecessary alignment. 4 Critical Necessary for RHR B loop SPC HX cooling. 5 Critical Necessary for RHR B loop SPC HX cooling. 6 Not Critical Unnecessary alignment. 7 Critical Necessary for RHR flow 8 Critical Necessary for RHR B loop SPC 9 Critical Necessary for RHR B loop SPC 10 Critical Necessary for RHR B loop SPC 11 Critical Necessary to terminate release 12 Not Critical Communication REVISION

SUMMARY

2 New template. Enhanced Standards for JPM steps Changed SCO to CRS Added Critical Step delineation 1 Converted format to Word using current JPM template, updated to match the current revision of the procedure. Changed title to LOT from LOR. 0 Initial Revision. (Provide sufficient detail for reviewers and evaluators to understand the scope of any technical and/or administrative changes). LOT-SIM-JP-017-A12 Page 8 of 10 Rev.2

Validation Time: 15 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: 2 Setting: In-Plant Simulator X Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes X No EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: O Comments reviewed with Performer Evaluator Signature: Date: LOT-SIM-JP-017-A12 Page 90110 Rev.2

TASK CONDITIONS:

1. A Resin Injection has occurred, currently raising power to burn off the resin.

2. RWCU has been isolated.

3. SRV C was not properly seated, subsequently it has been closed.

4. Torus temperature is >95 deg. F.

5. The A Loop of RHR has been started in Suppression Pool Cooling mode using CSW.

6. 2D RHR and 2D RHR SW Pumps are under clearance.

INITIATING CUE: You are directed to place the B Loop of RHR in Suppression Pool Cooling in accordance with the hard card using NSW and inform the Unit CRS when the required actions are complete. Page 10 of 10

RESIDUAL HEAT REMOVAL SYSTEM OPERATING 2OP-l 7 PROCEDURE Rev. 174 Page 330 of 334 ATTACHMENT 14 Page 1 of 1

           << Emergency Suppression Pool Cooling Using Loop B (2OP-17)>>

NOTE This attachment is not to be used for normal system operations C Start RHR SW B LOOP (NUC) Start RHR SW B LOOP CCONV) Open SW-V105 Open SW-ViOl Close SW-V143 11 Open SW-Vl02 Start CSW PUMPS AS NEEDED Close SW-V143 IF LOCA SIGNAL IS PRESENT, Start PUMPS ON NSW HDR AS NEEDED THEN place RHR SW BOOSTER W LOCA SIGNAL IS PRESENT, PUMPS B & D LOCA OVERRIDE THEN place RHR SW BOOSTER PUMPS SWITCH TO MANUAL OVERRIDE B & D LOCA OVERRIDE SWITCH TO MANUAL OVERRIDE Start RHR SW PMP StartRHRSWPMP Adjust El l-PDV-F068B Adjust El 1-PDV-F068B Establish CLG WTR TO VITAL HDR Establish CLG WTR TO VITAL HDR Start ADDITIONAL RHR SW PUMP Start ADDITIONAL RHR SW PUMP and adjust FLOW AS NEEDED U and adjust FLOW AS NEEDED Start RHR LOOP B IF LOCA SIGNAL IS PRESENT, THEN D Verify COOLING LOGIC IS MADE UP IF E1l-FOI5B IS OPEN, THEN close Ell-FO17B Start LOOP B RHR PMP Open Ell-F028B D Throttle El l-F024B Throttle Ell-F048B C Start ADDITIONAL LOOP A RHR PMP and adjust FLOW AS NEEDED U 2 2/1063 2 5/1064

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE SIM JPM E 2016 NRC INITIAL EXAM ROIISRO LESSON TITLE: VENT THE DRYWELL PER OP-JO WI STACK RAD MONITOR INCREASE >50% LESSON NUMBER: LOT-SIM-JP-O1 O-A02 REVISION NO: 8 Dan HuIgin 8/18/16 PREPARER I DATE Bob Bohn 9/06/16 TECHNICAL REVIEWER I DATE Grant Newton 9/06/16 VALIDATOR I DATE LINE SERVISOR I D 1 A TRAINING SUPERVISION APPROVAL I DATE LOT-SIM-JP-01 0-A02 Page 1 of 13 Rev.8

RELATED TASKS: 261 008B1 01 Perform Normal Primary Containment Venting KIA REFERENCE AND IMPORTANCE RATING: 261000 A4.01 Ability to manually operate and/or monitor in the Control Room Off site Release Rate

REFERENCES:

20P-1 0,Section 6.3.2-Venting Containment Via SBGT TOOLS AND EQUIPMENT: None SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): 9 Radioactivity Release SAFETY CONSIDERATIONS None LOT-SIM-JP-010-A02 Page 2 of 13 Rev.8

SETUP INSTRUCTIONS Recommended Initial Conditions IC-Il, 100% Power, BOC Reciuired Plant Conditions

1. Drywell Pressure above 0.5 psig SLOWLY rising or stable, AND below 1.8 psig.

Triqqers Trigger 1 Q6225LGT CAC-V23 Green Lamp = False. Malfunctions None Overrides Event Panel Tag Title Value Activate Deactivate (ramp rate) Time Time (sec) (sec) El XU-3 G5BO2G1 5 Main Stack Radiation 2.48 / 2 mm 0 SEC N/A Remotes None Special Instructions

1. Secure Drywell Coolers 2C and 2D Fans 1 and 2

2. Allow drywell pressure to rise to 0.6 psig as indicated on CAC-PI-2685-1 on XU 51.

3. Restart Drywell Coolers 2D Fan 2 and allow Drywell pressure to stabilize.

4. Override Drywell Cooler 2C Fans 1 and 2 and Drywell Cooler 2D Fan I control switches OFF LOT-SIM-JP-010-A02 Page 3 of 13 Rev.8

SAFETY CONSIDERATIONS:

1. None.

EVALUATOR NOTES: (Do not read to performer)

1. A marked up copy of 20P-1 0, Section 6.3.2 WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG 1021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM will be performed in the simulator on Unit Two.

Read the following to the ]PM performer. TASK CONDITIONS:

1. Drywell pressure is above normal due to a partial loss of Drywell Cooling.

2. Standby Gas Treatment System is in the Standby Alignment.

3. The plant stack radiation monitor is in service and CAC-CS-5519, CAC Purge Vent Isolation Override is in OFF.

4. ERFIS is unavailable.

INITIATING CUE: The Unit CRS directs you to vent the Drywell via Standby Gas Treatment, and inform him (her) when drywell pressure has been reduced below 0.5 psig. LOT-SIM-JP-0J0-A02 Page 4 of 13 Rev.8

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments requited for any step evaluated as UNSAT. Step 1 - Obtain copy of 20P-1O Standby Gas Treatment System Operating Procedure, Section 6.3.2. Copy of 20P-lO Standby Gas Treatment System Operating Procedure, Section 6.3.2 is obtained. SAT/U NSAT TIME START: Step 2 Record 2-D12-RR-R600B (Stack Rad Monitor) digital point display. Records the 2-D72-RR-R600B (Stack Rad Monitor) digital point display in the space provided on step 6.3.2.2a (value of 1.12 El or 11.2). SATIUNSAT Step 3 Multiply the value obtained in Step 2.a by 1.5 to obtain the value for a 50% rise in stack radiation monitor reading. Records the value obtained in step 6.3.2.2a (1.12 El or 77.2) in the space provided on step 6.3.2.2b. Multiplies the value by 1.5, and records the product (1.68 El or 16.8) in the space provided in step 6.3.2.2b. SATIUNSAT PROMPT: If asked, sign the procedure as the IV for the calculation. Do not change the value if calculated incorrectly. Step 4 Close 2-VA-2D-BFV-RB (Reactor Building SBGT Train 2A Inlet Valve). Reactor Building SBGT Train 2A Inlet Valve 2-VA-2D-BFV-RB switch is rotated counterclockwise to the close position and held (throttle valve) for 70 seconds after the red light is extinguished and the green light is illuminated at which time it can be released to the neutral position. 2-VA -2D-BFV-RB is closed* (*critical)

                                                                  **CRITICAL STEP**SATIUNSAT LOT-SIM-JP-010-A02                            Page 5 of 13                                          Rev.8

Step 5 Close 2-VA-2H-BFV-RB (Reactor Building SBGT Train 2B Inlet Valve). Reactor Building SBGT Train 28 Inlet Valve 2-VA -2H-BFV-RB switch is rotated counterclockwise to the close position and held (throttle valve) for 70 seconds after the red light is extinguished and the green light is illuminated at which time it can be released to the neutral position. 2-VA-2H-BFV-RB is closed* (*critical)

                                                                   **CRITICAL STEP**SATIUNSAT Step 6 Open 2-VA-2F-BFV-RB (SBGT DW Suct Damper).

SBGT DW Suction Damper 2-VA -2F-BFV-RB control switch is rotated clockwise to the open position and then released. Observes red light illuminates and the green light goes out. 2-VA -2F-BFV-RB is open* (*critical)

                                                                   **CRITICAL STEP**SATIUNSAT I PROMPT:    If asked as CRS, direct performer to vent the drywell only.                                I Step 7 Open 2-CAC-V9 (DW Purge Exh Vlv).

DWPurge Exh Vlv 2-CAC-V9 switch is rotated clockwise from the close position to the open position and then released to the neutral position. Observes the red light illuminates and the green light goes out. 2-CAC-V9 is open* (*critical)

                                                                   **CRITICAL STEP**SATIUNSAT SIM OP:    When CAC-V23 is opened, verify Trigger 1 initiates to ramp Main Stack Rad Monitor value.

Step S Open 2-CAC-V23 (DW Purge Exh Vlv). DWPurge Exh Vlv 2-CAC-V23 switch is rotated clockwise from the close position to the open position and then released to the neutral position. Observes the red light illuminates and the green light goes out. 2-CAC-V23 is open* (*criticaI)

                                                                   **CRITICAL STEP**SATIUNSAT LOT-SIM-JP-O1O-A02                          Page 6 of 13                                       Rev.8

PROMPT: If requested as CRS, inform performer that it is desired to vent from the drywell head (additional vent capacity is desired). Step 9 IF additional vent capacity is desired, THEN open 2-CAC-V49 ( DW Head Purge Exh Vlv). 2-CA C-V49 switch is rotated clockwise from the close position to the open position and then released to the neutral position. Observes the red light illuminates and the green light goes out. 2-CA C-V49 is open. SATIU NSAT Step 10 IF additional vent capacity is desired, THEN open 2-CAC-V50 ( DW Head Purge Exh Vlv). 2-CA C-V50 switch is rotated clockwise from the close position to the open position and then released to the neutral position. Observes the red light illuminates and the green light goes out. 2-CA C-V50 is open. SATIUNSAT Step 11 On Panel XU-3, monitor 2-Di 2-RR-R600B (Stack Rad Monitor) for a rise in activity during the performance of this procedure. Monitors 2-D12-RR-R6008 for a rise in activity, and determines Stack Rad Monitor reading has risen by 50%. SATIU NSAT NOTE: It is critical for at least one valve to be closed in each vent path that is open, i.e., CAC-V23 or CAC-V9, AND, CAC-V49 or CAC-V50, or that the primary containment suction valve VA-2F-BFV-RB is closed to isolate the release path. SIM OP: When the vent path has been isolated, delete the meter override on the Main Stack Rad Monitor. PROMPT: lithe examinee informs the Unit CRS that the Main Stack has risen by >50%, direct examinee as Unit CRS to perform requited actions for the increase. NOTE: Either Step 12 OR Step 13 is CRITICAL. PROMPT: Another operator is available to perform Independent Verifications.

                            **ALTERNATE PATH BEGINS AT STEP 12**

LOT-SIM-JP-010-A02 Page 7 of 13 Rev.8

Step 12 IF stack radiation rises to greater than the value determined in Section 6.3.2 Step 2.b, THEN perform the following to secure venting the drywell: Close 2-CAC-V23 (DW Purge Exh VIv). 2-CA C-V23 control switch is rotated counterclockwise to the close position. Observes the red light goes out and the green light illuminates. 2-CA C-V23 is closed* (*critical)

                                                                 **CRITICAL STEP**SATIUNSAT Step 13  IF stack radiation rises to greater than the value determined in Section 6.3.2 Step 2.b, THEN perform the following to secure venting the drywell Close 2-CAC-V9 (Drywell Purge Exh Vlv).

2-CA C-V9 control switch is rotated counterclockwise to the close position. Observes the red light goes out and the green light illuminates. 2-CAC-V9 is closed* (*critical)

                                                                 **CRITICAL STEP**SATIUNSAT I NOTE:       Either Step 14 OR Step 15 is CRITICAL if the 2-CAC-49 and V50 were opened.            I Step 14  IF stack radiation rises to greater than the value determined in Section 6.3.2 Step 2.b, THEN perform the following to secure venting the drywell Ensure 2-CAC-V49 (DW Head Purge Exh VIv) is CLOSED.

2-CA C-V49 control switch is rotated counterclockwise to the close position. Observes the red light goes out and the green light illuminates. 2-CA C-V49 is closed* (*critical)

                                                                 **CRITICAL STEP**SATIUNSAT Step 15  IF stack radiation rises to greater than the value determined in Section 6.3.2 Step 2.b, THEN perform the following to secure venting the drywell Ensure 2-CAC-V50 (DW Head Purge Exh Vlv) is CLOSED.

2-CA C-V50 control switch is rotated counterclockwise to the close position. Observes the red light goes out and the green light illuminates. 2-CA C-V50 is closed* (*critical)

                                                                 **CRITICAL STEP**SATIUNSAT LOT-SIM-JP-O1O-A02                         Page 8 of 13                                     Rev.8

PROMPT: If asked, inform examinee as Unit CR5 that E&RC has been notified to sample primary containment, and to reference E&RC 2020 Setpoint Determinations for Gaseous Radiation Monitors (Noble Gas Instantaneous Release Rate Determination). NOTE: Step 16 is not critical if either step 12 or 13 AND either l4orl5was completed SAT. Release path may be isolated by closing 1 valve in each vent path OR by closing the common isolation in Step 16. Step 16 CLOSE SBGT DW SUCT DAMPER, 2-VA-2F-BFV-RB. SBGT DW Suction Damper 2-VA -2F-BFV-RB control switch is rotated counter clockwise to the closed position and then released. Observes green light illuminates and the red light goes out. 2-VA -2F-BF V-RB is closed* (*critical)

                                                                   **CRITICAL STEP**SATIUNSAT NOTE:     The following valves would auto open on SBGT Initiation therefore Steps 17 and 18 are NOT critical.

Step 17 Open 2-VA-2H-BFV-RB (SBGT Train 2B Reactor Building Suction Valve). Verifies 2-VA -2H-BFV-RB indicates full open. Observes green light out and the red light lit. 2-VA -2H-BFV-RB is full open. SATIU NSAT Step 18 Open 2-VA-2D-BFV-RB (SBGT Train 2A Reactor Building Suction Valve). Verifies 2-VA -2D-BFV-RB indicates full open. Observes green light out and the red light lit. 2-VA -2D-BF V-RB is full open. SATIU NSAT PROMT: If asked, inform examinee as Unit CRS that another operator is standing by to perform OPT-02.3.lb, Suppression Pool to Drywell Vacuum Breaker Position Check. LOT-SIM-JP-010-A02 Page 9 of 13 Rev.8

Step 19 Inform Unit CRS that venting is secured due to increase of 50% in Main Stack Rad Monitor reading. Unit CRS is in formed venting is secured due to increase of 50% in Main Stack Rad Monitor reading. SATIU NSAT TERMINATING CUE: When Primary containment Venting has been secured and the Unit CRS is notified, this JPM is complete. TIME COMPLETED: COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. LOT-SIM-JP-010-A02 Page 10 of 13 Rev.8

Step Critical I Not Critical Reason 1 Not Critical Administrative 2 Not Critical Readings 3 Not Critical Readings 4 Critical Necessary for venting alignment 5 Critical Necessary for venting alignment 6 Critical Necessary for venting alignment 7 Critical Necessary for venting alignment 8 Critical Necessary for venting alignment 9 Not Critical Additional venting alignment 10 Not Critical Additional venting alignment 11 Not Critical Monitoring 12 Critical Communication 13 Critical Necessary for termination of release 14 Critical Necessary for termination of release 15 Critical Necessary for termination of release 16 Critical Necessary for termination of release 17 Not Critical Auto-action for securing vent 18 Not Critical Auto-action for securing vent 19 Not Critical Communication REVISION

SUMMARY

8 Enhanced Standards for JPM steps Added Critical Step delineation Fixed numbering 7 Updated to the new JPM template. (Provide sufficient detail for reviewers and evaluators to understand the scope of any technical and/or administrative changes). LOT-SIM-JP-010-A02 Page 11 of 13 Rev.8

Validation Time: 15 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: 2 Setting: In-Plant Simulator X Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes X No EVALUATION Performer: ]PM: Pass Fail Remedial Training Required: Yes No Comments: D Comments reviewed with Performer Evaluator Signature: Date: LOT-SIM-JP-010-A02 Page 12 of 13 Rev.8

TASK CONDITIONS:

1. Drywell pressure is above normal due to a partial loss of Drywell Cooling.

2. Standby Gas Treatment System is in the Standby Alignment.

3. The plant stack radiation monitor is in service and CAC-CS-5519, CAC Purge Vent Isolation Override is in OFF.

4. ERFIS is unavailable.

INITIATING CUE: The Unit CRS directs you to vent the Drywell via Standby Gas Treatment, and inform him (her) when drywell pressure has been reduced below 0.5 psig. Page 13 of 13

STANDBY GAS TREATMENT SYSTEM OPERATING 20P-10 PROCEDURE Rev. 81 Page 19 of 49 6.3.2 Venting Containment Via SBGT Date/Time Started________________

1. Confirm the following Initial Conditions are met:

• Drywell pressure has risen to greater than 0.15 psig
• SBGT System is in STANDBY in accordance with Section 6.1.1
• Qn of the following:

o Plant stack radiation monitor is in service and CAC-CS-551 9 (CAC Purge Vent Isol Ovrd) is in OFF O E&C has sampled the drywell atmosphere and has determined that it is suitable for release

• Unit CR5 approval is obtained prior to venting NOTE
• Backwashing an RWCU filter or AOG sampling may cause stack radiation to rise. Venting primary containment and RWCU backwash ing or AOG effluent sampling are NOT done concurrently E1
• To aid in monitoring for a 50% rise in stack radiation level, value monitoring alarm may be set up to monitor ERFIS point

2. lF2-D12-RR-R600B (Stack Rad Monitor) is OPERABLE THEN perform the following

a. Record 2-D12-RR-R600B (Stack Rad Monitor) digital point display 2-D1 2-RR-R600B:

b. Multiply the value obtained in Step 2.a by 1 .5 to obtain the value for a 50% rise in stack radiation monitor reading /

IV (1.5)X(___________ )= (Section 6.3.2 Step 2.a)

3. Close 2-VA-2D-BFV-RB (Reactor Building SBGT Train 2A Inlet Valve)

STANDBY GAS TREATMENT SYSTEM OPERATING 20P-10 PROCEDURE Rev. 81 Page 20 of 49 6.3.2 Venting Containment Via SBGT (continued)

4. Close 2-VA-2H-BFV-RB (Reactor Building SBGT Train 2B Inlet Valve)

5. Open 2-VA-2F-BFV-RB (SBGT DW Suct Damper)

CAUTION Simultaneous venting of the drywell and the suppression pool is NOT performed when the plant is in MODE 1,2, or3. {8.1.1} D

6. E venting the suppression chamber, THEN perform the following

a. Open 2-CAC-V172 (Supp Pool Purge Exh VIv)

b. Open 2-CAC-V22 (Torus Purge Exhaust Vlv)

c. IF 2-D12-RR-R600B (Stack Rad Monitor) is OPERABLE THEN perform the foIlowing (1) On Panel XU-3, monitor 2-Di 2-RR-R600B (Stack Rad Monitor) for a rise in activity during the performance of this procedure (2) IF stack radiation rises to greater than the value determined in Step 2.b, THEN perform the following to secure venting the suppression pool

• Close 2-CAC-V172 (Supp Pool Purge Exh Vlv)... I IV
• Close 2-CAC-V22 (Torus Purge Exhaust Vlv)

IV

STANDBY GAS TREATMENT SYSTEM OPERATING 20-10 PROCEDURE Rev. 81 Page 21 of49 6.3.2 Venting Containment Via SBGT (continued)

• Notify E&C to perform the following:

0 Sample primary containment 0 Refer to OE&RC-2020, Setpoint Determinations for Gaseous Radiation Monitors (Noble Gas Instantaneous Release Rate Determination) Person Notified

d. WHEN the desired suppression chamber pressure is reached, as indicated on Computer Point L128, THEN close the following valves

                    .      2-CAC-V172 (Supp Pool Purge Exh Vlv)                           I IV
                    .      2-CAC-V22 (Torus Purge Exhaust Vlv)                            I IV

7. IF venting the drywell, THEN perform the following

a. Open 2-CAC-V9 (DW Purge Exh Vlv)

b. Open 2-CAC-V23 (DW Purge Exh Vlv)

NOTE CAC-V49 (DW Head Purge Exh Vlv) and CAC-V50 (DW Head Purge Exh Vlv) are NOT normally opened, but are available if additional vent capacity is desired D

c. IF additional vent capacity is desired, THEN open 2-CAC-V49 f DW Head Purge Exh VIv)

d. IF additional vent capacity is desired, THEN open 2-CAC-V50 ( DW Head Purge Exh Vlv)

e. On Panel XU-3, monitor 2-D12-RR-R600B (Stack Rad Monitor) for a rise in activity during the performance of this procedure

STANDBY GAS TREATMENT SYSTEM OPERATING 20 P-i 0 PROCEDURE Rev. 81 Page 22 of 49 6.3.2 Venting Containment Via SBGT (continued) IF 2-D12-RR-R600B (Stack Rad Monitor) is OPERABLE THEN perform the following (1) On Panel XU-3, monitor 2-Di 2-RR-R600B (Stack Rad Monitor) for a rise in activity during the performance of this procedure (2) IF stack radiation rises to greater than the value determined in Section 6.3.2 Step 2.b, THEN perform the following to secure venting the drywell

• Close 2-CAC-V23 (DW Purge Exh Vlv) I IV
• Close 2-CAC-V9 (Drywell Purge Exh VIv)

IV

• Ensure 2-CAC-V49 (DW Head Purge Exh Vlv) is CLOSED IV
• Ensure 2-CAC-V50 (DW Head Purge Exh VIv) is CLOSED /

IV

• Notify E&C to perform the following:

0 Sample primary containment O Refer to OE&RC-2020, Setpoint Determinations for Gaseous Radiation Monitors (Noble Gas Instantaneous Release Rate Determination) Person Notified

g. WHEN the desired drywell pressure is reached as indicated on CAC-Pl-2685-i (Drywell Pressure) on Panel XU-5i, THEN perform the following

• Close 2-CAC-V23 (DW Purge Exh Vlv) /

IV

STANDBY GAS TREATMENT SYSTEM OPERATING 20P-1O PROCEDURE Rev 81 Page 23 of 49 6.3.2 Venting Containment Via SBGT (continued)

• Close 2-CAC-V9 (Drywell Purge Exh Vlv) I IV
• Ensure 2-CAC-V49 (DW Head Purge Exh Vlv) is CLOSED /

IV

• Ensure 2-CAC-V50 (DW Head Purge Exh Vlv) is CLOSED /

IV

8. Close 2-VA-2F-BFV-RB (SBGT DW Suct Damper) /

IV

9. Open 2-VA-2H-BFV-RB (SBGT Train 2B Reactor Building Suction Valve) /

IV

10. Open 2-VA-2D-BFV-RB (SBGT Train 2A Reactor Building Suction Valve) /

IV NOTE Technical Specification 3.6.1 .6.1 (MODES 1, 2, or 3) requires completion of OPT-02.3.IB, Suppression Pool To Drywell Vacuum Breaker Position Check, within 6 hours following an operation that causes any of the vacuum breakers to open. Due to the inability to detect a vacuum breaker opening and subsequently reclosing, other than observation from a dedicated Operator, the conservative approach is to perform OPT-02.3.1B, Suppression Pool To Drywell Vacuum Breaker Position Check following venting activities

11. IF in MODES 1, 2, or 3, THEN perform OPT-02.3.1 B, Suppression Pool To Drywell Vacuum Breaker Position Check, within 6 hours following an operation that causes any of the vacuum breakers to open

STANDBY GAS TREATMENT SYSTEM OPERATING 20P-1O PROCEDURE Rev. 81 Page 24 of 49 6.3.2 Venting Containment Via SBGT (continued) Date/Time Completed Performed By (Print) Initials Reviewed By Unit CRS/SRO

DUKE ENERGY BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE SIM 1PM F 2076 NRC INITIAL EXAM

          -                        - RO/ISRO LESSON TITLE:      Shifting Caswell Beach Lube Water Pumps From The RTGB LESSON NUMBER: LOT-SIM-JP-029-07 REVISION NO:       0 Dan Hulqin                                      8/18/16 PREPARER I DATE Bob Bolin                                       9/06/16 TECHNICAL REVIEWER I DATE Dwayne Wolf Grant Newton                                    9/06/16 VALIDATOR I DATE

( LINE SPRVISOR I DAtE TRAINI G S ERVISION A ROVAL I DATE LOT-SIM-JP-029-01 Page 1 of 10 Rev. 0

Shifting Caswell Beach Lube Water Pumps From The RTGB RELATED TASKS: 275002B101 Startup The Circulating Water System Per OP-29 K/A REFERENCE AND IMPORTANCE RATING: 400000 A4.01 Ability to manually operate and/or monitor in the control room: CCW indications and control.

REFERENCES:

20P-29, Section 6.3.27, Shifting Caswell Beach Lube Water Pumps From The RTGB TOOLS AND EQUIPMENT: None SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): 8 - Plant Service Systems (Component Cooling Water System) LOT-SIM-JP-029-01 Page 2 of 10 Rev. 0

Shifting Caswell Beach Lube Water Pumps From The RTGB SETUP INSTRUCTIONS SIMULATOR SETUP Initial Conditions

1. Recommended Initial Conditions IC-il

2. Required Plant Conditions None.

Malfunctions: None. Special Instructions None LOT-SIM-JP-029-Ol Page 3 of 10 Rev. 0

Shifting Caswell Beach Lube Water Pumps From The RTGB SAFETY CONSIDERATIONS:

1. None EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM will be performed in the simulator on Unit Two.

Read the following to the JPM performer. TASK CONDITIONS:

1. An Auxiliary operator is stationed at Caswell Beach.

2. All Section 5.0 prerequisites of 20P-29, Circulating Water System are met.

INITIATING CUE: You are directed by the Unit CRS to place Caswell Beach Bearing Lube Water pump 2B in service, and secure the Caswell Beach Bearing Lube Water pump 2A lAW 20P-29 Section 6.3.27, Shifting Caswell Beach Lube Water Pumps From The RTGB. Inform the CRS when complete. LOT-SIM-JP-029-01 Page 4 of 10 Rev. 0

Shiftinci Caswell Beach Lube Water Pumps From The RTGB PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. Step 0 May perform take a minute at job site prior to beginning task. Examinee may cover the following questions, as deemed necessary. What are the hazards in the area? What PPE is required? Am I using appropriate gloves? Tools/PPE inspected prior to use? Energy sources secured/isolated? Is Clearance/Tag Out sufficient? Whats the worst that can happen? Any ALARA concerns? Will I affect plant status? HU Tools needed? Step 1 - Obtain current revision of 20P-29 Section 6.3.27. Provide current revision of 20P-29 Section 6.3.27. SAT/UNSAT TIME START: PROMPT: When contacted as the Auxiliary Operator at Caswell Beach, acknowledge the communication, and report that you are standing by for the pump shift. Step 2 - Establish communication with Auxiliary Operator at Caswell Beach. Contacts the AO at Caswell Beach. SAT/UNSAT NOTE: When the Point Select push button is depressed it initiates a 10 second window for starting or stopping Lube Water pumps. Place keeping of the pump start steps and pump stop steps may be deferred until after the Lube Water pump is running or stopped. Step 3 Start the non-operating Bearing Lube Water pump as follows:

a. Depress Point Select push button for the selected Bearing Lube Water pump.

Depresses Point Select push button for the Caswell Beach Bearing Lube Water pump 2B on Panel XU-2 and the checkback lamp comes ON.

                                                               **CRITICAL STEP**           SAT/UNSAT LOT-SIM-JP-029-01                           Page 5 of 10                                           Rev. 0

Shifting Caswell Beach Lube Water Pumps From The RTGB Step 4 Start the non-operating Bearing Lube Water pump as follows:

b. WHEN checkback lamp in the selected point push button comes ON, THEN place the Bearing Lube Water pump control switch to START.

Places the control switch for the Cas well Beach Bearing Lube Water pump 2B to START, within 10 seconds of the checkback lamp for the Caswell Beach Bearing Lube Water pump 28 point push button coming ON and the Caswell Beach Bearing Lube Water pump 28 starts.

                                                             **CRITICAL STEP**   SAT/UNSAT PROMPT:       When contacted as the Auxiliary Operator at Caswell Beach, acknowledge the communication and report a good start on the Caswell Beach Bearing Lube Water pump 2B.

Step 5 Start the non-operating Bearing Lube Water pump as follows:

c. Confirm with the Auxiliary Operator that the pump starts without unusual noise or cavitation.

Confirms with the Auxiliary Operator that the Caswell Beach Bearing Lube Water pump 28 has started without unusual noise or cavitation. SAT/UNSAT Step 6 Stop the previously operating Bearing Lube Water pump as follows:

a. Depress Point Select push button for the selected Bearing Lube Water pump.

Depresses Point Select push button for the Cas well Beach Bearing Lube Water pump 2A on Panel XU-2 and the checkback lamp comes ON. CRITICAL STEP ** SAT/UNSAT Step 7 Stop the previously operating Bearing Lube Water pump as follows:

b. WHEN checkback lamp in the selected point push button comes ON, THEN place the Bearing Lube Water pump control switch to STOP.

Places the control switch for the Caswell Beach Bearing Lube Water pump 2A to STOP, within 70 seconds of the checkback lamp for the Caswell Beach Bearing Lube Water pump 2A point push button coming ON and the Cas well Beach Bearing Lube Water pump 2A stops. CRITICAL STEP ** SAT/UNSAT LOT-SIM-JP-029-O1 Page 6 of 10 Rev. 0

Shifting Caswell Beach Lube Water Pumps From The RTGB PROMPT: When contacted as the Auxiliary Operator at Caswell Beach, acknowledge the communication and report that the Caswell Beach Bearing Lube Water pump 2A has stopped, and that lube water flow is adequate for all CWOD pumps. Step 8 Stop the non-operating Bearing Lube Water pump as follows:

c. Confirm with the Auxiliary Operator that the pump stops.

Confirms with the Auxiliary Operator that the Caswell Beach Bearing Lube Water pump 2A has stopped. SAT/UNSAT Step 9 Informs CRS that the Caswell Beach Lube Water Pumps have been shifted with the 2B pump now running and the 2A pump secured. Informs CRS that the pump shift is complete. SAT/U NSAT TERMINATING CUE: Once Caswell Beach Bearing Lube Water pump 2B is running ,pump 2A is secured, and the CRS has been notified the JPM can be terminated. TIME COMPLETED: COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. Step Critical / Not Critical Reason 1 Not Critical Administrative 2 Not Critical Communication 3 Critical Action required to start pump 4 Critical Action required to start pump 5 Not Critical Communication 6 Critical Action required to stop pump 7 Critical Action required to stop pump 8 Not Critical Communication 9 Not Critical Communication LOT-SIM-JP-029-O1 Page 7 of 10 Rev. 0

Shifting Caswell Beach Lube Water Pumps From The RTGB REVISION

SUMMARY

0 New JPM written for 2016 Initial NRC exam. LOT-SIM-JP-029-O1 Page 8 of 10 Rev. 0

Shiftinci Caswell Beach Lube Water Pumps From The RTGB Validation Time: 10 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: 2 Setting: In-Plant Simulator X Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: O Comments reviewed with Performer Evaluator Signature: Date: LOT-SIM-JP-029-01 Page 9 of 10 Rev. 0

TASK CONDITIONS:

1. An Auxiliary operator is stationed at Caswell Beach.

2. All Section 5.0 prerequisites of 20P-29, Circulating Water System are met.

INITIATING CUE: You are directed by the Unit CRS to place Caswell Beach Bearing Lube Water pump 2B in service, and secure the Caswell Beach Bearing Lube Water pump 2A lAW 20P-29 Section 6.3.27,Shifting Caswell Beach Lube Water Pumps From The RTGB. Inform the CRS when complete. Page 10 of 10

CIRCULATING WATER SYSTEM 20P-29 Rev. 174 Page 130 of 211 6.3.27 Shifting Caswell Beach Lube Water Pumps From The RTGB

1. Confirm the following initial Conditions are met:

• All applicable prerequisites listed in Section 5.0, Prerequisites are met
• An Auxiliary Operator is available at Caswell Beach

2. Establish communication with Auxiliary Operator at Caswell Beach...

NOTE When the Point Select push button is depressed it initiates a 10 second window for starting or stopping Lube Water pumps. Place keeping of the pump start steps and pump stop steps may be differed until after the Lube Water pump is running or stopped D

3. Start the non-operating Bearing Lube Water pump as follows:

a. Depress Point Select push button for the selected Bearing Lube Water pump

b. WHEN checkback lamp in the selected point push button comes ON, THEN place the Bearing Lube Water pump control switch to START -

c. Confirm with the Auxiliary Operator that the pump starts without unusual noise or cavitation

4. Stop the previously operating Bearing Lube Water pump as follows:

a. Depress Point Select push button for the selected Bearing Lube Water pump

b. WHEN checkback lamp in the selected point push button comes ON, THEN place the Bearing Lube Water pump control switch to STOP

c. Confirm with the Auxiliary Operator that the pump stops

5. Ensure lube water flow is adequate for all CWOD pumps per 001-03.11, Auxiliary Operator U0 Outside Electronic Rounds (DSR & CS)

CIRCULATING WATER SYSTEM 20P-29 Rev. 174 Page 131 of 211 6.3.27 Shifting Caswell Beach Lube Water Pumps From The RTGB (continued) Date/Time Completed Performed By (Print) Initials Reviewed By Unit CRS/SRO

DUKE ENERGY BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE SIM JPM G -2076 NRC INITIAL EXAM - RO/ISRO LESSON TITLE: SUBSTITUTING A CONTROL ROD POSITION INTO THE RWM LESSON NUMBER: LOT-SIM-JP-007-B02 REVISION NO: 3 Dan Hulciin 08/18/16 PREPARER I DATE Bob Bolin 09/06/16 TECHNICAL REVIEWER I DATE Dwayne Wolf 09/06/16 VALIDATOR I DATE (%&% LINE SUIERVISOR i o TRAINING SUPERVISION APPROVAL I DATE LOT-SIM-JP-007-B02 Page 1 of 14 Rev.3

RELATED TASKS: 214202B1 01 Determine The RWM Substitute Rod Position For A Failed Reed Switch Position Indicator Per OP-07 K/A REFERENCE AND IMPORTANCE RATING: 201006 A4.06 3.2/3.2 Ability to manually operate and/or monitor in the control room: Selected rod position indication

REFERENCES:

20P-07, Section 6.3.11 Determination Of The RWM Substitute Position 001-53, Rod Worth Minimizer (NUMAC-RWM) TOOLS AND EQUIPMENT: None SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): 7 instrumentation SAFETY CONSIDERATIONS None LOT-SIM-JP-007-B02 Page 2 of 14 Rev.3

SETUP INSTRUCTIONS Recommended Initial Conditions IC-li, 100% Power, BOC Reciuired Plant Conditions Fail the Reed switch for Control Rod 22-03 position 48. Tricicje rs None Malfunctions System Tag Title Value RD RD179M Reed Switch Failure Rod 22-03 48 Overrides None Remotes None Special Instructions None LOT-SIM-JP-007-B02 Page 3 of 14 Rev.3

SAFETY CONSIDERATIONS:

1. None.

EVALUATOR NOTES: (Do not read to pertormer)

1. The applicable copy of 20P-07, Section 6.3.11 WILL be provided to the performer.

2. If requested, a copy of 001-53, WILL be provided to the performer.

3. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG 1021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

4. This JPM will be performed in the simulator on Unit Two.

Read the following to the JPM performer. TASK CONDITIONS:

1. Control rod 22-03 has just been positioned at position 46.

2. Prerequisites listed in section 5.0 of 20P-07 are met.

3. The on duty Reactor Engineer has been notified.

4. The Unit CRS has reviewed Technical Specifications tor applicability and has given permission to perform this procedure.

INITIATING CUE: You are directed to enter a substitute value for Control Rod 22 03 into the RWM, and inform the CRS when complete. LOT-SIM-JP-007-B02 Page 4 of 14 Rev.3

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. Step 1 - Obtain a copy of 20P-07, Reactor Manual Control System Operating Procedure, Section 6.3.11. Copy of 20P-07, Reactor Manual Control System Operating Procedure, Section 6.3.77 is obtained. SAT/U NSAT TIME START: PROMPT: Role play as a Concurrent Verification, if requested. DO NOT correct the performer. PROMPT: Role play as CRS, to initial Steps 6.3.11.2, 4 & 5 when asked. Step 2 Using Concurrent Verification, Inserl or withdraw control rod one additional notch to an operable control rod reed switch position indicator.

• Turns on Rod Select Power by placing the control switch to ON.
• Selects Control Rod 22-03 on the RTGB select matrix by depressing its Control Rod Select pushbutton
• Inserts Control Rod 22-03 to position 46 using the Rod Movement control switch in the IN position for one notch.
• Control Rod 22-03 is at position 46* (critical*).
• Control Rod 22-03 is selected* (critical*).

                                                                 **CRITICAL STEP**SAT/UNSAT Step 3  Record the OPERABLE control rod reed switch position below.

Records Control Rod 22-03 at position 46 for step 6.3.7 7.6.c in the space labeled Operable Control Rod Reed Switch Position. SAT/U NSAT LOT-SIM-JP-007-B02 Page 5 of 14 Rev.3

Step 4 Restore control rod to the position of the failed control rod reed switch position indicator.

• Withdraws Control Rod 22-03 to position 48 using the Rod Movement control switch.
• May use continuous withdraw and perform an over travel check -OR- May single notch out to 48 and then attempt to notch out past position 48.
• Control Rod 22-03 is at position 48* (critical*).

                                                                   **CRITICAL STEP**SAT/UNSAT Step 5   Ensure inferred position offered by the RWM deviates by one notch and in the correct direction from the position determined in Section 6.3.11 Step 6.c.

Verifies RWM infers substitute value of 48. SAT/U NSAT Step 6 Record the inferred rod position below. Records 48 for step 6.3.7 7.6.g in the space labeled Inferred Position. SAT/UNSAT I NOTE: 001-53 may be used for guidance in performance of step 7. I Step 7 IF a valid inferred position from the RWM OR valid rod position determined by appropriate methods (as identified in Section 6.1.1) has been obtained, THEN perform the following: Substitute the valid rod position into RWM.

• At the RWM Operators Console on the RTGB, depresses the ETC softkey to obtain the SUBSTITUTE OPTIONS softkey.
• Depresses the SUBSTITUTE OPTIONS softkey to change to the SUBSTITUTE OPTION screen.
• Verifies RWM offers an inferred position of 48 as the substitute position or depresses the increment/decrement softkey to adjust the substitute position to 48.
• Depresses the ENTER SUBSTITUTE softkey.
• Depresses the EXIT softkey to return to the main menu screen
• 48 substituted as the position for Control Rod 2203* (critical*).

                                                                   **CRITICAL STEP**SAT/UNSAT LOT-SIM-JP-007-B02                           Page 6 of 14                                      Rev.3

PROMPT: Inform trainee as the Unit CRS that Reactor Engineer will enter the substitute value in the PPC. Also another operator will execute a CORE MON, and make the Log entries. Step 8 Notify Unit CRS that a substitute value of 48 has been entered for Control Rod 22-03 per OP-7.0. Unit CRS notified. SAT/UNSAT TERMINATING CUE: When Control rod 22-03 has been given a substitute position of 48 in the RWM, and the CRS has been notified, this JPM is complete. TIME COMPLETED: COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. Step Critical I Not Critical Reason 1 Not Critical Administrative 2 Critical Rod selected is necessary to substitute rod position in step 7. Rod at position 46 is necessary for RWM to infer correct position. 3 Not Critical Recording value 4 Critical Rod at position 48 is necessary for RWM to infer correct position. 5 Not Critical Verification 6 Not Critical Recording value 7 Critical Necessary to substitute value. 8 Not Critical Communication LOT-SIM-JP-007-B02 Page 7 of 14 Rev.3

REVISION

SUMMARY

3 New Format. Critical Steps added based on necessity for RWM inferred position. Standards enhanced Critical Step delineation table added. Changed SCO to CRS. Corrected procedure section. Added additional Notes and Prompts Validation time changed to 15 minutes based on validators time. 2 Convert to Word, changed title from LOR to LOT (Provide sufficient detail for reviewers and evaluators to understand the scope of any technical and/or administrative changes). LOT-SIM-JP-007-B02 Page 8 of 14 Rev.3

Validation Time: 15 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: 2 Setting: In-Plant Simulator X Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: D Comments reviewed with Performer Evaluator Signature: Date: LOT-SIM-JP-007-B02 Page 9 of 14 Rev.3

ATTACHMENT 5 C Continuous Page 1 of 4 Use Control Rod Movement The purpose of this attachment is to document the rod pattern prior to power change. Enter the rod position information or attach Display 810 Edit. Unit: 2 51 48 48 48 48 48 47 48 48 48 48 48 48 48 48 48 43 48 48 48 48 48 36 48 48 48 48 48 39 48 48 48 48 48 48 48 48 48 48 48 35 48 48 48 48 08 48 00 48 08 48 48 48 48 31 48 48 48 48 48 48 46 48 48 48 48 48 48 27 48 48 36 48 00 48 48 48 00 48 36 48 48 23 48 48 48 48 48 48 48 48 48 48 48 48 48 19 48 48 48 48 08 48 00 48 08 48 48 48 48 15 48 48 48 48 48 48 48 48 48 48 48 11 48 48 48 48 48 36 48 48 48 48 48 07 48 48 48 48 48 48 48 48 48 03 48 48 48 48 48 02 06 10 14 18 22 26 30 34 38 42 46 50 Prepared by: %aCe Date Today Reactor Engineer Verified by: Zi t&e Date Today Reactor Engineer or SRO Approved by: 57a4e ea*e Date Today Unit CRS OENP-24.5 Rev. 9 Page 17 of 23

C AUACHMENT5 Continuous Page 2 of 4 Use Control Rod Movement Individual Rod Movement Instructions Sheet 1 of 1 SRO Initials: Q Control Rod Correct Rod Control Rod Licensed Overtravel Full Out Second Comments Selected/Verified (CV) Position Operator Check Position Licensed To Check Operator NOTE 3 NOTE 1 NOTE2 22-03 / / / 46 22-03 / / / 48

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                   /         /        /

OENP-24.5 Rev. 9 Page 18 of 23

AUACHMENT5 C Continuous Page 3 of 4 Use Control Rod Movement NOTE 1: WHEN a control rod is withdrawn to the Full Out position, either MAINTAIN the continuous withdrawal signal for at least 3 to 5 seconds OR APPLY a separate notch withdrawal signal, AND PERFORM the following rod coupling integrity check: CONFIRM ROD OVER TRAVEL (A-05 4-2) annunciator does NOT alarm. (SR 3.1.3.4) CONFIRM rod full out light is not lost. CONFIRM rod position indication on the four-rod display indicates position 48. CONFIRM ROD DRIFT (A-05 3-2) annunciator does NOT alarm. NOTE 2: VERIFY the rod reed switch position indicator corresponds to the control rod position indicated by the Full Out teed switch. NOTE 3: Concurrent Verification (CV) of rod selection required prior to rod movement. Additional (CV) signoffs for subsequent rod selection following a deselect. [OENP-24.5 Rev. 9 Page 19 of 23

ATTACHMENT 5 Page 4 of 4 Control Rod Movement Other Instructions: Performing movement of Control Rod 22-03 to determine RWM substitute position. Date/Time Completed Performed By (Print) Initials Reviewed By: Unit CRS OENP-24.5 Rev. 9 Page 20 of 23

TASK CONDITIONS:

1. Control rod 22-03 has just been positioned at position 48.

2. Prerequisites listed in section 5.0 of 20P-07 are met.

3. The on duty Reactor Engineer has been notified.

4. The Unit CRS has reviewed Technical Specifications for applicability and has given permission to perform this procedure.

INITIATING CUE: You are directed to enter a substitute value for Control Rod 22-03 into the RWM, and inform the CRS when complete. OENP-24.5 Rev. 9 Page 20 of 23

REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-07 PROCEDURE Rev. 105 Page 69 of 162 6.3.11 Determination Of The RWM Substitute Position Control Rod

1. Record Control Rod Number above

2. Confirm the following initial conditions are met:

               .       All applicable prerequisites in Section 5.0 are met
               .       Attachment I has been reviewed
               .       RWM requires a position to be substituted
               .       Unit CRS has referenced Technical Specification SR 3.1.3.1 CRS

3. IF control rod movement is NOT documented by an approved procedure THEN obtain copy of OGP-1 1, Second Operator Rod Sequence Checkoff Sheets to record rod movements

4. IF AT ANY TIME a control rod position must be substituted in the RWM or a control rod bypassed in the RWM, THEN reference Technical Specification 3.3.2.1 CRS

5. Obtain permission from the Unit CRS to perform this procedure section CRS NOTE

• 001-53, Rod Worth Minimizer, may be of assistance in performance of this procedure D The RWM inferred position is NOT valid for conditions of three or more failed reed switch positions which include two failed even positions (e.g, RWM inferred position is NOT valid for failed positions 06, 07, 08). If below the RWM LPSP, then a one notch withdrawal/insert error may be created to determine a valid inferred position. If below the RWM LPSP, and at the Insert/Withdrawal limit, then a one notch withdrawal/insert error may be created to determine a valid inferred position 0

6. IF a valid rod position has NOT been determined by appropriate methods (as identified in Section 6.1.1),

THEN determine a valid RWM inferred position by performing the following

REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-07 PROCEDURE Rev. 105 Page 70 of 162 6.3.11 Determination Of The RWM Substitute Position (continued)

a. IF less than or equal to LPSP AND at a Withdraw Limit, THEN bypass the affected control rod using Section 6.3.14 BEGIN R.M. LEVEL R21R3 REACTIVITY EVOLUTION

b. Using Concurrent Verification, Insert or withdraw control rod one additional notch to an operable control rod reed switch position indicator

c. Record the OPERABLE control rod reed switch position below Operable Control Rod Reed Switch Position U. Restore control rod to the position of the failed control rod reed switch position indicator

e. IF control rod was bypassed in the RWM, THEN unbypass affected control rod using Section 6.3.15

1. Ensure inferred position offered by the RWM deviates by one notch and in the correct direction from the position determined in Section 6.3.11 Step 6.c

g. Record the inferred rod position below Inferred Position

h. E necessary to ensure a valid inferred position from the RWM, THEN repeat Section 6.3.11 Step 6.a through Section 6.3.11 Step 6.f

REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-07 PROCEDURE Rev. 105 Page 71 of 162 6.3.11 Determination Of The RWM Substitute Position (continued)

7. E a valid inferred position from the RWM OR valid rod position determined by appropriate methods (as identified in Section 6.1 .1) has been obtained, THEN perform the following _______

a. Substitute the valid rod position into RWM

b. Obtain concurrence of Reactor Engineer and substitute affected control rod position in PPC in accordance with OOP-55, Plant Process and ERFIS Computer Systems Operating Procedure

c. Execute a CORE MON and confirm the correct control rod position substitution was made

d. Record the following in the Operators log:

• Control rod number
• Failed position switch
• Substituted position

e. Continue rod motion as directed by the Unit CRS

8. IF a valid rod position can NOT be determined AND reactor power is less than or equal to LPSP, THEN perform the following

a. Refer to Technical Specification Section 3.1.3 CRS

b. Bypass affected control rod in the RWM using Section 6.3.14

c. Record control rod bypassed in the Operators log

d. Fully insert affected control rod until reactor power is greater than LPSP

REACTOR MANUAL CONTROL SYSTEM OPERATING 20P-07 PROCEDURE Rev. 105 Page 72 of 162 6.3.11 Determination Of The RWM Substitute Position (continued)

9. IF a valid control rod position can NOT be determined AND reactor power is greater than LPSP, THEN position control rod to a valid control rod position indication in accordance with the Reactor Engineer END R.M. LEVEL R21R3 REACTIVITY EVOLUTION

10. WHEN affected control rod is moved to a position with an OPERABLE reed switch position indicator a valid rod position can be determined, THEN perform the following

a. Obtain concurrence of Reactor Engineer and remove the PPC control rod position substituted in Section 6.3.11 Step 7.b, in accordance with OOP-55, Plant Process and ERFIS Computer Systems Operating Procedure

b. IF the control rod was bypassed in Section 6.3.11 Step 8.a, THEN unbypass affected control rod using Section 6.3.15

c. Record the control rod returned to normal in the Operators log Date/Time Completed Performed By (Print) Initials Reviewed By Unit CRS/SRO

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE SIM JPM H 2016 NRC INITIAL EXAM

           -                           RO LESSON TITLE:          Test the Main Steam Isolation Valves LESSON NUMBER: LOT-SIM-JP-025-A04 REVISION NO:           0
             %o Sodet PREPARER I DATE 9o                                      9//5/O/5 TECHNICAL REVIEWER I DATE u44:   hO2d                             9/1112015 Vezeá Pccáez                           9/10/2015 VALIDATOR I DATE Pie7ce                            9/24/20/5 LINE SUPERVISOR I DATE Qct                                     9/25/20/5 TRAINING SUPERVISION APPROVAL I DATE LOT-SI M-J P-025-A04                     Page 1 of 10           Rev 0

RELATED TASKS: 239201 B201, Test Main Steam Isolation Valves per OPT-40.2.7 K/A REFERENCE AND IMPORTANCE RATING: 239001 A4.01 4.2/4.0 Ability to manually operate and/or monitor the MSIV5 in the Control Room

REFERENCES:

OPT-40.2.7, Testing of Main Steam Line Isolation Valves After Maintenance OPT-40.2.8, Main Steam Isolation Valve Closure Test TOOLS AND EQUIPMENT: Stop Watch SAFETY FUNCTION (from NUREG 1123): 3 Pressure Control SIMULATOR SETUP Initial Conditions: Reactor power <50 RTP% Place Feedwater Control Mode Select switch in 1 -ELEM per 20P-32. LOT-SIM-JP-025-A04 Page 2 of 10 Rev 0

SAFETY CONSIDERATIONS: None EVALUATOR NOTES: (Do not read to pertormer)

1. The applicable procedure section WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM may be performed on Unit 2.

4. Critical Step Basis

1. Prevents Task Completion

2. May Result in Equipment Damage

3. Affects Public Health and Safety

4. Could Result in Personal Injury

5. Provide copy of OPT-40.2.7, Acceptance Criteria, Prerequisites, Section 6.2, and Attachment 2, Post Maintenance Testing B21-F022A (Inboard MSIV A Vlv)

Read the following to the ]PM performer. TASK CONDITIONS:

1. Unit Two startup is in progress following a forced outage to repair MSIV 2B21-F022A, Inboard MSIV A Valve.

2. Conditions are such that steam flow can be stopped in the main steam line of the MSIVs being tested.

3. It is not required to stop steam flow in MSL A to perform the slow closure test of B21 -

F022A, Inboard MSIV A Valve.

4. No other tests or maintenance activities are in progress that could provide a half scram signal to the RPS logic.

5. Another operator has placed Feedwater Control Mode Select switch in 1 -ELEM per 20P-32, Condensate and Feedwater System Operating Procedure.

INITIATING CUE: You are directed by the Unit CRS to perform OPT-40.2.7, Testing of Main Steam Isolation Valve after Maintenance, for MSIV 2B21 -F022A, Inboard MSIV A Valve ONLY and inform the CRS if the stroke time meets the acceptance criteria. LOT-SIM-JP-025-A04 Page 3 of 10 Rev 0

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. Step 1 - Perform take a minute at job site prior to beginning task. Examinee should cover the following questions, as deemed necessary. What are the hazards in the area? What PPE is required? Tools/PPE inspected prior to use? Energy sources secured/isolated? Is Clearance/Tag Out sufficient? Whats the worst that can happen? Any ALARA concerns? Will I affect plant status? HU Tools needed? SAT/UNSAT TIME START: NOTE: The examinee should be provided a copy of OPT-40.2.7, Testing of Main Steam Isolation Valve after Maintenance, and given time to review and pre-mark appropriate sections. I PROMPT If asked, a Reactivity Management Team is in place for this test. I Step 2 Confirm Reactor power is less than 55% RTP Confirmed power less than 55% RTP. SAT/U NSAT Step 2a Confirm conditions are such that steam flow can be stopped in the main steam line of the MSIV being tested Confirms steam flow can be stopped in the A Main Steam Line. SAT/UNSAT PROMPT If asked, No other tests or maintenance activities are in progress that could provide a half scram signal to the RPS logic. Step 3 Confirm all MSIVs are open. Confirmed all MSIVs are open. SAT/UNSAT Step 4 Confirm Reactor Recircutation system is NOT in single loop operation (SLO) Confirmed Reactor Recirculation system not in single loop. SAT/U NSAT LOT-SIM-JP-025-A04 Page 4 of 10 Rev 0

NOTE: Have stop watch ready to give to Examinee. Step 5 Obtain a stopwatch and record calibration information. Stop watch obtained and calibration information recorded. SAT/UNSAT PROMPT If asked, As the CRS grant permission to perform the test. Step 5a Ensures all prerequisites are met. Verifies all steps in Section 5.0 are met. SAT/UNSAT Step 5b Ensures Feedwater Control Mode Select switch, in 1 -ELEM per 20P-32, Condensate and Feedwater System Operating Procedure. Verifies Feedwater Control Mode Select control switch is in 7 ELEM. SAT/U NSAT NOTE: IF AT ANY TIME while performing this test in MODE 1, annunciator A-05, 4-6, Main Steam Isol Vlv Not Full Open, is received, THEN suspend this test and determine its cause. Step 6 Ensure the following annunciators are clear:

• A-05, 4-6, Main Steam Isol Vlv Not Full Open
• A-05, 1-7, Reactor Auto Scram Sys A
• A-05, 2-7, Reactor Auto Scram Sys B Annunciators confirmed to be clear.

SAT/U NSAT NOTE: When this test is performed in MODE 1, reactor pressure, power level, and steam flow are monitored while closing the MSIVs. Any deviation from expected plant response is cause for suspension of this test and notification of the Unit CRS prior to proceeding. PROMPT It is NOT required to stop steam flow in Main Steam Line A. NOTE: Performer should NA step 6.2.2. PROMPT It IS required to perform slow closure (spring closure) test of B21-F022A. LOT-SIM-JP-025-A04 Page 5 of 10 Rev 0

Step 7 Depress and hold B21 -F022A (Inboard MSIV A Test) pushbutton until the valve goes CLOSED, approximately 45-60 seconds. B27-F022A (Inboard MSIVA Test) pushbutton depressed and held until the valve is CLOSED, green light on, red light off.

                                                             **CRITICAL STEP** SAT/UNSAT Step 8- Release B21-F022A (Inboard MSIV A Test) pushbutton and confirm the valve goes OPEN Pushbutton for B27-F022A released and valve open confirmed.
                                                             **CRITICAL STEP** SAT/UNSAT PROMPT If asked, stroke time testing is required.

NOTE: Operation with both MSIVs closed in a main steam line is minimized to reduce the severity of differential pressure transients when reopening the Outboard MSIV. Step 9 - Perform stroke time test as follows:

a. Ensure B21-F022A (Inboard MSIV A Vlv) OPEN.

B27-F022A verified open. SAT/UNSAT

b. Close B21-F022A (Inboard MSIV A Vlv) utilizing the pistol grip switch.

B27-F022A pistol grip switch taken to close.

                                                             **CRITICAL STEP** SAT/UNSAT

c. Record stroke time:

Stroke time recorded. SAT/U NSAT

d. Enter the measured stroke time from Section 6.2 Step 4.c and calculate the corrected stroke time (Stroke Time from Section 6.2, Step 4.c X 1.1 = Corrected Stroke Time)

Corrected stroke time calculated

                                                             **CRITICAL STEP** SAT/UNSAT LOT-SIM-JP-025-A04                           Page 6 of 10                                Rev 0

e. Record corrected stroke time on Attachment 1 or Attachment 2 Corrected Stroke Time recorded on Attachment 2 SAT/U NSAT NOTE: Step 6.2.5 is NA, as the B21 -F028A was not closed previously.

PROMPT If asked, it is required by plant conditions to open B21 -F022A. Step 10 IF required by plant conditions, THEN open B21-F022A (Inboard MSIV A Vlv). B27-F022A pistol grip switch taken to open. SAT/UNSAT NOTE: Step 6.2.7 is N/A NOTE: Annunciator A-7, 4-2, FW Sys CtrI Trbl, may alarm. Step 11 Informs CRS that the stroke time for the Inboard MSIV A is SAT Determines from Attachment 2 that the stroke time for A MS/V is within the Acceptance Criteria. SAT/U NSAT PROMPT Inform Examinee that another operator will complete the Restoration section of the PT. TERMINATING CUE: When the 2B21 -F022A, Inboard MSIV A Valve, has been re-opened after testing and the CRS is notified that the stroke time meets the Acceptance Criteria of the PT this JPM is complete. TIME COMPLETED: COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. LOT-SIM-JP-025-A04 Page 7 of 10 Rev 0

Step Critical I Not Critical Reason 1 Not Critical Administrative 2-6 Not Critical Verification of initial conditions and pre-requisites. 7-8 Critical Required actions to complete the test. 9a Not Critical Verification step. 9b Critical Action required to complete the test. 9c Not Critical Recording time not critical to test completion. 9d Critical Calculation of Corrected Stroke Time required to complete task. 9e Not Critical Recording required information. 10 Not Critical Re-opening valve not required to obtain results. REVISION

SUMMARY

0 NewJPM. LOT-SIM-JP-025-A04 Page 8 of 10 Rev 0

Validation Time: 15 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: 2 Setting: In-Plant Simulator X Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: D Comments reviewed with Performer Evaluator Signature: Date: LOT-SIM-JP-025-A04 Page 9 of 10 Rev 0

TASK CONDITIONS:

1. Unit Two startup is in progress following a forced outage to repair MSIV 2B21 -F022A, Inboard MSIV A Valve.

2. Conditions are such that steam flow can be stopped in the main steam line of the MSIVs being tested.

3. It is not required to stop steam flow in MSL A to periorm the slow closure test of B21 -

F022A, Inboard MSIV A Valve.

4. No other tests or maintenance activities are in progress that could provide a half scram signal to the RPS logic.

5. Another operator has placed Feedwater Control Mode Select switch in 1-ELEM per 20P-32, Condensate and Feedwater System Operating Procedure.

INITIATING CUE: You are directed by the Unit CRS to perform OPT-40.2.7, Testing of Main Steam Isolation Valve after Maintenance, for MSIV 2B21-F022A, Inboard MSIV A Valve ONLY and inform the CRS if the stroke time meets the acceptance criteria. Page 10 of 10

DUKE ENERGY Continuous Use BRUNSWICK UNIT 0 SURVEILLANCE TEST PROCEDURE OPT-40.2.7 TESTING OF MAIN STEAM ISOLATION VALVES AFTER MAINTENANCE REVISION 16 Special Considerations: This is a Reactivity Management Procedure.

TESTING OF MAIN STEAM ISOLATION VALVES OPT-40.2.7 AFTER MAINTENANCE Rev. 16 Page 2of39 REVISION

SUMMARY

PRR 635473 DESCRIPTION Revision 16 Editorial Revision to capitalize component positions, make consistent use of emphasis techniques, make component descriptions title case, update placekeeping aids, revise Steps in Section 6.5 for Acceptance Criteria review and update Attachment 4 to latest version of the Certification/Review form. Revised by Jim McCrary

TESTING OF MAIN STEAM ISOLATION VALVES OPT-40.2.7 AFTER MAINTENANCE Rev 16 Page 3 of 39 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE 4 2.0 SCOPE 4 3.0 PRECAUTIONS AND LIMITATIONS 4 4.0 ACCEPTANCE CRITERIA 5 5.0 PREREQUISITES 7 6.0 INSTRUCTIONS 7 6.1 General 7 6.2 Post Maintenance Testing B21-F022A (Inboard MSIV A VIv) 8 6.3 Post Maintenance Testing B21-F028A (Outboard MSIV A VIv) 11 6.4 Post Maintenance Testing B21-F022B (Inboard MSIV B VIv) 14 6.5 Post Maintenance Testing B21-F028B (Outboard MSIV B VIv) 17 6.6 Post Maintenance Testing B21-F022C (Inboard MSIV C VIv) 20 6.7 Post Maintenance Testing B21-F028C (Outboard MSIV C VIv) 23 6.8 Post Maintenance Testing B21-F022D (Inboard MSIV D VIv) 26 6.9 Post Maintenance Testing B21-F028D (Outboard MSIV D VIv) 29 6.10 Restoration 31 7.0 RECORDS 33

8.0 REFERENCES

33 ATTACHMENT 1 Unit 1 Nuclear Steam Supply System Valves Data 36 2 Unit 2 Nuclear Steam Supply System Valves Data 37 3 Valve Position Restoration Table 38 4 Certification And Review Form 39

TESTING OF MAIN STEAM ISOLATION VALVES OPT-40.2.7 AFTER MAINTENANCE Rev 16 Page4of39 1.0 PURPOSE This test demonstrates the OPERABILITY of the main steam isolation valves, MSIVs, after maintenance and provides direction for the slow closure test of MSIVs. 2.0 SCOPE

1. This test demonstrates each MSIVs ability to full stroke within the stroke times specified in Unit 1 (Unit 2) Technical Specifications SR 3.6.1.3.5. This satisfies the 1ST requirement in Technical Specification 5.5.6.

2. This test checks the MSIV Slow Closure function described in UFSAR Sections 5.4.5 and 7.3.1 .1 .5.

3. This test does NOT provide instructions for stroke adjustments subsequent to testing.

3.0 PRECAUTIONS AND LIMITATIONS

1. When isolating and unisolating a steam line in MODE 1 or 2 a small pressure change may occur causing a reactivity change. This reactivity change is classified as a Reactivity Manipulation (R2) per OPS-NGGC-1 306, Reactivity Management Program.

2. If this test is being performed in MODE 1, the RPS System will receive a partial trip signal that will NOT be annunciated as long as the remaining MSIVs are in the open position.

3. Annunciator A-05, 4-6, Main Steam Isol VIv Not Full Open, may alarm when a main steam line is isolated. This annunciator is received only when two or more MSIVs are closed.

4. Operation with both MSIVs closed in a main steam line is minimized to reduce the severity of differential pressure transients when reopening the Outboard MSIV. The section of pipe between the inboard and outboard MSIV will depressurize as it cools down or if any steam leaks are present (such as stem packing leak).

TESTING OF MAIN STEAM ISOLATION VALVES OPT-40.2.7 AFTER MAINTENANCE Rev 16 Page 5 of 39 3.0 PRECAUTIONS AND LIMITATIONS (continued)

5. An administrative band of 3.6 seconds to 4.4 seconds is applied when in MODE 2 or 3 due to temperature affects on stroke time. This is an administrative limit, and is NOT controlled by the 1ST program. More detail concerning these administrative limits is available in Section 8.7 Miscellaneous Document 3, EC# 86807, Evaluation of MSIV Stroke Time Criteria. If the corrected stroke time is satisfactory, but outside the Administrative range, it is to be adjusted to within the Administrative range per the applicable CHANNEL CALIBRATION.

4.0 ACCEPTANCE CRITERIA This test may be considered satisfactory with the successful completion of this procedure. NOTE This test demonstrates the slow closure design base function of the MSIVs described in UFSAR Sections 5.4.5 and 7.3.1.1.5. This function requires MSIVs to close on spring pressure alone. Slow closure is NOT a safety function D

1. Slow Closure Test

a. When an MSIV is given a close signal from the Control Room test pushbutton, the valve goes to the CLOSED position.

NOTE

• The Valve Stroke time test satisfies Unit 1 (Unit 2) Technical Specifications SR 3.6.1.3.5 and partially satisfies the 1ST requirement in Technical Specification 5.5.6 D
• Stroke time is measured from the time the control switch is repositioned to the time the valve is fully stroked by light indication
• For MSIVs, the measured stroke times are multiplied by a correction factor of 1 .1 to compensate for the position settings of the indicating light sensors of 10% and 100% open D

2. Valve Stroke Time

a. Corrected stroke times are within the Limiting range as specified by the minimum and maximum stroke times shown on Attachment 1, Unit 1 Nuclear Steam Supply System Valves Data or Attachment 2, Unit 2 Nuclear Steam Supply System Valves Data

TESTING OF MAIN STEAM ISOLATION VALVES OPT-40.2.7 AFTER MAINTENANCE Rev 16 Page 6 of 39 4.0 ACCEPTANCE CRITERIA (continued)

b. For tests where the corrected stroke time of the valve is less than the minimum or greater than the maximum Limiting stroke time or the valve disc or stem fail to exhibit the required change of position, the valve shall immediately be declared INOPERABLE.

NOTE This test partially satisfies the 1ST requirement in Technical Specification 5.5.6 D

3. Valve Fail-Safe Testing

a. The fail-safe test is considered satisfactory when the control switch is placed in the CLOSED position for fail-closed valves or OPEN position for fail-open valves, and the valve changes position in response to control switch movement.

TESTING OF MAIN STEAM ISOLATION VALVES OPT-40.2.7 AFTER MAINTENANCE Rev 16 Page 7of39 5.0 PREREQUISITES

1. Confirm Reactor power is less than 55% RTP

2. Confirm conditions are such that steam flow can be stopped in the main steam line of the MSIV being tested or steam flow exists

3. IF unit is in MODEl, THEN confirm the following:

             .       NO other tests or maintenance activities are in progress that could provide a half scram signal to the RPS logic
             .       All main steam isolation valves are OPEN

4. Confirm the Reactor Recirculation system is NOT in single loop operation (SLO)

5. Obtain a stopwatch and record information TEST EQUIPMENT Item ID No. Cal Date Cal Due Date Stopwatch 6.0 INSTRUCTIONS 6.1 General

1. Request permission from the Unit CRS to perform this test

2. Ensure all prerequisites in Section 5.0 are met

3. Ensure Feedwater Control Mode Select switch, in 1-ELEM per 1 OP-32 (20P-32) Condensate and Feedwatet System Operating Procedure

4. IF AT ANY TIME while performing this test in MODE 1 annunciator A-05, 4-6, Main Steam Isol Vlv Not Full Open, is received, THEN suspend this test and determine its cause

TESTING OF MAIN STEAM ISOLATION VALVES OPT-40.2.7 AFTER MAINTENANCE Rev 16 Page 8 of 39 NOTE When isolating and unisolating a steam line in MODE I or 2 a small pressure change may occur causing a reactivity change. This reactivity change is classified as a Reactivity Manipulation (R2) per OPS-NGGC-1 306, Reactivity Management Program D 6.2 Post Maintenance Testing B21-F022A (Inboard MSIV A VIv)

1. IF unit is in MODEl, THEN ensure the following annunciators CLEAR:

• A-05, 4-6, Main Steam Isol Vlv Not Full Open
• A-05, 1-7, Reactor Auto Scram Sys A
• A-05, 2-7, Reactor Auto Scram Sys B CAUTION When this test is performed in MODE 1, reactor pressure, power level, and steam flow are monitored while closing the MSIVs. Any deviation from expected plant response is cause for suspension of this test and notification of the Unit CR5 prior to proceeding D BEGIN R.M. LEVEL R2 REACTIVITY EVOLUTION

2. IF it is required to stop steam flow in Main Steam Line A, THEN perform the following:

a. Depress and hold B21 -F028A (Outboard MSIV A Test) pushbutton until the valve is CLOSED

b. Place pistol grip switch for B21-F028A (Outboard MSIV A Vlv) in CLOSE

3. IF performing slow closure (spring closure) test of B21-F022A (Inboard MSIV A Vlv),

THEN perform the following:

a. Depress and hold B21-F022A (Inboard MSIV A Test) pushbutton until the valve goes CLOSED, approximately 45-60 seconds

b. Release B21-F022A (Inboard MSIV A Test) pushbutton and confirm the valve goes OPEN

TESTING OF MAIN STEAM ISOLATION VALVES OPT-40.2.7 AFTER MAINTENANCE Rev. 16 Page 9 of 39 6.2 Post Maintenance Testing B21 -F022A (Inboard MSIV A Vlv) (continued) CAUTION Operation with both MSIVs closed in a main steam line is minimized to reduce the severity of differential pressure transients when reopening the Outboard MSIV D

4. Perform stroke time test as follows:

a. Ensure B21-F022A (Inboard MSIVAVIv) OPEN

b. Close B21-F022A (Inboard MSIVAVIv) utilizing the pistol grip switch

c. Record stroke time Stroke Time Seconds

d. Enter the measured stroke time from Section 6.2 Step 4.c and calculate the corrected stroke time IV seconds X1.1= seconds Stroke Time from Corrected Stroke Time Section 6.2 Step 4.c

e. Record corrected stroke time on Attachment 1 or Attachment 2.

5. IF B21 -F028A (Outboard MSIV A Vlv) pistol grip switch was placed in CLOSE in Section 6.2 Step 2, THEN place pistol grip switch in OPEN and confirm the valve goes OPEN

6. IF required by plant conditions, THEN open B21-F022A (Inboard MSIVAVIv)

END R.M. LEVEL R2 REACTIVITY EVOLUTION

TESTING OF MAIN STEAM ISOLATION VALVES OPT-40.2.7 AFTER MAINTENANCE Rev. 16 Page 10 of 39 6.2 Post Maintenance Testing B21-F022A (Inboard MSIV A Vlv) (continued)

7. j all the following conditions are met:

• Corrected stroke time is within the Limiting range
• Corrected stroke time is outside the Administrative range
• The unit is in MODE 2 or 3 with the Drywell/MSIV Pit NOT accessible, THEN generate an CR to adjust the valve stroke time to within the Administrative range during the next outage

TESTING OF MAIN STEAM ISOLATION VALVES I OPT-40.2.7 AFTER MAINTENANCE Rev. 16 Page 37 of 39 ATTACHMENT 2 Page 1 of 1 Unit 2 Nuclear Steam Supply System Valves Data ADMINISTRATIVE ACCEPTANCE VALVE STROKE PROC STEP REMOTE POSITION FAIL-STROKE RANGE CRITERIA NUMBER DIRECTION SECTION INDICATION TIME (MODE2&3ONLY) LIMITINGVALUE REF. SAFE VALVE (INITIALS) TEST MINIMUM MAXIMUM MINIMUM t MAXIMUM STROKE TEST SAT/ I IND. () () () TIME UNSAT STEM (SEC) (INITIALS) I LIGHTS 1 INBOARD_______ 2-B21-F022A OPEN Section 6.2 Step 4 *

• 2-B21-F022A CLOSED 3.6 44 3.0 5.0 4.0 2-B21-F022B OPEN Section 6.4 Step 4 *
• 2-B21-F022B CLOSED 3.6 44 3.0 5.0 4.0 2-B21-F022C OPEN Section 6.6 Step 4 *
• 2-B21-F022C CLOSED 3.6 44 3.0 5.0 4.0 2-B21-F022D OPEN Section 6.8 Step 4 *
• 2-B21-F022D CLOSED 3.6 44 3.0 5.0 4.0 OUTBOARD 2-B21-F028A OPEN Section 6.3 Step 4 *
• 2-B21-F028A CLOSED 3.6 44 3.0 5.0 4.0 2-B21-F028B OPEN Section 6.5 Step 4 *
• 2-B21-F028B CLOSED 3.6 44 3.0 5.0 4.0 2-B21-F028C OPEN Section 6.7 Step 4 *
• 2-B21-F028C CLOSED 3.6 44 3.0 5.0 4.0 2-B21-FO2SD OPEN Section 6.9 Step 4 *
• 2-B21-F028D CLOSED 3.6 44 3.0 5.0 4.0 If corrected stroke time is within the Limiting range and outside the Administrative range and the unit is in MODE 2 or 3, an CR shall be generated to schedule and track performance of 2MST-RPS22(A-H)R to be completed during the next outage to ensure the valve stroke time is adjusted to within the Administrative range.

Performed by (Signature) Date Performed by (Signature) Date Reviewed, 1ST Group (Signature) Date

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE IP JPM I 2076 NRC INITIAL EXAM RO/ISRO LESSON TITLE: ALTERNATE COOLANT INJECTION LEP-Ol HEATER DRAIN PUMPS LESSON NUMBER: AOT-OJT-JP-300-J13 REVISION NO: 6 Dan HuIgin 8/18/16 PREPARER I DATE Bob Bolin 9/07/16 TECHNICAL REVIEWER I DATE Hunter Morris 9/07/16 VALIDATOR I DATE LINE S1PERVISOR I DTE 2 20/ TRAINING SUPERVISION APPROVAL I DATE AOT-OJT-JP-300-J13 Page 1 of 9 Rev.6

RELATED TASKS: 200072B504 Perform Alternate Coolant Injection With Heater Drain Pumps Per LEP-01. K/A REFERENCE AND IMPORTANCE RATING: 295031 AA1 .08 3.8/3.9 Ability to operate alternate injection system systems as they apply to Reactor Water Level Low.

REFERENCES:

OEOP-01 -LEP-01, ALTERNATE COOLANT INJECTION TOOLS AND EQUIPMENT: CR1 04P key for Unit Trip Load Shed Selector switch. SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): 2 Inventory Control AOT-OJT-JP-300-J13 Page 2 of 9 Rev.6

SAFETY CONSIDERATIONS:

1. Notify SM/CRS of JPM performance prior to commencing In-plant JPM.

2. Determine actual radiological conditions and potentially contaminated areas to achieve ALARA.

3. Ensure all electrical safety requirements are observed.

4. Review Work Practices section prior to conduct of the JPM.

5. DO NOT OPERATE any plant equipment during performance of this JPM.

EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee. OEOP-01 -LEP-01, will be provided to the examinee when asked for.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM will be performed on Unit br Unit 2.

4. Consider starting this JPM in the Control Room due to the need to obtain a CR1 04P key for Unit Trip Load Shed Selector Switch as well as for obtaining permission to enter the 4 KV Switchgear area in the Turbine Building

5. Critical Step Basis a) Prevents Task Completion b) May Result in Equipment Damage c) Affects Public Health and Safety U) Could Result in Personal Injury Read the following to the JPM performer.

TASK CONDITIONS:

1. A low Reactor Water level condition exists on Unit

2. The CRS is executing the Reactor Vessel Control Procedure (EOP-01 -RVCP)

3. RVCP directs use of Alternate Coolant Injection per EOP-O1-LEP-01.

4. RPV Pressure is 450 psig.

5. The main condenser is under vacuum.

INITIATING CUE: You are directed to perform the Auxiliary Operator actions for Alternate Coolant Injection, Heater Drain Pump Injection per EOP-01 -LEP-Ol, Section 2.2, and inform the Control Room when all required Auxiliary Operator actions are complete. AOT-OJT-JP-300-J13 Page 3 of 9 Rev.6

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. I Step 1 - Perform take a minute at job site prior to beginning task. Examinee should cover the following questions, as deemed necessary. What are the hazards in the area? What PPE is required? Tools/PPE inspected prior to use? Energy sources secured/isolated? Is Clearance/Tag Out sufficient? Whats the worst that can happen? Any ALARA concerns? Will I affect plant status? HU Tools needed? SAT I UNSAT TIME START PROMPT: Inform examinee that LEP-Ol, Section 2.2 Steps 1 through 3 have been completed. Step 2 Maintain level in the heater drain tank: Place feedwater heater level controllers to MAN and decrease the air signal to 0% to open the associated feedwater heater level control valves: HD-LC-75 (Feedwater Heater 4A Level Controller). Places HD-LC-75 Auto/Manual Selector to MAN. Adjust Controller output to 0% using Manual Control Unit Thumb wheel. HD-LC-75 in MAN with air signal at Q%* (critical*). CRITICAL STEP ** SAT I UNSAT Step 3 Maintain level in the heater drain tank: Place feedwater heater level controllers to MAN and decrease the air signal to 0% to open the associated feedwater heater level control valves: HD-LC-83 (Feedwater Heater 5A Level Controller). Places HD-LC-83 Auto/Manual Selector to MAN. Adjust Controller output to 0% using Manual Control Unit Thumb wheel. HD-LC-83 in MAN with air signal at Q%* (critical*). CRITICAL STEP ** SAT I UNSAT Step 4 Maintain level in the heater drain tank: Place feedwater heater level controllers to MAN and decrease the air signal to 0% to open the associated feedwater heater level control valves: HD-LC-79 (Feedwater Heater 4B Level Controller). Places HD-LC-79 Auto/Manual Selector to MAN. Adjust Controller output to 0% using Manual Control Unit Thumbwheel. HD-LC-79 in MAN with air signal at Q%* (critical*). CRITICAL STEP SAT I UNSAT AOT-OJT-JP-300-J13 Page 4 of 9 Rev.6

Step 5 Maintain level in the heater drain tank: Place teedwater heater level controllers to MAN and decrease the air signal to 0% to open the associated feedwater heater level control valves: HD-LC-87 (Feedwater Heater 5B Level Controller). Places HD-LC-87 Auto/Manual Selector to MAN. Adjust Controller output to 0% using Manual Control Unit Thumb wheel. HD-LC-87 in MAN with air signal at Q%* (critical*). CRITICAL STEP SAT I UNSAT Step 6 - Ensure HD-LC-91 (Heater Drain Deaerator Level Controller) in AUTO. Verifies HD-LC-97 in AUTO (Auto/Manual Selector in AUTO). HD-LC-97 in AUTO. SAT I UNSAT Step 7 Unit 2 Only: Ensure HD-LC-97 (Heater Drain Deaerator Level Controller) in AUTO. Verifies HD-LC-97 in AUTO (Controller Mode (Manual or Auto) A displayed on the controller). HD-LC-97in AUTO SAT/UNSAT PROMPT: When informed that AC actions for step 4 are complete, inform examinee that LEP-Ol, Section 2 Step 5 through 6 have been completed. Inform Examinee that Heater Drain Pump 1 (2)A is to be started for alternate coolant injection. NOTE: A CR1 04P key for Unit Trip Load Shed Selector Switch is located in the RO Desk locked drawer. A key can also be found in the Control room or WCC key lockers. Heater Drain Pump 1 (2)A Unit Trip Load Selector Switch is on BOP Bus 1 (2)D. Permission is required to enter the 4 KV Switchgear area in the Turbine Building. STEP 8a is to be performed if this JPM is performed on Unit 1. STEP 8b is to be performed if this JPM is performed on Unit 2. Step 8a Place Unit Trip Load Shed Selector Switch for heater drain pump to be started in DISABLED: At Bus 1 D, Row Hi, Compt AD8 (Htr Drain Pump 1A). Heater Drain Pump JA Unit Trip Load Selector Switch is placed in DISABLED. CRITICAL STEP SAT I UNSAT AOT-OJT-JP-300-Ji3 Page 5 of 9 Rev.6

Step 8b Place Unit Trip Load Shed Selector Switch for heater drain pump to be started in DISABLED: At Bus 2D, Row Ii, Compt AD8 (Htr Drain Pump 2A). Heater Drain Pump 2A Unit Trip Load Selector Switch is placed in DISABLED. CRITICAL STEP SAT I UNSAT Step 10 Inform Control Room AD Actions for Alternate Coolant Injection using Heater Drain Pump Injection are complete. Control Room informed AO actions per LEP-Ol, Section 2.2 are complete. SAT / UNSAT TERMINATING CUE: When AC Actions for Alternate Coolant Injection using Heater Drain Pump Injection are complete, this JPM is complete. TIME COMPLETED COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. Step Critical I Not Critical Reason 1 Not Critical Administrative 2 Critical Necessary for HDP alternate coolant injection 3 Critical Necessary for HDP alternate coolant injection 4 Critical Necessary for HDP alternate coolant injection 5 Critical Necessary for HDP alternate coolant injection 6 Non Critical Verification 7 Non Critical Verification 8a Critical Necessary for HDP alternate coolant injection (UNIT 1) 8b Critical Necessary for HDP alternate coolant injection (UNIT 2) 9 Critical Necessary for HDP alternate coolant injection 10 Critical Communication AOT-OJT-JP-300-J13 Page 6 of 9 Rev.6

REVISION

SUMMARY

6 NewJPM template Critical Step Delineation table added Renumbered steps Corrected procedure section Enhanced standards 5 Added basis for critical steps Minor format changes to cover/signature page Change SCO title to CRS AOT-OJT-JP-300-J13 Page 7 of 9 Rev.6

Validation Time: 10 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Periormance: Simulate X Actual Unit: 1/2 Setting: In-Plant X Simulator Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: U Comments reviewed with Performer Evaluator Signature: Date: AOT-OJT-JP-300-J13 Page 8 of 9 Rev.6

TASK CONDITIONS:

1. A low Reactor Water level condition exists on Unit

2. The CRS is executing the Reactor Vessel Control Procedure (EOP-O1 -RVCP)

3. RVCP directs use of Alternate Coolant Injection per EOP-O1 -LEP-Ol.

4. RPV Pressure is 450 psig.

5. The main condenser is under vacuum.

INITIATING CUE: You are directed to perform the Auxiliary Operator actions for Alternate Coolant Injection, Heater Drain Pump Injection per EOP-01 -LEP-Ol, Section 2.2, and inform the Control Room when all required Auxiliary Operator actions are complete. Page 9 of 9

DUKE ENERGY BRUNSWICK UNIT 0 EMERGENCY OPERATING PROCEDURE OEOP-O1 -LEP-Ol ALTERNATE COOLANT INJECTION REVISION 34 Special Considerations: EOP-Protected procedure Any revision to this procedure should be reviewed by an EOP writer.

ALTERNATE COOLANT INJECTION OEOP-O1 -LEP-Ol Rev. 34 Page 2 of 47 REVISION

SUMMARY

PRR 756734 DESC RI PT 10 N Revision 34 adds use of the torus suction to Section 2.5, RCIC Injection Using Manual Valve Operations and revises Section 2.10, RCIC Restoration to reflect same. Revised by Mike English.

ALTERNATE COOLANT INJECTION OEOP-01-LEP-01 Rev. 34 Page 3 of 47 TABLE OF CONTENTS SECTION PAGE 1.0 ENTRY CONDITIONS 5 2.0 INSTRUCTIONS 5 2.1 SLC Pump Demineralized/Fire Water Injection 5 2.1.1 ManpowerRequired 5 2.1.2 Special Equipment 5 2.1.3 SLC Pump Actions 5 2.2 Heater Drain Pump Injection 9 2.2.1 Manpower Required 9 2.2.2 Special Equipment 9 2.2.3 Heater Drain Pump Actions 9 2.3 RHR Loop B Injection 13 2.3.1 Manpower Required 13 2.3.2 Special Equipment 13 2.3.3 RHR Loop B Actions 13 2.4 Demineralized Water Injection 19 2.4.1 Manpower Required 19 2.4.2 Special Equipment 19 2.4.3 Demineralized Water Actions 19 2.5 RCIC Injection Using Manual Valve Operations 23 2.5.1 Manpower Required 23 2.5.2 Special Equipment 23 2.5.3 RCIC Actions 24 2.6 SLC Restoration 35 2.6.1 Manpower Required 35 2.6.2 Special Equipment 35 2.6.3 SLC Restoration Actions 35 2.7 Heater Drain Restoration 38 2.7.1 Manpower Required 38 2.7.2 Special Equipment 38 2.7.3 Heater Drain Restoration Actions 38 2.8 RHR Loop B Restoration 40 2.8.1 Manpower Required 40 2.8.2 Special Equipment 40 2.8.3 RHR Loop B Restoration Actions 40

ALTERNATE COOLANT INJECTION OEOP-O1 -LEP-Ol Rev. 34 Page 4 of 47 TABLE OF CONTENTS (continued) SECTION PAGE 2.9 Demineralized Water Restoration 43 2.9.1 Manpower Required 43 2.9.2 Special Equipment 43 2.9.3 Demineralized Water Restoration Actions 43 2.10 RCIC Restoration 45 2.10.1 ManpowerRequired 45 2.10.2 Special Equipment 45 2.10.3 RCIC Restoration Actions 45

ALTERNATE COOLANT INJECTION 0EOP-01-LEP-01 Rev. 34 Page 5 of 47 1.0 ENTRY CONDITIONS

• As directed by Emergency Operating Procedures (EOPs)
• As directed by Severe Accident Management Guidelines (SAMGs)
• As directed by Security Events, OAOP-40.0 2.0 INSTRUCTIONS 2.1 SLC Pump DemineralizedlF ire Water Injection 2.1.1 Manpower Required
• 1 Reactor Operator
• I Auxiliary Operator 2.1.2 Special Equipment
• RO Desk Locked Drawer o 1 LEP toolbox key (LSV-1) C
• Reactor Building 80 LEP Toolbox 0 1 pipe wrench C O 1 Demineralized Water to SLC Jumper Hose C O 1 Fire Hose to SLC Connector C 2.1.3 SLC Pump Actions

1. IF it has been determined the reactor will remain shutdown under all conditions without boron, THEN continue in this procedure C RO

2. jf directed to inject demineralized water, THEN:

a. Unlock and close C41-F001 (SLC Storage Tank Outlet Isolation Valve) C AO

ALTERNATE COOLANT INJECTION OEOP-O1 -LEP-Ol Rev. 34 Page 6 of 47 2.1.3 SLC Pump Actions (continued)

b. Remove cap and connect one end of the Demineralized Water to SLC Jumper Hose to the threaded connection at C41-V5000 (SLC Demineralized Water Supply Isolation Valve) C AC

c. Remove cap and connect the other end of the Demineralized Water to SLC Jumper Hose to the threaded connection upstream of C41-F014 (SLC Test Tank Outlet Demineralized Water Supply Isolation Valve) C AC

d. Unlock and open C41-V5000 (SLC Demineralized Water Supply Isolation Valve) C AC

e. Unlock and open 041 -F014 (SLC Test Tank Outlet Demineralized Water Supply Isolation Valve) C AC

3. jf demineralized water NOT available AND fire protection water available, THEN:

a. Unlock and close 041-FOOl (SLC Storage Tank Outlet Isolation Valve) C AC NOTE Fire hose stations are located on north wall of CRD rebuild room and south wall of Reactor Building 80 C

b. Disconnect the nozzle from the fire hose selected for supplying SLC C AC

ALTERNATE COOLANT INJECTION OEOP-O1-LEP-O1 Rev. 34 Page 7of47 2.1.3 SLC Pump Actions (continued) Figure 1, Fire Hose to SLC Connector

c. Remove cap and connect one end of the Fire Hose to SLC Connector to the threaded connection upstream of C41-F014 (SLC Test Tank Outlet Demineralized Water Supply Isolation Valve) Figure 1, Fire Hose to SLC Connector D AO

d. Connect the other end of the Fire Hose to SLC Connector to the fire hose selected in Section 2.1.3 Step 3.b AO

e. Ensure vent valve on the Fire Hose to SLC Connector CLOSED Q AO

f. Open the hose reel angle valve for the fire hose selected in Section 2.1.3 Step 3.b AO

g. Open vent valve on the Fire Hose to SLC Connector D AO

h. WHEN a stream of water issues from the valve, THEN close vent valve on the Fire Hose to SLC Connector AO

ALTERNATE COOLANT INJECTION OEOP-O1-LEP-O1 Rev. 34 Page 8 of 47 2.1.3 SLC Pump Actions (continued) Unlock and open C41-F014 (SLC Test Tank Outlet Demineralized Water Supply Isolation Valve) D AC

4. Start SLC Pumps A and B from the RTGB D RO

5. WHEN SLC pumps NOT required for RPV injection, THEN:

a. Stop SLC Pumps A and B D RD

b. Close C41-F014 (SLC Test Tank Outlet Demineralized Water Supply Isolation Valve) D AD

c. Exit this section and go to Section 2.6 0 RD

ALTERNATE COOLANT INJECTION OEOP-01-LEP-01 Rev. 34 Page9of47 2.2 Heater Drain Pump Iniection 2.2.1 Manpower Required

• I Reactor Operator
• 1 Auxiliary Operator 2.2.2 Special Equipment
• RO Desk Locked Drawer o 1 CR1 04P key for Unit Trip Load Shed Selector Switch D 2.2.3 Heater Drain Pump Actions

1. Ensure EW-FV-1 77 (Feedwater Recirc To Condenser Vlv)

CLOSED RO

2. Open FW-V13 (REP Bypass Vlv) D RO

3. Close:

• COD-V49 (REP A Suction Vlv)

RO

• COD-V50 (REP B Suction Vlv)

RO

4. Maintain level in the heater drain tank:

a. Place feedwater heater level controllers to MAN and decrease the air signal to 0% to open the associated feedwater heater level control valves:

(1) HD-LC-75 (Eeedwater Heater 4A Level Controller) D AO (2) HD-LC-83 (Feedwater Heater 5A Level Controller) D AO (3) HD-LC-79 (Feedwater Heater 4B Level Controller) C AO (4) HD-LC-87 (Eeedwater Heater 5B Level Controller) C AO

ALTERNATE COOLANT INJECTION OEOP-01-LEP-01 Rev. 34 Page 10 of 47 2.2.3 Heater Drain Pump Actions (continued)

b. Ensure HD-LC-91 (Heater Drain Deaerator Level Controller) in AUTO AC

c. Unit 2 Only: Ensure HD-LC-97 (Heater Drain Deaerator Level Controller) in AUTO D AC

d. IF main condenser under vacuum, THEN continue in this section at Section 2.2.3 Step 5 D RO

e. IF Condensate Transfer System available AND condenser NOT under vacuum, THEN:

NOTE RFA-Ll-56 (RFB-Ll-57) (REP Cond Level) indicator on XU-2 reading on-scale indicates level in the hotwell is sufficient to drain to the heater drain tank D (1) Notify Radwaste to make up to the hotwell to maintain hotwell level above +16 inches D RO (2) WHEN hotwell level above +16 inches, THEN open HD-V57 (Deaerator Fill & Drain Vlv) to drain the hotwell to the heater drain tank RO

5. Ensure CLOSED:

• FW-V118 (FW Htr4A Inlet VIv) D RO
• FW-V119 (FW Htr4B Inlet Vlv) D RO
• FW-V6 (FW Htr 5A Outlet VIv)

RD

ALTERNATE COOLANT INJECTION OEOP-O1-LEP-O1 Rev. 34 Page 11 of 47 2.2.3 Heater Drain Pump Actions (continued)

               . FW-V8 (FW Htr 5B Outlet Vlv)                                        0 RO
               . FW-V120 fEW Htrs 4 & 5 Byp Vlv)                                     0 RO NOTE Heater drain deaerator level must be greater than or equal to 48 inches to start a heater drain pump                                                                         0

6. Circle heater drain pump to be started 0 RO A B C

7. Unit I Only: Place Unit Trip Load Shed Selector Switch for heater drain pump to be started in DISABLED:

• At Bus 1D, Row Hi, ComptAD8 fHtr Drain Pump 1A) 0 AO
• At Bus IC, Row Ill, ComptAC3 (Htr Drain Pump IB) 0 AO
• At Bus ID, Row Di, ComptAD4 (Htr Drain Pump 10) 0 AO

8. Unit 2 Only: Place Unit Trip Load Shed Selector Switch for heater drain pump to be started in DISABLED:

• At Bus 2D, Row Ii, Compt ADS (Htr Drain Pump 2A) 0 AO
• At Bus 2C, Row MM1, Compt AC3 (Htr Drain Pump 2B) 0 AO
• At Bus 2D, Row Hi, ComptAD7 fHtr Drain Pump 20) 0 AO

ALTERNATE COOLANT INJECTION OEOP-O1-LEP-O1 Rev. 34 Page 12 of 47 2.2.3 Heater Drain Pump Actions (continued)

9. Start selected heater drain pump D RO

10. Place FW-UC-3269 (Startup Level Control Valve) in M (Manual) and open D RO

11. WHEN heater drain pump injection Q longer required, THEN exit this section and go to Section 2.7 RO

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE IP JPM J 2016 NRC INITIAL EXAM

         -                     - ROIISROIUSRO LESSON TITLE:        REMOTE SHUTDOWN PANEL SRV OPERATION LESSON NUMBER: LOT-OJT-JP-300-J25 REVISION NO:         0 Dan Hulqin                                   8/18/16 PREPARER I DATE Bob Bolin                                    9/07/16 TECHNICAL REVIEWER I DATE Grant Newton Hunter Morris                                9/07/16 VALIDATOR I DATE (Z

LINE SIfPERVISOR I DAVE 9- 2 ?c/ TRAINING SUPERVISION APPROVAL I DATE LOT-OJT-J P-300-J25 Page 1 of 10 Rev.0

RELATED TASKS: 239006B501 - Perform Remote Shutdown Panel SRV Operation per OEOP-01-LEP-05. KIA REFERENCE AND IMPORTANCE RATING: 295016 AAI.08 4.0/4.0 Ability to operate and/or monitor Reactor Pressure as it applies to Control Room Abandonment.

REFERENCES:

OEOP-01-LEP-05, REMOTE SHUTDOWN PANEL SRV OPERATION TOOLS AND EQUIPMENT: None SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): 7 Instrumentation LOT-OJT-JP-300-J25 Page 2 of 10 Rev.0

SAFETY CONSIDERATIONS:

1. Notify SM/CRS of JPM performance prior to commencing In-plant JPM.

2. Determine actual radiological conditions and potentially contaminated areas to achieve ALARA.

3. Ensure all electrical safety requirements are observed.

4. Review Work Practices section prior to conduct of the JPM.

5. DO NOT OPERATE any plant equipment during performance of this JPM.

EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee. 0EOP-01-LEP-05, will be provided to the examinee when asked for.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM will be performed on Unit br Unit 2.

4. Critical Step Basis a) Prevents Task Completion b) May Result in Equipment Damage c) Affects Public Health and Safety d) Could Result in Personal Injury Read the following to the JPM performer.

TASK CONDITIONS:

1. Unit is executing OEOP-01-EDP.

2. Reactor Pressure is 500 psig

3. SRV operation from the Control Room is not successful.

INITIATING CUE: You are directed by the Control Room Supervisor to rapidly depressurize the RPV by opening SRVs B, E, and G from the Unit Remote Shutdown Panel (RSDP) lAW OEOP-01 -LEP-05. LOT-OJT-JP-300-J25 Page 3 of 10 Rev.0

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. Step I - Perform take a minute at job site prior to beginning task. Examinee should covet the following questions, as deemed necessary. What ate the hazatds in the atea? What PPE is required? Tools/PPE inspected prior to use? Energy sources secured/isolated? Is Clearance/Tag Out sufficient? Whats the worst that can happen? Any ALARA concerns? Will I affect plant status? HU Tools needed? SAT I UNSAT TIME START NOTE: Special Equipment is located in the following areas: RO Desk Locked Drawer: I LEP toolbox key (LSV-1) ASSD Toolbox in Control Room: I sound-powered phone with extension cord for RD Reactor Building 20 LEP Toolbox 4T112 keys I sound-powered phone with extension cord for RD Step 2 Notify CRS: RTGB level indicators B21-Ll-R604B and C32-PR-R609 (N026B) will be lost. Notifies the CRS that level indicators B21-Ll-R604B and C32-PR-R609 (N026B) will be lost. SAT I UNSAT Step 3 Notify CRS: Control of SRV B, E and G from RTGB will be lost. Notifies the CRS that Control of SRV B, E and G from RTGB will be lost. SAT I UNSAT Step 4 Establish communication between RSDP and Control Room. Establishes communication with the Control Room. SAT I UNSAT LOT-OJT-JP-300-J25 Page 4 of 10 Rev.0

Step 5 At the Remote Shutdown Panel: Ensure B21-FO13E (Manual Relief E Vlv Close/Open) control switch in CLOSE. Verifies B27-FOI3E Close/Open control switch is in CLOSE. SAT I UNSAT Step 6 At the Remote Shutdown Panel: Place B21-FO13E (Manual Relief E Vlv Normal/Local) control switch in LOCAL. Inserts Key and places B27-FO73E Normal/Local control switch in LOCAL. B21-FO73E Normal/Local control switch in LOCAL* (critical*). CRITICAL STEP SAT I UNSAT Step 7 At the Remote Shutdown Panel: Ensure B21 -FOl 3G (Manual Relief E Vlv Close/Open) control switch in CLOSE. Verifies B27-FOI3G Close/Open control switch is in CLOSE. SAT I UNSAT Step 8- At the Remote Shutdown Panel: Place B21-FO13G (Manual Relief E Vlv Normal/Local) control switch in LOCAL. Inserts Key and places B21-FOI3G Normal/Local control switch in LOCAL. B21-FOI3G Normal/Local control switch in LOCAL* (critical*). CRITICAL STEP ** SAT I UNSAT Step 9 At the Remote Shutdown Panel: Ensure B21-FO13B (Manual Relief B Vlv Close/Open) control switch in CLOSE. Verifies B21-FO13B Close/Open control switch is in CLOSE. SAT I UNSAT Step 10 At the Remote Shutdown Panel: Place B21-FO13B (Manual Relief E Vlv Normal/Local) control switch in LOCAL. Inserts Key and places Normal/Local control switch in LOCAL. B21-FO13B Normal/Local control switch in LOCAL* (critical*). CRITICAL STEP ** SAT I UNSAT LOT-OJT-JP-300-J25 Page 5 of 10 Rev.0

Step 11 - At the Remote Shutdown Panel: Place B21 -CS-3345 (Reactor Water Level Normal/Local Switch) in LOCAL to transfer level transmitter B21-LT-N026B output to B21 -Ll-R6O4BX. Inserts Key and places B21-CS-3345 Normal/Local control switch in LOCAL. B21-FOI3B Normal/Local control switch in LOCAL. SAT I UNSAT PROMPT: If asked, CAC-LI-3342 (Supp Pool Level) is -2 ft. PROMPT: If asked, B27-Ll-R6O4BX (Reactor Water Level) is 150 inches and reading is valid lAW Caution 1. Step 12 - Confirm torus water level is greater than -8 feet. Verifies torus water level greater than -8 feet on CAC-LI-3342 (Supp Pool Level) SAT I UNSAT Step 13- Monitor RPV level. Monitors RPV level on B21-Ll-R6O4BX (Reactor Water Level). SAT I UNSAT NOTE: SRV B, E, and G can be opened in any sequence. Each SRV being opened is a critical step. PROMPT: Once SRVs are opened and if asked the status of pressure on C32-Pl-3332 (Reactor Pressure). State pressure is lowering on C32-Pl-3332 (Reactor Pressure) and is currently 350 psig. Step 14 - Monitor and control RPV pressure using SRVs B, E and G as directed by Control Room. Places OPEN/CLOSE control switches to OPEN for SRV SAT UNSAT B E G Verifies Reactor Pressure is lowering on C32-Pl-3332 (Reactor Pressure) CRITICAL STEP SAT I UNSAT LOT-OJT-J P-300-J25 Page 6 of 10 Rev.0

Step 15 - Notify the Control Room SRVs B, E, and G are open and reactor pressure is lowering. Notifies the control room that SRVs B, E, and G are open, and RPV pressure is 350 psig and lowering.: SAT I UNSAT TERMINATING CUE: When SRVs B, E, and G have been opened, and the control room has been notified this ]PM is complete. TIME COMPLETED COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. LOT-OJT-JP-300-J25 Page 7 of 10 Rev.0

Step Critical I Not Critical Reason I Not Critical Administrative 2 Non Critical Communication related 3 Non Critical Communication related 4 Non Critical Communication related 5 Non Critical Verification 6 Critical Necessary to open SRV 7 Non Critical Verification 8 Critical Necessary to open SRV 9 Non Critical Verification 10 Critical Necessary to open SRV 11 Non Critical Line up for parameter monitoring 12 Non Critical Parameter monitoring 13 Non Critical Parameter monitoring 14 Critical Necessary to depressurize RPV 15 Non Critical Communication related REVISION

SUMMARY

0 NewJPM LOT-OJT-JP-300-J25 Page 8 of 10 Rev.0

Validation Time: 10 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate X Actual Unit: 1/2 Setting: In-Plant X Simulator Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: D Comments reviewed with Performer Evaluator Signature: Date: LOT-OJT-JP-300-J25 Page 9 of 10 Rev.0

TASK CONDITIONS:

1. Unit is executing OEOP-O1-EDP.

2. Reactor Pressure is 500 psig

3. SRV operation from the Control Room is not successful.

INITIATING CUE: You are directed by the Control Room Supervisor to rapidly depressurize the RPV by opening SRVs B, E, and G from the Unit Remote Shutdown Panel (RSDP) lAW OEOP-01 -LEP-05. Page 10 of 10

4% DUKE ENERGY BRUNSWICK UNIT 0 EMERGENCY OPERATING PROCEDURE OEOP-OJ -LEP-05 REMOTE SHUTDOWN PANEL SRV OPERATION REVISION 0 Special Considerations: EOP-Protected procedure Any revision to this procedure should be reviewed by an EOP writer.

REMOTE SHUTDOWN PANEL SRV OPERATION OEOP-O1-LEP-05 Rev. 0 Page 2 of 10 REVISION

SUMMARY

PRR 709061 DESCRIPTION New procedure for operation of SRVs at the Remote Shutdown Panel (RSDP) and EC 99559. Written by Mike English.

REMOTE SHUTDOWN PANEL SRV OPERATION 0EOP-01-LEP-05 Rev. 0 Page 3 of 10 TABLE OF CONTENTS SECTION PAGE 1.0 ENTRY CONDITIONS 4 2.0 INSTRUCTIONS 4 2.1 SRV Operation 4 2.1.1 ManpowerRequired 4 2.1.2 Special Equipment 4 2.1.3 Operator Actions 4 2.2 Restoration 7 2.2.1 Manpower Required 7 2.2.2 Special Equipment 7 2.2.3 Operator Actions 7 ATTACHMENT I Caution I Information for B21 LI-R6O4BX

                                    -                           10

REMOTE SHUTDOWN PANEL SRV OPERATION 0EOP-O1-LEP-05 Rev. 0 Page4oflo 1.0 ENTRY CONDITIONS As directed by Emergency Operating Procedures (EOPs) 2.0 INSTRUCTIONS 2.1 SRV Operation 2.1.1 Manpower Required

• I Reactor Operator 2.1.2 Special Equipment
• RD Desk Locked Drawer O 1 LEPtoolboxkey(LSV-1) D
• ASSD Toolbox in Control Room O 1 sound-powered phone with extension cord for RD D
• Reactor Building 20 LEP Toolbox O 4TIl2keys 0 0 1 sound-powered phone with extension cord for RD 0 2.1.3 Operator Actions

1. Notify CRS:

• RTGB level indicators B21-Ll-R604B and C32-PR-R609 (N026B) will be lost 0 RD
• Control of SRV B, E and G from RTGB will be lost 0 RD NOTE ASSD Unit 1(2) Train B sound powered phone circuit is located on the RSDP U

2. Establish communication between RSDP and Control Room U RD

REMOTE SHUTDOWN PANEL SRV OPERATION 0EOP-01-LEP-05 Rev. 0 Page 5 of 10 2.1.3 Operator Actions (continued)

3. At the Remote Shutdown Panel:

a. Ensure B21-FO13E (Manual Relief E Vlv Close/Open) control switch in CLOSE D RO

b. Place B21-FO13E (Manual Relief E Vlv Normal/Local) control switch in LOCAL RO

c. Ensure B21-F01 3G (Manual Relief G Vlv Close/Open) control switch in CLOSE D RO

d. Place B21-FO13G (Manual Relief G Vlv Normal/Local) control switch in LOCAL RO

e. Ensure B21-FO13B (Manual Relief B Vlv Close/Open) control switch in CLOSE RO

f. Place B21-FOI3B (Manual Relief B Vlv Normal/Local) control switch in LOCAL RO

g. Place B21-CS-3345 (Reactor Water Level Normal/Local Switch) in LOCAL to transfer level transmitter B21-LT-N026B output to B21-Ll-R6O4BX D RO NOTE CAC-LI-3342 (Supp Pool Level) is on the RSDP

4. Confirm torus water level is greater than -8 feet D RD

REMOTE SHUTDOWN PANEL SRV OPERATION OEOP-01-LEP-05 Rev. 0 Page 6 of 10 2.1.3 Operator Actions (continued) NOTE . B21-Ll-R6O4BX (Reactor Water Level) is on the RSDP . Attachment 1 provides Caution 1 requirements for B21-Ll-R6O4BX CAUTION RSDP RPV level indicator B21-Ll-5977 is calibrated for hot condftions and includes significant loop uncertainties. Therefore it is not expected to be consistent with Control Room indication and B21-Ll-R6O4BX should be used D

5. Monitor RPV level RO NOTE C32-Pl-3332 (Reactor Pressure) is on the RSDP CAUTION

. SRV indication on the Remote Shutdown Panel is control switch position NOT valve position . Power to the SRV acoustic monitors comes from Sub E6(E8). If DC power is lost to the RTGB SRV control circuits, SRV position indication lights on Panel XU-73 should be available L1

6. Monitor and control RPV pressure using SRVs B, E and G as directed by Control Room i:

RO

7. WHEN SRV operation, from RSDP, is [Q longer required, THEN exit this section and go to Section 2.2 D RO

2.2 Restoration 2.2.1 Manpower Required

• 2 Reactor Operators 2.2.2 Special Equipment
• RO Desk Locked Drawer 0 1 LEP toolbox key (LSV-1) 2.2.3 Operator Actions

1. Record Unit number Unit Number

2. AttheRTGB:

a. Ensure B21-FO13B (Manual Relief B Vlv) control switch in CLOSE /

CV

b. Ensure B21-FO13E (Manual Relief E Vlv) control switch in CLOSE /

CV

c. Ensure B21-FO13G (Manual Relief G Vlv) control switch in CLOSE /

CV

3. At the Remote Shutdown Panel:

a. Ensure B21-FO13E (Manual Relief E Vlv Close/Open) control switch in CLOSE I____

IV

b. Place B21-FOI3E (Manual Relief E Vlv Normal/Local) control switch in NORMAL and remove key /

IV

c. Ensure B21-FO1 3G (Manual Relief G Vlv Close/Open Control Switch) in CLOSE /____

IV

d. Place B21-FO13G (Manual Relief G Vlv Normal/Local) control switch in NORMAL and remove key I____

IV

REMOTE SHUTDOWN PANEL SRV OPERATION OEOP-01-LEP-05 Rev. 0 Page 8 of 10 2.2.3 Operator Actions (continued)

e. Ensure B21-F013B (Manual Relief B VIv Close/Open) control switch in CLOSE I____

IV

f. Place B21-FO13B (Manual Relief B Vlv Normal/Local) control switch in NORMAL and remove key /____

IV

g. if EOP-01-LEP-04 NOT in progress, THEN:

(1) Place B21-CS-3345 (Reactor Water Level Normal/Local Switch) in NORMAL and remove key to transfer level transmitter B21-LT-N026B output to the RTGB / IV (2) Confirm RTGB level indicators consistent with plant conditions:

• B21-LI-R604B /____

IV

• C32-PR-R609 (N026B) /____

IV

h. Confirm CLOSED at RTGB:

• B21-F0138 (Manual Relief B VIv) /

IV

• B21-FO13E (Manual Relief E VIv) /

IV

• B21-FOI3G (Manual Relief G Vlv) /

IV

4. Restore headset to Control Room ASSD equipment tool box

5. Perform a Control Room ASSD tool box inventory

6. Restore EOP equipment to RO desk locked drawer

7. Perform a Control Room EOP inventory

8. Restore EOP equipment to the tool boxes

REMOTE SHUTDOWN PANEL SRV OPERATION OEOP-01-LEP-05 Rev. 0 Page 9 of 10 2.2.3 Operator Actions (continued)

9. Perform a LEP tool box inventory

10. Exit this procedure and continue in procedure(s) in effect Date/Time Completed Performed By (Print) Initials Reviewed By:

CRS

REMOTE SHUTDOWN PANEL SRV OPERATION OEOP-O1-LEP-05 Rev. 0 PagelOoflO ATTACHMENT I Page 1 of I Caution I Information for B21-LI-R6O4BX Reactor Saturation Limit 600 550 L

            !     500 450 D

400 C, 350 w 0 z 300 U U U 250 200 100 300 500 700 900 1,100 200 400 600 800 1,000 1,200 0 RPV PRESSURE (PSIG) INSTRUM ENT: Wide Range Level Instruments B21 -Ll-R6O4BX fB21 -LT-N026B) Indicating Range 0-210 Inches Cold Reference Leg CONDITIONS FOR USE:

1. Temperature on Reactor Building 50 is below 140°F (B21-XY-5948AA2-4, B21-XY-5948B A2-4, ERFIS Computer Point B211A102 or ERFIS Computer Point 821TA103)

D _Q

2. If reference leg area drywell temperature (RSDP CAC-TR-778, Point 1 or RTGB) in UNSAFE D

region of Reactor Saturation Limit, indicated level is greater than 20 inches OR If reference leg area drywell temperature (RSDP CAC-TR-778, Point 1 or RTGB) in SAFE region of Reactor Saturation Limit, indicated level is greater than 10 inches

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE IP JPM K 2016 NRC INITIAL EXAM

        -                        - ROIISROIUSRO LESSON TITLE:       Racking In E6 Cross-Tie with Breaker Charging Spring Failure LESSON NUMBER: AOT-OJT-JP-303-13 REVISION NO:        6 Dan Hulgin                                    8/18/16 PREPARER I DATE Bob Bolin                                     9/07/16 TECHNICAL REVIEWER I DATE Hunter Morris                                  9/07/16 VALIDATOR I DATE
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LINE S1PERVISOR I DMt TRAINING SUPERVISION APPROVAL I DATE AOT-OJT-JP-303-1 3 Page 1 of 11 Rev.6

RELATED TASKS: 262605B104 - Rack in a 480 Volt Electrically Operated Breaker (K-3000) per 1(2)OP-50. KIA REFERENCE AND IMPORTANCE RATING: 295003 AAI.01 3.7/3.8 Ability to Operate and/or Monitor AC Electrical Distribution System as it applies to a partial or complete loss of A.C. power.

REFERENCES:

OEOP-01-SBO-07, 480V E-bus Crosstie TOOLS AND EQUIPMENT: Racking tool for 480V Breakers Manual charging handle for 480V Breaker SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): 6 (Electrical Distribution) Page 2 of 11 Rev.6 AOT-OJT-JP-303-13

SAFETY CONSIDERATIONS:

1. Notify SM/CRS of JPM performance prior to commencing In-plant JPM.

2. Determine actual radiological conditions and potentially contaminated areas to achieve ALARA.

3. Ensure all electrical safety requirements are observed.

4. Review Work Practices section prior to conduct of the JPM.

5. DO NOT OPERATE any plant equipment during performance of this JPM.

EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee. OEOP-01-SBO-07, Attachment 1, will be provided to the exam inee when asked for.

2. Prior to the first ]PM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM will be performed on Unit 1.

4. Critical Step Basis a) Prevents Task Completion b) May Result in Equipment Damage c) Affects Public Health and Safety d) Could Result in Personal Injury Read the following to the ]PM performer.

TASK CONDITIONS:

1. A complete Loss of Offsite Power has occurred on both Units.

ed.

2. OEOP-01-SBO-07 is being executed, and Step 2.1.3.11 is ready to be perform

3. A Flex DG is NOT supplying E6.

4. 480v Crosstie breaker on E5 has been racked in.

INITIATING CUE: You are directed by the Reactor Operator to complete the Auxiliary Operator actions associated with cross-tying 480V Substation E5 to E6 lAW OEOP-01-SBO-07, rs are Step 2.1.3.11, and inform the Control Room when the E5 to E6 cross-tie breake ready to be closed. Page 3 of 11 Rev.6 AOT-OJT-JP-303-1 3

PERFORMANCE CHECKLIST required for any step evaluated as UNSAT. NOTE: Sequence is assumed unless otherwise indicated, comments Step 1 - Perform take a minute at job site prior to beginning task. ary. Exam/flee should covet the following questions, as deemed necess inspected What are the hazards in the area? What PPE is required? Tools/PPE prior to use? Energy sources secured/isolated? Is Cleara nce/Ta g Out Will I sufficient? Whats the worst that can happen? Any ALARA concerns? affect plant status? HU Tools needed? SAT I UNSAT T STAR TIME A 480 V racking tool is contained in the DG Building 23 LEP Toolbo x. NOTE: ted, but PROMPT: Inform Examinee that use of electrical safety equipment may be simula Examinee that the examinee should state the location of this equipment. Inform that electrical equipment compartments are NOT to be breached. r. NOTE: If requested, pictures will be provided of the internals of the 480 V breake breaker Step 2 At E6, Row Fl, rack in Compt AXJ (Tie Breaker To E5): Confirm locally OPEN. by the (Tie breaker to E5) Compt AXI on Bus E6 verified open as indicated green open flag. SAT I UNSAT PROMPT: If asked, inform the examinee that the locking hasp position is as seen. I IF necessary, THEN depress Step 3 At E6, Row Fl, rack in Compt AXI (Tie Breaker To E5): locking hasp to allow opening of racking shutter. Compt. Locking hasp DEPRESSED or verified to already be depressed on E6 AX1.

                                                                **CRITICAL STEP**           SAT / UNSAT Page 4 of 11                                             Rev.6 AOT-OJT-JP-303-l 3

Step 4 At E6, Row Fl, rack in Compt AX1 (Tie Breaker To E5):Rotate racking crank clockwise until breaker stops. Breaker Compt AX7 on Bus E6 stops in the CONNECT position (tacked in and shutter window closes when the racking tool is removed). CRITICAL STEP ** SAT I UNSAT PROMPT: As requested, inform the examinee that the closing springs failed to charge as indicated by lack of charging noise when toggle switch turned on and/or lack of spring charged indicator at front of breaker.

                            **ALTERNATE PATH BEGINS AT STEP 5**

toggle Step 5 At E6, Row Fl, rack in Compt AX1 (Tie Breaker To E5): Place Charging Power switch to ON, determine springs failed to charge, and Attachment 1, Manually Charging 480v Breaker Charging Springs is required. Charging power switch for E6 Compt AXI placed to the ON position, springs determined not charged, and Attachment I determined to be used. SAT I UNSAT NOTE: A manual charging handle for the 480 VAC cross-tie breaker springs is located in the DG Building 23 LEP Toolbox. Step 6 Place charging power toggle switch to OFF. (attachment 1) Charging power toggle switch is OFF (down position). SAT I UNSAT PROMPT: Provide a picture to the trainee to identify the location of the manual charging lever. The manual charging lever is located at the bottom middle of the 480 VAC breaker. The equipment enclosure should NOT be breached. Step 7 - Describe the action to open breaker door, and insert manual charging handle behind the breaker compartment door, using 480v breaker pictures. Manual charging handle is inserted in the breaker. CRITICAL STEP SAT I UNSAT Page 5 of 11 Rev.6 AOT-OJT-JP-303-13

d (clicks Step 8 (Simulate) Pump manual charging handle until closing springs are charge or. into position) and confirm charge is satisfactory by Springs Charged indicat d Closing springs are fully charged, as indicated by the yellow springs charge indication. CRITICAL STEP SAT I UNSAT door. Step 9 (Simulate) Remove manual charging handle and close compartment Manual charging handle removed, compartment door closed. SAT I UNSAT Step 10 Place charging power toggle switch to ON. Charging power toggle switch is ON (up position). SAT I UNSAT IN0TE: Step 2.1 .3.12 is N/A, a Flex DG is not supplying E6. Step 11 Inform control room that E5-E6 tie breakers are ready to be closed. Control room contacted and told E5-E6 tie breakers are ready to be closed SAT / UNSAT be closed, [PROMPT: When contacted as control room that E5-E6 tie breakers are ready to exam inee to stand clear so the breakers can be closed. inform springs charged and TERMINATING CUE: Substation E6 crosstie breaker is racked in, closing is ready to be closed then this JPM is complete. TIME COMPLETED RITY. COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECU Page 6 of 11 Rev.6 AOT-OJT-JP-303-1 3

Step Critical I Not Critical Reason I Not Critical Admin istrativ e 2 Non Critical Verify only. No action required. 3 Critical Required to complete task. 4 Critical Required to complete task. 5 Non Critical Since charging springs are not charged, turning power on accomplishes nothing. 6 Non Critical Places system in original configuration, but does completes action. 7 Critical Action required to complete task. 8 Critical Action required to complete task. 9 Non Critical Actions not required to accomplish task. 10 Non Critical Actions not required to accomplish task. 11 Non Critical Communicates results of actions. REVISION

SUMMARY

6 Changed 2.1.3.lOto2.1.3.11 due to procedure numbering change (non-technical change). Changed wording on steps to match procedure verbiage (non-technical change). 5 Changed Duke logo. Revised from OAOP-36.2 to OEOP-01-SBO-07 Added pictures of the 480v Breaker Page 7o1 11 Rev.6 AOT-OJT-JP-303-1 3

Validation Time: 12 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING X Actual Unit: Performance: Simulate X Simulator Admin Setting: In-Plant Yes No X Time Limit N/A Time Critical: Alternate Path: Yes X No EVALUATION Performer: JPM: Pass Fail Yes No Remedial Training Required: Comments: D Comments reviewed with Performer Date: Evaluator Signature: 8 of 11 Rev.6 AOT-OJT-JP-303-1 3 Page

UJO 6 O6ed S 4 I

s-- FT CAUTJON+ a. ___ a I I t t I  %..aa. a ti..k L11. I L

TASK CONDITIONS:

1. A complete Loss of Offsite Power has occurred on both Units.

2. OEOP-01-SBO-07 is being executed, and Step 2.1.3.11 is ready to be performed.

3. A Flex DG is NOT supplying E6.

4. 480v Crosstie breaker on E5 has been racked in.

INITIATING CUE: You are directed by the Reactor Operator to complete the Auxiliary Operator actions associated with cross-tying 480V Substation E5 to E6 lAW OEOP-O1-SBO-07, Step 2.1 .3.111 and inform the Control Room when the E5 to E6 cross-tie breakers are ready to be closed. Page 11 of 11

480V E-BUS CROSSTIE 0EOP-01 -SBO-07 Rev. 001 Page4of37 1.0 ENTRY CONDITIONS

• As directed by Emergency Operating Procedures fEOP5) 2.0 INSTRUCTIONS 2.1 Energizing E5 From E6 21.1 Manpower Required
• 1 Reactor Operator
• 1 Auxiliary Operator 2.1.2 Special Equipment
• RO Desk Locked Drawer 0 1 LEP toolbox key (LSV-1)
• 1 flashlight V
• Diesel Building 23 LEP Toolbox O 1 Racking tool for 480V breakers
• Unit I Control Building 23 Stairwell LEP Toolbox 0 1 rope lOft. to tie open battery room door 2.1.3 Energizing E5 From E6 Actions

1. IF a FLEX DG supplying E6, THEN at MCC ICA ensure OFF:

• Row Al, Compt C03 (Ventilation NC Cndsr 1 D 1-VA-i D-CU-CB) NIA AO
• Row A2, Compt Ci 7 (Control Bldg Floor Drain Sump Pmp 1-CB-1E-1) EN/A AD
• Row A3, Compt Ci 6 (CB HVAC Ctl Sys lnstr Air Compr 2-VA-2B-AC-CB) Q N/A AD

480V E-BUS CRCSSTIE OECP-0l-SBC-07 Rev. 001 Page 5of37 2.1.3 Energizing E5 From E6 Actions (continued)

• Row El, Compt C08 (Control Bldg Floor Drain Sump Pmp l-CB-1E-2) DN/A AC
• Row B4, Compt C28 (FP FilUDemin JA Holding Pump l-G41-Z00i-7A) D N/A AC
• Row Ci, Compt C04 (Ventilation NC Supply Fan 1 D 1-VA-ID-SF-CE) D N/A AC
• Row C2, Compt C18 (Ventilation Exhaust Fan 1A 1-VA-1A-EF-CB) D N/A AC
• Row C3, Compt C19 (Ventilation Supply Fan IA 1-VA-1A-SF-CB) DN/A AC
• Row C4, Compt C20 (Ventilation Exhaust Fan 10 1-VA-JC-EF-CB) D N/A AC
• Row D2, Compt C21 (Ventilation Supply Fan IC l-VA-1C-SF-CB) 0 N/A AC
• Row E2, Compt C24, (SAMA Diesel Generator #2 Supply) 0 N/A AC
• Row E4, Compt 012 (Fdr Bkr For Ltg DistrXfmr 1CEI i-ICEI-XFMR) 0 N/A AC
• Row Fl, Compt Cli (FUr Bkr For Cony Distr Pnl Xfmr 1 G-CB l-JG-CB-XFMR) ON/A AC
• Row F2, Compt C09 (Control Room Heating Coil 1-VA-i A-EHE-CB) 0 N/A AC

480V E-BUS CROSSTIE OEOP-0i-SBO-07 Rev. 001 Page6of37 2.1.3 Energizing E5 From E6 Actions (continued)

• Row F4, Compt dO (RPS MG Set 1A Control Panel i-C71-SOO1A) D N/A AD
• Row F5, Compt C06 (Batt Charger IA-2 AC Input Normal Feed i-IA-2-I25VDC-CHRGR) N/A AC
• Row F6, Compt C07 (Primary UPS IA, i-UPS-lA) D N/A AD
• Row F7, Compt C05 (Batt Charger lA-i AC Input Normal Feed I-1A-1-I25VDC-CHRGR) N/A AC
• Row G2, Compt C29 (Primary UPS 1A Heater) N/A AD
• Row G4, Compt C27 (CB Electric Unit Heater 1-VA-I A-UH-CB) N/A AC
• Row G5, Compt C32 (CB Cndsr Unit Exh Booster Fan i-VA-1A-BF-CB)

AD

• Row H3, Compt XX5 (Transformer 480/208/I 20V 9KVA 30 i-I CA-DIST-PNL-XFMR) N/A AC

2. At El, Row Ni, Compt AF9 (Nuc Serv Wtr Pmp 1A):

a. Ensure breaker CPEN AD

b. Remove normal control power fuses AD

c. Remove alternate control power fuses AC

3. At E5 open:

• Row A2, Compt AU6 (Feeder To MCC DGA) 121 AC

480V E-BUS CROSSTIE OEOP-0J -SBO-07 Rev. 001 Page 7of37 2.1.3 Energizing E5 From E6 Actions (continued)

• Row SI, ComptAT7 (MCC JOG)

AC

• Row B2, Compt AT8 (Feeder To MCC 2XA-2)

AC

• Row B3, ComptAT9 (MCC 1XC)

AC

• Row Cl, ComptAU4 (Feeder Breaker MCC IXA)

AC

• Row C2, Compt AUJ (MCC 2XJ) 1YJ AC
• Row C3, Compt AU2 (MCC JXE)

AC

• Row C4, Compt AU3 (MCC I PA) 121 AC
• Row E3, ComptAVO (MCC 1XG) 121 AC

4. IF DG 2 NOT available, THEN at E5, Row D2, open Compt AU5 (MCC IXL)

AC

5. IF a FLEX DG supplying E6, THEN go to Section 2.1.3 Step 8 D N/A RC

6. At E6 open:

• Row B3, ComptAV5 (MCC 1XH) 121 AC
• Row Cl, ComptAV6 (Distr Pnl ElO)

AD

• Row 02, Compt AV7 (Distr Pnl E9) 121 AC
• Row C3, Compt AV8 (Feeder To MOO 2XB-2)

AC

480V E-BUS CROSSTIE OECP-0l-SBO-07 Rev. 001 Page 8 of 37 2.1.3 Energizing E5 From E6 Actions (continued)

• Row Dl, ComptAWO (Feeder To MCC IXB)

AC

• Row D2, Compt AWl (MCC 2XK)

AC

• Row El, Compt AW4 (MCC 20G)

AC

• Row F2, ComptAW6 (MCC 1XD)

AC

7. IF DG 2 NOT available, THEN at E6 open:

• Row C4, Compt AV9 (MCC DGB)

AC

• Row D3, ComptAW2 (MCC 1XF)

AC

• Row D4, ComptAW3 (MCC JPB)

AC

• Row E2, ComptAW5 (MCC JXM)

AC

8. Place control switch for E5 feeder breakers to TRIP and confirm CPEN 121 RC

• Breaker AU9 (Sub E5 480V Main Breaker)

RC

• Breaker AF8 (Bus El To Sub E5) 121 RC

9. [ breaker AU9 (Sub E5 480V Main Breaker) will NOT open electrically, THEN: D N/A RC At E5, Row E2, Compt AU9 (Main Breaker):

a. Depress TRIP pushbutton on Compt AU9 N/A AC

480V E-BUS CROSSTIE OEOP-01-SBC-07 Rev. 001 Page 9 of 37 2.1.3 Energizing E5 From E6 Actions (continued)

b. Confirm ComptAU9 OPEN QN/A AC

10. At E5, RowAl, rack in ComptAT4 (Tie BreakerTo E6):

a. Confirm locally breaker OPEN 121 AC

b. IF necessary, THEN depress locking hasp to allow opening of racking shutter AC NOTE Breaker CONNECTED when racking shutter drops when racking tool removed CAUTION Crank NOT to be forced after breaker stops

c. Rotate racking crank clockwise until breaker stops AC

d. Place Charging Power toggle switch to ON 121 AC

e. Confirm charge satisfactory by SPRINGS CHARGED indicator 121 AC

1. IF closing springs fail to charge, THEN manually charge per Attachment 1 and return N/A AC

11. At E6, Row El, rack in Compt AX1 (Tie Breaker To E5):

a. Confirm locally breaker OPEN AC

b. IF necessary, THEN depress locking hasp to allow opening of racking shutter D AC

480V E-BUS CROSSTIE OEOP-01 -SBO-07 Rev. 001 Page 10of37 2.1.3 Energizing E5 From E6 Actions (continued) NOTE Breaker CONNECTED when racking shutter drops when racking tool removed C CAUTION Crank NOT to be forced after breaker stops C

c. Rotate racking crank clockwise until breaker stops C AC

d. Place Charging Power toggle switch to ON C AC

e. Confirm charge satisfactory by SPRINGS CHARGED indicator C AC

1. f closing springs fail to charge, THEN manually charge per Attachment 1 and return C AC NOTE Crosstie breakers will NOT close electrically if a FLEX DG supplying E6 C

12. IF a FLEX DG supplying E6, THEN go to Section 2.1.3 Step 14 C AC

13. Place and hold control switch for Bus E5 Tie To Bus E6 cross-tie breakers in CLCSE until both AT4 (Mstr) and AXJ (Slave) indicate CLOSED C RC

14. IF Breaker AT4 (Tie Breaker To E6) will NOT close electrically, THEN C RC At E5, Row Al, Compt AT4 (Tie Breaker To E6):

a. Lift Manual Close lever on Compt AT4 C AC

b. Confirm Compt AT4 CLOSED C AC

480V E-BUS CROSSTIE OEOP-01 -SBO-07 Rev. 001 Page 11 of 37 2.1.3 Energizing E5 From E6 Actions (continued)

15. IF breaker AXI (Tie Breaker To E5) will NOT close electrically, THEN C RO At E6, Row Fl, ComptAXi (Tie Breaker To ES):

a. Lift Manual Close lever on Compt AX1 C AC

b. Confirm ComptAXi CLOSED C AC

16. ffi either Tie Breaker was manually closed, THEN notify Control Room E5 ENERGIZED from E6 C AC

17. IF a FLEX DG supplying E6, THEN C RC

a. At MCC JCA place ON:

• Row F5, Compt C06 (Batt Charger 1A-2 AC Input Normal Feed 1-1 A-2-1 25VDC-CHRGR) C AC
• Row F7, Compt C05 (Batt Charger lA-i AC Input Normal Feed i-iA-i -1 25VDC-CHRGR) C AC
• Row A3, Compt C16 (CB HVAC Ctl Sys lnstr Air Compr 2-VA-2B-AC-CB) C AC
• Row C4, Compt C20 (Ventilation Exhaust Fan 1C i-VA-1C-EF-CB) C AC
• Row D2, Compt C21 (Ventilation Supply Fan 1C i-VA-IC-SF-CB) C AC

480V E-BUS CROSSTIE OEOP-01-SBO-07 Rev. 001 Pagel2of37 2.1.3 Energizing E5 From E6 Actions (continued)

b. W I B battery room vent fans NOT OPERATING, THEN -

RO (I) At MCC 1CB ensure ON:

• Row Bi, Compt C43 (Vent Supply Fan 1 B, 1B-SF-CB) D AC
• Row B2, Compt C42 (Vent Exhaust Fan 1 B, 1 B-EF-CB) D AC (2) Start 1B-SF-CB and 1B-EF-CB (Battery Room lB Vent Fans) D RO

18. Start 1C-SF-CB and 1C-EF-CB (Battery Room IA Vent Fans)

RO WARNING Batteries generate hydrogen gas when charging D

19. IF battery room vent fans NOT operating, THEN D RO

a. Tie open door CTB-DR-EL023-1 18 between 1A Battery Room and Cable Spread D AC

b. Notify TSC batteries JA-l and IA-2 charging without battery room ventilation D RO

20. Confirm for Battery Charger IA-l:

• BAT-AM-6000 (D.C. Amperes) above 10 amps D AC
• BAT-VM-6009 (D.C. Volts) 130 to 140 volts AC

480V E-BUS CROSSTIE OEOP-01 -SBO-07 Rev. 001 PageJ3of37 2.1.3 Energizing E5 From E6 Actions (continued)

21. Confirm for Battery Charger 1A-2:

• BAT-AM-6001 (D.C. Amperes) above 10 amps AO
• BAT-VM-6010 (D.C. Volts) 130 to 140 volts D AO

22. W battery chargers NOT ENERGIZED, THEN:

a. Notify Control Room D AO

b. Assess upstream electrical alignment D RO

c. Notify CRS of assessment results D RO

480V E-BUS CROSSTIE 0EOP-01-SBO-Rev. 001 Page 37 of 37 ATTACHMENT 1 Page 1 of I

                << Manually Charging 480v Breaker Charging Springs>>

1.0 MANPOWER REQUIRED

• 1 Auxiliary Operator 2.0 SPECIAL EQUIPMENT

       . RD Desk Locked Drawer G       1 LEP toolbox key (LSV-l)                                      D
       . I flashlight                                                           E1

• Diesel Building 23 LEP Toolbox 0 1 Manual charging handle for 480V breakers D 3.0 OPERATOR ACTIONS CAUTION Manually overcharging closing springs may cause breaker to bind D 0

1. Place Charging Power toggle switch to OFF 0 0 AD

2. Open breaker compartment door and insert manual charging handle 0 0 AO

3. Pump manual charging handle until closing springs charged (clicks into position) and confirm charge satisfactory by SPRINGS CHARGED indicator 0 0 AD

4. Remove manual charging handle and close breaker compartment door 0 0 AD

5. Place Charging Power toggle switch to ON 0 0 AD

6. Return to crosstie section 0 0 AD

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE LESSON TITLE: PERFORM SJAE OFF-GAS RADIATION MONITORS CHANNEL CHECK CALCULATION LESSON NUMBER: LOT-ADM-JP-201 -015 REVISION NO: 0 Daniel Hulqin 09/06/1 6 PREPARER I DATE Bob Bolin 09/06/16 TECHNICAL REVIEWER! DATE Kyle Cooper 09/06/16 Dwayne Wolf 09/06/16 Hunter Morris 09/06/16 VALIDATOR I DATE

      óLZ LINE     PERVISOR ID     E EEE7 TRAINING SUPERVISIOF APPROVAL I DATE

RELATED TASKS: 299201 B201 Perform Daily Surveillance Report Per 01-3.1 or 01-3.2 K/A REFERENCE AND IMPORTANCE RATING: GEN 2.1.25 3.9/4.2 Ability to interpret reference materials, such as graphs, curves, tables, etc.

REFERENCES:

201-03.2, Reactor Operator Daily Surveillance Report ODCM TOOLS AND EQUIPMENT: Student may use calculator SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): Generic (Administrative) SETUP INSTRUCTIONS None LOT-ADM-JP-201-D15 Page 2 of 9 Rev.0

SAFETY CONSIDERATIONS:

1. None EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. Critical Step Basis a) Prevents Task Completion b) May Result in Equipment Damage c) Affects Public Health and Safety d) Could Result in Personal Injury Read the following to the JPM performer.

TASK CONDITIONS:

1. Readings for D12-RM-K6O1A, SJAE Off Gas Rad MonitorA, and D12-RM-K6O1B, SJAE Off Gas Rad Monitor B, have been recorded on the Unit 2 Dayshift RODSR for Saturday 0630-1230.

2. Main Condenser Air Ejector is in operation.

3. HP has reported a local survey reading of 300 mR INITIATING CUE:

RO, and SRO candidates: You are directed by the Control Room Supervisor to complete item 108, SJAE Off-Gas Radiation Monitors Channel Check, of 201-03.2, Reactor Operator Daily Surveillance Report, and state the status of the channel check.

• Channel check is SAT
• Channel check is UNSAT SRO ONLY:

Based on the above information, determine the required actions, if any. LOT-ADM-JP-201 -Dl 5 Page 3 of 9 Rev.0

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. TIME START: Step I Record SJAE OFF-Gas Radiation Monitor readings from item 104 & 106 in Table 1. 452 for D12-RM-K601A recorded in SATbiock of table 1, and 224 for D12-RM-K6OIB recorded in SAT block of table 1. SATIU NSAT Step 2 Determine D12-RM-K6O1A is the highest reading, and divide by 2. Value for D72-RM-K607A divided by 2 determined to be 226. SATIUNSAT Step 3 Compare lower reading monitor to value in step 3. Determines D12-RM-K6OIB value of 224 is < value in step 3 (226). Determines the channel check is not yet satisfactory.

                                                                **CRITICAL STEP** SATIUNSAT Step 4  Contact E&RC health physics to obtain a local reading with an appropriate survey instrument.

Determines In formation from the E&RC survey is needed. SATIU NSAT Step 5 Record local survey instrument reading Records 300 in local survey instrument reading SAT block in attachment 1. SATIU NSAT Step 6 Multiply local survey instrument reading by 0.75 Determines local instrument times 0.75 is 225. SATIU NSAT LOT-ADM-JP-201 -Dl 5 Page 4 of 9 Rev.0

Step 7 Compare lower reading monitor to local survey results Determines D12-RM-K6OIB value of 224 is <0.75 of the local survey results (), and therefore, the channel check is unsatisfactory.

                                                          **CRITICAL STEP** SATIUNSAT TERMINATING CUE: When the results of the survey have been compared to D12-RM-K6OIB and the evaluation of the channel check has been made, this JPM is complete for RO candidates.

TIME COMPLETED: LOT-ADM-JP-201-D15 Page 5 of 9 Rev.O

SRO Candidates ONLY: Step 8 Determines the deviation is non-conservative, and instrument is declared inoperable. Determines that ODCM 7.3.2 Condition A is entered immediately, and ODCM Condition / requires the following actions:

1. Gaseous Radwaste Treatment System is immediately verified not bypassed

2. Main stack effluent noble gas monitor is immediately verified operable

3. Grab sample taken once within 72 hours and every 4 hours thereafter and analyzed to verify that the noble gas gross gamma activity rate is 243,600 pCi/second

4. Channel restored to operable within 30 days

                                                            **CRITICAL STEP** SATIUNSAT TERMINATING CUE: When evaluation of the channel check has been made, and ODCM TIME COMPLETED COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY.

LOT-ADM-JP-201-D15 Page 6 of 9 Rev.O

Step Critical I Not Critical Reason I Not Critical Documentation of data in block 2 Not Critical Math not documented 3 Critical Error would prevent correct channel check 4 Not Critical From task conditions 5 Not Critical Documentation of data in block 6 Not Critical Math not documented 7 Critical Error would prevent correct channel check 8 Critical Correct ODCM actions required REVISION

SUMMARY

0 NewJPM (Provide sufficient detail for reviewers and evaluators to understand the scope of any technical and/or administrative changes). LOT-ADM-JP-201-D15 Page 7 of 9 Rev.0

Validation Time: 15 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: 2 Setting: In-Plant Simulator Admin X Time Critical: Yes No X Time Limit N/A Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: D Comments reviewed with Performer Evaluator Signature: Date: LOT-ADM-JP-201-D15 Page 8 of 9 Rev.O

TASK CONDITIONS:

1. Readings for D12-RM-K6O1A, SJAE Off Gas Rad MonitorA, and D12-RM-K6O1B, SJAE Off Gas Rad Monitor B, have been recorded on the Unit 2 Dayshift RODSR for Saturday 0630-1230.

2. Main Condenser Air Ejector is in operation.

3. HP has reported a local survey reading of 300 mR INITIATING CUE:

RO, and SRO candidates: You are directed by the Control Room Supervisor to complete item 108, SJAE Off-Gas Radiation Monitors Channel Check, of 201-03.2, Reactor Operator Daily Surveillance Report, and state the status of the channel check.

• Channel check is SAT
• Channel check is UNSAT SRO ONLY:

Based on the above information, determine the required actions, if any. Page 9of9

DUKE ENERGY 4111) DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE LESSON TITLE: DETERMINE PRIMARY CONTAINMENT WATER LEVEL AND EVALUATE PCPL-A LESSON NUMBER: LOT-ADM-J P-300-BOO REVISION NO: 4 Daniel Hulgin 09/6/16 PREPARER I DATE Bob Bolin 09/6/16 TECHNICAL REVIEWER I DATE Hunter Morris 09/06/16 Kyle Cooper 09/06/16 VALIDATOR I DATE LINE PERVISORI ATE TRAINING SUPERVISION APPROVAL IDATE LOT-ADM-J P-300-B00 Page 1 of 11 Rev.4

RELATED TASKS: 200602B501 Determine Primary Containment water level per EOP-O1-UG K/A REFERENCE AND IMPORTANCE RATING: GEN2.1.7 4.4/4.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

REFERENCES:

OEOP-O1 -UG OAOP-36. 1 TOOLS AND EQUIPMENT: Student may use calculator SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): Administrative Conduct Of Operations SETUP INSTRUCTIONS None LOT-ADM-JP-300-BOO Page 2 of 11 Rev.4

SAFETY CONSIDERATIONS:

1. Notify SM/CRS of JPM performance prior to commencing In-plant JPM.

2. Determine actual radiological conditions and potentially contaminated areas to achieve ALARA.

3. Ensure all electrical safety requirements are observed.

4. Review Work Practices section prior to conduct of the JPM.

5. DO NOT OPERATE any plant equipment during performance of this JPM.

EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM will be performed Unit 2

4. This is an administrative JPM designed to be administered in any setting and may be administered to multiple candidates simultaneously in a classroom setting

5. Critical Step Basis a) Prevents Task Completion b) May Result in Equipment Damage c) Affects Public Health and Safety d) Could Result in Personal Injury Read the following to the JPM performer.

TASK CONDITIONS:

1. An accident is in progress on Unit Two. The Unit CRS is directing actions of EOP RVCP and EOP-02-PCCP.

2. 480 VAC Substation E7 is de-energized due to a fault. All other electrical buses are energized.

3. ERFIS is unavailable

4. See Attachment 1 for the Containment parameter readings that are available on the RTGB.

INITIATING CUE: You are directed to determine Primary Containment water level per EOP-01-UG, Attachment 36. Determine the current region of operation (Safe/Unsafe) on Primary Containment Pressure Limit A (PCPL-A) LOT-ADM-JP-300-B00 Page 3 of 11 Rev.4

PERFORMANCE CHECKLIST [OTE: Sequence is assumed unless otherwise indicated, comments required for any step evuated asUNSAT TIME START: Step 1 Determine suppression pool water level instruments cannot be used to determine primary containment water level since CAC-LI-2601-1 is above +2 feet and CAC-LR 2602 is powered from E7 and is de-energized. Determine suppression pool water level instruments cannot be used to determine primary containment water. SATIU NSAT Step 2 Determine suppression chamber pressure instruments cannot be used to determine primary containment water level since CAC-Pl-1257-2B is not less than 75 psig and CAC-Pl-1257-2A is powered from E7 and is de-energized Determine suppression chamber pressure instruments cannot be used to determine primary containment water. SAT/UNSAT Step 3 Determine primary containment water level should be calculated using CAC-Pl-1 230 and CAC-Pl-41 76 since both instruments have power and suppression chamber pressure is not less than 75 psig, determine CAC-PR-1257-1 is powered from E7 and should not be used. Determine primary containment water level should be calculated using CAC-PI 7230 and CA C-Pl-4 176. SAT/U NSAT Step 4 Calculate primary containment water level to be 2.3 ft/psi (72.5 67.5) + 28.5 ft. Primary containment water level calculated to be 40 feet.

                                                                 **CRITICAL STEP** SAT/U NSAT Step 5   Determine operation to be in the safe region of PCPL-A using the PCPL-A graph, calculated primary containment water level and CAC-Pl-41 76 for drywell pressure, (or by using CAC-PI-1 230 reading <70 psig).

Determine PCPL-A is in the Safe region

                                                                 **CRITICAL STEP** SATIUNSAT LOT-ADM-JP-300-B00                         Page 4 of 11                                            Rev.4

TERMINATING CUE: When primary containment water level is calculated, and PCPL-A is determined to be in the Safe region, this JPM is complete TIME COMPLETED: COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. Step Critical I Not Critical Reason 1 Not Critical JPM can still be completed without performing this step. 2 Not Critical JPM can still be completed without performing this step. 3 Not Critical JPM can still be completed without performing this step. 4 Critical Calculation required to complete this JPM. 5 Critical Determination required to complete this JPM. REVISION

SUMMARY

4 New template incorporated. Modified torus press o read slightly >75 psig in att 1 Removed take a minute (step 1) reordered steps. 3 Changed Unit SCO to Unit CRS. No technical changes. 2 Revised to new JPM Template, Revision 3. No technical changes (Provide sufficient detail for reviewers and evaluators to understand the scope of any technical and/or administrative changes). LOT-ADM-JP-300-BOO Page 5 of 11 Rev.4

Validation Time: 15 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: Setting: In-Plant Simulator Admin . Yes No X Time Limit N/A Time Critical: Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: D Comments reviewed with Performer Evaluator Signature: Date: Page 6 of 11 Rev.4 LOT-ADM-JP-300-BOO

TASK CONDITIONS: of EOP

1. An accident is in progress on Unit Two. The Unit CRS is directing actions 01-RVCP and EOP-02-PCCP.

electrical buses are

2. 480 VAC Substation E7 is de-energized due to a fault. All other energized.

3. ERFIS is unavailable.

available on the

4. See Attachment I for the Containment parameter readings that are RTGB.

INITIATING CUE: EOP-01-UG, You are directed to determine Primary Containment water level per on Primary Attachment 36. Determine the current region of operation (Safe/Unsafe) Containment Pressure Limit A (PCPL-A). Page 7 of Ii Rev.4 LOT-ADM-JP-300-B00

Attachment I sup ss K] N P3OL LEVEL C.CLR262 E3-E7 Page 8 of Ii

Attachment I DfELL PRESSURE cci 7l Page 9 of 11

Attachment I e 245 75 U 4 p p 1 95 C 2 H 50 A 0 M 1 45 B L 2 E V 4

                       -95 F                                       LJ E

E 6 T

                        -45 8

10 o SU PP DRYELL sU PP P30 L POOL LEVEL I PRES3IJRE PRESSI]RE CkCU2CC11 1 I CAD P14176 CACPI12572 E4 t4t5 E3-E7 Page lOof 11

Attachment I 75 /5 71 R: 69 67 F P P L E S I P R C C E 25 S S U

   =-            R GAUGE ZERO     t o               o___

x DECC AIER CE/EL INDK CD DRiitLL FRESS INDfDN 2CPi23G SUPRESiON ROL PRtIRt

                                )\CPI12%71B 2X Cli 21          UF Page 11 of 11

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE LESSON TITLE: CALCULATE DRYWELL LEAKAGE RATE LESSON NUMBER: LOT-ADM-J P-201 -Dl 6 REVISION NO: 0 DaneI Hulcin 09/06/16 PREPARER I DATE Bob Bolin 09/06/16 TECHNICAL REVIEWER I DATE Hunter Morris 09/06/16 Kyle Cooper 09/06/16 Dwayne Wolf 09/06/16 VALIDATOR I DATE LOR? TRAINING SUPERVISION APPROVAL I DATE Page 1 of 9 Rev.0 LOT-ADM-JP-201 -Dl 6

RELATED TASKS: K/A REFERENCE AND IMPORTANCE RATING: GEN2.2.12 3.7/4.1 Knowledge of surveil lance procedures.

REFERENCES:

201-03.2 T.S 3.4.4 TOOLS AND EQUIPMENT: Student may use calculator SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): Generic (Administrative) SETUP INSTRUCTIONS None Page 2 of 9 Rev.0 LOT-ADM-JP-201-D16

SAFETY CONSIDERATIONS:

1. None EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee.

ed in NUREG-1021,

2. Prior to the first JPM of the 1PM set, provide the 1PM briefing contain (s).

Appendix E, or similar briefing (for non-regulated exams) to the trainee

3. Critical Step Basis a) Prevents Task Completion b) May Result in Equipment Damage c) Affects Public Health and Safety d) Could Result in Personal Injury Read the following to the ]PM performer.

TASK CONDITIONS:

1. This task will be performed on Unit 2.

2. Unit2isinMODEl.

tor reading of 457620, and the

3. The Equipment Drain Sump was manually pumped to an integra pump stopped at 2000 on Sunday Nightshift.

of 13944891, and the pump

4. The Floor Drain Sump was manually pumped to an integrator reading stopped at 2000 on Sunday Nightshift.

INITIATING CUE: RO, and SRO candidates: the 24 hour leak rate for the You are directed by the Control Room Supervisor to determine drywell lAW Attachment 1, equipment and floor drains, and the 24 hour total leak rate to the Surveillance Report, for Drywell Leakage Calculation, of 201-03.2, Reactor Operator Daily Sunday Nightshift at time 2000. SRO ONLY: is NOT met, identify the State if Tech Spec LCO 3.4.4 is or is NOT met, and If the LCO LATEST time Unit 2 is required to be in MODE 3. Page 3 of 9 Rev.0 LOT-ADM-JP-201-D16

PERFORMANCE CHECKLIST ents required for any step evaluated as UNSAT. NOTE: Sequence is assumed unless otherwise indicated, comm TIME START: on Saturday. Step I Calculate time interval from equipment drain manual pump at 2000 y (20) from Subtracts time equipment drain sump pump stops on 2000 Sunda for a value of time equipment drain sump pump stops on 2000 Saturday (20) 1440 minutes.

                                                                  **CRITICAL STEP** SATIUNSAT manual pump at 2000 Step 2   Calculate difference in equipment drain integrator reading from on Saturday.

0) from 2000 Subtracts 2000 Saturday equipment drain integrator reading (42378 a value of 33840 gal.

Sunday equipment drain integrator reading (457620) for

                                                                  **CRITICAL STEP** SAT/UNSAT Step 3    Calculate 24 hour equipment drain leak rate step 2) for a Divides leakage (value from step 3) by time interval (value from value of 23.5 gpm.
                                                                   **CRITICAL STEP** SAT/UNSAT Saturday.

Step 4 Calculate time interval from floor drain manual pump at 2000 on (20) from time floor Subtracts time floor drain sump pump stops on 2000 Sunday 1440 minutes. drain sump pump stops on 2000 Saturday (20) for a value of

                                                                   **CRITICAL STEP** SAT/UNSAT l pump at 2000 on Step 5    Calculate difference in floor drain integrator reading from manua Saturday.

from 2000 Subtracts 2000 Saturday floor drain integrator reading (13942587) of 2304 gal. Sunday floor drain integrator reading (13944891) for a value

                                                                    **CRITICAL STEP** SATIUNSAT Page 4 of 9                                            Rev.0 LOT-ADM-JP-201-D16

Step 6 Calculate 24 hour equipment drain leak rate. from step 5) for a Divides leakage (value from step 6) by time interval (value value 017.6 gpm.

                                                            **CRITICAL STEP** SATIUNSAT Step 7  Calculate 24 hour total leak rate to drywell.

Adds value from step 4 and step 7 for a value of 25.7 gpm.

                                                            **CRITICAL STEP** SAT/UNSAT equipment drain, and 24 TERMINATING CUE: When the results for the 24 hour floor drain,                       is complete hour total leak rate to drywell have been recorded, this JPM for RO candidates.

TIME COMPLETED: Page 5 of 9 Rev.O LOT-ADM-JP-201-D16

SRO Candidates ONLY: is NOT met, identify Step 8 State if Tech Spec LCO 3.4.4 is or is NOT met, and If the LCO the latest time Unit 2 is required to be in MODE 3. T.S. 3.4.4 limit, and Determines that the 24 hour total leak rate has exceeded the Monday. entry into MODE 3 would be required no later than 7600 on

                                                               **CRITICAL STEP** SATIUNSAT I

made, and the earliest time to TERMINATING CUE: When evaluation of the T.S. has been MODE 3 has been determined, this JPM is complete for SRO candidates. TIME COMPLETED FOR EXAM SECURITY. COLLECT AND CONTROL ALL JPM EXAM MATERIALS Critical I Not Critical Reason Step 1 Critical Math Critical for JPM solution 2 Critical Math Critical for JPM solution 3 Critical Math Critical for JPM solution 4 Critical Math Critical for JPM solution 5 Critical Math Critical for JPM solution 6 Critical Math Critical for JPM solution 7 Critical Math Critical for JPM solution 8 Critical IS determination is Critical REVISION

SUMMARY

0 NewJPM rstand the scope of any (Provide sufficient detail for reviewers and evaluators to unde technical and/or administrative changes). Page 6 of 9 Rev.0 LOI-ADM-JP-201 -Dl 6

Validation Time: 15 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Actual X Unit: 2 Performance: Simulate Simulator Admin X Setting: In-Plant Yes No X Time Limit N/A Time Critical: Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Yes No Remedial Training Required: Comments: D Comments reviewed with Performer Date: Evaluator Signature: 7 of 9 Rev.O Page LOT-ADM-JP-201-D16

NOT GIVE TO STUDENTS EXAM KEY DO Al IAUMMtoNI I Page 130160 Report (RODSR) Unit 2 Reactor Operator Daily Surveillance ION DRYWELL LEAKAGE CALCULAT 01 04 20 MUAL ruMP EUPMENT DRAIN SuMP - TME PLM STDPS LOV LEVEL TRP, cACLLATE TIME INTERVAL IMNUTESI FROM MANUA_ FUMP A SAME TIME ON PREVIOUS CA E2CRCURRENTINTEGRTCRREADNG 01014(51 7t6t210 CADULATE DIFFERENCE IN INTEGRAOR READING 333D SOM MNJAL PUMP AT SAME ME ON REVOUO DAY EAKAGE CA.CULATE 24 -CUF EQUPMENT DRAIN LEAK RATE DIVIDE LEAKAGE BY IME NTERVALI MANUALL FUMP FLOOR DRAIN SUMP USING ONE Ri GO DUMP RECORD TIME PUMR STOPS LOV LEVEL Tn CACULATE TIME INTERVAL iMNUTESi FROM MANUA UMP AT SAME TIME ON PREVIOUS CA RECOPDIN1tGRATERREADINS 113191414161911 CA_CUL&TE DIFFERENCE IN INTEOPATER READING 2303 FROM MANUAL PUMP AT SAME TIME ON PREVCIUS )A EAIAGE CA.CULATE 24 NOUR FLOOR DRAIN LE PATE DI.iDE

 .EAKA3E STME INTERVALi 24 IQUP E0LIMEN DPA N LEAK RAE                  2350 24 NDUR FOCR DRfiN LEA RATE                      I 6 24 OLR OAL LEAK RATE TO DRYWE.                   25 1 ONEC- TEC   SPEC 344 LEA ADE      MET             unsat NOT           GIVE        TO        STUDENTS EXAM                 KEY             DO Page 8 of 9                        Rev.O LOT-ADM-JP-201-D16

TASK CONDITIONS:

1. This task will be performed on Unit 2.

2. Unit 2 is in MODE 1.

integrator reading of 457620, and the

3. The Equipment Drain Sump was manually pumped to an pump stopped at 2000 on Sunday Nightshift.

reading of 13944891, and the pump

4. The Floor Drain Sump was manually pumped to an integrator stopped at 2000 on Sunday Nightshift.

INITIATING CUE: RO, and SRO candidates: mine the 24 hour leak rate for the You are directed by the Control Room Supervisor to deter rate to the drywell lAW Attachment 1, equipment and floor drains, and the 24 hour total leak ator Daily Surveillance Report, for Drywell Leakage Calculation, of 201-03.2, Reactor Oper Sunday Nightshift at time 2000. SRO ONLY: LCO is NOT met, identify the State if Tech Spec LCD 3.4.4 is or is NOT met, and If the LATEST time Unit 2 is required to be in MODE 3. Page 9 019

ATTACHMENT I Page 12 of 60 rt (RODSR) Unit 2 Reactor Operator Daily Surveillance Repo DRYWELL LEAK AGE CALC ULAT ION 00 04 20 MANUALLY PUMP EQUIPMENT DRAIN SUMP RECORD flMEPUMP STOPS LOW LEVEL TRiP

                                                                .4        ,                      I A A(                        1440 CALCULATE TIME INTERVAL (MINUTES) FROM MANUAL                                                       u i 44u PUMP AT SAME TIME ON PREVIOUS DAY.

0J 0j4 J4 1616 16 RECORDCuRRENT1NTEGRATOR READING. ol 0f4 12 19 Ii 18 Ii 13

                                                -    010 1412131718 lo CALCULATE DIFFERENCE IN INTEGRATOR READING                                                                                   30672 FROM MANUAL PUMP AT SAME TIME ON PREVIOUS DAY                    30240                           30384 (LEAKAGEJ.

CALCULATE 24 HOUR EQUIPMENT DRAIN LEAK RATE (DIVIDE LEAKAGE BY TIME INTERVAL). VU 04 MANUALLY PUMP FLOOR DRAIN SUMP USING ONE U 4 A Afl 14 I A A ri I CALCULATE TIME INTERVAL (MINUTES) FROM MANUAL I I PUMP AT SAME TIME ON PREVIOUS DAY. RECORDINTEGRATORREADING. 1 l I 12 12 oj2 CALCULATE DIFFERENCE IN INTEGRATOR READING 131 I f2 15 18 l 1 I 19 l Ii 19 l 18 924 720 851 FROM MANUAL PUMP AT SAME TIME ON PREVIOUS DAY CALCULATE 24 HOUR FLOOR DRAIN LEAK RATE (DIVIDE LEAKAGE BY TIME INTERVAL). 24HOUREQUIPMENTDRAINLEAKRATE. 21 1 21.3 21 24HOURFLOORDRAINLEAKRATE.* - 0.5 0.59 0.64 24 HOUR TOTAL LEAK RATE TO DRYWELL. 21.5 21.69 21.94 CHECK TECH SPEC 3.4.4 LEAKAGE LIMITS MET V ce SR 3.4.4.1 and AOP-14.O. SUmp leak calculations required in Modes 1, 2, and 3- referen gpm within the previous 24 hour

• previous 24-hour period or if in Mode 1 increases by 2 If floor drain leak rate exceeds 5 gpm averaged over the period, enter Tech Spec 3.4.4 and OAOP-14.O.
  • previous 24-hour period, enter Tech Spec 3.4.4.

If total leakage rate exceeds 25 gpm averaged over the SHIFT Saturday Nightshift December 10, 2016 Week Beginning Page 20 of 80 Rev. 135 201-03.2

ATTACHMENT 1 Page 13 of 60 Unit 2 nce Report (RODSR) Reactor Operator Daily Surveilla DRYWELL LEA KAG E CALCULATION 12 16 08 1 16

                                          - RECORD MANUALLY PUMP EQUIPMENT DRAIN SUMP 199             LiI                                                                     I A An                              1 A4 AL                     J44fl CALCULATE TIME INTERVAL (MINUTES) FROM MANU PUMP AT SAME TIME ON PREVIOUS DAY.

0 I RECORD CURRENT INTEGRATOR READING. 0 12 9 0 0 8 I I j0 I I 0 f I I ING 32832 CALCULATE DIFFERENCE IN INTEGRATOR READ 31968 DAY 31248 FROM MANUAL PUMP AT SAME TIME ON PREVIOUS p_ 22 2 22 8 CALCULATE 24 HOUR EQUIPMENT DRAIN LEAK RATE 21 .7 (DIVIDE LEAKAGE BY TIME INTERVAL). 12 16 MANUALLY PUMP FLOOR DRAIN SUMP USING ONE 08 AL AArI 1440 44 CALCULATE TIME INTERVAL (MINUTES) FROM MANU -i-PUMP AT SAME TIME ON PREVIOUS DAY. RECORD INTEGRATOR READING. I 1 ING I IkI I I I I 12 I 9 0 f 5 1 I I f4 I I I CALCULATE DIFFERENCE IN INTEGRATOR READ 1012 1 131 DAY FROM MANUAL PUMP AT SAME TIME ON PREVIOUS E 0 92 CALCULATE 24 HOUR FLOOR DRAIN LEAK RATE (DIVIDE 0 70 0 79 LEAKAGE BY TIME INTERVAL). 22.2 22.8 24 HOUR EQUIPMENT DRAIN LEAK RATE. 21 .7 0.70 0.79 0.92 24 HOUR FLOOR DRAIN LEAK RATE.

• 22.99 23.72 24 HOUR TOTAL LEAK RATE TO DRYWELL. 22.4 V V CHECK TECH SPEC 3.4.4 LEAKAGE LIMITS MET Sump leak calculations required in Modes 1, 2, and 3 reference SR 3.4.4.1 and AOP-14.O.

1, increases by 2 gpm within the previous 24 hour

• the previous 24-hour period or if in Mode If floor drain leak rate exceeds 5 gpm averaged over period, enter Tech Spec 3.4.4 and OAOP-14.O.

Spec 3.4.4.

• • the previous 24-hour period, enter Tech If total leakage rate exceeds 25 gpm averaged over SHIFT Sunday Dayshift Beginning December 10, 2016 Week Page 21 of 80 Rev. 135 201-03.2

ATTACHMENT I Page 14 of 60 Unit 2 Reactor Operator Daily Surveillance Report (RODSR) DRYWELL LEAKAGE CALCULATION 00 04 20 MANUALLY PUMP EQUIPMENT DRAIN SUMP RECORD JMPSTOPSjLOWVEL.Rj CALCULATE TIME INTERVAL (MINUTES) FROM MANUAL PUMP AT SAME TIME ON PREVIOUS DAY. RECORD CURRENT INTEGRATOR READING. CALCULATE DIFFERENCE IN INTEGRATOR READING FROM MANUAL PUMP AT SAME TIME ON PREVIOUS DAY CALCULATE 24 HOUR EQUIPMENT DRAIN LEAK RATE (DIVIDE LEAKAGE BY TIME INTERVAL). MANUALLY PUMP FLOOR DRAIN SUMP USING ONE CALCULATE TIME INTERVAL (MINUTES) FROM MANUAL PUMP AT SAME TIME ON PREVIOUS DAY. RECORDINTEGRATORREADING. 1111111 1111111 1111111 CALCULATE DIFFERENCE IN INTEGRATOR READING FROM MANUAL PUMP AT SAME TIME ON PREVIOUS DAY (LEAKAGE). CALCULATE 24 HOUR FLOOR DRAIN LEAK RATE (DIVIDE LEAKAGE BY TIME INTERVAL). 24 HOUR EQUIPMENT DRAIN LEAK RATE. 24 HOUR FLOOR DRAIN LEAK RATE. 24 HOUR TOTAL LEAK RATE TO DRYWELL. CHECK TECH SPEC 3.4.4 LEAKAGE LIMITS MET 2, and 3 reference SR 3.4.4.1 and AOP-14.O. Sump leak calculations required in Modes 1 24 hour 1, increases by 2 gpm within the previous

• over the previous 24-hour period or if in Mode If floor drain leak rate exceeds 5 gpm averaged period, enter Tech Spec 3.4.4 and OAO P-i 4.0.

Spec 3.4.4.

• • ed over the previous 24-hour period, enter Tech If total leakage rate exceeds 25 gpm averag SHIFT Sunday Nightshift December 10, 2016 Week Beginning Page 22 0180 Rev. 135 201-03.2

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE LESSON TITLE: Determine Stay Time Limitations in High Radiation Areas LESSON NUMBER: LOT-ADM-JP-102-A03 REVISION NO: 3 Daniel Hulqin 09/06/16 PREPARER I DATE Bob Bolin 09/06/16 TECHNICAL REVIEWER I DATE Hunter Morris 09/06/16 Kyle Cooper 09/06/16 VALIDATOR I DATE 1%& LINE VISORIDQ_ E

          %-rr TRAINING SUPERVISION APPROVAL I DATE LOT-ADM-JP-102-A03                    Page 1 of 7                                Rev 3

RELATED TASKS: None K/A REFERENCE AND IMPORTANCE RATING: Generic 2.3.4 3.2/3.7 Knowledge of Radiation Exposure Limits under normal or emergency conditions Generic 2.3.7 3.5/3/6 Ability to comply with radiation work permit requirements during normal and abnormal conditions

REFERENCES:

PD-RP-ALL-0001, Radiation Worker Responsibilities TOOLS AND EQUIPMENT: Calculator Radiation Survey Map of 50 Reactor Building SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): A.3 Radiation Control SETUP INSTRUCTIONS None LOT-ADM-JP-1 02-A03 Page 2 of 7 Rev 3

SAFETY CONSIDERATIONS:

1. None EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL NOT be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-l 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. This JPM may be performed on Unit 1 or Unit 2 as selected by the evaluator. Survey map must reflect correct unit.

4. Critical Step Basis a) Prevents Task Completion b) May Result in Equipment Damage c) Affects Public Health and Safety d) Could Result in Personal Injury Read the following to the JPM performer.

TASK CONDITIONS: Two workers will be performing a lube check and coupling alignment on the Unit 2 RWCU Pump 2A. Worker #1 has accumulated 800 mrem this year. Worker #2 has accumulated 970 mrem this year. The elevator is out of service The following times for each worker have been estimated for performance of the job.

1. Traversing Southeast stairwell 20 50 Rx Bldg:

6 minutes

2. Staging time in access area directly outside the RWCU room: 45 minutes

3. Staging time in area directly inside room access door: 20 minutes

4. Work time at the A RWCU pump: 2.5 hours

5. Following completion of the job, an additional 60 mrem per worker will be received during de-staging activities and transit back to the maintenance shop.

INITIATING CUE: Using the information above and the provided radiological survey using best ALARA practices:

1. Determine the total dose accumulated for both workers. (Assume the same task times for both workers).

2. Determine if any Brunswick administrative dose limitations will be exceeded.

LOT-AD M-]P-1 02-A03 Page 3 of 7 Rev 3

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. TIME START Step 1 - Determines dose for each worker as follows:

a. Traversing SE stairwell 20 50 Rx Bldg (SE is the lowest dose stairwell)

(6 mm) 0.1 Hr X 5 mr/hr = 0.5 mrem Estimate 0.5 mrem dose accumulation

                                                               **CRITICAL STEP**           SATIUNSAT

b. Staging time in access area directly outside the RWCU room (45 mm) 0.75 Hr X 20 mr/hr = 15 mrem Estimate 15 mrem dose accumulation

                                                               **CRJTICAL STEP**           SATIUNSAT

c. Staging time in area directly in side room access door (20 mm) 0.33 Hr X 80 mr/hr = 26.7 mrem Estimate 26.7 mrem dose accumulation.

                                                               **CRITICAL STEP**           SATIUNSAT

d. Work time at the A RWCU pump 2.5 Hrs X 200 mr/hr = 500 mrem Estimate 500 millirem dose accumulation

                                                               **CRITICAL STEP**           SATIUNSAT NOTE: An additional 60 mr will be accumulated once the job is done for de-staging activities.

e. Total = 0.5 + 75 + 26.7 ÷ 500 ÷ 60 = 602.2 mrem

                                                               **CRITICAL STEP**           SATIUNSAT LOT-ADM-JP-1 02-A03                          Page 4 of 7                                            Rev 3

Step 2 - Determines that neither worker would exceed the Brunswick administrative limit of 2 REM per calendar year if the estimated dose were accumulated. Worker #1: 800 mr + 602.2 mr = 1402.2 mr (< 2R limit) Worker #2: 970 mr + 602.2 mr = 1572.2 mr (< 2R limit)

                                                             **CRITICAL STEP**         SATIUNSAT TERMINATING CUE: When the total dose for each worker has been determined and the administrative limits addressed, the JPM is complete.

TIME COMPLETED: NOTE: Comments required for any step evaluated as UNSAT. Step Critical I Not Critical Reason 1a Critical Each calculation is critical to determine total dose for personnel safety. Ib Critical Each calculation is critical to determine total dose. Ic Critical Each calculation is critical to determine total dose. Id Critical Each calculation is critical to determine total dose. le Critical Each calculation is critical to determine total dose. 2 Critical Total calculation and knowledge of Admin Dose Limit is required to complete JPM. REVISION

SUMMARY

3 Removed take a minute-step 1. Reordered steps 2 Revised to new JPM Template Revised times so that calculations are different than previous versions. I Revised to new JPM Template, Revision 3. No technical changes. LOT-ADM-JP-102-A03 Page 5 017 Rev 3

Validation Time: 15 Minutes (approximate) Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate X Actual X Unit: 2 Setting: In-Plant Simulator Admin Time Critical: Yes No X Time Limit N/A Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: D Comments reviewed with Performer Evaluator Signature: Date: LOT-ADM-JP-1 02-A03 Page 6 of 7 Rev 3

TASK CONDITIONS: Two workers will be performing a lube check and coupling alignment on the Unit 2 RWCU Pump 2A. Worker #1 has accumulated 800 mrem this year. Worker #2 has accumulated 970 mrem this year. The elevator is out of service The following times for each worker have been estimated for performance of the job.

1. Traversing Southeast stairwell 20 50 Rx B lug:

6 minutes

2. Staging time in access area directly outside the RWCU room: 45 minutes

3. Staging time in area directly inside room access door: 20 minutes

4. Work time at the A RWCU pump: 2.5 hours

5. Following completion of the job, an additional 60 mrem per worker will be received during de-staging activities and transit back to the maintenance shop.

INITIATING CUE: Using the information above and the provided radiological survey using best ALARA practices:

1. Determine the total dose accumulated for both workers. (Assume the same task times for both workers).

2. Determine if any Brunswick administrative dose limitations will be exceeded.

Results: Page 7 of 7

DUKE ENERGY DUKE BRUNSWICK TRAINING SECTION ENERGY JOB PERFORMANCE MEASURE LESSON TITLE: CLASSIFY AN EMERGENCY PER PEP-02.1. LESSON NUMBER: SOT-ADM-JP-301-M6 REVISION NO: I Daniel Hulciin 09/06/1 6 PREPARER I DATE Bob Bolin 09/06/16 TECHNICAL REVIEWER I DATE Dwayne Wolf 09/06/16 Kyle Cooper 09/06/16 VALIDATOR I DATE (% PERVISOR LINE

               &% I DA 2

TRAINING SUPERVISION APPROVAL I DATE 9 2 - / SOT-ADM-JP-301-A16 Page 1 of 8 Rev.1

RELATED TASKS: 344256B502 Direct initial emergency actions including emergency classification per OPEP-02.1 K/A REFERENCE AND IMPORTANCE RATING: GEN 2.4.29 3.1/4.4 Knowledge of the Emergency Plan

REFERENCES:

OPEP-02. 1 TOOLS AND EQUIPMENT: None SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1): Admin Emergency Procedures I Plan SETUP INSTRUCTIONS None SOT-ADM-JP-301-A16 Page 2 of 8 Rev.1

SAFETY CONSIDERATIONS:

1. None EVALUATOR NOTES: (Do not read to performer)

1. The applicable procedure section WILL be provided to the trainee.

2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1 021, Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).

3. Critical Step Basis a) Prevents Task Completion b) May Result in Equipment Damage c) Affects Public Health and Safety U) Could Result in Personal Injury Read the following to the JPM performer.

TASK CONDITIONS:

1. Unit One is operating at 100% power.

2. Unit Two is operating at 100% power with DG4 under clearance when the following event occurs (consider all items that exceed EAL thresholds occur at the same time).

Unit Two Event Description

• A seismic event greater than the Operating Basis Earthquake results in a loss of the PBX Telephone System, Commercial Telephones, and NRC Emergency Telecommunications System
• A Manual Scram is inserted, the Mode Switch is in Shutdown, ARI is initiated, and reactor power indicates 20%.
• Driving rods lAW LEP-02, Alternate Control Rod Insertion, and SLC injection are in progress.
• Current indications: reactor power is I %, reactor water level maintained +60 to +90 inches, and reactor pressure is 945 psig on EHC.
• NO radiological releases in progress, and NO indications of an onsite security event.

INITIATING CUE: You are to evaluate the above event as the Control Room Site Emergency Coordinator (SEC) and determine the HIGHEST required classification and its EAL Identifier for Unit Two ONLY:

1. Write the required Classification and its associated EAL identifier in the table below.

2. Raise your hand when complete to have the evaluator stop the evaluation time and collect your cue sheet This JPM is TIME CRITICAL.

CLASSIFICATION EAL IDENTIFIER(s) SOT-ADM-JP-301-AI6 Page 3 of 8 Rev.1

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless otherwise indicated, comments required for any step evaluated as UNSAT. PROMPT: Ensure a clock is visible for candidates. Announce and Write the Start Time on the board. Add 15 minutes to the Start Time and write that in the ]PM Completion Time. If all candidates have not Declared a classification, this is the time to STOP all work, put pencils/pens down, and collect all remaining cue I NOTE: Declaration of event must be made in 15 minutes from the Start Time. TIME START: NOTE: Loss of PBX Telephone System, Commercial Telephones, and NRC Emergency Telecommunications System does not reach EAL classification threshold Step 1 Determine required Classification threshold and associated EAL Number(s). o Unusual Event HU2. I

                -Seismic event> OBE per OAOP-13.O SATIUNSAT NOTE: Candidate may base the ALERT on SA8.1 if they determine the ATWS was a result of the earthquake. Critical Step is that the ALERT is based on EITHER SA8.1 or SA6.1 EITHER Step 2 OR 3 is critical.

SOT-ADM-JP-301-A16 Page 4 of 8 Rev.1

Step 2 Determine required Classification threshold and associated EAL Number(s). o Alert-SAB.7

            -The occurrence of any Table S-4 hazardous event (Seismic event).

AND EITHER

            -Event damage has caused indications of degraded performance in at least one train of a safety system needed for the current operating mode (A TW$).
            -The event has caused visible damage to a safety system component or structure needed for the current operating mode.
                                                             **CRITJCAL STEP** SATIUNSAT Step 3  Determine required Classification threshold and associated EAL Number(s).

o Alert SA6.1

            -An automatic or manual scram fails to reduce reactor power <2% (APRM downscale).

AND

            -Manual scram actions taken at the reactor control console (Manual PBs, Mode Switch, ARI) are not successful in shutting down the reactor as indicated by reactor power> 2% (note 8)

Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving control rods or boron injection strategies.

                                                             **CRITICAL STEP** SATIUNSAT Step 4  Classification made within required the required time (Declaration Time minus Start Time < 15 minutes).

o Classification declared < 15 minutes of Start Time.

                                                             **CRITICAL STEP** SATIUNSAT TERMINATING CUE: When the event is classified with applicable EAL identifier(s) in the table, this JPM is complete.

TIME COMPLETED: SOT-ADM-JP-301-A16 Page 5 of 8 Rev.1

COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY. Step Critical I Not Critical Reason 1 Not Critical This EAL is not the highest classification 2 or 3 Critical Highest EAL classification and EAL designator 4 Critical Time to declare is critical REVISION

SUMMARY

1 Incorporated new format. Modified initial conditions as follows:

• Changed earthquake magnitude to OBE per AOP13
• Changed Voicenet to PBX
• Removed turbine trip
• Changed reactor power to 1% following SLC and driving rods Changed initiating cue to include EAL identifier(s) instead of EAL identifier Changed step 1 to reflect new standard instead of obtaining a procedure Step 2 EAL criteria updated to new PEP2.1 criteria Step 3 EAL criteria updated to new PEP2.1 criteria. This is now a critical step. Its EAL designation is now part of the highest classification Step 4 3 EAL criteria updated to new PEP2.1 criteria. This is now an ALERT and no longer a SAE Removed take a minute (step 1) and reordered steps.

o NewJPM (Provide sufficient detail for reviewers and evaluators to understand the scope of any technical and/or administrative changes). SOT-ADM-JP-301-A16 Page 6 of 8 Rev.1

Validation Time: 15 Minutes (approximate). Time Taken: Minutes APPLICABLE METHOD OF TESTING Performance: Simulate Actual X Unit: 2 Setting: In-Plant Simulator Admin X Time Critical: Yes X No Time Limit 15 mm Alternate Path: Yes No X EVALUATION Performer: JPM: Pass Fail Remedial Training Required: Yes No Comments: D Comments reviewed with Performer Evaluator Signature: Date: SOT-ADM-JP-301-A16 Page 7 of 8 Rev.1

This JPM is TIME CRITICAL. TASK CONDITIONS:

1. Unit One is operating at 100% power.

2. Unit Two is operating at 100% power with DG4 under clearance when the following event occurs (consider all items that exceed EAL thresholds occur at the same time).

Unit Two Event Description

• A seismic event greater than the Operating Basis Earthquake results in a loss of the PBX Telephone System, Commercial Telephones, and NRC Emergency Telecommunications System
• A Manual Scram is inserted, the Mode Switch is in Shutdown, ARI is initiated, and reactor power indicates 20%.
• Driving rods lAW LEP-02, Alternate Control Rod Insertion, and SLC injection are in progress.
• Current indications: reactor power is 1 %, reactor water level maintained +60 to +90 inches, and reactor pressure is 945 psig on EHC.
• NO radiological releases in progress, and NO indications of an onsite security event.

INITIATING CUE: You are to evaluate the above event as the Control Room Site Emergency Coordinator (SEC) and determine the HIGHEST required classification and its EAL Identifier for Unit Two ONLY:

1. Write the required Classification and its associated EAL identifier in the table below.

2. Raise your hand when complete to have the evaluator stop the evaluation time and collect your cue sheet CLASSIFICATION EAL IDENTIFIER This JPM is TIME CRITICAL.

Page 8 of 8}}