ML17137A186

WATERFORD-2017-03 Draft Written Examination - Do Not Release Until 4-30-19

ML17137A186
Person / Time
Site:	Waterford
Issue date:	04/05/2017
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
References
Download: ML17137A186 (540)

Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 (RO Exam) Date of Exam: March 27, 2017 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

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

Abnormal 2 2 2 1 N/A 1 2 N/A 1 9 4 Plant Evolutions Tier Totals 5 5 4 4 5 4 27 10 1 3 3 2 3 2 2 2 2 3 3 3 28 5 2.

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

3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 7 Categories 3 3 2 2 10 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-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 CE/E02, EA1.2: Ability to operate and 000007 (BW/E02&E10; CE/E02) Reactor X / or monitor the following as they 3.3 1 Trip - Stabilization - Recovery / 1 apply to the (Reactor Trip Recovery):

Operating behavior characteristics of the facility (CFR: 41.7 / 45.5 / 45.6)

AK2.01: Knowledge of the 000008 Pressurizer Vapor Space X interrelations between the Pressurizer 2.7 2 Accident / 3 Vapor Space Accident and the following: Valves (CFR 41.7 / 45.7)

EK3.04: Knowledge of the reasons for 000009 Small Break LOCA / 3 X the following responses as the apply 4.1 3 to the small break LOCA: Starting additional Charging Pumps (CFR 41.5 /

41.10 / 45.6 / 45.13)

EA2.03: Ability to determine or 000011 Large Break LOCA / 3 X interpret the following as they apply 3.7 4 to a Large Break LOCA: Consequences of managing LOCA with loss of CCW (CFR 43.5 / 45.13)

AK2.07: Knowledge of the 000015/17 RCP Malfunctions / 4 X interrelations between the Reactor 2.9 5 Coolant Pump Malfunctions (Loss of RC Flow) and the following: RCP seals (CFR 41.7 / 45.7)

AK1.03: Knowledge of the operational 000022 Loss of Rx Coolant Makeup / 2 X implications of the following concepts 3.0 6 as they apply to Loss of Reactor Coolant Makeup: Relationship between charging flow and PZR level (CFR 41.8 / 41.10 / 45.3) 2.1.27: Knowledge of system purpose 000025 Loss of RHR System / 4 X and/or function. (CFR: 41.7). 3.9 7 AA2.01: Ability to determine and 000026 Loss of Component Cooling X interpret the following as they apply 2.9 16 Water / 8 to the Loss of Component Cooling Water: Location of a leak in the CCW System. (CFR: 43.5 / 45.13) 2.2.44: Ability to interpret control 000027 Pressurizer Pressure Control X room indications to verify the status 4.2 8 System Malfunction / 3 and operation of a system, and understand how operator actions and directives affect plant and system conditions (CFR: 41.5 / 43.5 / 45.12)

EA1.12: Ability to operate and 000029 ATWS / 1 X monitor the following as they apply to 4.1 9 a ATWS: M/G set power supply and reactor trip breakers (CFR 41.7 / 45.5

/ 45.6)

EK3.02: Knowledge of the reasons for 000038 Steam Gen. Tube Rupture / 3 X the following responses as the apply 4.4 10 to the SGTR: Prevention of secondary PORV cycling (CFR 41.5 / 41.10 / 45.6

/ 45.13)

CE/E05, EK1.2: Knowledge of the 000040 (BW/E05; CE/E05; W/E12) X operational implications of the 3.2 11 Steam Line Rupture - Excessive Heat following concepts as they apply to Transfer / 4 the (Excess Steam Demand): Normal, abnormal and emergency operating procedures associated with (Excess Steam Demand)(CFR: 41.8 / 41.10 /

45.3)

CE/E06, EK1.1: Knowledge of the 000054 (CE/E06) Loss of Main X operational implications of the 3.2 12 Feedwater / 4 following concepts as they apply to the (Loss of Feedwater): Components, capacity, and function of emergency systems. (CFR: 41.8 / 41.10 / 45.3)

EA2.03: Ability to determine or 000055 Station Blackout / 6 X interpret the following as they apply 3.9 13 to a Station Blackout: Actions necessary to restore power (CFR 43.5 /

45.13) 2.4.18: Knowledge of the specific 000056 Loss of Off-site Power / 6 X bases of the EOPs (CFR: 41.10 / 43.1 / 3.3 14 45.13) 000057 Loss of Vital AC Inst. Bus / 6 AK3.01: Knowledge of the reasons for 000058 Loss of DC Power / 6 X the following responses as they apply 3.4 15 to the Loss of DC Power: Use of dc control power by D/Gs (CFR 41.5,41.10

/ 45.6 / 45.1) 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 X AA1.01: Ability to operate and / or 2.7 17 monitor the following as they apply to the Loss of Instrument Air: Remote manual loaders W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 AK2.07: Knowledge of the 000077 Generator Voltage and Electric X interrelations between Generator 3.6 18 Grid Disturbances / 6 Voltage and Electric Grid Disturbances and the following: Turbine/Generator Control (CFR: 41.4, 41.5, 41.7, 41.10

/ 45.8)

K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18/

6

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 X AK2.01: Knowledge of the 2.7 20 interrelations between Emergency Boration and the following:

Valves (CFR 41.7 / 45.7) 000028 Pressurizer Level Malfunction / 2 X AA2.12: Ability to determine and 3.1 21 interpret the following as they apply to the Pressurizer Level Control Malfunctions: Cause for PZR level deviation alarm:

controller malfunction or other instrumentation malfunction (CFR:

43.5 / 45.13) 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 X AK1.02: Knowledge of the 3.5 23 operational implications of the following concepts as they apply to Steam Generator Tube Leak:

Leak rate vs. pressure drop (CFR 41.8 / 41.10 / 45.3) 000051 Loss of Condenser Vacuum / 4 X AK3.01: Knowledge of the reasons 2.8 22 for the following responses as they apply to the Loss of Condeser Vacuum: Loss of Steam Dump capability upon loss of condenser vacuum. (CFR 41.5,41.10

/ 45.6 / 45.13) 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 X 2.4.31: Knowledge of annunciator 4.2 24 alarms, indications, or response procedures (CFR: 41.10 / 45.3) 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 AK2.03 Knowledge of the 000068 (BW/A06) Control Room Evac. / 8 X 2.9 25 interrelations between the Control Room Evacuation and the following: Controllers and positioners (CFR 41.7 / 45.7) 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 X EA1.01: Ability to operate and 4.2 26 monitor the following as they apply to a Inadequate Core Cooling: RCS Water Inventory (CFR 41.7 / 45.5 / 45.6) 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4

W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 X AA2.2: Ability to determine and 3.0 27 interpret the following as they apply to the (RCS Overcooling):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments (CFR: 43.5

/ 45.13)

AK1.1: Knowledge of the CE/A16 Excess RCS Leakage / 2 X 3.2 19 operational implications of the following concepts as they apply to the (Excess RCS Leakage):

Components, capacity, and function of emergency systems (CFR 41.8 / 41.10 / 45.3)

CE/E09 Functional Recovery K/A Category Point Totals: 2 2 1 1 2 1 Group Point Total: 9/4

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 K6.14: Knowledge of the 003 Reactor Coolant Pump X effect of a loss or 2.6 28 malfunction on the following will have on the RCPS: Starting requirements (CFR: 41.7 / 45/5)

A4.07: Ability to manually 004 Chemical and Volume X operate and/or monitor in the 3.9 29 Control control room:

Boration/dilution (CFR: 41/7

/ 45.5 to 45.8)

X K5.30: Knowledge of the 3.8 51 operational implications of the following concepts as they apply to the CVCS:

Relationship between temperature and pressure in CVCS components during solid plant operation A4.02: Ability to manually 005 Residual Heat Removal X X operate and/or monitor in the 3.4 30 control room: Heat exchanger bypass flow control (CFR: 41.7 / 45.5 to 45.8) 2.1.20: Ability to interpret 4.6 31 and execute procedure steps.

(CFR: 41.10 / 43.5 / 45.12)

K4.17: Knowledge of ECCS 006 Emergency Core Cooling X design feature(s) and/or 3.8 32 interlock(s) which provide for the following: Safety Injection valve interlocks (CFR: 41.7)

A2.03: Ability to (a) predict 007 Pressurizer Relief/Quench X the impacts of the following 3.6 33 Tank malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Overpressurization of the PZR (CFR: 41.5 / 43.5 / 45.3 /

45.13)

K1.05: Knowledge of the 008 Component Cooling Water X X physical connections and/or 3.0 34 cause-effect relationships between the CCWS and the following systems: Sources of makeup water(CFR: 41.2 to 41.9 /

45.7 to 45.9)

A3.08: Ability to monitor automatic operation of the 3.6 35 CCWS, including: Automatic actions associated with the CCWS that occur as a result of a safety injection signal (CFR: 41.7 / 45.5)

K6.01: Knowledge of the 010 Pressurizer Pressure Control X X effect of a loss or 2.7 36 malfunction of the following will have on the PZR PCS:

Pressure detection systems (CFR: 41.7 / 45.7) 2.8 37 A1.01: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including:

PZR and RCS boron concentrations (CFR: 41.5 /

45.5)

K1.02: Knowledge of the 012 Reactor Protection X physical connections and/or 3.4 38 cause effect relationships between the RPS and the following systems: 125V dc system (CFR:

41.2 to 41.9 / 45.7 to 45.8)

K2.01: Knowledge of bus power 013 Engineered Safety Features X X supplies to the following: 3.6 39 Actuation ESFAS/safeguards equipment control (CFR: 41.7)

A2.02: Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based 4.3 40 Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Excess steam demand (CFR: 41.5 / 43.5 /

45.3 / 45.13)

K3.02: Knowledge of the 022 Containment Cooling X effect that a loss or 3.0 41 malfunction of the CCS will have on the following: Containment instrumentation readings (CFR: 41.7 / 45.6) 025 Ice Condenser K2.01: Knowledge of bus power 026 Containment Spray X supplies to the following: 3.4 42 Containment spray pumps (CFR:

41.7)

K5.08: Knowledge of the 039 Main and Reheat Steam X operational implications of 3.6 43 the following concepts as the apply to the MRSS:

Effect of steam removal on reactivity (CFR: 441.5 /

45.7)

K4.18: Knowledge of MFW 059 Main Feedwater X design feature(s) and/or 2.8 44 interlock(s) which provide for the following: Automatic feedwater reduction on plant trip (CFR: 41.7)

K2.02: Knowledge of bus power 061 Auxiliary/Emergency X X supplies to the following: 3.7 46 Feedwater AFW electric drive pumps(CFR:

41.7)

A3.03: Ability to monitor automatic operation of the 3.9 47 AFW, including: AFW S/G level control on automatic start (CFR: 41.7 / 45.5)

A3.05: Ability to monitor 062 AC Electrical Distribution X automatic operation of the ac 3.5 48 distribution system, including: Safety-related indicators and controls (CFR:

41.7 / 45.5)

A4.02: Ability to manually 063 DC Electrical Distribution X operate and/or monitor in the 2.8 49 control room: Battery voltage indicator (CFR: 41.7 / 45.5 to 45.8)

K4.10: Knowledge of ED/G 064 Emergency Diesel Generator X system design feature(s) 3.5 50 and/or interlock(s) which provide for the following:

Automatic load sequencer:

blackout (CFR: 41.7) 073 Process Radiation Monitoring X 2.4.4: Ability to recognize 4.5 52 abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. l (CFR: 41.10 / 43.2 / 45.6)

K1.19: Knowledge of the 076 Service Water X physical connections and/or 3.6 53 cause- effect relationships between the SWS and the following systems: SWS Emergency heat loads (CFR:

41.2 to 41.9 / 45.7 to 45.8)

X A1.02: Ability to predict 2.6 45 and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: Reactor and turbine building closed cooling water temperatures K3.01: Knowledge of the 078 Instrument Air X effect that a loss or 3.1 54 malfunction of the IAS will have on the following: Containment air system (CFR: 41.7 / 45.6) 2.2.44: Ability to interpret 103 Containment X control room indications to 4.2 55 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)

K/A Category Point Totals: 3 3 2 3 2 2 2 2 3 3 3 Group Point Total: 28/5

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive X K5.06: Knowledge of the 3.8 56 following operational implications as they apply to the CRDS: Effects of control rod motion on axial offset(CFR:

41.5/45.7)

K6.02: Knowledge of the effect 002 Reactor Coolant X or a loss or malfunction on the 3.6 57 following RCS components: RCP(CFR: 41.7 /

45.7) 011 Pressurizer Level Control 014 Rod Position Indication K4.07: Knowledge of NIS design 015 Nuclear Instrumentation X feature(s) and/or interlock(s) 3.7 58 provide for the following:

Permissives (CFR: 41.7) 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor X A4.02: Ability to manually 3.8 59 operate and/or monitor in the control room: Temperature values used to determine RCS/RCP operation during inadequate core cooling (i.e., if applicable, average of five highest values)

(CFR: 41.7 / 45.5 to 45.8) 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling X A1.01: Ability to predict and/or 2.7 60 monitor changes in parameters(to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including: Spent fuel pool water level (CFR: 41.5 /

45.5) 034 Fuel Handling Equipment 035 Steam Generator A3.02: Ability to monitor 041 Steam Dump/Turbine Bypass X automatic operation of the SDS, 3.3 61 Control including: RCS Pressure, RCS Temperature, and reactor power (CFR: 41.7 / 45.5)

045 Main Turbine Generator X A2.08: Ability to (a) predict 2.8 62 the impacts of the following malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Steam dumps are not cycling properly at low load, or stick open at higher load (isolate and use atmospheric reliefs when necessary) (CFR:

41.5 / 43.5 / 45.3 / 45.5)

K3.01: Knowledge of the effect 055 Condenser Air Removal X that a loss or malfunction of 2.5 63 the CARS will have on the following: Main condenser (CFR:

41.7 / 45.6) 056 Condensate 068 Liquid Radwaste K1.04: Knowledge of the physical 071 Waste Gas Disposal X connections and/or cause-effect 2.7 64 relationships between the Waste Gas Disposal System and the following systems: Station ventilation (CFR: 41.2 to 41.9 /

45.7 to 45.8) 072 Area Radiation Monitoring 2.1.30: Ability to locate and 075 Circulating Water X operate components, including 4.4 65 local controls. (CFR: 41.7 /

45.7) 079 Station Air 086 Fire Protection K/A Category Point Totals: 1 0 1 1 1 1 1 1 1 1 1 Group Point Total: 10/

3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Waterford 3 (RO) Date of Exam: September 14, 2015 Category K/A # Topic RO SRO-Only IR # IR #

Knowledge of shift or short-term relief turnover practices.

2.1.3 3.7 66 (CFR: 41.10 / 45.13)

Ability to use plant computers to evaluate system or 2.1.19 3.9 67 component status. (CFR: 41.10 / 45.12)

Knowledge of RO duties in the control room during fuel 2.1.44 3.9 68

1. handling, such as responding to alarms from the fuel Conduct of handling area, communication with the fuel storage Operations facility, systems operated from the control room in support of fueling operations, and supporting instrumentation. (CFR: 41.10 / 43.7 / 45.12) 2.1.

Subtotal Knowledge of the process for making changes to 2.2.6 3.0 69 procedures. (CFR: 41.10 / 43.3 / 45.13)

Ability to determine Technical Specification Mode of

2. 2.2.35 3.6 70 Operation. (CFR: 41.7 / 41.10 / 43.2 / 45.13)

Equipment Control 2.2.39 Knowledge of less than or equal to one hour Technical 3.9 71 Specification action statements for systems.(CFR: 41.7 /

41.10 / 43.2 / 45.13)

Subtotal Knowledge of radiological safety procedures pertaining to 2.3.13 3.4 72 licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 43.4 / 45.9 /

3. 45.10)

Radiation Knowledge of radiation monitoring systems, such as fixed 2.3.15 2.9 73 Control radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR:

41.12 / 43.4 / 45.9) 2.3.

Subtotal Knowledge of how abnormal operating procedures are 2.4.8 3.8 74 used in conjunction with EOPs.(CFR: 41.10 / 43.5 /

45.13)

4. Knowledge of fire protection procedures.

Emergency 2.4.25 3.3 75 (CFR: 41.10 / 43.5 / 45.13)

Procedures /

Plan 2.4.

2.4.

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 1/1 009 EK3.04 009 EK 3.17 rejected because suitable distractors were not available for knowledge of the reason for a containment RO 3 isolation. All the distractors we came up with were not plausible.

1/1 056 2.4.18 056 2.4.45 rejected because we could not develop a question with plausible distractors for a loss of offsite power that would RO 14 require a prioritization of annunciators (first part of the K/A) 1/1 026 AA2.01 062 AA2.02 The only possible loss of ACCW is a pump trip.

The only ACCW pump trips is overcurrent and undervoltage.

RO 16 A question was developed on the ACCW pump failure to start on the EDG sequencer but discovered it overlapped a question on one of the previous two exams. Could not develop a question that did not involve the EDG sequencer that would have plausible distractors.

1/1 065 AA1.01 065 AA1.04 is related to an Emergency Air Compressor for a Loss of Instrument Air. W3 does not have an Emergency Air RO17 Compressor.

1/1 077 AK2.07 077 AK2.02 Could not link breakers, relays with generator voltage and grid disturbance that could be supported with RO 18 adequate technical reference.

1/2 CE/A16 AK1.1 0003 AK1.19 rejected because we could not identify plausible distractors when creating a question. There is no references RO 19 in OPS procedures that discussed differential rod worth during a dropped CEA.

1/2 051 AK3.01 032 AK3.01 rejected because we could not develop a loss of source range question that did not duplicate existing RO RO 22 questions on this exam (RO68) or the previous two exams.

1/2 074 EA1.01 074 EA1.13 and the K/A for question number 59 were both asking about subcooling monitors. Rejected the K/A for RO RO 26 26 because we could not develop independent questions.

2/1 003 K6.14 003 K6.04 rejected because we could not develop a question where a fault of a CIV would affect a RCP. CBO flow has two RO 28 CIVs but isolating them has no effect due to flow being directed to another location, and the CCW valves to the RCPs fail in the open (no effect) position.

2/1 012 K1.02 012 K1.04 rejected. The rod position indicating system does not have a direct physical connection with the W3 RPS RO 38 system. Could not develop a question that effectively matched the K/A.

2/1 076 A1.02 059 A1.03 rejected because a question could not be developed that did not require a reference and a direct look RO 45 up.

2/1 004 K5.30 073 K5.01 rejected because we could only find questions for this K/A that pertained to radiation theory. We could not apply RO 51 this theory to the Process Radiation Monitors and have adequate technical references.

2/1 076 K1.19 076 K1.05 rejected because the ACCW system does not have a physical connection with the EDG system.

RO 53 2/2 041 A3.02 041 A3.05 rejected. Two independent Main Steam crossover pressures would have to fail such that the SDS is affected RO 61 (permissive and demand). The question developed for this K/A was not plausible.

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000007 (CE/E02)EA1.2 Importance Rating 3.3 K/A Statement 007 Reactor Trip Recovery, EA1.2: Ability to operate and / or monitor the following as they apply to the (Reactor Trip Recovery): Operating behavior characteristics of the facility.

Proposed Question: RO 1 Rev: 0 Given:

Circulating Water Pump (CWP) A is unavailable and danger tagged Circulating Water Pumps B, C and D are running Plant tripped due to feedwater heater perturbations The auto transfer to the Startup Transformers was not successful for the 1B and 2B electrical busses Following the plant trip, the plant is left with ____(1)____ Circulating Water Pump(s) running. The diagnostic flowchart will direct the crew to _____(2)_____ Recovery Procedure.

(1) (2)

A. one OP-902-006, Loss of Feedwater B. one OP-902-001, Reactor Trip C. zero OP-902-001, Reactor Trip D. zero OP-902-006, Loss of Feedwater Revision 0 Facility: Waterford 3 Page 1 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. With three CW pumps running and two trip, the remaining CW pump is designed to trip. The EOPs have been revised such that with AFW and /or 100% EFW flow equivalent available, the Reactor Trip Recovery Procedure is entered. Loss of MFW Recovery Procedure would have been correct prior to the EOP revision.

B. Incorrect. With three CW pumps running and two trip, the remaining CW pump is designed to trip. Part 2 is correct.

C. CORRECT: This question is from a W3 event (CR-WF3-2015-3565). With three CW pumps running and two trip, the remaining CW pump is designed to trip. Since CWP A is initially tagged out, the plant is left with zero CW pumps running. The plant is left without Main Feed Pumps due to an eventual loss of Condenser Vacuum. In this event, the crew diagnosed into the Loss of Main Feedwater Recovery Procedure. The EOPs have since been revised such that with AFW and /or 100% EFW flow equivalent available, the Reactor Trip Recovery Procedure (OP-902-001) is entered.

D. Incorrect. Part 1 is correct. The EOPs have been revised such that with AFW and /or 100%

EFW flow equivalent available, the Reactor Trip Recovery Procedure is entered. Loss of MFW Recovery Procedure would have been correct prior to the EOP revision.

Technical Reference(s): TG-OP-902-001 page 24 Rev. 306 (Attach if not previously provided) OP-902-001 pages 8 and 16 Rev. 16 (including version/revision number) OP-902-009 Appendix 1 Rev. 315 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 12,16 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 2 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000008 AK2.01 Importance Rating 2.7 K/A Statement 008 Pressurizer Vapor Space Accident, AK2.01: Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Valves.

Proposed Question: RO 2 Rev: 0 Given:

A Pressurizer Relief Valve spuriously opens Reactor Coolant System pressure is currently 1500 PSIA and steady Quench Tank temperature rises to 165°F and stabilizes QSPDS indicates saturation conditions in the reactor vessel head The Pressurizer Relief Valve is currently ____(1)____.

During the event, Pressurizer Level will ________(2)________.

(1) (2)

A. closed rise B. closed lower C. open lower D. open rise Revision 0 Facility: Waterford 3 Page 3 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Based on quench tank temperature stabilizing below the saturation temperature of containment, the applicant can determine that the rupture disc on the quench tank is intact and since temperature is steady, the relief valve is closed. With saturation conditions in the vessel head, pressurizer level will rise due to bubble formation in the head and no RCPs running.

B. Incorrect. Part 1 is correct. Pressurizer level will rise due to bubble formation in the head.

Pressurizer level lowering is plausible because there is a mass loss of the RCS out of the relief valve.

C. Incorrect. Quench tank temperature would continue to rise if the valve were still open.

Quench tank temperature would eventually stabilize if the relief valve ruptured. The temperature would stabilize at saturation temperature for containment pressure once the QT diaphragm ruptured. Pressurizer level will rise due to bubble formation in the head.

Pressurizer level lowering is plausible because there is a mass loss of the RCS out of the relief valve.

D. Incorrect. Quench tank temperature would continue to rise if the valve were still open. Part 2 is correct. Quench tank temperature would eventually stabilize if the relief valve ruptured. The temperature would stabilize at saturation temperature for containment pressure once the QT diaphragm ruptured Technical Reference(s): Steam Tables (Attach if not previously provided) TG-OP-902-002 page 10 Rev. 19 (including version/revision number) WLP-OPS-PPE02 slide 58 Rev. 17 Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: WLP-OPS-RCS00 obj. 2 (As available)

WLP-OPS-PPE02 obj. 3 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 3 55.43 Revision 0 Facility: Waterford 3 Page 4 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000009 EK3.04 Importance Rating 4.1 K/A Statement 009 Small Break LOCA, EK3.04: Knowledge of the reasons for the following responses as they apply to the small break LOCA: Starting additional Charging Pumps Proposed Question: RO 3 Rev: 1 Given:

A SBLOCA has occurred RCS pressure is 1600 psia and lowering Pressurizer level is 15% and lowering Charging Pump AB Assignment switch is in the NORM position Two CEAs failed to insert into the core on the reactor trip The ATC informs the CRS that only Charging Pump A is running.

For this condition, the crew will verify all available charging pumps running by starting Charging Pump(s) ____(1)____ .

The crew started the additional charging pump(s) to ensure ______(2)_____.

(1) (2)

A. B only adequate RCS makeup is available B. B only shutdown margin requirements met C. B and AB shutdown margin requirements met D. B and AB adequate RCS makeup is available Revision 0 Facility: Waterford 3 Page 5 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Since the Charging Pump AB Assignment switch is in Norm and a SIAS is initiated due to RCS pressure < 1684 psia, Charging Pumps A and B should be running and Charging Pump AB will be secured. OI-038-000 step 5.4.72 defines all available charging pumps with a SIAS as two running. With RCS pressure above HPSI shutoff head, charging pumps are required to maintain pressurizer level.

B. Incorrect. Part 1 is correct. Two CEAs stuck out requires emergency boration to meet shutdown margin. Only one Charging pump >40 gpm is required to meet Shutdown Margin.

C. Incorrect. If there was no SIAS, all three charging pumps would be required to be running due to a level deviation. Two CEAs stuck out requires emergency boration to meet shutdown margin. Only one Charging pump >40 gpm is required to meet Shutdown Margin.

D. Incorrect. If there was no SIAS, all three charging pumps would be required to be running due to a level deviation. Part 2 is correct.

Technical Reference(s): OI-038-000 step 5.4.47 Rev 13 (Attach if not previously provided) OP-902-002 page 69 Rev. 20 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam none Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 5, 10 55.43 Comment: Free review question. Initial question did not meet the K/A. Changed part 1 such that an additional Charging pump is started. Changed part 2 to indicate the reason.

Revision 0 Facility: Waterford 3 Page 6 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000011 EA2.03 Importance Rating 3.7 K/A Statement 011 Large Break LOCA, EA2.03: Ability to determine or interpret the following as they apply to a Large Break LOCA: Consequences of managing LOCA with loss of CCW.

Proposed Question: RO 4 Rev: 0 Given:

A large break LOCA occurred A SIAS, CIAS, MSIS and CSAS has occurred Essential Chillers A and B are in service The crew has split CCW headers due to only one CCW pump available The crew has ___(1)___ train(s) of Auxiliary Component Cooling Water to manage the event.

The crew has ___(2)___ train(s) of Low Pressure Safety Injection available to manage the event.

(1) (2)

A. two two B. two one C. one one D. one two Revision 0 Facility: Waterford 3 Page 7 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Part 1 is correct. Only one train of LPSI is available because the C/S for one of the pumps is in OFF.

B. CORRECT: OP-902-009 directs the crew to remove the EDG, LPSI pump, HPSI Pump, and CS pump from service on the affected CCW train. The ACCW pump is not on the list therefore two trains are available. This is important because the next step in Appendix 35 is to regain an Essential Chiller using ACCW. Only one train of LPSI is available because the C/S for one of the pumps is in OFF.

C. Incorrect: OP-902-009 directs the crew to remove the EDG, LPSI pump, HPSI Pump, and CS pump from service on the affected CCW train. The ACCW pump is not on the list therefore two trains are available. Part 2 is correct.

D. Incorrect. OP-902-009 directs the crew to remove the EDG, LPSI pump, HPSI Pump, and CS pump from service on the affected CCW train. The ACCW pump is not on the list therefore two trains are available. Only one train of LPSI is available because the C/S for one of the pumps is in OFF.

Technical Reference(s): OP-902-009 Appendix 35 Rev. 315 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 8, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 8 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000015/17 AK2.07 Importance Rating 2.9 K/A Statement 015/17 RCP Malfunctions, AK2.07: Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: RCP seals.

Proposed Question: RO 5 Rev: 0 The ATC notes the following indications on RCP 1A:

• Vapor Seal is 45 PSIG

• Upper Seal is 1180 PSIG

• Middle Seal is 1250 PSI

• RCS pressure is 2250 PSIA Which ONE of the following RCP 1A seals has failed or is degraded?

A. Vapor B. Upper C. Middle D. Lower Revision 0 Facility: Waterford 3 Page 9 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Normal pressure range is 25 to 45 PSIG. Pressure is high in the band due to a failed middle seal.

B. Incorrect. Normal pressure range is 585 to 915 PSIG. Upper seal pressure is outside of the band which is indicative of a failed middle seal. The upper seal is working properly.

C. CORRECT: The D/P between the middle seal and upper seal pressures are less than 100 psid. The D/P is outside of the range given in OP-901-130. These parameters are indicative of a middle seal failure.

D. Incorrect. Normal Pressure is 1237 to 1815 psig for middle seal pressure. Pressure is low in the band due to the failed middle seal.

Technical Reference(s): OP-901-130 section E1 note Rev. 11 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank # RO4 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2009 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 3 55.43 Revision 0 Facility: Waterford 3 Page 10 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000022 AK1.03 Importance Rating 3.0 K/A Statement 022 Loss of Rx Coolant Makeup, AK1.03: Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: Relationship between charging flow and PZR level.

Proposed Question: RO 6 Rev: 0 Given:

Plant operating at 100% power Charging Pump B is running with Charging Pumps A and AB in standby Upon a trip of Charging Pump B, the first backup charging pump is designed to start at a pressurizer level of __(1)__ %. The crew will monitor __(2)__ heat exchanger outlet temperature for automatic isolation of letdown.

(1) (2)

A. 51.7 Regenerative B. 51.7 Letdown C. 53.1 Letdown D. 53.1 Regenerative Revision 0 Facility: Waterford 3 Page 11 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: The PLCS is designed such that the first Backup Charging Pump will start at a Pressurizer level deviation of 2.5%. The second Charging Pump will start at a level deviation of 3.9% which would equate to a level of 51.7% at 100% power. Part 2 is correct.

B. Incorrect. The PLCS is designed such that the first Backup Charging Pump will start at a Pressurizer level deviation of 2.5%. The second Charging Pump will start at a level deviation of 3.9% which would equate to a level of 51.7% at 100% power. A loss of Charging flow will raise Regenerative HX outlet temperature. If temperature reaches 470ºF, letdown will isolate. The letdown HX is downstream of the Regenerative HX and is cooled by CCW.

C. Incorrect. Part 1 is correct. A loss of Charging flow will raise Regenerative HX outlet temperature. If temperature reaches 470ºF, letdown will isolate. The letdown HX is downstream of the Regenerative HX and is cooled by CCW.

D. CORRECT: The PLCS is designed such that the first Backup Charging Pump will start at a Pressurizer level deviation of 2.5%. The applicant must know that Pressurizer level at 100% power is programmed at 55.6%. Charging flow is the cooling medium for the Regenerative HX. Therefore, a loss of Charging flow will raise Regenerative HX outlet temperature. If temperature reaches 470ºF, letdown will isolate.

Technical Reference(s): OP-901-112 pages 5 and 8. Revision 6 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CVC00 objs. 3, 6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 12 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000025 2.1.27 Importance Rating 3.9 K/A Statement 025 Loss of RHR System, 2.1.27: Knowledge of system purpose and/or function.

Proposed Question: RO7 Rev: 0 Given:

• The plant is in Mode 5 with Shutdown Cooling Train A is in service

• Component Cooling Water (CCW) flow has been lost to Shutdown Cooling Train A due to a loss of CCW pumps The crew will implement which of the following procedures?

A. OP-901-131, Shutdown Cooling Malfunction only B. OP-901-510, Component Cooling Water System Malfunction only C. OP-901-131, Shutdown Cooling Malfunction and OP-901-510, Component Cooling Water System Malfunction concurrently D. OP-901-131, Shutdown Cooling Malfunction and the Emergency Operating procedures concurrently Revision 0 Facility: Waterford 3 Page 13 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Plausible if the applicant is unaware that OP-901-131 does not contain all of the steps required to restore SDC heat removal capability, which is why it states to perform both procedures concurrently.

B. Incorrect. Plausible if the applicant is unaware that OP-901-131 has a section devoted to a loss of heat removal capability C. CORRECT: The purpose of OP-901-131 (section A of OP-901-131) is to provide guidance on a loss of flow, system leakage, loss of cooling, and a system malfunction in Mode 4. On a loss of CCW flow to a SDC heat exchanger, both OP-901-131 and OP-901-510 has a step to perform each other concurrently.

D. Incorrect. If the plant was in Mode 4 with a CS Train operable, the EOPs would be performed concurrently as indicated in the system malfunction in Mode 4 section of OP-901-131. Containment Spray and SDC share heat exchangers so if both SDC trains are in service, CS cannot be in service.

Technical Reference(s): OP-901-131 section A and section E3, Rev. 304 (Attach if not previously provided) OP-901-510 page 29 Rev. 303 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ21 obj. 1 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 14 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000027 2.2.44 Importance Rating 4.2 K/A Statement 027 Pressurizer Pressure Control System Malfunction, 2.2.44: 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.

Proposed Question: RO 8 Rev: 0 Given:

Plant is at 100% power RC-301A, Pressurizer Spray Valve indicates open Pressurizer Pressure is 2230 psia and dropping In accordance with OP-901-120, Pressurizer Pressure Control Malfunction, the CRS will direct the ATC to place the ____(1)____ controller in manual to restore pressurizer pressure. The ATC will then place the spray valve selector switch to the _____(2)_____

position.

(1) (2)

A. Pressurizer Pressure 1A B. Pressurizer Pressure 1B C. Pressurizer Spray 1B D. Pressurizer Spray 1A Revision 0 Facility: Waterford 3 Page 15 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. OP-901-120 directs the crew to take the Pressurizer Spray controller to manual and adjust output to 0%. The procedure directs the crew to take manual control of the Pressurizer pressure controller for other PPC malfunctions, but not for a failed open spray valve. The spray valve selector switch will de-energize the solenoid for the position it is not selected to.

B. Incorrect. OP-901-120 directs the crew to take the Pressurizer Spray controller to manual and adjust output to 0%. The procedure directs the crew to take manual control of the Pressurizer pressure controller for other PPC malfunctions, but not for a failed open spray valve. Part 2 is correct.

C. CORRECT: The applicant must determine that the spray valve should not be open with Pressurizer pressure at 2230 psia. OP-901-120 directs the crew to take the Pressurizer Spray controller to manual and adjust output to 0%. OP-901-120 will then direct the crew to take the spray valve selector switch to the 1B position which will de-energize the 1A spray valve.

D. Incorrect. Part 1 is correct. The spray valve selector switch will de-energize the solenoid for the position it is not selected to.

Technical Reference(s): OP-901-120 section E0 and E3 Rev. 302 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 7,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 16 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000029 EA1.12 Importance Rating 4.1 K/A Statement 029 ATWS, EA1.12: Ability to operate and monitor the following as they apply to a ATWS: M/G set power supply and reactor trip breakers.

Proposed Question: RO 9 Rev: 0 Given:

Plant is at 100% power An event occurs and automatic reactor trips fail Manual Reactor trip at CP-2 fails Manual Diverse Reactor Trip fails The crew will attempt to de-energize the Motor Generator Sets by opening the

___(1)___ feeder breakers. If this method of tripping the reactor is unsuccessful, the ATC will direct an NAO to the ___(1)___ safety switchgear room to locally open all Reactor Trip breakers.

(1) (2)

A. 31A and 31B A B. 31A and 31B B C. 32A and 32B A D. 32A and 32B B Revision 0 Facility: Waterford 3 Page 17 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The 31A(B) bus feeder breakers have control switches next to the 32A(B) feeder breakers on CP-1, but they do not supply power to the MG sets. The A Switchgear is located on the same elevation as the B switchgear but is not where the reactor trip switchgear breakers are located.

B. Incorrect:. The 31A(B) bus feeder breakers have control switches next to the 32A(B) feeder breakers on CP-1, but they do not supply power to the MG sets. Part 2 is correct.

C. Incorrect. Part 1 is correct. The A Switchgear is located on the same elevation as the B switchgear but is not where the reactor trip switchgear breakers are located.

D. CORRECT: If all attempts to trip the reactor at CP-2 fail, the Standard Post Trip actions direct the crew to de-energize the MG sets by opening the 32A and 32B feeder breakers. If this fails, the crew will direct an NAO to the B switchgear room to operate and monitor all Reactor Trip breakers open.

Technical Reference(s): SD-480 Figure 5 Rev. 5 (Attach if not previously provided) OP-902-000 step 1 Rev 16 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CED00 objs. 3 and 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6,10 55.43 Revision 0 Facility: Waterford 3 Page 18 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000038 EK3.02 Importance Rating 4.4 K/A Statement 038 Steam Gen. Tube Rupture, EK3.02: Knowledge of the reasons for the following responses as they apply to the SGTR: Prevention of secondary PORV cycling.

Proposed Question: RO 10 Rev: 0 Given:

A Steam Generator Tube Rupture and Loss of Offsite Power is in progress The crew is cooling down the RCS to less than 520ºF Thot in accordance with OP-902-007, Steam Generator Tube Rupture The crew will perform the cooldown using ____(1)____ Atmospheric Dump Valve(s).

Operation of the ADV(s) is more desirable than operation of the Main Steam Safety Valves because the ADV(s) is a(n) _____(2)_____ release path.

(1) (2)

A. one monitored B. both isolable C. one isolable D. both monitored Revision 0 Facility: Waterford 3 Page 19 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Step 11 of OP-902-007 has the crew use both ADVs to cooldown the RCS to less than 520ºF Thot. ONE ADV is plausible because after the initial cooldown, the crew is directed to use the ADV on the least affected generator (OP-901-202) to continue the cooldown. The ADV is considered an unmonitored release path at W3.

B. CORRECT: The MSSV is the W3 equivalent to the secondary PORV mentioned in the K/A.

Step 11 of OP-902-007 has the crew using both ADVs to cooldown the RCS to less than 520ºF Thot. The Tech Guide for this same step states that the ADV is more controlled because it can be manually operated and is isolable.

C. Incorrect. Step 11 of OP-902-007 has the crew use both ADVs to cooldown the RCS to less than 520ºF Thot. ONE ADV is plausible because after the initial cooldown, the crew is directed to use the ADV on the least affected generator (OP-901-202) to continue the cooldown. Part 2 is correct.

D. Incorrect. Part 1 is correct. . The ADV is considered an unmonitored release path at W3.

Technical Reference(s): OP-902-007 step 11 Rev. 17 (Attach if not previously provided) TG-OP-902-007 step 11 Rev. 308 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE07 obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 10,12 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 20 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000040 CE/E05 EK1.2 Importance Rating 3.2 K/A Statement CE/E05 Steam Line Rupture, EK1.2: Knowledge of the operational implications of the following concepts as they apply to the (Excess Steam Demand): Normal, abnormal and emergency operating procedures associated with (Excess Steam Demand).

Proposed Question: RO 11 Rev: 0 OP-902-004, Excess Steam Demand Recovery Procedure, requires verifying no more than two Reactor Coolant Pumps running if Tcold is less than ___(1)___ °F to

__________(2)___________.

(1) (2)

A. 380 prevent generating excessive core uplift forces B. 380 aid in stabilizing Reactor Coolant System temperature after Steam Generator dryout C. 202 prevent generating excessive core uplift forces D. 202 aid in stabilizing Reactor Coolant System temperature after Steam Generator dryout Revision 0 Facility: Waterford 3 Page 21 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Per OP-902-004, Tc < 380 requires 1 RCP in each loop to be secured.

TG-OP-902-004 states that the PT limit curve specifies that one RCP should be tripped although OP-902-004 directs the crew to secure one RCP in each loop for even flow between the loops. The reason for this step identified in the Tech Guide is to prevent a core uplift problem.

B. Incorrect. The temperature is correct. The basis is incorrect per TG-OP-902-004; however it is plausible that elimination of heat addition to the RCS would aid in stabilizing RCS temperature since Waterfords ADVs are not sufficient size to prevent needing to have a minimum EFW flow to aid in RCS temperature stabilization.

C. Incorrect. The temperature is incorrect. The temperature of 202°F is the minimum temperature for three pump configuration per OP-001-002. The basis is correct per TG-OP-902-004.

D. Incorrect. The temperature is incorrect. The temperature of 202°F is the minimum temperature for three pump configuration per OP-001-002. The basis is incorrect per OP-902-004; however it is plausible that elimination of heat addition to the RCS would aid in stabilizing RCS temperature since Waterfords ADVs are not sufficient size to prevent needing to have a minimum EFW flow to aid in RCS temperature stabilization Technical Reference(s): OP-902-004 step 10 Rev. 16 (Attach if not previously provided) TG-OP-902-004 Rev. 307 (including version/revision number) OP-001-002 limitation 3.1.6 Rev. 22 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE04 obj. 4 (As available)

Question Source: Bank # RO11 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Revision 0 Facility: Waterford 3 Page 22 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000054 CE/E06 EK1.1 Importance Rating 3.2 K/A Statement CE/E06 Loss of Main Feedwater, EK1.1: Knowledge of the operational implications of the following concepts as they apply to the (Loss of Feedwater):

Components, capacity, and function of emergency systems.

Proposed Question: RO 12 Rev: 0 Given:

The reactor is tripped due to an Inadvertent Main Steam Isolation Signal (MSIS)

Emergency Feedwater (EFW) Pump B is danger tagged out Both Steam Generators are 43% NR level and dropping A subsequent loss of ____(1)____ would require that all running Reactor Coolant Pumps be tripped. This action is required to ____(2)____.

(1) (2)

A. EFW Pump A prevent lifting Pressurizer safety valves B. EFW Pump AB reduce the heat input into the RCS C. EFW Pump AB prevent lifting Pressurizer safety valves D. EFW Pump A reduce the heat input into the RCS Revision 0 Facility: Waterford 3 Page 23 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. EFW Pump AB supplies 100% capacity to the Steam Generators, therefore, the crew will not be required to enter OP-902-006 if EFW Pump AB is the only EFW Pump running. The applicant must determine that OP-902-006 entry is required. Part 2 is plausible because RCS temperature/pressure will rise in an inadvertent MSIS and it can be assumed that securing RCPs will reduce the pressure rise and prevent lifting Pressurizer safeties. Per the Tech guide, this is not the reason RCPs are secured during a loss of Main Feedwater event.

B. CORRECT: EFW motor driven pumps (EFW Pumps A and B) are 50% capacity. Therefore, if the only feedwater available is ONE motor driven EFW pump, all RCPs must be secured to reduce the heat input into the RCS. The applicant must determine that OP-902-006 entry is required The inadvertent MSIS in the stem eliminated FW sources from MFW pumps and the AFW pump.

C. Incorrect. Part 1 is correct. Part 2 is plausible because RCS temperature/pressure will rise in an inadvertent MSIS and it can be assumed that securing RCPs will reduce the pressure rise and prevent lifting Pressurizer safeties. Per the Tech guide, this is not the reason RCPs are secured during a loss of Main Feedwater event.

D. Incorrect. EFW Pump AB supplies 100% capacity to the Steam Generators, therefore, the crew will not be required to enter OP-902-006 if EFW Pump AB is the only EFW Pump running. The applicant must determine that OP-902-006 entry is required. Part 2 is correct.

Technical Reference(s): TG-OP-902-006 step 6 Rev. 18 (Attach if not previously provided) OP-902-006 step 6 Rev. 18 SD-EFW page 6, Rev. 12, OP-902-009 Att. 1 Rev.

(including version/revision number) 315 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE06 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 24 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000055 EA2.03 Importance Rating 3.9 K/A Statement 055 Station Blackout, EA2.03: Ability to determine or interpret the following as they apply to a Station Blackout: Actions necessary to restore power.

Proposed Question: RO 13 Rev: 0 Given:

A Station Blackout has occurred Offsite power has been restored to buses 3A and 3B A Probable Maximum Precipitation (PMP) event just commenced To restore a DCT Sump Pump, power must be restored to the non-safety section of Motor Control Center _____(1)_____.

OP-902-009 Appendix 20, Operation of DCT Sump Pumps, directs the crew to align at least one Dry Cooling Tower Motor Driven Sump Pump within ___(2)___ of the PMP event.

(1) (2)

A. 312A or B 30 minutes B. 314A or B 30 minutes C. 312A or B 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> D. 314A or B 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Revision 0 Facility: Waterford 3 Page 25 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Wrong MCCs, however, the 312 MCCs also have safety to non-safety tie breakers.

B. CORRECT: Per OP-902-009, requires restoration of a DCT Motor Driven Sump Pump within 30 minutes from the start of a PMP event. These pumps are powered from MCCs 314A and 314B.

C. Incorrect. Wrong MCCs, however, the 312 MCCs also have safety to non-safety tie breakers Wrong time frame. Three hours correlates to the time required to have a diesel driven sump pump aligned for operation from the DCT sumps.

D. Incorrect. Correct MCCs. Three hours correlates to the time required to have a diesel driven sump pump aligned for operation from the DCT sumps.

Technical Reference(s): OP-902-005 step 5 Rev. 20 (Attach if not previously provided) OP-902-009 App. 20 Rev. 315 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE05 obj. 7 (As available)

Question Source: Bank # X Question #13 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 26 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000056 2.4.18 Importance Rating 3.3 K/A Statement 056 Loss of Off-site Power, 2.4.18: Knowledge of the specific bases for the EOPs.

Proposed Question: RO 14 Rev: 0 Given:

Loss of offsite power has occurred Crew has entered OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery The sequencers for both Emergency Diesel Generators are timed out The CRS has directed the BOP to Verify Containment Cooling The BOP will ____(1)____ . This action is required to ____ (2) .

(1) (2)

A. secure one Containment Fan ensure emergency diesel Cooler generator fuel oil consumption is minimized B. align the Containment Fan ensure emergency diesel Coolers to slow speed generator fuel oil consumption is minimized C. align the Containment Fan prevent damage to Containment Coolers to slow speed Cooling ductwork D. secure one Containment Fan prevent damage to Containment Cooler Cooling ductwork Revision 0 Facility: Waterford 3 Page 27 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Part 1 is correct. Securing one CFC will help EDG fuel oil consumption, but with no SIAS, fuel oil consumption is not a concern.

B. Incorrect. The only time four CFCs should be running together in slow speed is following a SIAS or surveillance testing. Neither of these events/evolutions are in progress, so going to slow speed is not required. Aligning the CFCs to slow speed will help EDG fuel oil consumption, but with no SIAS, fuel oil consumption is not a concern.

C. Incorrect. The only time four CFCs should be running together in slow speed is following a SIAS or surveillance testing. Neither of these events/evolutions or in progress, so going to slow speed is not required. Part 2 is correct.

D. CORRECT: Following a LOOP without a SIAS, all CFCs will start in Fast speed. If all four CFCs start in Fast there is a concern for damaging duct work.

Technical Reference(s): OP-902-003 step 13 Rev. 10 (Attach if not previously provided) TG-OP-902-003 step 13 Rev. 306 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE05 obj. 6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8,10 55.43 Revision 0 Facility: Waterford 3 Page 28 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000058 AK3.01 Importance Rating 3.4 K/A Statement 058 Loss of DC Power, AK3.01: Knowledge of the reasons for the following responses as they apply to the Loss of DC Power: Use of dc control power by D/Gs.

Proposed Question: RO 15 Rev: 0 Given:

A loss of the 125 Volt DC bus A-DC has occurred The BOP will direct the NAO to secure the Emergency Diesel Generator (EDG) A Fuel Oil Transfer pump by ____(1)____. This action is required to ____(2)____.

(1) (2)

A. taking the local control prevent overflowing the Fuel Oil Day Tank switch to off B. taking the local control prevent pump overheating switch to off C. opening the pump supply prevent pump overheating breaker D. opening the pump supply prevent overflowing the Fuel Oil Day Tank breaker Revision 0 Facility: Waterford 3 Page 29 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect: On a loss of DC, the Fuel Oil Day Tank level switch fails low, causing the fuel oil transfer pump to run continuously. The local switch is an auto/on switch, therefore the pump cannot be secured via the c/s. Part 2 is correct.

B. Incorrect: On a loss of DC, the Fuel Oil Day Tank level switch fails low, causing the fuel oil transfer pump to run continuously. The local switch is an auto/on switch, therefore the pump cannot be secured via the c/s. The fuel oil transfer pump requires 10 gpm recirc flow to prevent overheating. The discharge flowpath is not isolated during normal or abnormal operations, therefore overheating is not a reason to secure the pump in this instance.

C. Incorrect. Part 1 is correct. The fuel oil transfer pump requires 10 gpm recirc flow to prevent overheating. The discharge flowpath is not isolated during normal or abnormal operations, therefore overheating is not a reason to secure the pump in this instance D. CORRECT: The loss of DC off-normal, OP-901-313, directs the crew to open the EDG A Fuel Oil Transfer Pump breaker, EGF-312A-3F on a loss of A DC. This same step states that this action is performed to prevent overflowing the Fuel Oil Day Tank.

Technical Reference(s): OP-901-313 page 19 Rev. 305 (Attach if not previously provided) SD-EDG page 70 Rev. 24 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO30 obj. 3 (As available)

Question Source: Bank #

Modified Bank # RO52 (Note changes or attach parent)

New Question History: Last NRC Exam 2007 RO Makeup Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Revision 0 Facility: Waterford 3 Page 30 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000026 AA2.01 Importance Rating 2.9 K/A Statement 026 Loss of Component Cooling Water, AA2.01: Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: Location of a leak in the CCW System.

Proposed Question: RO 16 Rev: 1 Given:

A leak is occurring in the Component Cooling Water System Initially CCW Surge Tank Level A and B lowered to 0%

CCW Surge Tank level A remains at 0%

CCW Surge Tank level B recovered to ~ 55% and stabilized Which of the following is a potential location of the leak?

A. Supply pipe to Waste Gas Compressor A B. Supply pipe to Shutdown Cooling HX A C. Return pipe from Spent Fuel Pool HX A D. Return pipe from CEDM Fan Cooler A Revision 0 Facility: Waterford 3 Page 31 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. The Waste Gas compressor is an NNS Loop load. This pipe would have been automatically isolated by the AB loop isolation valves and level in both sides of the CCW surge tank would have recovered.

B. CORRECT: Shutdown HX A is a Safety Loop A load. Safety Loops A and B would be split out such that the leak no longer affects Loop B until CCW Surge Tank Level B exceeds 55% which is the top of the baffle separating A and B sides of the CCW surge tank.

C. Incorrect. SFP HX A is an AB Loop load. This pipe would have been automatically isolated by the AB Loop isolation valves and level in both sides of the CCW surge tank would have recovered.

D. Incorrect. CEDM Fan Cooler A is an AB Loop load. This pipe would have been automatically isolated by the AB Loop isolation valves and level in both sides of the CCW surge tank would have recovered Technical Reference(s): OP-901-510 page 6 Rev. 303 (Attach if not previously provided) SD-CC Fig. 2 Rev. 21 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank # RO 45 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 7 55.43 Comments: Free review question. Randomly selected a new K/A.

Revision 0 Facility: Waterford 3 Page 32 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000065 AA1.01 Importance Rating 2.7 K/A Statement 065 Loss of Instrument Air, AA1.01: Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air: Remote manual loaders.

Proposed Question: RO 17 Rev: 0 Given:

Plant is at 100% power An instrument air leak has occurred Crew has entered OP-901-511, Instrument Air Malfunction The BOP will direct the NAO to ____(1)____ the set point for SA-125, SA Backup Supply for IA Press Cntl Valve, to force it open. When instrument air pressure lowers to 95 psig, the crew will _____(2)_____ .

(1) (2)

A. raise verify IA-123, Instrument Air Dryers Bypass solenoid, opens B. raise commence a plant shutdown in accordance with OP-010-005, Plant Shutdown C. lower verify IA-123, Instrument Air Dryers Bypass solenoid, opens D. lower commence a plant shutdown in accordance with OP-010-005, Plant Shutdown Revision 0 Facility: Waterford 3 Page 33 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: The set point on SA-125 must be raised to force the valve open. The set point is the pressure at which the valve will start to open. Usually set at 105 psig.

Raising the set point will make the valve open sooner. IA-123 is designed to open and bypass the IA dryers at 95 psig.

B. Incorrect. Part 1 is correct. The crew will commence a plant shutdown in accordance with OP-010-005, Plant Shutdown when Instrument Air pressure reaches 80 psig.

C. Incorrect. The set point on SA-125 must be raised to force the valve open. The set point is the pressure at which the valve will start to open. Usually set at 105 psig.

Raising the set point will make the valve open sooner. Part 2 is correct.

D. Incorrect. The set point on SA-125 must be raised to force the valve open. The set point is the pressure at which the valve will start to open. Usually set at 105 psig.

Raising the set point will make the valve open sooner. The crew will commence a plant shutdown in accordance with OP-010-005, Plant Shutdown when Instrument Air pressure reaches 80 psig.

Technical Reference(s): OP-901-511 page 6 and 7 Rev. 15 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 34 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000077 AK2.07 Importance Rating 3.6 K/A Statement 077 Generator Voltage and Electric Grid Disturbances, AK2.07: Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Turbine/Generator Control.

Proposed Question: RO 18 Rev: 0 Given:

Reactor power is 100%

A large load connected to the grid is lost causing a grid disturbance MVAR load changed from 60 MVAR out to 100 MVAR in The grid disturbance caused grid voltage to __(1)__.

To restore MVAR load to 60 MVAR out, the crew will adjust the Main Generator voltage adjust regulator control switch to _______(2)______.

(1) (2)

A. rise raise B. rise lower C. lower raise D. lower lower Revision 0 Facility: Waterford 3 Page 35 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Losing a large load on the grid will cause grid voltage to rise because Reactive Power (VAR) decreases. To change MVAR from leading to lagging the voltage regulator C/S must be taken to raise.

B. Incorrect. Part 1 is correct. To change MVAR from leading to lagging the voltage regulator C/S must be taken to raise.

C. Incorrect. Losing a large load on the grid will cause grid voltage to rise. Part 2 is correct.

D. Incorrect. Losing a large load on the grid will cause grid voltage to rise. To change MVAR from leading to lagging the voltage regulator C/S must be taken to raise.

Technical Reference(s): WLP-OPS-TYC05 slide 247, 257, 258, 275 Rev. 4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-TYC05 obj. 20 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 5, 10 55.43 Revision 0 Facility: Waterford 3 Page 36 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # CE/A16 AK1.1 Importance Rating 3.2 K/A Statement CE/A16 Excess RCS Leakage, AK1.1: Knowledge of the operational implications of the following concepts as they apply to the (Excess RCS Leakage):

Components, capacity, and function of emergency systems.

Proposed Question: RO 19 Rev: 0 Given:

Plant is at 100% power AB Charging pump is out of service with Charging Pumps A and B running An RCS leak is in progress and the crew is performing the required actions of OP-901-111, RCS Leak Subsequently, the STA reports that RCS leakage has been verified to be 90 gpm Which of the following is the appropriate action(s) to be performed?

A. Remain in OP-901-111, Reactor Coolant System Leak, and attempt to locate the leak.

B. Commence a normal shutdown in accordance with OP-010-005, Plant Shutdown.

C. Commence a rapid plant shutdown in accordance with OP-901-212, Rapid Plant Power Reduction.

D. Initiate a manual reactor trip, SIAS/CIAS, and go to OP-902-000, Standard Post Trip Actions.

Revision 0 Facility: Waterford 3 Page 37 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. RCS leakage exceeds the capacity of two charging pumps. This is an option only if pressurizer level can be maintained (leak within the capacity of two charging pumps.

B. Incorrect: OP-901-111 step 17 directs a normal shutdown, if required, in accordance with OP-010-005. In this case, RCS leakage exceeds the capacity of the Charging Pumps, therefore a Reactor trip is required.

C. Incorrect. RCS leakage exceeds the capacity of two charging pumps. This is a valid shutdown option in other conditions; however, OP-901-111 specifically ignores this option based on the possible inability to control pressurizer level on the shutdown.

D. CORRECT: The Charging Pump capacity is 44 gpm per pump. With two Charging Pumps running, there is 88 gpm capacity. Based on the total inventory loss from the RCS exceeding Charging pump capacity, Pressurizer level cannot be maintained.

OP-901-111 requires a reactor trip and a manual initiation of SIAS/CIAS be performed.

Technical Reference(s): OP-901-111 page 7 Rev. 303 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank # RO65 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 38 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000024 AK2.01 Importance Rating 2.7 K/A Statement 024 Emergency Boration, AK2.01: Knowledge of the interrelations between Emergency Boration and the following: Valves.

Proposed Question: RO 20 Rev: 0 Given:

Plant is in MODE 3 Emergency Diesel Generator A is Danger Tagged out Plant has experienced a steam line leak resulting in an uncontrolled cooldown Loss of power to the 3A safety bus has occurred CRS directs ATC operator to commence emergency boration Which of the following will be the emergency boration flow path aligned by the ATC?

A. BAM Tanks via BAM-143, Direct Boration Valve.

B. BAM Tanks via BAM-113 A and B, Boric Acid Gravity Feed Valves.

C. Boric Acid Makeup (BAM) Tank B via BAM Pump B and BAM-133, Emergency Boration Valve.

D. Refueling Water Storage Pool (RWSP) via CVC-507, RWSP to Charging Pump Suction Isolation.

Revision 0 Facility: Waterford 3 Page 39 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. BAM-143 is the direct boration valve from the BAMTs. BAM-143 is not an optional emergency boration flowpath per OP-901-103, Emergency Boration..

B. CORRECT: The loss of the 3A bus resulted in a loss of the BAM pumps. Emergency boration must be aligned from the BAMTs via BAM-113A and B (both gravity feed valves are B train powered).

C. Incorrect. The loss of the 3A bus resulted in a loss of both Boric Acid Makeup (BAM) pumps (both BAM pumps are A train powered). Emergency boration must be aligned from the BAMTs via BAM-113A and B (both gravity feed valves are B train powered).

D. Incorrect. Emergency boration from the RWSP is only aligned when the BAMTs are

< 15%.

Technical Reference(s): OP-901-103, Emergency Boration, Rev. 3 OP-002-005, Chemical And Volume Control (Attach if not previously provided) System, Rev. 56 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CVC00 obj. 3,5 (As available)

Question Source: Bank # RO20 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2012 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 6,10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 40 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000028 AA2.12 Importance Rating 3.1 K/A Statement 028 Pressurizer Level Malfunction, AA2.12: Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: Cause for PZR level deviation alarm: controller malfunction or other instrumentation malfunction.

Proposed Question: RO 21 Rev: 0 Given:

Plant is at 100% power The Pressurizer Level Control Channel Select Switch is in the Channel Y position The ATC reports that the Pressurizer Level Hi/Lo and Pressurizer Level Hi/Hi annunciators are in alarm One Charging Pump is running Letdown flow is 40 gpm and steady This event could be caused by which of the following failing high?

A. Reactor Regulating System RCS temperature Loop 1 hot leg indicator B. Pressurizer Level Control Channel X C. Pressurizer Level Control Channel Y D. Pressurizer Level Controller output Revision 0 Facility: Waterford 3 Page 41 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. At 100% power, the pressurizer programmed level setpoint cannot go greater than 55.6%. The hot leg indicator is an input to the level setpoint but since the cap for pressurizer level is 55.6%, the setpoint is not affected.

B. CORRECT: The annunciators given in the stem will still annunciate on the non-selected channel failing high because the annunciator inputs are located upstream of the channel select switch. Since channel X is the non-selected channel, the inputs to Charging and Letdown are not affected.

C. Incorrect. Pressurizer Level Channel Y failing high would bring in the annunciators located in the stem but would also affect Charging and Letdown. Charging pumps running would remain at one running but letdown flow would go to maximum. The stem indicated normal letdown flow and steady.

D. Incorrect. The Pressurizer Level controller failing high would not affect Charging Pumps initially but would raise letdown flow. Letdown flow in the stem indicates normal letdown flow and steady.

Technical Reference(s): SD-PLC page 25 Rev. 10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 42 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 051 AK3.01 Importance Rating 2.8 K/A Statement 051 Loss of Condenser Vacuum, AK3.01: Knowledge of the reasons for the following responses as they apply to the Loss of Condenser Vacuum and the following: Loss of steam dump capability upon loss of condenser vacuum.

Proposed Question: RO 22 Rev: 0 Given:

A Loss of offsite power has occurred and the crew is performing actions of OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery Condenser Vacuum has lowered to less than 14 INHG The BOP will close the Main Steam Isolation Valves to ensure ______(1)______.

The Steam Bypass Control Valves are designed to close at a Condenser vacuum of

__(2)__ INHG.

(1) (2)

A. steam is isolated to the Main 6.4 Feedwater Pumps B. steam is isolated to the Main 3.4 Feedwater Pumps C. the Main Condenser is protected 3.4 from over-pressurization D. the Main Condenser is protected 6.4 from over-pressurization Revision 0 Facility: Waterford 3 Page 43 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The Main Feed Pumps (MFPs) are designed to trip at 14 INHG. The step for Protecting the Main Condenser is also performed at 14 INHG. Isolating MSIVs will isolate steam to the MFPs, which makes this answer plausible but not the reason we close MSIVs. The SBCVs are designed to auto close at 3.4 INHG. The reset value that the SBCVs will come back open is a Main Condenser Vacuum of 6.4 INHG B. Incorrect. The Main Feed Pumps (MFPs) are designed to trip at 14 INHG. The step for Protecting the Main Condenser is also performed at 14 INHG. Isolating MSIVs will isolate steam to the MFPs, which makes this answer plausible but not the reason we close MSIVs. Part 2 is correct.

C. CORRECT: The Main Steam Isolation Valves (MSIVs) are closed to isolate steam going through the Steam Bypass Control Valves (SBCVs) to the main condenser.

This action is performed in Step 9 of OP-902-003, (Protect the Main Condenser) and the tech guide for this step states that the reason is to protect the Main Condenser from over-pressurization. The SBCVs are designed to auto close at 3.4 INHG.

D. Incorrect. Part 1 is correct. The SBCVs are designed to auto close at 3.4 INHG. The reset value that the SBCVs will come back open is a Main Condenser Vacuum of 6.4 INHG.

Technical Reference(s): OP-902-003 step 9 Rev. 10 (Attach if not previously provided) TG-OP-902-003 step 9 Rev. 306 (including version/revision number) OP-500-005 Att. 4.63 Rev. 14 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE05 obj. 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 44 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000037 AK1.02 Importance Rating 3.5 K/A Statement 037 Steam Generator Tube Leak, AK1.02: Knowledge of the operational implications of the following concepts as they apply to Steam Generator Tube Leak: Leak rate vs. pressure drop Proposed Question: RO 23 Rev: 1 With a Steam Generator Tube leak in progress it is ultimately desired to depressurize the RCS to within ___(1)___ psid of the ruptured Steam Generator to ___(2)___ and minimize the potential for Steam Generator overfill.

(1) (2) minimize the potential release to the A. 50 environment 50 prevent a loss of subcooled margin B.

minimize the potential release to the C. 100 environment prevent a loss of subcooled margin D. 100 Revision 0 Facility: Waterford 3 Page 45 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Per step 21.5 of OP-901-202, 50 psid is the correct pressure differential. This minimizes pri-to-sec leakage which minimizes release magnitude and helps to prevent S/G overfill. The basis for 50 psid is located in TG-OP-902-007. The first bullet of this same step (maintain pressure within the PT curves) is the action required to maintain subcool margin.

B. Incorrect. Part 1 is correct. The bullet within step 21.5 provides instructions for maintaining RCS pressure within the P/T curve limits prevents loss of subcooled margin and allows continued operation of RCPs.

C. Incorrect. Per step 21.5 of OP-901-202, 50 psid is the correct pressure differential. Part 2 is correct.

D. Incorrect. Per step 21.5 of OP-901-201, 50 psid is the correct pressure differential. The substep within step 12 that provides instructions for maintaining RCS pressure within the P/T curve limits prevents loss of subcooled margin and allows continued operation of RCPs.

Technical Reference(s): OP-901-202 page 18 Rev. 15 (Attach if not previously provided) TG-OP-902-007 page 33 Rev. 308 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO2 obj. 3 (As available)

Question Source: Bank # RO10 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2014 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5, 10 55.43 Comments:

Free review question. Developed a new question with the same K/A.

Revision 0 Facility: Waterford 3 Page 46 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000060 2.4.31 Importance Rating 4.2 K/A Statement 060 Accidental Gaseous Radwaste Rel., 2.4.31: Knowledge of annunciator alarms, indications, or response procedures.

Proposed Question: RO 24 Rev: 0 Given:

Gas Decay Tank(s) discharge is in progress in accordance with OP-007-003, Gaseous Waste Management The RM-11 alarms and the following annunciator on CP-4 is received:

WASTE GAS DISCH RAD HIGH DRYER/MONITOR TROUBLE The ATC will verify the validity of this annunciator by checking Waste Gas Flow and Rad Recorder (GWMIFRR0648) located on __(1)__ .

If the GWM radiation monitor indicates a hi rad condition, the GDT discharge will __(2)__

(1) (2)

A. LCP-42 be isolated manually B. CP-4 be isolated manually C. LCP-42 auto isolate D. CP-4 auto isolate Revision 0 Facility: Waterford 3 Page 47 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. LCP-42 is the location for the Waste Gas Decay Tanks to Plant Vent Flow Indicator as mentioned in the ARP. This is not the indicator the off-normal directs use of to verify the validity of the annunciator. On high radiation during a GDT release, GWM-309 will automatically close. The off-normal procedure gives further direction to secure the release after verifying auto actions occur.

B. Incorrect. Part 1 is correct. On high radiation during a GDT release, GWM-309 will automatically close. The off-normal procedure gives further direction to secure the release after verifying auto actions occur.

C. Incorrect. LCP-42 is the location for the Waste Gas Decay Tanks to Plant Vent Flow Indicator as mentioned in the ARP. This is not the indicator the off-normal directs use of to verify the validity of the annunciator. Part 2 is correct.

D. CORRECT: From the annunciator response and off-normal procedure, the validity of this annunciator is verified by checking Waste Gas Flow and Rad Recorder (GWMIFRR0648). This recorder is located on CP-4. On high radiation during a GDT release, GWM-309 will automatically close.

Technical Reference(s): OP-500-007 Att. 4.50 Rev. 16 (Attach if not previously provided) OP-901-413, page 4 and 6 Rev. 2 (including version/revision number) SD-GWM page 13 Rev. 6 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO40 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10, 13 55.43 Revision 0 Facility: Waterford 3 Page 48 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000068 AK2.03 Importance Rating 2.9 K/A Statement 068 Control Room Evac., AK2.03 Knowledge of the interrelations between the Control Room Evacuation and the following: Controllers and positioners.

Proposed Question: RO 25 Rev: 0 Given:

A fire developed in the Control Room, requiring the Control Room to be evacuated The crew is performing OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown Prior to leaving the Control Room, the BOP will place ____(1)____ to manual and lower output to 0% and ____(2)____ .

(1) (2)

A. the ADV controller that manually initiate a Main has spuriously opened Steam Isolation Signal B. the ADV controller that close both Main Steam has spuriously opened Isolation Valves using its C/S C. both ADV controllers close both Main Steam Isolation Valves using its C/S D. both ADV controllers manually initiate a Main Steam Isolation Signal Revision 0 Facility: Waterford 3 Page 49 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Part 1 is correct. The MSIS would close the MSIVs. An MSIS is actuated later in the procedure, but not until controls are established at LCP-43.

B. CORRECT: If evacuating the control room due to a fire, the BOP immediate action is to take the ADV controller to manual and lower output to 0% for any ADV that is acting spuriously. The next action for the BOP is to close the MSIVs.

C. Incorrect. The action is to manually close the ADV if it is acting spuriously. The MSIS would close the MSIVs. Part 2 is correct.

D. Incorrect. The action is to manually close the ADV if it is acting spuriously. The MSIS would close the MSIVs. An MSIS is actuated later in the procedure, but not until controls are established at LCP-43.

Technical Reference(s): OP-901-502 step 2.4 Rev. 33 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO51 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 50 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000074 EA1.01 Importance Rating 4.2 K/A Statement 074 Inad. Core Cooling, EA1.01: Ability to operate and monitor the following as they apply to a Inadequate Core Cooling: RCS Water Inventory.

Proposed Question: RO 26 Rev: 0 Given:

Small Break LOCA occurred one hour ago Pressurizer level is 0%

RCS is saturated The Reactor Vessel Level Monitoring System can be monitored using QSPDS and recorders on ____(1)____.

To meet the requirements to perform Hot and Cold Leg Injection, the CRS will direct the ATC to report when vessel plenum level is less than 80%. The equivalent level will be indicated on the RVLMS QSPDS mimic when level ___(2)___ indicates voided.

(1) (2)

A. CP-8 6 B. CP-7 6 C. CP-7 5 D. CP-8 5 Revision 0 Facility: Waterford 3 Page 51 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. RVLMS can be read on recorders located on CP-7, a large portion of the accident indicators are located on CP-8, but not RVLMS. The applicant needs to be aware that RVLMS numbering starts at the top of the head is numbered from 1-8.

Level 6 is plausible since it is a level in the plenum that is not at the top or bottom.

B. Incorrect. Part 1 is correct. The applicant needs to be aware that RVLMS numbering starts at the top of the head is numbered from 1-8. Level 6 is plausible since it is a level in the plenum that is not at the top or bottom.

C. CORRECT: RVLMS can be read on recorders located on CP-7. Level 5 being voided is an indication that Reactor Vessel level is less than 80% in the plenum.

D. Incorrect. RVLMS can be read on recorders located on CP-7, a large portion of the accident indicators are located on CP-8, but not RVLMS. Level 5 being voided is an indication that Reactor Vessel level is less than 80% in the plenum.

Technical Reference(s): OP-902-002 step 53 Rev. 20 (Attach if not previously provided) SD-QSP page 13 Rev. 5 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02 obj. 17 (As available)

WLP-OPS-QSP00 obj. 6 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 52 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # CE/A11 AA2.2 Importance Rating 3.0 K/A Statement CE/A11 RCS Overcooling, AA2.2: Ability to determine and interpret the following as they apply to the (RCS Overcooling): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Proposed Question: RO 27 Rev: 0 Given:

A steam line rupture has occurred on Steam Generator #1 and the crew is performing the actions of OP-902-004, Excess Steam Demand Recovery Procedure RCS pressure is 1600 PSIA Containment pressure is 18 PSIA Steam Generator #1 has been isolated in accordance with OP-902-004, Excess Steam Demand Recovery Procedure The crew is ready to perform a controlled plant cooldown Which ONE of the following describes the method and maximum rate allowed to perform the RCS cooldown in accordance with OP-902-004?

Cooldown using A. ADVs at no greater than 50°F per hour B. SBCS at no greater than 100°F per hour C. SBCS at no greater than 50°F per hour D. ADVs at no greater than 100°F per hour Revision 0 Facility: Waterford 3 Page 53 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: The Steam Bypass Control System will be unavailable due to the MSIS signal (occurred at 17.1 psia). ADVs will be required. The cooldown limit for natural circulation conditions with an assymetric S/G is 50ºF/hr. An assymetric S/G cooldown occurs when one S/G is used for the cooldown and the other S/G is not being steamed by either the SBCS or the ADV. S/G #1 has been isolated per OP-902-004, therefore, the SBCS is not available and the ADV is in manual and closed.

B. Incorrect. The Steam Bypass Control System will be unavailable due to the MSIS signal (occurred at 17.1 psia). ADVs will be required. The cooldown limit for natural circulation conditions without an assymetric S/G is 100 ºF/hr.

C. Incorrect. The Steam Bypass Control System will be unavailable due to the MSIS signal (occurred at 17.1 psia). ADVs will be required. The cooldown limit for natural circulation conditions with an assymetric S/G is 50ºF/hr.

D. Incorrect. The Steam Bypass Control System will be unavailable due to the MSIS signal (occurred at 17.1 psia). ADVs will be required. The cooldown limit for natural circulation conditions without an assymetric S/G is 100 ºF/hr.

Technical Reference(s): TG-OP-902-004 step 49 revision 307 OP-902-004 steps 16, 49 and note prior to step 49 (Attach if not previously provided) Rev. 16 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE04 objs. 7 and 8 (As available)

Question Source: Bank # RO41 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2008 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 5, 10 55.43 Revision 0 Facility: Waterford 3 Page 54 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 003 K6.14 Importance Rating 2.6 K/A Statement 003 Reactor Coolant Pump, K6.14: Knowledge of the effect of a loss or malfunction on the following will have on the RCPS: Starting requirements.

Proposed Question: RO 28 Rev: 0 To prevent motor overload in excess of the nameplate horsepower rating, the minimum RCS cold leg temperature for RCP operation is 350ºF when operating in three pump configuration using RCP ____(1)____.

This limitation is applicable since this pump requires more brake horsepower because it

___(2)___ of the four reactor coolant pumps.

(1) (2)

A. 1A has the largest flywheel B. 2A has the largest flywheel C. 2A develops the highest head D. 1A develops the highest head Revision 0 Facility: Waterford 3 Page 55 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. RCP 1A is plausible because OI-038-000, Expectations for EOPs, has the crew trip RCP 1A and 2A first so that we dont exceed this limitation in the question and keep one RCP in each loop running. Plausible because RCPs have a flywheel and a larger one would require more horsepower.

B. Incorrect. Part 1 is correct. Plausible because RCPs have a flywheel and a larger one would require more horsepower.

C. CORRECT: This question is an administrative start interlock (starting the third pump if it is RCP 2A). Per OP-001-002, To prevent motor overload in excess of the nameplate horsepower rating, the minimum RCS cold leg temperature for RCP operation is 350ºF when operating in three pump configuration using RCP2A. This is because RCP 2A develops higher head and requires more horsepower. Important concept for the applicant to know since this is the reason for the securing sequence directed in the EOPs.

D. Incorrect. RCP 1A is plausible because OI-038-000, Expectations for EOPs, has the crew trip RCP 1A and 2A first so that we dont exceed this limitation in the question and keep one RCP in each loop running Technical Reference(s): OP-001-002 Precaution 3.1.5 Rev. 22 (Attach if not previously provided) OI-038-000 step 5.4.71 Rev 13 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-RCP00 obj. 10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 56 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 004 A4.07 Importance Rating 3.9 K/A Statement 004 Chemical and Volume Control, A4.07: Ability to manually operate and/or monitor in the control room: Boration/dilution.

Proposed Question: RO 29 Rev: 0 Given:

The plant is at 90% power VCT Level Instrument, CVC-ILT-0227 fails low The crew will verify the Charging Pump suction swaps to the __(1)__ and __(2)__ to match Tavg and Tref.

(1) (2)

A. Refueling Water Storage Pool lower turbine load B. Refueling Water Storage Pool raise turbine Load C. Boric Acid Makeup Tanks lower turbine load D. Boric Acid Makeup Tanks raise turbine Load Revision 0 Facility: Waterford 3 Page 57 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: CVC-ILT-0227 failing low will cause RWSP to Charging Pumps, CVC-507, to open and VCT to the Charging Pumps, CVC-183, to close. The RWSP is kept at higher boron concentration than the VCT so Tave will drop. The crew will be required to reduce Turbine load (less steam) to raise Tave.

B. Incorrect. Part 1 is correct. The RWSP is kept at higher boron concentration than the VCT so Tave will drop. The crew will be required to reduce Turbine load (less steam) to raise Tave.

C. Incorrect. CVC-ILT-0227 failing low will cause RWSP to Charging Pumps, CVC-507, to open and VCT to the Charging Pumps, CVC-183, to close. The BAMTs can be aligned to the Charging Pumps automatically (SIAS). Part 2 is correct.

D Incorrect. CVC-ILT-0227 failing low will cause RWSP to Charging Pumps, CVC-507, to open and VCT to the Charging Pumps, CVC-183, to close. The BAMTs can be aligned to the Charging Pumps automatically (SIAS). The RWSP is kept at higher boron concentration than the VCT so Tave will drop. The crew will be required to reduce Turbine load (less steam) to raise Tave.

Technical Reference(s): OP-901-113, step 1 Rev. 302 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 OBJ. 3 (As available)

Question Source: Bank #

Modified Bank # RO30 (Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 6, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 58 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 005 A4.02 Importance Rating 3.4 K/A Statement 005 Residual Heat Removal, A4.02: Ability to manually operate and/or monitor in the control room: Heat exchanger bypass flow control.

Proposed Question: RO 30 Rev: 0 Given:

The plant is in Mode 4 Shutdown Cooling Train A is in service RCS temperature is 204°F and steady Component Cooling Water temperature is dropping through the night as ambient temperature drops To prevent entry into Mode 5, the SDC Train A Temperature Control Valve (SI-415A) must be throttled ____(1)____. This will result in the SDC Train A Flow Control Valve (SI-129A) automatically throttling ______(2)______.

(1) (2)

A. open open B. open closed C. closed closed D. closed open Revision 0 Facility: Waterford 3 Page 59 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. To prevent entry into Mode 5 (RCS temp <200°F), less SDC flow is required through the SDC HX to compensate for CCW temperature dropping. This is done by closing the TCV (SI-415A). Part 2 is correct B. Incorrect. To prevent entry into Mode 5 (RCS temp <200°F), less SDC flow is required through the SDC HX to compensate for CCW temperature dropping. This is done by closing the TCV (SI-415A). To maintain flow constant, the FCV (SI-129A) will automatically open since there is now less flow going through the HX.

C. Incorrect. Part 1 is correct. SI-129A will open to maintain flow constant since there is now less flow going through the SDC HX.

D. CORRECT: To prevent entry into Mode 5 (RCS temp <200°F), less SDC flow is required through the SDC HX to compensate for CCW temperature dropping. This is done by closing the TCV (SI-415A). To maintain flow constant, the FCV (SI-129A) will automatically open since there is now less flow going through the HX. SI-129A is the bypass around Shutdown Cooling HX A.

Technical Reference(s): SD-SDC pp. 5,6,13 revision 8 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SDC00 obj 4 (As available)

Question Source: Bank # RO31 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2015 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 60 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 005 2.1.20 Importance Rating 4.6 K/A Statement 005 Residual Heat Removal, 2.1.20: Ability to interpret and execute procedure steps.

Proposed Question: RO 31 Rev: 0 Given:

Plant is in MODE 5 The RCS was drained to 14.5 ft. MSL LPSI pumps have been secured due to RCS leakage HPSI Pump B has been started in accordance with OP-901-131, Shutdown Cooling Malfunction RCS level has been raised to and is being maintained at 16 ft. MSL To restore a Shutdown Cooling train to service in accordance with OP-901-131, the crew will vent and start LPSI Pump ____(1)____ because HPSI Pump B is injecting to

_____(2)_____ .

(1) (2)

A. A Hot Leg 1 B. A Hot Leg 2 C. B Hot Leg 1 D. B Hot Leg 2 Revision 0 Facility: Waterford 3 Page 61 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. The crew will start LPSI pump A because HPSI Pump B is injecting to the suction of LPSI pump A. HPSI Pump B injects to Hot leg #2 (not hot leg #1). LPSI pump A takes a suction from hot leg 2.

B. CORRECT: The crew will start LPSI pump A because HPSI Pump B is injecting to the suction of LPSI pump A. HPSI Pump B injects to Hot leg #2. LPSI pump A takes a suction from hot leg 2.

C. Incorrect. The crew will start LPSI pump A because HPSI Pump B is injecting to the suction of LPSI pump A. HPSI Pump B injects to Hot leg #2 (not hot leg #1). LPSI pump A takes a suction from hot leg 2.

D. Incorrect. The crew will start LPSI pump A because HPSI Pump B is injecting to the suction of LPSI pump A. HPSI Pump A (not running) injects to Hot leg #1. LPSI pump A takes a suction from hot leg 2.

Technical Reference(s): OP-901-131 section E1 Rev. 304 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ21 obj. 6 (As available)

Question Source: Bank # RO26 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2014 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 62 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 006 K4.17 Importance Rating 3.8 K/A Statement 006 Emergency Core Cooling, K4.17: Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following: Safety Injection valve interlocks Proposed Question: RO 32 Rev: 0 Given:

A Loss of Cooling Accident (LOCA) has occurred A Recirculation Actuation Signal (RAS) has been received Which ONE of the following valves automatically changes position due to the Recirculation Actuation Signal?

A. LPSI Cold Leg Injection Flow Control Valve, SI-138A B. SI Pumps Recirc Isolation Valve SI-120B C. ESF Pumps Suction RWSP Valve, SI-106B D. ESF Pumps Suction SI Sump Valve, SI-602A Revision 0 Facility: Waterford 3 Page 63 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. SI-138 opens on a SIAS but does not reposition on a RAS. SI-138 may be manually positioned upon a RAS if the LPSI Pump fails to secure.

B. Incorrect. SI-120B does not reposition on a RAS. This valve is required to be manually closed from CP-8 within two minutes of a RAS.

C. Incorrect. SI-106 valves do not reposition on a RAS (the open signal from the SIAS is removed). This valve is required to be manually closed upon a RAS after SI-602 valves are verified open.

D. CORRECT: SI-602 valves automatically open on a RAS.

Technical Reference(s): SD-SI page 42 Rev. 15 (Attach if not previously provided) OP-902-002 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SI00 obj. 4 (As available)

Question Source: Bank # RO5 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2008 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Revision 0 Facility: Waterford 3 Page 64 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 007 A2.03 Importance Rating 3.6 K/A Statement 007 Pressurizer Relief/Quench Tank, A2.03: Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Overpressurization of the PZR.

Proposed Question: RO 33 Rev: 0 Given:

An Excess Steam Demand occurred inside containment SIAS, CIAS, MSIS and CSAS have been initiated Pressurizer level is 2% and starting to rise Representative CET temperature and RCS pressure start to rise Based on the current conditions, the crew will take actions to control Pressurizer level and prevent challenging Pressurizer safeties by ____(1)____. The Pressurizer Safety Valves are designed to lift at ____(2)____ psia.

(1) (2)

A. stabilizing RCS pressure above 2500 HPSI Pump shutoff head B. stabilizing RCS pressure above 2750 HPSI Pump shutoff head C. securing HPSI Pumps 2500 D. securing HPSI Pumps 2750 Revision 0 Facility: Waterford 3 Page 65 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: OP-902-004 step for stabilizing RCS temperature directs the crew to maintain RCS pressure within 1500-1600 psia of SG pressure. The shutoff head of the HPSI pump is just below 1500 psia. The lift set point of the Pressurizer safeties is 2500 psia. This part of the question is what meets the predict the impact part of the K/A.

B. Incorrect. Part 1 is correct. The safety limit for RCS pressure is 2750 psia, but the pressurizer safety valves lift at 2500 psia.

C. Incorrect. With pressurizer level at 2%, HPSI throttle criteria is not met. HPSI pumps cannot be secured until Pressurizer level has risen above 7%. There is a step in 902-004 to secure HPSI pumps one at a time once HPSI throttle criteria is met. Part 2 is correct.

D. Incorrect. With pressurizer level at 2%, HPSI throttle criteria is not met. HPSI pumps cannot be secured until Pressurizer level has risen above 7%. There is a step in 902-004 to secure HPSI pumps one at a time once HPSI throttle criteria is met. The safety limit for RCS pressure is 2750 psia, but the pressurizer safety valves lift at 2500 psia.

Technical Reference(s): OP-902-004 steps 18 and 20 Rev. 16 (Attach if not previously provided) SD-RCS page 55 Rev. 21 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE04 objs. 4 and 7 (As available)

WLP-OPS-RCS00 obj. 4 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 3, 10 55.43 Comments: Submitted for free review. No changes are required to the question.

Revision 0 Facility: Waterford 3 Page 66 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 008 K1.05 Importance Rating 3.0 K/A Statement 008 Component Cooling Water, K1.05: Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems:

Sources of makeup water.

Proposed Question: RO 34 Rev: 0 The normal source of makeup to the Auxiliary Component Cooling Water (ACCW) Wet Cooling Towers is the _______(1)_______. The normal source of makeup water to the Component Cooling Water Surge Tank is the _______(2)_________.

(1) (2)

A. Condensate Storage Pool Demineralized Water Storage Tank B. Condensate Storage Pool Condensate Storage Pool C. Demineralized Water Storage Tank Condensate Storage Pool D. Demineralized Water Storage Tank Demineralized Water Storage Tank Revision 0 Facility: Waterford 3 Page 67 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Wrong makeup source for WCTs. Correct Makeup source for CCW Surge Tank.

B. Incorrect. Wrong Makeup source to the WCTs and wrong source to the CCW Surge Tank. The CSP is the backup makeup source to the CCW Surge Tank.

C. Incorrect. The normal source of makeup to the WCTs is the DWST; however, the CSP is not the normal makeup source to the CCW Surge Tank.

D. CORRECT: The normal source of makeup to the WCTs and the CCW Surge Tank is the Demineralized Water Storage Tank.

Technical Reference(s): SD-CMU figure 2, Rev. 8 (Attach if not previously provided) SD-CC page 11,13 Rev. 22 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00 obj. 2,3 (As available)

Question Source: Bank # RO34 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2012 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Revision 0 Facility: Waterford 3 Page 68 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 008 A3.08 Importance Rating 3.6 K/A Statement 008 Component Cooling Water, A3.08: Ability to monitor automatic operation of the CCWS, including: Automatic actions associated with the CCWS that occur as a result of a safety injection signal Proposed Question: RO 35 Rev: 0 Given:

A Steam Line Break has occurred outside containment The crew has manually initiated a MSIS, CIAS and SIAS As a result, Component Cooling Water Flow has been lost to ________________.

A. the Spent Fuel Pool Heat Exchanger B. Emergency Diesel Generator B C. the Reactor Coolant Pumps D. Train B Dry Cooling Tower Revision 0 Facility: Waterford 3 Page 69 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: CCW flow is lost to the Spent Fuel Pool Heat Exchanger due to CC-620, Spent Fuel Pool Heat Exchanger TCV, going closed on a SIAS.

B. Incorrect. EDG B is located on the safety train and CCW flow is not isolated by the SIAS.

C. Incorrect. The reactor coolant Pumps are located on the CCW AB header. Flow will be isolated on the B train going to the AB header but will still be supplied from the A train going to the AB header.

D. Incorrect. Train B Dry Cooling Tower will isolate in the event of a low level in the CCW surge tank but flow is maintained through the tower upon a receipt of a SIAS with no low level.

Technical Reference(s): SD-CC page 37, 46, figure 22 Rev. 22 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00 obj. 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 7 55.43 Revision 0 Facility: Waterford 3 Page 70 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 010 K6.01 Importance Rating 2.7 K/A Statement 010 Pressurizer Pressure Control, K6.01 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: Pressure Detection Systems Proposed Question: RO 36 Rev: 0 Given:

Plant is at 100% power Pressurizer Pressure Channel X/Y recorder (RC-IPR-0100) indicates that the in-service Pressurizer Pressure Control Channel instrument has failed low The Pressurizer Pressure Controller output will ____(1)____. The Pressurizer Pressure control system is designed to de-energize all heaters at ___(2)___ psia.

(1) (2)

A. rise 2270 B. lower 2270 C. lower 2300 D. rise 2300 Revision 0 Facility: Waterford 3 Page 71 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. The in-service pressurizer pressure control channel failing low will cause the PPC system to see actual pressure less than setpoint. The pressurizer pressure controller output will lower to raise pressure (energize heaters). Part 2 is correct.

B. CORRECT: The in-service pressurizer pressure control channel failing low will cause the PPC system to see actual pressure less than setpoint. The pressurizer pressure controller output will lower to raise pressure (energize heaters). A high pressure heater cutout will de-energize all heaters at 2270 psia through the low level heater cutout switch.

C. Incorrect. Part 1 is correct. A high pressure heater cutout will de-energize all heaters at 2270 psia through the low level heater cutout switch. A pressure of 2300 psia is plausible since this is the pressure at which the Pressurizer spray valves are full open on rising pressure.

D. Incorrect. The in-service pressurizer pressure control channel failing low will cause the PPC system to see actual pressure less than setpoint. The pressurizer pressure controller output will lower to raise pressure (energize heaters). A pressure of 2300 psia is plausible since this is the pressure at which the Pressurizer spray valves are full open on rising pressure.

Technical Reference(s): SD-PLC pages 29,37,38 Rev. 10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 7, 10 55.43 Revision 0 Facility: Waterford 3 Page 72 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 010 A1.01 Importance Rating 2.8 K/A Statement 010 Pressurizer Pressure Control, A1.01: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including: PZR and RCS boron concentrations.

Proposed Question: RO 37 Rev: 0 Given:

A plant shutdown is commenced in accordance with OP-010-005, Plant Shutdown (assume that Pressurizer and RCS boron concentrations are matched before the shutdown)

The crew will commence Boron Equalization to ____(1)____ Pressurizer boron concentration. OP-010-005, Plant Shutdown, states that Boron Equalization may be secured when RCS and Pressurizer Boron concentration is within ____(2)____ ppm of each other.

(1) (2)

A. raise 50 B. lower 10 C. raise 10 D. lower 50 Revision 0 Facility: Waterford 3 Page 73 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Part 1 is correct. The note in OP-010-005 states Boron Equalization may be secured when the boron concentration difference between the RCS and the Pressurizer is <10 ppm. Boron equalization is required to be initiated if the change in RCS concentration is expected to exceed 50 ppm.

B. Incorrect. During a plant startup, part 1 would be correct. Part 2 is correct.

C. CORRECT: Boron equalization is commenced during a Plant Shutdown to keep the RCS and the Pressurizer boron concentrations within 10 ppm. The RCS concentration will rise due to direct boration to the RCS, therefore, boron equalization will raise pressurizer boron concentration. Boron equalization is performed using Pressurizer heaters and spray.

D. Incorrect. During a plant startup, part 1 would be correct. The note in OP-010-005 states Boron Equalization may be secured when the boron concentration difference between the RCS and the Pressurizer is <10 ppm. Boron equalization is required to be initiated if the change in RCS concentration is expected to exceed 50 ppm.

Technical Reference(s): OP-010-005 note before step 9.13.1 Rev. 328 (Attach if not previously provided) OP-010-005 note before step 9.13.3 Rev. 328 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPN02 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 6, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 74 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 012 K1.02 Importance Rating 3.4 K/A Statement 012 Reactor Protection, K1.02: Knowledge of the physical connections and/or cause effect relationships between the RPS and the following systems: 125V dc system Proposed Question: RO 38 Rev: 0 Given:

The plant is at 100% power A loss of the A-DC bus occurs Reactor Trip Switchgear breakers _____(1)_____ open and a reactor trip

_______(2)______.

(1) (2)

A. 1, 2, 5, and 6 occurs B. 1, 2, 5, and 6 does not occur C. 1, 3, 5, and 7 does not occur D. 1, 3, 5, and 7 occurs Revision 0 Facility: Waterford 3 Page 75 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. This arrangement of breakers opening is the same as losing an instrument SUPS. With this combination of breakers, the reactor does not trip. Part 2 is correct.

B. Incorrect. This arrangement of breakers opening is the same as losing an instrument SUPS. With this combination of breakers, the reactor does not trip.

C. Incorrect. Part 1 is correct. . The arrangement of the reactor trip switchgear breakers is such that this combination of RTSG breakers will de-energize the CEDM coils.

D. CORRECT: A Reactor trip will occur on a loss of the A-DC bus. The arrangement of the reactor trip switchgear breakers is such that this combination of RTSG breakers will de-energize the CEDM coils. There are arrangements such that four RTSG breakers will open and not trip the reactor. This question is comprehensive since the applicant must know what breakers are affected and then apply it to the RTSG flow path to the CEDM coils.

Technical Reference(s): OP-901-313 page 13 Rev. 305 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO3 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 6, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 76 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 K2.01 Importance Rating 3.6 K/A Statement 013 Engineered Safety Features Actuation, K2.01: Knowledge of bus power supplies to the following: ESFAS/safeguards equipment control.

Proposed Question: RO 39 Rev: 0 What is the power supply to MS-401A, EFW Pump Turbine Steam Supply Valve?

A. A-DC B. AB-DC C. SUPS A D. SUPS MC Revision 0 Facility: Waterford 3 Page 77 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. MS-401A motor operated valve receives power from the AB-DC bus.

B. CORRECT: MS-401A motor operated valve receives power from the AB-DC bus.

C. Incorrect. SUPS A (120VAC) supplies various loads including ESFAS operated valves. MS-401A motor operated valve receives power from the AB-DC bus.

D. Incorrect. SUPS MC (120VAC) is an A side powered SUPS that supplies various loads including ESFAS panels. MS-401A motor operated valve receives power from the AB-DC bus Technical Reference(s): OP-005-004 page 73 Rev. 33 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EFW00 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 78 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 A2.02 Importance Rating 4.3 K/A Statement 013 Engineered Safety Features Actuation, A2.02: Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Excess steam demand.

Proposed Question: RO 40 Rev: 0 Given:

An Excess Steam Demand event has occurred The crew has entered OP-902-004, Excess Steam Demand Recovery Procedure Following the Main Steam Isolation Signal (MSIS), indications are:

Steam Generator 1 pressure is 400 PSIA Steam Generator 2 pressure is 475 PSIA Steam Generator 1 level is 9% NR and lowering Steam Generator 2 level is 9% NR and lowering Currently, ____(1)____ initiated.

Per OI-038-000, EOP Operations Expectations/Guidance procedure, the minimum criteria to Verify MSIS Actuated is to _____(2)________.

(1) (2)

A. only EFAS-2 is verify Main Feedwater and Main Steam Isolation valves closed B. only EFAS-2 is verify trip path lights not illuminated on CP-7 for MSIS C. neither EFAS-1 nor verify trip path lights not illuminated on EFAS-2 are CP-7 for MSIS D. neither EFAS-1 nor verify Main Feedwater and Main Steam EFAS-2 are Isolation valves closed Revision 0 Facility: Waterford 3 Page 79 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Neither EFAS 1 nor EFAS 2 will be initiated because S/G #1 and S/G #2 pressures are less than 666 PSIA and S/G #2 pressure is not 123 psid greater than S/G #1 pressure. Part 2 is correct.

B. Incorrect. Neither EFAS 1 nor EFAS 2 will be initiated because S/G #1 and S/G #2 pressures are less than 666 PSIA and S/G #2 pressure is not 123 psid greater than S/G #1 pressure. Verifying trip path lights not illuminated on CP-7 for MSIS is the minimum requirement to verify MSIS initiation.

C. Incorrect. Part 1 is correct. Verifying trip path lights not illuminated on CP-7 for MSIS is the minimum requirement to verify MSIS initiation.

D. CORRECT: Neither EFAS 1 nor EFAS 2 will be initiated because S/G #1 and S/G #2 pressures are less than 666 PSIA and S/G #2 pressure is not 123 psid greater than S/G #1 pressure. OI-038-000 states that the minimum requirement for verify MSIS actuated is to verify Main Feedwater and Main Steam Isolation valves closed.

Technical Reference(s): OI-038-000 step 5.4.113, 5.2.7 Rev. 13 (Attach if not previously provided) TG-OP-902-004 step 8 Rev. 16 (including version/revision number) SD-PPS page 36 Rev. 17 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPS obj. 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 7, 10 55.43 Revision 0 Facility: Waterford 3 Page 80 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 022 K3.02 Importance Rating 3.0 K/A Statement 022 Containment Cooling, K3.02: Knowledge of the effect that a loss or malfunction of the CCS will have on the following: Containment instrumentation readings.

Proposed Question: RO 41 Rev: 0 Given:

A Small Break LOCA has occurred The BOP reports that Component Cooling Water has been lost to one train of Containment Fan Coolers and containment temperature is rising Per OI-038-000, EOP Operations Expectations/Guidance Procedure, ____(1)____

parameter values will be used if a harsh environment exists in containment . A harsh environment is defined as Containment Temperature greater than or equal to a minimum temperature of _____(2)_____ ºF.

(1) (2)

A. underlined 220 B. underlined 200 C. bracketed 200 D. bracketed 220 Revision 0 Facility: Waterford 3 Page 81 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Part 1 is plausible because some numbers in the EOPs are underlined but are not harsh environment numbers. 220 ºF is the minimum temperature to meet Condition 1 criteria for the CTPC SFSC in OP-902-002, LOCA Recovery Procedure.

B. Incorrect. . Part 1 is plausible because some numbers in the EOPs are underlined but are not harsh environment numbers. Part 2 is correct.

C. CORRECT: The effect of losing CCW flow to a train of CFCs is that containment temperature will rise and cause inaccuracies of some containment instruments. The EOPs account for this inaccuracy by supplying the operators with bracketed parameter values that must be used when containment temperature is greater than or equal to 200ºF.

D. Incorrect. Part 1 is correct. 220 ºF is the minimum temperature to meet Condition 1 criteria for the CTPC SFSC in OP-902-002, LOCA Recovery Procedure.

Technical Reference(s): OI-038-000 page 6 Rev. 13 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 82 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 026 K2.01 Importance Rating 3.4 K/A Statement 026 Containment Spray, K2.01: Knowledge of bus power supplies to the following: Containment spray pumps.

Proposed Question: RO 42 Rev: 0 Containment Spray (CS) Pump B is powered from which bus?

A. SWGR 31B B. SWGR 3B C. SWGR 1B D. SWGR 2B Revision 0 Facility: Waterford 3 Page 83 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect: Containment Spray (CS) Pump B is powered from the 3B bus. The 31B is a 480 V safety bus fed from the 3B bus.

B. CORRECT: Containment Spray (CS) Pump B is powered from the 3B bus.

C. Incorrect: Containment Spray (CS) Pump B is powered from the 3B bus. The 1B bus is a 6.9 Kv which carries large loads.

D. Incorrect: Containment Spray (CS) Pump B is powered from the 3B bus. The 2B bus is a 4.16 Kv non-safety bus.

Technical Reference(s): OP-009-001 page 40 Rev. 306 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-ED00 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Revision 0 Facility: Waterford 3 Page 84 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 039 K5.08 Importance Rating 3.6 K/A Statement 039 Main and Reheat Steam, K5.08: Knowledge of the operational implications of the following concepts as the apply to the MRSS: Effect of steam removal on reactivity.

Proposed Question: RO 43 Rev: 0 Given:

A reactor startup has commenced following a reactor trip from 100% power that occurred 4 days ago 250 EFPD SBCS valves are maintaining Tave Which ONE of the following will cause actual critical rod position to be LOWER than estimated critical rod position?

A. SBCS master controller setpoint fails upward 50 PSI B. SBCS permissive setpoint fails upward 50 PSI C. SBCS master controller setpoint fails downward 50 PSI D. SBCS permissive setpoint fails downward 50 PSI Revision 0 Facility: Waterford 3 Page 85 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Raising the master controller setpoint will result in a higher Tave and add negative reactivity due to negative MTC (MOC conditions). Critical rod height would be higher than estimated.

B. Incorrect. Raising the permissive setpoint by 50 PSI simply moves the permissive closer to the demand setpoint. It does not change the modulation point of the valves.

C. CORRECT: Lowering the master controller setpoint will result in SBCS valves opening sooner, reducing RCS temperature (250 EFPD ensures a negative MTC).

Colder moderator will produce more thermal neutrons, lowering the required critical rod position.

D. Incorrect. Lowering the permissive setpoint will bring the permissive signal in earlier, but the master controller setpoint will be the same, meaning no change in RCS temperature.

Technical Reference(s): SD-SBC page 9 and figure 3. Rev. 10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SBC00 obj. 8 (As available)

Question Source: Bank # RO15 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2008 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 14 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 86 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 059 K4.18 Importance Rating 2.8 K/A Statement 059 Main Feedwater, K4.18: Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: Automatic feedwater reduction on plant trip.

Proposed Question: RO 44 Rev: 0 Given:

Power is 100%

SG-ILR-1111, Steam Generator 1 Narrow Range level indicator is reading 0% on the CP-1 recorder SG-ILR-1105, Steam Generator 1 Narrow Range level indicator is reading 68%

on the CP-1 recorder All Narrow Range Steam Generator levels on CP-8 are indicating 68%

With the above conditions, a reactor trip occurs.

Main Feedwater Pump A speed controller ___(1)___ . Main Feedwater Regulating Valve A and Startup Feedwater Regulating Valve A Controllers __(2)___.

(1) (2)

A. lowers to 3900 RPM remain at their pre-trip positions B. lowers to 3900 RPM reposition to their Reactor Trip Positions C. remains at 4500 RPM remain at their pre-trip positions D. remains at 4500 RPM reposition to their Reactor Trip Positions Revision 0 Facility: Waterford 3 Page 87 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: All controllers on the FWCS #1 side are in manual due to a level deviation. Upon a reactor trip, the design of the FWCS is to return the MFP controller to auto for 5 seconds such that MFP minimum speed is obtained (3900 RPM). The Main Feedwater Reg and Startup Feedwater Reg Valve controllers do not have this feature and will remain in their pre-trip position.

B. Incorrect. Part 1 is correct. All controllers on the FWCS #1 side are in manual due to a level deviation. The Main Feedwater Reg Valve and Startup Feedwater Reg Valve controllers will remain in their pre-trip position.

C. Incorrect. All controllers on the FWCS #1 side are in manual due to a level deviation. Upon a reactor trip, the design of the FWCS is to return the MFP controller to auto for 5 seconds such that MFP mimimum speed is obtained (3900 RPM). Part 2 is correct.

D. Incorrect. All controllers on the FWCS #1 side are in manual due to a level deviation. Upon a reactor trip, the design of the FWCS is to return the MFP controller to auto for 5 seconds such that mimimum speed is obtained (3900 RPM). The Main Feedwater Reg and Startup Feedwater Reg Valve controllers do not have this feature and will remain in their pre-trip position.

Technical Reference(s): SD-FWC00 page 24,25 Rev. 13 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-FWC00 obj. 5 (As available)

Question Source: Bank # RO 45 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 88 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 076 A1.02 Importance Rating 2.6 K/A Statement 076 Service Water, A1.02: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: Reactor and turbine building closed cooling water temperatures Proposed Question: RO 45 Rev: 0 Given:

The BOP reports rising Turbine Cooling Water (TCW) temperatures The crew enters OP-901-512, Loss of Turbine Cooling Water The BOP will lower TCW temperature using TC-147, Turbine Cooling Water Temperature Control, by taking the control switch to the ____(1)____ position.

To maintain TCW system pressure constant, the control switch for TC-135, Turbine Cooling Water Pressure Control, will be taken to the ___(2)___ position.

(1) (2)

A. less more B. less less C. more less D. more more Revision 0 Facility: Waterford 3 Page 89 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Part 1 is correct. Opening TC-147 will raise system pressure, TC-135 is required to be closed to lower system pressure.

B. CORRECT: To lower TCW temperature, the C/S for TC-147 must be taken to LESS (less temperature). Opening TC-147 will raise system pressure, the C/S for TC-135 must be taken to LESS (will close the valve) to restore system pressure to the previous value.

C. Incorrect. The individual TCVs for components served by TCW have less/more switches which correlates to the valve being more open or less open. The opposite of how the TC-147 works.

D. Incorrect. The individual TCVs for components served by TCW have less/more switches which correlates to the valve being more open or less open. The opposite of how the TC-147 works. Opening TC-147 will raise system pressure, TC-135 is required to be closed to lower system pressure.

Technical Reference(s): OP-901-512 section E3 Rev. 3 (Attach if not previously provided) SD-TCW page 11, Figure 1 Rev. 8 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-TC00 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 90 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 061 K2.02 Importance Rating 3.7 K/A Statement 061 Auxiliary/Emergency Feedwater, K2.02: Knowledge of bus power supplies to the following: AFW electric drive pumps Proposed Question: RO 46 Rev: 0 The Auxiliary Feedwater Pump (AFW Pump) is powered from which bus?

A. 1A B. 2A C. 1B D. 2B Revision 0 Facility: Waterford 3 Page 91 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The Auxiliary Feedwater (AFW) Pump is a 6.9 KV pump, but is powered from the 1B bus.

B. Incorrect. The Auxiliary Feedwater (AFW) Pump is a 6.9 KV (not 4.16 KV) motor powered from the 1B bus C. CORRECT: The Auxiliary Feedwater (AFW) Pump is a 6.9 KV motor powered from the 1B bus.

D. Incorrect. The Auxiliary Feedwater (AFW) Pump is a 6.9 KV (not 4.16 KV) motor powered from the 1B bus.

Technical Reference(s): OP-003-035 page 20, Rev. 305 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-ED00 obj. 1 (As available)

Question Source: Bank # RO47 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2012 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 92 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 061 A3.03 Importance Rating 3.9 K/A Statement 061 Auxiliary/Emergency Feedwater, A3.03: Ability to monitor automatic operation of the AFW, including: AFW S/G level control on automatic start Proposed Question: RO 47 Rev: 0 A Loss of MFW event has occurred. After a manual Reactor trip, the following conditions are observed:

SG 1 pressure is 980 PSIA SG 2 pressure is 960 PSIA SG 1 level is 51% WR level and lowering SG 2 level is 38% WR level and lowering Pressurizer pressure 1900 PSIA Containment pressure 15.3 PSIA With SG 2 EFW flow transmitter failed high, which of the following describes the EFW system response for SG 2?

A. Primary FCV and Backup FCV remain closed.

B. Primary FCV remains closed and backup FCV opens to 175 GPM.

C. Primary FCV opens to preset valve position and Backup FCV remains closed.

D. Primary FCV opens to preset valve position and Backup FCV opens to 400 GPM flow value.

Revision 0 Facility: Waterford 3 Page 93 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. At 55% WR level, the Primary FCV opens to a preset valve position. The EFW flow transmitter failing high has no effect on the Primary FCV at this level.

B. Incorrect. At 55% WR level, the Primary FCV opens to a preset valve position. The EFW flow transmitter failing high has no effect on the Primary FCV at this level. Part 2 is plausible because the backup FCV would open to 175 gpm if the Primary FCV did not open and there was no failed transmitter.

C. CORRECT: At 55% WR level, the Primary FCV opens to a preset valve position. The EFW flow transmitter failing high has no effect on the Primary FCV at this level. At 45% WR, the Backup FCV would have opened to 400 gpm, but with the EFW flow transmitter failed high, the Backup FCV will remain closed D. Incorrect. This would be the correct answer if the EFW flow transmitter was not failed high. At 45% WR, the Backup FCV would have opened to 400 gpm, but with the EFW flow transmitter failed high, the Backup FCV will remain closed.

Technical Reference(s): SD-EFW page 7, Figure 19 Revision 13 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EFW obj. 6 (As available)

Question Source: Bank # RO46 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2012 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 4,8 55.43 Revision 0 Facility: Waterford 3 Page 94 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 062 A3.05 Importance Rating 3.5 K/A Statement 062 AC Electrical Distribution, A3.05: Ability to monitor automatic operation of the ac distribution system, including: Safety-related indicators and controls.

Proposed Question: RO 48 Rev: 0 Given:

• EDG A is running paralleled in TEST mode for surveillance OP-903-068, Emergency Diesel Generator and Subgroup Relay Operability Verification

• A SIAS occurred with NO Loss of Offsite Power condition The EDG output breaker will (1) and the HPSI Pumps will be started (2) .

(1) (2)

A. OPEN by the sequencer B. OPEN immediately C. stay CLOSED by the sequencer D. stay CLOSED immediately Revision 0 Facility: Waterford 3 Page 95 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: EDG output breaker OPENs and after 2 seconds the sequencer loading begins since NO undervoltage conditions existed on the safety bus.

B. Incorrect. EDG output breaker OPENs and after 2 seconds the sequencer loading begins since NO undervoltage conditions existed on the safety bus to ensure NO excess loading if IMMEDIATELY started.

C. Incorrect. EDG output breaker OPENs and after 2 seconds the sequencer loading begins since NO undervoltage conditions existed on the safety bus.

D. Incorrect. EDG output breaker OPENs and after 2 seconds the sequencer loading begins since NO undervoltage conditions existed on the safety bus to ensure NO excess loading if IMMEDIATELY started.

Technical Reference(s): SD-EDG page 40, 49 Rev. 25 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SEQ00 obj. 1 and 2 (As available)

Question Source: Bank # RO12 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 8 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 96 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 063 A4.02 Importance Rating 2.8 K/A Statement 063 DC Electrical Distribution, A4.02: Ability to manually operate and/or monitor in the control room: Battery voltage indicator.

Proposed Question: RO 49 Rev: 0 Given:

An event occurred and the crew is performing Appendix 1, Diagnostic Flow Chart The BOP has determined that neither 125 VDC Safety Bus is energized The crew will diagnose to ____(1)____.

The 125VDC Safety Bus Voltage indicators are located on ___(2)___ .

(1) (2)

A. OP-902-005, Station CP-1 Blackout B. OP-902-005, Station CP-7 Blackout C. OP-902-008, Functional CP-1 Recovery Procedure D. OP-902-008, Functional CP-7 Recovery Procedure Revision 0 Facility: Waterford 3 Page 97 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect: Part 1 is plausible because OP-902-005 is the optimal procedure that contains the step for a loss of all AC and one vital DC bus, but not both DC busses.

Part 2 is correct.

B. Incorrect: Part 1 is plausible because OP-902-005 is the optimal procedure that contains the step for a loss of all AC and one vital DC bus, but not both DC busses.

Part 2 is plausible since CP-7 is the location of the 120V vital AC indicators which is also checked during the diagnostic flow chart C. CORRECT: The second block in the diagnostic flow chart asks the crew is at least one 125 VDC Safety Bus energized?. If the answer is no, the crew is directed to go to OP-902-008, Functional Recovery Procedure. The vital DC Bus indicators are located on CP-1. The second part of the question is meeting the K/A portion that will demonstrate the applicants ability to monitor the indicator in the Control Room (find its location).

D. Incorrect: Part 1 is correct. Part 2 is plausible since CP-7 is the location of the 120 V vital AC indicators which is also checked during the diagnostic flow chart.

Technical Reference(s): OP-902-009 Appendix 1 Rev. 315 (Attach if not previously provided) SD-DC page 13 Rev. 9 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE05 obj. 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 10 55.43 Revision 0 Facility: Waterford 3 Page 98 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 K4.10 Importance Rating 3.5 K/A Statement 064 Emergency Diesel Generator, K4.10: Knowledge of ED/G system design feature(s) and/or interlock(s) which provide for the following: Automatic load sequencer: blackout Proposed Question: RO 50 Rev: 0 Given:

The crew is restoring Emergency Diesel Generator (EDG) B following a Station Blackout Emergency Diesel Generator B has started and the output breaker is closed The Input Transformer to MCC 315B develops a direct short to ground upon being re-energized by the 1 second Sequencer Load Block (S1 relay)

An undervoltage condition exists on 4160 Volt Bus 3B Which of the following conditions describes the Sequencer response?

A. Immediately stops the automatic loading process due to Under Voltage Override (UVO) condition.

B. Immediately stops the automatic loading process due to Sequencer Lockout (SLO) condition.

C. Stops the automatic loading process when the 7 second Load Block (S3 relay) is reached.

D. Stops the automatic loading process when the 17 second Load Block (S4 relay) is reached.

Revision 0 Facility: Waterford 3 Page 99 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The Undervoltage Override (UVO) of the Sequencer occurs at 0.5 seconds. UVO is the interlock that allows the sequencer to remain running between 0.5 and 17 seconds while large transformer loads are sequenced on.

B. Incorrect. Sequencer Lockout is blocked between the 0.5 second Sequencer Load block and the 17 Second load block (due to the UVO feature). The short occurs on the 1 Second load block so the sequencer will not stop immediately C. Incorrect. A SLO condition occurs at 17 seconds not 7 seconds as stated in the distractor. Sequencer Lockout is active at the 7 second load block but is blocked by the UVO feature.

D. CORRECT: A SLO condition will occur at Load Block S4 due to the undervoltage condition on Bus 3A.

Technical Reference(s): SD-EDG pg. 53 Rev. 25 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SEQ obj. 2 (As available)

Question Source: Bank # RO23 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 100 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 004 K5.30 Importance Rating 3.8 K/A Statement 004 Chemical and Volume Control, K5.30: Knowledge of the operational implications of the following concepts as they apply to the CVCS: Relationship between temperature and pressure in CVCS components during solid plant operation Proposed Question: RO 51 Rev: 0 Given:

Solid Plant operations are in progress RCS temperature is 120°F and stable Letdown Back Pressure Controller is operating in AUTO CCW flow is reduced to the in-service SDC Heat Exchanger RCS pressure will ____(1)____ until the Letdown Back Pressure Control Valve throttles

_____(2)____.

(1) (2)

A. lower closed B. rise closed C. lower open D. rise open Revision 0 Facility: Waterford 3 Page 101 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Lowering CCW flow will cause temperature and RCS pressure to rise. To restore RCS pressure to setpoint the letdown backpressure valve will be opened (not closed) to raise letdown flow to lower RCS pressure to setpoint.

B. Incorrect. Lowering CCW flow will cause temperature and RCS pressure to rise. To restore RCS pressure to setpoint the letdown backpressure valve will be opened (not closed) to raise letdown flow to lower RCS pressure to setpoint.

C. Incorrect. Lowering CCW flow will cause temperature and RCS pressure to rise. To restore RCS pressure to setpoint the letdown backpressure valve will be opened to raise letdown flow to lower RCS pressure to setpoint.

D. CORRECT: Per OP-010-005, when setting up solid operation the Letdown Backpressure valves are set to control pressure at ~ 100 psig in AUTO. Lowering CCW flow to the in-service SDC Heat Exchanger when RCS temperature is initially stable will cause temperature and RCS pressure to rise. To restore RCS pressure to setpoint the letdown backpressure valve must open to raise letdown flow to lower RCS pressure to setpoint.

Technical Reference(s): OP-001-001 pg. 17 Revision 34 (Attach if not previously provided) SD-SDC Figure 2 Revision 2 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CVC00 obj. 3 (As available)

Question Source: Bank # RO29 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2012 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 5 55.43 Revision 0 Facility: Waterford 3 Page 102 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 073 2.4.4 Importance Rating 4.5 K/A Statement 073 Process Radiation Monitoring, 2.4.4: Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Proposed Question: RO 52 Rev: 0 Given:

A Waste Condensate Tank is being discharged to Circulating Water Upon rising activity, the ATC will verify which of the following close automatically to terminate the release?

A. BM-547, Boron Management Discharge to Circ Water Auto Isolation, only.

B. LWM-441, Liquid Waste to Circ Water Shutoff, only.

C. BM-547, Boron Management Discharge to Circ Water Auto Isolation, and BM-549, Boron Management Discharge to Circ Water Flow Control D. LWM-441, Liquid Waste to Circ Water Shutoff and LWM-442, Liquid Waste to Circ Water Control Revision 0 Facility: Waterford 3 Page 103 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The BM valve would be applicable if a Boric Acid Condensate Tank was being discharged. BM-549 is a control valve and has only a controller on CP-4 (no C/S), but it still gets an isolation signal.

B. Incorrect. The automatic feature is that both LWM-441 and LWM-442 will close.

LWM-442 is a control valve and has only a controller on CP-4 (no C/S), but it still gets an isolation signal.

C. Incorrect. This answer would be correct if a Boric Acid Condensate Tank to Circulating Water discharge was occurring.

D. CORRECT: Waste Condensate Tank release to Circ Water will be automatically terminated upon a high radiation indicated by the LWM Rad Monitor. The automatic feature is that both LWM-441 and LWM-442 will close. This event is an entry level condition for off-normal, OP-901-412, Liquid Waste Discharge High Radiation.

Technical Reference(s): OP-901-412 page 4 revision 2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO40 obj. 3 (As available)

Question Source: Bank # RO59 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2007 RO Makeup Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10, 11 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 104 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 076 K1.19 Importance Rating 3.6 K/A Statement 076 Service Water, K1.19: Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems: SWS emergency heat loads Proposed Question: RO 53 Rev: 0 When aligning Auxiliary Component Cooling Water (ACCW) flow to the Essential Chillers the ACCW valves open___(1)____ .

To prevent void formation during the transition to ACCW, the ACCW ___(2)____ valves to the Essential Chillers are interlocked to open first.

(1) (2)

A. first, then the CCW valves inlet close B. after the CCW valves outlet close C. after the CCW valves inlet close D. first, then the CCW valves outlet close Revision 0 Facility: Waterford 3 Page 105 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Part 1 is plausible because the applicant could assume that flow is maintained to prevent the Essential Chiller from tripping on low flow. The Essential Chillers have a CCW low flow trip. Part 2 is correct.

B. Incorrect. Part 1 is correct. The ACCW outlet valves to the Essential Chillers will open when the inlet valve is 5 to 10% open to keep the lines full.

C. CORRECT: When aligning the Essential Chillers to ACCW (Wet Mode), the ACCW valves are opened after the CCW valves close to prevent cross-connecting the two systems. The ACCW outlet valves to the Essential Chillers will open when the inlet valve is 5 to 10% open to keep the lines full.

D. Incorrect. Part 1 is plausible because the applicant could assume that flow is maintained to prevent the Essential Chiller from tripping on low flow. The Essential Chillers have a CCW low flow trip. The ACCW outlet valves to the Essential Chillers will open when the inlet valve is 5 to 10% open to keep the lines full.

Technical Reference(s): SD-CC page 29 Rev. 22 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CHW obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 7 55.43 Revision 0 Facility: Waterford 3 Page 106 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 078 K3.01 Importance Rating 3.1 K/A Statement 078 Instrument Air, K3.01: Knowledge of the effect that a loss or malfunction of the IAS will have on the following: Containment air system.

Proposed Question: RO 54 Rev: 0 Given:

Instrument Air has been lost to Containment The crew is performing the actions of OP-901-511, Loss of Instrument Air CEDM COOLING COILS CCW INLET HDR ISOL, CC-646, is failed ____(1)____.

Containment Fan Cooler discharge air flow to the ring header is ___(2)___ .

(1) (2)

A. open unaffected B. closed unaffected C. open bypassed D. closed bypassed Revision 0 Facility: Waterford 3 Page 107 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Part 1 is correct. CCS-102A and CCS-102B, CNTMT COOLING HVAC SAFETY DISCH DPR A(B) will open bypassing discharge flow to the CFC ring header.

B. Incorrect. CC-646, CEDM COOLING CCW INLET HDR ISOL, will fail open on a loss of instrument air. CCS-102A and CCS-102B, CNTMT COOLING HVAC SAFETY DISCH DPR A(B) will open bypassing discharge flow to the CFC ring header C. CORRECT: CC-646, CEDM COOLING CCW INLET HDR ISOL, will fail open on a loss of instrument air allowing full CCW flow to the CEDM coolers. CCS-102A and CCS-102B, CNTMT COOLING HVAC SAFETY DISCH DPR A(B) will open bypassing discharge flow to the CFC ring header.

D. Incorrect. CC-646, CEDM COOLING CCW INLET HDR ISOL, will fail open on a loss of instrument air. Part 2 is correct.

Technical Reference(s): OP-901-511 Attachment 7 Rev. 15 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 10 55.43 Revision 0 Facility: Waterford 3 Page 108 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 103 2.2.44 Importance Rating 4.2 K/A Statement 103 Containment, 2.2.44: 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.

Proposed Question: RO 55 Rev: 0 If Containment to ____(1)____ differential pressure reaches 8.5 INWD, Containment Vacuum Relief Valves, CVR-101 and CVR-201, open, and will ______(2)_____ when differential pressure lowers to the reset value.

(1) (2)

A. annulus automatically close B. ambient automatically close C. ambient require manual closure D. annulus require manual closure Revision 0 Facility: Waterford 3 Page 109 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Part 1 is correct. CVR-101 and CVR-201 must be manually closed when the reset value is reached.

B. Incorrect. Containment to Ambient D/P can be read on CP-18 but is not the D/P switch that opens CVR-101 and CVR-201. CVR-101 and CVR-201 must be manually closed when the reset value is reached.

C. Incorrect. Containment to Ambient D/P can be read on CP-18 but is not the D/P switch that opens CVR-101 and CVR-201. Part 2 is correct.

D. CORRECT: CVR-101 and CVR-201 open automatically when annulus pressure is greater than containment pressure by 8.5 INWD (Containment to annulus D/P).

CVR-101 and CVR-201 must be manually closed when the reset value is reached.

Technical Reference(s): SD-CB pages 8 and 9 Rev. 11 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CB00 obj. 2 (As available)

Question Source: Bank #

Modified Bank # RO55 (Note changes or attach parent)

New Question History: Last NRC Exam 2012 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 4 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9 55.43 Revision 0 Facility: Waterford 3 Page 110 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 001 K5.06 Importance Rating 3.8 K/A Statement 001 Control Rod Drive, K5.06: Knowledge of the following operational implications as they apply to the CRDS: Effects of control rod motion on axial offset.

Proposed Question: RO 56 Rev: 0 Given:

The crew is performing a down power in accordance with OP-901-212, Rapid Plant Power reduction Due to the down power, ASI will be trending in the ____(1)____ direction. To compensate, the ATC will ____(2)____ control rods.

(1) (2)

A. negative insert B. negative withdraw C. positive withdraw D. positive insert Revision 0 Facility: Waterford 3 Page 111 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: As power is lowered, ASI will move in the negative direction (top of the core). To compensate the ATC will insert control rods.

B. Incorrect. Part 1 is correct. Plausible because the applicant believe that withdrawing control rods will make ASI more positive.

C. Incorrect. As power is lowered, ASI will move in the negative direction (top of the core). Withdrawing control rods would make ASI more positive.

D. Incorrect. As power is lowered, ASI will move in the negative direction (top of the core). Part 2 is correct.

Technical Reference(s): OP-010-005 step 9.10.2.1 Rev. 328 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPN02 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 1, 6, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 112 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 002 K6.02 Importance Rating 3.6 K/A Statement 002 Reactor Coolant, K6.02: Knowledge of the effect of a loss or malfunction on the following RCS components: RCP Proposed Question: RO 57 Rev: 0 To provide for detection and isolation of a Reactor Coolant leak through the Reactor Coolant Pump 1A seal cooler, the 1A RCP Seal Cooler (CC-679A/CC-6651A) isolation valves will close on a Component Cooling Water (CCW) return temperature of

____(1)____ ºF.

If CCW flow is lost to a Reactor Coolant Pump (RCP) seal and cannot be restored within

____(2)____ minutes, then trip the Reactor and secure the affected RCP.

(1) (2)

A. 145 three B. 145 ten C. 155 three D. 155 ten Revision 0 Facility: Waterford 3 Page 113 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. CCW Seal return temperature of 145 ºF is the value at which the valves will remain open and not auto close again 100 seconds after isolation. Part 2 is correct.

B. Incorrect. CCW Seal return temperature of 145 ºF is the value at which the valves will remain open and not auto close again 100 seconds after isolation. Ten minutes is the time limit in which CCW cannot be restored to a RCP if isolated.

C. CORRECT: If CCW Seal return temperature reaches 155ºF, the CCW inlet and outlet valves to the seal coolers auto close. CCW flow must be restored to the RCP within 3 minutes are the operator must trip the reactor and secure the affected RCP.

D. Incorrect. Part 1 is correct. Ten minutes is the time limit in which CCW cannot be restored to a RCP if isolated.

Technical Reference(s): OP-901-130 page 5, 11 Rev. 11 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00 obj. 12 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 114 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 015 K4.07 Importance Rating 3.7 K/A Statement 015 Nuclear Instrumentation, K4.07: Knowledge of NIS design feature(s) and/or interlock(s) provide for the following: Permissives.

Proposed Question: RO 58 Rev: 0 Given:

A plant Shutdown is in progress in accordance with OP-010-005, Plant Shutdown At 1X10-4 % power and dropping, the CPC trips are (can be)___(1)___ bypassed and the High Log Power Trips are (can be) __(2)___ enabled.

(1) (2)

A. manually manually B. automatically manually C. automatically automatically D. manually automatically Revision 0 Facility: Waterford 3 Page 115 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Part 1 is correct. The design feature for the permissive on the NI log channel is that the High Log Power Trip is automatically enabled on the 1X10-4 bistable.

B. Incorrect. The CPC trips must be manually bypassed (by procedure) at this power.

The CPC trip bypass is automatically removed on the 1X10-4 bistable. The design feature for the permissive on the NI log channel is that the High Log Power Trip is automatically enabled on the 1X10-4 bistable.

C. Incorrect. The CPC trips must be manually bypassed (by procedure) at this power.

The CPC trip bypass is automatically removed on the 1X10-4 bistable. Part 2 is correct.

D. CORRECT: The design feature for the permissive on the NI log channel is that the High Log Power Trip is automatically enabled on the 1X10-4 bistable. The CPC trips must be manually bypassed (by procedure) at this power.

Technical Reference(s): OP-010-005 step 9.1.42 Rev. 328 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-ENI00 obj. 16 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 116 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 017 A4.02 Importance Rating 3.8 K/A Statement 017 In-Core Temperature Monitor, A4.02: Ability to manually operate and/or monitor in the control room: Temperature values used to determine RCS/RCP operation during inadequate core cooling (i.e., if applicable, average of five highest values)

Proposed Question: RO 59 Rev: 0 Given:

Plant is at 100% power Loss of offsite power has occurred The CRS has directed the ATC to verify the RCS Inventory Control Safety Function per OP-902-000, Standard Post Trip Actions The ATC will verify RCS subcooling using ____(1)____ subcooling instruments. These instruments can be obtained from __(2)__.

(1) (2)

A. Thot CP-2 and CP-8 B. Thot QSPDS and CP-7 C. CET QSPDS and CP-7 D. CET CP-2 and CP-8 Revision 0 Facility: Waterford 3 Page 117 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. OI-038-000 step 5.2.4 states that RCS subcooling is checked using Thot subcooling for forced circulation (CP-2 and CP-8) and CET subcooling for natural circulation (QSPDS and CP-7).

B. Incorrect. OI-038-000 step 5.2.4 states that RCS subcooling is checked using Thot subcooling for forced circulation (CP-2 and CP-8) and CET subcooling for natural circulation (QSPDS and CP-7)

C. CORRECT: OI-038-000 step 5.2.4 states that RCS subcooling is checked using Thot subcooling for forced circulation (CP-2 and CP-8) and CET subcooling for natural circulation (QSPDS and CP-7). The applicant will need to determine that the RCS is on natural circulation due to the loss of offsite power given in the stem.

D. Incorrect. OI-038-000 step 5.2.4 states that RCS subcooling is checked using CET subcooling for natural circulation. CET subcooling instruments are located on QSPDS and CP-7, not on CP-2 and CP-8.

Technical Reference(s): OI-038-000 step 5.2.4 Rev. 13 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 2, 10 55.43 Revision 0 Facility: Waterford 3 Page 118 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 033 A1.01 Importance Rating 2.7 K/A Statement 033 Spent Fuel Pool Cooling, A1.01: Ability to predict and/or monitor changes in parameters(to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including: Spent fuel pool water level.

Proposed Question: RO 60 Rev: 0 Given:

The RAB watch reports lowering Spent Fuel Pool level The crew enters OP-901-513, Spent Fuel Pool Cooling Malfunction The crew will makeup to the Spent Fuel Pool from the ____(1)____. The ATC will verify the Spent Fuel Pool Cooling Pumps trip at __(2)__ MSL.

(1) (2)

A. Condensate Storage Pool only 416 B. Refuel Water Storage Pool or 416 Condensate Storage Pool C. Refuel Water Storage Pool or 439 Condensate Storage Pool D. Condensate Storage Pool only 439 Revision 0 Facility: Waterford 3 Page 119 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Makeup to the SFP is done frequently on shift. The makeup is always from the non-borated CSP to make-up for evaporated losses. However, the off-normal procedure allows makeup from either source. Part 2 is correct.

B. CORRECT: Makeup can be aligned to either the CSP or the RWSP. The Fuel Pool Cooling Pumps trip at a Spent Fuel Pool level of 416 MSL. The ATC will be required to trip the SFP cooling pumps if it does not happen at this level.

C. Incorrect. Part 1 is correct. Due to Single Point Vulnerability, an additional level switch is required (2/2) to trip the SFP Cooling Pumps. The level switch that was added sends its trip signal at 439 MSL. This will make up half the logic but not trip the SFP Cooling Pumps.

D. Incorrect. Makeup to the SFP is done frequently on shift. The makeup is always from the non-borated CSP to make-up for evaporated losses. However, the off-normal procedure allows makeup from either source. Due to Single Point Vulnerability, an additional level switch is required (2/2) to trip the SFP Cooling Pumps. The level switch that was added sends its trip signal at 439 MSL. This will make up half the logic but not trip the SFP Cooling Pumps Technical Reference(s): OP-901-513 page 4 and 6 Rev. 20 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9, 10 55.43 Revision 0 Facility: Waterford 3 Page 120 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 041 A3.02 Importance Rating 3.3 K/A Statement 041 Steam Dump/Turbine Bypass Control, A3.02: Ability to monitor automatic operation of the SDS, including: RCS Pressure, RCS Temperature, and reactor power.

Proposed Question: RO 61 Rev: 0 Given:

Plant is initially at 100% power RCS Tavg is 573 F Reactor Power Cutback system is out of service Reactor trip on Turbine trip is enabled Turbine trip occurs Which of the following describes the Steam Bypass Control system response immediately following the Turbine trip?

A. All valves quick open.

B. Only valves 1 through 5 quick open.

C. Only valves 1 through 3 quick open.

D. No valves quick open.

Revision 0 Facility: Waterford 3 Page 121 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. All SBCVs do not quick open on a reactor trip. SBC valve #6 quick open is blocked any time there is a reactor trip.

B. CORRECT: SBC valves 1-5 will quick open on a reactor trip with plant power at 100% power. SBC valve #6 quick open is blocked any time there is a reactor trip C. Incorrect. The quick open mode for the SBC system operates in two groups of 3 valves. Only one group of SBCVs quick opening will not handle a trip from 100%

power.

D. Incorrect. RCS Tave less than 561 F would block all SBCVs from quick opening.

The initial conditions in this event has Tave at 573 F.

Technical Reference(s): SD-SBC page 25 Rev. 10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-SBC00 (As available)

Question Source: Bank # RO 63 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2012 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 122 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 045 A2.08 Importance Rating 2.8 K/A Statement 045 Main Turbine Generator, A2.08: Ability to (a) predict the impacts of the following malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Steam dumps are not cycling properly at low load, or stick open at higher load (isolate and use atmospheric reliefs when necessary).

Proposed Question: RO 62 Rev: 0 Given:

Plant is at 100% power One Steam Bypass Control Valve has failed open The crew has entered OP-901-221, Secondary System Transient The BOP will attempt to close the failed open Steam Bypass Control Valve from CP-1 by placing the Valve Mode (Permissive) Select Switch to the ____(1)____ position. If the Steam Bypass fails to close, the crew will reduce reactor power to less than 100% power by ____(2)____.

(1) (2)

A. OFF adjusting main turbine load B. MANUAL adjusting main turbine load C. OFF commencing direct boration to the RCS D. MANUAL commencing direct boration to the RCS Revision 0 Facility: Waterford 3 Page 123 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: To remove the permissive (to open) going to the failed open SBCS valve, the procedure directs the crew to take the permissive select switch to OFF.

OP-901-221 directs the crew to adjust Main Turbine load to reduce Reactor Power less than 100%.

B. Incorrect. Taking the Permissive switch to MANUAL will lock in a permissive to open the SBC Valve. Part 2 is correct.

C. Incorrect. Part 1 is correct. OP-901-221 directs the crew to adjust Main Turbine load to reduce Reactor Power less than 100% and to match Tave and Tref. Direct Boration to the RCS would reduce Reactor Power but will not assist in the excessive cooldown and would not be an immediate effect.

D. Incorrect. Taking the Permissive switch to MANUAL will lock in a permissive to open the SBC Valve. OP-901-221 directs the crew to adjust Main Turbine load to reduce Reactor Power less than 100%. Direct Boration to the RCS would reduce Reactor Power but will not assist in the excessive cooldown and would not be an immediate effect.

Technical Reference(s): OP-901-221 steps 4 and 6 Rev. 4 (Attach if not previously provided) SD-SBC page 24 Revision 10 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 7, 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 124 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 055 K3.01 Importance Rating 2.5 K/A Statement 055 Condenser Air Removal, K3.01: Knowledge of the effect that a loss or malfunction of the CARS will have on the following: Main condenser.

Proposed Question: RO 63 Rev: 0 Given:

Plant is at 100% power Condenser Vacuum Pump B has tripped To maintain a suction path for Condenser Air Evacuation PIG Rad Monitor (PRM-IRE-0004), the BOP will verify Condenser Vacuum Pump ____(1)____

is running. Condenser Vacuum Pumps are designed to automatically start at a Main Condenser Vacuum of ____(2)____ inches Hg.

(1) (2)

A. A 25 B. A 26 C. C 26 D. C 25 Revision 0 Facility: Waterford 3 Page 125 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Condenser Vacuum Pump B or C must be operating to maintain a suction path for Condenser Air Evacuation PIG Rad Monitor (PRM-IRE-0004). OP-901-220, Loss of Condenser Vacuum, requires the crew to commence a rapid plant down power at a Main Condenser Vacuum of 25 inches Hg.

B. Incorrect. Condenser Vacuum Pump B or C must be operating to maintain a suction path for Condenser Air Evacuation PIG Rad Monitor (PRM-IRE-0004). Part 2 is correct.

C. CORRECT: Condenser Vacuum Pump B or C must be operating to maintain a suction path for Condenser Air Evacuation PIG Rad Monitor (PRM-IRE-0004).

Standby Condenser Vacuum Pumps will auto start at 26 inches Hg on lowering Main Condenser Vacuum.

D. Incorrect. Part 1 is correct. OP-901-220, Loss of Condenser Vacuum, requires the crew to commence a rapid plant down power at a Main Condenser Vacuum of 25 inches Hg.

Technical Reference(s): OP-003-001 page 15 and 20 Rev. 20 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-AE00 obj. 6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10, 11 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 126 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 071 K1.04 Importance Rating 2.7 K/A Statement 071 Waste Gas Disposal, K1.04: Knowledge of the physical connections and/or cause effect relationships between the Waste Gas Disposal System and the following systems: Station ventilation Proposed Question: RO 64 Rev: 0 Given:

The crew is scheduled to perform Section 6.4, Discharging Gas Decay Tank, per OP-007-003, Gaseous Waste Management Discharging ____(1)____ is the preferred method. The crew is required to secure the release if at least one RAB normal __(2)__ fan is not running.

(1) (2)

A. all three Gas Decay exhaust Tanks simultaneously B. all three Gas Decay supply Tanks simultaneously C. one Gas Decay Tank at a supply time D. one Gas Decay Tank at a exhaust time Revision 0 Facility: Waterford 3 Page 127 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Discharging all GDTs simultaneously is the preferred method (CR 1291) due to the potential leakage of GDT inlet valves. At least one RAB Normal Exhaust Fan shall be operating while discharging. This guidance is in the caution statement before the release is commenced.

B. Incorrect. Part 1 is correct. At least one RAB Normal Exhaust Fan shall be operating while discharging. Integrated knowledge is required here because the applicant will need to know that the supply fan cannot run alone because it would trip if it does not see required exhaust flow.

C. Incorrect. Part 1 is correct. At least one RAB Normal Exhaust Fan shall be operating while discharging. Integrated knowledge is required here because the applicant will need to know that the supply fan cannot run alone because it would trip if it does not see required exhaust flow.

D. Incorrect. Discharging all GDTs simultaneously is the preferred method (CR 1291) due to the potential leakage of GDT inlet valves. Part 2 is correct.

Technical Reference(s): OP-007-003 Caution on page 14 Rev. 307 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-GWM00 obj. 9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Comments: Submitted for free review. No changes are required.

Revision 0 Facility: Waterford 3 Page 128 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 075 2.1.30 Importance Rating 4.4 K/A Statement 075 Circulating Water, 2.1.30: Ability to locate and operate components, including local controls.

Proposed Question: RO 65 Rev: 0 Given:

A Shutdown of the Circulating Water System is in progress Circulating Water Pumps C and D are operating The BOP takes the control switch for Circulating Water Pump C to stop. Circulating Water Pump D ____(1)____.

If required, manual operation of Circ Water Pump C or D Discharge Isolation Valves (CW-103C(D)) will be performed ___(2)___ .

(1) (2)

A. can be stopped after in the turbine generator building 100 seconds B. will trip immediately at the intake structure C. will trip immediately in the turbine generator building D. can be stopped after at the intake structure 100 seconds Revision 0 Facility: Waterford 3 Page 129 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect: When securing the first two Circ Water pumps (A and B in this case), a 100 second delay interlock is required before tripping them, the last two pumps secure simultaneously. The Circulating Water Pump discharge (inlet to the condenser) is located in the turbine generator building but is not the valve to be verified closed per OP-003-006.

B. CORRECT: With only two circulating water pumps running, stopping any one pump will cause both pumps to trip simultaneously. Circ Water pump discharge valves are located at the intake structure. The discharge valves should have closed automatically but W3 has OE requiring manual control of these valves.

C. Incorrect. Part 1 is correct. The Circulating Water Pump discharge (inlet to the condenser) is located in the turbine generator building but is not the valve to be verified closed per OP-003-006.

D. Incorrect. With only two circulating water pumps running, stopping any one pump will cause both pumps to trip simultaneously. Part 2 is correct.

Technical Reference(s): OP-003-006 page 65 Rev. 320 (Attach if not previously provided) OP-003-006 page 105 Rev. 320 (including version/revision number) SD-CW page 16 Rev. 17 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CW00 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 130 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.3 Importance Rating 3.7 K/A Statement 2.1.3 Knowledge of shift or short-term relief turnover practices.

Proposed Question: RO 66 Rev: 0 A Reactor Operator has not stood an on-shift watch in the previous four weeks.

Per OI-042-000, Watch Station Processes, the minimum required Station Logs to be reviewed prior to turnover is ________ week(s).

A. one B. two C. three D. four Revision 0 Facility: Waterford 3 Page 131 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect: OI-042-000, Attachment 6.4 first bullet, states prior to turnover, review the Station Log (since the last shift or two weeks minimum). Not a subset because the question is asking what the procedure states.

B. CORRECT: OI-042-000, Attachment 6.4 first bullet, states prior to turnover, review the Station Log (since the last shift or two weeks minimum). In this case, the last shift stood was 4 weeks ago, therefore two weeks of logs must be reviewed.

C. Incorrect. OI-042-000, Attachment 6.4 first bullet, states prior to turnover, review the Station Log (since the last shift or two weeks minimum). Three weeks would be OK but is not the minimum required presented in OI-042-000.

D. Incorrect. Even though the RO has not stood a watch in four weeks, OI-042-000 only requires him/her to review two weeks of logs prior to turnover.

Technical Reference(s): OI-042-000 Attachment 6.4, rev. 45 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 132 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.19 Importance Rating 3.9 K/A Statement 2.1.19 Ability to use plant computers to evaluate system or component status.

Proposed Question: RO 67 Rev: 0 Which of the following control board annunciators would be in alarm if the Plant Monitoring Computer fails?

A. CEA DISABLED B. CEDMCS TIMER FAILURE C. CEDMCS MAINTENANCE ERROR D. POWER DEPENDENT INSERTION LIMIT Revision 0 Facility: Waterford 3 Page 133 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. This alarm is generated by contacts for CEA Circuit Breakers in OFF.

B. Incorrect. This alarm is generated by ACTM Micro-processer Card.

C. Incorrect. This alarm is generated by CEDMCS when trying to place more than one subgroup on hold bus.

D. CORRECT. This alarm is generated by the PMC.

OP-500-008 Atts. 4.88, 4.100, 4.99, 4.78 Rev.

Technical Reference(s): 28,40 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PMC00 obj. 3 (As available)

Question Source: Bank # RO 67 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 RO Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 134 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.44 Importance Rating 3.9 K/A Statement 2.1.44 Knowledge of RO duties in the control room during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

Proposed Question: RO 68 Rev: 0 Given:

Plant is in a refueling outage Refueling Group is preparing to withdraw the first fuel assembly from the Reactor Vessel The Source Range Neutron Flux Monitors shall be operable and operating with (1) channel(s) operable with continuous visible indication in the control room and (2) channel(s) operable with audible indication in the containment and the control room.

(1) (2)

A. two one B. one one C. two two D. one two Revision 0 Facility: Waterford 3 Page 135 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: The RO will take Tech Spec logs on the Source range neutron flux monitors. They are verified operable by ensuring at least 2 channels operable with continuous visible indication in the control room and 1 channel operable with audible indication in the containment and control room. (TS 3.9.2)

B. Incorrect. Source range neutron flux monitors are verified operable by ensuring at least 2 channels operable with continuous visible indication in the control room.

C. Incorrect. TS 3.9.2 requires only 1 channel operable with audible indication in the containment and control room.

D. Incorrect. Source range neutron flux monitors are verified operable by ensuring at least 2 channels operable with continuous visible indication in the control room and 1 channel operable with audible indication in the containment and control room. (TS 3.9.2)

Technical Reference(s): TS 3.9.2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ04 obj. 2 (As available)

Question Source: Bank # RO 62 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2012 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2,6 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 136 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.6 Importance Rating 3.0 K/A Statement 2.2.6 Knowledge of the process for making changes to procedures.

Proposed Question: RO 69 Rev: 0 Per OI-019-000, Operations Procedure Administration, a Procedure Improvement Request (PIR) must be verified to be complete, accurate, and proper justification provided by obtaining a review from ____(1)____ prior to forwarding the PIR to the

______(2)_____.

(1) (2)

A. the requestors immediate Operations Procedure supervisor Administrative Group (OPAG)

B. a peer evaluator Operations Procedure Administrative Group (OPAG)

C. a peer evaluator Work Management Center (WMC)

D. the requestors immediate Work Management Center supervisor (WMC)

Revision 0 Facility: Waterford 3 Page 137 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: OI-019-000 step 5.1.2.1 directs the PIR requestors immediate supervisor to verify the PIR is complete, accurate and has proper justification provided. Once the verification is complete, step 5.1.2.1.2 directs the PIR to be submitted to OPAG.

B. Incorrect. OI-019-000 step 5.1.2.1 directs the PIR requestors immediate supervisor to verify the PIR is complete, accurate and has proper justification provided. Part 2 is correct.

C. Incorrect. OI-019-000 step 5.1.2.1 directs the PIR requestors immediate supervisor to verify the PIR is complete, accurate and has proper justification provided. The WMC has many operations functions but it is not where OI-019-000 directs the PIR to be submitted.

D. Incorrect. Part 1 is correct. The WMC has many operations functions but it is not where OI-019-000 directs the PIR to be submitted.

Technical Reference(s): OI-019-000 page 8 rev. 306 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA obj. 2 (As available)

Question Source: Bank #

Modified Bank # RO 69 (Note changes or attach parent)

New Question History: Last NRC Exam 2012 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 138 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.35 Importance Rating 3.6 K/A Statement 2.2.35 Ability to determine Technical Specification Mode of Operation.

Proposed Question: RO 70 Rev: 0 Given the following plant conditions:

RCS average temperature: 425 F All control rods: Fully inserted What is the current plant Mode as defined in Technical Specifications for these conditions?

A. MODE 1 B. MODE 2 C. MODE 3 D. MODE 4 Revision 0 Facility: Waterford 3 Page 139 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Mode 1 is defined as Keff > .99 and > 5% power. These parameters would require control rods to be withdrawn.

B. Incorrect. Shutdown Bank control rods would have to be withdrawn for Mode 2. W3 enters mode 2 when regulating groups are initially withdrawn.

C. CORRECT: Mode 3 is defined as Keff <.99, tavg >350F.

D. Incorrect. Mode 4 is defined as Keff <.99 and Tave between 200F and 350F.

Technical Reference(s): TS Definitions Table 1.2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-TS00 (As available)

Question Source: Bank # RO66 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2007 RO Makeup Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 140 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.39 Importance Rating 3.9 K/A Statement 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems.

Proposed Question: RO 71 Rev: 0 In accordance with OP-100-014, Technical Specification and Technical Requirements Compliance, the crew is required to enter the appropriate cascading Technical Specifications upon declaring a(an) ___(1)____ inoperable.

The crew will be required to ____(2)____ within one hour.

(1) (2)

A. Emergency Diesel Generator complete OP-903-066, Electrical Breaker Alignment Check B. Component Cooling Water Train complete OP-903-066, Electrical Breaker Alignment Check C. Component Cooling Water Train verify Emergency Feedwater Pump AB operable D. Emergency Diesel Generator verify Emergency Feedwater Pump AB operable Revision 0 Facility: Waterford 3 Page 141 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect: Even though the EDG supplies power to various loads during an emergency, W3 does not cascade due to an EDG being inoperable. Part 2 is correct.

B. CORRECT: OP-100-014, Technical Specification and Technical Requirements Compliance designates systems that required cascading Tech Specs, CCW is one of those systems. The Electrical Breaker Lineup Check is required to be completed w/i 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with TS 3.8.1.1b.

C. Incorrect. Part 1 is correct. TS 3.8.1.1d (verify EFW Pump AB operable) is required during cascading Tech Specs but must be performed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> D. Incorrect. Even though the EDG supplies power to various loads during an emergency, W3 does not cascade due to an EDG being inoperable. TS 3.8.1.1d (verify EFW Pump AB operable) is required during cascading Tech Specs but must be performed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Technical Reference(s): OP-100-014 page 43 Revision 336 (Attach if not previously provided) TS 3.8.1.1 actions b and d (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00 obj. 9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Revision 0 Facility: Waterford 3 Page 142 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.13 Importance Rating 3.4 K/A Statement 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Proposed Question: RO 72 Rev: 0 Given:

The plant is at 10% Power A containment entry is desired Which of the following areas inside containment are forbidden from being entered?

A. Pressurizer Cubicle below +21 elevation B. +46 elevation at the Quench Tank C. Main Steam Line Crossovers on the +46 elevation D. RCS Cold Leg 1A penetration through the D Ring Wall Revision 0 Facility: Waterford 3 Page 143 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. This is an area known to have high radiation levels, but is not listed in HP-001-213 as being forbidden in MODE 1 (step 5.2.2). Also, not an area listed as needing RP Manager approval to enter (step 5.2.1).

B. Incorrect. This area is in close proximity to the Reactor Cavity but is a sufficient distance away that it is not forbidden (step 5.2.2), or need RP Manager approval to enter (step 5.2.1).

C. Incorrect. This is an exception to the requirement for obtaining RP Manager approval for going above the actual +46 elevation in Containment (step 5.2.1). It is not listed as forbidden (step 5.2.2).

D. CORRECT. Per HP-001-213, Step 5.2.2 and Attachment 7.1, this is a forbidden area in MODE 1. (> 5% RTP).

Technical Reference(s): HP-001-213, Rev. 305 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 (As available)

Question Source: Bank # RO 71 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2011 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9,12 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 144 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.15 Importance Rating 2.9 K/A Statement 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: RO 73 Rev: 0 Per OI-038-000, Emergency Operating Procedures Operations Expectations/Guidance, following a momentary loss of power or voltage dip, some radiation monitors needed for Emergency Plan may require ___________?

A. a restart of the sample pump B. verification of proper setpoint C. reset of their power supply D. local purge operation Revision 0 Facility: Waterford 3 Page 145 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: OI-038-000, step 5.3.4.1, states that some radiation monitors sample pumps needed for Emergency Plan may require restarting following a loss of power or voltage dips. Failure to restart sample pumps could delay the ED classification (W-3 specific issue)

B. Incorrect: OI-038-000, step 5.3.4.1, states that some radiation monitors sample pumps needed for Emergency Plan may require restarting following a loss of power or voltage dips. This distractor is plausible because it is an action that can be performed on a rad monitor per OP-004-001, Radiation Monitors C. Incorrect: OI-038-000, step 5.3.4.1, states that some radiation monitors sample pumps needed for Emergency Plan may require restarting following a loss of power or voltage dips. This distractor is plausible because it is feasible that a voltage dip could trip the local supply breaker.

D. Incorrect: OI-038-000, step 5.3.4.1, states that some radiation monitors sample pumps needed for Emergency Plan may require restarting following a loss of power or voltage dips. This distractor is plausible because it is an action that can be performed on a rad monitor per OP-004-001, Radiation Monitors Technical Reference(s): OI-038-000 step 5.3.4.1 Rev. 13 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 4 (As available)

Question Source: Bank # RO 72 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2014 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10, 11 55.43 Revision 0 Facility: Waterford 3 Page 146 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.8 Importance Rating 3.8 K/A Statement 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Proposed Question: RO 74 Rev: 0 Which of the following is prohibited by OP-100-017, Emergency Operating Procedure Implementation Guide?

A. Simultaneous performance of more than one safety function concurrently while in the Functional Recovery Procedure.

B. Simultaneous performance of an Optimal Recovery Procedure and an Off-Normal Procedure C. Simultaneous performance of more than one Optimal Recovery Procedure D. Simultaneous performance of an Optimal Recovery Procedure and a System Operating Procedure Revision 0 Facility: Waterford 3 Page 147 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The simultaneous performance of more than one Safety Function concurrently while in the Functional Recovery procedure is allowed per OP-100-016, step 5.16.4.

B. Incorrect. Simultaneous performance of an Optimal Recovery Procedure and an off-normal procedure is allowed.

C. CORRECT: Step 5.11.1 of OP-100-017, Emergency Operating Procedure Implementation Guide, states that simultaneous performance of more than one Optimal Recovery Procedure is prohibited.

D. Incorrect. Simultaneous performance of an Optimal Recovery Procedure and a system operating procedure is allowed.

Technical Reference(s): OP-100-017 step 5.11.1, and 5.16.4 Rev.4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 148 of 150

2017 NRC Exam RO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.25 Importance Rating 3.3 K/A Statement 2.4.25 Knowledge of fire protection procedures.

Proposed Question: RO 75 Rev: 0 A site Fire Brigade of at least __(1)__ members shall be maintained on site at all times.

The fire brigade composition may be less than the minimum requirement for a period of time not to exceed a maximum of ___(2)___ hour(s).

(1) (2)

A. four one B. four two C. five one D. five two Revision 0 Facility: Waterford 3 Page 149 of 150

2017 NRC Exam RO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The applicant will select four fire brigade members if he does not include the fire brigade leader. The fire brigade limit of 5 is inclusive of the fire brigade leader. The fire brigade composition may be less than the minimum requirement for a period of time not to exceed a maximum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. Incorrect. The applicant will select four fire brigade members if he does not include the fire brigade leader. The fire brigade limit of 5 is inclusive of the fire brigade leader. Part 2 is correct.

C. Incorrect. Part 1 is correct. The fire brigade composition may be less than the minimum requirement for a period of time not to exceed a maximum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. CORRECT: A site Fire Brigade of at least five members shall be maintained on site at all times. The fire brigade composition may be less than the minimum requirement for a period of time not to exceed a maximum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Technical Reference(s): FP-001-020 step 3.1 Rev. 311 (Attach if not previously provided) OI-042-000 page 22 Rev. 45 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 obj. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Revision 0 Facility: Waterford 3 Page 150 of 150

2017 NRC Written Examination Waterford 3 Reactor Operator and Senior Reactor Operator

1. C 26. C

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ES-401 Site-Specific RO Written Examination Form ES-401-7 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:

Date: April 5, 2017 Facility/Unit: Waterford 3 Region: IV Reactor Type: CE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results Examination Value 75 Points Applicants Score Points Applicants Grade Percent

2017 NRC Written Examination Waterford 3 Reactor Operator

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Technical Guide for TG-OP-902-001 Reactor Trip Recovery Procedure Revision 306 Step Number 12 Protect Main Condenser Objective The intent of this step is to prevent over-pressurizing the Main Condenser in the event of a loss of condenser vacuum. The most likely cause would be a total or partial loss of Circulating Water due to a loss of electrical power to the CW pumps. However, this guidance also applies to any event that could lead to a loss of condenser vacuum. This step also ensures that the operator has control of RCS cooldown rate and SG inventory.

Instructions The Operator verifies that the Main Condenser has sufficient cooling provided by circulating Water and Condenser Vacuum is being maintained. If cooling to the condenser is not available or the condenser vacuum is not maintained then implement the contingency and isolate the condenser. . The Circulation Water system procedure is reference to provide instructions to restore CW system operation if needed.

In the event that a single train of off-site power is lost the circulating water pumps (SWGR 1A or SWGR 1B power) on that train will be deenergized. Also the discharge valves (MCC 221A or MCC 221B power) associated with those pumps will be deenergized and open. If the CW Pumps were in a 4 of 4 operating configuration, the open discharge valves will result in some of the circulating water flow from the two operating CW pumps to recirculate back into the bay. If the CW Pumps are in a 3 of 4 operating configuration and the train of off-site power that is lost was carrying two operating CW pumps, then the last operating CW pump, on the train of off-site power that is energized, will also trip.

The following OE was obtained from the Reactor Trip (CR-WF3-2015-03565) June 03, 2015: The reactor tripped due to a feedwater casualty. The A CWP was tagged out and the other 3 CW Pumps were operating. The B side power did not transfer to the SUT and both the 1B and 2B buses de-energized. Since the 1B bus de-energized all 3 CWPs tripped. If 3 CWPs are operating and 2 trip the third pump trips. For CWP B & D this also meant that the discharge valve motors were de-energized and the discharge valves did not close setting up the CW bay to short cycle. Since CWP C discharge valve was energized and the valve did go closed on the pump trip. Condenser vacuum slowly lowered until CWP C was restarted and then condenser vacuum held steady. The discharge valve only got to 45% open and jammed; this is believed to be due to the DP set up by the short cycling recirculation back to the bay. The valve then had to be manually opened locally; several personnel were required to assist. Once the discharge valve was fully opened vacuum restored to 25 to 26 inches. While only one CWP was in operation the amps on the pump were above normal until B train power was restored and other CW pumps were brought back in service.

24

WATERFORD 3 SES OP-902-001 Revision 016 Page 8 of 19 REACTOR TRIP RECOVERY PROCEDURE INSTRUCTIONS CONTINGENCY ACTIONS Protect Main Condenser

• 12. Perform BOTH of the following: 12.1 Perform the following:

Verify CW System in operation. a. Verify BOTH MSIVs are closed:

REFER TO OP-003-006, Circulating Water. MS 124A, MSIV 1 Check Condenser vacuum greater MS 124B, MSIV 2 than 14 Hg.

b. Verify ALL Steam Generator Blowdown Isolation valves are closed:

BD 102A, STM GEN 1 (IN)

BD 102B, STM GEN 2 (IN)

BD 103A, STM GEN 1 (OUT)

BD 103B, STM GEN 2 (OUT)

WATERFORD 3 SES OP-902-001 Revision 016 Page 16 of 19 REACTOR TRIP RECOVERY PROCEDURE SAFETY FUNCTION

6. RCS Heat Removal PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. At least ONE Steam 10 - 76% NR level Generator capable of and MFW steaming available

b. RCS TC 530 - 550 F Condition 2

a. At least ONE Steam MFW or EFW Generator capable of restoring level to steaming 55 - 70% NR.

b. RCS TC 530 - 550 F Condition 3

a. At least ONE Steam 100% capacity Generator capable of EFW available to steaming restore level

b. RCS TC 530 - 550 F

Technical Guide for Loss of TG-OP-902-002 Coolant Accident Recovery Procedure Revision 019 Trending of Key Parameters (contd)

Pressurizer Level Pressurizer level may decrease or increase. For breaks not located in the pressurizer, the pressurizer will empty and, depending on the size of the break, not refill during the course of the accident. Breaks located in the pressurizer may lead to increased pressurizer level since water from the hot leg flows into the pressurizer surge line while significant voiding of the RCS loop is occurring. If there is a break on or near the pressurizer level instruments, this may cause this instrument to be grossly inaccurate and misrepresent pressurizer level (high or low).

For small break L0CAs where the pressurizer refills as a result of safety injection, pressurizer level may not be representative of RCS inventory or core coverage. As indicated above, the depressurization associated with a leak in the RCS will usually result in the formation of voids in RCS hot spots (reactor vessel head, hot legs, S/G tube bundle). The growth or persistence of these voids, after refill of the pressurizer by the safety injection system, may cause pressurizer level to increase or remain constant in spite of continuing loss of inventory through the break.

Steam Generator Pressure Steam generator pressure may increase or remain constant in the short-term if the break is small. However, for all sized L0CAs, steam generator pressure will usually decrease in the long term as a result of operator action.

Steam Generator Level Steam generator level will decrease rapidly following the reactor trip and then increase to the hot standby level. Level may then remain constant or increase somewhat based on automatic or manual control of feedwater.

10

LOCA in the Pressurizer What are the differences between a Pressurizer LOCA and a LOCA in the RCS Loops?

Pressurizer level will rise rapidly vice lower due to the depressurization of the steam space and the Steam Void shift to the Reactor Vessel Head The lifting of a Pressurizer Safety Valve will cause the downstream tailpipe temperature to match Tsat for the Quench Tank pressure One HPSI pumps flow will keep the core covered RCS Decay heat and temperature are low enough to regain subcooled margin on the discharge head of the HPSI pump Pzr Safety Opening is 0.0273 ft2.

58

Operating Instruction OI-038-000 Emergency Operating Procedures Revision 013 Operations Expectations / Guidance 5.4.45 Maintain Least Affected SG Level None 5.4.46 Maintain Long Term Cooling None 5.4.47 Maintain Pressurizer Level Following an SIAS, all available charging pumps is defined as two selected pumps operating. This is the normal system response following an SIAS. It is not necessary to start the standby pump. If power is available to only one charging pump following an SIAS, that charging pump is considered all available.

If SIAS has not initiated, all available charging pumps are defined as three charging pumps operating. The operator should start and stop charging pumps as necessary to maintain Safety Functions. It is not necessary to emergency borate with three charging pumps if it results in jeopardizing other safety functions.

Water solid operations of the pressurizer should be avoided unless minimum subcooling (28 F) cannot be maintained in the RCS. If the RCS is solid, closely monitor any makeup or draining and any system heatup or cooldown, to avoid any unfavorable rapid pressure excursions.

Once a pressurizer cooldown has begun, pressurizer level indication decalibration will occur. The indication on the normal pressurizer level indication will begin to deviate from the true pressurizer level. To find true pressurizer water level, the operator should use correction curves in the Plant Data Book.

5.4.48 Maintain RCS Pressure If there is a conflict between maintaining adequate core cooling and complying with the RCP operation limits, then maintaining adequate core cooling will be given the higher priority. Subcooling of 28 F has precedence over reducing leakage through the Steam Generator Tubes.

5.4.49 Maintain RCS within RCS PT Limits During insurge to the pressurizer, saturation conditions may be lost.

Pressurizer heaters are required to regain pressure control.

25

WATERFORD 3 SES OP-902-002 Revision 020 Page 69 of 83 LOSS OF COOLANT ACCIDENT RECOVERY 5.0 SAFETY FUNCTION STATUS CHECK SAFETY FUNCTION:

1. Reactivity Control PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. Reactor Power Dropping

b. SUR Negative

c. ONE of the following:

CEAs < 2 NOT fully inserted Emergency boration 40 gpm Shutdown margin Verified Condition 2

a. Reactor Power < 10-4% stable or dropping

b. ONE of the following:

CEAs < 2 NOT fully inserted Emergency boration 40 gpm Shutdown margin Verified

WATERFORD 3 SES OP-902-009 Revision 315 Page 198 of 206 STANDARD APPENDICES Appendix 35 Page 1 of 2 35.0 Single CCW Pump Operation INSTRUCTIONS

1. IF only ONE CCW pump operating, THEN split CCW headers as follows:

a. IF CCW Pump A is operating, THEN close CCW Suction and Discharge Header Tie Valves:

CC 126A/114A, AB TO A CC 200B/563, B TO AB

b. IF CCW Pump AB is operating, AND replacing CCW Pump A, THEN close CCW Suction and Discharge Header Tie Valves:

CC 126B/114B, AB TO B CC 200B/563, B TO AB

c. IF CCW Pump B is operating, THEN close CCW Suction and Discharge Header Tie Valves:

CC 127B/115B, AB TO B CC 200A/727, A TO AB

d. IF CCW Pump AB is operating, AND replacing CCW Pump B, THEN close CCW Suction and Discharge Header Tie Valves:

CC 127A/115A, AB TO A CC 200A/727, A TO AB

WATERFORD 3 SES OP-902-009 Revision 315 Page 199 of 206 STANDARD APPENDICES Appendix 35 Page 2 of 2 INSTRUCTIONS

2. IF an isolated CCW Train does NOT have an operating CCW pump, THEN perform ALL of the following for the AFFECTED CCW Train:

Pull EDG A(B) Overspeed Trip Device Place HPSI A(B) to OFF.

Place LPSI A(B) to OFF.

Place CS Pump A(B) to OFF.

IF HPSI AB aligned for service, THEN place Assignment Switch to NORM.

3. IF an isolated CCW Train does NOT have an operating CCW pump, AND the associated 4.16 KV Safety Bus is energized, THEN evaluate starting BOTH of the following for the affected CCW Train:

a. ACCW A(B) in WET Mode

b. Essential Chiller A(B)(AB)

End of Appendix 35

Off Normal Procedure OP-901-130 Reactor Coolant Pump Malfunction Revision 011 E1 SEAL FAILURE PLACEKEEPER START DONE N/A NOTE

1. RCP Seal pressure and Control Bleedoff temperature and flow are normally as follows (assuming normal operating RCS temperature and pressure):

Vapor Seal pressure: 25 to 45 PSIG Upper Seal pressure: 585 to 915 PSIG Middle Seal pressure: 1237 to 1815 PSIG CBO temperature: 135° to 190°F CBO flow: 1.2 to 1.8 GPM

2. The following parameters are indicative of RCP seal failure:

Any seal pressure equal to RCS pressure Two or more seal pressures approximately equal to each other Controlled Bleedoff flow greater than 2.0 GPM Inability to maintain Seal CCW Cooler Return Temperature <145°F A failed stage is indicated by a differential pressure of less than 100 psid across the stage

3. If only one Reactor Coolant Pump Seal has failed on a Reactor Coolant Pump, then pump operation may continue provided the seal package is monitored for further degradation.

1. Inform System Engineer of Reactor Coolant Pump Seal failure.

9

Off Normal Procedure OP-901-112 Charging or Letdown Malfunction Revision 006 C AUTOMATIC ACTIONS

1. IF Pressurizer level rises above program level, THEN the following occur:

1.1 Letdown flow rises to a maximum of 126 gpm.

1.2 IF level rises 4.0 above program level, THEN the following occur:

Backup Charging Pumps receive backup stop signal.

IF Pressurizer pressure 2270 PSIA, THEN Backup Heaters energize.

2. IF Pressurizer level drops below program level, THEN the following occur:

2.1 Letdown flow drops to a minimum of 28 gpm.

2.2 IF level drops 2.5 below program level, THEN first backup Charging Pump starts.

2.3 IF level drops 3.9 below program level, THEN second backup Charging Pump starts.

2.4 IF level drops 6.0 below program level, THEN BOTH backup Charging Pumps receive backup start signal.

2.5 IF level drops to 28 , THEN ALL Pressurizer Heaters de-energize.

3. IF REGEN HX TUBE OUTLET temperature 470 F, THEN LETDOWN STOP VALVE (CVC 101) closes.

4. IF LETDOWN HX TUBE OUTLET temperature >140 F, THEN the ION EXCHANGER BYPASS (CVC 140) bypasses Ion Exchangers

5. IF VCT level 5.5%, THEN the following occur:

RWSP TO CHARGING PUMPS (CVC 507) opens VCT DISCH VALVE (CVC 183) closes 5

Off Normal Procedure OP-901-112 Charging or Letdown Malfunction Revision 006 E1 CHARGING MALFUNCTION PLACEKEEPER START DONE N/A NOTE If all Charging Pumps are secured, then LETDOWN STOP VALVE (CVC 101) will close on high REGEN HX TUBE OUTLET temperature if RCS is 470 F.

CAUTION THE REACTOR COOLANT SYSTEM WILL BE BORATED IF A CHARGING PUMP IS STARTED WITH THE RWSP AS THE MAKEUP WATER SOURCE.

1. IF Charging Pumps have tripped, THEN perform the following:

1.1 Verify open EITHER VCT DISCH VALVE (CVC 183) OR RWSP TO CHARGING PUMP (CVC 507).

1.2 IF Letdown has NOT isolated, THEN attempt to restart Charging Pump(s).

1.3 IF the Charging Pump can NOT be restarted, THEN verify closed LETDOWN STOP VALVE (CVC 101).

1.4 IF the reason for the Charging pump trip is corrected AND Pressurizer level is in normal operating band, THEN place Charging and Letdown in service in accordance with Attachment 2.

8

Off Normal Procedure OP-901-131 Shutdown Cooling Malfunction Revision 304 A. PURPOSE

1. Provide instructions for Shutdown Cooling System leakage.

2. Provide instructions for Loss of Shutdown Cooling flow.

3. Provide instructions for Loss of Shutdown Cooling heat removal capability.

2

Off Normal Procedure OP-901-131 Shutdown Cooling Malfunction Revision 304 E3. LOSS OF SHUTDOWN COOLING HEAT REMOVAL CAPABILITY PLACEKEEPER START DONE

1. IF CCW Pump has tripped, THEN start standby CCW pump.

2. Verify shutdown cooling train with operating CCW pump is in service in accordance with OP-009-005, SHUTDOWN COOLING SYSTEM.

3. Implement OP-901-510, COMPONENT COOLING WATER SYSTEM MALFUNCTION, concurrently with this procedure.

CAUTION RAPID RESTORATION OF CCW FLOW WHEN SHUTDOWN COOLING HEAT EXCHANGER TEMPERATURE IS >200 F MAY CAUSE DAMAGE TO CCW PIPING.

4. IF CCW flow is NOT restored to affected Shutdown Cooling Heat Exchanger within 5 minutes AND RCS temperature is

>200 F, THEN perform the following:

4.1 Verify associated Shutdown Cooling Heat Exchanger Outlet valve in SETPNT:

SHDN HX A OUTLET valve (CC 963A)

OR SHDN HX B OUTLET valve (CC 963B) 4.2 Close Temperature Control valve for affected Train:

SDCS LOOP 2 TEMPERATURE CONTROL (SI 415A)

OR SDCS LOOP 1 TEMPERATURE CONTROL (SI 415B) 4.3 Place associated LPSI Header Flow controller to MAN AND adjust output to 10%.

LPSI B DISCH HDR FLOW controller (SI IFIC 0306)

OR LPSI A DISCH HDR FLOW controller (SI IFIC 0307) 31

Component Cooling Water OP-901-510 System Malfunction Revision 303 E2 LOSS OF CCW PUMP(S)

PLACEKEEPER START DONE

1. IF CCW is lost to in-service Shutdown Cooling train, THEN implement OP-901-131, SHUTDOWN COOLING MALFUNCTION, AND perform concurrently with this procedure.

2. IF Component Cooling Water Pump AB has tripped, THEN Start standby CCW Pump.

2.1 Place CCW ASSIGNMENT Switch to NORM Position.

3. IF Component Cooling Water Pump A has tripped, THEN align CCW Pump AB for Operation as follows:

3.1 Place CCW ASSIGNMENT Switch to A position.

3.2 Verify Open the following valves:

CC-126A/CC-114A CCW SUCT & DISCH HEADER TIE VALVES AB TO A CC-127A/CC-115A CCW SUCT & DISCH HEADER TIE VALVES AB TO A 3.3 Start CC-0001AB, CCW PUMP AB.

3.4 Evaluate AB Electrical Bus alignment for Technical Specification Operability requirements.

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Page 43 of 48 WATERFORD 3 SES OP-902-000 Revision 016 Page 5 of 14 STANDARD POST TRIP ACTIONS 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

---

NOTE ----------------------------------------------------------

Steps 1 and 2 are immediate actions and satisfy Reactivity Control.

Verify Reactivity Control

1. Determine Reactivity Control acceptance criteria are met:

a. Check Reactor power is a.1 Perform the following as necessary to dropping. insert CEAs:

1) Manually trip the Reactor.

2) Manually initiate DIVERSE REACTOR TRIP.

3) Open BOTH of the following breakers for 5 seconds and THEN close:

SST A32 FEEDER SST B32 FEEDER

4) Locally open ALL Reactor Trip breakers.

b. Check Startup Rate is negative.

c. Check less than TWO CEAs are c.1 Commence emergency boration.

NOT fully inserted.

WATERFORD 3 SES OP-902-007 Revision 017 Page 12 of 55 STEAM GENERATOR TUBE RUPTURE RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

• 10. (continued) (continued) 10.3 IF CCW Pump AB is the faulted pump, THEN perform the following:

a. Place the CCW ASSIGNMENT switch to the neutral position.

b. Start CCW Pump A(B).

c. IF CCW flow is NOT restored, THEN pull the affected EDG overspeed trip.

10.4 IF only ONE CCW pump operating, THEN split CCW headers using Appendix 35, Single CCW Pump Operation.

Cooldown RCS to Less Than 520°F T H

11. Commence a rapid RCS cooldown to 11.1 Commence a rapid RCS cooldown to less than 520°F TH using the Steam less than 520°F TH using BOTH Bypass Control valves. Atmospheric Dump valves.

11.2 IF necessary, THEN locally control ADVs using Appendix 22, Local Operation of the Atmospheric Dump Valves.

Technical Guide for Steam Generator Tube TG-OP-902-007 Rupture Recovery Procedure Revision 308 Step Number 11 Cooldown RCS to Less Than 520 F T Hot Objective The intent of this step is to ensure that the temperature of the RCS is lowered prior to isolating the affected steam generator to prevent opening the atmospheric dump valve (ADV) or lifting the main steam safety valves (MSSVs) on the affected steam generator after isolation.

Instructions This step is most easily accomplished when RCPs are operating and when one or more steam generators are providing cooling. If all RCPs have been tripped and natural circulation is the heat removal process, then it is necessary to cooldown both steam generators to provide uniform RCS cooling.

Under natural circulation flow conditions, Thot will rise as the core differential temperature rises as a result of the change from two loop to one loop heat removal. The temperature in the isolated SG will be essentially equivalent to Thot since it is no longer being used as a heat sink. The first MSSV opens at 1085 psia which corresponds to a saturation temperature of approximately 555°F. Allowing a margin below the corresponding saturation temperature and accounting for the rise in Thot that results in a value for isolating the affected SG. For forced flow conditions, the increase in Th is negligible. Thus, the value selected to cover natural circulation conditions will cover forced circulation conditions. If RCS hot leg temperature is not less than the desired value then the RCS should be cooled down to this value.

Natural circulation cooldown of the RCS is not effective for cooling the reactor vessel head region. If natural circulation cooling provides the reduction of Thot to less than saturation temperature corresponding to ADV setpoint, heat transfer to the steam generator from the RCS loops will not cause lifting of the ADV. However, the energy stored in the reactor vessel head region and pressurizer has to be dealt with to bring RCS pressure close to steam generator pressure to minimize leakage into the steam generator and to preclude the ADV opening due to filling the steam generator with high RCS pressure. Controlling RCS pressure with the pressurizer and with an un-cooled reactor vessel head region is addressed in later actions.

This action may be performed by feeding the steam generators with feedwater and dumping steam to the condenser via manual control of the steam bypass valves.

31

Technical Guide for Steam Generator Tube TG-OP-902-007 Rupture Recovery Procedure Revision 308 Step Number 11 Cooldown RCS to Less Than 520 F T Hot (contd)

Contingency Actions If the condenser or steam bypass system is not available, then the atmospheric dump valves may be used. However, it is less desirable to use the atmospheric dump valves to cooldown the RCS because of the unmonitored release of activity to the environment.

Operation of the ADVs is more desirable than operation of the MSSVs for two reasons.

First, the MSSVs may stick open or leak by. MSSVs can not be isolated, and the ADVs can. Secondly, the ADVs provide a more controlled release than do the MSSVs.

A contingency was added to reference a standard appendix for local operation of ADVs.

This provides additional flexibility for operation of the ADVs which provide a more controlled release than do the MSSVs.

Justification for Deviations There are no deviations.

References

1. PEIR OM-102

2. ECS98-001 T.10 32

WATERFORD 3 SES OP-902-004 Revision 016 Page 9 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Verify RCP Operating Limits

• 10. IF RCPs are operating, THEN perform the following:

a. Verify CCW available to RCPs. a.1 IF CCW is lost to RCPs, AND is NOT restored within 3 minutes, THEN stop the affected pumps.

a.2 IF CCW Flow Low to RCPs, AND Instrument Air is available, THEN place IA 909 to close and THEN open.

b. IF a CSAS is initiated, THEN stop ALL RCPs.

c. IF RCS TC is less than 380ºF [384ºF],

THEN verify ONE RCP in each loop is stopped.

d. Check RCP operating parameters: d.1 IF ANY RCP operating limit exceeded, NPSH, REFER TO Attachment THEN stop affected RCP(s).

2A-D, RCS P-T Limits Bearing temperatures less than or equal to 225°F Bleed Off temperature less than 200°F Cooling Coils Return CCW temp less than 155°F At Least Two Seals per RCP operable

Technical Guide for Excess Steam TG-OP-902-004 Demand Recovery Procedure Revision 307 Step Number 10 Verify RCP Operating Limits Objective The intent of this step is to verify that RCPs are operating within their acceptable operating conditions.

It is important to verify RCP operating limits in the EOPs given the off-normal conditions of the RCS so that any RCP that is not operating within their operating limits is stopped prior to being damaged.

Instructions The operator is expected to verify CCW flow to the RCPs. The CCW system provides cooling for the RCPs. Without proper cooling the likelihood of a RCP seal failure is increased.

If CCW is available to containment and CCW flow to RCPs is low an additional contingency is provided to check instrument air available to containment. A CIAS closes IA-909. An SIAS trips the CEDM fans. When the CEDM fans are secured, CC-646 the inlet to the CEDM coolers goes closed. The design CCW flow to the CEDM coolers is 700 gpm and it is in parallel with the RCP motor and seal coolers. CC-646 is an air operated valve that fails open. If CC-646 comes open due to a lack of instrument air, CCW flow will be diverted from the RCPs and thru the CEDM cooler. This may result in low flow to the RCP coolers. This is only a check to prompt the Operator to situational awareness and respond to accordingly.

A substep is included to secure all RCPs if a containment pressure has reached the CSAS setpoint. This step is performed for protection of the RCPs, since CCW, which provides cooling to the RCPs, is isolated upon CSAS actuation.

A plant specific step has been developed to verify no more than two RCPs are operating when Tcold is less than 380 F [384 F]. The RCS PT Limit Curve specifies that one RCP should be tripped when Tcold is less than 380 F [384 F] to prevent a core uplift problem.

A check of RCP operating parameters is provided. This step provides guidance to monitor various operating parameters such as seal staging, motor amps, pump DP, and associated annunciators clear to ensure any RCP that is operating outside their operating limits is stopped prior to being damaged. The limits provided are the same RCP operating limits provided in OP-901-130 RCP Malfunction. This step places all of the limits in one step for the Operators convenience.

28

WATERFORD 3 SES OP-902-006 Revision 018 Page 6 of 37 LOSS OF FEEDWATER RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Restore Operation of DCT Sump Pumps

• 5. IF power has been interrupted to either 3A or 3B Safety bus, THEN perform Appendix 20, "Operation of DCT Sump Pumps."

Stop RCPs

• 6. Stop all RCPs.

Technical Guide for Loss of Feedwater TG-OP-902-006 Recovery Procedure Revision 018 Step Number 7 Stop RCPs Objective The intent of this step is to cease adding pump heat to the RCS.

Instructions A Loss of Feedwater results in a reduction of the ability of the steam generators to remove heat from the RCS. Since natural circulation heat removal is adequate to remove the decay heat generated in the core, the RCPs are stopped to eliminate their heat input to the RCS.

The Operator is directed to secure all RCPs. The presumption is that this procedure has been entered as a result of inadequate feedwater and that reducing heat load is appropriate.

Historically if Main Feedwater is lost and only one motor driven Emergency Feedwater pump is available the operator is directed to secure all RCPs within the following 30 minutes. The 30 minute time clock for tripping RCPs began when there is a loss of main feedwater concurrent with the condition of the only available feedwater pump is one motor driven EFW pump. Waterford has now reduced the step to stop all RCPs to simplify the step. The Operator should not delay in tripping RCPs when arriving at this step. By stopping all RCPs the Operator meets the commitment to stop all RCPs within 30 minutes of operating with one MDEFW Pump. This step should be completed within 30 minutes of an inadequate feedwater condition. This does not mean to wait 30 minutes and then trip RCPs it means that RCPs are tripped when the Operator arrives at this step.

This step is the first major mitigating action taken in a LOF event.

Contingency Actions None 19

SD-EFW setpoint plus 3% tolerance. At this head, the turbine-driven EFW pump is at 4347 rpm and 655 GPM flow (575 plus 80 GPM recirculation flow).

SYSTEM DESCRIPTION The EFW system consists of the Condensate Storage Pool (CSP), two 50% capacity motor-driven pumps (EFW Pump A and B), a 100% capacity steam turbine-driven pump (EFW Pump AB) and the associated piping and valves required to supply water to either the intact steam generator or both steam generators.

The EFW system supplies water to the steam generators from the CSP to maintain a heat sink for the decay heat generated by the reactor core following emergency shutdowns. Failure to remove this decay heat will result in the temperature and pressure of the Reactor Coolant System (RCS) increasing. This would cause loss of inventory from the RCS if the Pressurizer Code Safety Valves reach their lift setting.

Excessive loss of water inventory in the RCS could cause overheating of the reactor core and possibly core damage.

As shown in Fig. 1, the three pumps take suction on the CSP via two suction lines and discharge to a common supply point. Each EFW pump can be isolated by suction and discharge valves which are normally locked open. The suction and discharge lines of the pumps are normally cross-connected by locked-open manual valves. Water is then delivered to each Main Feedwater header through an arrangement of four pneumatically-operated control and isolation valves.

The Wet Cooling Tower Basins of the Auxiliary Component Cooling Water (ACCW)

System are available to the EFW pumps as an alternate source of water. This source would only be used if the CSP became depleted.

Indicating lights are provided to monitor equipment status. All EFW system motors and remotely operated valves have status indicating lights in the main control room on CP-

8. Most are also duplicated on the Remote Shutdown Panel, LCP-43. Fig. 2 illustrates the EFW system instrumentation.

The discharge of the EFW pumps can supply either or both steam generators. Venturi flow detectors on each feed line provide indication on CP-8 and LCP-43 and through the plant monitoring computer. The EFW flow indicators on LCP-43 are qualified reliable instruments for a Case II fire condition. The meters provide indication of flowrate to each steam generator from 0 to 800 GPM.

Two safety-related flow indicating switches (with ranges of 0-1650 GPM) are provided on each line. In the event excessive flow exists on a feedline, these switches cause an Extremely High Flow annunciator to sound on CP-18's associated safety-related annunciator panels (SA and SB). This annunciation will alert the operator during a Revision 13 Page 6 of 94

WATERFORD 3 SES OP-902-005 Revision 020 Page 6 of 43 STATION BLACKOUT RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Restore Operation of DCT Sump Pumps

• 5. Perform Appendix 20, "Operation of DCT Sump Pumps."

---

NOTE ----------------------------------------------------------

If a Temporary Diesel generator is available and ready to load within 30 minutes of SBO onset, then load stripping via the associated Attachment 7-A or 7-B should not be necessary.

Performing Attachment 7-A and 7-B will remove DC control power from the EDGs, including loss of EDG annunciator panels and CP-1 indicating lights. An EDG will not start with control power removed.

Reduce Switchgear Battery Loads

• 6. IF AC power is NOT restored, THEN direct NAOs to perform ALL of the following within 30 minutes from the onset of SBO:

Attachment 7-A, "Switchgear Room A Removable Loads" Attachment 7-B, "Switchgear Room B Removable Loads" Attachment 7-C, "Switchgear Room AB Removable Loads"

WATERFORD 3 SES OP-902-009 Revision 315 Page 126 of 206 STANDARD APPENDICES Appendix 20 Page 1 of 2 20.0 Operation of DCT Sump Pumps INSTRUCTIONS

---

NOTE -----------------------------------------------------------

This attachment should be performed following any power interruption to either the 3A or 3B Safety bus (as directed from EOPs).

If a Probable Maximum Precipitation (PMP) event is in progress and any Dry Cooling Tower (DCT) Motor Driven Sump pump is unavailable, then both of the following shall be performed for the affected DCT sump to prevent flooding of associated 315A(B) Motor Control Center and Transformer within time frames as listed:

One DCT Motor Driven Sump pump is aligned for operation within 30 minutes of the PMP event.

The DCT Portable Sump pump (diesel driven) is aligned for operation within three hours of the PMP event.

1. At MCC-314A, place BOTH of the following switches to BYPASS:

DCT #1 Sump Pump A Radiation Monitor Bypass switch DCT #2 Sump Pump A Radiation Monitor Bypass switch

2. At MCC-314B, place BOTH of the following switches to BYPASS:

DCT #1 Sump Pump B Radiation Monitor Bypass switch DCT #2 Sump Pump B Radiation Monitor Bypass switch

WATERFORD 3 SES OP-902-009 Revision 315 Page 127 of 206 STANDARD APPENDICES Appendix 20 Page 2 of 2 INSTRUCTIONS

3. IF MCC-314A is energized, AND breaker SSD-EBKR-314A-2M, MCC-314A Safety to Non-Safety Tie is open, THEN perform the following:

a. Verify EDG A SEQUENCER has timed out.

b. Open ALL MCC-314A Non-Safety Load breakers.

c. Close SSD-EBKR-314A-2M, MCC-314A Safety to Non-Safety Tie.

d. Close BOTH of the following Supply breakers:

SP-EBKR-314A-4F, West Dry Cooling Tower Sump Pump A SP-EBKR-314A-5F, East Dry Cooling Tower Sump Pump A

4. IF MCC-314B is energized, AND breaker SSD-EBKR-314B-2M, MCC-314B Safety to Non-Safety Tie is open, THEN perform the following:

a. Verify EDG B SEQUENCER has timed out.

b. Open ALL MCC-314B Non-Safety Load breakers.

c. Close SSD-EBKR-314B-2M, MCC-314B Safety to Non-Safety Tie.

d. Close BOTH of the following Supply breakers:

SP-EBKR-314B-4F, West Dry Cooling Tower Sump Pump B SP-EBKR-314B-5F, East Dry Cooling Tower Sump Pump B

5. IF a PMP event is in progress, THEN align DCT Portable Sump Pump A(B) using OP-003-024, Sump Pump Operation.

End of Appendix 20

WATERFORD 3 SES OP-902-003 Revision 010 Page 12 of 34 LOSS OF OFFSITE POWER/LOSS OF FORCED CIRCULATION RECOVERY Restore Instrument Air

• 12. Verify Instrument Air is available:

a. Check BOTH of the following are a.1 Align Potable water to the operating: Instrument Air compressors using Appendix 18, "Aligning Potable TCW pump Water to Instrument Air Compressors."

CW pump

b. Check Instrument Air pressure is b.1 Dispatch an operator to start ALL greater than 95 psig. available Air compressors.

c. Check IA 909, CNTMT c.1 IF Instrument Air pressure is ISOLATION INSTRUMENT AIR greater than 95 psig, valve is open. THEN perform the following:

1) Open IA 909, CNTMT ISOLATION INSTRUMENT AIR valve.

2) IF Instrument Air pressure NOT restoring or maintaining greater than 95 psig, THEN close IA 909, CNTMT ISOLATION INSTRUMENT AIR valve.

Verify Containment Cooling

• 13. Verify Containment Cooling a maximum of three Containment Fan Coolers are operating in the normal mode:

Technical Guide for Loss of Offsite Power/ TG-OP-902-003 Loss of Forced Circulation Recovery Procedure Revision 306 Step Number 13 Verify Containment Cooling Objective The intent of this step is to verify that following a LOOP event the containment fan coolers have restarted and aligned in a normal configuration. The step was placed near the front of the procedure after steps to check condition of the plant and after the step to restore Instrument Air.

Instructions Direction is given to verify a maximum of three Containment Fan Coolers are operating in the normal mode. To prevent vibration alarms and damage to Containment Cooling unit duct work, configuration is limited to only three of four Containment Fan Coolers operating at one time when in fast speed. On restoration of power both CFCs per train will start in Normal or Fast Mode if there is no SIAS.

Contingency Actions None Justification for Deviations The step is not in the EPG. Following a LOOP without SIAS, all available CFCs will start in fast speed. There are four CFCs and if all four start then there is a concern for damaging duct work and causing vibration alarms.

References

1. OP-008-003, Containment Cooling System

2. ECE90-006 EDG Loading and Fuel Oil Consumption 27

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E1 LOSS OF 125 VOLT DC BUS A-DC PLACEKEEPER START DONE N/A

1. Verify Automatic Actions (Section C) take place as designed.

Continuous

2. Verify a Reactor Trip occurs, refer to OP-902-000, Standard Post Trip Actions, and perform concurrently with this procedure.

CAUTION IF CONTAINMENT SPRAY HEADER A ISOLATION, CS-125A, IS OPEN WHILE SHUTDOWN COOLING TRAIN A IS OPERATING, THEN CONTAINMENT SPRAY A RISER MAY FILL AND POSSIBLY SPRAY WATER INTO CONTAINMENT, DUE TO LEAKAGE PAST CONTAINMENT SPRAY PUMP A DISCHARGE STOP CHECK, CS-117A

3. If Shutdown Cooling (SDC) is in service, then restore SDC with LPSI Pump B in accordance with OP-009-005, Shutdown Cooling System.

4. If Shutdown Cooling Train A is in service and SDCS Loop 2 Suction Isolation Downstream Inside SI 405A closes, then locally Trip LPSI Pump A breaker SI-EBKR-3A-5 to prevent pump damage. Refer to OP-901-131, Shutdown Cooling Malfunction.

5. Locally Open the EG A Fuel Oil Transfer Pump breaker EGF-EBKR-312A-3F to prevent overflowing the Fuel Oil Day Tank.

19

SD-EDG Rev. 25 EC 16864 superseded EC 11723 for max frequency to assure a 7 day supply of oil.

This EC requires us to adjust emergency mode frequency if <59.77Hz or > 60.1 Hz.

PME will need to adjust a local setpoint potentiometer to obtain this range if necessary.

EC 24379 determined the current fuel oil margin to meet Technical Specification OPERABILITY. It determined if we operate at 60.1 Hz, we have 29 gallons of margin in Storage Tank A and 5 gallons of margin in Storage Tank B.

EC25302 installed six tie-ins to the existing EDG fuel oil piping at the existing A and B transfer pumps, EGF-MPMP-0001A and B, and at the gravity drain, 7EG1-44, from the EDG B fuel injector drain header to support future installation of additional equipment to increase diesel fuel oil storage capacity (under EC26937). See Fuel Oil Transfer Pump description below for more on EC25302.

FUEL OIL TRANSFER PUMP This radial, horizontal, centrifugal pump is designed to pump No. 2 fuel oil at a rated flow of 50 gpm at 150 psig discharge pressure at a design temperature of 125°F. It is driven by a 7.5-hp, 60-hertz, 3600-rpm electric motor powered from 480VAC MCC A312 (B). Its normal operating temperature band is 50-120°F. It required NPSH is 6 feet, and its available NPSH is 37.7 feet. It has a local ON-AUTO control switch. In ON, the pump runs continuously. In AUTO, it starts when the Fuel Oil Feed Tank level decreases to -30 in. (EGF-ILS-6907A (B)) and stops when level increases to -6 in.

(EGF-ILS-6908A (B)), where levels are referenced to the overflow line.

The pump's minimum recirculation requirement is 10 gpm via an orifice. Its UNID is EGF-MPMP-0001A (B).

Cross connect piping allows either FOTP to fill either Feed Tank from either Storage Tank. EC 16280 gives the minimum Storage Tank levels to support cross connect operation as: 38% for Tank A and 82% for Tank B.

EC25302 installed a transfer switch for the EDG-A Fuel Oil Transfer Pump Motor, EGF-EMTR-312A-3F, to allow the transfer of the electrical power source to a new future redundant pump.

The transfer Switch has three positions. It is lockable in all positions. The positions of the transfer switch are:

A1 for alignment to the existing motor (A Fuel Oil Transfer Pump)

OFF A2 for alignment to a future motor (under EC26937)

The transfer switch is not suitable for nor is it required to be a live transfer.

PAGE 69 OF 170

Component Cooling Water OP-901-510 System Malfunction Revision 303 C AUTOMATIC ACTIONS (contd)

NOTE (1) CCW Suct & Disch Header Tie Valves AB to A remain Open when CCW ASSIGNMENT Switch is aligned to A.

(2) CCW Suct & Disch Header Tie Valves AB to B remain Open when CCW ASSIGNMENT Switch is aligned to B.

4. At 16% dropping CCW A(B) Surge Tank level, the following valves Close, to split the A AND B CCW trains:

CC-126A/CC-114A CCW SUCT & DISCH HEADER TIE VALVES AB TO A CC-127A/CC115A CCW SUCT & DISCH HEADER TIE VALVES AB TO A CC-126B/CC-114B CCW SUCT & DISCH HEADER TIE VALVES AB TO B CC-127B/CC-115B CCW SUCT & DISCH HEADER TIE VALVES AB TO B CC-200A/CC-727 CCW SUCT & DISCH HEADER TIE VALVES A TO AB CC-200B/CC-563 CCW SUCT & DISCH HEADER TIE VALVES B TO AB

5. IF loss of CCW Pump(s) occurs due to loss of Vital Bus, THEN applicable pump will start 7 seconds after bus is energized by the Emergency Diesel Generator.

6. WHEN CCW temperature 92 F, THEN CCW Dry Cooling Tower Fans in Auto start sequentially in Slow speed at 60 second intervals.

7. WHEN ALL CCW Dry Cooling Tower Fans in AUTO are operating in Slow speed AND CCW temperature 92 F, THEN CCW Dry Cooling Tower Fans shift sequentially to Fast speed at 60 second intervals.

8. WHEN CCW temperature >100 F, then applicable Auxiliary Component Cooling Water Pump Starts AND ALL CCW Dry Cooling Tower Fans in Auto shift to Fast speed.

6

PAGE 98 OF 123 SD-CC-02

Off Normal Procedure OP-901-511 Instrument Air Malfunction Revision 015 E SUBSEQUENT OPERATOR ACTIONS E0 GENERAL PLACEKEEPER START DONE N/A

1. If Instrument Air pressure drops to 65 PSIG, then trip the Reactor and perform OP-902-000, Standard Post Trip Actions, concurrently with this procedure.

2. Dispatch an operator to the Air Compressors and verify the following:

All Instrument Air and Station Air Compressors running loaded with normal separator levels SA is making up to IA by performing the following, as applicable:

a) Verify SA Backup Supply for IA Press Cntl Valve (SA-125) opened automatically.

b) Raise the setpoint for SA Backup Supply for IA Press Cntl Valve (SA-125) to force it open.

c) Throttle open SA to IA Pressure Regulator Bypass (SA-127) to desired position.

If Instrument Air pressure is less than 95 PSIG, then Instrument Air Dryers Bypass Solenoid valve (IA-123) Opens

3. If all of the actions of step 2 have occurred and Instrument Air pressure is still dropping, then, using the Plant Paging System, announce the following two times:

"Attention Station Personnel, Attention Station Personnel. The plant is experiencing a loss of Instrument Air Pressure. Discontinue use of Instrument Air and Station Air. Report all air usage or any air leaks to the Control Room".

4. SI-129A(B) fails open on loss of Instrument Air. If Shutdown Cooling is in service or LPSI is replacing CS, then manually throttle or force closed SI-129A(B) in accordance with OP-009-008, Safety Injection System.

6

Off Normal Procedure OP-901-511 Instrument Air Malfunction Revision 015 E0 GENERAL (CONTD)

PLACEKEEPER START DONE N/A

5. If Instrument Air pressure can not be maintained above 80 PSIG, then consideration should be given to commence a Plant Shutdown in accordance with OP-010-005, Plant Shutdown.

6. Complete Attachment 4, Safety Related Valve Accumulator Checks.

7. If all Instrument and Station Air Compressors are running and air pressure is dropping, then perform the following:

7.1 Attempt to determine cause of low pressure condition.

7.2 If low pressure condition is due to line break or leakage, then after notifying the Control Room of location, isolate leak or cause of problem.

7.3 Restore as much of the air system as possible to normal pressure.

7.4 If Condensate Polisher backwash is in progress, then secure backwash of Condensate Polisher in accordance with OP-003-031, Condensate Polisher/Backwash Treatment.

7

247 GENERATOR GENERATOR 1 2 PLANT A Load AREA LOAD Sharing GRID PLANT B GENERATOR GENERATOR 1 2 257

Manual Voltage Regulation Initial Conditions: System stable. Generators operating in parallel, equal reactive and true power loads. No automatic voltage regulators.

GENERATOR GENERATOR 1 2 AREA LOAD GRID 258

POWER GRID Single Generator Parallel with a Grid POWER GRID 275

Off Normal Procedure OP-901-111 Reactor Coolant System Leak Revision 303 E SUBSEQUENT OPERATOR ACTIONS E0 GENERAL PLACEKEEPER START DONE N/A NOTE Where multiple indications for one parameter exist, more than one instrument should be used to obtain a particular reading.

1. If Shutdown Cooling Train is in service or Shutdown Cooling System Suction Relief valve (LTOP) is aligned, then go to OP-901-131, Shutdown Cooling Malfunction.

NOTE Pressurizer level indication may be inaccurate due to steam space leakage, reference leg flashing, or reference leg failure.

2. If lowering Pressurizer level cannot be terminated, then perform the following:

2.1 Manually trip Reactor.

2.2 Manually initiate Safety Injection Actuation (SIAS) and Containment Isolation Actuation (CIAS).

2.3 Go to OP-902-000, Standard Post Trip Actions.

7

Off Normal Procedure OP-901-103 Emergency Boration Revision 003 D. IMMEDIATE OPERATOR ACTIONS

1. If Charging is available, then perform the following:

1.1 Place Makeup Mode selector switch to MANUAL.

1.2 Align borated water source by performing one of the following (a. or b.):

a. Initiate Emergency Boration using Boric Acid Pump as follows:

Open Emergency Boration Valve, BAM-133.

Start one Boric Acid Pump.

Close recirc valve for Boric Acid Pump started:

BAM-126A Boric Acid Makeup Pump Recirc Valve A or BAM-126B Boric Acid Makeup Pump Recirc Valve B OR

b. Initiate Emergency Boration using Gravity Feed as follows:

Open the following Boric Acid Makeup Gravity Feed valves:

BAM-113A Boric Acid Makeup Gravity Feed Valve A BAM-113B Boric Acid Makeup Gravity Feed Valve B 1.3 Close VCT Disch Valve, CVC-183.

1.4 Verify at least one Charging Pump operating and Charging Header flow 40 GPM.

2. If Charging is not available and RCS pressure is <1400 psia, then use HPSI to inject into the RCS as follows:

2.1 Verify one HPSI Pump which is aligned for injection is running.

2.2 Throttle at least one of the following valves to establish 40 GPM flow:

SI-225A(B), HPSI Header A(B) to RC Loop 1A Flow Control SI-226A(B), HPSI Header A(B) to RC Loop 1B Flow Control SI-227A(B), HPSI Header A(B) to RC Loop 2A Flow Control SI-228A(B), HPSI Header A(B) to RC Loop 2B Flow Control 5

11.3 BORIC ACID MAKEUP SYSTEM STANDBY BREAKER LINEUP COMPONENT REQUIRED PERFORMED BY IV BY COMPONENT DESCRIPTION LOCATION NUMBER POSITION (INITIAL/DATE) (INITIAL/DATE)

BORIC ACID MAKEUP FLOW BAM-EBKR-014AB-8 RAB+21 11A&K ON BAMIFT0210-Y TRANSMITTER BORIC ACID BATCH TANK RAB+21 10A&J BAM-EBKR-1AB-17 ON TEMPERATURE CONTROLLER (SUPS AB)

BORIC ACID MAKEUP TANK A BAM-EBKR-311A-12B RAB+21 10A&H ON HEATERS 1-1 (BAMEHTR311A)

BORIC ACID MAKEUP TANK B BAM-EBKR-311A-12D RAB+21 10A&H ON HEATERS 1-2 (BAMEHTR311B)

EMERGENCY BORATION VALVE BAM-EBKR-311A-7C RAB+21 10A&H ON (BAM-133)

BORIC ACID MAKEUP TANK A BAM-EBKR-311B-12B RAB+21 10A&K ON HEATERS 1-2 (BAMEHTR311B)

BORIC ACID MAKEUP TANK B BAM-EBKR-311B-12D RAB+21 10A&K ON HEATERS 2-2 (BAMEHTR311B)

BORIC ACID MAKEUP TANK B BAM-EBKR-311B-7C RAB+21 10A&K ON GRAVITY FEED (BAM-113B)

BORIC ACID MAKEUP PUMP B BAM-EBKR-312A-2D RAB+21 11A&H ON (BAMMPMP0001-B)

BORIC ACID BATCH TANK BAM-EBKR-312AB-2M RAB+21 11A&J OFF HEATERS (BAMEHTR312AB)

BORIC ACID MAKEUP TANK A BAM-EBKR-312B-2J RAB+21 9A&K ON GRAVITY FEED (BAM-113A)

BORIC ACID MAKEUP PUMP A BAM-EBKR-313A-3D RAB+21 10A&G ON (BAMMPMP0001-A)

BORIC ACID BATCH TANK BAM-EBKR-64AB-9 RAB+21 11A&J ON MIXER (BAMEMTR64AB)

BORIC ACID MAKEUP PUMPS BAM-EBKR-90A-10 RAB+35 11A&G ON A & B CONTROL POWER BORIC ACID MAKEUP PUMP A BAM-EBKR-90A-5 RAB+35 11A&G ON RECIRC VALVE (BAM-126A)

OP-002-005 Revision 056 Attachment 11.3 (1 of 2) 190

BORIC ACID MAKEUP SYSTEM STANDBY BREAKER LINEUP (CONTD)

COMPONENT REQUIRED PERFORMED BY IV BY COMPONENT DESCRIPTION LOCATION NUMBER POSITION (INITIAL/DATE) (INITIAL/DATE)

BORIC ACID MAKEUP PUMP B BAM-EBKR-90A-7 RAB+35 11A&G ON RECIRC VALVE (BAM-126B)

BORIC ACID MAKEUP HEADER BAM-EBKR-96AB-13 RAB+35 11A&H ON ACID FCV (BAM-141)

DIRECT BORATION VALVE BAM-EBKR-96AB-21 RAB+35 11A&H ON (BAM-143)

OP-002-005 Revision 056 Attachment 11.3 (2 of 2) 191

SD-PLC Level Channel 110X Fails Low, Level Channel 110X Controlling Symptoms:

Pressurizer Lo-Lo and Hi-Low level alarms received.

Pressurizer level channel X indication will be 0%.

Pressurizer heaters cut out.

Pressurizer level control M/A station will indicate low process level and controller output will be 0%.

Both backup charging pumps running.

Letdown heat exchanger outlet pressure low alarm.

Letdown flow decreases to 30 GPM.

Pressurizer level recorder indicates 0% level with setpoint remaining constant.

Increase in Pressurizer level channel Y indication.

Actions:

Actions will be the same as a high channel failure, with the additional action of selecting the unaffected channel on the heater cutout selector switch, and re-energizing the heaters.

EMERGENCY OPERATION There are no emergency operating procedures applicable to Pressurizer Level Control.

REVISION 10 PAGE 25 of 69

A-13 E0703 CWD 1690 4.63 STM BYPASS SYSTEM CNDSR VACUUM FAIL (A-13) REV 014 INITIATING DEVICE SETPOINT CD-IPS-1904A(B, C) 3.4 in. Hg. (R: 6.4 in. Hg.)

POSSIBLE EFFECTS

1. SBCS is non-functional. No steam bypass capacity is available.

CONTROL ROOM INDICATIONS LOCAL INDICATIONS Cndsr Interlock control switch green OFF NONE light illuminated CD-IPI-1902B2, Condenser Vac WR SBCS TEST drawer (CP-5)

NOTE When condenser vacuum is restored, Condenser Interlock 1 and 2 must be manually Reset with reset switch on CP-1 or CP-5.

POSSIBLE CAUSES RECOMMENDED ACTIONS

1. Loss of Condenser Vacuum. 1.1 Go to OP-901-220, Loss of Condenser Vacuum.

OP-500-005 Revision 014 Attachment 4.63 (1 of 1)

WATERFORD 3 SES OP-902-003 Revision 010 Page 10 of 34 LOSS OF OFFSITE POWER/LOSS OF FORCED CIRCULATION RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Check RCP Seal Cooling

• 8. Check CCW flow to RCPs. 8.1 IF NO CCW flow to ANY RCP(s),

THEN perform the following:

a. IF CCW lost for greater than 10 minutes, THEN do not restore CCW flow to affected RCP(s).

b. IF CCW lost for less than 10 minutes, THEN attempt to restore CCW to ANY affected RCP(s)

Protect Main Condenser

• 9. Perform BOTH of the following: 9.1 Perform the following:

Verify CW System in operation. a. Verify BOTH MSIVs are closed:

REFER TO OP-003-006, Circulating Water. MS 124A, MSIV 1 Check Condenser vacuum greater MS 124B, MSIV 2 than 14 Hg.

b. Verify ALL Steam Generator Blowdown Isolation valves are closed:

BD 102A, STM GEN 1 (IN)

BD 102B, STM GEN 2 (IN)

BD 103A, STM GEN 1 (OUT)

BD 103B, STM GEN 2 (OUT)

Technical Guide for Loss of Offsite Power/ TG-OP-902-003 Loss of Forced Circulation Recovery Procedure Revision 306 Step Number 9 Protect Main Condenser Objective The intent of this step is to prevent over-pressurizing the Main Condenser in the event of a loss of condenser vacuum. The most likely cause would be a total or partial loss of Circulating Water due to a loss of electrical power to the CW pumps. However, this guidance also applies to any event that could lead to a loss of condenser vacuum. This step also ensures that the operator has control of RCS cooldown rate and SG inventory.

Instructions The Operator verifies that the Main Condenser has sufficient cooling provided by circulating Water and Condenser Vacuum is being maintained. If cooling to the condenser is not available or the condenser vacuum is not maintained then implement the contingency and isolate the condenser. . The Circulation Water system procedure is reference to provide instructions to restore CW system operation if needed.

In the event that a single train of off-site power is lost the circulating water pumps (SWGR 1A or SWGR 1B power) on that train will be deenergized. Also the discharge valves (MCC 221A or MCC 221B power) associated with those pumps will be deenergized and open. If the CW Pumps were in a 4 of 4 operating configuration, the open discharge valves will result in some of the circulating water flow from the two operating CW pumps to recirculate back into the bay. If the CW Pumps are in a 3 of 4 operating configuration and the train of off-site power that is lost was carrying two operating CW pumps, then the last operating CW pump, on the train of off-site power that is energized, will also trip.

The following OE was obtained from the Reactor Trip (CR-WF3-2015-03565) June 03, 2015: The reactor tripped due to a feedwater casualty. The A CWP was tagged out and the other 3 CW Pumps were operating. The B side power did not transfer to the SUT and both the 1B and 2B buses de-energized. Since the 1B bus de-energized all 3 CWPs tripped. If 3 CWPs are operating and 2 trip the third pump trips. For CWP B & D this also meant that the discharge valve motors were de-energized and the discharge valves did not close setting up the CW bay to short cycle. Since CWP C discharge valve was energized and the valve did go closed on the pump trip. Condenser vacuum slowly lowered until CWP C was restarted and then condenser vacuum held steady. The discharge valve only got to 45% open and jammed; this is believed to be due to the DP set up by the short cycling recirculation back to the bay. The valve then had to be manually opened locally; several personnel were required to assist. Once the discharge valve was fully opened vacuum restored to 25 to 26 inches. While only one CWP was in operation the amps on the pump were above normal until B train power was restored and other CW pumps were brought back in service.

19

Technical Guide for Loss of Offsite Power/ TG-OP-902-003 Loss of Forced Circulation Recovery Procedure Revision 306 Step Number 9 Protect the Main Condenser (contd)

Contingency Actions The operator closes the MSIVs to isolate the major main steam supply to the condenser.

The operator also closes the Blowdown isolation valves to the condenser.

Justification for Deviations WF3 expands on the EPG instruction to verify circulating water flow by specifying that the circulating water system is in operation and that condenser vacuum is maintaining above a minimum value. A minimum condenser vacuum value is specified for the case where the CW system is operating and the condenser vacuum is degrading due to other reasons (such as a rupture or loss of gland seal steam).

References

1. OP-003-006 Circulating Water

2. CR-WF3-2015-05365 Reactor Trip due to loss of feed heaters

3. ECS98-001 P.36

4. OP-901-220 Loss of Condenser Vacuum 20

Off Normal Procedure OP-901-202 Steam Generator Tube Leakage or High Activity Revision 015 E0 GENERAL (CONTD)

PLACEKEEPER START DONE N/A Continuous 21.5 Maintain pressurizer pressure within all of the following criteria:

Within the limits of Technical Specification 3.4.8.1, Figure 3.4-3.

Less than 950 PSIA.

Within 50 PSID of the most affected steam generator pressure NOTE The Blowdown Radiation Monitor is normally aligned to both Steam Generators. The monitor may be aligned locally to sample either Steam Generator.

22. Determine the Steam Generator to be isolated by considering all of the following:

Main Steam Line 1 and 2 Radiation Monitor indications Steam Generator Blowdown Radiation Monitor indication Activity in each Steam Generator as indicated by Chemistry samples 18

Technical Guide for Steam Generator Tube TG-OP-902-007 Rupture Recovery Procedure Revision 308 Step Number 12 Depressurize the RCS Objective The intent of this step is to establish control of RCS pressure. The general goals associated with RCS pressure control are:

Providing subcooling to support the core heat removal process Avoiding overpressure situations for PTS and RT NDT considerations Minimizing the pressure differential between the steam generator and the RCS to minimize the leakage Deliberately creating a primary to secondary differential pressure to establish backflow to control SG level rise or reduce SG pressure/temperature Controlling RCS pressure below the atmospheric dump valve or main steam safety valve lift pressure to prevent uncontrolled release of radioactivity to the environment This RCS depressurization step establishes the depressurization goals of the EOP for the Operator. All goals will not be met instantaneously at 520 degrees F Thot. Goals can be achieved in coordination with a plant cooldown. The Operator should reduce pressure as much as possible while maintaining subcooled margin and operating within the PT curves. If RCPs are operating RCP NPSH pressure must be maintained to continue operating the RCPs. Operating RCPs allows for a greater rate of cooldown and will allow the Operator to arrive at SDC temperature and pressure limits sooner than if RCPs were not operating. Therefore maintaining RCPs is preferable to natural circulation cooldown. Pressure reduction should continue along with cooldown while maintaining RCP NPSH and PT curves as applicable.

Instructions Maintaining the RCS pressure below the lift setpoint of the ADV, within the PT limits and approximately equal to the isolated steam generator pressure [+/-50 psi] will minimize the loss of primary fluid to the secondary side and the possibility of overfilling the isolated SG. This action will minimize the potential for release of radiation to the environment by minimizing RCS to steam generator leakage.

33

E-10 G0510 CWD 690 4.50 W ASTE GAS DISCH RAD HIGH DRYER/MON TRBL (E-10) REV 16 INITIATING DEVICE SETPOINT PRMIRE0648 K1 Relay (Fail Alarm) De-energized PRMIRE0648 K2 Relay (High Alarm) Variable GWMIPS0648, GWM Rad Monitor Dryer 6 PSIG (R: 3 PSIG)

Outlet Pressure POSSIBLE EFFECTS

1. Termination of Gas Decay Tank Discharge (GWM-309 closes).

CONTROL ROOM INDICATIONS LOCAL INDICATIONS RM-11, Grid EFL, GWM RAB -4, Gaseous PRMIRE0648, GWM Rad Monitor local Waste System Noble Gas Monitor indication (PRMIRE0648)

GWMIFRR0648, Waste Gas Flow and GWMIFI6712, Waste Gas Decay Tanks to Rad Recorder Plant Vent Flow Indicator, at LCP-42A Waste Gas Discharge Flow Control Valve, Waste Gas Discharge Flow Control Valve, GWM-309, position. GWM-309, local indication PMC Point ID A41300, Gas to Stack Flow Power Supplies:

PRMEBKR014AB 10, Waste Gas Discharge Rad Detector LWMEBKR45AB 6, Waste Mgmnt Disch Flow and Rad Recorder POSSIBLE CAUSES RECOMMENDED ACTIONS

1. HI activity 1.1 Refer to OP-901-413, Waste Gas Discharge High Radiation.

1.2 Verify discharge auto isolates.

1.3 Direct Chemistry to sample GDT to determine validity of activity levels.

1.4 Verify correct setpoint in PRMIRE0648, Waste Gas Rad Monitor.

OP-500-007 Revision 017 83 Attachment 4.50 (1 of 2)

Off Normal Procedure OP-901-413 Waste Gas Discharge High Radiation Revision 002 C AUTOMATIC ACTIONS

1. Waste Gas Discharge Flow Control GWM-309 closes to terminate gaseous waste release.

4

SD-GWM GWM-311, GWM Waste Gas Flow Ind. Trans. Isolation Valve, is a 1" needle valve that provides throttling capability for waste gas release within the limits of the Gas Release Permit. GWM-3094, Waste Gas to Waste Gas Collection Header Pressure Valve, provides a set backpressure in the GDT common discharge header to maintain a constant differential pressure across GWM-311 to ensure a constant release flow rate to the VGCH.

LCP-42A provides the following indications for the GDTs (refer to Figure 2):

• Temperature (GWM-ITI-6709, 6710, 6711), with ranges of 40 - 250 °F

• Pressure (GWM-IPI-6709, 6710, 6711), with ranges of 0 - 450 psig

• GDT A, (B), (C) high pressure alarm (340 psig).

• Waste Gas Release Rate (GWM-IFI-6712), with a range of 0 - 50 scfm.

CP-4 provides for indication of the waste gas release rate (0 - 50 scfm) and the waste gas release activity (10 102 mci/cc) on recorder GWM-IFRR-0648. A totalizer is also provided to indicate the total standard cubic feet of waste gas released.

CONTAINMENT ISOLATION VALVES GWM-104 and GWM-105, Inside and Outside Containment Isolation Valves, are operated by individual CLOSE-OPEN control switches located on CP-4. They auto isolate on a CIAS. Once the isolation signal clears and the CIAS is reset, the valve control switches must be taken to CLOSE to reset the operating logic "94" relays. This allows the valves to be reopened. There is no CIAS override associated with these valves. GWM-104 and GWM-105 are powered from PDP 95-SB and PDP 94-SA respectively. These valves are normally closed.

Revision 6 13 of 39

Off Normal Procedure OP-901-502 Evacuation of Control Room and Subsequent Plant Shutdown Revision 033 D. IMMEDIATE OPERATOR ACTIONS (CONTD)

PLACEKEEPER START DONE N/A

2. BOP Operator: Perform the following:

2.1 Verify Turbine trip:

Governor valves Closed Throttle valves Closed 2.2 Verify Generator trip:

Exciter Field Breaker Tripped Generator Breaker A Tripped Generator Breaker B Tripped 2.3 Reset Moisture Separator Reheater controls.

2.4 If evacuating the Control Room due to fire, then perform the following:

2.4.1 If either of the following valves has spuriously Opened, then place the applicable controller(s) in MANUAL and lower the output to zero:

MS-116A SG 1 Atmospheric Dump MS-116B SG 2 Atmospheric Dump 2.4.2 Close the following valves:

MS-124A Main Steam Isol Valve #1 MS-124B Main Steam Isol Valve #2 2.5 Obtain assigned Operations Security Key Ring and proceed to RAB +35 Relay Room.

6

WATERFORD 3 SES OP-902-002 Revision 020 Page 46 of 83 LOSS OF COOLANT ACCIDENT RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Hot and Cold Leg Injection

• 53. IF elapsed time from the start of the event is between 2 and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, AND ANY of the following conditions exist:

RCS subcooling is less than 28°F based on representative CET temperature PZR level is less than 7% [23%]

Reactor vessel level indicates at least ONE of the following:

QSPDS REACTOR VESSEL LEVEL 5 is voided VESSEL LEVEL PLENUM less than 80%

THEN establish simultaneous Hot and Cold Leg injection using Appendix 15, "Hot and Cold Leg Injection."

SD-QSP Figure 6 shows the arrangement of an individual HJTC sensor with a HJTC-UHJTC pair. The two thermocouples in a pair are separated by several inches distance and by a splash shield, which surrounds the heated thermocouple. The splash shield protects the HJTC from spurious cooling by water running down the sensor sheath or by entrained water droplets.

The figure also shows that there are two separator tubes located in the region above the fuel alignment plate. Each tube contains a probe, and each probe has eight HJTC sensors at discrete locations. The probe arrangement is called a "split probe" configuration because the separator tube is divided into independent upper and lower portions by a divider disk. Each portion has eight 13/64-inch diameter holes around both the top and bottom for drainage and pressure equalization. Each portion can have its own collapsed water level, independent of the other portion's water level. The probes are located near the core periphery and away from the hot legs.

Each probe assembly is housed in a stainless steel guide tube (separator tube) which protects it from hydraulic loads and serves as a guide path for the probe. A third tube located between the upper guide structure support plate and the fuel alignment plate provides additional support and attaches the entire assembly to a control element assembly (CEA) shroud. The guide and support tubes are perforated along their entire length with 3/8-inch holes. Additionally, slots in these tubes are positioned relative to the holes in the separator tube to prevent steam bubbles from entering the probe at the bottom and entrained water droplets from entering the probe at the top.

A sensor heater power control system is used to protect the HJTC from damage due to overheating. When an increasing HJTC temperature or sensor T exceeds a predetermined setpoint, the heater power is gradually reduced until an acceptable stable temperature is reached. The power always remains high enough, however, so that all sensors are still capable of providing an "uncovered" signal. Figure 7 shows the heater power control logic.

The signals from each HJTC and heater circuits are carried from the reactor vessel to the QSPDS panels where the temperature and T signals are processed and analyzed for display and control purposes. QSPDS determines a thermocouple pair to be uncovered if either the T is greater than 200°F or the UHJTC temperature is greater than 700°F. Refer to Figure 8.

The percent level in each portion of the separator tube (upper and lower) is determined by the number of covered HJTCs versus the number of uncovered HJTCs. There are two two-pen Vessel Level recorders on CP-7. The recorder's input channels are RC-IUR-2103-AS1 (AS2, BS1, BS2).

REVISION 5 Page 13 of 57

WATERFORD 3 SES OP-902-004 Revision 016 Page 16 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Isolate Most Affected SG with ESD

• 16. Isolate the MOST AFFECTED SG as follows:

Steam Generator 1

a. Verify MS 124A, MSIV 1 is closed.

b. Verify FW 184A, MFIV 1 is closed.

c. Verify MS 116A, ADV 1 is closed and the controller in manual.

d. Verify EFW Isolation valves are closed:

EFW 228A, SG 1 PRIMARY EFW 229A, SG 1 BACKUP

e. Place EFW FLOW CONTROL valves in MAN and THEN close:

EFW 224A, SG 1 PRIMARY EFW 223A, SG 1 BACKUP

f. Close MS 401A, PUMP AB TURB STM SUPPLY SG 1.

g. Close Main Steam Line Drains:

MS 120A, NORMAL MS 119A, BYPASS (continue)

WATERFORD 3 SES OP-902-004 Revision 016 Page 39 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

---

NOTE ----------------------------------------------------------

The following RCS cooldown rates apply:

RCS 100°F/hr RCS on Natural Circulation with Asymmetric Steam Generator: RCS 50°F/h Pressurizer 200°F/hr Attachment 3-A, "Pressurizer/RCS Cooldown Log may be required during the cooldown and depressurization.

Perform Controlled Cooldown

49. Cooldown the RCS to less than 350°F 48.1 IF necessary, TH or CET temperature using the Intact THEN locally control intact SG ADV SG ADV. using Appendix 22 Local Operation of Atmospheric Dump Valves.

Technical Guide for Excess Steam TG-OP-902-004 Demand Recovery Procedure Revision 307 Step Number 49 Perform Controlled Cooldown Note Objective The intent of this Note is to ensure the operators are made aware of the cooldown rates prior to starting the RCS cooldown. The Note also includes information on forms that may be required during the cooldown.

Justification for Deviations The EPG does not include this Note. This Note is included to inform the operators of additional information which is not required to maintain the plant, but information that may add value to maintenance of the overall plant.

An Asymmetric Steam Generator Cooldown is when one SG is used for cooldown and the other steam generator is not being steamed by either the SBCS or the ADV.

References

1. ECS98-001 R.01

2. ECS98-001 R.02

3. ECS98-001 R.03 97

Technical Guide for Excess Steam TG-OP-902-004 Demand Recovery Procedure Revision 307 Step Number 49 Perform Controlled Cooldown Objective The intent of this step is to initiate a controlled cooldown of the plant. This action will reduce the pressure and temperature of the plant. In the case of a LOCA, SGTR, or ESD this limits the associated break flow by reducing the stored energy in the plant.

A plant cooldown will change the reactivity conditions in the reactor core. Therefore, whenever a cooldown is performed the operators must consider the effects of the cooldown on reactivity control and take appropriate actions to maintain the shutdown margin. This may include borating and sampling the RCS, both prior to and during the cooldown. This step assumes that emergency boration is in progress due to an SIAS and therefore no direction is given to commence boration.

Instructions The operator has initiated MSIS early in the procedure, therefore steam bypass in not available. The atmospheric dump valves (ADVs) are used to cooldown the plant.

Contingency Actions A contingency was added to direct use of local operation of ADVs by using a standard appendix.

Justification for Deviations The EPG list the steam bypass system as the preferred method to cooldown the plant and list the ADVs as the alternate method. The MSIVs are closed early in this procedure which isolates the steam bypass system; therefore steam bypass is not a viable option.

References None 98

Operating Instruction OI-038-000 Emergency Operating Procedures Revision 013 Operations Expectations / Guidance 5.4.65 Place CS Pumps in OFF None 5.4.66 Place Hydrogen Analyzers in Service If a CIAS has occurred the operators must override the containment isolation valves for the Hydrogen Analyzers.

5.4.67 Protect the Main Condenser If electrical power is not available, then the control switches for the isolation valves should be placed in the closed position in preparation for the return of electrical power.

5.4.68 RAS Initiation Criteria Do not prematurely initiate RAS. Manual action should not be taken unless an automatic RAS did not function when required and an adequate containment sump level exists.

During recirculation, at least one HPSI pump should be operating at all times unless HPSI throttle criteria is met.

5.4.69 RCP Restart Criteria If Component Cooling Water is lost to RCP seals for 10 minutes, restoring Component Cooling Water to the RCPs may result in seal failure.

During a SGTR, consideration of the possibility of an un-borated slug of water in the affected RCS loop should be taken into account when considering the desirability of restarting RCPs.

5.4.70 RCP Start When restarting RCPs, pressurizer level and pressure may drop due to coolant shrinkage and void collapse.

If restarting Reactor Coolant Pumps, consideration should be given to choosing pump combinations that will maximize pressurizer spray flow.

5.4.71 RCP Trip Strategy Tripping RCP 1A and 2A is the preferred combination of pumps to secure due to the higher head of RCP 2A.

29

Off Normal Procedure OP-901-113 Volume Control Tank Makeup Control Malfunction Revision 302 E SUBSEQUENT OPERATOR ACTIONS E0 GENERAL NOTE Failure low of VCT level instrument CVC-ILT-0227(PID A39401) will cause RWSP to Charging Pumps (CVC 507) to open and VCT Disch Valve (CVC 183) to close. Failure of VCT level instrument CVC-ILT-0226 (PID A39400) affects CP-4 level indication and auto makeup to the VCT.

1. If a VCT level instrument fails, then perform the following:

1.1 If level instrument CVC-ILT-0227 fails low causing Charging Pump suction source to swap to RWSP, then perform the following:

1.1.1 Simultaneously secure all Charging Pumps and Close Letdown Stop Valve (CVC 101).

1.1.2 Operate Charging Pumps as necessary to maintain Pressurizer Level above minimum level for operation in accordance with Attachment 1, Pressurizer Level Versus Tave Curve.

1.1.3 Match Tavg and Tref by adjusting Turbine Load.

1.1.4 Initiate corrective action to repair level instrument.

1.1.5 When level instrument CVC-ILT-0227 is repaired, then restore Charging and Letdown in accordance with OP-002-005, Chemical and Volume Control.

1.2 If level instrument CVC-ILT-0227 fails high causing the VCT Inlet/Bypass Valve (CVC 169) to divert to BMS, then perform the following:

1.2.1 Align VCT Inlet/Bypass Valve (CVC 169) to VCT.

1.2.2 Make up to VCT as required to restore level in accordance with OP-002-005, Chemical and Volume Control.

1.2.3 Initiate corrective action to repair level instrument.

1.2.4 When level instrument CVC-ILT-0227 is repaired, then restore VCT Inlet/Bypass Valve (CVC 169) to AUTO.

6

SD-SDC REFUELING WATER LEVEL INDICATING SYSTEM (RWLIS) is designed to perform the following functions:

Monitor the water level in the Reactor Coolant System and the refueling pool during refueling operations.

Monitor the water level in the RCS hot leg during maintenance evolutions requiring the RCS water levels at elevations within the range of the hot leg.

Provide indication of the level locally in the containment building and remotely in the control room.

Provide an alarm in the control room when the water level in the refueling pool exceeds a predetermined level or when the water level drops below a predetermined level in the range of the hot leg.

The Reactor Coolant Shutdown Level Measurement System (RCSLMS) is designed to perform the following functions:

Monitor the water level in the Reactor Coolant System (RCS) during non-power operation.

Monitor the water level in the RCS hot leg during maintenance evolutions requiring the RCS water levels at elevations within the range of the hot leg.

Provide indication of the level locally in the containment building and remotely in the control room.

Provide an alarm in the control room when water level drops below a predetermined level in the range of the hot leg.

SYSTEM DESCRIPTION The SDC System consists of two independent and duplicate trains to achieve the required redundancy. Each shutdown cooling loop contains a LPSI pump, a vacuum priming pump, a SDCHX, two safety injection cold leg nozzles, flow control valves, and the controls and instrumentation necessary to provide for proper system operation.

One loop (train) operating alone is capable of bringing the RCS to refueling temperature although the cooldown, with one loop, would be considerably longer. With the most limiting single active failure in the SDC System, RCS temperature can be brought to 200°F in approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following shutdown and to the refueling temperature of 140°F in approximately 193 hours0.00223 days <br />0.0536 hours <br />3.191138e-4 weeks <br />7.34365e-5 months <br /> using only one Low Pressure Safety Injection (LPSI)

Pump and one SDCHX.

During shutdown cooling, reactor coolant is circulated by the LPSI Pump(s) through the SDCHX(s) to the LPSI headers and is returned to the RCS cold legs through the safety injection nozzles. The cooldown rate is manually controlled by adjusting the flow rate Revision 8 Page 5 of 52

SD-SDC through the heat exchangers with SDCHX A(B) Temperature Control Valve, SI-415A(B),

on the discharge of each heat exchanger.

The desired total shutdown cooling flow rate is automatically controlled at setpoint by adjustment of the amount of coolant which bypasses the shutdown cooling heat exchangers through LPSI Pump A(B) Discharge Flow Control Valve, SI-129A(B). The required minimum SDC System flow rates for time after shutdown can be found in the System Operations section of this system description.

A warm-up recirculation line is provided in the SDC System to limit thermal stress in the piping and components that would occur if a step change in temperature from ambient to reactor coolant temperature were permitted during system lineup.

Positive isolation from the RCS is provided whenever the RCS is above the shutdown cooling initiation pressure of 392 PSIA (Pressurizer). Isolation valves with appropriate interlocks are provided on the SDCS suction lines for this purpose.

The SDCS is protected from overpressure of system components, where design pressure and temperature are less than the RCS design limits, by overpressure protection devices. Each SDC suction line is equipped with three isolation valves in a series arrangement. Valves SI-401A(B) and SI-405A(B) are located inside containment, while valves SI-407A(B) are located outside containment. With this arrangement, a redundant parallel SDC path is available should a single failure preclude the availability of one of the SDC trains. Each valve inside containment is provided with an interlock to prevent opening until RCS pressure is less than 386 PSIA. In addition, during normal operation, the circuit breakers for motor operated valves SI-401A and SI-401B are locked in the open position to prevent inadvertent opening.

To assure availability of the SDC System and positive isolation of the RCS pressure boundary, RCS isolation valves SI-405A and SI-405B are provided with pneumatic operators. These operators are connected to Instrument Air and have back-up accumulators in containment to ensure air is available to operate the valve. These valves must have at least 55 psig of air to operate, or they will fail AS-IS.

Each SDC train is powered from their respective buses to assure that power failure in one train will not jeopardize availability of the other train. The containment isolation valves (SI-407A and SI-407B) are motor operated and are powered from their respective buses.

A pressure relief in each SDC suction line protects the RCS from overpressurization during system operation when the suction valves inside containment are open. The valves are sized considering transients due to inadvertent SIAS or an RCP start with secondary temperature greater than primary temperature. A LOOP 1(2) SDC RELIEF VLV ACTIVE annunciator will alarm on CP-8 in the Control Room if RCS pressure is greater than 392 PSIA with SI-401A(B) or SI-405A(B) off the closed seat.

Revision 8 Page 6 of 52

SD-SDC Shutdown Cooling Temperature Control Valves (SI-415A(B))

SI-415A and B are 10 motor operated butterfly valves used to set the desired flow through the SDCHX for temperature control. The valves are controlled from CP-8 and LCP-43. The valves have a two position (MORE/LESS) keylock switch on CP-8. The key is removable only when the switch is in the LESS position. The valves have a MORE/LESS control switch at LCP-43. The seating ring for SI-415A(B) has been removed and the valves pass 800 GPM when the valve is fully closed. The valves work in conjunction with SI-129A(B) to maintain desired SDC System total flow and RCS temperature.

Shutdown Cooling Flow Control Valves (SI-129A(B))

The Shutdown Cooling Flow Control valves are 10 butterfly valves pneumatic piston operated (Fail Open) with a three-way solenoid used to control total SDC flow during normal plant shutdown. The valves are controlled by handswitches on CP-8 and LCP-

43. The key switch on CP-8 has positions of OPEN/AUTO with the key removable only in the OPEN position. The handswitch has positions of OPEN/AUTO on LCP-43.

Automatic operation of the valves is controlled by an electro-pneumatic converter, SI-IE/P-129A(B), that receives a flow signal from SI-IFIC-0307(0306) via PAC. The CP-8 flow controllers are used regardless of the location of operation, since there are no flow controllers on LCP-43. The valves have status lights on CP-8 and LCP-43. An annunciator is located on CP-8 to indicate LPSI PUMP A(B) FLOW LOST.

Shutdown Line Warm-up Valves (SI-135A(B))

The Shutdown Line Warm-up valves are 8 motor operated gate valves used to warm up the system to within 100°F of RCS temperature prior to placing SDC in service to the RCS loops. The warm-up valve control switches have positions of MORE/LESS and are spring return to normal. The valves are controlled from CP-8 or LCP-43.

SI - Flow Control Valves (SI-138A(B) and SI-139A(B))

SI-138A(B) and SI-139A(B) are 6 motor-operated globe valves used for throttling LPSI Pump flow during SDC operations below 18 ft MSL, for filling the Refueling Cavity, and also can be used for containment isolation. These valves are flow balanced to ensure equal flow in each line. The valves can be operated by handswitches on CP-8 and LCP-43. On CP-8, the handswitch has positions of MORE/LESS with spring return to neutral. These valves open automatically upon receiving an SIAS and will bypass the overload relay. An SIAS open signal can be overridden by placing the handswitch in the MORE position and then back to the LESS position. Position indication is provided on CP-8 and LCP-43. Along with the OPEN/CLOSE indication on CP-8 is a 0 - 100%

position indication. The valves are supplied from 480 VAC MCC 311A(B).

Revision 8 Page 13 of 52

Off Normal Procedure OP-901-131 Shutdown Cooling Malfunction Revision 304 E1. SYSTEM LEAKAGE PLACEKEEPER START DONE

1. IF ANY of the following LPSI Pump cavitation/Air binding indications occur, THEN stop affected LPSI Pump:

Dropping OR erratic ammeter indication Dropping OR erratic Shutdown Cooling flow Steady low flow AND amperage less than expected for system configuration Local observation.

CAUTION (1) IF RCS IS OPEN FOR MAINTENANCE, THEN FILLING MAY RESULT IN DISCHARGE FROM OPENING. THIS COULD ENDANGER PERSONNEL IN OR AROUND OPENINGS. CLOSED RCS IMPLIES NO MAJOR OPENINGS THAT WOULD PREVENT FILLING RCS HOT LEG. RCS IS CONSIDERED CLOSED WITH RCP SEAL REMOVED.

(2) IF STEAM VOIDS ARE PRESENT IN THE RCS OR SDC PIPING, THEN MAKEUP SHOULD BE PERFORMED SLOWLY TO MINIMIZE WATER HAMMER CONCERNS.

2. IF RCS makeup is required, THEN using Shutdown Cooling Train A or B perform the following:

2.1 For Shutdown Cooling Train A:

a. Start Train B HPSI Pump.

b. Close HPSI HEADER ORIFICE BYPASS VALVE (SI 219B).

c. Open HOT LEG 2 INJECTION ISOLATION VALVE (SI 502B).

10

Off Normal Procedure OP-901-131 Shutdown Cooling Malfunction Revision 304 E1 SYSTEM LEAKAGE (CONTD)

PLACEKEEPER START DONE

d. Throttle the following valves as necessary to restore inventory:

HPSI COLD LEG INJECTION 1A (SI 225B)

HPSI COLD LEG INJECTION 1B (SI 226B)

HPSI COLD LEG INJECTION 2A (SI 227B)

HPSI COLD LEG INJECTION 2B (SI 228B)

HOT LEG 2 INJECTION FLOW CONTROL (SI 506B).

OR 2.2 For Shutdown Cooling Train B:

a. Start Train A HPSI Pump.

b. Close HPSI HEADER ORIFICE BYPASS (SI 219A).

c. Open HOT LEG 1 INJECTION ISOLATION (SI 502A).

d. Throttle the following valves as necessary to restore inventory:

HPSI COLD LEG INJECTION 1A (SI 225A)

HPSI COLD LEG INJECTION 1B (SI 226A)

HPSI COLD LEG INJECTION 2A (SI 227A)

HPSI COLD LEG INJECTION 2B (SI 228A)

HOT LEG 1 INJECTION FLOW CONTROL (SI 506A)

3. Restore AND maintain RCS level >15.13 feet, top of RCS Hot Leg.

4. Monitor RCS Hot Leg for saturation conditions AND determine RCS heatup rate using EITHER:

CETs OR IF CETs NOT available, THEN refer to Attachment 2:

Calculated RCS Time to Boil.

11

Off Normal Procedure OP-901-131 Shutdown Cooling Malfunction Revision 304 E1 SYSTEM LEAKAGE (CONTD)

PLACEKEEPER START DONE

5. IF NO LPSI Pump is operating, THEN perform the following:

a. Vent suction piping of LPSI Pump that will take suction on Hot Leg with operating HPSI Pump.

b. Place Shutdown Cooling Train in service in accordance with OP-009-005, SHUTDOWN COOLING SYSTEM.

c. Continue venting until all air is removed.

6. IF source of system leakage is known AND isolated, THEN go to step 17.

7. IF additional RCS makeup is required, THEN start second available HPSI Pump in accordance with step 2.

7.1 IF additional RCS makeup is required, THEN start third available HPSI Pump.

8. Available Charging Pumps may be started to provide additional RCS makeup.

CAUTION IF RCS IS OPEN FOR MAINTENANCE, THEN FILLING MAY RESULT IN DISCHARGE FROM OPENING. THIS COULD ENDANGER PERSONNEL IN OR AROUND OPENINGS.

CLOSED RCS IMPLIES NO MAJOR OPENINGS THAT WOULD PREVENT FILLING RCS HOT LEG. RCS IS CONSIDERED CLOSED WITH RCP SEAL REMOVED.

9. IF RCS level is NOT being restored by available HPSI AND Charging Pumps, THEN align LPSI Pump A or B for injection as follows:

9.1 For Train A:

a. Verify LPSI Pump A secured.

b. Locally open LPSI PUMP A SUCTION ISOLATION VALVE FROM RWSP (SI 109A).

12

Off Normal Procedure OP-901-131 Shutdown Cooling Malfunction Revision 304 E1. SYSTEM LEAKAGE (CONTD)

PLACEKEEPER START DONE

c. Verify closed the following valves:

LPSI FLOW CONTROL COLD LEG 2A (SI 139A)

LPSI FLOW CONTROL COLD LEG 2B (SI 138A)

SDCS LOOP 2 WARMUP VALVE (SI 135A)

SDCS LOOP 2 SUCTION ISOL DOWNSTREAM OUTSIDE (SI 407A)

d. Locally Open AND Lock LPSI PUMP A MINIMUM FLOW RECIRC STOP CHECK (SI 116A).

e. Open LPSI A DISCH HDR FLOW CONTROL VALVE (SI 129A).

f. Locally vent LPSI Pump A suction piping.

g. Start LPSI Pump A.

h. Throttle open LPSI Flow Control valves:

LPSI FLOW CONTROL COLD LEG 2A (SI 139A)

LPSI FLOW CONTROL COLD LEG 2B (SI 138A)

i. Continue venting LPSI Pump A suction piping until all air is removed.

j. Restore AND maintain RCS level >15.13 feet, top of Hot Leg.

OR 9.2 For Train B:

a. Verify LPSI Pump B secured.

b. Locally open LPSI PUMP B SUCTION ISOLATION VALVE FROM RWSP (SI 109B).

13

Off Normal Procedure OP-901-131 Shutdown Cooling Malfunction Revision 304 E1 SYSTEM LEAKAGE (CONTD)

PLACEKEEPER START DONE

c. Verify closed the following valves:

LPSI FLOW CONTROL COLD LEG 1A (SI 139B)

LPSI FLOW CONTROL COLD LEG 1B (SI 138B)

SDCS LOOP 1 WARMUP VALVE (SI 135B)

SDCS LOOP 1 SUCTION ISOL DOWNSTREAM OUTSIDE (SI 407B)

d. Locally Open AND Lock LPSI PUMP B MINIMUM FLOW RECIRC STOP CHECK (SI 116B).

e. Open LPSI B DISCH HDR FLOW CONTROL VALVE (SI 129B).

f. Locally vent LPSI Pump B suction piping.

g. Start LPSI Pump B.

h. Throttle open LPSI Flow Control valves:

LPSI FLOW CONTROL COLD LEG 1A (SI 139B)

LPSI FLOW CONTROL COLD LEG 1B (SI 138B)

i. Continue venting LPSI Pump B suction piping until all air is removed.

j. Restore AND maintain RCS level >15.13 feet, top of Hot Leg.

NOTE If RCS is open and no Steam Generator is available, then RCS cooling is provided by steaming out of opening and makeup water from HPSI or LPSI Pumps.

10. IF EITHER Steam Generator is available AND RCS cooling is required, THEN perform the following:

10.1 Provide a feed path for ONE OR BOTH available Steam Generators using EITHER Condensate Pumps, Auxiliary Feedwater Pumps, OR Emergency Feedwater Pumps.

10.2 Provide a steaming path for ONE OR BOTH available Steam Generators using EITHER Atmospheric Dump valve OR Steam Bypass Control System.

14

SD-SI Specifically, a SIAS will cause the following actions relative to the SI system:

Starts two HPSI pumps (AB only if substituted for A or B)

Starts both LPSI pumps Starts associated HPSI and LPSI pump room coolers Starts up CVAS system Starts accident RAB HVAC systems Stops normal RAB HVAC systems Sends open signal to eight HPSI cold leg injection valves (225-228A (B))

Sends open signal to four LPSI cold leg flow control valves (138-139A (B))

Sends open signal to four SIT isolation valves (331-332A (B))(normally open)

Sends open signal to two RWSP suction valves (106A (B))

Sends close signals to four SIT leakage drain valves (303-304A (B))

Sends close signal to two HPSI hot leg injection drain valves (301, 302)

Sends close signal to two SI Sump suction valves (602A (B))

Short Term Recirculation (Figure 10)

During the injection mode of operation 55°F to 100°F water is being supplied to the RCS by the SIS to replace the coolant being lost through the break. Eventually the supply of water in the RWSP will be exhausted and the system will respond by automatically transferring system suction to the SI sump. This action occurs at an RWSP level of 10%.

Upon receipt of a RAS the system automatically responds in the following manner:

The SI sump outlet isolation valves, SI-602A(B), to the SIS supply header open.

Both LPSI pumps stop.

After the automatic actions have been verified, the following valves must be shut:

HPSI and LPSI pump suction valves (SI-106A and B) from the RWSP, Train A and B SI pumps recirc isolation valves, SI-120A(B) and SI-121A(B).

Revision 15 Page 42 of 83

WATERFORD 3 SES OP-902-002 Revision 020 Page 38 of 83 LOSS OF COOLANT ACCIDENT RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS CAUTION SI 120A(B) and SI 121A(B), SI PUMPS RECIRC ISOL should be closed within two minutes of receipt of RAS to prevent recirculating SI sump water to the RWSP.

RAS Initiation Criteria

• 45. IF the break is inside containment, AND RWSP level is less than 10%,

THEN perform the following:

a. Check RAS is initiated. a.1 Manually actuate RAS using Appendix 34, RAS Manual Actuation.

b. Verify that BOTH LPSI pumps are b.1 IF LPSI pump continues to run, stopped. THEN continue with Instruction Step 45.c

c. Verify that ESF PUMPS SUCTION SI SUMP valves are open:

SI 602A SI 602B

d. Close the SI PUMPS RECIRC ISOL VALVES within two minutes of receipt of RAS:

SI 120A SI 120B SI 121A SI 121B (continued)

WATERFORD 3 SES OP-902-004 Revision 016 Page 20 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

---

NOTE ----------------------------------------------------------

Actions to stabilize RCS temperature following an excess steam demand event should be initiated when BOTH of the following parameters are met:

CET temperatures rise Pressurizer pressure rise Stabilize RCS Temperature with SG ESD

• 18. Stabilize RCS temperature within PT Curves by performing the following:

a. For the LEAST AFFECTED SG:

1) Place the ADV to manual and fully open ADV.

2) Manually initiate EFAS.

3) Place the EFW Flow Control valve to manual and commence feeding.

4) Perform ANY of the following as necessary to establish RCS pressure and temperature control:

Throttle associated SG ADV.

Adjust associated SG EFW flow.

b. IF RCS pressure is greater than or equal b.1 IF RCS pressure is less than to 1500 psia, 1500 psia, THEN stabilize RCS pressure at a value THEN stabilize RCS not to exceed 1600 psid between the pressure at greater than RCS and the lowest SG pressure. HPSI shutoff head (1500 -

1600 psia).

c. REFER TO Step 20, HPSI Throttle Criteria.

WATERFORD 3 SES OP-902-004 Revision 016 Page 22 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS HPSI Throttle Criteria

• 20. IF HPSI pumps are operating, AND ALL of the following conditions are satisfied:

RCS subcooling is greater than or equal to 28ºF PZR level is greater than 7% [23%]

and controlled Verify at least ONE SG is available for RCS heat removal and level is being maintained or restored to within 55 to 70% NR

[60 to 80% NR] using EFW in auto or manual.

RVLMS indicates level higher than Hot Leg by at least ONE of the following:

QSPDS REACTOR VESSEL LEVEL 5 NOT voided VESSEL LEVEL PLENUM greater than or equal to 80%

THEN perform ANY of the following:

Throttle HPSI flow.

Stop ONE HPSI pump at a time.

SD-RCS TABLE 2 EQUIPMENT DATA PRESSURIZER Dimensions Overall length, including skirt and spray 441 in.

nozzle Outside diameter 106-1/2 in.

Inside diameter 96 in.

Cladding thickness (minimum) 1/8 in.

Dry weight, including heaters 203,300 lbm Flooded weight, including heaters 297,700 lbm PRIMARY SAFETY VALVE Design pressure 2500 PSIA Design temperature 700°F Set pressure 2500 PSIA +/-1%

Capacity at set pressure, each 460,000 lbm/hr Type Spring loaded safety-balanced bellows. Enclosed bonnet.

Accumulation 3%

Backpressure Max buildup/max. superimposed 500 PSIG/130 PSIG Blowdown 4%

REVISION 21 Page 55 of 117

SD-CC The CCW Non-Nuclear Safety (NNS) Header supplies reactor auxiliaries necessary for normal plant operation. Reactor auxiliaries and other services located on the NNS Header are:

Two Boric Acid Concentrators Primary and secondary sample coolers Blowdown System radiation monitor coolers Blowdown System metal transport sample cooler Two Waste Gas Compressors CCW corrosion rate monitor CCW Chemical Feed Tank CCW filters The air-operated supply and return isolation valves for the NNS Header are open during normal operations.

The CCW Surge Tank is located approximately 70 feet above the pumps and is connected to the two return headers to the CCW pumps common suction. It provides required net positive suction head (NPSH) for the CCW Pumps and accommodates fluid expansion and contraction of the system due to temperature changes. It also provides an immediate makeup source for system fluid changes. The normal makeup water source is from the Condensate Makeup (CMU) System through an air-operated level control valve. In addition, two seismic, category I CCW Makeup Pumps take suction from the Condensate Storage Pool (CSP) and are capable of providing makeup for the CCW Surge Tank.

If leakage from the CCW System causes CCW Surge Tank level to lower, level switches on the surge tank first cause the CCW Makeup Pumps to start, and the Dry Cooling Towers to bypass and isolate. If level continues to lower, the CCW System splits into two independent trains by closing the eight air-operated cross-connect valves on the pump suction and discharge headers and closing the four air-operated supply and return isolation valves for the CCW AB Header. Flow is lost to the AB Header and the NNS Header at this point. Since flow to the Reactor Coolant Pumps (RCPs) is lost, operator action is required to either quickly restore flow to the AB Header or to trip the Reactor and secure all four RCPs.

If a Safety Injection Actuation Signal (SIAS) occurs, the CCW System automatically splits into two independent trains by closing the CCW pump suction and discharge cross-connect valves and the Train B supply and return valves to the AB Header. The two selected CCW pumps receive start signals on SIAS. The NNS Header supply and return valves automatically close. Flow to the AB Header (i.e. RCPs) continues to be supplied from Train A. In addition, CCW flow is automatically initiated to Shutdown Cooling Heat Exchanger B. SIAS occurs in response to a design basis accident, such as a LOCA or MSLB.

REVISION 22 PAGE 11 OF 123

SD-CC Normal WCT Basin makeup is from the CMU System utilizing air-operated level control valves. A manual backup source of makeup is from the CW System.

If a SIAS occurs, both trains of ACCW automatically start, if not already running. The setpoint temperature maintained by ACC-126A(B) is automatically raised to 115°F.,

unless WCT Basin temperature is low. SIAS occurs in response to a design basis accident, such as a LOCA or MSLB.

WCT Basin water may also be aligned as a secondary source of Emergency Feedwater (EFW) if the inventory of the Condensate Storage Pool (CSP) is depleted during an accident.

Inventory from one WCT Basin may be transferred to the other by gravity drain using normally closed, air-operated cross-connect valves ACC-138A and B.

The non-safety WCT Basin Filtration System (Figure 26) is completely independent of the safety related ACCW System. It serves to circulate, filter, and chemically treat the water stored in the WCT basins. Each basin contains suction manifolds, pump, filter, discharge spargers, and chemical injection equipment.

Component Cooling Water Makeup Subsystem Flowpath CCW Makeup subsystem contains two independent, 100% capacity trains. Suction is taken from the Condensate Storage Pool (CSP) by two high-capacity centrifugal pumps, which normally remain in standby. The headers downstream of the pump discharge check valves are normally pressurized with CMU from the Condensate Storage Tank (CST) Pumps. A low level condition in either the EDG Jacket Water Standpipes or the Essential Chiller Expansion Tanks utilizes this CMU source from the headers for normal makeup. As a backup, a low level condition in the CCW Surge Tank will automatically start the CCW Makeup Pumps and initiate flow to the CCW Surge Tank from the CSP.

CCW Makeup can be manually aligned to the Spent Fuel Pool to provide pure water as makeup. The CCW Makeup Pumps can be manually started if necessary to fill either the CHW Expansion Tank or the Emergency Diesel Generator Jacket Water Standpipe if normal makeup is not sufficient. In addition, CCW Makeup may be manually aligned to pump the CSP to the CW System.

REVISION 22 PAGE 13 OF 123

SD-CC Fuel Pool Temperature Control Valve, CC-620 The Fuel Pool Temperature Control Valve (CC-620) is located in the Fuel Handling Building (FHB) +1 ft elevation. It modulates position to control CCW flow based on controlling Fuel Pool Cooling and Purification System (FS) temperature at the common outlet of the Fuel Pool Heat Exchangers below 120°F (Figure 12).

Table 1.19 contains valve data.

CC-620 is a piston air-operated, 12 inch butterfly valve; that needs power (and air) to close. CC-620 fails as-is in the event of TOTAL loss of air to the actuator and its associated accumulator. However, on lowering air pressure, a 75% trip-block pilot valve aligns air from an accumulator to close CC-620. This trip-block pilot valve also prevents air from bleeding off the actuator once CC-620 is closed. This action blocks CC-620 in the closed position on a lowering supply air pressure, and makes the valve operate as a FAIL-CLOSED valve for loss of IA. CC-620 control is restored when IA supply pressure is restored. (See TD-F130.0025 page 201 for actuator information.)

There are two CLOSE/CONTROL control switches (A and B train) on CP-8. Both control switches must be in the CONTROL position to allow CC-620 to control temperature to setpoint of FS-ITIC-5172. Each control switch controls a 3-way solenoid valve that will close CC-620 if either is energized (and air pressure is present). The solenoid valves, control circuits, and position indicating lights are powered from:

Train A CC-ISV-0620-1 PDP-90A Train B CC-ISV-0620-2 PDP-91B The A powered solenoid receives a signal to energize, thereby closing CC-620, under either of the following two conditions:

AB Loop Supply and Return isolation valves, CC-200A and CC-727, are fully closed. (These valves close on CSAS Train A / Low CCW Surge Tank level),

or Receipt of SIAS Train A The B powered solenoid receives a signal to energize, thereby closing CC-620, if AB Loop Supply and Return isolation valves, CC-200B and CC-563, are fully closed.

(These valves close on SIAS Train B / Low CCW Surge Tank level).

An override feature exists for the auto-closures discussed above. Taking both the control switches to CLOSE and then back to CONTROL accomplish this override.

Temperature controller FS-ITIC-5172 is a pneumatic type controller located in the FHB

+1 ft elevation in the Fuel Pool Heat Exchanger Room. The controller does not require electrical power. A thermobulb senses the temperature of the FS flow in the common header downstream of both fuel pool heat exchangers. The sensed temperature is REVISION 22 PAGE 37 OF 123

SD-CC On SIAS the CCW System splits into two independent trains by automatically closing the following (Figure 22):

All eight CCW pump suction and discharge cross-connect valves Train B supply to and return from the AB Header, CC-200B and CC-563 NNS Header Supply and Return Valves, CC-501 and CC-562 The AB Header continues to be supplied from Train A. The following CCW System components also actuate on an SIAS:

CC-963B, Shutdown Heat Exchanger B Outlet Valve, opens Two selected CCW pumps get start signals from Sequencers All CFC containment isolation valves fail open CC-835 A and B, CFC Temperature Control Valves, fail open CC-620, Fuel Pool Temperature Control Valve, closes On CSAS the CCW System further protects the independent trains, safety headers, and containment isolation by automatically closing the following (Figure 23):

Train A supply to and return from the AB Header, CC-200A and CC-727 CCW Containment Isolation Valves, CC-641, CC-710, and CC-713 REVISION 22 PAGE 46 OF 123

PAGE 119 OF 123 SD-CC-22

SD-PLC are on-off heaters called back-up heaters. The back-up heaters function automatically when Pressurizer pressure drops significantly below setpoint, or on a large insurge of coolant to the Pressurizer.

Both the proportional heaters and the backup heaters are automatically turned off if either of the following occurs:

Pressurizer level drops to 28%, or Pressurizer pressure increases to 2270 PSIA.

Proportional Heaters (Figure 12)

There are two identical banks of proportional heaters. Bank 1 is powered from 480VAC Bus 32A. Bank 2 is powered from 480VAC Bus 32B. Each bank is comprised of 3 heaters at 50 kW per heater, for a total of 150 kW per bank. Each heater is provided with a local circuit breaker which can isolate it in the event of an electrical fault.

Heater output is dependent on electrical power input. The power supply to the proportional heaters is regulated by a silicon control rectifier (SCR) known as the proportional power controller which varies the power to the proportional heaters in response to the control signal from the pressure indicator controller.

A silicon control rectifier regulates power to the heaters by triggering 480 VAC pulses (full line voltage is applied at all times) using a time proportioning control device. The time proportioning control system fires 480VAC pulses on a set number of half cycles proportional to the control signal supplied to it. As the control signal strength decreases, the on time pulse frequency increases, always firing at 0 line voltage and power always delivered at 480 VAC.

The result of using this system of power control is a change in effective heater voltage (RMS voltage) while always supplying the heaters with the same voltage and current when they are on.

Example: If the heaters are on for 40 seconds over a one minute period, their RMS voltage applied was 40/60 (2/3) of line voltage, giving them an effective voltage supply of 320 VAC.

The proportional heater control switches actuate breakers which supply power to the proportional power controllers. The switches have positions of OFF/ON and are spring return to normal. They supply signals to a logic circuit which ensures that certain prerequisites are met before allowing the breaker to shut. Regardless of control switch position, the heaters are protected from low Pressurizer level (28% actual level), high Pressurizer pressure (2270 PSIA) undervoltage, or overcurrent by signal gates downstream of the switches.

REVISION 10 PAGE 29 of 69

SD-PLC SYSTEM OPERATIONS Note:

This discussion is for training information only. Refer to current approved procedures for actual system requirements.

NORMAL OPERATION Under steady state conditions, heat must be added to the Pressurizer to compensate for the heat loss through the vessel walls, and for the heat defect introduced by continuous spray through the spray bypass valves. The proportional heaters have been designed to provide enough heat at half capacity to make up for total heat loss from the Pressurizer. Therefore, when Pressurizer heat inputs and losses are balanced the signal supplied by the Pressurizer control panel must maintain the proportional heaters at half power.

Automatic Operation When the controller setpoint is at 2250 PSIA, the proportional heaters will be at full power at 2225 PSIA and zero power at 2275 PSIA. The spray valves will start to open at 2280 PSIA and be fully open at 2300 PSIA. The pressures given for the operating range of the proportional heaters and the spray valves are good only if the setpoint of the pressure controller is 2250 PSIA and the setpoint on the spray controller is set at 75%. Changing the controller setpoint will cause the operating pressures of the proportional heaters and the spray valves to change by the same amount as the setpoint change.

Backup heater control is independent of pressure controller output. The heaters will always energize at 2200 PSIA decreasing, and de-energize at 2225 PSIA increasing.

The Pressurizer pressure controller cannot always hold the pressure at setpoint because it is a proportional controller. The pressure will equal the setpoint only if it takes one-half power to the proportional heaters to make up for heat loss and the spray bypass flow. Let's assume that the spray bypass flow and heat loss is such that it requires three-quarters power to the proportional heaters to hold the pressure constant.

We would then find that the pressure would have to be 12.5 PSIA below setpoint to get a controller output of 16.5%. At 16.5% output, the proportional heaters would be at 3/4 power. If the controller setpoint was 2250 PSIA, the steady state pressure would be 2237.5 PSIA. If it was desired to have pressure at 2250 PSIA, the operator could raise REVISION 10 PAGE 37 of 69

SD-PLC the setpoint to approximately 2267.5 PSIA. Pressure would stay 12.5 PSIA below setpoint, resulting in a pressure of 2250 PSIA.

Manual Operation Setpoint Adjustment The Pressurizer pressure controllers on CP-2 and LCP-43 have raise and lower pushbuttons which control the pressure setpoint. The setpoint meter has a range from 1500 to 2500 PSIA. When the controller is set for manual output adjustment, setpoint will track process to minimize controller bump on the return to automatic control. The operator must then reset the setpoint to desired pressure.

Controller Output Adjustment The operator has the ability to take manual control of the Pressurizer pressure controller output, which will allow the proportional heaters and spray valves to function normally on the output signal from the pressure controller. To place the controller in MANUAL, the MANUAL pushbutton is depressed, which makes controller output dependent on the operator depressing the raise/lower buttons on the M/A station. Note that raising the controller output will tend to lower system pressure and conversely, lowering the controller output tends to raise system pressure.

Control of Proportional Heater Output Control of the proportional heaters requires a signal from the pressure controller.

Therefore, the only method of manually controlling heater output is variance of the pressure indicator controller output between 0 and 67%. Note that lowering the controller output will increase heater output.

Manual Control of the Spray Valves Manual operation of the spray valves is accomplished by depressing the MANUAL pushbutton on the spray valve control station. This allows the operator to regulate the controller output current to the spray valves. The setpoint potentiometer may also be adjusted to lower the input signal needed to open the spray valve.

Manual Operation of the Backup Heaters The backup heaters are operated manually by placing the control switch for that bank in the ON position.

REVISION 10 PAGE 38 of 69

9.13 BORON EQUALIZATION NOTE To prevent Pressurizer heater cutout, avoid operating with Pressurizer pressure near the heater cutout pressure of 2270 psia while on Boron Equalization. [CR-WF3-2012-01861]

NOTE At any time the plant is performing a significant power change, Boron Equalization should be performed to prevent an unequal balance of boron concentration between the Pressurizer and the Reactor Coolant System. However, if a change in RCS boron concentration of >50 PPM is anticipated, then Boron Equalization shall be initiated to maintain RCS and Pressurizer boron concentrations within 10 PPM.

9.13.1 Since this evolution affects reactivity the following practices should be observed:

SM/CRS should be informed of this evolution Operator should minimize distractions while performing and receive a peer check Monitor reactor power and temperature for changes after performing 9.13.2 Perform Boron Equalization as follows:

9.13.2.1 Place available Pressurizer Backup Heaters control switches to On.

9.13.2.2 Reduce Pressurizer Spray Valve Controller (RC-IHIC-0100) setpoint potentiometer to establish spray flow and maintain RCS Pressure 2250 PSIA (2175 - 2265).

NOTE Boron Equalization may be secured when the Boron concentration difference between the RCS and the Pressurizer is <10 PPM, but should continue until the plant has been returned to a Steady State condition and subsequent samples of the RCS and the Pressurizer show that Boron concentrations are not changing.

9.13.3 Secure Boron Equalization as follows:

9.13.3.1 Place Pressurizer Backup Heater control switches to Auto.

9.13.3.2 Set Pressurizer Spray Controller setpoint potentiometer to approximately 75%.

OP-010-005 Revision 328 Attachment 9.13 (1 of 1) 87

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 C AUTOMATIC ACTIONS

1. Loss of 125 Volt DC Bus A-DC Reactor Trip Switchgear Breakers 1, 3, 5 and 7 open causing a Reactor Trip Loss of control power to 7KV bus 1A Startup Transformer supply breaker (7KV-EBKR-1A-4)

Loss of control power to 7KV bus 1A Auxiliary Transformer supply breaker (7KV-EBKR-1A-1)

Loss of control power to all Safety Train A 4KV and 480V equipment SDCS Loop 2 Suction Isolation Downstream Inside SI-405A fails closed, if gagging device not installed RC Loop 2 SDC Suct HDR PRESS Equalizing and INSD CNTMNT Isol SI-4052A fails closed NOTE 480V breakers overcurrent protection is available from Electronic Current Sensing Device (ECS).

The 4KV and 480V switchgear in Safety Train A are left without control power. All remote manual control and automatic protection of the switchgear and the associated connected components are disabled.

Safety Train A load shedding and load sequencing capabilities are disabled.

The following Emergency Feedwater valves fail open:

EFW to SG1 Primary Isolation (EFW-228A)

EFW to SG2 Backup Isolation (EFW-229B)

Loss of Emergency Diesel Generator A control power.

Loss of EDG A Fuel Oil Booster Pump.

Loss of EDG A Field Flashing power.

Emergency power supply (A1-DC) to Control Room Annunciator Logic Panel is lost.

Diesel Generator A Standpipe Level Cntrl Vlv (CMU 524A) fails open. The EDG A Standpipe overflows to the -35 RAB oil sump.

CCW RCP OUTLET INSIDE ISOL (CC 710) fails open. Lose CSAS protection for this valve.

Emergency power to SUPS A, SUPS MA and SUPS MC is lost.

Containment Spray Header A Isolation (CS 125A) fails open.

Loss of Control Room Emergency Lighting Panel (ELP 322-PA).

13

MAIN STEAM SYSTEM STANDBY BREAKER LINEUP (CONTD)

COMPONENT COMPONENT DESCRIPTION LOCATION REQUIRED PERFORMED BY IV BY NUMBER POSITION (INITIAL/DATE) (INITIAL/DATE)

MS 2 BYPASS TO COND A MS-EBKR-313AB-9C TGB +15 9T&E ON N/A DRIPPOT S/U DRN (MS-308A)

MS 1 BYPASS TO COND A MS-EBKR-313AB-9F TGB +15 9T&E ON N/A DRIPPOT S/U DRN (MS-307A)

MS 1 BYPASS TO COND B MS-EBKR-313AB-9J TGB +15 9T&E ON N/A DRIPPOT S/U DRN (MS-307B)

MS 1 BYPASS TO COND C MS-EBKR-313AB-9M TGB +15 9T&E ON N/A DRIPPOT S/U DRN (MS-307C)

TURBINE REHEAT CONTROL MS-EBKR-84A-14 RAB +21 11A&G ON N/A (MSISV0031A&B)

EFW PUMP AB STEAM MS-EBKR-39B-19 RAB -35 8A&L ON LINE DRAIN (MS-409)

EFW PUMP AB STOP VALVE MS-EBKR-62AB-9 RAB +21 11A&J ON B/U POSITION IND (MS-416)

EFW PUMP AB TURBINE MS-EBKR-311AB-3C RAB +21 10A&H ON STOP VALVE (MS-416)

TURBINE DRAIN VALVES MS-EBKR-67A-2 TGB +40 7T&E ON N/A MSISV0161B & MSISVO163B STEAM BYPASS CONTROL SYS MS-EBKR-87A-5 RAB +21 12A&H ON N/A CONTROL POWER SUPPLY CP-5 TURB BYPS VLV 1C TO CNDSR C MS-EBKR-TGB-A38 TGB +15 9T&E ON N/A (MS-319C) CONTROL POWER TURB BYPS VLV 1B TO CNDSR B MS-EBKR-TGB-B34 TGB +15 10T&E ON N/A (MS-319B) CONTROL POWER TURB BYPS VLV 2B TO CNDSR B MS-EBKR-TGB-B36 TGB +15 10T&E ON N/A (MS-320B) CONTROL POWER TURB BYPS VLV 2C TO CNDSR C MS-EBKR-TGB-B38 TGB +15 10T&E ON N/A (MS-320C) CONTROL POWER EFW PUMP AB MAIN STEAM MS-EBKR-AB-21 RAB +21 10A&J ON SUPLY ISOL (MS-401A)

OP-005-004 Revision 033 Attachment 11.7 (3 of 5) 73

Operating Instruction OI-038-000 Emergency Operating Procedures Revision 013 Operations Expectations / Guidance 5.2.6 Verify Core Heat Removal The operator should check RCPs are operating by checking control switch indication, motor current and delta-p indication on CP-2. If no RCPs are operating, then the CRS is not required to obtain additional information for the Core Heat Removal Safety Function. The CRS should go to the RCS Heat Removal Safety Function.

The loop delta-T should be checked only for loops with an operating RCP.

The operator may use control board indications or the PMC to obtain this indication.

TH subcooling should be used for forced circulation (CP-2 & CP-8).

5.2.7 Verify RCS Heat Removal Main Feedwater in RTO is acceptable to satisfy feedwater control requirements.

The operator should allow EFW system to operate in automatic. It is not required to manually initiate an EFAS signal.

Verifying Feedwater flow is not excessive is accomplished by verifying FWCS is in RTO.

If an ESD is indicated and the faulted Steam Generator has emptied during the performance of the SPTAs, then the operators should perform the following:

Place the ADV in manual and fully open the ADV on the unaffected Steam Generator Manually initiate EFAS for the unaffected Steam Generator Commence feeding the unaffected Steam Generator and continue to feed to maintain the NR Steam Generator level band Ensure RCS pressure is above or rises above HPSI shutoff head and control pressure with Pzr spray so that the differential pressure between the RCS and the lowest Steam Generator pressure does not exceed 1600 psid.

When CET temperatures have stabilized, the ADV should be throttled to maintain temperature Check MSIS initiated by ensuring the Trip Path lights are not illuminated on CP-7 for MSIS and that the MSIS Train Logic Initiated annunciators (CP-2 panel K L-19 & L-20) are in alarm. Other expectations for this step include verifying the Main Steam and Main Feedwater Isolation valves are closed.

15

Operating Instruction OI-038-000 Emergency Operating Procedures Revision 013 Operations Expectations / Guidance 5.4.112 Verify Most Affected SG Isolated None 5.4.113 Verify MSIS Actuated The minimum criteria for verifying MSIS actuation is verifying Main Feedwater and Main Steam Isolation valves closed.

5.4.114 Verify MSIS Actuation The minimum criteria for verifying MSIS actuation is verifying Main Feedwater and Main Steam Isolation valves closed.

5.4.115 Verify ONE RCP in Each Loop Secured Tripping RCP 1A and 2A is the preferred combination of pumps to secure due to the higher head of RCP 2A.

5.4.116 Verify Operation of Ventilation Systems None 5.4.117 Verify Proper CCW Operation None 5.4.118 Verify RCP Operating Limits Tripping RCP 1A and 2A is the preferred combination of pumps to secure due to the higher head of RCP 2A.

Verifying CCW available to the RCPs includes checking CCW containment isolation valves on CP-8, CCW pumps operating, and seal cooler isolation valves on CP-2.

5.4.119 Verify RCS is Not Water Solid None 5.4.120 Verify RCS Temperature Control None 37

SD-PPS signals actuate the reactor trip breakers, the ESFAS trip signals actuate various ESF system components, such as valves, pumps, fans, dampers, etc. The ESFAS bistable outputs (or ESFAS actuation signals) interface with these plant components through the Auxiliary Relay Cabinets located on RAB +21 elevation. The Auxiliary Relay Cabinets contain a multitude of relays that are actuated by ESFAS signals to operate various plant components as assumed in the safety analyses. See FSAR table 7.3-5 for a complete list of components operated by ESFAS actuations.

Table 10 provides a list of inputs to the ESFAS Subsystem.

Table 3 and Figure 32 summarize the ESFAS signals, their coincidence and setpoints, and their bases. Refer to Figure 33 for the following general discussion of how an ESFAS signal is generated. Sensor inputs are sent to trip normally energized bistables in CP-10. These bistables use the same type of bistable comparator cards and bistable relay cards that the RPS bistables do. Like the RPS, the ESFAS bistables de-energize to actuate ladder contacts in six different matrices with each matrix having four normally energized matrix output relays that open normally closed trip paths. Each trip path consists of six contacts in series, with each contact being fed from a 2/4 coincidence matrix. An open trip path actuates redundant initiation relays that send independent ESFAS signals to relays in each of the two Auxiliary Relay Cabinets. The Auxiliary Relay Cabinet relays then independently actuate their two trains of ESF system components. Figure 34 shows the ESF Status Panel mimic on CP-10. The darkened lights show that they are normally energized in an at-power, not tripped condition.

Unlike the RPS, all ESFAS signals (except EFAS), once actuated, de-energize a "lockout" relay that opens a contact in the trip path to maintain the ESFAS actuation relays de-energized even if the actuating bistables reset and the matrix relays later re-energize. This ensures that the ESF system remains in its safeguards lineup until deliberate manual action is taken to reset its lockout relay and restore the associated ESF system lineup.

Emergency Feedwater Actuation Signal (EFAS)

This signal is generated by low SG water level at 27.4% (Reset 68.4%) narrow range (affected SG) coincident with either (1) SG pressure in the affected SG above 666 PSIA (variable) OR (2) the affected SG steam pressure being above the other SG's steam pressure by at least 123 PSID. There are two separate actuation signals, one for SG 1 and one for SG 2. EFAS signals will actuate the following equipment:

Start the EFW pumps Close the SG blowdown isolation valves Open the affected SG EFW isolation valves Send a permissive open signal to the affected SG FCVs, which will then cycle on SG water level.

Revision 17 Page 36 of 158

Technical Guide for Excess Steam TG-OP-902-004 Demand Recovery Procedure Revision 307 Step Number 8 Verify MSIS Actuation Objective The intent of this step is to ensure that all MSIVs are closed very early in the procedure in an attempt to isolate the break. Closing the MSIVs effectively separates the two steam generators which facilitates determining which SG is most affected if the break is upstream of the MSIVs.

The step is placed early in the procedure to ensure that MSIVs are closed.

Instructions The operator is directed to verify that a MSIS actuated.

Contingency Actions None Justification for Deviations At Waterford, the minimum criteria for verifying a MSIS actuation is verifying Main Feedwater and Main Steam Isolation Valves are closed. Thus, Waterford differentiates from the EPG by verifying MSIVs are closed by verifying a MSIS has actuated.

References None 25

Operating Instruction OI-038-000 Emergency Operating Procedures Revision 013 Operations Expectations / Guidance 5.0 PROCEDURE 5.1 GENERAL EXPECTATIONS Where multiple indications for one parameter exist, more than one instrument should be used to obtain a particular reading. Accidents may cause irregularities in particular instrument readings.

Systems should not normally be placed in manual unless it is determined that manual operation will provide better control based on the plant status or mis-operation in automatic is apparent. Systems placed in manual should be monitored to ensure proper operation.

Safety Functions, as specified on the Safety Function Status Checklist, shall be continuously monitored throughout the use of this procedure.

When responding to a request for a parameter, the operator should voice the parameter as well as the trend of the parameter.

Verification of an RCS temperature response to a plant change during natural circulation cannot be accomplished until approximately 5 to 15 minutes following the action due to increased cycle times. CETs should be used during natural circulation conditions.

Hot leg temperatures and Cold leg temperatures may be influenced by Safety Injection flow. Multiple indications and CET temperatures should be used to determine the Reactor Coolant System temperature.

Parameter values that are bracketed should be used if a harsh environment in containment exists. A harsh environment is defined as Containment Temperature greater than or equal to 200 F.

6

11.2 CONTAINMENT SPRAY SYSTEM STANDBY BREAKER LINEUP COMPONENT COMPONENT DESCRIPTION LOCATION REQUIRED PERFORMED BY IV BY NUMBER POSITION (INITIAL/DATE) (INITIAL/DATE)

CONTAINMENT SPRAY RISER CS-EBKR-213A-2M RAB +21 8A&H OFF LEVEL PUMP A CONTAINMENT SPRAY RISER CS-EBKR-213B-2M RAB +21 10A&L OFF LEVEL PUMP B CS-EBKR-3A-6 CONTAINMENT SPRAY PUMP A RAB +21 9A&G RACKED IN CONTAINMENT SPRAY PUMP A CS-EBKR-3A-6A RAB +21 9A&G ON MOTOR HEATER CS-EBKR-3B-5 CONTAINMENT SPRAY PUMP B RAB +21 10A&K RACKED IN CONTAINMENT SPRAY PUMP B CS-EBKR-3B-5A RAB +21 10A&K ON MOTOR HEATER CONTAINMENT SPRAY HDR A CS-EBKR-1A-6 RAB +21 11A&H ON ISOLATION (CS-125A) (DC)

CONTAINMENT SPRAY HDR B CS-EBKR-1B-6 RAB +21 10A&J ON ISOLATION (CS-125B) (DC)

CNTMT SPRAY HDR A RISER CS-EBKR-60A-9 RAB +21 11A&H ON CHK VLV BYPASS (CS-129A)

CNTMT SPRAY HDR B RISER CS-EBKR-61B-9 RAB +21 9A&K ON CHK VLV BYPASS (CS-129B)

SHUTDOWN COOLING HX A & B CS-EBKR-90A-28 RAB +35 11A&G ON TEMPERATURE RECORDER OP-009-001 Revision 306 Attachment 11.2 (1 of 1) 40

SD-SBC is demanded. These setpoints are obtained from a curve (Figure 3) that is a linear fit to the normal steady state main steam header pressure that will exist from 20% to 100%

power (as sensed by steam flow). The curve has an upper limit on pressure below 5%

power that is equal to hot zero power steam pressure (approximately 970 PSIA). The setpoint curve is programmed above expected steam pressure based on steam flow so that the SBCS valves will not open on a 10% step decrease in load. Per CR-WF3-2005-2799, the inputs have uncertainty and the square root module is not perfect. At low steam flows of less than 90,000 lbm/hr per steam generator, the SBCS 970 psia control setpoint extends out beyond 5% steam load and occasionally beyond 10%

steam load. At some point between 5% and approximately 10% steam load, the control setpoint will rapidly converge with the ideal setpoint ramp as a function of increasing steam load.

The steam flow input for SP1 generation is steam flow (W S1) from steam generator 1 (MS-IFT-1011) as transmitted from Feedwater Control System 1. The SP2 program receives its steam flow (W S2) from S/G 2 (MS-IFT-1021). Both steam flow signals are made to pass through a lag compensation network to increase system stability and response. This circuit retards the motion of the setpoint away from its original value, so that SBCS has a quicker response to pressure increases caused by load rejection events. In addition, this lag circuit will filter high frequency noise, if it is present.

To enable the SBCS to respond quickly to disturbances originating in the reactor coolant system, the setpoint is biased by a signal that is proportional to the deviation of the Pressurizer pressure (PP1 and PP2) from its normal operating value. This bias will lower or raise the pressure setpoint 0.2 PSI for every 1 PSI the Pressurizer pressure increases or decreases from the normal operating pressure (2250 PSIA).

The setpoint networks sum the setpoint value based on steam flow with the bias value based on deviation from normal Pressurizer pressure, and the resultant signals are termed SP1 and SP2.

"Permissive" Pressure Setpoint (Figure 4) - The permissive demand channel selects the lowest of the two independently generated setpoint signals (SP1 and SP2). An adjustable bias is provided to ensure that the permissive setpoint is at least a minimum value (15 psi) below the Main Pressure Setpoint. The difference between the permissive setpoint and the main pressure setpoint accounts for any dynamic differences between the two channels and to provide smoother transient behavior in the transition from Quick Opening to Modulation Control.

"Main" Pressure Setpoint (Figure 5) - The main pressure setpoint channel selects the highest of the two independently generated setpoint (SP1 and SP2) and forwards the high signal to the Master Pressure Controller (MS-IPIC-1010).

Automatic Permissive Signal Generation (Figure 4) - The higher of the two main steam pressure inputs (MS-IPT-1010 or MS-IPT-1020) is compared to the permissive setpoint Revision 10 Page 9 of 68

SD-FWC The RTO condition is initiated by a reactor trip through four undervoltage relay status signals received from the Control Element Drive Mechanism Control System (CEDMCS). They are input to FWCS as follows:

UV 1 and 2 are direct inputs to FWCS 1.

UV 3 and 4 are direct inputs to FWCS 2.

FWCS 1 sends UV 1 and 2 to FWCS 2.

FWCS 2 sends UV 3 and 4 to FWCS 1.

As shown on Figure 23, at least 1 UV signal directly from CEDMCS coupled with one UV signal from the other FWCS is required to initiate RTO. This increases system reliability by requiring both FWCSs to recognize a reactor trip before starting control action.

If one or both pairs of UV signals are present a bistable output is generated to the three time delay networks. The time delay networks will hold an output for 10 seconds after receiving an input signal. This assures that the FWCS will maintain the reactor trip status long enough to switch to RTO. An output from two of three time delay networks generates the RTO logic.

Figure 24 shows that when the RTO logic signal exists, the following limitations are placed on the feedwater regulating valves and the pump speed program:

Flow demand to the main feedwater pump is limited to 0 percent, holding it at minimum speed. If the M/A station is in Manual, RTO will force the controller to Auto for 5 seconds. This allows the speed to be reduced to minimum speed. After 5 seconds the Operator can place the M/A station in manual as required Flow demand to the Startup Regulating valve is limited to ~ 3.5 percent, holding it at its minimum opening.

Flow demand to the Main Regulating valve is limited to ~ 0 percent, causing it to close.

With the Feedwater Turbine in Local Manual Control upon receipt of a Reactor Trip Override signal the local station will be inhibited for 5 seconds to allow speed to go to 3900 RPM REVISION 13 Page 24 of 79

SD-FWC When the reactor trips, a large level decrease in the SGs will occur due to shrink. This large level deviation will result in the FWCS generating a large FLOW DEMAND signal.

With the RTO present, the FWCS should only supply enough feedwater to make up for decay heat removal and to slowly return the SG downcomer water level to normal. This is accomplished as follows:

There are three comparators that are constantly comparing the actual FLOW DEMAND signal with minimum flow signal (~ 3.5 percent of rated flow).

If the FLOW DEMAND signal is greater than ~ 3.5 percent signal, these three comparators seal in the RTO logic signal.

When the SG levels are returned to a level that results in the FLOW DEMAND signal being less than ~ 3.5 percent, the RTO condition is reset and the FWCS functions as normal in AUTO.

The RTO programs are only activated for the respective M/A Valve Controller Stations that are in the AUTO mode of operation.

Level Deviation Condition A deviation between the two level measurement channels on the same SG indicates an instrument problem and could result in the FWCS controlling at an improper level. On a deviation greater than 13 in. (approximately 7.2 percent), the following actions occur:

SG LEVEL DEVIATION alarm - CP-1.

Main regulating valve M/A Station shifts to MANUAL and holds last automatic output.

Startup Regulating Valve M/A Station shifts to MANUAL.

Pump speed M/A Station to MANUAL.

These actions occur regardless of the position of the Level Channel Selector Switch.

If the level deviation alarm cannot be cleared and the AUTOMATIC mode of operation is desired, a jumper must be placed across two front terminal points in the FWCS calculator to allow the M/A Stations to be returned to AUTO.

REVISION 13 Page 25 of 79

MAJOR VALVES Cooling Water Temperature Control Valve (TC-147)

NOTE: DDC operation (auto mode) is not available and has been permanently removed.

Air-operated temperature control butterfly valve, TC-147, is located at the heat exchanger shell side common outlet header. The valve controls the amount of cooling water flow through the heat exchanger, to maintain temperature. TC-147 fails open on a loss of air pressure or fails as is on a loss of electrical power. The valve is powered from PDP-48A.

The valve position is controlled by a three-position, MORE/LESS/NORMAL spring return to NORMAL, control switch on CP-1. Turning the switch to MORE or LESS will control the volume of water flow through the cooler. With the MORE position meaning more temperature or the valve moves in the closing direction. An indicator (TC-IZI-147A) for valve position is located on CP-1 and aids manual operation. The operator maintains the temperature of the water leaving the cooler between 40°F - 110°F.

Cooling Water Pressure Control Valve (TC-135)

Air-operated pressure control butterfly valve, TC-135, is located in the bypass line of the heat exchangers. The valve is used to bypass the heat exchangers to balance system pressure as the TCV changes the flow through the heat exchanger to maintain 70-100 psig/design. TC-135 closed on a loss of air pressure or fails as is on a loss of electrical power. The valve is powered from PDP-48A.

The pressure transmitter (TC-IPI-9277) controls the PCVs position based on system pressure downstream of the heat exchangers. In low heat load or cold circulating water conditions, the TCV (TC-147) reduces flow through the heat exchangers. This causes downstream system pressure to drop, causing the PCV (TC-135) to open. The pressure transmitter (TC-IPI-9277) located at the common outlet header downstream of this valve, keeps the valve in a partially open position during normal operation.

The MANUAL position enables the operator to control the valve position from the Control Room. The valve position is controlled by a three-position, MORE/LESS/NORMAL spring return to NORMAL, control switch on CP-1. Turning the switch to MORE or LESS will control the pressure of water in the system. With the valve in the MORE position, pressure will increase. An indicator (TC-IZI-0135B) for valve position is located on CP-1 and aids manual operation. The operator maintains the system pressure between 70-100 psig/design.

Revision 8 Page 11 of 35

Revision 8 Page 29 of 35 11.2 AUXILIARY FEEDWATER SYSTEM STANDBY BREAKER LINEUP COMPONENT COMPONENT DESCRIPTION LOCATION REQUIRED PERFORMED BY NUMBER POSITION (INITIAL/DATE)

AFW-EBKR-1B-10 AUXILIARY FEEDWATER PUMP TGB +15 10T&B RACKED IN AUXILIARY FEEDWATER PUMP AFW-EBKR-1B-10B TGB +15 10T&B ON MOTOR HEATER AFW PRESS CONTROL AFW-125 &

AFW-EBKR-68B-47 TGB +40 3T&E ON AFW PUMP RECIRC AFW-113 AUXILIARY FEEDWATER SUPPLY AFW-EBKR-215B-5C TGB +40 10T&C ON HDR ISOL (AFW-144) 7KV-EBKR-1B-2 SWITCHGEAR 1B UV COIL POWER TGB +15 10T&B ON OP-003-035 Revision 305 Attachment 11.2 (1 of 1) 20

SD-EFW postulated feedwater or main steam line break that isolation is necessary to prevent excessive cooldown of the RCS.

The EFW system is normally lined up in standby during power plant operation. System operation will actuate automatically upon receipt of an Emergency Feedwater Actuation Signal (EFAS) which can indicate any of the following conditions:

Loss of Main Feedwater Flow Loss of Offsite and/or Onsite Electrical Power (AC)

Main Steam Line or Main Feedwater Line Break EFAS signals are generated by the Engineered Safety Features Actuation System section of the Plant Protection System (PPS) and are covered in detail in System Description SD-PPS.

There are two EFAS signals. An EFAS-1 initiates emergency feed to Steam Generator 1; EFAS-2 initiates emergency feed to Steam Generator 2. The EFAS-1 initiations are the following:

Steam Generator 1 level low (27.4% Narrow Range)

AND Steam Generator 1 pressure not low (greater than 666 PSIA)

OR Steam Generator 1 level low (27.4% Narrow Range)

AND Differential pressure between the two steam generators greater than 123 PSID with the higher pressure in Steam Generator 1.

An EFAS-2 is generated when these respective conditions exist for Steam Generator 2.

The EFAS logic prevents feeding a faulted steam generator. Two-out-of-four logic prevents EFAS initiation from a single failed instrument. The EFAS initiation logic is shown in Fig. 3 and a composite EFAS-1 signal flowpath is illustrated in Fig. 22.

Upon receipt of either EFAS, the EFW system logic will start the EFW pumps, determine which steam generator(s) is/are intact (if applicable) and open the EFW isolation valves to the intact steam generator(s), and close the containment isolation valves for the Steam Generator Blowdown System. The signals also provide an open permissive signal to the intact steam generator(s) control valves, but flow will not automatically initiate until the CRITICAL LEVEL (55% WR) is reached.

Revision 13 Page 7 of 94

REVISION 0 SD-EFW-19 SD-EDG Rev. 25 The EDG output breaker (3A-14 or 3B-15) will open on any of the following sets of conditions:

Manual Trip (local EDG Panel or RTGB) o Control switch placed to TRIP position Bus 3A(B) bus undervoltage or degraded voltage (3/3 phases) with 3-to-2 bus tie breaker closed (3A(B)-11)

SIAS with 3-to-2 bus tie breaker closed (3A(B)-11)

Protective trip signal from EDG control circuit Local Manual Operation o Manual mechanical trip button depressed Local Test Operation o PERM/PB pushbutton held depressed o Local CS/69 control switch momentarily taken to TRIP Static Exciter and Voltage Regulator (SEVR) System Overview To generate output voltage and hence, electrical power, it is necessary to supply the generator's field with DC current (called "exciting" current). This process is called "excitation" or "exciting the field". During a diesel start, the EDG field is flashed (excited in timed pulses). Flashing is not absolutely necessary since residual magnetism in the field can last as long as a year; however, it is provided to ensure a rapid buildup of EDG terminal output voltage to meet the 10-second availability requirement. As the EDG comes up to rated speed and voltage, the flashing of the field is discontinued, and the field exciting current is then supplied by a device called a "static exciter". It is called "static" because it has no moving or rotating parts in its power section. It derives its power from the generator output through a voltage stepdown control system. Hence, though the exciter can handle a flashless start by utilizing residual magnetism in the field, the field flashing circuit assists in "self-starting" the exciter by expediting the EDG terminal output voltage buildup, after which the exciter has a power supply to tap for its own field exciting current. The static exciter has an advantage over brush and brushless type exciters in that it has a faster response time.

PAGE 40 OF 170

SD-EDG Rev. 25 Each timer is associated with specific dedicated loads or load circuits, which it starts at a predetermined time in the load sequencer's 200-second interval. The sequencer circuit consists of a detection circuit, a test circuit, and a display circuit. The detection circuit senses safety bus undervoltage (UV), safety bus degraded voltage (DV), and/or SIAS conditions and initiates the operation of the load sequencer as appropriate. The test circuit allows sequencer testing during normal plant conditions, and the display circuit indicates the status of sequencer circuit operation.

The test and display circuit controls and indications are on CP-1 in the Main Control Room. These include a momentary-contact, sequencer RESET-TEST switch that spring returns to center position and twelve indicating lights (shown below):

.5 17 168 1 29 200 1.5 41 TEST 7 110 LOCKOUT The ten step timing lights come on when their associated TDPU timer relays energize, the TEST light comes on when the Sequencer RESET-TEST switch is taken to TEST momentarily (spring returns to center position), and the LOCKOUT light comes on if the SLO relay energizes. An actual sequencer actuation signal during sequencer testing will always override the test mode. All the sequencer relays are located in Auxiliary Relay Cabinets 1 and 2.

An undervoltage or degraded voltage on the 4160V bus will cause the associated 3-to-2 bus tie breaker to open, and start the associated EDG. When the EDG reaches rated speed (600 rpm) and voltage (4160V), its output breaker will close and energize the safety bus transformers and loads.

If a 4160V and/or 480V safety bus undervoltage condition occurs without an SIAS, the undervoltage condition will interrupt power to the timers, resetting them. However, the sequencer won't start until power is available to the safety buses (4.16 KV and 480 VAC). Once power is restored to the safety buses from either the EDGs (within 10 seconds) or offsite power, the UV detector circuits will reset and the timers will begin their sequence at that time. If an SIAS occurs without a safety bus undervoltage condition, the timers in the sequencer circuit reset, and since no undervoltage condition exists, the sequencer begins loading the plant loads after two seconds per its program.

The two-second time delay allows time for each relay's field flux to collapse. If SIAS and safety bus undervoltage occur simultaneously, the sequencer will not start until after the EDG energizes the bus approximately 10 seconds after the undervoltage condition initially occurred.

PAGE 49 OF 170

SD-DC Each battery is connected via a two pole, non-automatic battery disconnect to an insulated bus in a metal enclosed load center. The A and B battery disconnects are located in their respective switchgear rooms on the +21 ft elevation of the RAB. The AB disconnect is located in the AB battery room. The TGB battery disconnect is located in an enclosure just south of the battery room.

The TGB Battery Disconnect is equipped with a position indicator that provides indication of actual switch blades position by the color showing in the indicator window.

The color code is as follows:

Green - The switch is open Red - The switch is closed Yellow - WARNING, THE SWITCH IS PARTIALLY OPEN IF ANY YELLOW APPEARS IN THE WINDOW, IMMEDIATE ACTION IS REQUIRED, this indicates that the switch blades are not fully closed (or open) and the bolted pressure mechanism is not fully engaged. This condition may cause over heating or severe damage to occur. A thourough examination of the entire switch should be made as soon as possible.

Each of the two safety-related chargers is connected to the bus through a circuit breaker. Molded case circuit breakers are used for all outgoing feeder circuits in load centers A, B, and TGB. Load center AB employs fused breakers for all feeders.

Each load center is provided with a ground detector and a ground detector relay because the DC systems operate ungrounded. The ground detector is a DC type with a negative 0.5 to 0 to positive-0.5 ma output and a negative 150 to 0 to positive 150 volt voltmeter. A positive indication on the zero-centered voltmeter scale means a ground has occurred on the negative leg of a circuit. Conversely, a negative indication means a ground has occurred on a positive leg of a circuit. Since a second ground on the same train may result in a direct positive to negative short circuit and the loss of the entire DC train on which the short occurred, the initial ground should be sought and isolated, using ground isolation procedures. One ground detector relay is provided for each load center. A bus voltmeter, a battery ammeter, and an undervoltage relay are also provided in each load center. Figure 9 depicts the indications for each load center.

Voltage for the A, B, and AB buses is also indicated on CP-1 in the Control Room and is monitored by the plant computer. There are various annunciators on CP-35 to identify battery, battery charger, or PDP troubles. These are listed in Table 1.

13 of 46

SD-EDG Rev. 25 Undervoltage Override An undervoltage override (UVO) feature prevents cycling of the sequencer due to momentary drops in the safety bus voltage when various safety transformers are being connected to the bus. It is active only if the 3-to-2 bus tie breaker, 3A-11, is open (EDG is only power supply to the bus), and even then, it is only active during the time interval between the 0.5-second and 17-second load blocks. When the sequencer initiates, the UVO relay is energized at step 0 (S0X, 0.5 seconds). It opens contacts to de-energize the 4160V and 480V bus auxiliary relays (27-1X, 27-2X, 27-3X,27-11X, 27-21X, and 27-31X in the figures). This blocks any subsequent undervoltage/degraded voltage actuation signals, in the event bus voltage should again drop below the undervoltage/degraded voltage actuation setpoints during bus loading. It remains energized until step 4 (S4X), 17 seconds after the EDG re-energizes the bus. At this time, the UVO relay de-energizes and no longer prevents subsequent undervoltage/degraded voltage actuation signals.

Sequencer Lockout A sequencer lockout (SLO) feature will terminate automatic loading of the EDG in the event EDG voltage cannot be maintained. This will most likely be due to an equipment malfunction. Like the undervoltage override feature, the sequencer lockout feature is only utilized when the EDG is supplying the bus alone (3-to-2 tie breaker 3A-11 open).

Since the UVO feature blocks undervoltage signals during the time interval between 0.5 to 17 seconds, the SLO circuit is actually only enabled during the interval between 17 and 200 seconds. After step 4 (17 seconds), if 3/3 phases on either a 4.16 KV or 480 VAC safety bus have undervoltage/degraded voltage conditions, then a UV contact will close (27-1 for 4160V; 27-3 for 480V) to energize the SLO relay. When the SLO relay energizes, one SLO contact opens to de-energize (reset) all TDPU timer relays, one closes to turn on the LOCKOUT indicating light on CP-1, and one closes to seal-in the SLO relay through a RESET-TEST switch reset contact. Once the SLO circuit is actuated, it can only be reset manually from CP-1 by momentarily taking the sequencer RESET-TEST switch to the RESET position after the UV condition has cleared. This will drop out the SLO seal-in contact and de-energize the SLO reset to allow sequencer actuator to begin again. Thus, if an undervoltage/degraded voltage condition occurs after the undervoltage override times out, the loads previously loaded onto the bus will be respired (load shed) due to the undervoltage/degraded voltage relays, extinguishing all load block lights, and resetting all timer relays to time 0. The sequencer lockout feature will prevent automatic sequencing from resuming until the Sequencer is manually reset. Faulted load(s) should be removed from the associated 4160V or 480V AC bus prior to resetting the Sequencer.

PAGE 53 OF 170

System Operating Procedure OP-001-001 Reactor Coolant System Fill and Vent Revision 034 CAUTION (1) VENT RIGS FOR RC-101, RC-10111, AND RC-309 SHALL NOT BE FITTED WITH ANY TEMPORARY VALVE OR FLOW BLOCKING OR FLOW LIMITING DEVICE.

(2) VENT RIGS FOR RC-101, RC-10111, AND RC-309 SHALL BE UNOBSTRUCTED AND WILL VENT AND FLOW FREELY.

6.1.22 When a solid stream of water issues from Pressurizer Spray Line Vent RC-309, then Close RC-309.

6.1.23 Prior to exceeding 150 PSIA in the RCS, verify the following valves are Locked Open:

SI-109A LPSI Pump A Suction Valve SI-109B LPSI Pump B Suction Valve 6.1.23.1 Document on Attachment 11.9, RCS Fill and Vent Valve Lineup Restoration.

6.1.24 When RCS pressure reaches approximately 150 PSIA, then secure LPSI fill by performing the following:

6.1.24.1 Close the LPSI Flow Control Valve being used for filling RCS:

SI-138A LPSI Header to RC Loop 2B Flow Control SI-138B LPSI Header to RC Loop 1B Flow Control SI-139A LPSI Header to RC Loop 2A Flow Control SI-139B LPSI Header to RC Loop 1A Flow Control 6.1.24.2 Close LPSI Pump Flow Control Valve Controller, SI-IFIC-306(307).

6.1.24.3 Secure LPSI Pump being used for filling.

6.1.25 Verify SDC Train used for fill is placed in standby in accordance with OP-009-005, Shutdown Cooling System.

6.1.26 If possible, then stabilize RCS solid pressure at approximately 150 PSIA with Backpressure Regulator, CVC-IPIC-0201.

17

Off Normal Procedure OP-901-412 Liquid Waste Discharge High Radiation Revision 002 C AUTOMATIC ACTIONS

1. Liquid Waste to Circulating Water Shutoff Valve (LWM-441) and Liquid Waste to Circulating Water Control Valve (LWM-442) close to terminate Liquid Waste Management discharge.

2. Boron Management Discharge to Circulating Water Auto Isolation Valve (BM-547) and Boron Management Discharge to Circulating Water Flow Control Valve (BM-549) close to terminate Boron Management discharge.

4

SD-CC and return isolation valves are open while ACCW supply and return isolation valves are closed. The switches are normally maintained in the AUTO position.

With the CCW COOLING MODE selector switches in AUTO position, PAC monitors the associated CCW Heat Exchanger outlet temperature. If temperature exceeds 102°F.,

ACCW is automatically aligned to the associated chiller. The transfer from hot to cooler water protects the chiller, which should not be operated above 105°F. This is particularly true during a LOCA, where the CCW temperature might be as high as 115°F. When temperature drops below 95°F., PAC automatically aligns CCW to the associated chiller.

N2 Accumulators 1 and 2 provide backup motive force on a loss of instrument air pressure. EC-41095 was implemented during RF-18 to add Air accumulators to N2 accumulators 1 and 2.

The valves are interlocked such that one set of ACCW valves will not open until the CCW valves for that train are closed. The interlock also works in the opposite direction, preventing the CCW valves from opening until the ACCW valves for that train are closed.

Limit switches for each train are installed to interrupt the actuation of ACC-139A(B) opening until valve ACC-112A(B) is 5% to 10% open. The plate on which the open and closed limit switches are mounted was replaced with a mounting plate for all three limit switches. The valve position indicator plate (valve actuator turning plate) for ACC-112A(B) actuator has a cam which opens the new limit switch within 5% to 10% of valve closure; the cam prevents the limit switch from closing until valve ACC-112A(B) has opened 5% to 10%. The change delays the opening of ACC-139A(B) by less than one second and does not affect closure of ACC-139A(B).

A fire isolation switch (FR-5) is provided for the Train B valves only to permit isolation of circuits that are suspect during fire in the cable vault or control room. Operation of this switch ensures that fire related damage to circuits would not interfere with operation of the B train valves. Fire isolation switch FR-5 is located in the transfer panels in the RAB

+35 ft elevation Relay Room.

Normally, two essential chillers are in operation, one for each safety-related train. The CCW/ACCW trains are maintained split, not cross-connected, at the chillers. Manual valves at the inlet and outlet of each chiller are used to direct CCW/ACCW flow to the two running chillers while maintaining train separation. When Essential Chiller AB is not replacing another chiller, all CCW cross connect valves are closed except the return valves on the side the AB electrical busses are aligned to. If misaligned, the manual CCW cross connect valves at the essential chillers would couple both independent trains of CCW together.

REVISION 22 PAGE 29 OF 123

Off Normal Procedure OP-901-511 Instrument Air Malfunction Revision 015 Page 3 of 4 ATTACHMENT 7: INSTRUMENT AIR VALVES INSIDE CONTAINMENT (CONTD)

WASTE MANAGEMENT Valve Component Failure Effect on System BM-109 RDT OUTLET INSIDE CNTMT ISOL CLOSED RC-323 QUENCH TANK VENT TO CNTMT VENT CLOSED Can not drain tank HDR RC-325 QUENCH TANK DRAIN TO RDT CLOSED SP-105 CNTMT SUMP HDR INSIDE CNTMT ISOL CLOSED GWM-104 CNTMT VENT HDR INSIDE CNTMT ISOL CLOSED Can not vent GWM-101 RDT VENT TO CNTMT VENT HDR OPEN Can not vent NG-515 NITROGEN REGULATOR TO RDT CLOSED Can not fill with N2 HVAC Valve Component Failure Effect on System CAP-104 CNTMT PURGE INLET INSIDE CNTMT ISOL CLOSED CAP-201 CNTMT PURGE REFUELING EXHAUST CLOSED Can not DAMPER CAP-202 CNTMT PURGE NORMAL EXHAUST OPEN purge DAMPER CAP-203 CNTMT PURGE EXHAUST INSIDE CNTMT CLOSED ISOL CAR-200B CARS CNTMT PRESS EXHAUST INLET CLOSED Can not run CARS ISOL CCS-102A CNTMT COOLING HVAC SAFETY DISCH OPEN Cntmt Ventilation DPR A CCS-102B CNTMT COOLING HVAC SAFETY DISCH OPEN Ring Hdr DPR B bypassed CC-646 CEDM COOLING COILS CCW INLET HDR OPEN No effect ISOL 31

SD-CB Rev. 11 openings. The inside radius of the hemispherical dome is 70 ft 15/32 in. with a dome plate of 0.95 in. thick connected to the cylindrical portion of the shell at the tangent line by means of full penetration weld. The containment spray piping is attached to the dome by means of welded clips as in the dome inspection walkway and platforms. The Containment Vessel is protected from external missiles by the Shield Building.

Protection from internal missiles is provided by the primary and secondary shield walls and other containment internal structures.

CONTAINMENT VACUUM RELIEF SYSTEM The containment vessel is designed for an external pressure differential of 0.65 psig (approximately 18 in. WG) at 120°F with the containment being at the lowest pressure.

During normal operation, the containment vessel is vented as required to maintain it at atmospheric pressure. The shield building annulus is maintained at a negative pressure by the annulus negative pressure system. However, an inadvertent actuation of the Containment Spray system can cool the containment atmosphere to the extent of lowering containment pressure. This could create a differential pressure between the containment and the shield building which is higher than the design pressure difference.

The Containment Vacuum Relief system (Figure 2) protects the containment vessel by maintaining the pressure differential across the vessel lower than the design value.

The system consists of two redundant penetrations that connect the shield building annulus with the containment atmosphere. The penetrations provide a flow path for air to pass from the annulus to the containment. Each flowpath is provided with its own set of isolation valves which remain closed during normal operation. In the event that the air pressure inside containment drops below that of the annulus, system instrumentation opens the automatic isolation valves (CVR-101 and CVR-201) which allow air pressure in the annulus to bleed into the containment.

Vacuum Relief Automatic Isolation Valves, (CVR-101 and CVR-201)

Each of the redundant trains making up the Containment Vacuum Relief system is functionally independent from the other. Pneumatic operated, fail closed, butterfly valves (CVR-101 and CVR-201) are installed on the annulus side of the containment penetrations These valves serve as both automatic relief valves and containment isolation valves. The valves are provided with backup air accumulators that will allow valve operation in the event of a loss of instrument air. CVR-101 and CVR-201 solenoids receive power from 120 VAC PDPs 90A and 91B, respectively. CVR-101 and CVR-201 must be manually closed after automatic opening.

SD-CB Rev. 11 The butterfly valves are opened automatically in response to signals from the following pressure switches:

At a DP of 8.5 in. of water, CVR-IDPIS-5221AS and BS provide the signals to open CVR-101 and CVR-201 respectively. Reset 5.5 inwc CVR-IDPIS-5220AS and BS provide a backup signal to open the valves at 8.5 in. of water DP. The backup signal resets at 5.5 in of water decreasing.

The Train A instruments provide a signal to CVR-201 while the Train B instruments serve CVR-101.

CVR-101 and CVR-201 can be operated by control switches on CP-18 in the main control room. The control switches have positions of CLOSE/OPEN and are spring return to normal. Valve position indication is provided on the control switch for each valve and on the plant monitoring computer.

Vacuum Relief Check Valves (CVR-102 and CVR-202)

Check valves are installed on the containment side of the penetrations and serve as containment isolation valves. The valves have magnetic latches that hold the valve in the close position but the valves will open quickly against a small differential pressure.

INTERNAL STRUCTURES The steel containment encloses several structures (Figure 3) and structural components which comprise the internal structures of the containment. The main components that comprise the internal structures are as follows:

Primary Shield Wall Secondary Shield Wall Refueling Canal Pressurizer Enclosure Regenerative Heat Exchanger Enclosure Reactor Vessel Support Steam Generator Supports Reactor Coolant Pump Supports

9.10.2 SHORT TERM POWER REDUCTION CONTROL OF ASI CAUTION (1) TO PREVENT EXCEEDING THE TRANSIENT INSERTION LIMITS OF TECHNICAL SPECIFICATION 3.1.3.6 AND THE SHUTDOWN MARGIN REQUIREMENTS OF TECHNICAL SPECIFICATION 3.1.1.1, REGULATING GROUP 5 CEAS SHALL NOT BE INSERTED TO LESS THAN 145 INCHES WITHDRAWN FOR ASI CONTROL ABOVE 80% POWER.

(2) BELOW 80% POWER, REGULATING GROUP 5 CEAS MAY BE INSERTED WITHIN THE BOUNDS OF THE TRANSIENT INSERTION LIMIT AS LONG AS REGULATING GROUP 6 CEAS ARE INSERTED FIRST, AND MAINTAINED AT LEAST 15 INCHES BELOW REGULATING GROUP 5 CEAS.

(3) AT ANY TIME WITH REACTOR POWER 20% AND REGULATING GROUP 6 OR GROUP P CEAS ARE <120 INCHES WITHDRAWN OR REGULATING GROUP 5 CEAS ARE <145 INCHES WITHDRAWN FOR ASI CONTROL, THEN THE AMOUNT OF TIME SHALL BE LOGGED IN ACCORDANCE WITH OP-903-001, TECHNICAL SPECIFICATION SURVEILLANCE LOGS.

9.10.2.1 Power reduction should be initiated by boration. As power begins to lower, ASI should move in the negative direction. Insert CEAs to maintain the ASI in a small band about the target ESI as recommended by Reactor Engineering.

9.10.2.2 Continue to maintain ASI at target ESI 0.05. Dilution or TAVG deviations may be required in response to CEA insertion, or as a result of Xenon buildup to reduce the rate of power reduction, or to level power at the final desired level.

9.10.2.3 When final power level is obtained following a rapid power reduction using CEAs, then the CEAs should be borated out and used to control ASI at the target ESI 0.05.

9.10.2.4 If power operation is to continue for >72 hours at a reduced power level, then Reactor Engineering should be consulted to determine if it is desirable to return to an ARO configuration. If this is recommended, then a target ESI for the reduced power level must be recommended by Reactor Engineering.

9.10.2.5 When a return to the pre-reduction power level is desired, then initiate the power increase by dilution. Withdraw CEAs to maintain ASI as follows:

If power was maintained below 75% for greater than 27 days, then maintain ASI at the target ESI +/-0.03 during power ascension above 50%.

If power was maintained above 75% or below 75% for less than 27 days, then maintain ASI at the target ESI +/-0.05 during power escalation above 50%.

9.10.2.6 Contact Reactor Engineering if problems are encountered.

OP-010-005 Revision 328 Attachment 9.10 (3 of 3) 84

Off Normal Procedure OP-901-130 Reactor Coolant Pump Malfunction Revision 011 Alarms (contd):

RCP 1B CCW FLOW LOST (Cabinet SA, A-2), (Cabinet SB, A-7)

RCP 2A CCW FLOW LOST (Cabinet SA, A-3), (Cabinet SB, A-8)

RCP 2B CCW FLOW LOST (Cabinet SA, A-4), (Cabinet SB, A-9)

RCP CONTL BLEEDOFF HEADER PRESS HI-HI (Cabinet G, M-5)

RCP CONTL BLEEDOFF HEADER PRESSURE HI (Cabinet G, N-5)

2. Indications:

RCP Controlled Bleedoff temperature rising RCP Controlled Bleedoff flow rising RCP Bearing temperature rising Seal Water Cooler CCW Outlet temperature rising RCP Motor Lube Oil Reservoir Cooler(s) CCW return temperature rising RCP Motor Lube Oil Reservoir(s) temperature rising RCP Motor Oil Cooler(s) differential temperature rising RCP Oil Reservoir level dropping RCP Seal pressures outside of normal parameters RCP Anti-Reverse Rotation Device temperature(s) rising PMC PID A13003, RCP-1A X-PLANE PUMP SHAFT PMC PID A13004, RCP-1A Y-PLANE PUMP SHAFT PMC PID A13010, RCP-2A X-PLANE PUMP SHAFT PMC PID A13011, RCP-2A Y-PLANE PUMP SHAFT PMC PID A13017, RCP-1B X-PLANE PUMP SHAFT PMC PID A13018, RCP-1B Y-PLANE PUMP SHAFT PMC PID A13024, RCP-2B X-PLANE PUMP SHAFT PMC PID A13025, RCP-2B Y-PLANE PUMP SHAFT 4

Off Normal Procedure OP-901-130 Reactor Coolant Pump Malfunction Revision 011 E1 SEAL FAILURE (CONTD)

PLACEKEEPER START DONE N/A CAUTION (1) CCW TEMPERATURES OF <75 F COULD LEAD TO ESSENTIAL CHILLER TRIPS ON EVAPORATOR LOW REFRIGERANT PRESSURE.

(2) CCW TEMPERATURE SHOULD BE CHANGED AT A RATE OF 10 F IN ONE HOUR TO PREVENT DEGREDATION OF THE REACTOR COOLANT PUMP SEALS.

2. If Controlled Bleedoff temperature is rising, then lower Component Cooling Water temperature by any of the following:

Start Dry Cooling Tower Fans.

Start Auxiliary Component Cooling Water Pump(s) and associated Wet Cooling Tower Fans.

Start Auxiliary Component Cooling Water Pump(s) and lower ACC-126A(B) setpoint.

3. If two or more seals fail in rapid succession, (within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) then perform the following:

3.1 Trip the Reactor.

3.2 Secure affected Reactor Coolant Pump.

3.3 Go to OP-902-000, Standard Post Trip Actions.

4. If a second seal fails on the same Reactor Coolant Pump, then commence a controlled Plant shutdown in accordance with OP-010-005, Plant Shutdown.

10

(Initial/Date) 9.1.42 When Reactor Power drops below 10-4%, then perform the following:

NOTE

1) The LOCAL POWER DENSITY/DNBR BY-PASS Annunciator (D-11 Cabinet K) will alarm when all CPCs are in Bypass.

2) Reactor Coolant Loop 2 Hot Leg Temperature Indicators, RC ITI 0102HA (HB), will fail HI once out of range.

9.1.42.1 Bypass the CPC Trips by positioning each CPC Trip /

Bypass keyswitch to On.

9.1.42.1.1 Check the On light illuminates for each CPC. /

9.1.42.2 Verify the High Log Power Trip is Enabled on all /

four PPS channels.

9.1.42.2.1 Check the Off light illuminates for each PPS ROM. /

CAUTION 3 OUT OF 4 HI LOG POWER CHANNELS MUST BE OPERABLE PRIOR TO CLOSING REACTOR TRIP BREAKER WITH MG SETS OPERATING.

9.1.43 Prior to closing Reactor Trip Breakers with MG sets operating, /

verify the surveillances on Attachment 9.14 are current.

[P-3016]

9.1.44 Verify at 10-6% power Startup (S/U) Channels Energize and /

read 104 counts per second.

9.1.45 Between 30 to 60 minutes after Reactor Shutdown, verify /

Boron Dilution Monitors setpoint is adjusted in accordance with the COLR and continue to adjust at the frequency specified by the COLR. [T.S. 4.1.2.9.5]

9.1.45.1 Document setpoint verifications in accordance with OP-903-001,Technical Specification Surveillance Logs.

9.1.46 Verify OP-903-101, Startup Channel Functional Test Startup /

Channel __1 and __2, has been performed within the previous 31 days. [T.S. 4.1.2.9.2]

OP-010-005 Revision 328 Attachment 9.1 (12 of 14) 33

Operating Instruction OI-038-000 Emergency Operating Procedures Revision 013 Operations Expectations / Guidance 5.2.3 (continued)

The CRS should ask the NPO the status of Maintenance of Vital Auxiliaries on the substep level. The NPO will report the status of Maintenance of Vital Auxiliaries to the CRS on the substep level.

The CRS should ensure the RO has reported the status of the Maintenance of Vital Auxiliaries acceptance criteria. The CRS may prompt the operator to obtain the necessary information.

5.2.4 Verify RCS Inventory Control HPSI pumps and charging pumps may be secured as necessary in SPTA and RTR if the HPSI Throttle Criteria are met and there is a probability of the Pressurizer going solid.

The NPO checks Pressurizer level indicators. The NPO should check charging pump breaker indications, charging pump flow and letdown indications. The number of charging pumps in service is dependent on Pressurizer level, electrical power, and if an SIAS has occurred. The NPO should verify that charging flow is consistent with the number of the charging pumps in service and that in most instances letdown flow is at minimum.

RCS subcooling is checked using the following criteria. Use TH subcooling for forced circulation (CP-2 & CP-8) and CET subcooling for natural circulation (QSPDS and CP-7). The CRS should also note that if RCS subcooling is less than the minimum required, the minimum required RCP NPSH requirement may be exceeded and the RCPs should be stopped.

The NPO should first attempt to restore Pressurizer level if it is outside the expected post trip band by using the PLCS. This requires the NPO to place the Pzr Level controller (CP-2) in manual and adjust controller output.

The NPO may start and stop charging pumps as necessary to control Pressurizer level if the PLCS is malfunctioning. The NPO may operate the Letdown or Backpressure flow controllers to gain control of letdown flow.

The CRS may also elect to isolate Letdown to minimize the inventory loss from the RCS.

13

Off Normal Procedure OP-901-513 Spent Fuel Pool Cooling Malfunction Revision 020 C AUTOMATIC ACTIONS

1. Fuel Pool Pumps trip requires both Low Spent Fuel Pool level of 439 MSL (FS-ILS-2000-A1) and Low-Low Spent Fuel Pool level of 41'6" MSL (FS-ILS-2000-A2).

4

Off Normal Procedure OP-901-513 Spent Fuel Pool Cooling Malfunction Revision 020 D IMMEDIATE ACTIONS NONE 5

Off Normal Procedure OP-901-513 Spent Fuel Pool Cooling Malfunction Revision 020 E SUBSEQUENT OPERATOR ACTIONS E0 GENERAL PLACEKEEPER START DONE N/A

1. If Spent Fuel Pool Cooling malfunction is due to a CCW System Malfunction, then perform OP-901-510, Component Cooling Water System Malfunction concurrently with this procedure.

2. If Spent Fuel Pool level < 43'9" MSL (low level alarm setpoint), then perform the following:

2.1 Makeup to Spent Fuel Pool (SFP) from Refueling Water Storage Pool or Condensate Storage Pool in accordance with OP-002-006, Fuel Pool Cooling and Purification.

2.2 If normal means of makeup to the SFP are unavailable, then makeup to the SFP from the Fire Protection System per Attachment 1, Fire Protection Makeup to the Spent Fuel Pool.

2.3 If none of the above methods of SFP makeup are available, then implement the SFP makeup guidance in S-SAMG-01, Loss of Large Areas of the Plant Due to Fire/Explosion, Section 7.0, Spent Fuel Pool Actions.

2.4 If a Spent Fuel Pool leak is indicated by abnormal Spent Fuel Pool level loss or frequent makeup, then attempt to locate source of leakage by verifying the following:

2.4.1 Perform a walkdown of SFP purification piping and monitor SFP level for signs of leakage.

2.4.2 Secure Spent Fuel Pool Purification in accordance with OP-002-006, Fuel Pool Cooling and Purification.

6

SD-SBC There are three lights for each bypass valve on CP-1. The signal to operate the lights comes from the valve control circuits. The lights provide valve position (valve not fully open, valve not fully closed, Permissive status, and Q.O. indication.

INTERLOCKS Quick Opening Block (Figure 21)

Following a reactor trip from a low or medium reactor power level (Tave <561°F), it is not desirable to allow a "quick opening" of the steam bypass valves, because the NSSS has the additional heat storage to accommodate decay and reactor coolant pump heat.

The modulation speed of the valves minimizes pressure and temperature swings and provides a smoother transition to hot zero-power conditions.

The "quick opening block" signal generation is based on the receipt of Control Element Drive Mechanism Control System (CEDMCS) undervoltage signals transmitted from the Feedwater Control System (FWCS) or a Loss of Feed Pump condition transmitted from RXC. Two coincident reactor tripped signals (reactor tripped 1 and 3 or 2 and 4) are required.

All steam bypass valves are blocked from quick opening any time the SBCS senses a Loss of Feed Pump condition (one or both feed pumps). Turbine bypass valve 6 is blocked from quick opening any time that the SBCS senses a reactor trip. This is done because the flow capacity of five valves is the maximum required for all transients and accidents analyzed that result in a reactor trip. Steam bypass valves 1-5 are blocked from quick opening following a reactor trip with average reactor coolant temperature (TAVG) below 561 F. TAVG is input to the SBCS from the active reactor regulating system.

Automatic Withdrawal Prohibit (Figure 22)

The SBCS generates an "automatic withdrawal prohibit" whenever a demand signal exists for opening a turbine bypass valve, if the SBCS is in "operate" and not in "emergency off". The objective of this AWP is to prevent a reactor trip from high Pressurizer pressure due to an uncontrolled automatic control element assembly withdrawal.

The AWP signal is sent to CEDMCS to block its response to the reactor regulating system's demand to withdraw CEAs. The AWP is generated when any one of the following conditions exist:

"Master controller demand" and "automatic modulation permissive".

Group quick opening demands "X1" and "X2".

Revision 10 Page 25 of 68

SD-SBC The AMI threshold controller (MS-IHIC-1010) is used to manually set the AMI threshold used in the SBCS.

Output Adjust Knob - This is a ten-turn potentiometer with a position indicating dial.

Scale is 0-100%.

Output Meter - The movable pointer displays the AMI threshold setpoint being used in the SBCS. This displayed setpoint is the lower of:

The setpoint from MS-IHIC-1010 as shown on the output adjust dial.

The sum of SBCS available valve capacity and turbine first stage pressure.

Permissive Switches Each Bypass valve has a permissive switch associated with its operation. Each switch is a three position, rotary snap switch. The positions are labeled OFF/AUTO/MANUAL.

OFF - a permissive signal cannot be sent to the valve control circuit, thus the valve will not operate.

MANUAL - the permissive signal will be sent to the valve control circuit to enable valve operation unless Emergency Off is initiated or a Condenser Interlock is present. This switch position is used for manual control of the bypass valves.

AUTO - the permissive signal is sent to the valve control circuit as determined by the SBCS logic.

Valve Interfacing Relays Since the permissive and Q.O. signals leaving the SBCS Cabinet Assembly are 120 VAC, and the bypass valve control circuits require 125 VDC signals, interfacing relays are required. Upon receiving a 120 VAC signal from the cabinet, the respective relay energizes to close contacts which send a 125 VDC signal to the valve control circuit.

Condenser Interlock Relay The Condenser Interlock relay located remotely is normally energized from the SBCS Cabinet Assembly. When a Condenser Interlock signal is produced, the relay is de-energized causing contacts to open in the modulation and Q.O. paths to each turbine bypass valve control circuit.

CP-1 Indicating Lights (Figure 19)

Revision 10 Page 24 of 68

Off Normal Procedure OP-901-221 Secondary System Transient Revision 004 E SUBSEQUENT OPERATOR ACTIONS E0 GENERAL PLACEKEEPER START DONE N/A NOTE (1) Some steps of this procedure may not be applicable due to plant conditions. In these cases SM/CRS may NA the step.

(2) Steps within this procedure may be performed concurrently or out of sequence with SM/CRS concurrence.

Continuous

1. If Reactor trip occurs, then go to OP-902-000, Standard Post Trip Actions.

Continuous

2. If Reactor Power Cutback occurs, then perform OP-901-101, Reactor Power Cutback, concurrently with this procedure.

Continuous

3. If an Atmospheric Dump Valve fails or begins to fail Open, then place the respective controller to MANUAL with minimum output.

Continuous

4. If a Steam Bypass Valve fails or begins to fail Open, then perform any of the following (in preferred order) to close the valve.

Place the respective Valve Mode Select switch to OFF.

Place the respective valve controller to MANUAL with minimum output.

6

Off Normal Procedure OP-901-221 Secondary System Transient Revision 004 E0 GENERAL (CONTD)

PLACEKEEPER START DONE N/A

6. If Main Turbine is available, then adjust Turbine load as necessary to maintain the following:

Reactor Power 100%

Match Tavg with Tref FWPT Suction Pressure > 300 PSIG (monitored on CP-1 via CD IPI1280, IP Htrs Outlet Hdr)

RCS Tcold 536F - 549F

7. If needed, then concurrently perform OP-901-212, Rapid Plant Down Power, until a power level is reached in which the plant can be stabilized.

8. If a loss of Feedwater preheating occurs, then go to E1, Loss of Feedwater Preheating.

9. If a loss of secondary inventory occurs (Condensate, Feedwater or Steam leak), then go to E2, Loss of Secondary Inventory.

10. If a Steam Bypass Control Valve has failed open, then manually isolate the affected valve and determine impact on the Reactor Power Cutback System in accordance with OP-004-015, Reactor Power Cutback System.

END 9

System Operating Procedure OP-003-001 Condenser Air Evacuation System Revision 020 NOTE Energizing an AE Pump automatically starts its Seal Water Recirculation Pump.

6.4.2.2 Check the Seal Water Pump for the Condenser Air Evacuation Pump started is operating.

NOTE Condenser Vacuum Pump B or C must be left operating to provide a suction path for the Condenser Air Evacuation Pump Discharge radiation monitor, PRM-IRE-0004.

6.4.3 Perform the following for the AE Pump being Stopped:

6.4.3.1 If applicable, Close the following:

PRM-029-002 AE B Sample Supply Inlet Isolation Valve PRM-029-001 AE C Sample Supply Inlet Isolation Valve 6.4.3.2 Stop AE Pump A(B)(C) by taking its control switch to STOP.

6.4.3.3 Close the following valves:

CMU-7121A(B)(C) Cond Air Evacuation Drain Line Strainer Inlet Isolation CMU-7122A(B)(C) Cond Air Evacuation Drain Line Strainer Outlet Isolation 15

System Operating Procedure OP-003-001 Condenser Air Evacuation System Revision 020 9.0 AUTOMATIC FUNCTIONS 9.1 Transfer Condenser Vacuum Pumps A, B, and C from Hogging mode to Hold mode at vacuum rising.

AE-IPS-1941A(B)(C). ................................................................... 25" Hg (23-24" Hg) 9.2 Standby Condenser Vacuum Pumps Auto-Start at vacuum lowering. AE-IPS-1940A(B)(C). ................................................... 26" Hg 20

Operating Procedure OP-007-003 Gaseous Waste Management Revision 307 CAUTION

1. IF DISCHARGE ACTIVITY EXCEEDS THE GASEOUS RELEASE PERMIT SETPOINT THEN OP-901-413, WASTE GAS DISCHARGE HIGH RADIATION, SHALL BE ENTERED.

2. AT LEAST ONE RAB EXHAUST FAN SHALL BE OPERATING WHILE DISCHARGING.

3. THE RELEASE SHOULD BE TERMINATED IF METEOROLOGICAL CONDITIONS ARE OUTSIDE THE PERMISSIBLE LIMITS.

4. THE RELEASE SHALL BE TERMINATED IF THE IN-SERVICE GDT PRESSURE BEGINS TO DECREASE UNTIL ADDITIONS TO THE GAS SURGE HEADER HAVE BEEN SECURED. IF NO NEW GASES HAVE BEEN INTRODUCED AND A WR HAS BEEN WRITTEN ON THE LEAKING INLET VALVE, THEN DISCHARGING CAN CONTINUE.

5. IF DISCHARGING ALL THREE GDTs SIMULTANEOUSLY, THEN ADDITIONS TO THE GAS SURGE TANK FROM THE VCT, GAS SURGE HEADER OR THE CONTAINMENT VENT HEADER SHOULD NOT BE MADE DURING GDT RELEASES DUE TO POTENTIAL LEAKAGE OF GDT INLET VALVES. DISCHARGING ALL GDTS SIMULTANEOUSLY IS THE PREFERRED METHOD. [CR-98-1291]

6.4.9 If the Gaseous Waste Discharge Radiation Monitor, PRM-IRE-0648, is operable and sample flow has risen to >2 scfm as seen locally and documented on Attachment 11.4, then continue to throttle Open Waste Gas Discharge GWM-IFIT-0648 Outlet Isolation Valve, GWM-311, to establish desired flow within limit indicated on Gaseous Release Permit.

6.4.10 If Gaseous Waste Discharge Radiation Monitor, PRM-IRE-0648, is not operable, then commence discharging by Throttling Open Waste Gas Discharge GWM-IFIT-0648 Outlet Isolation Valve, GWM-311, to establish the desired flow rate within the limit indicated on the Gaseous Release Permit.

6.4.11 Record the 0-hour data readings on the Gaseous Release Permit.

14

11.1 CIRCULATING WATER SYSTEM STANDBY VALVE LINEUP COMPONENT COMPONENT DESCRIPTION LOCATION REQUIRED PERFORMED BY NUMBER POSITION (INITIAL/DATE)

CIRC WATER PUMP A DISCH CW-101A INTAKE STRUCT OPEN CW IPS8811A ROOT CIRC WATER PUMP B DISCH CW-101B INTAKE STRUCT OPEN CW IPS8811B ROOT CIRC WATER PUMP C DISCH CW-101C INTAKE STRUCT OPEN CW IPS8811C ROOT CIRC WATER PUMP D DISCH CW-101D INTAKE STRUCT OPEN CW IPS8811D ROOT CIRC WATER PUMP A DISCHARGE CW-102A INTAKE STRUCT CLOSED PX ROOT CIRC WATER PUMP B DISCHARGE CW-102B INTAKE STRUCT CLOSED PX ROOT CIRC WATER PUMP C DISCHARGE CW-102C INTAKE STRUCT CLOSED PX ROOT CIRC WATER PUMP D DISCHARGE CW-102D INTAKE STRUCT CLOSED PX ROOT CIRC WATER PUMP A DISCHARGE CW-103A INTAKE STRUCT CLOSED ISOLATION CIRC WATER PUMP B DISCHARGE CW-103B INTAKE STRUCT CLOSED ISOLATION CIRC WATER PUMP C DISCHARGE CW-103C INTAKE STRUCT CLOSED ISOLATION CIRC WATER PUMP D DISCHARGE CW-103D INTAKE STRUCT CLOSED ISOLATION RIVER WATER SUPPLY PUMP CW-1041 INTAKE STRUCT CLOSED DISCHARGE TEST CONN ISOL RIVER WATER SUPPLY PUMP CW-105 INTAKE STRUCT OPEN DISCHARGE ISOLATION RIVER WATER SUPPLY PUMP CW-106 INTAKE STRUCT OPEN DISCH CW IPI9112 RT OP-003-006 Revision 320 Attachment 11.1 (1 of 11) 105

SD-CW 13 demonstrate the CW Pump start logic and the CW Pump Discharge Valve opening logic.

A time delay prevents the starting of the second, third and fourth pumps until the previous pump's discharge valve has been fully opened (or fully closed, in the event that the associated pump failed to start) for 100 seconds. The other CW Pumps are also started by turning their respective control switches to START. These pumps discharge valves open to the full open position without stopping at 20%; however, the pump will still start when the valve reaches its 20% open position. After another 100 second time delay the next CW Pump may be started. During normal operation, two, three or all four CW Pumps may be operating, depending on conditions.

To stop a CW Pump during three or four pump operation, the operator turns the control switch to STOP. The pump discharge valve begins to close. The associated pump trips when the valve passes through the 90% open position. A delay of 100 seconds, after the valve is closed, is required before the next pump can be stopped. If the pump fails to trip when its discharge valve reaches the 90% open position, the valve will reopen, and the trip cannot be attempted again remotely until after a 100 second time delay.

To stop the last two pumps, either pump control switch may be turned to STOP. This causes both CW Pump motor circuit breakers to trip. Both pumps discharge valves will remain open for 5 minutes after their pumps have stopped. After 5 minutes they will automatically close sequentially. Figures 14 and 15 demonstrate the CW Pump stop logic and the CW Pump Discharge Valve closing logic.

Precautions are taken to prevent hydraulic shock in the event that the CW Pumps trip abnormally. For example, all of the pumps will stop in the following events:

Four pumps are operating and three trip.

Three pumps are operating and two trip.

This precaution is also taken during system shutdown. The last two pumps will always shut down simultaneously when the operator turns any one control switch to STOP.

It should be noted however, if only two pumps are operating and one pump trips or if four pumps are operating and two trip, from an event that is not preceded by initial closure of its respective discharge valve (i.e., motor fault or loss of the bus); the remaining pump will not trip. The vacuum breakers in this case will open to prevent a water hammer and maintain system stability and remain open for 100 seconds after the last discharge valve is closed. Condenser flow will decrease, but will not stop or reverse direction.

The controls for each discharge valve are powered by a 120 volt ac control transformer located in the motor starter. The master logic circuit is powered by 120 VAC REVISION 17 PAGE 16 OF 59

6.4 CONTROL ROOM TURNOVER SHEET AND CHECKLIST (Typical)

Date: ____/____/________

Prior to turnover, review the Station Log (since last shift or two weeks minimum), and Critical Parameters in allowable limits (Modes 1 and 2):

Pressurizer Level 33-56%

Pressurizer Pressure 2125-2275 psia Steam Generator Level 60-70%

Steam Generator 800-970 psia Pressure Tavg 541-575°F Reactor Power 100% (Mode 2 5%)

As soon as possible after turnover review the following:

Parameter: Positions required to review:

Daily Instructions NPO, CRS, SM Waterford 3 Watch Station Deficiency Database [P-23974] NPO, CRS, SM Equipment Out of Service Log NPO, CRS, SM ODMI Index NPO, CRS, SM Control Room Cleanliness CRS NAO Turnover Sheets CRS Clearance Logs / Active Tagouts CRS, SM Weekly Surveillance Schedule CRS, SM TAR Log SM Key Log SM RCS Perturbation Log Index (Applicable in Modes 5 and 6) SM Containment Impairment Log SM EOOS (Risk assessment program) Verify current plant SM safety index and current equipment removed from service.

Perform a Control Board Walkdown in accordance with EN-OP-115. [P-24954]

Operations Spotlight Issue:

OI-042-000 Revision 045 Attachment 6.4 (1 of 6) 72

H-8 H0808 CWD 112 4.78 POWER DEPENDENT INSERTION LIMIT (H-8) REV 026 INITIATING DEVICE SETPOINT Plant Computer Derived from the following:

PID C24567, Biased Plant Power (PPMAX)

CEA Pulse Counters POSSIBLE EFFECTS

1. Power Dependent Insertion Limit alarm indicates Transient Insertion Limit being exceeded. [T.S. 3.1.3.6]

2. Insertion of subgroup below Power-Dependent Insertion Limit means that there may be inadequate Shutdown Margin.

CONTROL ROOM INDICATIONS LOCAL INDICATIONS CEA CRT NONE CEAC 1 CEAC 2 Following PMC Computer Points:

R35105 REG GRP 1 PDIL R35205 REG GRP 2 PDIL R35305 REG GRP 3 PDIL R35405 REG GRP 4 PDIL R35505 REG GRP 5 PDIL R35605 REG GRP 6 PDIL R35705 GROUP P PDIL OP-500-008 Revision 028 Attachment 4.78 (1 of 2)

H-8 (cont'd)

POWER DEPENDENT INSERTION LIMIT POSSIBLE CAUSES RECOMMENDED ACTIONS

1. One or more CEA Regulating subgroups 1.1 Refer to Technical Specification. 3.1.3.6.

or Group P inserted below the Transient Insertion Limit NOTE Technical Specification 4.1.3.6 requires CEA group positions verification every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2. PDIL alarm inoperable 2.1 Refer to OP-901-501, PMC or Core Operating Limit Supervisory System Inoperable.

2.2 Contact Electrical Maintenance.

3. PMC/MUX hardware or software 3.1 Refer to OP-901-501, PMC or Core problem Operating Limit Supervisory System Inoperable.

3.2 Direct CS&S Maintenance Group to repair failed equipment..

4. Possible dilution occurring 4.1 Inform SM/CRS.

Secure any dilution in progress.

4.3 Refer to OP-901-104, Inadvertent Positive Reactivity Addition.

5. Failed instrument 5.1 Direct PMI to repair failed equipment.

6. CEA/CEDMCS malfunction 6.1 Go to OP-901-102, CEA or CEMCS malfunction.

OP-500-008 Revision 028 Attachment 4.78 (2 of 2)

K-8 H0908 CWD 2749 CEA DISABLED INITIATING DEVICE SETPOINT CEA Circuit Breakers Breaker Off POSSIBLE EFFECTS

1. Dropped CEA

2. Possible Reactor trip CONTROL ROOM INDICATIONS LOCAL INDICATIONS D36604 CEDMCS - CEA Disabled CEA Disabled lights on CEDMC Cabinet C-4, Bays A and B CEAC 1 & 2 position indication CEA CRT Rod bottom lights POSSIBLE CAUSES RECOMMENDED ACTIONS

1. CEA/PLCEA Ckt Bkr placed 1.1 Go to OP-901-102, CEA or in Off CEDMCS Malfunction.

OP-500-008 Revision 026 Attachment 4.88 (1 of 1)

L-9 H1009 CWD 2733 CEDMCS MAINTENANCE ERROR INITIATING DEVICE SETPOINT Subgroup Maintenance toggle switch N/A on CEDMCS Maintenance and Supply lamp panel.

POSSIBLE EFFECTS NONE CONTROL ROOM INDICATIONS LOCAL INDICATIONS NONE Maintenance Error red lamp indicator on Maintenance and Supply Lamp panel.

POSSIBLE CAUSES RECOMMENDED ACTIONS NOTE (1) Only one subgroup may be assigned to Hold Bus at a time. If more than one is selected, subsequent subgroup(s) actuate a Maintenance Error alarm through SGRI Annunciator Relay Card#2.

(2) Subgroup on Hold Bus will have a red light above its toggle switch. The others will not.

(3) Second subgroup will remain on the Active (motive) Bus and will move on any action demand.

1. Attempt made to assign more than 1.1 Remove all but one subgroup one subgroup to Hold Bus from Hold Bus.

OP-500-008 Revision 026 Attachment 4.99 (1 of 1)

L-10 H1010 CWD 2733 CEDMCS TIMER FAILURE NOTE This alarm has two modes of logic when alarming:

Flash -Annunciator (L-10 on CP-2) will alarm then Flash.

Locked in - Annunciator (L-10 on CP-2) will alarm and remain locked in and not clear.

INITIATING DEVICE SETPOINT ACTM Microprocessor Card ACTM alarm output Counter did not Reset in five (5) seconds POSSIBLE EFFECTS

1. If high voltage is not removed, then coils will burn open within 20-30 minutes and may result in a dropped CEA.

CONTROL ROOM INDICATIONS LOCAL INDICATIONS D36603 CEDMCS - Timing Logic Failure Timer Failure lamp illuminated.

Card Status lamps illuminated (Logic Housing)

ACTM Card lamps illuminated Timer Failure lamps illuminated (Supervisory Panel)

OP-500-008 Revision 026 Attachment 4.100 (1 of 2)

REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE and operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room.

APPLICABILITY: MODE 6.

ACTION:

a. With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of Technical Specification 3.9.1.

b. With both of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL CHECK in accordance with the Surveillance Frequency Control Program,

b. A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initial start of CORE ALTERATIONS, and

c. A CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program.

WATERFORD - UNIT 3 3/4 9-2 AMENDMENT NO. 185, 249

Operating Instruction OI-019-000 Operations Procedure Administration Revision 306 5.1.1.6.2 Submit the PIR, along with any supporting documentation, to immediate supervisor for review.

5.1.2 The Requestor's immediate supervisor performs the following:

5.1.2.1 Verify the following:

The PIR is complete.

The PIR is accurate.

Proper justification is provided.

5.1.2.1.1 If the PIR is incomplete, inaccurate, or not properly justified, then do one of the following:

Void the PIR and exit the Procedure Improvement process, or Resolve inadequacies with the Requestor.

5.1.2.1.2 If the PIR is complete, accurate, and properly justified, then approve the PIR and submit the PIR, along with any supporting documentation, to OPAG as follows:

5.1.2.1.2.1 Check all three corresponding check boxes for (1) completeness and accuracy, (2) required documents attached and (3) proper justification provided.

5.1.2.1.2.2 Click the SUPERVISOR APPROVAL button.

5.1.2.1.2.3 In the SUPERVISOR Approval window select your user name as the Supervisor, enter your Password, then click the Approve button.

5.1.3 The OAC performs the following:

NOTE Incomplete or incorrect PIRs may be resolved informally to correct deficiencies prior to rejecting the PIR. Also, rejected PIRs (that have been converted to inactive PAIs) may be subsequently approved (activated) if deficiencies have been corrected. Step 5.1.3.1.2 may be performed in either case to activate the PAI.

5.1.3.1 If the PIR is not valid, is incomplete, or is adequately addressed by an existing PAI or by a more appropriate process, then reject the PIR as follows:

5.1.3.1.1 In the Comments section, provide specific justification for rejection or information indicating what was duplicated by the Request (e.g., the number of a duplicate Procedure Action Item).

8

2. STARTUP ~:. < 5%

< 350°F> Tavg>200 °F

< 0 <

6. REFUELING*~ < 0.95 < 140°F

SPECIFIC SYSTEM GUIDELINES (CONTD)

CC/ACC Applicable Affected Systems and Tech Component Required Action(s)

Mode(s) Specs Component Cooling Water 1, 2, 3, 4 CC/ACC - 3.7.3 Declare affected systems Inoperable and enter

/ Auxiliary Component CHW - 3.7.12 appropriate cascading Tech Specs Cooling Water Train A(B) SVS - TRM 3.7.13 Complete OP-903-066, Electrical Breaker Alignment

[CC/ACC] (1) AC Sources - 3.8.1.1 Check, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and at least every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> UHS - 3.7.4 thereafter in accordance with TS 3.8.1.1b.

CCS - 3.6.2.2 CS - 3.6.2.1 Complete TS 3.8.1.1 Action d within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> CREAFS - 3.7.6.1 If an opposite train component becomes Inoperable, CR Air Temp. - 3.7.6.3 then evaluate Tech Specs for both trains being CR Air Temp - 3.7.6.4 Inoperable.

Charging Pumps - 3.1.2.4 EFW - 3.7.1.2 CR Air Temperature-Shutdown TS 3.7.6.4 is ECCS - 3.5.2 applicable to cascading TS even in modes 1-4 when moving fuel or loads over fuel is being performed.

Component Cooling Water 5, 6 Tech Specs to Review Review affected Tech Spec LCO Actions for

/ Auxiliary Component (not necessarily to enter) applicability.

Cooling Water Train A(B) Declare affected systems Inoperable and enter AC Sources - 3.8.1.2 (1) appropriate Tech Specs if the CCW/ACCW systems

[CC/ACC] CREAFS - 3.7.6.1 are unavailable to provide cooling.

CR Air Temp. - 3.7.6.4 SDC Trains -3.9.8.1, 3.9.8.2, 3.4.1.3, 3.4.1.4, 3.4.1.5 (1) If any ACC Header A(B) To Wet Clg Tower Inlet Isol, ACC-132A(B), ACC-133A(B), ACC-134A(B), or ACC-135A(B) is Closed, then that train of ACC is Inoperable.

OP-100-014 Revision 336 Attachment 6.6 (4 of 31) 43

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and

b. Two separate and independent diesel generators, each with:

1. Diesel oil feed tanks containing a minimum volume of 339 gallons of fuel, and

2. A separate diesel generator fuel oil storage tank, and

3. A separate fuel transfer pump.

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

ACTION:

a. With one offsite circuit of 3.8.1.1a inoperable, demonstrate the OPERABILITY of the remaining offsite A.C. circuit by performing Surveillance Requirement 4.8.1.1.1a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. Restore the offsite A.C. circuit to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With one diesel generator of 3.8.1.1b inoperable:

(1) Demonstrate the OPERABILITY of the remaining A.C. circuits by performing Surveillance Requirements 4.8.1.1.1a (separately for each offsite A.C. circuit) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. If the diesel generator became inoperable due to any cause other than an inoperable support system, an independently testable component, or preplanned maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel generator (unless it has been successfully tested in the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) by performing Surveillance Requirement 4.8.1.1.2a.4 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> unless the absence of any potential common mode failure for the remaining diesel generator is demonstrated.

(2) Restore the diesel generator to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, unless the following condition exists:

WATERFORD - UNIT 3 3/4 8-1 AMENDMENT NO. 23,92,126,166,199,216

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

(a) The requirement for restoration to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be extended to 10 days if a temporary emergency diesel generator is verified available, and (b) If at any time the temporary emergency diesel generator availability cannot be met, either restore the temporary emergency diesel generator to available status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (not to exceed 10 days from the time the permanent plant EDG originally became inoperable), or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With one offsite A.C. circuit and one diesel generator of the above required A.C.

electrical power sources inoperable, demonstrate the OPERABILITY of the remaining offsite A.C. circuit by performing Surveillance Requirement 4.8.1 .1.1 a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; and, if the diesel generator became inoperable due to any cause other than an inoperable support system, an independently testable component, or preplanned maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel generator by performing Surveillance Requirement 4.8.1.1 .2a.4 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (unless it is already operating) unless the absence of any potential common mode failure for the remaining diesel generator is demonstrated. Restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Restore the other A.C. power source (offsite A.C.

circuit or diesel generator) to OPERABLE status in accordance with the provisions of ACTION statement a or b, as appropriate, with the time requirement of that ACTION statement based on the time of initial loss of the remaining inoperable A.C. power source. A successful test of diesel generator OPERABILITY per Surveillance Requirement 4.8.1 .1 .2a.4 performed under this ACTION statement satisfies the diesel generator test requirement of ACTION statement a or b.

d. With one diesel generator inoperable, in addition to ACTION b. or c. above, verify that:

(1) All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and (2) When in MODE 1, 2, or 3, the steam-driven emergency feed pump is OPERABLE.

If these conditions are not satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

WATERFORD - UNIT 3 3/4 8-2 AMENDMENT NO. 23,120, 166

Operating Instruction OI-038-000 Emergency Operating Procedures Revision 013 Operations Expectations / Guidance 5.3 REACTOR TRIP RECOVERY 5.3.1 Reset ESFAS Actuation If an EFSAS has occurred, the operator should reset the actuation and secure the applicable components using the Standard Appendix or the off-normal procedure if additional guidance is necessary.

This step should be completed prior to exiting the EOPs.

5.3.2 Sample the RCS The operator should notify chemistry to sample the RCS to comply with the required Technical Specification sample.

This step should be completed prior to exiting the EOPs.

5.3.3 Calculate Shutdown Margin If emergency boration is in progress, it is necessary for the operator to calculate Shutdown Margin and secure the boration prior to exiting this procedure.

5.3.4 Radiation Monitor Restoration 5.3.4.1 Some radiation monitors sample pumps needed for Emergency Plan may require restarting following a loss of power or voltage dips. Examples are Plant Stack WRGM, Plant Stack PIGs, Condenser WRGM and Fuel Handling Building PIGs. OP-004-001, Radiation Monitoring, limitation 3.2.1 provides a more comprehensive list of Radiation Monitors that will require their pumps to be restarted.

17

Administrative Procedure OP-100-017 Emergency Operating Procedure Implementation Guide Revision 004 5.11 USE OF MULTIPLE EMERGENCY OPERATING PROCEDURES 5.11.1 Simultaneous performance of more than one optimal recovery procedure is prohibited. The FRP is used for multiple events (except as noted below).

5.11.2 Certain events do not require offsite power in order to adequately mitigate the effects of the event. For this reason, if a loss of offsite power occurs concurrently with any of the following events: LOCA, ESD, LOF, or SGTR then the diagnosed ORP should be entered. LOOP/LOFC should only be entered when it is the only event in progress.

16

Administrative Procedure OP-100-017 Emergency Operating Procedure Implementation Guide Revision 004 5.16 FUNCTIONAL RECOVERY PROCEDURE 5.16.1 The Standard Post Trip Actions (SPTAs) are performed prior to entry into the FRP for an event initiated from Mode 1 or Mode 2. The FRP may be entered directly after completion of the SPTAs if a diagnosis of a single event is not possible. The FRP might also be entered from an ORP if an ORP had been initially selected but was subsequently found to be inadequate. If the Safety Function Acceptance Criteria are not satisfied at any time, then the CRS should evaluate the need to implement the FRP.

5.16.2 The FRP may be entered directly from a Mode 3 or Mode 4 event if the Entry Conditions are met.

5.16.3 The CRS reviews the status of all safety functions using the Resource Assessment Trees and the Safety Function Tracking Sheet. The CRS should obtain input and concurrence from the Balance of Plant Operator and the At The Controls Operator while determining which success path to implement. The CRS will implement success paths for all safety functions based on equipment availability and current plant conditions. The STA should independently perform a check of each Safety Function using the Safety Function Status Checklist to determine the appropriate success path to enter. The STA should concur with the CRS to validate the CRS assessment of the appropriate success path to enter.

5.16.4 More than one safety function may be pursued concurrently if plant conditions warrant.

5.16.5 If a significant change in plant status has occurred as determined by the CRS, then the CRS should reevaluate the success paths in use using the Resource Assessment Trees and the Safety Function Tracking Sheet. The STA should independently validate the CRS assessment of the success paths.

22

Fire Protection FP-001-020 Fire Emergency / Fire Report Revision 311 3.0 DEFINITIONS 3.1 Fire Brigade: A five (5) man team of plant personnel assigned to firefighting duties on a shift basis and under the direction of the SM/CRS, Emergency Coordinator or TSC Lead Communicator. A minimum of (2) two brigade members and the leader shall have knowledge of plant safety-related systems.

3.2 Fire Brigade Leader: An individual who through training and demonstrated experience, directs the brigade during firefighting activities and is designated as "in charge" of the fire scene, the brigade, and is responsible for directing firefighting activities at the location of the fire for both the brigade and offsite support. As a minimum, this will be a fully qualified Level A Nuclear Auxiliary Operator (NAO).

3.3 Fire Watch: A person trained in accordance with NTP-202, Fire Protection Training, and equipped to sound an alarm and attempt to extinguish fires that may occur during the work process.

3.4 Off-site Fire Department: The off-site fire department is the Hahnville Volunteer Fire Department, which has mutual aid agreements with other St. Charles Parish volunteer fire departments.

5

Operating Instruction OI-042-000 Watch Station Processes Revision 045 5.1.4 PERSONNEL/SHIFT STAFFING 5.1.4.1 Shift Staffing 5.1.4.1.1 All Modes

a. The SM/CRS is responsible for ensuring minimum shift staffing for their shift.

b. Radiation Protection and the Fire Brigade shall be manned as follows:

A Health Physics Technician shall be on site when fuel is in the reactor. [P-6081]

Radiation Protection will be available 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day to process personnel radiation dosimetry, monitor for radiation contamination, and decontaminate affected personnel. [P-7472]

A site Fire Brigade of at least five members shall be maintained on site at all times. The Fire Brigade shall not include the Shift Manager, the Shift Technical Advisor, the Control Room Supervisor, the Emergency Communicator, nor the two other members (ATC and BOP) of the minimum shift crew necessary for a safe shutdown of the unit nor any personnel including the Remote Shutdown Operator positions (Field Operator and Switchgear Operator) required for other essential functions during a fire emergency. [P-16395]

The site Fire Brigade shall consist of one Fire Brigade Leader (normally a senior NAO) and four site personnel (two of which must be safety related system trained). The Emergency Communicator and Remote Shutdown Operators (Field Operator and Switchgear Operator) shall not fill a position on the Fire Brigade. [P-13950]

The Health Physics Technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence, provided that immediate action is taken to fill the required positions. [P-16395, P-2657]

22

ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 (SRO) Date of Exam: March 27, 2017 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 18 3 3 6 Emergency &

Abnormal 2 N/A N/A 9 2 2 4 Plant Evolutions Tier Totals 27 5 5 10 1 28 3 2 5 2.

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 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-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 X EA2.14: Ability to determine and interpret the 4.4 76 following as they apply to a small break LOCA:

Actions to be taken if PTS limits are violated (CFR 43.5) 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 2.2.22: Knowledge of limiting conditions for 000022 Loss of Rx Coolant Makeup / 2 X operations and safety limits.(CFR: 41.5 / 43.2 / 4.7 81 45.2) 000025 Loss of RHR System / 4 X 2.4.9: Knowledge of low power/shutdown 4.2 77 implications in accident (LOCA or loss of heat removal) mitigation strategies (CFR 41.10, 43.5, 45.13) 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 000038 Steam Gen. Tube Rupture / 3 AA2.01 Ability to determine and interpret the 000040 (BW/E05; CE/E05; W/E12) X following as they apply to the Steam Line Rupture: 4.7 78 Steam Line Rupture - Excessive Heat Occurrence and location of a steam line rupture Transfer / 4 from pressure and flow indications (CFR 43.5,45.13) 000054 (CE/E06) Loss of Main Feedwater / 4 2.4.6: Knowledge of EOP mitigation strategies.

000055 Station Blackout / 6 X (CFR: 41.10 / 43.5 / 45.13) 4.7 79 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 X AA2.03: Ability to determine and interpret the following as they apply to the Loss of DC Power:

DC loads lost; impact on ability to operate and 3.9 80 monitor plant systems(CFR: 43.5) 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4

BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 3 3 Group Point Total: 18/

6

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 X 3.9 82 AA2.05: Amount of boron to add to achieve required shutdown margin (CFR 43.5) 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 X 2.3.12 Knowledge of radiological safety 3.7 84 principles pertaining to licensed-operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high rad areas, aligning filters, etc. (CFR 43.7) 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 X AA2.17: Ability to determine and interpret 4.3 85 the following as they apply to plant fire on site: systems that may be affected by the fire (CFR 43.5) 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery X 2.2.37: Ability to determine operability and/or 4.6 83 availability of safety related equipment.

(CFR 43.5)

K/A Category Point Totals: 2 2 Group Point Total: 9/4 ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump A2.15: Ability to (a) predict the impacts 004 Chemical and Volume X of the following malfunctions or 3.7 86 Control operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High or low PZR level (CFR 43.5) 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water X 2.2.40: Ability to apply Technical 4.7 87 specifications for a system.(CFR: 43.2 /

43.5) 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features X A2.06: Ability to (a) predict the impacts 4.0 88 Actuation of the following malfunctions or operations on the ESFAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertent ESFAS Actuation (CFR 43.5) 022 Containment Cooling 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency X A2.04: Ability to (a) predict the impacts 3.8 90 Feedwater of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: pump failure or improper operation (CFR 43.5) 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator X 2.2.25 Knowledge of the bases in 4.2 89 Technical Specifications for limiting conditions for operation and safety limits (CFR 43.2) 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air

103 Containment K/A Category Point Totals: 3 2 Group Point Total: 28/

5

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication X 2.4.11 Knowledge of abnormal condition 4.2 91 procedures.(CFR: 41.10 / 43.5 / 45.13) 015 Nuclear Instrumentation 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge X A2.03: Ability to (a) predict the impacts of 3.1 92 the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Startup operations and the associated required valve lineups. (CFR 43.5) 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air X 2.4.8 Knowledge of how abnormal 4.5 93 operating procedures are used in conjunction with EOPs (CFR 43.5) 086 Fire Protection

K/A Category Point Totals: 1 2 Group Point Total: 10/

3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Waterford 3 (SRO) Date of Exam: September 14, 2015 Category K/A # Topic RO SRO-Only IR # IR #

Knowledge of individual licensed operator responsibilities 2.1.4 3.8 94 related to shift staffing, such as medical requirements, no solo operation, maintenance of active license status,

1. 10CFR55, etc. (CFR 43.2)

Conduct of Ability to identify and interpret diverse indications to 2.1.45 4.3 95 Operations validate the response of another indication (CFR 43.5) 2.1.

Subtotal Knowledge of tagging and clearance procedures 2.2.13 4.3 96 2.

2.2.12 Knowledge of surveillance procedures (CFR 43.5) 4.1 97 Equipment Control 2.2.

Subtotal Ability to control releases (CFR 43.4) 2.3.11 4.3 98 3.

2.3.

Radiation Control 2.3.

Subtotal Knowledge of the parameters and logic used to assess 2.4.21 4.6 99 the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant integrity, containment conditions, radioactivity release control, etc.

4. (CFR 43.4)

Emergency Knowledge of events related to system operation/status 2.4.30 4.1 100 Procedures / that must be reported to internal organizations or external Plan agencies, such as the State, the NRC, or the transmission system operator (CFR 43.5) 2.4.

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 1/1 040 AA2.01 038 EA2.14 rejected. Could not make an SRO question from the original K/A that did not conflict with the K/A for RO10.

SRO 3 They were too much alike and one of the K/As had to be (Q78) rejected.

1/1 058 AA2.03 058 AA2.02 rejected. Could not develop an SRO only question associated with the determination and interpretation SRO 5 of a low DC voltage.

(Q80) 1/1 022 2.2.22 077 2.2.22 rejected. There is a 2015 RO question (RO18) written on voltage and Electrical Grid Disturbance basis SRO 6 information. Could not devise an SRO question independent (Q81) of that one.

1/2 CE/E09 033 2.2.37 rejected. The intermediate range detectors at W3 are the log channel instruments. There is a failed log channel SRO 8 on the audit simulator exam (scenario 1). Was not able to (Q83) create a question that would not create overlap with the audit exam.

2/1 004 A2.15 004 A2.05 rejected. RCP seal failures do not have an effect on CVCS other than CBO going to the VCT. Could not SRO 11 develop a question based solely on rising CBO temperature (Q86) or flow rate.

2/1 013 A2.06 013 A2.02 rejected. This K/A was an exact duplicate of the K/A for RO40.

SRO 13 (Q88) 2/1 064 2.2.25 062 2.2.25 rejected. The basis section for AC distribution (TRM 3.8.1.1) was used as an RO question on the 2015 RO SRO 14 exam (RO18). Could not develop a question that did not (Q89) duplicate the RO question.

2/1 061 A2.04 103 A2.03 rejected. W3 does not utilize a Phase A and Phase B Isolation system.

SRO15 (Q90) 2/2 029 A2.03 017 A2.02 rejected. Could not locate enough procedural guidance on the effects that core damage would have on SRO 17 Core Exit Thermocouples to develop an SRO test question.

(Q92) 3/2 2.2.13 2.2.5 rejected. The steps in the W3 procedure for making design changes to the facility are cumbersome and designing SRO21 a specific question on these detailed steps resulted in (Q96) questions that are considered minutia.

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000009 EA2.14 Importance Rating 4.4 K/A Statement 009 Small Break LOCA, EA2.14: Ability to determine or interpret the following as they apply to the Small Break LOCA: Actions to be taken if PTS limits are violated Proposed Question: SRO 1 (Q76) Rev: 0 Given:

A Small Break LOCA has occurred SIAS, CIAS, and MSIS are present The crew is performing actions of OP-902-002, Loss of Coolant Accident Recovery Procedure The ATC informs the CRS that the upper limits of the RCS Pressure and Temperature Curve (PT curve) are being exceeded To restore the RCS within the limits of the Pressure-Temperature (PT) curve, the CRS will direct the crew to take actions to lower RCS ____(1)____ .

To maintain the RCS within PT limits, the crew will re-energize the 32A and 32B busses in accordance with OP-902-009, Standard Appendices, Appendix ______(2)______.

(1) (2)

A. temperature 12, Electrical Restoration B. pressure 12, Electrical Restoration C. pressure 25, Restore Pressurizer Heater Control D. temperature 25, Restore Pressurizer Heater Control Revision 0 Facility: Waterford 3 Page 1 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The applicant is required to know exceeding the upper limit of the PT curve means that the RCS is over sub-cooled. OP-902-002 step 30 requires the crew to stop the cooldown and lower pressure. Appendix 12 has various sections for restoring electrical power, but guidance to close in the 32 feeder breakers with a SIAS is located in Appendix 25 B. Incorrect. Part 1 is correct. Appendix 12 has various sections for restoring electrical power, but guidance to close in the 32 feeder breakers with a SIAS is located in Appendix 25.

C. CORRECT: If the upper limit of the PT curve is exceeded, the concern is PTS. To restore the RCS to within limits, the crew will need to lower RCS pressure. OP-902-002 step 30 directs the crew to Appendix 25, Restore Pressurizer Heater Control guidance for energizing the 32A(B) busses and restoring pressurizer heaters.

D. Incorrect. The applicant is required to know exceeding the upper limit of the PT curve means that the RCS is over sub-cooled. OP-902-002 step 30 requires the crew to stop the cooldown and lower RCS pressure. Part 2 is correct.

Technical Reference(s): OP-902-002 step 30 Rev. 20 (Attach if not previously provided) OP-902-009 Appendix 25 Rev. 315 (including version/revision number) OP-902-009 Attachment 2 Rev. 315 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 2 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000025 2.4.9 Importance Rating 4.2 K/A Statement 025 Loss of RHR System, 2.4.9: Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Proposed Question: SRO 2 (Q77) Rev: 0 Given:

Shutdown Cooling Trains A and B is in service An Inadvertent RAS has occurred The crew will enter ____(1)____. The crew will first ______(2)______.

(1) (2)

A. OP-901-504, Inadvertent ESFAS restart LPSI pumps Actuation only B. OP-901-504, Inadvertent ESFAS reset RAS initiation Actuation and actuation relays C. OP-901-131, Shutdown Cooling reset RAS initiation Malfunction and OP-901-504, and actuation relays Inadvertent ESFAS Actuation D. OP-901-131, Shutdown Cooling restart LPSI pumps Malfunction and OP-901-504, Inadvertent ESFAS Actuation Revision 0 Facility: Waterford 3 Page 3 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The guidance to restart LPSI pumps is located in OP-901-504. OP-901-504 directs the crew to perform OP-901-131 concurrently. Part 2 is correct.

B. The guidance to restart LPSI pumps is located in OP-901-504. OP-901-504 directs the crew to perform OP-901-131 concurrently. LPSI pumps can be restarted 1 second after an RAS signal. OP-901-504 directs the crew to reset RAS initiation and actuation relays after starting LPSI pumps so that SI-602A(B) can be closed.

C. Incorrect. Part 1 is correct. LPSI pumps can be restarted 1 second after an RAS signal. OP-901-504 directs the crew to reset RAS initiation and actuation relays after starting LPSI pumps so that SI-602A(B) can be closed.

D. CORRECT: A loss of Shutdown Cooling flow has occurred due to both LPSI pumps tripping on an Inadvertent RAS. The guidance to restart LPSI pumps is located in OP-901-504. OP-901-504 directs the crew to perform OP-901-131 concurrently. LPSI pumps can be restarted 1 second after an RAS signal.

Technical Reference(s): OP-901-131 section E2 Rev. 304 (Attach if not previously provided) OP-901-504 section E3 Rev. 10 (including version/revision number) SD-SI page 27 Rev. 15 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 4 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000040 AA2.01 Importance Rating 4.7 K/A Statement 040 Steam Line Rupture, AA2.01: Ability to determine and interpret the following as they apply to the Steam Line Rupture: Occurrence and location of a steam line rupture from pressure and flow indications Proposed Question: SRO 3 (Q78) Rev: 1 Given:

The ATC reports that RCS Tcold is dropping rapidly A MSIS, SIAS, and CIAS has occurred Following the ESFAS actuations, the BOP reports that both Steam Generator Pressures are 500 psia and dropping The crew will go to ____(1)____ Recovery Procedure. The next action the CRS will direct is to ____(2)____.

(1) (2)

A. OP-902-008, Functional close MS-401A and MS-401B, Pump AB TURB STM Supplies B. OP-902-008, Functional stabilize RCS temperature within the PT curves C. OP-902-004, Excess Steam close MS-401A and MS-401B, Demand Pump AB TURB STM Supplies D. OP-902-004, Excess Steam stabilize RCS temperature within Demand the PT curves Revision 0 Facility: Waterford 3 Page 5 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect: The MSIS did not isolate the leak to any steam generator. The applicant could incorrectly determine that two events are going on or that parameters do not support any optimal recovery procedure entry; either case would put the crew into OP-902-008. Part 2 is correct.

B. Incorrect. The MSIS did not isolate the leak to any steam generator. The applicant could incorrectly determine that two events are going on or that parameters do not support any optimal recovery procedure entry, either case would put the crew into OP-902-008. Stabilize RCS temperature within the PT curves is a step that must be performed but will not be performed next because both S/G pressures are still dropping.

C. CORRECT: The MSIS should have either isolated the steam leak, or contained the steam leak to one Steam Generator unless the leak is in the common lines going to EFW Pump AB.

With S/G pressures still dropping and equal to each other, OP-902-004 directs the crew to close MS-401A and B to attempt to isolate the leak. Nothing in the stem indicates that two events are occurring, the crew should go to OP-902-004.

D. Incorrect. Part 1 is correct. Stabilize RCS temperature within the PT curves is a step that must be performed but will not be performed next because both S/G pressures are still dropping Technical Reference(s): OP-902-004 step 15 and step 18 Rev. 16 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE04 obj. 7 and 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Free review question. Changed the stem to give indications that an ESD is occurring.

This meets the part of the K/A for determining the occurrence of an ESD.

Revision 0 Facility: Waterford 3 Page 6 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000055 2.4.6 Importance Rating 4.7 K/A Statement 055 Station Blackout, 2.4.6: Knowledge of EOP mitigation strategies.

Proposed Question: SRO 4 (Q79) Rev: 1 Given:

A Station Blackout has occurred The CRS has been informed that restoration of one Emergency Diesel Generator is expected in five hours and the duration for restoration of offsite power is unknown The CRS will go to ____(1)____.

To ensure station battery deep load shed time requirements are met, a decision to declare an Extended Loss of AC Power (ELAP) event must occur within a maximum of

____(2)____ hour(s) of the onset of the Station Blackout .

(1) (2)

A. OP-902-005, Station Blackout one Recovery Procedure only B. OP-902-005, Station Blackout two Recovery Procedure only C. OP-902-005, Station Blackout two Recovery Procedure and declare an Extended Loss of AC Power (ELAP) event D. OP-902-005, Station Blackout one Recovery Procedure and declare an Extended Loss of AC Power (ELAP) event Revision 0 Facility: Waterford 3 Page 7 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Entry into OP-902-005 only would be correct if the Loss of all AC power was determined to be less than four hours. Part 2 is correct.

B. Incorrect. Entry into OP-902-005 only would be correct if the Loss of all AC power was determined to be less than four hours. Part 2 is plausible because two hours is less than the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping time for W3 on the station batteries with the loss of all AC power.

C. Incorrect. Part 1 is correct. Part 2 is plausible because two hours is less than the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping time for W3 on the station batteries with the loss of all AC power.

D. CORRECT: When the Loss of all AC power may exceed the License Basis SBO duration of four hours, the ELAP must be declared within one hour of the onset of the SBO to meet ELAP time requirements. The caution prior to step 17 of OP-902-005 states that the ELAP must be declared within one hour of the onset of a SBO to ensure that the Station Battery deep load shed time requirements are met. Once an ELAP is declared, the crew will commence FIG-001, Extended Loss of AC Power. This is a new SBO step for W3 and is a result of the latest revision to the EOPs.

Technical Reference(s): OP-902-005 Step 17 Rev. 20 (Attach if not previously provided) TG-OP-902-005 step 17 Rev. 309 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE05 obj. 6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1, 5 Comments:

Free review question. Changed the distractor hours to declare an ELAP to less than the coping time of station batteries.

Revision 0 Facility: Waterford 3 Page 8 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 000058 AA2.03 Importance Rating 3.9 K/A Statement 058 Loss of DC Power, AA2.03: Ability to determine and interpret the following as they apply to the Loss of DC Power: DC loads lost; impact on ability to operate and monitor plant systems Proposed Question: SRO 5 (Q80) Rev: 0 Given:

A Loss of all Offsite Power has occurred The crew has entered OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery Procedure In preparation for securing TGB-DC loads, the CRS will direct the crew to vent the Main Generator in accordance with the guidance located in ____(1)____.

Reducing TGB-DC loads will preserve the remote operation of the ____(2)____ bus feeder breakers.

(1) (2)

A. OP-902-009, Attachment 33-B, 2 Generator Auxiliary Operations-LOOP B. OP-902-009, Attachment 33-B, 1 Generator Auxiliary Operations-LOOP C. OP-901-313, section E4, Loss of 1 125 Volt DC Bus TGB-DC D. OP-901-313, section E4, Loss of 2 125 Volt DC Bus TGB-DC Revision 0 Facility: Waterford 3 Page 9 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: OP-902-003 directs the crew to perform OP-902-009, Attachment 33-B, Generator Auxiliary Operations-LOOP. This attachment will provide the guidance to vent the main generator in preparation for securing the DC seal oil pump. The TGB-DC is preserved to maintain important loads such as the remote operation of the 2 bus feeder breakers.

B. Incorrect. Part 1 is correct. The feeder breakers for the 1 bus is powered from Safety related DC (A-DC and B-DC) for backup overcurrent protection for the RCPs.

C. Incorrect. OP-901-313 is plausible because a section exists for a loss of TGB-DC but it provides no guidance to vent the main generator because AC power is assumed to be available. The feeder breakers for the 1 bus is powered from Safety related DC (A-DC and B-DC) for backup overcurrent protection for the RCPs.

D. Incorrect. OP-901-313 is plausible because a section exists for a loss of TGB-DC but it provides no guidance to vent the main generator because AC power is assumed to be available. Part 2 is correct.

Technical Reference(s): OP-902-009 Attachment 33-B Rev. 315 (Attach if not previously provided) TG-OP-902-009 page 86 Rev. 311 (including version/revision number) OP-901-313 section E4 Rev. 305 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 10 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 00022 2.2.22 Importance Rating 4.7 K/A Statement 022 Loss of Rx Coolant Makeup, 2.2.22: Knowledge of limiting conditions for operations and safety limits.

Proposed Question: SRO 6 (Q81) Rev: 0 Given:

A loss of Volume Control Tank (VCT) level has occurred Charging parameters indicate gas intrusion into the Charging Pumps The crew will perform the actions of ____(1)____. The restoration of Charging Pump(s) will require the entry into TS ____(2)____ .

(1) (2)

A. OP-901-110,Pressurizer 3.1.2.4, Charging Level Control Malfunction Pumps-Operating B. OP-901-112, Charging or 3.1.2.4, Charging Letdown Malfunction Pumps-Operating C. OP-901-112, Charging or 3.0.3 Letdown Malfunction D. OP-901-110,Pressurizer 3.0.3 Level Control Malfunction Revision 0 Facility: Waterford 3 Page 11 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The Pressurizer Level Control Malfunction does not provide guidance on Gas bound Charging pumps even though the loss of Charging will result in a Pressurizer level drop. TS 3.1.2.4 has a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action if only one charging pump is inoperable. With two trains inoperable, there is no action available, therefore TS 3.0.3 entry is required B. Incorrect. Part 1 is correct. TS 3.1.2.4 has a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action if only one charging pump is inoperable. With two trains inoperable, there is no action available, therefore TS 3.0.3 entry is required.

C. CORRECT: Section E3 GAS BOUND CHARGING PUMPS of OP-901-112, Charging or Letdown Malfunction, provides the guidance for gas binding in the Charging Pumps. This guidance directs the crew to place all three charging pumps in the OFF position and to enter TS 3.0.3. The applicant is required to know that all three Charging Pumps are taken to off.

Some off-normal strategies for VCT level control require the crew to leave the third charging pump selected to auto such that only TS 3.1.2.4 entry is required. This is not the case for gas intrusion. This question does not address a safety limits because W3 has no safety limits associated with a loss of charging.

D. Incorrect. The Pressurizer Level Control Malfunction does not provide guidance on Gas bound Charging pumps even though the loss of Charging will result in a Pressurizer level drop. Part 2 is correct.

Technical Reference(s): OP-901-112 page 20 Rev. 6 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank # RO5 (Note changes or attach parent)

New Question History: Last NRC Exam 2012 RO NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 55.43 2, 5 Comments:

Free review question. No changes required.

Revision 0 Facility: Waterford 3 Page 12 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000024 AA2.05 Importance Rating 3.9 K/A Statement 024 Emergency Boration, AA2.05: Amount of boron to add to achieve required shutdown margin.

Proposed Question: SRO 7 (Q82) Rev: 0 Given:

The crew has entered OP-902-004, Excess Steam Demand Recovery Procedure due to a Main Steam Line break.

The Emergency Boration Termination criteria for an excessive cooldown is located in

___(1)___ .

This guidance states that the crew must emergency borate until the cooldown has stopped ___(2)___ Cold Shutdown boron concentration is reached .

(1) (2)

A. OP-902-004, Excess Steam or Demand Recovery Procedure B. OP-901-103, Emergency or Boration C. OP-902-004, Excess Steam and Demand Recovery Procedure D. OP-901-103, Emergency and Boration Revision 0 Facility: Waterford 3 Page 13 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect.. Emergency boration termination criteria for an uncontrolled cooldown is not located in OP-902-004. The operator is expected to know the emergency boration termination criteria located in OP-901-103 when making the decision to secure emergency boration. Part 2 is correct.

B. CORRECT: The emergency boration termination criteria is located in OP-901-103, Emergency Boration. During an uncontrolled cooldown event, the crew will emergency borate until the cooldown has stopped or Cold Shutdown boron concentration is reached.

C. Incorrect. During an uncontrolled cooldown event, the crew will emergency borate until the cooldown has stopped or Cold Shutdown boron concentration is reached. Emergency boration termination criteria for an uncontrolled cooldown is not located in OP-902-004. The operator is expected to know the emergency boration termination criteria located in OP-901-103 when making the decision to secure emergency boration D. Incorrect. Part 1 is correct. During an uncontrolled cooldown event, the crew will emergency borate until the cooldown has stopped or Cold Shutdown boron concentration is reached.

Technical Reference(s): OP-901-103 step 6.2 Rev. 3 (Attach if not previously provided) OP-902-004 Rev 16 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 14 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # CE/E09 2.2.37 Importance Rating 4.6 K/A Statement CE/E09 Functional Recovery, 2.2.37: Ability to determine operability and/or availability of safety related equipment.

Proposed Question: SRO 8 (Q83) Rev: 0 Given:

The crew has entered OP-902-008, Safety Function Recovery Procedure.

The CRS will identify success paths to implement for each safety function using the

___(1)___ .

OI-038-000, EOP Operations Expectations/Guidance procedure, states that the operating crew will monitor Safety Functions ___(2)___ .

(1) (2)

A. Safety Function Status every 15 minutes Checklist B. Safety Function Status continuously Checklist C. Resource Assessment every 15 minutes Trees (RATs)

D. Resource Assessment continuously Trees (RATs)

Revision 0 Facility: Waterford 3 Page 15 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The Safety Function Status Checklist is plausible because OP-100-017, step 5.16.3, directs the STA (not the CRS) to use this checklist to select success paths. The 15 minute time requirement is plausible because this is the time requirement for the STA to check safety functions being met.

B. Incorrect. The Safety Function Status Checklist is plausible because OP-100-017, step 5.16.3, directs the STA (not the CRS) to use this checklist to select success paths. Part 2 is correct.

C. Incorrect. Part 1 is correct. . The 15 minute time requirement is plausible because this is the time requirement for the STA to check safety functions being met.

D. CORRECT: Per OP-100-017, step 5.16.3, the CRS will select success paths for all safety functions based on equipment availability and current plant conditions using the RATs. Per OP-0100-017, step 5.20.3, the operating crew will continuously monitor safety functions being met. This question is SRO only because it requires knowledge of administrative procedures to coordinate the implementation of the safety function recovery procedure.

Technical Reference(s): OP-100-017 step 5.16.3, 5.20.2, 5.20.3 Rev. 4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 16 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 000036 2.3.12 Importance Rating 3.7 K/A Statement 036 Fuel Handling Accident, 2.3.12 Knowledge of radiological safety principles pertaining to licensed-operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high rad areas, aligning filters, etc.

Proposed Question: SRO 9 (Q84) Rev: 0 Given:

A Core offload is in progress A loss of the Reactor Cavity Pool Seal has occurred The Refuel SRO will expect level to lower in the ___(1)___ . The loss of Refuel Cavity water level guidelines that contain the safe locations to place a fuel assembly in transit during the loss of Reactor Cavity level can be found in ___(2)___ .

(1) (2)

A. Reactor Cavity only OP-901-405, Fuel Handling Incident B. Reactor Cavity only OP-010-006, Outage Operations C. Reactor Cavity and Spent OP-901-405, Fuel Fuel Pool Handling Incident D. Reactor Cavity and Spent OP-010-006, Outage Fuel Pool Operations Revision 0 Facility: Waterford 3 Page 17 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The Spent Fuel Pool is not normally connected to the Refuel Cavity.

However, during a core offload the two will be connected due to FHS-201 open and the blind flange removed. A loss of the cavity pool seal will affect both. OP-901-405, Fuel Handling Incident does not address the loss of Reactor Cavity Level.

B. Incorrect. The Spent Fuel Pool is not normally connected to the Refuel Cavity.

However, during a core offload the two will be connected due to FHS-201 open and the blind flange removed. A loss of the cavity pool seal will affect both. Part 2 is correct.

C. Incorrect. Part 1 is correct. OP-901-405, Fuel Handling Incident does not address the loss of Reactor Cavity Level.

D. CORRECT: The Spent Fuel Pool is aligned to the Refuel Cavity during an core offload. A loss of the Cavity Pool Seal will cause water level in the Spent Fuel Pool and Refuel Cavity to lower. To prevent radiation levels from rising in containment, OP-010-006, Outage Operations, has the guidance on where to place fuel assemblies upon a loss of Refuel Cavity level.

Technical Reference(s): OP-010-006 Attachment 9.23 Rev. 329 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-REQ04 obj. 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 55.43 5,7 Comments:

Revision 0 Facility: Waterford 3 Page 18 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 00067 AA2.17 Importance Rating 4.3 K/A Statement 067 Plant Fire on Site, AA2.17: Ability to determine and interpret the following as they apply to plant fire on site: systems that may be affected by the fire.

Proposed Question: SRO 10 (Q85) Rev: 0 Given:

The fire brigade leader reports a fire in the Relay Room that requires activation of the Fire Brigade team.

Within 20 minutes of the fire, the CRS will direct the crew to perform actions to electrically lockout the Reactor Coolant Pumps because the location of this fire could affect ____(1)____. This step will be performed in accordance with____(2)____ .

(1) (2)

A. the backup overcurrent OP-901-524, Fires in areas protection for the RCPs affecting Safe Shutdown B. the backup overcurrent OP-901-502, Evacuation of protection for the RCPs Control Room and Subsequent Plant Shutdown C. components which provide OP-901-524, Fires in areas cooling to the RCP seals affecting Safe Shutdown D. components which provide OP-901-502, Evacuation of cooling to the RCP seals Control Room and Subsequent Plant Shutdown Revision 0 Facility: Waterford 3 Page 19 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. Plausible because there is a TS action to de-energize the RCPs if the primary and backup overcurrent protection is inoperable (TS 3.8.4.1). Part 2 is correct.

B. Incorrect: Plausible because there is a TS action to de-energize the RCPs if the primary and backup overcurrent protection is inoperable (TS 3.8.4.1). OP-901-502 would be entered and a Control Room Evacuation performed if the fire was in the cable spreading room which is adjacent to the relay room.

C. CORRECT: For a fire in the relay room, the crew will enter OP-901-524, Fires in areas affecting Safe Shutdown. This procedures directs the crew to trip the RX, secure RCPs and electrically lockout the RCPs per Attachment 2 within 20 minutes of the fire.

D. Incorrect. Part 1 is correct. OP-901-502 would be entered and a Control Room Evacuation performed if the fire was in the cable spreading room which is adjacent to the relay room.

Technical Reference(s): OP-901-524 pages 7,8, and 106 Rev. 15 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 20 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 004 A2.15 Importance Rating 3.7 K/A Statement 004 Chemical and Volume Control, A2.15: Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High or low PZR level Proposed Question: SRO 11 (Q86) Rev: 0 Given:

Reactor Power is 100% with Tavg and Tref matched Pressurizer Level Channel X is 40% and slowly lowering Pressurizer Level Channel Y is 60% and slowly rising Channel X is selected for service The crew has entered OP-901-110, Pressurizer Level Control Malfunction The CVCS system configuration will be changed by (1) .

The crew will implement section (2) .

(1) (2)

A. both backup charging pumps E1, Pressurizer Level Control Channel starting Malfunction B. letdown flow rising to 126 gpm E1, Pressurizer Level Control Channel Malfunction C. both backup charging pumps E3, Pressurizer Level Controller Malfunction starting D. letdown flow rising to 126 gpm E3, Pressurizer Level Controller Malfunction Revision 0 Facility: Waterford 3 Page 21 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: The level setpoint for 100% is normally ~ 56%. With Selected pressurizer level indicating 40% both backup charging pumps would have received start signals and letdown should be at minimum (~28 gpm). Pressurizer level is diverging between channels. This is indicative of a failing channel, therefor E1 is correct section to implement.

B. Incorrect. If Channel Y was the selected channel, then letdown flow rising to 126 gpm would be the correct answer. Part 2 is correct.

C. Incorrect. Part 1 is correct. Pressurizer level is diverging between channels. This is indicative of a failing channel. If Pressurizer level channels were tracking together it would be indicative of a Pressurizer controller or setpoint malfunction.

D. Incorrect. If Channel Y was the selected channel, then letdown flow rising to 126 gpm would be the correct answer. Pressurizer level is diverging between channels.

This is indicative of a failing channel. If Pressurizer level channels were tracking together it would be indicative of a Pressurizer controller or setpoint malfunction.

Technical Reference(s): OP-901-110 pages 4-7 Rev. 9 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank # SRO1 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2010 NRC SRO Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 22 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 008 2.2.40 Importance Rating 4.7 K/A Statement 008 Component Cooling Water, 2.2.40: Ability to apply Technical specifications for a system.

Proposed Question: SRO 12 (Q87) Rev: 0 Given:

A Tornado Watch is in affect for St. Charles Parish The crew has entered OP-901-521, Severe Weather and Flooding Due to the changing weather conditions, the CRS will verify additional operability requirements are met for the Dry Cooling Tower (DCT) fans by verifying DCT fans

___(1)___ are operable. The Dry Cooling Tower fans are considered operable if their control switches are in the ___(2)___ position(s).

(1) (2)

A. 7 through 15 auto only B. 1 through 9 auto only C. 1 through 9 auto or fast D. 7 through 15 auto or fast Revision 0 Facility: Waterford 3 Page 23 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Part 1 is correct. The basis for TS 3.7.4 action c states that the DCT fan operability is maintained by operating in fast or auto mode.

B. Incorrect. TS 3.7.4 action c states with a tornado watch in effect, all 9 DCT fans under the missile protected portion of the DCT shall be operable. The 9 DCT fans under the missile shield are DCT fans 7-15. The basis for this TS action states that the DCT fan operability is maintained by operating in fast or auto mode.

C. Incorrect. TS 3.7.4 action c states with a tornado watch in effect, all 9 DCT fans under the missile protected portion of the DCT shall be operable. The 9 DCT fans under the missile shield are DCT fans 7-15. Part 2 is correct.

D. CORRECT: TS 3.7.4 action c states with a tornado watch in effect, all 9 DCT fans under the missile protected portion of the DCT shall be operable. The 9 DCT fans under the missile shield are DCT fans 7-15. The basis for TS 3.7.4 action c states that the DCT fan operability is maintained by operating in fast or auto mode.

Technical Reference(s): OP-901-521 section E2 step 3 Rev. 322 (Attach if not previously provided) TS 3.7.4 and basis (including version/revision number) SD-CC page 19 Rev. 22 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-CC00 obj. 10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Revision 0 Facility: Waterford 3 Page 24 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 A2.06 Importance Rating 4.0 K/A Statement 013 Engineered Safety Features Actuation, A2.06: Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertent ESFAS actuation.

Proposed Question: SRO 13 (Q88) Rev: 0 Given:

The plant is at 100% power An Inadvertent Containment Spray Actuation has occurred The crew will enter ____(1)____ . The CRS will direct the crew to establish Component Cooling Water Flow to the Reactor Coolant Pumps by opening _____(2)_____.

(1) (2)

A. OP-901-504, Inadvertent ESFAS CC-641, CC-710 and CC-713, Actuation only RCP Inlet and Outlet isolation valves B. OP-901-504, Inadvertent ESFAS CC-200A and CC-727, Suct. &

Actuation only Disch Heater Tie Valves A to AB C. OP-901-504, Inadvertent ESFAS CC-641, CC-710 and CC-713, Actuation and OP-902-000, RCP Inlet and Outlet isolation Standard Post Trip Actions valves D. OP-901-504, Inadvertent ESFAS CC-200A and CC-727, Suct. &

Actuation and OP-902-000, Disch Heater Tie Valves A to AB Standard Post Trip Actions Revision 0 Facility: Waterford 3 Page 25 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: To get an automatic CSAS, ESFAS requires a valid SIAS and 17.7 psia in containment. The SIAS would result in an automatic reactor trip and entry into OP-902-000.

In this case, a CSAS inadvertently actuated due to de-energizing the CSAS actuation relays without a SIAS, no trip would occur. The offnormal procedure is the only required entry. CC-200 A and CC-727 closed due to a CSAS but CC-200B and CC-563 are available to align CCW to the RCPs, but it wont because downstream valves CC-641, 710 and 713 are closed.

Opening CC-641, 710 and 713 will restore CCW to the Reactor Coolant Pumps.

B. Incorrect. Part 1 is correct. CC-200 A and CC-727 closed due to a CSAS but CC-200B and CC-563 are available to align CCW to the RCPs, but it wont because downstream valves CC-641, 710 and 713 are closed C. Incorrect. To get an automatic CSAS, ESFAS requires a valid SIAS and 17.7 psia in containment. The SIAS would result in an automatic reactor trip and entry into OP-902-000.

In this case, a CSAS inadvertently actuated due to de-energizing the CSAS actuation relays without a SIAS, no trip would occur. Part 2 is correct.

D. Incorrect. To get an automatic CSAS, ESFAS requires a valid SIAS and 17.7 psia in containment. The SIAS would result in an automatic reactor trip and entry into OP-902-000.

In this case, a CSAS inadvertently actuated due to de-energizing the CSAS actuation relays without a SIAS, no trip would occur. CC-200 A and CC-727 closed due to a CSAS but CC-200B and CC-563 are available to align CCW to the RCPs, but it wont because downstream valves CC-641, 710 and 713 are closed.

Technical Reference(s): OP-901-504 Section E2 Rev. 10 (Attach if not previously provided) SD-PPS page 83 Rev. 17 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO50 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 26 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 2.2.25 Importance Rating 4.2 K/A Statement 064 Emergency Diesel Generator, 2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operation and safety limits.

Proposed Question: SRO 14 (Q89) Rev: 0 Given:

Emergency Diesel Generator (EDG) A is inoperable due to unplanned maintenance 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> into the unplanned EDG A outage, the Temporary Emergency Diesel (TED) Generator is verified available and aligned as a backup for EDG A 8 days (192 hours0.00222 days <br />0.0533 hours <br />3.174603e-4 weeks <br />7.3056e-5 months <br />) from when EDG A first became inoperable, the Temporary Emergency Diesel (TED) Generator becomes unavailable and cannot be restored Per TS 3.8.1.1, from the time the Temporary Emergency Diesel (TED) Generator becomes unavailable, the time limit for restoring Emergency Diesel Generator A to service is ________ hours.

A. 28 B. 48 C. 68 D. 72 Revision 0 Facility: Waterford 3 Page 27 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> would be the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> AOT subtracted by the 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> that it took to get the TEDG available.

B. CORRECT: Technical Specification 3.8.1.1 basis states if the EDG becomes inoperable for pre-planned maintenance, the allowed outage time can be extended from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 10 days, if a TEDG is available and aligned. The same basis states,if the TEDG becomes unavailable during the 10 day AOT and cannot be restored, the EDG AOT reverts back to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, not to exceed 10 days. Therefore, 192 hour0.00222 days <br />0.0533 hours <br />3.174603e-4 weeks <br />7.3056e-5 months <br />s=8 days, 2 days (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) are available such that 10 days is not exceeded.

C. Incorrect. 68 hours7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br /> would be 10 days from the time the TEDG was placed in service.

D. Incorrect. If the TEDG becomes unavailable during the 10 day AOT and cannot be restored, the EDG AOT reverts back to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, not to exceed 10 days. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> would exceed the 10 day AOT.

Technical Reference(s): TS 3.8.1.1 basis (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-EDG00 obj. 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Free review question, no changes required.

Revision 0 Facility: Waterford 3 Page 28 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 061 A2.04 Importance Rating 3.8 K/A Statement 061 Auxiliary/ Emergency Feedwater, A2.04: Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: pump failure or improper operation.

Proposed Question: SRO 15 (Q90) Rev: 0 Given:

The crew is performing OP-902-000, Standard Post Trip Actions due to a loss of offsite power EFW Pump A is tagged out EFW Pump B has tripped on overcurrent A loss of the AB-DC Bus is imminent due to a loss of 3AB bus The crew will diagnose into ____(1)____ . The CRS will direct the crew to establish local control of Emergency Feedwater (EFW) Pump AB to obtain desired flow. EFW Pump AB speed will be monitored _____(2)_____.

(1) (2)

A. OP-902-008, Functional Recovery locally using a strobe scope Procedure B. OP-902-006, Loss of Feedwater at CP-8 Recovery Procedure C. OP-902-008, Functional Recovery at CP-8 Procedure D. OP-902-006, Loss of Feedwater locally using a strobe scope Recovery Procedure Revision 0 Facility: Waterford 3 Page 29 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The applicant may determine that due to the loss of AB-DC and the imminent overspeed of EFW Pump AB, OP-902-008 entry would be required to reference the attachment. This appendix and contingency action for EFW Pump AB was added in the latest revision to the EOPs. Part 2 is correct.

B. Incorrect. Part 1 is correct. . A loss of AB-DC will also cause a loss of the AB EFW Pump on CP-8, thereby requiring the crew to monitor speed locally using a strobe scope C. Incorrect. The applicant may determine that due to the loss of AB-DC and the imminent overspeed of EFW Pump AB, OP-902-008 entry would be required to reference the attachment. This appendix and contingency action for EFW Pump AB was added in the latest revision to the EOPs. . A loss of AB-DC will also cause a loss of the AB EFW Pump on CP-8, thereby requiring the crew to monitor speed locally using a strobe scope.

D. CORRECT: The crew will diagnose to OP-902-006 on a loss of feedwater. The appendix to take manual control of EFW Pump AB is referenced from OP-902-006. The procedure directs the crew to take manual control of EFW Pump AB if the loss of the AB-DC bus is imminent.

This is because EFW Pump AB will trip on overspeed due to the governor failing open. A loss of AB-DC will also cause a loss of the AB EFW Pump on CP-8, thereby requiring the crew to monitor speed locally using a strobe scope.

Technical Reference(s): OP-902-006 step 9 Rev. 18 (Attach if not previously provided) OP-902-009 Appendix 1, 36 Rev. 315 (including version/revision number) SD-EFW page 17 Rev. 13 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE06 obj. 3 and 9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 30 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 014 2.4.11 Importance Rating 4.2 K/A Statement 014 Rod Position Indication, 2.4.11 Knowledge of abnormal condition procedures.

Proposed Question: SRO 16 (Q91) Rev: 0 Given:

Power is 100%

The ATC reports erratic indication for CEA #20 on the CP-2 CEAC CRT To obtain the PMC Pulse Counter CEA position, the CRS will request indication using the ____(1)____ .

The CRS will direct the ATC to remove a CEAC from service using the guidance in

_____(2)_____.

(1) (2)

A. CPC that CEA 20 is targeted to OP-901-102, CEA or CEDMCS Malfunction B. CPC that CEA 20 is targeted to OP-004-006, Core Protection Calculator System C. CP-2 digital meter for CEA 20 OP-004-006, Core Protection Calculator System D. CP-2 digital meter for CEA 20 OP-901-102, CEA or CEDMCS Malfunction Revision 0 Facility: Waterford 3 Page 31 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. The CEA indication on CPCs are from Reed Switch Position Transmitters.

The PMC (pulse counter) position indication is only located on CP-2. Part 2 is correct.

B. Incorrect. The CEA indication on CPCs are from Reed Switch Position Transmitters.

The PMC (pulse counter) position indication is located on CP-2. The procedure for removing a CEAC are in OP-901-102 and OP-004-003 (CEAC), not in OP-004-006 (CPC). Plausible because applicants have a tendency to go to the CPC normal operating procedure when there is an issue with the CEAC because the CEAC is physically embedded in CPC B and C.

C. Incorrect. Part 1 is correct. The procedure for removing a CEAC can be found in OP-901-102 and OP-004-003 (CEAC), not in OP-004-006 (CPC). Plausible because applicants have a tendency to go to the CPC procedure when there is an issue with the CEAC because the CEAC is physically embedded in CPC B and C.

D. CORRECT: Pulse counter (PMC) CEA position indication is found on the CP-2 digital meter. OP-901-102, CEA or CEDMCS Malfunction has a section to remove a CEAC from service. The section has been added recently due to issues incurred while try to remove a CEAC from service using the normal operating procedure.

Technical Reference(s): OP-901-102 section E5 Rev. 304 (Attach if not previously provided) WLP-OPS-PMC slide 78 and 79 Rev. 10 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPO10 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 32 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 029 A2.03 Importance Rating 3.1 K/A Statement 029 Containment Purge System, A2.03: Ability to (a) predict the impacts of the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Startup operations and the associated required valve lineups.

Proposed Question: SRO 17 (Q92) Rev: 0 Given:

A Plant Startup from a refueling outage is in progress in accordance with OP-010-003 Prior to establishing Containment Integrity for Mode 4, the crew will demonstrate operability of each Containment Purge Supply and Exhaust Isolation valve by verifying the mechanical stops are ____(1)____. When mode 4 is established, the cumulative hours these valves are opened will be tracked using the attachment in _____(2)_____.

(1) (2)

A. removed OP-903-001, Technical Specification Surveillance Logs B. installed OP-903-001, Technical Specification Surveillance Logs C. installed OP-100-014, Technical Specification and Technical Requirements Compliance D. removed OP-100-014, Technical Specification and Technical Requirements Compliance Revision 0 Facility: Waterford 3 Page 33 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. The Containment Purge Supply and Exhaust Isolation valve mechanical stops are removed during refuel outages but must be re-installed prior to establishing containment integrity. Part 2 is correct.

B. CORRECT: Prior to establishing Containment Integrity for Mode 4, OP-010-003 states to demonstrate operability of each Containment Purge Supply and Exhaust Isolation valve by verifying the mechanical stops limit valve opening to < 52º open position. The cumulative purge hours are tracked using an attachment in OP-903-001. These hours are tracked to ensure compliance with Tech Specs.

C. Incorrect. Part 1 is correct. The purpose of OP-100-014 is to ensure compliance with TS and TRMs. Logging the containment purge hours are Tech Spec required but are not located here.

D. Incorrect. The Containment Purge Supply and Exhaust Isolation valve mechanical stops are removed during refuel outages but must be re-installed prior to establishing containment integrity. The purpose of OP-100-014 is to ensure compliance with TS and TRMs. Logging the containment purge hours are Tech Spec required but are not located here.

Technical Reference(s): OP-903-001 Attachment 11.6 Rev. 66 (Attach if not previously provided) OP-010-003 step 9.1.31 Rev. 342 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPN01 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2,5 Comments:

Revision 0 Facility: Waterford 3 Page 34 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 079 2.4.8 Importance Rating 4.5 K/A Statement 079 Station Air, 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Proposed Question: SRO 18 (Q93) Rev: 1 Given:

Instrument Air pressure is 60 psig and steady due to a leak The BOP has closed the Main Feed Isolation Valves due to low pressure alarms The crew will implement which of the following?

A. OP-902-006, Loss of Feedwater Recovery Procedure only B. OP-902-006, Loss of Feedwater Recovery Procedure and OP-901-511, Instrument Air Malfunction concurrently C. OP-902-001, Reactor Trip Recovery only D. OP-902-001, Reactor Trip Recovery and OP-901-511, Instrument Air Malfunction concurrently Revision 0 Facility: Waterford 3 Page 35 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: D Explanation: (Optional)

A. Incorrect. Loss of Main Feedwater Recovery Procedure (OP-902-006) is plausible since both MFIVs are shut isolating MFW from the S/Gs. The EOPs were recently revised such that closing both MFIVs does not put the crew into OP-902-006 B. Incorrect. OP-901-511 step 1 states to enter OP-902-000 concurrently with OP-901-511. The EOPs were recently revised such that closing both MFIVs does not put the crew into OP-902-006.

C. Incorrect. OP-901-511 step 1 states to enter OP-902-000 concurrently with OP-901-511.

D. CORRECT: The K/A system is Station Air. To have a complete loss of instrument air there must also be a loss of station air since the systems are cross-connected. OP-901-511 step 1 states to enter OP-902-000 concurrently with OP-901-511. The MFIVs were closed but there is still 100% capacity available from the EFW system. The crew will diagnose to OP-902-001, Reactor Trip Recovery. OP-100-017, EOP Implementation Guide, states that an off-normal procedure can be used to supplement an Optimal Recovery Procedure at CRS discretion. To answer the question, the applicant must have knowledge of an administrative procedure that provides guidance on coordination of abnormal and emergency operating procedures.

Technical Reference(s): OP-901-511, pages 6-9 Rev. 15 (Attach if not previously provided) OP-100-017 step 5.10.2 Rev. 4 (including version/revision number) OP-902-009 appendix 1 Rev. 315 Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Free review question. Changed the given conditions such that the applicant must determine that OP-901-511 entry conditions are met.

Revision 0 Facility: Waterford 3 Page 36 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.4 Importance Rating 3.8 K/A Statement 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no solo operation, maintenance of active license status, 10CFR55, etc.

Proposed Question: SRO 19 (Q94) Rev: 0 Per OI-024-000, Maintaining Active SRO/RO Status, to maintain your Senior Reactor Operator license in an ACTIVE status, you must stand a minimum of ______(1)_______

12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts in a calendar quarter. To upgrade from INACTIVE status to ACTIVE status, a license holder must stand a minimum of _______(2)________ hours of under instruction watches.

(1) (2)

A. five 40 B. seven 40 C. five 60 D. seven 60 Revision 0 Facility: Waterford 3 Page 37 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: The minimum number of watches (5) is correct per OI-024-000. To upgrade from inactive status to active status, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of under instruction watch is required.

B. Incorrect. Seven watches would be the minimum required in a quarter if Waterford SROs stood 8 hr shifts, per 10CFR55. Waterford SROs stand and only take credit for 12 hr shift. To upgrade from inactive status to active status, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of under instruction watch is required.

C. Incorrect. The minimum number of watches is correct per OI-024-000. To upgrade from inactive status to active status, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of under instruction watch is required.

60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> is consistent with the number of hours required to maintain active status.

D. Incorrect. Seven watches would be the minimum required in a quarter if Waterford ROs stood 8 hr shifts, per 10CFR55. Waterford SROs stand and only take credit for 12 hr shift. 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> is consistent with the number of hours required to maintain active status.

Technical Reference(s): OI-024-000 pages 9,10 Rev. 314 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPA00 obj. 3 (As available)

Question Source: Bank # RO66 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2011 NRC RO Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1,2 Comments:

Revision 0 Facility: Waterford 3 Page 38 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.45 Importance Rating 4.3 K/A Statement 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication.

Proposed Question: SRO 20 (Q95) Rev: 0 Regarding EOP usage, to verify the RCS is not water solid with Pressurizer level indicating 100%, the crew will verify that ____(1)____ . If the RCS is considered water solid, the crew will _____(2)_____.

(1) (2)

A. RCS inventory or temperature remain in the selected Optimal Recovery changes do not produce Procedure and control RCS temperature severe pressure responses and throttle RCS makeup B. RCS inventory or temperature transition to the Functional Recovery changes do not produce Procedure and establish a bubble in the severe pressure responses Pressurizer C. Reactor vessel plenum level is remain in the selected Optimal Recovery less than 100% Procedure and control RCS temperature and throttle RCS makeup D. Reactor vessel plenum level is transition to the Functional Recovery less than 100% Procedure and establish a bubble in the Pressurizer Revision 0 Facility: Waterford 3 Page 39 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: Optimal Recovery procedures (902-002, 004 and 007) and OI-038-000 (Operations Expectations procedure) includes guidance for verifying the RCS is not water solid. RCS inventory or temperature changes not producing severe pressure responses is listed as a diverse indication that the RCS is not water solid. Controlling RCS temperature and throttling HPSI flow is the contingency action for the RCS is water solid.

Therefore the crew can remain in the selected ORP and transitioning to OP-902-008 is not required.

B. Incorrect. Part 1 is correct. Controlling RCS temperature and throttling HPSI flow is the contingency action for the RCS is water solid. Therefore, transitioning to OP-902-008 is not required.

C. Incorrect. The crew is required to verify RV head level is less than 100% to check RCS is not water solid, not plenum level. Plausible because the applicant would have to know RVLMS gives both head and plenum level on QSPDS. RCS inventory or temperature changes not producing severe pressure responses is listed as a diverse indication that the RCS is not water solid. Part 2 is correct.

D. Incorrect. The crew is required to verify RV head level is less than 100% to check RCS is not water solid, not plenum level. Plausible because the applicant would have to know RVLMS gives both head and plenum level on QSPDS. Controlling RCS temperature and throttling HPSI flow is the contingency action for the RCS is water solid. Therefore, transitioning to OP-902-008 is not required.

Technical Reference(s): OP-902-004 step 31 revision 16 (Attach if not previously provided) TGOP-902-004 step 31 Rev. 307 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE04 obj. 7 (As available)

Question Source: Bank # SRO20 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2014 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 40 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.13 Importance Rating 4.3 K/A Statement 2.2.13 Knowledge of tagging and clearance procedures.

Proposed Question: SRO 21 (Q96) Rev: 0 When clearing the Main Steam Administrative (MS-Admin) tagout, the ____(1)____ must sign off as the tagout holder. The (MS-Admin) tagout must be cleared ____(2)____ all tagouts that use the MS-ADMIN tagout are cleared.

(1) (2)

A. outage manager before B. designated operations after supervisor C. outage manager after D. designated operations before supervisor Revision 0 Facility: Waterford 3 Page 41 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. The outage management responsibility per EN-OP-102 is to communicate to site personnel when MS-ADMIN tagout is hung, cleared or changes made. The MS-ADMIN tagout cannot be cleared until all tagouts using it are cleared.

B. CORRECT: The designated operations supervisor signs on and must sign off the MS-ADMIN tagout before it can be cleared. The MS-ADMIN tagout can be cleared once all tagouts using it are cleared.

C. Incorrect. The outage management responsibility per EN-OP-102 is to communicate to site personnel when MS-ADMIN tagout is hung, cleared or changes made. Part 2 is correct.

D. Incorrect. Part 1 is correct. The MS-ADMIN tagout cannot be cleared until all tagouts using it are cleared.

Technical Reference(s): EN-OP-102 step 5.22[1][2k][6] Rev. 18 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: FLP-OPS-ESOMS obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 42 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.12 Importance Rating 4.1 K/A Statement 2.2.12 Knowledge of surveillance procedures Proposed Question: SRO 22 (Q97) Rev: 0 OP-903-001, Technical Specification Surveillance Logs, provide the means for compliance with Operations surveillance requirements that have intervals with a maximum time period of < ___(1)____ days.

The taking of data for the daily day shift Technical Specification Logs should begin no earlier than _____(2)_____.

(1) (2)

A. 7 1200 PM B. 7 1100 AM C. 30 1100 AM D. 30 1200 PM Revision 0 Facility: Waterford 3 Page 43 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: B Explanation: (Optional)

A. Incorrect. Part 1 is correct. The dayshift logs are indicated on the log attachment as the 1200 logs. Allowance has been given to start them one hour early.

B. CORRECT: The purpose section of OP-903-001, Technical Specification Surveillance Logs (step 1.1), states that this procedure provides the means for compliance with Operations surveillance requirements that have intervals of < one week. Step 7.1.1 and 7.1.2 states that the daily day shift Tech Spec logs should begin no earlier than 1100 AM.

C. Incorrect. OP-903-001, Technical Specification Surveillance Logs, has an attachment for daily logs and 7 day logs. 30 day surveillance requirements are common but are tracked via work orders. Part 2 is correct.

D. Incorrect. OP-903-001, Technical Specification Surveillance Logs, has an attachment for daily logs and 7 day logs. 30 day surveillance requirements are common but are tracked via work orders. The dayshift logs are indicated on the log attachment as the 1200 logs. Allowance has been given to start them one hour early.

Technical Reference(s): OP-903-001 step 1.1, 7.1.1 and 7.2.1 Rev. 66 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-TS00 obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 44 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.11 Importance Rating 4.3 K/A Statement 2.3.11 Ability to control radiation releases Proposed Question: SRO 23 (Q98) Rev: 0 Which of the following are required actions per TRM 3.3.3.10, Radioactive Liquid Effluent, to release a Boric Acid or Waste Condensate Tank with its associated discharge Radiation Monitor Inoperable?

A. Ensure release rate calculations have been verified by two technically qualified personnel prior to the discharge and the discharge valve lineup is independently verified.

B. Ensure release rate calculations have been verified by two technically qualified personnel prior to the discharge and obtain grab samples every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during discharge.

C. Calculate release flow rate every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the discharge and the discharge valve lineup is independently verified.

D. Calculate release flow rate every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the discharge and obtain grab samples every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during discharge.

Revision 0 Facility: Waterford 3 Page 45 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: A Explanation: (Optional)

A. CORRECT: These actions are required per TRM 3.3.3.10 action 1b for an inoperable BM or LWM Radiation Monitor.

B. Incorrect. Part 1 is correct. Independent samples are required prior to the discharge, not during the discharge (action 1a)

C. Incorrect. Part 1 is for an inoperable flow instrument (action 2). Part 2 is correct.

D. Incorrect. Part 1 is for an inoperable flow instrument (action 2). Independent samples are required prior to the discharge but not every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the discharge (action 1a).

Technical Reference(s): TRM 3.3.3.10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-BM00 obj. 9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 4 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Revision 0 Facility: Waterford 3 Page 46 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.21 Importance Rating 4.6 K/A Statement 2.4.21 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 integrity, containment conditions, radioactivity release control, etc.

Proposed Question: SRO 24 (Q99) Rev: 0 Given:

The CRS is prioritizing safety functions in accordance with OP-902-008, Functional Recovery Procedure A SIAS, CIAS, MSIS and CSAS are present Two CEAs have failed to insert The ATC reports that Reactor Power is 10-5% and dropping The CRS is evaluating the Reactivity Control Safety Function

____(1)____ safety function acceptance criteria are met. The CRS will implement success path _____(2)_____ for Reactivity control.

(1) (2)

A. Only RC-1 RC-1 B. Only RC-1 RC-2 C. RC-1 and RC-2 RC-2 D. RC-1 and RC-2 RC-1 Revision 0 Facility: Waterford 3 Page 47 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. RC-1(Condition 2) is met due to power <10-4% and dropping. RC-2(Condition 1) is met because emergency boration is in progress due to the SIAS. Even though RC-1 criteria is met, RC-2 is implemented because RC-1 has no guidance for emergency boration and termination.

B. Incorrect. RC-1(Condition 2) is met due to power <10-4% and dropping. RC-2(Condition 1) is met because emergency boration is in progress due to the SIAS. Part 2 is correct.

C. CORRECT: RC-1(Condition 2) is met due to power <10-4% and dropping. RC-2(Condition 1) is met because emergency boration is in progress due to the SIAS. Even though RC-1 criteria is met, RC-2 is implemented because RC-1 has no guidance for emergency boration and termination. This question is TIER 3 because the logic is from OI-038-000, EOP Operations/Expectations Guidance and this logic is the same when prioritizing safety functions with more than one CEA stuck for any type of event.

D. Incorrect. Part 1 is correct. Even though RC-1 criteria is met, RC-2 is implemented because RC-1 has no guidance for emergency boration and termination.

Technical Reference(s): OI-038-000 step 5.4.32 Rev. 13 (Attach if not previously provided) OP-902-008 RC-1 and RC-2 SFSC Rev. 26 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: WLP-OPS-PPE01 obj. 4 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 48 of 50

2017 NRC Exam SRO Written Exam Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.30 Importance Rating 4.1 K/A Statement 2.4.30 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.

Proposed Question: SRO 25 (Q100) Rev: 0 Given:

An valid automatic reactor trip signal has failed to trip the reactor The manual reactor trip pushbuttons and Diversified Reactor Trip System (DRTS) also failed to trip the reactor The BOP has tripped the reactor by de-energizing the Motor Generator Sets The Emergency Planning classification for this event is a(n) ____(1)____ . The NRC shall be notified no later than _____(2)_____ of declaring the emergency.

(1) (2)

A. Alert one hour B. Alert 15 minutes C. Site Area Emergency one hour D. Site Area Emergency 15 minutes Revision 0 Facility: Waterford 3 Page 49 of 50

2017 NRC Exam SRO Written Exam Worksheet Proposed Answer: C Explanation: (Optional)

A. Incorrect. The required information for making this classification is located in the basis document for SS3. If the applicant does not know that the 32 feeder breakers are not part of the reactor console he/she would answer alert. Part 2 is correct B. Incorrect. The required information for making this classification is located in the basis document for SS3. If the applicant does not know that the 32 feeder breakers are not part of the reactor console he/she would answer alert. 15 minutes is the time limit for notifying Operation Hotline Members and Waterford 1 and 2 for an event classification C. CORRECT: The basis for SS3 states that tripping the reactor using the 32 feeder breakers is a Site Area Emergency (not part of the reactor control console). The NRC shall be notified within one hour of the classification. This question is TIER 3 because the emergency classification criteria is the same for any automatic trip failure. Only one page of the E-plan will be given because the applicant is being tested on his/her knowledge of information located in the basis document for SS3 (not his ability to locate SS3).

D. Incorrect. Part 1 is correct. 15 minutes is the time limit for notifying Operation Hotline Members and Waterford 1 and 2 for an event classification.

Technical Reference(s): EP-001-001 page 152 Rev 32 (Attach if not previously provided) EP-002-010 step 5.2.1.5 Rev 314 (including version/revision number)

Proposed references to be provided to applicants during examination: EP-001-001 page 22 Rev 32 Learning Objective: WLP-OPS-EP02 objs. 9 and 17 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Revision 0 Facility: Waterford 3 Page 50 of 50

ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

Date: April 5, 2017 Facility/Unit: Waterford 3 Region: IV Reactor Type: CE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with a 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results RO/SRO-Only/Total Examination Values 75 / 25 / 100 Points Applicants Score / / Points Applicants Grade / / Percent

2017 NRC Written Examination Waterford 3 Senior Reactor Operator

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2017 NRC Written Examination Waterford 3 Senior Reactor Operator

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WATERFORD 3 SES OP-902-002 Revision 020 Page 27 of 83 LOSS OF COOLANT ACCIDENT RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Maintain RCS within RCS PT Limits

• 30. Maintain the RCS within the limits of Attachments 2A-D, "RCS Pressure and Temperature Limits" by performing ANY of the following:

a. Control Main or Auxiliary PZR spray.

b. Control PZR heaters. b.1 Restore PZR heaters using Appendix 25, "Restore Pressurizer Heater Control."

c. IF HPSI throttle criteria are met (Step 23),

THEN perform ANY of the following:

Control Charging and Letdown.

Throttle HPSI flow as necessary.

d. Control RCS cooldown rate within d.1 Lower the cooldown rate as allowed limits. necessary to restore within limits.

e. Check RCS pressure less than the e.1 Stop the cooldown.

upper limits of PT curves.

e.2 Lower RCS pressure using Main or Auxiliary spray.

e.3 IF HPSI throttle criteria met, THEN control charging, letdown, or HPSI flow as necessary.

e.4 IF necessary, THEN operate PZR vents using Appendix 38, Pressurizer Vent Operations.

e.5 WHEN RCS pressure within limits, THEN continue cooldown.

WATERFORD 3 SES OP-902-009 Revision 315 Page 7 of 206 STANDARD APPENDICES Attachment 2-A Page 1 of 1 2.0 Figures Attachment 2-A: RCS Pressure and Temperature Limits End of Attachment 2-A

WATERFORD 3 SES OP-902-009 Revision 315 Page 8 of 206 STANDARD APPENDICES Attachment 2-B Page 1 of 1 Figures -B: RCS Pressure and Temperature Limits (Non Harsh Expanded)

End of Attachment 2-B

WATERFORD 3 SES OP-902-009 Revision 315 Page 9 of 206 STANDARD APPENDICES Attachment 2-C Page 1 of 1 Figures -C: RCS Pressure and Temperature Limits (Harsh)

End of Attachment 2-C

WATERFORD 3 SES OP-902-009 Revision 315 Page 10 of 206 STANDARD APPENDICES Attachment 2-D Page 1 of 1 Figures -D: RCS Pressure and Temperature Limits (Harsh Expanded)

End of Attachment 2-D

WATERFORD 3 SES OP-902-009 Revision 315 Page 147 of 206 STANDARD APPENDICES Attachment 25-A Page 1 of 2 25.0 Restore Pressurizer Heater Control Attachment 25-A: Restore Train A Pressurizer Heaters INSTRUCTIONS

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NOTE -------------------------------------------------------------

SIAS actuation locks out the 32 Bus Feeder breaker close signal while the Sequencer is operating.

When the EDG is energizing the 3 Bus, the 32 Bus Feeder breaker will only auto close during Sequencer operation.

1. Verify Pressurizer level is greater than 33% [50%].

2. IF the 32A Bus Feeder breaker is open, THEN perform the following:

a. Check SEQUENCER NOT operating or has timed out.

b. IF 3A Bus energized from Off-Site Power, THEN close SST A32 FEEDER breaker at CP-1.

c. IF 3A Bus energized from EDG A, THEN perform the following:

1) Monitor EDG loading to ensure EDG does not exceed 4 MW.

2) IF SIAS has actuated, THEN reset SIAS using Appendix 5, ESFAS Reset.

3) Place SEQUENCER A switch to TEST and THEN release.

(continue)

WATERFORD 3 SES OP-902-009 Revision 315 Page 149 of 206 STANDARD APPENDICES Attachment 25-B Page 1 of 3 Restore Pressurizer Heater Control Attachment 25-B: Restore Train B Pressurizer Heaters INSTRUCTIONS

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NOTE -------------------------------------------------------------

SIAS actuation locks out the 32 Bus Feeder breaker close signal while the Sequencer is operating.

When the EDG is energizing the 3 Bus, the 32 Bus Feeder breaker will only auto close during Sequencer operation.

1. Verify Pressurizer level is greater than 33% [50%].

2. IF the 32B Bus Feeder breaker is open, THEN perform the following:

a. Check SEQUENCER NOT operating or has timed out.

b. IF 3B Bus energized from Off-Site Power, THEN close SST B32 FEEDER breaker at CP-1.

c. IF 3B Bus energized from EDG B, THEN perform the following:

1) Monitor EDG loading to ensure EDG does not exceed 4 MW.

2) IF SIAS has actuated, THEN reset SIAS using Appendix 5, ESFAS Reset.

3) Place SEQUENCER B switch to TEST and THEN release.

(continue)

Off Normal Procedure OP-901-131 Shutdown Cooling Malfunction Revision 304 E2. LOSS OF SHUTDOWN COOLING FLOW CAUTION DO NOT START LPSI PUMP UNLESS PROPERLY VENTED.

PLACEKEEPER START DONE

1. IF ANY Shutdown Cooling Loop Suction Isolation valves close on the operating Shutdown Cooling train, THEN perform the following:

1.1 Stop affected LPSI Pump.

1.2 Start standby Shutdown Cooling Train in accordance with OP-009-005, SHUTDOWN COOLING SYSTEM.

2. IF ONE LPSI Pump was operating AND tripped, THEN place standby Shutdown Cooling Train in service in accordance with OP-009-005, SHUTDOWN COOLING SYSTEM.

3. IF BOTH LPSI Pumps are operating AND ONE trips, THEN control RCS temperature using in service Shutdown Cooling Train.

4. Locally vent LPSI Pump suction piping until all air is removed.

5. IF the loss of Shutdown Cooling flow is the result of a loss of electric power, THEN implement one of the following concurrently with this procedure:

OP-901-310, LOSS OF 4160 VOLT SAFETY BUS A OP-901-311, LOSS OF 4160 VOLT SAFETY BUS B OP-902-005, STATION BLACKOUT RECOVERY

6. IF ANY of the following LPSI Pump cavitation/Air binding indications occur, THEN stop affected LPSI Pump:

Dropping OR erratic ammeter indication Dropping OR erratic Shutdown Cooling flow Steady low flow AND amperage less than expected for system configuration Local observation 26

Off Normal Procedure OP-901-504 Inadvertent ESFAS Actuation Revision 010 E3 INADVERTENT RAS PLACEKEEPER START DONE N/A

1. If LPSI Pump(s) has tripped during Shutdown Cooling Operation, then perform the following:

1.1 Restart LPSI Pump(s).

1.2 Implement OP-901-131, Shutdown Cooling Malfunction, concurrently with this procedure.

2. Verify RWSP Level > 10%.

3. Investigate cause of inadvertent RAS and initiate corrective action.

4. Advise SM/CRS to initiate a Condition Report in accordance with EN-LI-102, Corrective Action Process.

5. Reset RAS Initiation relays on all four channels as follows:

5.1 On CP-10 place RESET PERMISSIVE key switch to UNLK.

5.2 Depress RAS Reset pushbutton.

5.3 Verify initiation relay indicator Illuminated on ENGINEERED SAFETY FEATURES SYSTEM mimic on CP-10:

Channel A (Red Lamp)

Channel B (Yellow Lamp)

Channel C (Green Lamp)

Channel D (Blue Lamp) 5.4 Place Reset Permissive switch to LK position.

21

SD-SI LPSI pump flow is limited to > 100 GPM but < 2000 GPM for 3 consecutive hours contact the system engineer for guidance on pump monitoring. Prolonged operation <

2000 GPM can increase thrust bearing degradation.

LPSI Pump Controls and Interlocks The LPSI pumps receive actuation signals from the Engineered Safety Features (ESF) actuation system. Upon receipt of a SIAS actuation the pumps auto start on the 17 second sequencer block. If operating in response to a SIAS and a loss of off-site power occurs the pumps will automatically restart when power is restored by the emergency diesel generators.

Pump status is monitored in the control room utilizing the same red/green/amber lighting scheme as the HPSI pumps. The pumps can be controlled remotely from CP-8 or from the remote shutdown panel (LCP-43, See Figure 5). The remote shutdown panel control features are only available if the auxiliary panel control transfer switch is in the AUX CONTRL PANEL position.

The control switches on CP-8 have positions of OFF/START OVRD RAS and are spring return to midposition from the START OVRD RAS position. These control switches are maintain-contact in the OFF position. An ammeter is provided with each control switch on CP-8. These meters are used to determine if the pumps are cavitating during Shutdown Cooling operations. The control switches on LCP-43 are different in that they have positions of STOP/START OVRD RAS and are spring return to midposition from both positions. Also, no ammeters are provided on LCP-43.

An RAS will stop the LPSI pumps. At the minimum time at which an RAS can occur (20 minutes) one HPSI pump is capable of providing adequate core cooling. The recirculation valves are manually shut by the operator following an RAS. In order to protect the LPSI pumps from operating at less than minimum flow rates and to prevent vortexing in the SI Sump they are automatically stopped. The RAS signal can be overridden by taking the pump control switch to the START RAS OVRD position after 1 second.

Minimum Flow Recirculation Valves, SI-1161A(B)

As previously described the safety injection pumps are provided with a recirculation flow path back to the RWSP. For the LPSI pumps the minimum recirculation flow rate is 100 GPM. In addition to the common motor operated recirculation header stop valves utilized by the HPSI, LPSI, and CS systems the LPSI pumps are provided with separate solenoid operated stop valves SI-1161A(B).

Minimum Flow Recirc Stop Check Valve, SI-116A(B)

Revision 15 Page 27 of 83

WATERFORD 3 SES OP-902-004 Revision 016 Page 15 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS EFW Pump AB Steam Line Break

• 15. IF BOTH SGs are equally affected, AND suspect that the break is in the EFW Pump AB Steam Supply line, THEN perform the following:

a. Close BOTH Steam Supply to EFW Pump AB Isolation valves:

MS 401A, PUMP AB TURB STM SUPPLY SG 1 MS 401B, PUMP AB TURB STM SUPPLY SG 2

b. Monitor SG pressures to determine if break has been isolated.

c. IF closing Steam Supply to EFW Pump AB Isolation valves isolates the break, THEN GO TO Step 19.

d. IF closing Steam Supply to EFW Pump AB Isolation valves has had no effect on SG pressures, THEN open BOTH Steam Supply to EFW Pump AB Isolation valves:

MS 401A, PUMP AB TURB STM SUPPLY SG 1 MS 401B, PUMP AB TURB STM SUPPLY SG 2

WATERFORD 3 SES OP-902-004 Revision 016 Page 16 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Isolate Most Affected SG with ESD

• 16. Isolate the MOST AFFECTED SG as follows:

Steam Generator 1

a. Verify MS 124A, MSIV 1 is closed.

b. Verify FW 184A, MFIV 1 is closed.

c. Verify MS 116A, ADV 1 is closed and the controller in manual.

d. Verify EFW Isolation valves are closed:

EFW 228A, SG 1 PRIMARY EFW 229A, SG 1 BACKUP

e. Place EFW FLOW CONTROL valves in MAN and THEN close:

EFW 224A, SG 1 PRIMARY EFW 223A, SG 1 BACKUP

f. Close MS 401A, PUMP AB TURB STM SUPPLY SG 1.

g. Close Main Steam Line Drains:

MS 120A, NORMAL MS 119A, BYPASS (continue)

WATERFORD 3 SES OP-902-004 Revision 016 Page 17 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

• 16. (continued)

h. Verify SG Blowdown Isolation valves are closed:

BD 103A, STM GEN 1 (OUT)

BD 102A, STM GEN 1 (IN)

i. Check SG1 West Side Main Steam Safety valves are closed.

Steam Generator 2

a. Verify MS 124B, MSIV 2 is closed.

b. Verify FW 184B, MFIV 2 is closed.

c. Verify MS 116B, ADV 2 is closed and the controller in manual.

d. Verify EFW Isolation valves are closed:

EFW 228B, SG 2 PRIMARY EFW 229B, SG 2 BACKUP

e. Place EFW FLOW CONTROL valves in MAN and THEN close:

EFW 224B, SG 2 PRIMARY EFW 223B, SG 2 BACKUP

f. Close MS 401B, PUMP AB TURB STM SUPPLY SG 2.

(continue)

WATERFORD 3 SES OP-902-004 Revision 016 Page 18 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

• 16. (continued)

g. Close Main Steam Line Drains:

MS 120B, NORMAL MS 119B, BYPASS

h. Verify SG Blowdown Isolation valves are closed:

BD 103B, STM GEN 2 (OUT)

BD 102B, STM GEN 2 (IN)

i. Check SG2 East Side Main Steam Safety valves are closed.

WATERFORD 3 SES OP-902-004 Revision 016 Page 19 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Verify Correct SG Isolated

• 17. Verify the most affected SG is isolated 17.1 IF the wrong SG was isolated, by evaluating ALL of the following: THEN restore feeding and steaming capability to the isolated SG.

SG pressures 17.2 WHEN RCS Heat Removal has been SG levels re-established on the least affected SG, THEN REFER TO step 16 and isolate RCS Cold Leg temperatures the most affected SG.

WATERFORD 3 SES OP-902-004 Revision 016 Page 20 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

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NOTE ----------------------------------------------------------

Actions to stabilize RCS temperature following an excess steam demand event should be initiated when BOTH of the following parameters are met:

CET temperatures rise Pressurizer pressure rise Stabilize RCS Temperature with SG ESD

• 18. Stabilize RCS temperature within PT Curves by performing the following:

a. For the LEAST AFFECTED SG:

1) Place the ADV to manual and fully open ADV.

2) Manually initiate EFAS.

3) Place the EFW Flow Control valve to manual and commence feeding.

4) Perform ANY of the following as necessary to establish RCS pressure and temperature control:

Throttle associated SG ADV.

Adjust associated SG EFW flow.

b. IF RCS pressure is greater than or equal b.1 IF RCS pressure is less than to 1500 psia, 1500 psia, THEN stabilize RCS pressure at a value THEN stabilize RCS not to exceed 1600 psid between the pressure at greater than RCS and the lowest SG pressure. HPSI shutoff head (1500 -

1600 psia).

c. REFER TO Step 20, HPSI Throttle Criteria.

WATERFORD 3 SES OP-902-005 Revision 020 Page 17 of 43 STATION BLACKOUT RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

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NOTE ----------------------------------------------------------

Declaration of an ELAP event should be considered when recovery time from a Loss of All AC Power is unknown or may exceed Design License Basis SBO of four hours.

CAUTION In order to preserve a Station Battery to meet ELAP requirements, an ELAP must be declared and FIG-001 entered within one hour of the onset of SBO to ensure that the Station Battery deep load shed time requirements are met.

ELAP Event in Progress

• 17. IF a Loss of All AC power event may exceed the License Basis SBO of four hours, THEN within ONE hour of the onset of SBO, perform BOTH of the following:

a. Declare an ELAP event in progress.

b. Commence FIG-001, Extended Loss of AC Power in conjunction with this procedure.

Technical Guide for Station Blackout TG-OP-902-005 Recovery Procedure Revision 309 Step Number 17 ELAP Event in Progress Note and Caution Objective The intent of the note and caution is to inform the crew that an ELAP event should be considered when a Loss of All AC power may exceed the License Basis SBO of four hours. An ELAP must be commenced within one hour of the SBO to meet the ELAP time requirements.

Instructions This step contains no instructions.

Contingency Actions This note contains no contingency actions.

Justification for Deviations This note notifies the crew that if it is determined that the loss of all AC power event may exceed the License Basis SBO of four hours then an ELAP should be considered. An ELAP must be declared and commenced within one hour of a loss of all AC to meet ELAP time requirements. This note is provided to assist the Operator with the determination of an ELAP event. If an ELAP event is determined to be in progress then it is acceptable to declare an ELAP and commence implementation before the before the one hour point is reach. This will provide for events where extensive damage has occurred and restoration of AC power is not likely to happen until well past the one hour time frame. Timelier implementation of the FLEX strategy will aid in being able to maintain the key safety functions of core cooling, spent fuel pool cooling and containment integrity.

References

1. NRC Order EA- 12-049, Issuance Of Order To Modify Licenses With Regard To Requirements For Mitigation

2. Engineering Report WF3-SA-14-00002, Waterford 3 FLEX Strategy Development

3. EC 48142, Waterford 3 Engineering Change, FLEX Strategy Development

4. FLEX Implementing Guideline FIG-001, Extended Loss of AC Power

5. FSAR 8.1.4.1 Offsite Power System

6. FSAR Appendix 8.1A Station Blackout Evaluation

7. NEI 12-06 Diverse and Flexible Coping Strategies (FLEX) Implementation Guide 44

Technical Guide for Station Blackout TG-OP-902-005 Recovery Procedure Revision 309 Step Number 17 ELAP Event in Progress (contd)

Consistent with 10 CFR 50.54 (x), a licensee may take reasonable action that departs from a license condition or a technical specification in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with the license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent. Declaration of an ELAP event will place the plant outside of the existing design basis, allowing emergency actions to protect public health and safety and may place the plant in a condition where it cannot comply with certain Technical Specifications, and, as such may warrant invocation of 10 CFR 50.54(x).

A three phase flexible coping strategy has been developed with the goal of preventing damage to the fuel in the reactor and spent fuel pool and to maintain the containment function. The first phase (Phase 1) relies on installed equipment that has been determined to be robust. The second phase (Phase 2) utilizes on-site FLEX equipment to supplement installed plant equipment. The third phase (Phase 3) credits off-site equipment to obtain additional capabilities and redundancy.

Instructions This step provides the direction for implementing the FLEX strategy. Waterford 3 Site Specific Assumptions listed in WF3-SA-14-00002 assume that ELAP will be declared within approximately 60 minutes in order to enable actions which place the plant outside of the current design and licensing basis.

Contingency Actions None Justification for Deviations The EPG incorporated the FLEX strategy steps into the SBO guideline. Waterford developed a FLEX Implementation Guideline (FIG-001, Extended Loss of AC Power) which contains the FLEX strategy for controlling the plant and to initiate the applicable FSGs based on current plant conditions. This meets the intent of the EPG and merely changes the way the information is presented.

46

WATERFORD 3 SES OP-902-009 Revision 315 Page 187 of 206 STANDARD APPENDICES Attachment 33-B Page 1 of 5 Generator Auxiliary Operations Attachment 33-B: Generator Auxiliary Operations - LOOP (Both Trains)

INSTRUCTIONS

---

NOTE -----------------------------------------------------------

With turbine building battery chargers not energized and no Operator action, the maximum discharge time for the Turbine Building battery is 90 minutes.

1. Vent Hydrogen from the Main generator as follows:

a. Close HG-405, Generator Hydrogen Regulator Outlet Isol valve.

b. Close GG-101, Generator Gas Hydrogen Supply Isolation valve.

c. Verify GG-103, Generator Gas Low Point Vent Isolation valve closed.

d. Open GG-102, Generator Gas High Point Vent Isolation valve.

2. Verify ALL of the following DC Oil pumps are operating:

Main Turbine Emergency Oil Pump (DC) at CP-1 Air Side Seal Oil Backup Pump (DC)

FWPT A Emergency Oil Pump (DC)

FWPT B Emergency Oil Pump (DC)

3. Secure the Main Feed Pump Emergency Oil pumps as follows:

WHEN Main Feed Pump A speed is 0 RPM, THEN open LOF-EBKR-TGB-9, FWPT A Emergency Oil Pump (DC) breaker.

WHEN Main Feed Pump B speed is 0 RPM, THEN open LOF-EBKR-TGB-10, FWPT B Emergency Oil Pump (DC) breaker.

4. WHEN Generator Hydrogen pressure is less than 2 psig (GG-IPI-3008, GCM Dschg),

THEN open SO-EBKR-TGB-6, Air Side Seal Oil Backup Pump (DC) breaker.

Technical Guide for Standard Appendices TG-OP-902-009 Revision 311 Appendix 33 Generator Auxiliary Operations Objective The intent of this Standard Appendix is to provide instructions to protect the main generator and preserve the TGB batteries by securing the major DC powered motor loads in the event that the TGB battery chargers are not energized. This attachment was developed following INPO IER L2-12-27 and CR-WF3-2012-1929. With a loss of off site power the TGB battery chargers are lost and turbine auxiliary DC operated pumps will be started to protect equipment. Based on the information provided in ECE93-004 without operator action the TGB batteries could reach maximum discharge in 90 minutes. If main generator hydrogen is being maintained by the DC seal oil pump and the pump does not function than the hydrogen will vent thru the seals. Therefore with a loss of off-site power action needs to be taken to transfer seal oil to AC powered pumps or to vent the main generator and secure the DC seal oil pump. These actions are performed to reduce the risk of a hydrogen explosion in the event that all main turbine seal oil is lost.

Preserving the TGB Battery preserves remote operation of the 2 bus feeder breakers and the 1 and 2 bus load breakers which allows for off-site power to be more efficiently and safely restored. Off-site power is provided to the safety 4.16 KV buses through the non-safety 2 buses. Additional important loads that are supported by the TGB battery are the instrument air compressor un-loader valves and CMU makeup valves. The functioning of these components is required for production of instrument air. The maintenance of instrument air will assist the operator with restoring the plant to a normal operating configuration and help ensure a smoother transition out of EOPs.

This appendix will provide direction to secure the MFWPT DC emergency oil pumps, the DC seal oil backup pump and the main turbine emergency bearing oil pump. This appendix has been divided into 3 conditions: Station Blackout, Loss of off-site power (both trains) and Loss of a single train of off-site power.

The steps in each attachment are presented in the order that they are most likely to be performed however some steps can be performed when the conditions for the step are met. These steps should be pulled forward and performed when resources and time is available.

86

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC CAUTION (1) THE MAIN TURBINE AND MAIN FEEDWATER PUMP TURBINE (FWPTS) DO NOT HAVE DC LUBE OIL PUMPS AVAILABLE DUE TO LOSS OF BUS TGB-DC.

EXTREME CARE MUST BE TAKEN TO ENSURE AC LUBE OIL PUMPS DO NOT LOSE POWER UNTIL TURBINE SHAFTS STOP. POWER SUPPLIES FOR AC LUBE OIL PUMPS ARE:

(2) ALL NON-SAFETY 7KV, 4KV SWITCHGEAR ARE LEFT WITHOUT CONTROL POWER. ALL REMOTE MANUAL CONTROL AND AUTOMATIC PROTECTION OF THESE LOAD CENTERS AND CONNECTED COMPONENTS ARE DISABLED.

MAIN TURBINE AC BEARING OIL PUMP LOG-EBKR-313AB-2M FWPT A MOP 1 LOF-EBKR-211A-5D FWPT B MOP 1 LOF-EBKR-212A-5D FWPT A MOP 2 LOF-EBKR-212B-5D FWPT B MOP 2 LOF-EBKR-211B-5D (3) 480V BREAKERS WILL HAVE OVERCURRENT PROTECTION FROM ECS, BUT ARE LEFT WITHOUT CONTROL POWER.

(4) IF TGB-DC POWER CANNOT BE RESTORED, THEN ALL MAIN STEAM BYPASS VALVES CANNOT BE OPENED FROM CP-1 AND WILL NOT OPEN ON A REACTOR TRIP OR CUTBACK.

PLACEKEEPER START DONE N/A

1. Verify Automatic Actions (Section C) take place as designed.

2. Start Emergency Diesel Generators A and B on CP-1.

3. Manually Close Condensate Pump A, B and C recirculation valves CD 137A, CD 137B, and CD 137C.

33

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC (CONTD)

NOTE Buses 2A and 2B will not transfer to Startup Transformers from CP-1 due to the loss of control power to their supply breakers.

PLACEKEEPER START DONE N/A

4. Transfer bus 1A supply from Unit Auxiliary Transformer (UAT) A to Startup Transformer (SUT) A by positioning Bus A Transfer selector switch (CP-1) to SUT.

5. Transfer bus 1B supply from UAT B to SUT B by positioning Bus B Transfer selector switch (CP-1) to SUT.

6. Verify Main Generator Breaker A and Main Generator Breaker B (CP-1) indicate closed.

7. Transfer bus 2A supply from UAT A to SUT A locally at 2A switchgear as follows:

7.1 Close Startup Transformer A 4KV Isolation breaker (4KV-EBKR-2A-4) by depressing the MANUAL CLOSE pushbutton.

7.2 Open Feeder From Unit Auxiliary Transformer A breaker, 4KV-EBKR-2A-1, by depressing MANUAL TRIP pushbutton.

8. Transfer bus 2B supply from UAT B to SUT B locally at 2B switchgear as follows:

8.1 Close Startup Transformer B 4KV Isolation breaker, 4KV-EBKR-2B-4, by depressing MANUAL CLOSE pushbutton 8.2 Open Feeder From Unit Auxiliary Transformer B breaker, 4KV-EBKR-2B-1, by depressing MANUAL TRIP pushbutton.

34

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC (CONTD)

CAUTION MAIN FEEDWATER PUMP AUTOMATIC TRIPS, WITH THE EXCEPTION OF MECHANICAL OVERSPEED, ARE LOST. MAIN FEEDWATER PUMP GOVERNOR CONTROL VALVE ON CP-1 WILL INDICATE GREEN (LESS); HOWEVER, MAIN FEEDWATER PUMP WILL NOT BE IN MANUAL.

PLACEKEEPER START DONE N/A Continuous

9. Maintain discharge pressure of both Main Feedwater Pumps approximately 100 psia greater than Steam Generator pressure by adjusting SGFP Speed Controller FW-IHIC-1107, FW-IHIC-1108 in Manual using indications on CP-1.

Continuous

10. When the cause of the TGB-DC bus loss is found and corrected, then restore TGB-DC in accordance with OP-006-003, 125V DC electrical distribution system operations and perform the following:

10.1 Transfer buses 1A and 2A from the Startup Auxiliary Transformer (SUT) to Unit Auxiliary Transformer (UAT) by positioning the Bus A Transfer selector switch (CP-1) to UAT.

10.2 Transfer buses 1B and 2B from the Startup Auxiliary Transformer (SUT) to Unit Auxiliary Transformer (UAT) by positioning the Bus B Transfer selector switch (CP-1) to UAT.

10.3 Secure Emergency Diesel Generators A and B in accordance with OP-009-002, Emergency Diesel Generators.

10.4 Restore the Condensate Pump A, B, and C recirculation valves to automatic.

10.5 Restore both Main Feedwater Pumps SGFP Speed Controllers SG-IHIC-1107, FW-IHIC-1108 to automatic.

35

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC (CONTD)

CAUTION AUTO VOLTAGE REGULATOR OPERATION MAY BE ERRATIC OR LOST. INFORM PINE BLUFF SYSTEM OPERATIONS CENTER THAT WF3 WILL BE COMING OFF LINE AND MAY EXPERIENCE AUTO VOLTAGE REGULATOR PROBLEMS.

PLACEKEEPER START DONE N/A 10.6 Verify all equipment serviced by TGB-DC bus is restored.

10.7 Exit this procedure.

11. If bus TGB-DC cannot be re-energized, then conduct a Rapid Plant Power Reduction in accordance with OP-901-212 and cooldown to Mode 5 in accordance with OP-010-005, Plant Shutdown, in conjunction with this procedure.

Continuous

12. If a Reactor Trip occurs, initiate a MSIS, then refer to OP-902-000, Standard Post Trip Actions, and perform concurrently with this procedure.

CAUTION (1) POWER WILL BE LOST TO ALL STATION AIR COMPRESSORS AND INSTRUMENT AIR COMPRESSORS UNLOADER VALVES. THIS WILL RESULT IN A LOSS OF INSTRUMENT AIR AND STATION AIR.

(2) THE MAIN TURBINE CANNOT BE TRIPPED FROM CP-1 AND CAN ONLY BE TRIPPED LOCALLY.

(3) THE MAIN FEEDWATER PUMPS WILL NOT GO ON THE TURNING GEAR.

(4) IF TGB-DC POWER CANNOT BE RESTORED, THEN ALL MAIN STEAM BYPASS VALVES CANNOT BE OPENED FROM CP-1 AND WILL NOT OPEN ON A REACTOR TRIP OR CUTBACK.

Continuous

13. If Instrument Air pressure drops to 65 psig, then Trip the Reactor, initiate a MSIS, and refer to OP-902-000, Standard Post Trip Actions, and perform concurrently with this procedure.

36

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC (CONTD)

PLACEKEEPER START DONE N/A Continuous

14. If CCW is lost to Reactor Coolant Pumps for more than 3 minutes, then Trip the Reactor, initiate a MSIS, and secure Reactor Coolant Pumps. Refer to OP-902-000, Standard Post Trip Actions, and perform concurrently with this procedure:

14.1 Place control switches for the following RCP Seal Cooler isolation valves to closed to prevent thermal shock to RCP Seals when Instrument Air is restored:

RCP 1A Seal Cooler Inlet/Outlet Isol Valves CC-6651A and CC-679A RCP 1B Seal Cooler Inlet/Outlet Isol Valves CC-6651B and CC-679B RCP 2A Seal Cooler Inlet/Outlet Isol Valves CC-666A and CC-680A RCP 2B Seal Cooler Inlet/Outlet Isol Valves CC-666B and CC 680B.

CAUTION TECHNICAL SPECIFICATIONS PRESSURIZER LEVEL LIMITS, MODES 1-3 ARE 26% TO 62.5%. [T.S. 3.4.3.1]

Continuous

15. If Letdown isolation valves CVC 101, CVC 103, or CVC 109 fail closed, then cycle Charging pump(s) as necessary to maintain Pressurizer level 28% to 62% or as specified by SM/CRS.

37

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC (CONTD)

NOTE Temperature Control Valves for Turbine Cooling Water (TCW) System fail open on loss of Instrument Air.

PLACEKEEPER START DONE N/A Continuous

16. Throttle TCW discharge valves for the following Heat Exchangers to maintain desired temperature:

Main Turbine Lube Oil Reservoir TC-302 Hydrogen Coolers TC-350 Stator Coolers TC-334 EHC Heat Exchangers TC-315 Main Feedwater Pump Turbine Oil Coolers TC-319A and TC-319B.

NOTE Reactor Coolant Pumps (RCPs) 1B and 2B will be the two RCPs left running.

17. If Reactor is shutdown, then, at SM/CRS discretion, Open the following breakers on 7KV bus 1A:

Circulating Water Pump A, CW-EBKR-1A-5 Circulating Water Pump C, CW-EBKR-1A-6 Reactor Coolant Pump 1A, RC-EBKR-1A-7A Reactor Coolant Pump 2A, RC-EBKR-1A-8A Condensate Pump A, CD-EBKR-1A-9A Condensate Pump C, CD-EBKR-1A-10.

38

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC (CONTD)

PLACEKEEPER START DONE N/A

18. At SM/CRS discretion Open the following breakers on 4KV bus 2A:

Heater Drain Pump A, FHD-EBKR-2A-5 Heater Drain Pump C, FHD-EBKR-2A-6 Turbine Cooling Water Pump A, TC-EBKR-2A-9 Switchgear 22A Feeder, SSD-EBKR-2A-10 Switchgear 4A Feeder, 4KV-EBKR-2A-11.

CAUTION DEENERGIZING MCC 215A WILL CAUSE A LOSS OF THE AIR SIDE SEAL OIL PUMP IF IT IS ALIGNED TO THE NORMAL SOURCE POWER SUPPLY IN ACCORDANCE WITH OP-003-023, SEAL OIL.

19. At SM/CRS discretion Open the following breakers on 480V bus 21A:

MCC 213A Feeder, SSD-EBKR-21A-5A MCC 215A Feeder, SSD-EBKR-21A-7A Condenser Vacuum Pump C, AE-EBKR-21A-7B1 Condenser Vacuum Pump A, AE-EBKR-21A-7C1 MCC 214A Feeder, SSD-EBKR-21A-8C.

39

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC (CONTD)

PLACEKEEPER START DONE N/A

20. Verify the following breakers Closed on MCC 211A and Open all others:

SUT A Normal Cooling, ST-EBKR-211A-3BR Lighting Panel 30 Feeder, LTN-EBKR-211A-4BL FWPT A Main Oil Pump (MOP) 1, LOF-EBKR-211A-5D.

21. Verify the following breakers Closed on MCC 212A and Open all others:

UAT B Alternate Cooling, ST-EBKR-212A-3BR Lighting Panel 32 Feeder, LTN-EBKR-212A-4B FWPT B Main Oil Pump, LOF-EBKR-212A-5D.

22. At SM/CRS discretion Open the following breakers on 7KV bus 1B:

Circulating Water Pump B, CW-EBKR-1B-5 Circulating Water Pump D, CW-EBKR-1B-6 Condensate Pump B, CD-EBKR-1B-9.

23. At SM/CRS discretion Open the following breakers on 4KV bus 2B:

Heater Drain Pump B, FHD-EBKR-2B-5 Turbine Cooling Water Pump B, TC-EBKR-2B-6 Switchgear 22B Feeder, SSD-EBKR-2B-9 Switchgear 4B Feeder, 4KV-EBKR-2B-10.

40

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC (CONTD)

CAUTION DEENERGIZING MCC 215B WILL CAUSE A LOSS OF THE AIR SIDE SEAL OIL PUMP IF IT IS ALIGNED TO THE EMERGENCY SOURCE POWER SUPPLY IN ACCORDANCE WITH OP-003-023, SEAL OIL.

PLACEKEEPER START DONE N/A

24. At SM/CRS discretion Open MCC 215B Feeder, SSD-EBKR-21B-7A.

25. Open the following breakers on 480V bus 21B:

MCC 213B Feeder, SSD-EBKR-21B-5A Turbine Building Crane, CRN-EBKR-21B-5C Makeup Demineralizer Caustic Dilution Water Tank Heater, SSD-EBKR-21B-6C Condenser Vacuum Pump B, AE-EBKR-21B-7C1 MCC 214B Feeder SSD-EBKR-21B-8C.

26. Verify the following breakers on 480V MCC 211B Closed and Open all others:

SUT B Normal Cooling, ST-EBKR-211B-3BR Lighting Panel 31 Feeder, LTN-EBKR-211B-4BL FWPT B Main Oil Pump 2, LOF-EBKR-211B-5D.

27. Verify the following breakers on 480V MCC 212B Closed and Open all others:

SUT A alternate cooling, ST-EBKR-212B-3BL FWPT A Main Oil Pump 2, LOF-EBKR-212B-5D.

41

Off Normal Procedure OP-901-313 Loss of a 125 Volt DC Bus Revision 305 E4 LOSS OF 125 VOLT DC BUS TGB-DC (CONTD)

PLACEKEEPER START DONE N/A

28. Verify the following breakers on 480V MCC 313AB Closed and Open all others:

Main Turbine Bearing Lift Oil Pump, LOG-EBKR-313AB-4D Turbine Turning Gear, TUR-EBKR-313AB-4M Main Turbine Seal Oil Backup Pump, LOG-EBKR-313AB-3D Main Turbine AC Bearing Oil Pump, LOG-EBKR-313AB-2M.

29. When the cause of the TGB-DC bus loss is found and corrected, then restore TGB-DC in accordance with OP-006-003, 125V DC Electrical Distribution System Operation.

END 42

Off Normal Procedure OP-901-112 Charging or Letdown Malfunction Revision 006 E3 GAS BOUND CHARGING PUMPS PLACEKEEPER START DONE N/A NOTE Multiple indications and SM/CRS discretion should be applied to diagnosing Charging Pump gas intrusion. Indications for gas intrusion are the following:

Charging flow OR discharge pressure fluctuating OR lower than normal for the number of operating charging pumps.

AND EITHER of the Following:

Charging Suction Level Lost (a VCT, BAMT, or RWSP instrument malfunction may mask an actual low level condition).

Gas intrusion into the Charging Pumps caused by pulsation dampener failure, maintenance activities, OR a crack in the suction path of the Charging Pumps.

Similar symptoms on multiple charging pumps is also a good indication of a common affect like suction line gas binding.

CAUTION (1) GAS BINDING OF ONE CHARGING PUMP MAY INDICATE THAT THE COMMON SUCTION HEADER FOR ALL CHARGING PUMPS IS VOIDED. STANDBY CHARGING PUMPS SHOULD NOT BE STARTED UNTIL THE SUCTION SOURCE HAS BEEN RECOVERED.

(2) IN MODES 1 - 4, ENTRY INTO TECH SPEC 3.0.3 IS REQUIRED WITH ALL THREE CHARGING PUMP CONTROL SWITCHES IN OFF.

1. IF Charging parameters indicate gas intrusion into the Charging Pumps, THEN perform the following:

1.1 Simultaneously perform the following:

Close CVC-101, Letdown Stop Valve AND Stop the operating Charging Pump(s) 1.2 Place the standby Charging Pump control switches to OFF.

20

Off Normal Procedure OP-901-103 Emergency Boration Revision 003 E0 GENERAL (CONTD)

PLACEKEEPER START DONE N/A

6. Check the following Emergency Boration termination criteria:

6.1 If Emergency Boration was initiated due to inadequate Shutdown Margin, then continue to Emergency Borate until Shutdown Margin satisfies the following:

5.15% k/k with Tavg >200 F and any full length CEA is fully or partially withdrawn.

2.0% k/k with Tavg 200 F and any full length CEA is fully or partially withdrawn.

Greater than or equal to that required by Technical Specification 3.1.1.2 (applicable to Modes 3, 4, and 5, with all full length CEAs fully inserted).

6.2 If Emergency Boration was initiated due to uncontrolled cooldown, then continue to Emergency Borate until cooldown is stopped or Cold Shutdown boron concentration is reached.

6.3. If Emergency Boration was initiated due to uncontrolled positive reactivity addition, then continue Emergency Boration until Reactor is in a safe condition.

6.4 If Emergency Boration was initiated during Mode 6, then continue to Emergency Borate until Keff 0.95 or RCS boron concentration 2050 ppm, whichever is more restrictive.

8

WATERFORD 3 SES OP-902-004 Revision 016 Page 1 of 53 EXCESS STEAM DEMAND RECOVERY Emergency Operating Procedure OP-902-004 Excess Steam Demand Recovery OSRC Meeting No.:

Reviewed by:

OSRC Chairman: Print/Sign Approved by:

General Manager - Plant Operations: Print/Sign Approval Date Effective Date CONTINUOUS USE

WATERFORD 3 SES OP-902-004 Revision 016 Page 2 of 53 EXCESS STEAM DEMAND RECOVERY 1.0 PURPOSE This procedure provides operator actions which must be accomplished in the event of an Excess Steam Demand (ESD) event. The actions in this procedure are necessary to ensure the plant is placed in a safe, stable condition.

The goal of the procedure is to safely establish the plant in a condition which will allow the implementation of an appropriate existing procedure if the break has been isolated; or a procedure provided by the Technical Support Center. Radiological releases to the environment will be minimized and adequate core cooling will be maintained by following this procedure.

End of Section 1.0

WATERFORD 3 SES OP-902-004 Revision 016 Page 3 of 53 EXCESS STEAM DEMAND RECOVERY 2.0 SYMPTOMS AND ENTRY CONDITIONS This procedure is entered when an ESD event is evident, as indicated by ANY of the following conditions being met:

Loud noise indicative of a high energy steam line break or stuck open MSSV Lowering RCS temperature caused by the rise in RCS heat removal Rise in feedwater flow until main feedwater isolation valves are closed on Main Steam Isolation Signal (MSIS)

Possible rise in containment temperature, pressure, humidity, and sump level AND ONE of the following conditions being met:

The Standard Post Trip Actions have been performed.

Event initiated from Mode 3 or Mode 4.

End of Section 2.0

WATERFORD 3 SES OP-902-004 Revision 016 Page 4 of 53 EXCESS STEAM DEMAND RECOVERY 3.0 EXIT CONDITIONS This procedure may be exited when ANY of the following conditions are met:

1. Excess Steam Demand event has been misdiagnosed.

2. ANY of the ESD SFSC is NOT met.

3. The procedure has accomplished its purpose by satisfying ALL of the following:

ALL ESD SFSCs are satisfied Shutdown Cooling entry conditions are met.

OR Plant cooldown is NOT required and maintaining Mode 3 or 4 conditions is desired An appropriate procedure to implement has been provided and administratively approved.

End of Section 3.0

WATERFORD 3 SES OP-902-004 Revision 016 Page 5 of 53 EXCESS STEAM DEMAND RECOVERY 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

---

NOTE -----------------------------------------------------------

The Shift Chemist should be notified if a SIAS or CIAS has occurred. The secondary sampling containment isolation valves should not be opened following a SIAS or CIAS until directed by the Shift Chemist.

Confirm Diagnosis

• 1. Confirm diagnosis of an ESD:

a. Monitor the SFSCs and check a.1 GO TO ONE of the following:

Safety Function Status Check Acceptance criteria are satisfied. Appendix 1, Diagnostic Flowchart OP-902-008, "Functional Recovery Procedure"

b. IF SG sample path is available, b.1 IF SG can NOT be sampled, THEN direct Chemistry to sample THEN monitor for:

BOTH Steam Generators for activity. Unexplained rise in Secondary activity Unexplained rise in SG levels Announce the Event

• 2. Announce an Excess Steam Demand event is in progress using the plant page.

WATERFORD 3 SES OP-902-004 Revision 016 Page 6 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Classify the Event

• 3. Advise the Shift Manager to implement the Emergency Plan using EP-001-001, "Recognition & Classification of Emergency Condition."

Implement Placekeeping

• 4. REFER TO Section 6.0, "Placekeeper" and record the time of the reactor trip.

Restore Operation of DCT Sump Pumps

• 5. IF power has been interrupted to either 3A or 3B Safety bus, THEN perform Appendix 20, "Operation of DCT Sump Pumps."

Verify SIAS Actuation

• 6. Check SIAS has actuated. 6.1 IF PZR pressure is less than 1684 psia, OR CNTMT pressure greater than or equal to 17.1 psia, THEN verify SIAS is actuated.

WATERFORD 3 SES OP-902-004 Revision 016 Page 7 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Optimize Safety Injection

• 7. IF SIAS has actuated, THEN perform the following:

a. Verify Safety Injection pumps have started.

b. Check Safety Injection flow is b.1 Perform ANY of the following to restore within BOTH of the following: Safety Injection flow:

Attachment 2-E, "HPSI 1) Verify electrical power to Safety Flow Curve" Injection pumps and valves.

Attachment 2-F, "LPSI 2) Verify the following valves are Flow Curve" open:

HPSI Cold Leg injection valves LPSI Flow Control valves

3) Start additional Safety Injection pumps as needed until flow is within the following:

Attachment 2-E, "HPSI Flow Curve" Attachment 2-F, "LPSI Flow Curve"

4) IF HPSI flow is NOT within Attachment 2-E, THEN start the Standby HPSI pump using Appendix 26, "Aligning the Standby HPSI Pump."

(continue)

WATERFORD 3 SES OP-902-004 Revision 016 Page 8 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

• 7. (continued)

c. Verify ALL available Charging pumps are operating.

d. IF RWSP on Purification, THEN isolate RWSP using Appendix 40 Isolate RWSP from Purification.

Verify MSIS Actuation

8. Verify MSIS Actuation.

RCP Trip Strategy

• 9. IF PZR pressure is less than 1621 psia, AND SIAS is actuated, THEN perform the following:

a. Verify ONE RCP in each loop is stopped.

b. Check Pressurizer pressure is b.1 Verify ALL RCPs stopped.

greater than the minimum RCP NPSH of Attachment 2A-D, "RCS Pressure and Temperature Limits."

WATERFORD 3 SES OP-902-004 Revision 016 Page 9 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Verify RCP Operating Limits

• 10. IF RCPs are operating, THEN perform the following:

a. Verify CCW available to RCPs. a.1 IF CCW is lost to RCPs, AND is NOT restored within 3 minutes, THEN stop the affected pumps.

a.2 IF CCW Flow Low to RCPs, AND Instrument Air is available, THEN place IA 909 to close and THEN open.

b. IF a CSAS is initiated, THEN stop ALL RCPs.

c. IF RCS TC is less than 380ºF [384ºF],

THEN verify ONE RCP in each loop is stopped.

d. Check RCP operating parameters: d.1 IF ANY RCP operating limit exceeded, NPSH, REFER TO Attachment THEN stop affected RCP(s).

2A-D, RCS P-T Limits Bearing temperatures less than or equal to 225°F Bleed Off temperature less than 200°F Cooling Coils Return CCW temp less than 155°F At Least Two Seals per RCP operable

WATERFORD 3 SES OP-902-004 Revision 016 Page 10 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Protect Main Condenser

• 11. Perform BOTH of the following to protect 11.1 Perform the following:

the Main condenser:

a. Verify BOTH MSIVs are closed:

Verify CW System in operation.

REFER TO OP-003-006, Circulating MS 124A, MSIV 1 Water.

MS 124B, MSIV 2 Check Condenser vacuum greater than 14 inches Hg. b. Verify ALL Steam Generator Blowdown Isolation valves are closed:

BD 102A, STM GEN 1 (IN)

BD 102B, STM GEN 2 (IN)

BD 103A, STM GEN 1 (OUT)

BD 103B, STM GEN 2 (OUT)

WATERFORD 3 SES OP-902-004 Revision 016 Page 11 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Verify Proper CCW Operation

• 12. Check CCW operation:

a. Check a CCW pump is operating for a.1 IF the AB electrical bus is aligned to the EACH energized 4.16 KV Safety side with the faulted CCW pump, bus: THEN start CCW Pump AB as follows:

3A Safety bus 1) Place the CCW ASSIGNMENT switch to the desired position to 3B Safety bus replace the faulted pump.

2) Verify open the CCW SUCT &

DISCH HEADER TIE VALVES for the faulted CCW pump:

Train A CC 126A/114A CC 127A/115A Train B CC 126B/114B CC 127B/115B

3) Start CCW Pump AB.

4) IF CCW flow is NOT restored, THEN pull the affected EDG overspeed trip.

(continue) (continue)

WATERFORD 3 SES OP-902-004 Revision 016 Page 12 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

• 12. (continued) (continued) a.2 IF the AB electrical bus is NOT aligned to the side with the faulted CCW pump, AND the Sequencer has timed out, THEN start CCW Pump AB as follows:

1) Place the CCW ASSIGNMENT switch to the desired position to replace the faulted pump.

2) Verify open the CCW SUCT &

DISCH HEADER TIE VALVES for the faulted CCW pump:

Train A CC 126A/114A CC 127A/115A Train B CC 126B/114B CC 127B/115B

3) Start CCW Pump AB.

4) IF CCW flow is NOT restored, THEN pull the affected EDG overspeed trip.

(continue) (continue)

WATERFORD 3 SES OP-902-004 Revision 016 Page 13 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

• 12. (continued) (continued) a.3 IF CCW Pump AB is the faulted pump, THEN perform the following:

1) Place the CCW ASSIGNMENT switch to the neutral position.

2) Start CCW Pump A(B).

3) IF CCW flow is NOT restored, THEN pull the affected EDG overspeed trip.

b. IF only ONE CCW pump operating, THEN split CCW headers using Appendix 35, Single CCW Pump Operation.

c. Check an Essential chiller is c.1 Verify at least ONE 4KV Safety bus has operating for EACH energized an associated CCW pump and Essential 4.16 KV Safety bus: chiller operating as required.

3A Safety bus 3B Safety bus

WATERFORD 3 SES OP-902-004 Revision 016 Page 14 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Has MSIS Isolated the ESD

• 13. Check ESD break flow is still in 13.1 IF MSIS has isolated the ESD, progress. THEN GO TO Step 19.

Determine Most Affected SG with ESD

• 14. Determine the MOST AFFECTED SG by considering ALL of the following:

High steam flow from SG Dropping SG pressure Dropping SG level Dropping RCS Cold Leg temperature

WATERFORD 3 SES OP-902-004 Revision 016 Page 15 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS EFW Pump AB Steam Line Break

• 15. IF BOTH SGs are equally affected, AND suspect that the break is in the EFW Pump AB Steam Supply line, THEN perform the following:

a. Close BOTH Steam Supply to EFW Pump AB Isolation valves:

MS 401A, PUMP AB TURB STM SUPPLY SG 1 MS 401B, PUMP AB TURB STM SUPPLY SG 2

b. Monitor SG pressures to determine if break has been isolated.

c. IF closing Steam Supply to EFW Pump AB Isolation valves isolates the break, THEN GO TO Step 19.

d. IF closing Steam Supply to EFW Pump AB Isolation valves has had no effect on SG pressures, THEN open BOTH Steam Supply to EFW Pump AB Isolation valves:

MS 401A, PUMP AB TURB STM SUPPLY SG 1 MS 401B, PUMP AB TURB STM SUPPLY SG 2

WATERFORD 3 SES OP-902-004 Revision 016 Page 16 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Isolate Most Affected SG with ESD

• 16. Isolate the MOST AFFECTED SG as follows:

Steam Generator 1

a. Verify MS 124A, MSIV 1 is closed.

b. Verify FW 184A, MFIV 1 is closed.

c. Verify MS 116A, ADV 1 is closed and the controller in manual.

d. Verify EFW Isolation valves are closed:

EFW 228A, SG 1 PRIMARY EFW 229A, SG 1 BACKUP

e. Place EFW FLOW CONTROL valves in MAN and THEN close:

EFW 224A, SG 1 PRIMARY EFW 223A, SG 1 BACKUP

f. Close MS 401A, PUMP AB TURB STM SUPPLY SG 1.

g. Close Main Steam Line Drains:

MS 120A, NORMAL MS 119A, BYPASS (continue)

WATERFORD 3 SES OP-902-004 Revision 016 Page 17 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

• 16. (continued)

h. Verify SG Blowdown Isolation valves are closed:

BD 103A, STM GEN 1 (OUT)

BD 102A, STM GEN 1 (IN)

i. Check SG1 West Side Main Steam Safety valves are closed.

Steam Generator 2

a. Verify MS 124B, MSIV 2 is closed.

b. Verify FW 184B, MFIV 2 is closed.

c. Verify MS 116B, ADV 2 is closed and the controller in manual.

d. Verify EFW Isolation valves are closed:

EFW 228B, SG 2 PRIMARY EFW 229B, SG 2 BACKUP

e. Place EFW FLOW CONTROL valves in MAN and THEN close:

EFW 224B, SG 2 PRIMARY EFW 223B, SG 2 BACKUP

f. Close MS 401B, PUMP AB TURB STM SUPPLY SG 2.

(continue)

WATERFORD 3 SES OP-902-004 Revision 016 Page 18 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

• 16. (continued)

g. Close Main Steam Line Drains:

MS 120B, NORMAL MS 119B, BYPASS

h. Verify SG Blowdown Isolation valves are closed:

BD 103B, STM GEN 2 (OUT)

BD 102B, STM GEN 2 (IN)

i. Check SG2 East Side Main Steam Safety valves are closed.

WATERFORD 3 SES OP-902-004 Revision 016 Page 19 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Verify Correct SG Isolated

• 17. Verify the most affected SG is isolated 17.1 IF the wrong SG was isolated, by evaluating ALL of the following: THEN restore feeding and steaming capability to the isolated SG.

SG pressures 17.2 WHEN RCS Heat Removal has been SG levels re-established on the least affected SG, THEN REFER TO step 16 and isolate RCS Cold Leg temperatures the most affected SG.

WATERFORD 3 SES OP-902-004 Revision 016 Page 20 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

---

NOTE ----------------------------------------------------------

Actions to stabilize RCS temperature following an excess steam demand event should be initiated when BOTH of the following parameters are met:

CET temperatures rise Pressurizer pressure rise Stabilize RCS Temperature with SG ESD

• 18. Stabilize RCS temperature within PT Curves by performing the following:

a. For the LEAST AFFECTED SG:

1) Place the ADV to manual and fully open ADV.

2) Manually initiate EFAS.

3) Place the EFW Flow Control valve to manual and commence feeding.

4) Perform ANY of the following as necessary to establish RCS pressure and temperature control:

Throttle associated SG ADV.

Adjust associated SG EFW flow.

b. IF RCS pressure is greater than or equal b.1 IF RCS pressure is less than to 1500 psia, 1500 psia, THEN stabilize RCS pressure at a value THEN stabilize RCS not to exceed 1600 psid between the pressure at greater than RCS and the lowest SG pressure. HPSI shutoff head (1500 -

1600 psia).

c. REFER TO Step 20, HPSI Throttle Criteria.

WATERFORD 3 SES OP-902-004 Revision 016 Page 21 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Stabilize RCS Temperature With ESD Isolated by MSIS

• 19. IF ESD isolated by MSIS, THEN stabilize RCS temperature within PT Curves by performing the following:

a. Operate ADVs for BOTH SGs as necessary.

b. Check EFAS actuated.

c. IF EFAS actuated, THEN verify EFW operating in automatic or manual to restore BOTH SG levels.

d. Stabilize RCS Pressure within PT curves using Main or Auxiliary Spray valves.

WATERFORD 3 SES OP-902-004 Revision 016 Page 22 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS HPSI Throttle Criteria

• 20. IF HPSI pumps are operating, AND ALL of the following conditions are satisfied:

RCS subcooling is greater than or equal to 28ºF PZR level is greater than 7% [23%]

and controlled Verify at least ONE SG is available for RCS heat removal and level is being maintained or restored to within 55 to 70% NR

[60 to 80% NR] using EFW in auto or manual.

RVLMS indicates level higher than Hot Leg by at least ONE of the following:

QSPDS REACTOR VESSEL LEVEL 5 NOT voided VESSEL LEVEL PLENUM greater than or equal to 80%

THEN perform ANY of the following:

Throttle HPSI flow.

Stop ONE HPSI pump at a time.

WATERFORD 3 SES OP-902-004 Revision 016 Page 23 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS HPSI Pump Restart Criteria

• 21. IF ANY of the HPSI throttle criteria can NOT be maintained, THEN perform the following:

a. Raise HPSI flow.

b. Start HPSI pumps as necessary.

LPSI Pump Stop Criteria

• 22. IF PZR pressure is greater than 200 psia, AND controlled, THEN perform the following:

a. Stop LPSI pumps.

b. Close the LPSI Flow Control valves.

LPSI Pump Restart Criteria

• 23. IF PZR pressure lowers to less than 200 psia, AND is NOT controlled, THEN perform the following:

a. Start LPSI pumps as necessary.

b. Open the LPSI Flow Control valves.

WATERFORD 3 SES OP-902-004 Revision 016 Page 24 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Maintain SG Level

• 24. Verify at least ONE SG is available AND ALL SGs used for steaming are being maintained or restored to within 55 to 70% NR [60 to 80% NR] using MFW or EFW in auto or manual.

WATERFORD 3 SES OP-902-004 Revision 016 Page 25 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Maintain RCS within RCS PT Limits

• 25. Maintain the RCS within the limits of Attachments 2A-D, "RCS Pressure and Temperature Limits" by performing ANY of the following:

a. Control Main or Auxiliary PZR spray.

b. Control PZR heaters. b.1 Restore PZR heaters using Appendix 25, "Restore Pressurizer Heater Control."

c. IF HPSI throttle criteria are met (Step 20),

THEN perform ANY of the following:

Control Charging and Letdown.

Throttle HPSI flow as necessary.

d. Control RCS cooldown rate within d.1 Lower the cooldown rate as allowed limits. necessary to restore within limits.

e. Check RCS pressure less than the upper e.1 Stop the cooldown.

limits of PT curves.

e.2 Lower RCS pressure using Main or Auxiliary spray.

e.3 IF HPSI throttle criteria met, THEN control charging, letdown, or HPSI flow as necessary.

e.4 IF necessary, THEN operate PZR vents using Appendix 38, Pressurizer Vent Operations.

e.5 WHEN RCS pressure within limits, THEN continue cooldown.

WATERFORD 3 SES OP-902-004 Revision 016 Page 26 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Verify Containment Isolation and Cooling

• 26. IF CNTMT pressure greater than 17.1 psia, THEN perform the following:

a. Check CIAS is initiated. a.1 Manually actuate CIAS.

a.2 Verify that an Isolation valve is closed for each containment penetration required to be closed.

b. Verify ALL available Containment b.1 IF ANY Containment Fan Cooler Fan Coolers operating in is NOT operating, emergency mode. THEN close the associated Containment Fan Cooler CCW Isolation valves using Attachment 21-B, "CFC CCW Override."

Verify Containment Spray Actuation

• 27. IF CNTMT pressure is greater than 17.7 psia, THEN perform the following:

a. Check CSAS is initiated. a.1 Manually actuate CSAS.

a.2 Verify that CSAS components actuated.

b. Verify ALL operating Containment b.1 IF ANY CS-125, Containment Spray pumps are delivering flow Spray Header Isolation is open, greater than 1750 gpm. AND the associated CS pump is NOT operating, THEN close the valve using Attachment 21-A, "CS-125 Override."

WATERFORD 3 SES OP-902-004 Revision 016 Page 27 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS CAUTION The Containment Vacuum Relief Valves may open if Containment pressure lowers to less than 15.0 [15.7] psia.

Containment Spray Termination

• 28. IF at least ONE CS pump is operating, AND BOTH of the following conditions are satisfied:

Containment pressure is less than 16.7 psia and stable or lowering Containment spray is not required for containment cooling THEN reset CSAS actuation using Attachment 5-E, "CSAS Reset Procedure."

WATERFORD 3 SES OP-902-004 Revision 016 Page 28 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Restore Instrument Air

• 29. Verify Instrument Air is available by performing ALL of the following:

a. Check BOTH of the following are a.1 Align Potable water to the operating: Instrument Air compressors using Appendix 18, "Aligning Potable TCW pump Water to Instrument Air Compressors."

CW pump

b. Check Instrument Air pressure is b.1 Dispatch an operator to start ALL greater than 95 psig. available Air compressors.

c. Check IA 909, CNTMT ISOLATION c.1 IF Instrument Air pressure is INSTRUMENT AIR valve is open. greater than 95 psig, THEN perform the following:

1) Open IA 909, CNTMT ISOLATION INSTRUMENT AIR valve.

2) IF Instrument Air pressure NOT restoring or maintaining greater than 95 psig, THEN close IA 909, CNTMT ISOLATION INSTRUMENT AIR valve.

WATERFORD 3 SES OP-902-004 Revision 016 Page 29 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Restore Letdown

• 30. IF Letdown is isolated, AND BOTH of the following conditions exist:

HPSI throttle criteria are met (Step 20)

Letdown is needed or desired THEN restore Letdown using Appendix 9, "Letdown Restoration."

Verify RCS is NOT Water Solid

• 31. Check the RCS is NOT in a water solid 31.1 IF the RCS is water solid, condition as indicated by the following: THEN maintain the RCS within Attachments 2A-D, "RCS Pressure and

a. RCS pressure responds slowly to Temperature Limits" by ANY of the RCS inventory and temperature following:

changes.

Control RCS temperature.

b. At least ONE of the following conditions met: IF HPSI throttle criteria are met (Step 20),

PZR level less than 100% THEN perform ANY of the following:

Reactor Vessel Head level Control Charging and Letdown.

less than 100%

Throttle HPSI flow as necessary.

WATERFORD 3 SES OP-902-004 Revision 016 Page 30 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Draw a Bubble in Pressurizer

• 32. IF it is desired to draw a bubble in the PZR, THEN establish a pressurizer bubble using Appendix 17, "Establish Pressurizer Bubble."

Maintain Pressurizer Level

• 33. IF HPSI throttle criteria are met 33.1 IF the normal Charging path is NOT (Step 20), available, THEN maintain PZR level 7% [23%] to AND the HPSI pumps are NOT 60% by performing ANY of the operating, following: THEN start Charging pumps using Appendix 30, "Charging to the RCS via Control Charging and Letdown. the HPSI Header."

Throttle HPSI flow as necessary.

Energize AB Safety Buses

• 34. Verify the AB Safety buses are energized. REFER TO ANY applicable attachment:

Attachment 12-E, "Energize AB Safety buses from the A Side" Attachment 12-F, "Energize AB Safety buses from the B Side"

WATERFORD 3 SES OP-902-004 Revision 016 Page 31 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Secure TGB Battery Loads

• 35. IF Off-Site Power has been lost, AND resources permit, THEN secure TGB Battery DC loads.

REFER TO ANY applicable attachment:

Attachment 33-B, Generator Auxiliary Operations - LOOP (Both Trains)

Attachment 33-C, Generator Auxiliary Operations - Loss of Single Train Off-Site Power Restore Electrical Power

• 36. IF Off-Site Power has been lost, AND resources permit, THEN restore power to plant loads.

REFER TO ALL applicable attachments:

Attachment 12-A, "Energize 1A(B) and 2A(B) from Off-Site Power" Attachment 12-B, "Energize 3A(B) from 2A(B)"

Attachment 12-C, Transfer 3A(B) from EDGA(B) to 2A(B)"

Attachment 12-D, "Lighting and In-Plant Load Restoration" Attachment 12-G, "Energize 312AB and 313AB Non-Safety Buses"

WATERFORD 3 SES OP-902-004 Revision 016 Page 32 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS RCP Restart Criteria

• 37. IF RCP restart is desirable, THEN perform the following:

a. Verify RCS pressure and temperature are within Attachments 2A-D, "RCS Pressure and Temperature Limits" for RCP operation.

b. Verify at least ONE SG is available for RCS heat removal and level is being maintained within 55 to 70% NR

[60 - 80% NR] using MFW or EFW in auto or manual.

c. Verify PZR level is within 45 to 60%.

d. Verify electrical power is available to the RCPs.

e. Check ANY loss of CCW duration was less than 10 minutes.

f. IF RCS TC less 200 F, THEN verify the secondary water temperature of each SG is within 100 F of the RCS TC temperatures.

RCP Start

• 38. IF RCP restart is desired, AND RCP restart criteria are satisfied, THEN start RCPs using Appendix 19, "RCP Restart."

WATERFORD 3 SES OP-902-004 Revision 016 Page 33 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Check Single Phase Natural Circulation

• 39. IF NO RCPs are operating, 39.1 Verify proper SG feeding and steaming THEN check natural circulation flow in at by verifying operating loop SG pressure least ONE loop by ALL of the following: is approximately at saturation pressure for the existing TC.

RCS subcooling greater than or equal to 28°F based on representative CET temperature Loop T less than 58°F Hot and Cold Leg temperature constant or lowering TH and representative CET temperature T is less than 10°F

WATERFORD 3 SES OP-902-004 Revision 016 Page 34 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Place Hydrogen Analyzers in Service

40. IF the ESD is inside containment, THEN place Hydrogen Analyzers in service by performing the following:

Train A

a. Place Train A H2 ANALYZER CNTMT ISOL VALVE keyswitch to "OPEN." (Key 216)

b. Place H2 ANALYZER A POWER to "ON."

c. Check H2 ANALYZER A PUMPS indicate on.

Train B

a. Place Train B H2 ANALYZER CNTMT ISOL VALVE keyswitch to "OPEN." (Key 217)

b. Place H2 ANALYZER B POWER to "ON."

c. Check H2 ANALYZER B PUMPS indicate on.

WATERFORD 3 SES OP-902-004 Revision 016 Page 35 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Evaluate CNTMT Hydrogen Level

• 41. Evaluate the need for Containment Hydrogen Purge as follows:

a. Request TSC/EOF to evaluate the need to perform Containment Hydrogen purge.

b. Perform Containment Hydrogen purge as directed by TSC/EOF using Appendix 16, Operate CARS.

Secure Operating EDGs

• 42. IF ANY EDG is operating unloaded, AND is NO longer needed, THEN secure the EDG.

Monitor Spent Fuel Pool

• 43. Monitor SFP temperature, level, and 43.1 For SFP alarm or abnormal conditions, radiation: REFER TO OP-901-513, SFP Malfunction.

a. IF SFP cooling NOT in service, THEN restore SFP cooling with TSC concurrence.

b. Maintain SFP inventory with TSC concurrence.

WATERFORD 3 SES OP-902-004 Revision 016 Page 36 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Perform Post Trip Plant Alignment Appendix

• 44. REFER TO Appendix 24, "Post Trip Plant Alignment" and perform the applicable steps.

WATERFORD 3 SES OP-902-004 Revision 016 Page 37 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

---

NOTE ---------------------------------------------------------

The DWST or CST is the preferred source of Condensate Makeup rather than the Wet Cooling Tower.

Loss of power to either MCC-214A or MCC-214B will result in loss of BOTH Hotwell Transfer pump and Condensate Transfer pump due to I/O Panel losing power.

Makeup to CSP

• 45. Monitor Condensate Storage Pool level:

a. Maintain CSP level greater than a.1 IF CSP level can NOT be 92% by locally throttling CMU-141, maintained greater than 92%,

Condensate Storage Pool LCV THEN request the TSC to evaluate Bypass valve. Condensate Makeup restoration.

b. IF CSP level is less than 25%,

THEN transfer EFW Pump suction to ONE operating train of ACCW using Appendix 10, "Transferring EFW Pump Suction."

Evaluate Need for Cooldown

• 46. Evaluate the need for a plant cooldown based on ALL of the following:

Plant status Auxiliary systems availability IF EFW pumps are supplying the SGs, THEN verify Condensate inventory is greater than minimum required for plant cooldown using Attachment 2-G, "Condensate Inventory Calculation."

WATERFORD 3 SES OP-902-004 Revision 016 Page 38 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Cooldown NOT Desired

47. IF a plant cooldown is NOT desired, THEN perform the following:

a. Verify Section 3.0 Exit Criteria are met

b. WHEN BOTH of the following conditions are met:

At least ONE RCP is operating MFW is available to the SGs, THEN perform the following:

1) Maintain plant conditions stable.

2) IF ANY ESFAS actuations have initiated, THEN reset ESFAS actuations using Appendix 5, "ESFAS Reset."

3) GO TO OP-010-005, Plant Shutdown."

Maintain Shutdown Margin During the Cooldown

• 48. Maintain Shutdown Margin during cooldown as follows:

a. Calculate Shutdown Margin Boron concentration for desired RCS temperature using OP-903-090, "Shutdown Margin."

b. Borate the RCS to maintain Shutdown Margin throughout the cooldown.

WATERFORD 3 SES OP-902-004 Revision 016 Page 39 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

---

NOTE ----------------------------------------------------------

The following RCS cooldown rates apply:

RCS 100°F/hr RCS on Natural Circulation with Asymmetric Steam Generator: RCS 50°F/h Pressurizer 200°F/hr Attachment 3-A, "Pressurizer/RCS Cooldown Log may be required during the cooldown and depressurization.

Perform Controlled Cooldown

49. Cooldown the RCS to less than 350°F 49.1 IF necessary, TH or CET temperature using the Intact THEN locally control intact SG ADV SG ADV. using Appendix 22 Local Operation of Atmospheric Dump Valves.

WATERFORD 3 SES OP-902-004 Revision 016 Page 40 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

---

NOTE ----------------------------------------------------------

The low steam generator pressure signal provides an input to the MSIS and the EFAS logic. Failure to reset the low steam generator pressure signal may prevent automatic feeding of the steam generators by the EFAS logic.

Reset MSIS Initiation Setpoints

• 50. IF MSIS (low SG pressure) has NOT actuated, AND a controlled cooldown and depressurization is in progress, THEN lower MSIS initiation setpoints as necessary.

Reset SIAS Initiation Setpoints

• 51. IF SIAS (low PZR pressure) has NOT actuated, AND a controlled cooldown and depressurization is in progress, THEN lower SIAS initiation setpoints as necessary.

Letdown Control Valves

• 52. WHEN RCS pressure less than 1200 psia, AND Letdown is in service, THEN verify BOTH Letdown Flow control and BOTH Back pressure Control valves in service using Appendix 9, Letdown Restoration.

WATERFORD 3 SES OP-902-004 Revision 016 Page 41 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Void Elimination

• 53. IF the RCS fails to depressurize, AND voiding is suspected, THEN eliminate voids using Appendix 11, "Void Elimination."

SIT Isolation and Venting

• 54. WHEN PZR pressure less than 1000 psia, AND controlled, THEN isolate the SITs using Appendix 14, "SIT Isolation and Venting."

Verify ESFAS Actuations

• 55. IF additional personnel available, THEN verify ESFAS actuations using Appendix 4, "ESFAS Auto Actions."

Place LTOP In Service

• 56. WHEN RCS TC is less than 350°F, AND RCS pressure is less than 392 psia [358 psia],

THEN place LTOPs in service using OP-009-005, "Shutdown Cooling System."

WATERFORD 3 SES OP-902-004 Revision 016 Page 42 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Reset Safety Systems

• 57. IF any ESFAS actuations have initiated, AND are NO longer needed, THEN reset ESFAS actuations using Appendix 5, "ESFAS Reset."

SDC Entry Conditions

• 58. IF ALL of the following SDC entry conditions are established:

PZR level is greater than 33%

RCS subcooling is greater than or equal to 28°F PZR pressure is less than 392 psia

[358 psia]

RCS TH is less than 350°F THEN place SDC in service using OP-009-005, "Shutdown Cooling System."

Exit ESD

59. WHEN Section 3.0 Exit Conditions are satisfied, THEN GO TO OP-010-005, Plant Shutdown" or other approved procedure.

End of Section 4.0

WATERFORD 3 SES OP-902-004 Revision 016 Page 43 of 53 EXCESS STEAM DEMAND RECOVERY 5.0 SAFETY FUNCTION STATUS CHECK SAFETY FUNCTION:

1. Reactivity Control PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. Reactor Power Dropping

b. SUR Negative

c. ONE of the following:

CEAs < 2 NOT fully inserted Emergency boration 40 gpm Shutdown margin Verified Condition 2

a. Reactor Power < 10-4% stable or dropping

b. ONE of the following:

CEAs < 2 NOT fully inserted Emergency boration 40 gpm Shutdown margin Verified

WATERFORD 3 SES OP-902-004 Revision 016 Page 44 of 53 EXCESS STEAM DEMAND RECOVERY SAFETY FUNCTION:

2. Maintenance of Vital Auxiliaries (AC and DC Electrical Power)

PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. Bus A3, A-DC and BOTH associated vital AC Instrument Channels Energized Condition 2

a. Bus B3, B-DC and BOTH associated vital AC Instrument Channels Energized

WATERFORD 3 SES OP-902-004 Revision 016 Page 45 of 53 EXCESS STEAM DEMAND RECOVERY SAFETY FUNCTION:

3. RCS Inventory Control PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. Pressurizer level > 7% [23%]

b. RCS subcooling 28 F

c. ONE of the following:

RVLMS LEVEL PLENUM 80%

QSPDS Vessel Level Level 5 NOT voided Condition 2

a. HPSI flow Attachment 2-E, HPSI Flow Curve

b. LPSI flow Attachment 2-F, LPSI Flow Curve

c. RVLMS LEVEL PLENUM 20%

WATERFORD 3 SES OP-902-004 Revision 016 Page 46 of 53 EXCESS STEAM DEMAND RECOVERY SAFETY FUNCTION:

4. RCS Pressure Control PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. Pressurizer pressure Attachments 2A-D, RCS Pressure and Temperature Limits Condition 2

a. HPSI flow Attachment 2-E, HPSI Flow Curve

b. LPSI flow Attachment 2-F, LPSI Flow Curve SAFETY FUNCTION:

5. Core Heat Removal PARAMETER CRITERIA CRITERIA SATISFIED

a. RCS TH and Representative CET temperature < 650 F

b. RCS subcooling 28 F

WATERFORD 3 SES OP-902-004 Revision 016 Page 47 of 53 EXCESS STEAM DEMAND RECOVERY SAFETY FUNCTION:

6. RCS Heat Removal PARAMETER CRITERIA CRITERIA SATISFIED

---

NOTE ----------------------------------------------------------

The automatic system response of the EFW system is acceptable to satisfy the RCS Heat Removal requirements. Manual feeding is not required until Steam Generator level is less than 55% WR.

Condition 1

a. Steam Generator level Level is being maintained or restored 55 - 70% NR

[60 - 80% NR] in at least ONE SG using MFW or with EFW in MANUAL

b. RCS TC Stable or lowering Condition 2

a. Steam Generator level Level is being maintained or restored 55 - 70% NR

[60 - 80% NR] in at least ONE SG using EFW in AUTOMATIC

b. RCS TC Stable or lowering

WATERFORD 3 SES OP-902-004 Revision 016 Page 48 of 53 EXCESS STEAM DEMAND RECOVERY SAFETY FUNCTION:

7. Containment Isolation PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. Containment pressure < 17.1 psia

b. Containment Area No alarms or Radiation monitors unexplained rise

c. Steam Plant Activity No alarms or monitors unexplained rise Condition 2

a. CIAS Initiated

b. Containment Area No alarms or Radiation monitors unexplained rise

c. Steam Plant Activity No alarms or monitors unexplained rise

WATERFORD 3 SES OP-902-004 Revision 016 Page 49 of 53 EXCESS STEAM DEMAND RECOVERY SAFETY FUNCTION:

---

NOTE ----------------------------------------------------------

Hydrogen concentration acceptance criteria may be omitted until a hydrogen analyzer is in service.

8. Containment Temperature and Pressure Control PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. Containment temperature < 220 F

b. Containment pressure < 17.7 psia

c. Hydrogen < 0.6%

Condition 2

a. Containment Fan 1 operating in Coolers emergency mode

b. Containment Spray 1 operating with pump flow 1750 gpm

c. Containment pressure < 50 psia

d. Hydrogen < 3.4%

Condition 3

a. Containment Spray 2 operating with pumps flow 1750 gpm

b. Containment pressure < 50 psia

c. Hydrogen < 3.4%

End of Section 5.0

WATERFORD 3 SES OP-902-004 Revision 016 Page 50 of 53 EXCESS STEAM DEMAND RECOVERY 6.0 PLACEKEEPER Time of Reactor Trip _________

STEP INSTRUCTIONS START DONE

• 5. Restore Operation of DCT Sump Pumps .......................................... Cont.

• 6. Check SIAS Actuation ....................................................................... Cont.

• 7. Optimize Safety Injection ................................................................... Cont.

8. Verify MSIS Actuation........................................................................

• 9. RCP Trip Strategy .............................................................................

• 10. Verify RCP Operating Limits.............................................................. Cont.

• 11. Protect Main Condenser ....................................................................

• 12. Verify Proper CCW Operation ...........................................................

• 13. Has MSIS Isolated the ESD ..............................................................

• 14. Determine Most Affected SG with ESD .............................................

• 15. EFW Pump AB Steam Line Break .....................................................

• 16. Isolate Most Affected SG with ESD ...................................................

• 17. Verify Correct SG Isolated ................................................................

• 18. Stabilize RCS Temperature with SG ESD .........................................

• 19. Stabilize RCS Temperature with ESD Isolated by MSIS ...................

• 20. HPSI Throttle Criteria ........................................................................

WATERFORD 3 SES OP-902-004 Revision 016 Page 51 of 53 EXCESS STEAM DEMAND RECOVERY STEP INSTRUCTIONS START DONE

• 21. HPSI Pump Restart Criteria...............................................................

• 22. LPSI Pump Stop Criteria ................................................................... Cont.

• 23. LPSI Pump Restart Criteria ............................................................... Cont.

• 24. Maintain SG Level ............................................................................. Cont.

• 25. Maintain RCS within RCS PT Limits .................................................. Cont.

• 26. Verify Containment Isolation and Cooling..........................................

• 27. Verify Containment Spray Actuation..................................................

• 28. Containment Spray Termination ........................................................

• 29. Restore Instrument Air.......................................................................

• 30. Restore Letdown ...............................................................................

• 31. Verify RCS is NOT Water Solid .........................................................

• 32. Draw a Bubble in Pressurizer ............................................................

• 33. Maintain Pressurizer Level ................................................................ Cont.

• 34. Energize AB Safety Buses ................................................................

• 35. Secure TGB Battery Loads................................................................

• 36. Restore Electrical Power ...................................................................

• 37. RCP Restart Criteria..........................................................................

• 38. RCP Start ..........................................................................................

• 39. Check Single Phase Natural Circulation ............................................ Cont.

WATERFORD 3 SES OP-902-004 Revision 016 Page 52 of 53 EXCESS STEAM DEMAND RECOVERY STEP INSTRUCTIONS START DONE

40. Place Hydrogen Analyzers in Service................................................

41. Evaluate CNTMT Hydrogen Level .....................................................

• 42. Secure Operating EDGs....................................................................

• 43. Monitor Spent Fuel Pool .................................................................... Cont.

• 44. Perform Post Trip Plant Alignment Appendix ....................................

• 45. Makeup to CSP .................................................................................

• 46. Evaluate Need for Cooldown .............................................................

47. Cooldown NOT Desired.....................................................................

• 48. Maintain Shutdown Margin During the Cooldown.............................. Cont.

49. Perform Controlled Cooldown ...........................................................

• 50. Reset MSIS Initiation Setpoints .........................................................

• 51. Reset SIAS Initiation Setpoints..........................................................

• 52. Letdown Flow and Back Pressure Control Valves.............................

• 53. Void Elimination.................................................................................

• 54. SIT Isolation and Venting ..................................................................

• 55. Verify ESFAS Actuations ...................................................................

• 56. Place LTOP In Service ......................................................................

• 57. Reset Safety Systems .......................................................................

WATERFORD 3 SES OP-902-004 Revision 016 Page 53 of 53 EXCESS STEAM DEMAND RECOVERY STEP INSTRUCTIONS START DONE

• 58. SDC Entry Conditions........................................................................

59. Exit ESD ............................................................................................

End of Section 6.0

[LAST PAGE]

Administrative Procedure OP-100-017 Emergency Operating Procedure Implementation Guide Revision 004 5.16 FUNCTIONAL RECOVERY PROCEDURE 5.16.1 The Standard Post Trip Actions (SPTAs) are performed prior to entry into the FRP for an event initiated from Mode 1 or Mode 2. The FRP may be entered directly after completion of the SPTAs if a diagnosis of a single event is not possible. The FRP might also be entered from an ORP if an ORP had been initially selected but was subsequently found to be inadequate. If the Safety Function Acceptance Criteria are not satisfied at any time, then the CRS should evaluate the need to implement the FRP.

5.16.2 The FRP may be entered directly from a Mode 3 or Mode 4 event if the Entry Conditions are met.

5.16.3 The CRS reviews the status of all safety functions using the Resource Assessment Trees and the Safety Function Tracking Sheet. The CRS should obtain input and concurrence from the Balance of Plant Operator and the At The Controls Operator while determining which success path to implement. The CRS will implement success paths for all safety functions based on equipment availability and current plant conditions. The STA should independently perform a check of each Safety Function using the Safety Function Status Checklist to determine the appropriate success path to enter. The STA should concur with the CRS to validate the CRS assessment of the appropriate success path to enter.

5.16.4 More than one safety function may be pursued concurrently if plant conditions warrant.

5.16.5 If a significant change in plant status has occurred as determined by the CRS, then the CRS should reevaluate the success paths in use using the Resource Assessment Trees and the Safety Function Tracking Sheet. The STA should independently validate the CRS assessment of the success paths.

22

Administrative Procedure OP-100-017 Emergency Operating Procedure Implementation Guide Revision 004 5.20 MONITORING SAFETY FUNCTIONS 5.20.1 The STA performs the initial assessment of the Safety Function upon entering the ORP and when directed to in FRP.

5.20.2 The STA is required to perform a check of the SFSC every 15 minutes and notify the CRS of completion of the check.

5.20.3 The Safety Functions shall be continuously monitored by the operating crew using the Safety Function Status Checklist.

26

9.23 LOSS OF REFUEL CAVITY W ATER LEVEL GUIDELINES NOTE During the course of a Refueling Outage the potential exists for an inadvertent loss of refueling pool level due to various leakage paths.

CAUTION R THE FOLLOWING SECTION HAS THE POTENTIAL TO AFFECT CORE REACTIVITY.

[INPO 06-006]

9.23.1. Potential Leakage Paths:

9.23.1.1 Loss of Nozzle Dam: This condition, if uncorrected will result in a final water level of 13.5 ft MSL in the core and a 20 ft. MSL water level in Spent Fuel Pool and north end of the Refueling Cavity.

9.23.1.2 Loss of Pool Seal: This condition will result in a rapid loss of level with a final water level of 20 ft MSL at all locations.

9.23.1.3 Leak in the Shutdown Cooling System: This condition will result in a similar situation to step 9.23.1.1 above.

NOTE Cask Pit floor drain has been welded over and is no longer a potential leak path.

9.23.1.4 Leakage Through Floor Drains or Cavity Drain/Cooling System: This condition will result in a loss of complete level in the area of the leak (i.e.,

spent fuel canal, cask pit, refueling canal deep end). Reactor Vessel will drain to 20 ft. MSL, Spent Fuel Pool will drain to 20 ft. MSL and all other areas will drain below an acceptable level for safe fuel storage.

OP-010-006 Revision 329 Attachment 9.23 (1 of 2) 104

Off Normal Procedure OP-901-524 Fire in Areas Affecting Safe Shutdown Revision 015 E SUBSEQUENT OPERATOR ACTIONS E0 GENERAL PLACEKEEPER START DONE N/A NOTE Actions in this procedure should be taken based on SM/CRS judgment of the impact the fire is having on the plant. The SM/CRS may use that judgment to deem steps Not Applicable.

However, tripping the RCPs (step E0.4.2) and performance of Attachment 2, Electrical Lockout of Reactor Coolant Pumps, is required to be done within 20 minutes of a fire which can potentially affect components which provide cooling to the RCP seals in Fire Zones RAB 7A, 7B, 7C, 7D, 8A, 8B, 8C and TGB outside of the TGB Switchgear.

Continuous

1. Perform this procedure concurrently with FP-001-020, Fire Emergency / Fire Report.

Continuous

2. If the fire is in the Control Room (Fire Zone RAB-1A) or Cable Spreading Room (Fire Zone RAB-1E),

and the SM/CRS determines that the Control Room must be abandoned then go to OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

Continuous

3. Monitor all safety-related instrumentation and controls for abnormal indications, spurious actuations, or loss of control of equipment.

7

Off Normal Procedure OP-901-524 Fire in Areas Affecting Safe Shutdown Revision 015 E0 GENERAL (CONTD)

PLACEKEEPER START DONE N/A NOTE A review of the performance of Standard Post Trip Actions shall not prevent completion of the time critical action to trip and electrically lock out the Reactor Coolant Pumps.

4. If the SM/CRS determines the fire is of sufficient size to affect components which provide cooling to the RCP seals in Fire Zones RAB 7A, 7B, 7C, 7D, 8A, 8B, 8C and TGB outside of the TGB Switchgear, then perform the following:

4.1 Trip the Reactor and perform OP-902-000, Standard Post Trip Actions, concurrently with this procedure.

4.2 Trip and electrically lock out the Reactor Coolant Pumps as follows:

[20 minutes]

4.2.1 Trip all Reactor Coolant Pumps from CP-2 4.2.2 Direct the Switchgear Operator to perform Attachment 2, Electrical Lockout of Reactor Coolant Pumps.

8

Off Normal Procedure OP-901-524 Fire in Areas Affecting Safe Shutdown Revision 015 Page 2 of 8 ATTACHMENT 1: FIRE AREAS (CONTD)

FIRE ELEV. ROOM NAME SECTION DOOR AREA NUMBER NUMBER RAB-2 +46 Hatch Area E7 D86

+46 H&V Room E7 D86

+46 Passage E7 D98

+46 Airlock E7 D261, D262

+46 Intake Plenum E7 D263

+69 Air Plenum E7 D272

+91 H&C CHW Expansion Tank Room E7 D277

+46 Intake Plenum E7 D263

+69 Air Exchange Structure E7 D274

+69 H&V Fan Room E7 D271

+69 Air Filters Access Room E7 D271

+69 Air Exchange Structure E7 D275

+91 CCW Surge Tank Room E7 D277

+91 Fresh Air Intake E7 D277 RAB-3 +35 Vestibule E8 D176

+35 Corridor, North - South E8 D228 to D176

+46 H&V Equipment Room E8 D81

+46 H&V Room E8 D72, D75 RAB-3A +7 I&C Room E9 D121

+7 Toilet E9 D122

+7 H&V Duct Space E9 D126

+21 Vestibule E9 D49

+69 H&V Room E9 D184

+69 Elevator Machine Room E9 D186 RAB-5 +35 Electrical Penetration Area "B" E10 D174 RAB-6 +35 Electrical Penetration Area "A" E11 D179 RAB-7A +35 Relay Room A E12 D219, D209B RAB-7B +35 Relay Room B E13 D218, D209A RAB-7C +35 Relay Room Isolation Panel E14 D208B RAB-7D +35 Relay Room AB E15 D209A, D209B RAB-8A +21 Switchgear Room "A" E16 D7

+21 Computer Battery Room E16 D45

+21 Computer Equipment Area E16 D9, D51, D53A RAB-8B +21 H&V Fan Room E17 D50

+21 Switchgear Room "B" E17 D9

+21 CEA MG Set Room E17 D16

+21 Electrical Penetration Area E17 D9

+21 3A32 Switchgear E17 D53A RAB-8C +21 Switchgear Room "AB" E18 D9, D51 RAB-9 +21 Auxiliary Control Panel E19 D44 106

Off Normal Procedure OP-901-110 Pressurizer Level Control Malfunction Revision 009 C AUTOMATIC ACTIONS

1. If Pressurizer level rises above program level, then the following should occur:

1.1 Letdown flow rises to a maximum of 126 GPM.

1.2 If level rises 4.0 above program level, then the following should occur.

Backup Charging Pumps receive backup stop signal If Pressurizer pressure 2275 PSIA, then Backup Heaters energize

2. If Pressurizer level drops below program level, then the following should occur:

2.1 Letdown flow drops to a minimum of 28 GPM.

2.2 If level drops 2.5 below program level, then first backup Charging Pump starts.

2.3 If level drops 3.9 below program level, then second backup Charging Pump starts.

2.4 If level drops 6.0 below program level, then both backup Charging Pumps receive backup start signal.

2.5 If level drops to 28 , then all Pressurizer Heaters de-energize.

END 4

Off Normal Procedure OP-901-110 Pressurizer Level Control Malfunction Revision 009 E SUBSEQUENT OPERATOR ACTIONS:

E0 GENERAL

1. Stop Turbine load changes.

2. If malfunction is due to failure of Letdown Flow Control valve, then go to OP-901-112, CHARGING/LETDOWN MALFUNCTION.

3. If malfunction is due to failure of Pressurizer Level Control Channel (incorrect readings on either RC-ILI-0110X or RC-ILI-0110Y), then go to Subsection E1, Pressurizer Level Control Channel Malfunction.

4. If malfunction is due to failure of Pressurizer Level Setpoint (RC-ILIC-0110), then go to Subsection E2, Pressurizer Level Setpoint Malfunction.

5. If malfunction is due to failure of Pressurizer Level Controller (RC-ILIC-0110), then go to Subsection E3, Pressurizer Level Controller Malfunction.

END 6

Off Normal Procedure OP-901-110 Pressurizer Level Control Malfunction Revision 009 E1 PRESSURIZER LEVEL CONTROL CHANNEL MALFUNCTION NOTE Selecting the non-faulted channel may cause automatic actions to occur if actual level is not at program level.

1. Place Pressurizer Level Controller (RC-ILIC-0110) in MAN and adjust OUTPUT to slowly adjust Letdown flow to restore Pressurizer level.

2. Transfer Pressurizer Level Control CHANNEL SELECT switch to non-faulted channel.

3. Transfer Pressurizer CHANNEL SELECT LO LEVEL HEATER CUTOFF switch to non-faulted channel.

4. Verify desired backup Charging pumps in AUTO.

5. Verify all PROPORTIONAL and BACKUP HEATER BANKS reset.

6. Place Pressurizer Level Controller (RC-ILIC-0110) in AUTO and verify Pressurizer Level is being restored to setpoint.

7. Verify Pressurizer level controlling at program setpoint in accordance with Attachment 1, Pressurizer Level Versus Tave Curve.

8. Refer to the following Technical Specifications for Operability determination.

3.2.8, Power Distribution Systems, Pressurizer Pressure 3.3.3.5, Instrumentation, Remote Shutdown Instrumentation 3.3.3.6, Instrumentation, Accident Monitoring Instrumentation 3.4.3.1, Reactor Coolant System, Pressurizer END 7

Off Normal Procedure OP-901-521 Severe Weather and Flooding Revision 322 E2 TORNADO W ATCH/W ARNING (CONTD)

PLACEKEEPER START DONE N/A

3. Review OP-100-010, Equipment Out of Service and expedite restoration of vital plant systems and components to service, including performing the following:

Verify both Emergency Diesel Generators are Operable.

Verify at least one battery charger Operable for each battery bank.

Verify operability of Main Turbine DC Oil Pumps.

Verify operability of Steam Generator Feed Pumps DC Oil Pumps.

Verify availability of Portable Boiler.

Verify operability of required Cooling Tower fans located within the missile protected area.

Verify both Diesel Driven Fire Pumps are Operable.

Continuous

4. If current or expected weather conditions challenge offsite power availability and core alterations are in progress, then perform the following:

4.1 Place all refueling equipment in parked position or an otherwise safe condition, as applicable.

4.2 When weather conditions permit, then refueling activities may be resumed.

15

SD-CC cells. Each cell consists of two 40 foot vertical cooling bundles containing finned tubes.

The two cooling bundles in a cell are arranged in a V-shape. Each bundle is a four-pass heat exchanger with installed vent and drain valves. Each bundle has normally open manual butterfly isolation valves. The heat from the CCW flow passing through the tubes is dissipated to atmosphere by forcing air across the finned tubes using fans.

Each of the five cells contains three 40 Hp fans for a total of 15 fans per DCT. These fans run as required to remove heat from the system.

The DCTs are operated in three modes, dependent upon the cooling demand situation.

In the normal power operations mode the cooling towers remove heat from the CCW system. Each tower removes 27.25 million Btu/hr when both towers are in operation with a CCW flow of 6500 gpm and a maximum water outlet temperature of 100.65 F.

When ambient air temperature is 93°F or above the wet towers assist the dry towers, through the CCW Heat Exchangers, in maintaining CCW temperature below 100 F.

During normal shutdown mode, the wet and dry towers operate in a fashion identical to the normal mode. Three pump CCW operation may be required for high outdoor ambient temperature during shutdown. In this instance the flow would be approximately 8000 gpm.

During a LOCA (accident mode), both dry towers will operate in conjunction with the CCW heat exchangers to remove all heat required for a safe shutdown. By design, the DCTs remove approximately 60% of the total heat load rejected by the ultimate heat sink during an accident. The wet cooling towers will operate to assist CCW in heat removal until the cooling load is reduced to a level that the dry towers can handle alone.

With only one train of CCW/ACCW operating during an accident, the DCT would remove 113.38 million Btu/hr. The design assumptions are that CCW flow is 6900 gpm, outside air temperature (dry bulb) is 102 F, and 5% of the tubes are not in service.

Maximum water outlet temperature would be 131.1 F.

Sixty percent of each DCT is protected from tornado generated missiles by grating located above the tower. This grating, or missile shield, directly protects the cells containing DCT Fans 7 through 15 (Figure 4).

REVISION 22 PAGE 19 OF 123

LIMITING CONDITION FOR OPERATION trains of ultimate heat sink (UHS)

WATERFORD 3/4 712 Amendment No. 0$, 123

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

c. With a Tornado Watch in effect, all 9 DCT fans under the missile protected portion of the DCT shall be OPERABLE. If the number of fans OPERABLE is less than required, restore the inoperable fan(s) to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d. With any UHS fan inoperable, determine the outside ambient temperature at least once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and verify that the minimum fan requirements of Table 3.7-3 are satisfied (required only if the associated UHS is OPERABLE).

SURVEILLANCE REQUIREMENTS 4.7.4. Each train of UHS shall be determined OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying the average water temperature and water level to be within specified limits.

b. In accordance with the Surveillance Frequency Control Program, by verifying that each wet tower and dry tower fan that is not already running, starts and operates for at least 15 minutes.

WATERFORD - UNIT 3 3/4 7-13 AMENDMENT NO. 95, 123, 208, 249

TABLE 3.7-3 ULTIMATE HEAT SINK MINIMUM FAN REQUIREMENTS PER TRAIN DRY COOLING TOWER AMBIENT CONDITION DRY BULB > 97°F < 97°F DRY BULB > 91°F < 91°F DRY BULB Fan 15 14* 12*

Requirements(1)

WET COOLING TOWER Fan Requirements - 8 (1)

With any of the above required DCT Fans inoperable comply with ACTION d.

• With a tornado watch in effect, all 9 DCT fans under the missile protected portion of the DCT shall be OPERABLE.

WATERFORD - UNIT 3 3/4 7-14 AMENDMENT NO. 95,123,139, 237

PLANT SYSTEMS BASES 3/4.7.4 ULTIMATE HEAT SINK The limitations on the ultimate heat sink level, temperature, and number of fans ensure that sufficient cooling capacity is available to either (1) provide normal cooldown of the facility, or (2) to mitigate the effects of accident conditions within acceptable limits.

The UHS consists of two dry cooling towers (DCTs), two wet cooling towers (WCTs),

and water stored in WCT basins. Each of two 100 percent capacity loops employs a dry and wet cooling tower.

>(EC-38632, Ch. 72)

Each DCT consists of five separate cells. Cooling air for each cell is provided by 3 fans, for a total of 15 per DCT. Dry cooling tower fan operability is maintained by operating in fast or auto mode. The cooling coils on three cells of each DCT (i.e. 60%) are protected from tornado missiles by grating located above the coils and capable of withstanding tornado missile impact.

With a Tornado Watch in effect and the number of fans OPERABLE within the missile protected area of a DCT less than that required by Table 3.7-3, ACTION c requires the restoration of inoperable fans within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or plant shutdown as specified. This ACTION is based on FSAR analysis (subsection 9.2.5.3.3) that assumes the worst case single failure as, 1 emergency diesel generator coincident with a loss of offsite power. This failure occurs subsequent to a tornado strike and 60% cooling capacity of a DCT is assumed available.

<(EC-38632, Ch. 72)

>(DRN 04-1243, Ch. 38)

Each WCT has a basin which is capable of storing sufficient water to bring the plant to safe shutdown under all design basis accident conditions. Item a of LCO 3/4.7.4 requires a minimum water level in each WCT basin of 97% (-9.86 ft MSL). When the WCT basin water level is maintained at -9.86 ft MSL, each basin has a minimum capacity of 174,000 gallons.

This minimum WCT basin capacity contains enough volume to account for water evaporation and drift losses expected during a LOCA. Additional volume is needed from the second WCT basin to handle the non-essential load of fuel pool cooling during the LOCA. (The WCTs can be manually interconnected through a Seismic Category I line.) The WCT basin is also credited as a source of Emergency Feedwater (EFW). The WCT minimum capacity bounds the amount of EFW required from the WCT basin for all design basis accidents. Each WCT consists of two cells, each cell is serviced by 4 induced draft fans, for a total of 8 per WCT. There is a concrete partition between the cells that prevents air recirculation between the fans of each cell.

<(DRN 04-1243, Ch. 38)

Table 3.7-3 specifies increased or decreased fan OPERABILITY requirements based on outside air temperature. The table provides the cooling tower fan OPERABILITY requirements that may vary with outside ambient conditions. Fan OPERABILITY requirements are specified for each controlling parameter (i.e., dry bulb temperatures for DCT fans. The calculated temperature values (EC-M95-009) associated AMENDMENT NO. 95, 123, 139 WATERFORD - UNIT 3 B 3/4 7-4 CHANGE NO. 4, 38, 72

Off Normal Procedure OP-901-504 Inadvertent ESFAS Actuation Revision 010 E2 INADVERTENT CSAS PLACEKEEPER START DONE N/A CAUTION SHUTDOWN COOLING HX B OUTLET STOP CHECK, CS-117B, MAY LEAK BY WHILE OPERATING CONTAINMENT SPRAY HDR B ISOLATION, CS-125B, WHICH MAY CAUSE WATER TO FLOW OUT OF THE CONTAINMENT SPRAY RISER IF LPSI PUMP B IS ALSO RUNNING.

1. Secure both Containment Spray Pumps by placing control switch(es) to OFF.

CAUTION IF COMPONENT COOLING WATER IS LOST TO REACTOR COOLANT PUMP SEALS FOR >10 MINUTES, THEN RESTORING COMPONENT COOLING WATER TO REACTOR COOLANT PUMPS MAY RESULT IN SEAL FAILURE.

2. Within 3 minutes restore CCW flow to Reactor Coolant Pumps as follows:

2.1 Open the following valves:

CC-713 RCP OUTLET OUTSIDE ISOL CC-641 RCP INLET OUTSIDE ISOL CC-710 RCP OUTLET INSIDE ISOL 16

Off Normal Procedure OP-901-504 Inadvertent ESFAS Actuation Revision 010 E2 INADVERTENT CSAS (CONTD)

PLACEKEEPER START DONE N/A NOTE Manual override is accomplished by positioning control switch to CLOSED, then to OPEN. If after 100 seconds CCW Return temperature is not less than 145 F, Seal Cooler will isolate.

2.2 Verify the following Reactor Coolant Pump CCW Isolation valves Open:

CC-679A/CC-6651A 1A RCP SEAL COOLER CC-679B/CC-6651B 1B RCP SEAL COOLER CC-680A/CC-666A 2A RCP SEAL COOLER CC-680B/CC-666B 2B RCP SEAL COOLER NOTE Component Cooling Water flow can be raised to the Reactor Coolant Pumps by taking the control switch for CC-963A, SHDN HX A OUTLET, to Setpoint.

2.3 Verify Open CC-200A & CC-727, SUCT & DISCH HEADER TIE VALVES A TO AB.

17

Off Normal Procedure OP-901-504 Inadvertent ESFAS Actuation Revision 010 E2 INADVERTENT CSAS (CONTD)

PLACEKEEPER START DONE N/A

3. If in Mode 1 or 2 and CCW flow cannot be restored to Reactor Coolant Pumps within 3 minutes, then perform the following:

3.1 Trip the Reactor.

3.2 Stop the affected Reactor Coolant Pumps.

3.3 Go to OP-902-000, Standard Post Trip Actions.

4. If in Mode 3 or 4 and CCW flow cannot be restored to Reactor Coolant Pumps within 3 minutes, then Stop the affected Reactor Coolant Pumps.

Continuous

5. Monitor Reactor Coolant Pumps for degradation and contact Duty Engineering for guidance concerning continued RCP operation.

6. If CVR-101 and CVR-201 Opened on High Containment to Annulus Differential Pressure, then verify Containment Vacuum Relief resets as follows:

When computer point D51403 indicates NT HI, then Close CVR-101.

When computer point D51402 indicates NT HI, then Close CVR-201.

7. Investigate cause of inadvertent CSAS and initiate corrective action.

18

SD-PPS TABLE 3- ESFAS SIGNALS SETPOINT AND TRIP SIGNAL COINCIDENCE BASIS Safety Injection Containment high pressure Protect cladding and RCS Actuation Signal >17.1 PSIA (2/4) boundaries (SIAS) Pressurizer low pressure 1684 PSIA (variable)(2/4)

Containment Containment high pressure Protect Containment Isolation Actuation >17.1 PSIA (2/4) boundary and minimize Signal (CIAS) Pressurizer low pressure <1684 release of radioactivity to PSIA (variable)(2/4) environment Containment Containment high-high pressure Protect Containment Spray Actuation >17.7 PSIA boundary and minimize Signal (CSAS) AND release of radioactivity to Auto SIAS (2/4) environment Main Steam SG low pressure <666 PSIA Protect Containment Isolation Signal (variable) (2/4) boundary and minimize (MSIS) Containment high pressure RCS cooldown reactivity

>17.1 PSIA (2/4) addition Emergency SG low level <27.4% (2/4) Protect against loss of Feedwater AND heat sink to protect Actuation Signal Either of cladding and RCS (EFAS-1) - SG high DP >123.0 PSID (2/4) boundaries. Logic selects (EFAS-2) OR intact SG.

- SG pressure >666 PSIA (variable) (NOT low) (2/4)

Recirculation RWSP low level <10% Ensures long term cooling Actuation Signal capability (RAS)

Manual Manual (2/2 pushbuttons CP-7) Operator judgment Manual (2/2 pushbuttons CP-8)

Revision 17 Page 83 of 158

3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion l7 of Appendix A to 10 CFR Part 50.

(DRN 04-1243, Ch. 38; EC-1735, Ch. 55; EC-10725, Ch. 56)

(DRN 04-1243, Ch. 38; EC-1735, Ch. 55; EC-10725, Ch. 56)

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single failure of the other onsite A.C. source. When one diesel generator is inoperable to perform either preplanned maintenance (both preventive and corrective) or unplanned corrective maintenance work, the allowed-outage-time (AOT) can be extended from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 10 days, if a temporary emergency diesel generator (TEDG) is verified available and aligned for backup operation to the permanent plant EDG removed from service. The TEDG will be available prior to removing the permanent plant EDG from service for the extended preplanned maintenance work or prior to exceeding the 72-hour AOT for the extended unplanned corrective maintenance work. A Configuration Risk Management Program (CRMP) is implemented to assess risk of this activity when applying this ACTION. The TEDG availability is verified by: (1) starting the TEDG and verifying proper operation; (2) verifying 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> onsite fuel supply; and (3) ensuring the TEDG is aligned to supply power through a 4.16 kV non-safety bus to the 4.16kV safety bus. A status check for TEDG availability will also be performed at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the initial TEDG availability verification. The status check shall consists of: (1) verifying the TEDG equipment is mechanically and electrically ready for manual operation; (2) verifying 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> onsite fuel supply; and (3) ensuring the TEDG is aligned to supply power through a 4.16 kV non-safety bus to the 4.16 kV safety bus. If the TEDG becomes unavailable during the 10 day AOT and cannot be restored to available status, the EDG AOT reverts back to 72-hours. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> begins with the discovery of the TEDG unavailability, not to exceed a total of 10 days from the time the EDG originally became inoperable. The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, "Availability of Electrical Power Sources,"

December 1974. When one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE. This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term verify as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons. It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component.

AMENDMENT NO. 92, 166, WATERFORD - UNIT 3 B 3/4 8-1 CHANGE NO. 38, 55, 56

WATERFORD 3 SES OP-902-006 Revision 018 Page 11 of 36 LOSS OF FEEDWATER RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Restore Steam Generator Inventory

• 9. Restore feedwater to at least ONE SG by performing ANY of the following:

• a. Perform ANY of the following to a.1 Perform the following:

establish EFW to at least ONE SG:

IF local control of EFW Pump Start ALL available EFW Pumps. AB necessary, THEN establish local control IF EFW Pump AB has tripped, of EFW Pump AB using THEN reset as follows: Appendix 36, Local Manual Control of EFW Pump AB

1) Close BOTH PUMP AB TURB Turbine.

STM SUPPLY valves:

IF ANY EFW Pump available, MS 401A AND NOT operating properly, THEN perform ANY of the MS 401B following as necessary:

2) Locally depressurize Steam 1) Verify adequate suction Supply header using MS-407, source aligned.

EFW Pump AB Drip Pot Normal Drain Bypass. 2) Verify Discharge flowpath aligned to feed

3) Close MS 416, EFW Pump AB SG(s).

Turbine Stop valve.

3) Consider venting pump

4) Verify Mechanical overspeed casing/suction piping.

reset.

5) Open MS 416, EFW Pump AB Turbine Stop valve.

6) Open BOTH PUMP AB TURB STM SUPPLY valves:

MS 401A MS 401B (continue) (continue)

WATERFORD 3 SES OP-902-009 Revision 315 Page 6 of 206 STANDARD APPENDICES Appendix 1 Page 1 of 1 1.0 Diagnostic Flow Chart START GO TO Is Are FRP Is RCS BOTH OP-902-008 at least CONSIDER Subcooling No SG pressures No CONSIDER ONE RCP No ESD Has Is LOOP/LOFC < 28 F AND stable > 885 psia AND running? OP-902-004 ANY event Reactivity OP-902-003 or lowering? stable or No Control met? No been rising?

(Note 1) diagnosed?

Yes Yes Yes Yes Is Yes SGTR No Does Indicated?

at least CONSIDER (Note 3) Can a Is ONE SG have No at least LOF single event No adequate FW? be diagnosed?

ONE 125 VDC No OP-902-006 Yes (Note 2) CONSIDER (Note 4)

SAFETY bus energized? CONSIDER LOCA Yes SGTR OP-902-002 OP-902-007 Yes GO TO Yes Is Appropriate PZR EOP pressure No Is Are

> 1750 psia AND CNTMT CNTMT Is stable or pressure No Radmonitor(s) No CONSIDER Note 4: The LOCA, SGTR, at least rising?

CONSIDER < 16.4 psia AND in alarm or ESD ESD, and LOF events do ONE 4.16 KV No NOT require Off-Site Power LOOP/LOFC stable or unexplained OP-902-004 NON SAFETY bus Yes for mitigation and should be lowering? rise?

energized? OP-902-003 considered as a single event Is when LOOP also occurs.

Yes Yes PZR Note 5: If SG pressures Yes level > 7% CONSIDER No good due to MSIS isolating AND LOCA leak, then consider ESD.

stable or OP-902-002 Is rising?

at least CONSIDER Note 2: Adequate feedwater is ANY of ONE 4.16 KV No CONSIDER the following to at least ONE SG: ESD SBO Yes MFW/AFW available with OP-902-004 SAFETY bus Is OP-902-005 10 - 76% NR SG level energized? MFW/EFW restoring SG to SGTR Are 55 - 70% NR Indicated? No ALL 100% EFW capacity available to (Note 3) acceptance No Yes restore level criteria Are satisfied?

Note 3: Indications of SGTR BOTH Note 1: Reactivity Control Criteria Unexplained rise in Steam Plant Yes CONSIDER SG pressures No Reactor Power lowering Startup rate negative activity SGTR > 885 psia AND Yes GO TO Maximum of ONE CEA NOT fully Steam Plant Activity monitor alarms OP-902-007 stable or rising? RTR inserted OR Emergency Boration Unexplained level rise in ANY SG (Note 5)

Unexplained loss of RCS inventory OP-902-001 in progress Reactor Power < 10-4% (15 min)

Yes

WATERFORD 3 SES OP-902-009 Revision 315 Page 200 of 206 STANDARD APPENDICES Appendix 36 Page 1 of 2 36.0 Local Manual Control of EFW Pump AB Turbine INSTRUCTIONS

---

NOTE -------------------------------------------------------------

This appendix is implemented when power has failed to EFW Pump AB Turbine governor or power failure to the governor is imminent.

1. Obtain dedicated strobe scope stored in +35 Relay Room NFPA 805 Locker.

2. IF EFW Pump AB Turbine has tripped, THEN reset as follows:

a. Open EFW-EBKR-AB-37, Emergency Feedwtr Pump AB Turbine Gov (MS-417) (DC) breaker.

b. Open MS-EBKR-311AB-3C, EFW Pump AB Turbine Stop valve (MS-416) breaker.

c. Using local Handwheel verify MS-416, EFW Pump AB Turbine Stop valve closed,

d. Verify EFW Pump AB Turbine mechanical overspeed trip assembly is Reset.

e. Monitor EFWPT shaft speed using strobe scope.

CAUTION Exceeding 4800 RPM shaft speed could result in a mechanical overspeed trip.

f. Slowly throttle MS-416, EFW Pump AB Turbine Stop valve, to obtain desired EFW flow.

WATERFORD 3 SES OP-902-009 Revision 315 Page 201 of 206 STANDARD APPENDICES Appendix 36 Page 2 of 2 INSTRUCTIONS

3. IF EFW Pump AB Turbine is operating, THEN control as follows:

a. Open MS-EBKR-311AB-3C, EFW Pump AB Turbine Stop valve (MS-416) breaker.

b. Monitor EFW Pump AB Turbine shaft speed using strobe scope.

c. Throttle MS-416, EFW Pump AB Turbine Stop valve until governor no longer controls speed.

d. Open EFW-EBKR-AB-37, Emergency Feedwtr Pump AB Turbine Gov (MS-417) (DC) breaker.

CAUTION Exceeding 4800 RPM shaft speed could result in a mechanical overspeed trip.

e. Slowly throttle MS-416, EFW Pump AB Turbine Stop valve, to obtain desired EFW flow.

End of Appendix 36

SD-EFW Steam inlet pressure to the EFW pump turbine is indicated on CP-8. The indicator has a range of 0-1300 PSIG. This steam pressure indication is also sent to the Plant Monitoring Computer. Indication is not provided on LCP-43. This pressure indication is processed through PAC and will not be available to the operator on CP-8 or LCP-43 if the AB buses lose AC power.

Indicators of turbine speed are located on CP-8 in the Control Room and locally next to the turbine on the governor control panel. The speed monitor located in the local panel sends a DC voltage signal that is proportional to the turbine shaft RPM to these indicators. The indicators have a range of 0-6000 RPM. An additional speed monitor which does not indicate turbine speed is provided to generate the electrical overspeed trip at 110% rated speed (4895 RPM). This speed indication on CP-8 will not be available unless there is AC power available to the AB buses.

The turbine is controlled by the turbine stop valve and turbine governor valve. The operation and control of these valves is presented later in this SD along with the respective valves.

The worm gear-driven pump on the turbine supplies oil to the turbine governor actuator and lubricating oil to the bearings. The oil pump takes suction from the equalizer pipe and routes the oil to the bearings and governor through a filter and a cooler. Duplex Cuno (25 micron) filter elements are used. All oil added to the governor oil sump must be pre-filtered prior to addition. This extremely clean oil is required due to the close tolerances within the governor control system.

The pump is a six stage, horizontal split-case, centrifugal unit manufactured by Bingham-Willamette. The pump design capacity is 780 GPM (includes 80 GPM recirculation flow) at 1163 PSIG to ensure 100% of adequate cooling capability for the core. Runout flow for the pump is 1000 GPM. The recirculation line for EFW Pump AB is similar to the motor-driven pump lines. Actual pump recirculation flow indication is provided on the plant computer. Cooling water for the pump seals, packing, and turbine oil cooler is supplied from an extraction point on the first stage of the pump thus providing a guaranteed source of cooling water under all operating conditions.

The inboard and outboard pump and turbine bearing temperatures are monitored by the plant computer. The pump and turbine are monitored for high vibration using a local panel with input to the plant computer. The high vibration alarm must be reset locally. The turbine casing is protected by a safety valve which will lift if excessive backpressure exists. A local temperature gage is provided on the pump discharge line to detect potential check valve backleakage. Monitoring and indication of turbine-driven pump discharge pressure is similar to the monitoring and indication of the motor-driven Pumps discharge pressure.

Revision 13 Page 17 of 94

Off Normal Procedure OP-901-102 CEA or CEDMCS Malfunction Revision 304 E5 CEA POSITION INDICATION MALFUNCTION

[P-16410]

PLACEKEEPER START DONE N/A

1. If Plant up-power or down-power is in progress, then stop CEA movement and load changes.

2. Match Tavg and Tref by performing the following:

Adjust Turbine load in accordance with OP-010-004, Power Operations Adjust RCS boron concentration in accordance with OP-002-005, Chemical and Volume Control.

3. Verify CEA position by use of CEAC Reed Switch Position Transmitters and CEA Pulse Counter indications.

4. If any CEA is >7 inches out of alignment within its group, then go to section E1, CEA Misalignment Greater Than 7 Inches.

5. If Plant is in Modes 1 or 2, then verify that 2 of the 3 position indications agree within 5 inches in accordance with Technical Specifications.

6. If Plant is in Modes 3, 4, or 5, then verify at least one CEAC Reed Switch Position Transmitter is Operable for each CEA not fully inserted or immediately Open all Reactor Trip Circuit Breakers.

CAUTION FAILURE TO EXPEDITIOUSLY REMOVE A MALFUNCTIONING CEAC FROM SERVICE COULD RESULT IN A REACTOR TRIP.

7. If CEAC(s) position indication is behaving erratically or CEAC(s) malfunctioning then evaluate the need to remove the malfunctioning CEAC(s) from service per Attachment 2.

8. If CEAC Sensor Failure has occurred, then perform steps of Sensor Failures Section of OP-004-003, Control Element Assembly Calculation System.

19

Off Normal Procedure OP-901-102 CEA or CEDMCS Malfunction Revision 304 E5 CEA POSITION INDICATION MALFUNCTION (CONTD)

PLACEKEEPER START DONE N/A

9. If the CEAC Sensor Failure was cleared by a warm restart performed per OP-004-003, Control Element Assembly Calculation System, then perform the following:

9.1 Note warm restart in the remarks section of OP-903-001, Technical Specification Surveillance Log 9.2 Note warm restart in the Station Log.

9.3 Verify CEA positions.

9.4 If CEA positions are satisfactory, then exit this procedure.

10. If the CEAC Sensor Failure did not clear, then continue with this procedure.

11. If Reed Switch Position Transmitter for a single CEA is determined to be in error, then, at SM/CRS discretion, perform the following:

11.1 Insert affected CEA 5 inches.

11.2 Compare change in indication for suspected Reed Switch Position Transmitter against other indications for same CEA.

11.3 Return CEA to its original position.

12. If Reed Switch Position Transmitter responded properly, then note in remarks section of OP-903-001, Technical Specification Surveillance Logs, and in Station Logs.

20

Off Normal Procedure OP-901-102 CEA or CEDMCS Malfunction Revision 304 E5 CEA POSITION INDICATION MALFUNCTION (CONTD)

PLACEKEEPER START DONE N/A NOTE CPC Channel Operability is determined based on channel(s) which are affected by the failed RSPT.

13. If CEAC Sensor Failure is in a CEA group 140" withdrawn, then evaluate the targeted CPC for operability using Attachment 1.

14. If CEAC Sensor Failure is in a CEA group <140" withdrawn, then place affected CEAC channel in CEAC INOP on all CPC channels in accordance with OP-004-003, Control Element Assembly Calculation System and declare CPC with affected CEA as target CEA Inoperable.

14.1 Refer to the Technical Specification 3.2.4.b for maximum Allowed Power Limit with COLSS Operable and both CEACs Inoperable.

14.1.1 If a power reduction is required, then reduce power in accordance with OP-010-004, Power Operations.

15. If malfunction of CEA Pulse Counter indication is suspected, then enter correct CEA position in PMC database.

16. Notify maintenance and RE for assistance in identifying or correcting malfunction.

END 21

OTHER PMC COMPONENTS

• Control Room Digital Displays on CP2

- COLSS Power Margin

- CEA Group Position

- CEA Individual Position 78

Control Room Digital Displays on CP2 Margin to POL Selected Individual CEA Position Selected CEA Group Position 79

11.6 CONTAINMENT PURGE CUMULATIVE HOURS CALCULATION 11.6.1 If a Purge is performed in Modes 1-4, then perform the following on the Containment Purge Tracking Hours Calculation Data Sheet:

11.6.1.1 Enter the date and mode in the column Date/Mode.

11.6.1.2 Enter the time the Purge was Initiated in the column Time Initiated. If the Purge was in service at 0000, then enter 0000.

11.6.1.3 Enter the time the Purge was Terminated in the column Time Terminated. If the Purge was continued at 2400, then enter 2400.

11.6.1.4 Subtract column Time Initiated from column Time Terminated and enter the result in the column Time.

11.6.1.5 Initial for performance of all of the entries in column Performed.

11.6.1.6 Verifier check all entries and initial column Verified.

11.6.2 If a Purge was performed with the Plant in Modes 1-4, then complete the following on the Containment Purge Cumulative Hours/Daily Calculation Data Sheet, for the previous day, on the 19-07 shift:

11.6.2.1 Enter the date and mode in the column Date/Mode.

11.6.2.2 From the Containment Purge Cumulative Hours Tracking Data Sheet add all the entries in the column Time for the previous day and record in the column Purge Time.

11.6.2.3 Add column Purge Time to column Accumulated Purge Time and enter the result in column Sub Total.

11.6.2.4 Find the date 365 days past and record the value from column Purge Time from that date in column Purge Time Last YRS Date for the present Date.

11.6.2.5 Subtract column Purge Time Last YRS Date from column Sub Total and record this value in column Accumulated Purge Time.

11.6.2.6 Verify column Accumulated Purge Time result is <90 hours and record YES/NO in column <90 HRS.

11.6.2.7 Initial for performance of all of the entries in column Performed.

11.6.2.8 Verifier check all entries and initials in column Verified.

OP-903-001 Revision 066 Attachment 11.6 (1 of 4) 98

11.6.3 If a Containment Purge was not performed on the previous day, or the Plant was in Mode 5 or 6 on the previous day, then perform the following on the Containment Purge Cumulative Hours/Daily Calculation Data Sheet, on the 19-07 shift.

11.6.3.1 Enter the previous date and mode in the column Date/Mode.

11.6.3.2 Enter 0 in column Purge Time.

11.6.3.3 Add column Purge Time to the previous value in column Accumulated Purge Time and enter the results in column Sub Total.

11.6.3.4 Find the date 365 days and record the value from column Purge Time from that date in column Purge Time Last YRS Date for the present Date.

11.6.3.5 Subtract column Purge Time Last YRS Date from column Sub Total and record this value in column Accumulated Purge Time.

11.6.3.6 Verify column Accumulated Purge Time result is <90 hours and record YES/NO in column <90 HRS.

11.6.3.7 Initial for performance of all the entries in column Performed.

11.6.3.8 Verifier check all entries and initials in column Verified.

OP-903-001 Revision 066 Attachment 11.6 (2 of 4) 99

Containment Purge Cumulative Hours Tracking Data Sheet (Typical)

TIME TIME PERFORMED VERIFIED DATE/MODE TIME INITIATED TERMINATED (Initial) (Initial)

OP-903-001 Revision 066 Attachment 11.6 (3 of 4) 100

Containment Purge Cumulative Hours Tracking Data Sheet (Typical)

PURGE TIME ACCUMULATED <90 HRS PERFORMED VERIFIED DATE/MODE PURGE TIME SUB TOTAL LAST YRS DATE PURGE TIME YES/NO (Initial) (Initial)

OP-903-001 Revision 066 Attachment 11.6 (4 of 4) 101

( Initial / Date )

9.1.30 If Containment Purge Valves manual stops were removed ______ / ______

during Cold Shutdown, then reinstall manual stops in accordance with OP-002-010, RAB HVAC and Containment Purge.

NOTE Step 9.1.31 requirements will be satisfied by performance of step 9.1.30; however, step 9.1.31 should still be performed if manual stops were not removed and Cold Shutdown exceeds 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to satisfy Technical Specification 4.6.1.7.3. In both cases, step 9.1.31 should be completed to document CAP valve opening position. [P-1538]

9.1.31 If Cold Shutdown exceeded 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then prior to establishing ______ / ______

Containment Integrity, demonstrate operability of each Containment Purge Supply and Exhaust Isolation valve by verifying mechanical stops limit valve opening to 52 Open position. [TS 4.6.1.7.3, P-1538]

Performed IV

( Initial / Date )

( Initial / Date )

CAP-103, CAP M/U Annulus Isol Vlv ______ / ______ ______ / ______

CAP-104, CAP Makeup Cntmt Isol Vlv ______ / ______ ______ / ______

CAP-203, CAP Exhaust Cntmt Isol Vlv ______ / ______ ______ / ______

CAP-204, CAP Exhaust Ann. Isol Vlv ______ / ______ ______ / ______

9.1.32 Verify Containment Cooling Fans in operation in accordance ______ / ______

with OP-008-003, Containment Cooling System.

CAUTION OPERATION WITH RCP CONTROL BLEEDOFF PRESSURE GREATER THAN 65 PSIG MAY INCREASE RCP SEAL FACE WEAR WITHOUT SUFFICIENT RCS PRESSURE.

REFER TO RCS PRESSURE AND TEMPERATURE LIMITS GRAPH (ATTACHMENT 11.1)

IN OP-001-002, REACTOR COOLANT PUMP OPERATION.

RX 9.1.33 Verify Pressurizer level is 50% Pressurizer Level Cold ______ / ______

(RC-ILI-0103) and establish Reactor Coolant System pressure of 350 PSIA (335 - 365 PSIA).

OP-010-003 Revision 342 Attachment 9.1 (9 of 17) 30

Off Normal Procedure OP-901-511 Instrument Air Malfunction Revision 015 B SYMPTOMS B1 ALARMS

1. INST AIR PRESS BACKUP VLV OPEN (CABINET L, H-5)

2. INST AIR CMPRSR A TRIP/TROUBLE (CABINET L, L-6)

3. INST AIR CMPRSR B TRIP/TROUBLE (CABINET L, L-7)

4. INST AIR CMPRSR A SEPARATOR LVL HI/LO (CABINET L, K-6)

5. INST AIR CMPRSR B SEPARATOR LVL HI/LO (CABINET L, K-7)

6. INST AIR DRYER A TROUBLE (CABINET L, H-6)

7. INST AIR DRYER B TROUBLE (CABINET L, G-7)

8. INST AIR DRYERS BYPASSED (CABINET L, H-7)

9. VALVE OPERATORS NITROGEN BACKUP ACTUATED/TROUBLE (CABINET L, G-5)

10. INST AIR RECEIVER PRESSURE HI/LO (CABINET E, F-5)

11. TOXIC GAS MONITOR TROUBLE CHANNEL 1 (CABINET L, D-9)

12. TOXIC GAS MONITOR TROUBLE CHANNEL 2 (CABINET L, D-10)

B2 INDICATIONS

1. Instrument Air pressure >95 PSIG but <100 PSIG for >30 minutes as indicated on IA-IPI-9700 on CP-1.

2. Instrument Air pressure <95 PSIG as indicated on IA-IPI-9700 on CP-1.

3. Valves supplied by Instrument Air become sluggish, unresponsive or shift to their fail position.

3

Off Normal Procedure OP-901-511 Instrument Air Malfunction Revision 015 C AUTOMATIC ACTIONS NOTE Safety class valves listed in Attachment 1, Valves Supplied with N2/Instrument Air, and , Valves Supplied with Air Accumulators, have accumulators for operation after failure of Instrument Air System. Nitrogen (N2) accumulators have provisions for nitrogen makeup for extended periods of operation without instrument air.

1. The following are occurring as Instrument Air pressure is dropping:

SA Backup Supply for IA Press Cntl valve (SA-125) opens at SA-IPIC-9821 setpoint (normally >105 PSIG)

Instrument Air Compressor selected for standby starts at 105 PSIG Instrument Air Receiver pressure Instrument Air Dryers Bypass Solenoid valve (IA-123) Opens at 95 PSIG Instrument Air Dryer Outlet pressure (IA-IPS-9719).

4

Off Normal Procedure OP-901-511 Instrument Air Malfunction Revision 015 D IMMEDIATE OPERATOR ACTIONS NONE 5

Off Normal Procedure OP-901-511 Instrument Air Malfunction Revision 015 E SUBSEQUENT OPERATOR ACTIONS E0 GENERAL PLACEKEEPER START DONE N/A

1. If Instrument Air pressure drops to 65 PSIG, then trip the Reactor and perform OP-902-000, Standard Post Trip Actions, concurrently with this procedure.

2. Dispatch an operator to the Air Compressors and verify the following:

All Instrument Air and Station Air Compressors running loaded with normal separator levels SA is making up to IA by performing the following, as applicable:

a) Verify SA Backup Supply for IA Press Cntl Valve (SA-125) opened automatically.

b) Raise the setpoint for SA Backup Supply for IA Press Cntl Valve (SA-125) to force it open.

c) Throttle open SA to IA Pressure Regulator Bypass (SA-127) to desired position.

If Instrument Air pressure is less than 95 PSIG, then Instrument Air Dryers Bypass Solenoid valve (IA-123) Opens

3. If all of the actions of step 2 have occurred and Instrument Air pressure is still dropping, then, using the Plant Paging System, announce the following two times:

"Attention Station Personnel, Attention Station Personnel. The plant is experiencing a loss of Instrument Air Pressure. Discontinue use of Instrument Air and Station Air. Report all air usage or any air leaks to the Control Room".

4. SI-129A(B) fails open on loss of Instrument Air. If Shutdown Cooling is in service or LPSI is replacing CS, then manually throttle or force closed SI-129A(B) in accordance with OP-009-008, Safety Injection System.

6

Administrative Procedure OP-100-017 Emergency Operating Procedure Implementation Guide Revision 004 5.10 REFERENCE TO OTHER PROCEDURES 5.10.1 The EOPs are designed to minimize the interface with other procedures. The EOPs include standard appendices that operators may need during recovery.

This provides the control room staff with an easily located set of instructions while minimizing the simultaneous use of other procedures. Where standard appendices are needed, the required procedure will be referenced in the body of the EOP.

5.10.2 Other procedures (Normal Operating Procedures, Alarm Response, Off-Normals) are usually not normally needed to supplement the EOPs, but may be used as directed by the CRS.

15

WATERFORD 3 SES OP-902-009 Revision 315 Page 6 of 206 STANDARD APPENDICES Appendix 1 Page 1 of 1 1.0 Diagnostic Flow Chart START GO TO Is Are FRP Is RCS BOTH OP-902-008 at least CONSIDER Subcooling No SG pressures No CONSIDER ONE RCP No ESD Has Is LOOP/LOFC < 28 F AND stable > 885 psia AND running? OP-902-004 ANY event Reactivity OP-902-003 or lowering? stable or No Control met? No been rising?

(Note 1) diagnosed?

Yes Yes Yes Yes Is Yes SGTR No Does Indicated?

at least CONSIDER (Note 3) Can a Is ONE SG have No at least LOF single event No adequate FW? be diagnosed?

ONE 125 VDC No OP-902-006 Yes (Note 2) CONSIDER (Note 4)

SAFETY bus energized? CONSIDER LOCA Yes SGTR OP-902-002 OP-902-007 Yes GO TO Yes Is Appropriate PZR EOP pressure No Is Are

> 1750 psia AND CNTMT CNTMT Is stable or pressure No Radmonitor(s) No CONSIDER Note 4: The LOCA, SGTR, at least rising?

CONSIDER < 16.4 psia AND in alarm or ESD ESD, and LOF events do ONE 4.16 KV No NOT require Off-Site Power LOOP/LOFC stable or unexplained OP-902-004 NON SAFETY bus Yes for mitigation and should be lowering? rise?

energized? OP-902-003 considered as a single event Is when LOOP also occurs.

Yes Yes PZR Note 5: If SG pressures Yes level > 7% CONSIDER No good due to MSIS isolating AND LOCA leak, then consider ESD.

stable or OP-902-002 Is rising?

at least CONSIDER Note 2: Adequate feedwater is ANY of ONE 4.16 KV No CONSIDER the following to at least ONE SG: ESD SBO Yes MFW/AFW available with OP-902-004 SAFETY bus Is OP-902-005 10 - 76% NR SG level energized? MFW/EFW restoring SG to SGTR Are 55 - 70% NR Indicated? No ALL 100% EFW capacity available to (Note 3) acceptance No Yes restore level criteria Are satisfied?

Note 3: Indications of SGTR BOTH Note 1: Reactivity Control Criteria Unexplained rise in Steam Plant Yes CONSIDER SG pressures No Reactor Power lowering Startup rate negative activity SGTR > 885 psia AND Yes GO TO Maximum of ONE CEA NOT fully Steam Plant Activity monitor alarms OP-902-007 stable or rising? RTR inserted OR Emergency Boration Unexplained level rise in ANY SG (Note 5)

Unexplained loss of RCS inventory OP-902-001 in progress Reactor Power < 10-4% (15 min)

Yes

Operating Instruction OI-024-000 Maintaining Active SRO/RO Status Revision 314 5.0 PROCEDURE 5.1 MAINTAINING CURRENT SRO/RO STATUS 5.1.1 Active SRO Status is maintained by standing a minimum of five complete 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts per calendar quarter while actively performing the functions of a SRO. Only shifts stood as Shift Manager or Control Room Supervisor may be credited for Active SRO Status. Watches stood as Fuel Handling Supervisor may only be credited for Active SRO Status as Fuel Handling Supervisor. [P-13844]

5.1.2 Active RO Status is maintained by standing a minimum of five complete 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts per calendar quarter while actively performing the functions of an RO. Only shifts stood as ATC or BOP are credited for Active RO Status. Shifts stood as Administrative RO may not be credited. Only CRSs that had been previously qualified ROs can maintain RO proficiency by maintaining their SRO proficient.

Instant SROs and Shift Managers may not stand RO nor establish active or proficient RO status. [P-13844]

5.1.3 Standing the required number of proficiency watches in one calendar quarter validates Active Status to the licensee for the following calendar quarter.

5.1.4 Newly licensed personnel attain Active Status for the current and following calendar quarter upon receipt, by the facility, of their Valid License Number.

5.1.5 Inactive Status licensees who complete training watch requirements are authorized Active Status for the current and following calendar quarter.

5.1.6 SRO licensees should not routinely stand RO watches.

5.1.6.1 Instant SROs and Shift Managers may not stand RO watches.

5.1.6.2 A qualified SRO may stand RO if they have previously been qualified as an RO and one of the following conditions is met: [CR-WF3-2014-03650]

If filling a planned RO vacancy, then prior to standing the RO watch he/she should be observed in the simulator to ensure competency as a RO.

OR If filling an unplanned RO vacancy, then authorization must be obtained from the Senior Operations Manager or the Operations Manager-Shift.

5.1.7 The licensee's name and shift position must be logged in the Station Log or Refueling Supervisor Logbook.

5.1.8 All conditions of Attachment 6.4 and all administrative requirements of Attachment 6.5 must be met for an individual to stand watch in a Reactor Operator or Senior Reactor Operator position. [CR-WF3-1997-00831]

8

Operating Instruction OI-024-000 Maintaining Active SRO/RO Status Revision 314 5.2 UPGRADING FROM INACTIVE TO ACTIVE STATUS NOTE Only one trainee can be assigned to a qualified watchstander for the purpose of completing an under instruction watch.

5.2.1 RO and SRO licensees must stand a minimum of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> as a trainee on shift in the highest qualified position. A complete plant tour and all required shift turnover requirements must be completed during this 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. The 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> should be started and completed within a 30 day period. [P-13838]

5.2.1.2 The RO/SRO licensee trainee must stand watch under the onshift watchstander.

5.2.2 RO licensees must stand parallel watches as ATC or BOP trainee.

5.2.3 SRO licensee trainees must stand watches in the position to which the individual will be assigned. (i.e. Shift Manager to stand as SM and Control Room Supervisor to stand as CRS.) [P-13838]

5.2.4 For SRO licensees upgrading to Active SRO Status as Fuel Handling Supervisor only, a minimum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as Fuel Handling Supervisor trainee while conducting core alterations or fuel movement is required in addition to completion of a plant tour of all accessible areas. [P-13838]

5.2.5 Operations Manager-Shift authorization must be obtained prior to standing trainee watches. This is documented on the Authorization to Activate line of Attachment 6.1 5.2.6 All conditions of Attachment 6.4 and all administrative requirements of Attachment 6.5 must be met for an individual to resume Active Status in a Reactor Operator or Senior Reactor Operator position.

[CR-WF3-1997-00831]

5.2.7 Use Attachment 6.1, Checklist for Upgrading from Inactive to Active Status, and Attachment 6.2, Plant Tour Checklist to record the completion of necessary trainee requirements.

5.2.8 SROs re-establishing their Active SRO status also re-establish their RO Active status only if previous licensed as a RO. Instant SROs and Shift Managers may not stand RO nor establish Active RO status.

9

WATERFORD 3 SES OP-902-004 Revision 016 Page 29 of 53 EXCESS STEAM DEMAND RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS Restore Letdown

• 30. IF Letdown is isolated, AND BOTH of the following conditions exist:

HPSI throttle criteria are met (Step 20)

Letdown is needed or desired THEN restore Letdown using Appendix 9, "Letdown Restoration."

Verify RCS is NOT Water Solid

• 31. Check the RCS is NOT in a water solid 31.1 IF the RCS is water solid, condition as indicated by the following: THEN maintain the RCS within Attachments 2A-D, "RCS Pressure and

a. RCS pressure responds slowly to Temperature Limits" by ANY of the RCS inventory and temperature following:

changes.

Control RCS temperature.

b. At least ONE of the following conditions met: IF HPSI throttle criteria are met (Step 20),

PZR level less than 100% THEN perform ANY of the following:

Reactor Vessel Head level Control Charging and Letdown.

less than 100%

Throttle HPSI flow as necessary.

Technical Guide for Excess Steam TG-OP-902-004 Demand Recovery Procedure Revision 307 Step Number 31 Verify RCS Steam Bubble Objective The intent of this step is to determine if a steam bubble exists in the RCS. Since it is expected that a steam bubble exists in the RCS, the step is worded to expect that condition.

Instructions The EOPs are designed to address water solid conditions separately from non-water solid conditions. In order to initiate the appropriate actions, the operator must first evaluate the RCS to determine if a steam bubble is present or if it is water solid. The intent of this step is to aid the operator in making that determination. By definition in the EOPs, water solid conditions refer to the RCS being full of subcooled water, with possible exception of some small pockets of non-condensable gases in high points of the system. Conversely, non-water solid conditions imply that there is a steam bubble somewhere in the RCS. The steam bubble could be located in the pressurizer, reactor vessel head, or the steam generator U-tubes. Therefore, in this step, the operator is prompted to look at all RCS indications that may provide evidence that the RCS is not in a water solid condition. The following is a minimum list of RCS indications that should be considered:

No exaggerated or severe pressure response Observation of an abnormal or exaggerated Pressurizer pressure response due to operation of heaters, changes in RCS inventory, or changes in RCS temperature are good indicators the RCS is solid. Operator training and simulator exercises emphasize recognition of normal pressure response (with steam bubble) versus abnormal pressure response (water-solid).

66

Technical Guide for Excess Steam TG-OP-902-004 Demand Recovery Procedure Revision 307 Step Number 31 Verify RCS Steam Bubble (contd)

Pressurizer level less than 100%

Pressurizer level indicating on scale, i.e., < 100%, is a good indication that there is a saturated steam-water interface in the Pressurizer. Indicated level may not be representative of actual level, if instrument uncertainties are not considered, if the instrument has decalibrated due to a change in temperature, if leak is located in the top of the Pressurizer, or if harsh containment conditions have exaggerated instrument uncertainties in a non-conservative direction. All available channels of level instrumentation should be checked to corroborate the level reading. One channel of level indication should not be relied upon alone to confirm the existence of a steam bubble in the Pressurizer. Pressurizer level alone should not be relied upon to confirm that the RCS is not solid. RVLMS, the indications of saturated conditions in the RCS and RCS pressure response, must also be considered.

RVLMS less than 100%

RVLMS may be used to identify or corroborate a bubble in the reactor vessel head region. Level indication alone may not confirm that the bubble is saturated steam.

Head temperature and Pressurizer pressure can be used to validate that saturated conditions exist.

Contingency Actions The contingency action is intended to guide the operator in controlling the RCS while it is water solid. The goal of the step is to maintain the RCS within the limits of the RCS PT curves by controlling RCS temperature and pressure. When the RCS is intact and water-solid, the RCS pressure is sensitive to changes in RCS inventory. If the RCS is intact and inventory and temperature are not controlled carefully, there is a potential for a pressure excursion which could over pressurize the RCS, resulting in challenging the primary safety valves, or pressurized thermal shock of the RCS.

RCS pressure changes may result from changes in the heat input or removal from the RCS. Therefore, changes in steam generator feed rate and steaming rate should be slow and deliberate, and require the operator to be constantly aware of the effect of the change in Pressurizer pressure.

67

Technical Guide for Excess Steam TG-OP-902-004 Demand Recovery Procedure Revision 307 Step Number 31 Verify RCS Steam Bubble (contd)

The operator must also anticipate an exaggerated Pressurizer response when starting or stopping charging pumps, when increasing or decreasing letdown flow, when starting or stopping HPSI pumps, when throttling SI flow, when securing RCP bleedoff, when initiating or securing sampling, and when isolating RCS leak paths. For best control, changes should be made slowly, allowing time to monitor the effect of the actions taken.

Multiple or simultaneous changes should be avoided.

Water solid operations may continue as required within the guidelines described above, depending on plant conditions and the event recovery strategy. As long as the operators have established control of RCS temperature and inventory, there is no urgency to re-establish a bubble in the Pressurizer. In fact, the operators may conclude that it does not make sense to intentionally draw a bubble, and instead they may choose to proceed with the cooldown with the existing conditions.

Justification for Deviations There are no deviations.

References

1. EC-S98-001 L.05

2. EC-S98-001 L.06 68

NUCLEAR QUALITY RELATED EN-OP-102 REV. 18 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 44 OF 99 Protective and Caution Tagging 5.22 Main Steam Administrative Tagout

[1] The process for using a Main Steam Administrative Tagout:

Operations will Tagout the Main Steam system with a Main Steam Administrative Tagout. This Tagout will contain a MS-ADMIN component.

Once the Main Steam Boundary is tagged, then other Tagouts (e.g. Turbine Tagout) can share the MS-ADMIN tag and take credit for the Main Steam boundary.

Operations can change the Main Steam Administrative Tagout as required to support the outage. This could be moving the Boundaries from the Outboard MSIVs to the Steam Plugs or other applicable boundary. The new Main Steam boundary shall be completed prior to clearing the old Main Steam boundary.

Once all the Tagouts that use the MS-ADMIN tag are cleared then the Main Steam Administrative Tagout can be cleared.

This process is used to limit the number of revised Tagouts that maintenance must sign on and off during an outage.

Outage management should ensure that the Current Main Steam Administrative Tagout and changes to the Main Steam Administrative Tagout are appropriately communicated to site personnel.

[2] Requirements for Using the Main Steam Administrative Tagout (a) Main Steam Administrative Tagout allows work to be conducted downstream of the MSIVs, with either a proper combination of Main Steam Isolation Valves and drains, Main Steam Plugs or Reactor Coolant Pump breakers AND Pressurizer Heaters as applicable.

(b) Only work that is downstream of the MSIVs may utilize this Tagout.

(c) The plant shall be in Cold Shutdown, Refuel or Defueled modes except as specified below.

IF desired to use the Main Steam Administrative Tagout above cold shutdown THEN the only acceptable isolation boundaries are the Main Steam Isolation and bypass valves (including Main Steam Line Drain Valves and/or Main Steam Leak Collection Systems).

(d) Work Orders will not be directly assigned to this Tagout.

(e) The Main Steam Administrative Tagout shall contain a MS-ADMIN component. The Operations Manager will designate where this tag will be hung.

NUCLEAR QUALITY RELATED EN-OP-102 REV. 18 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 45 OF 99 Protective and Caution Tagging 5.22[2] cont (f) The Main Steam Administrative Tagout should only tag the acceptable boundaries identified below. Additional boundaries are not permitted on this Tagout.

(g) Acceptable boundaries:

(1) For BWRs

a. Inboard MSIVs, outboard MSIVs or MOVs (including Main Steam Line Drain Valves and/or Main Steam Leak Collection Systems) or Main Steam Line Plugs.

(2) For PWRs

a. MSIVs and MSIV Bypass Valves, or Reactor Coolant Pump Breakers AND Pressurizer Heaters. (RCP Breakers AND Pressurizer Heaters may include any boundary which would need to be included to keep personal and equipment safe downstream of the MSIVs if they are stroked. Examples include S/G Nitrogen isolation valves, and Feed water isolation valves, etc.)

(h) The boundaries may change from one acceptable boundary to another via revisions of this Tagout.

(i) In order to ensure worker safety, the new boundaries shall be in-place prior to clearing the original Tagout.

(j) The Main Steam Administrative Tagout SHALL NOT be Temporary Lifted.

(k) In the details of the Main Steam Administrative Tagout add the following note:

The Operations supervisor is required to sign on to this Tagout as the Tagout Holder.

[3] Issuing Main Steam Administrative Tagout (a) Generate a Main Steam Administrative Tagout with required components. The MS-ADMIN tag shall be sequenced last.

(b) Issue the Main Steam Administrative Tagout.

(c) Once the Initial boundary is tagged the MS-ADMIN tag will be hung.

(d) A designated Operations Supervisor will sign on to this Tagout.

(e) Notify Outage Management of boundary.

NUCLEAR QUALITY RELATED EN-OP-102 REV. 18 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 46 OF 99 Protective and Caution Tagging

[4] Tagout Development downstream of the Main Steam Administrative Tagout NOTE The MS-ADMIN component is used instead of Steam Plugs, MSIVs, Reactor Coolant Pump breakers and Pressurizer Heaters on each Tagout.

(a) MS-ADMIN listed as tagged component.

(b) Other Tagout Boundaries as required.

(c) Work Order(s) needed for the Tagout.

(d) In the details of this Tagout add the following note This Tagout uses the Main Steam Administrative Tagout for all or a portion of its boundaries. Prior to signing on this Tagout, refer to the current Main Steam Administrative Tagout for the Main Steam boundaries.

[5] Changing Boundaries within the Main Steam Administrative Tagout.

(a) Ensure plant conditions support installing the new Main Steam Administrative Tagout.

(b) Generate a revised Main Steam Administrative Tagout with appropriate components tagged.

(c) Authorize and establish the new Main Steam Administrative Tagout Boundaries.

(d) Verify the new Main Steam Administrative Tagout boundary is established (Boundaries holding and drained) to ensure the workers safety.

(e) Notify Outage Management of the changed boundaries.

(f) The designated Operations Supervisor will sign on to the new Main Steam Administrative Tagout.

(g) The designated Operations Supervisor will sign off the previous Main Steam Administrative Tagout.

(h) Clear the previous Main Steam Administrative Tagout.

NUCLEAR QUALITY RELATED EN-OP-102 REV. 18 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 47 OF 99 Protective and Caution Tagging

[6] Clearing the Main Steam Administrative Tagouts.

(a) Verify no other Tagouts are utilizing the MS-ADMIN tag.

(b) Notify the Outage Management that the administrative Tagout will be cleared.

(c) The designated Operations Supervisor will sign off the Main Steam Administrative Tagout.

(d) Clear the Main Steam Administrative Tagout.

5.23 Switch Yard & Turbine Generator Administrative Tagouts

[1] Switch Yard Administrative Tagout (a) The process for using a Switch Yard Administrative Tagout.

Operations in combination with the Authority Having Jurisdiction of the switch yard will tagout back feeding energy from the transmission system to the Main Transformer(s) or Main Generator with a Switch Yard Administrative Tagout. This tagout will contain a SW-YRD-ADMIN component.

Once the Switch Yard Boundary is tagged, then other tagouts (e.g. Main Transformer(s) tagout, Main Generator tagout, etc.) can share the SW-YRD-ADMIN tag and take credit for the Switch Yard boundary.

Operations along with the Authority Having Jurisdiction of the switch yard can change the Switch Yard Administrative Tagout as required to support the outage provided it meets an acceptable boundary as delineated by this procedure. The new Switch Yard boundary shall be completed prior to clearing the old Switch Yard boundary.

Once all the tagouts that use the SW-YRD-ADMIN tag are cleared then the Switch Yard Administrative Tagout can be cleared.

This process is used to limit the number of revised Tagouts that maintenance must sign on and off during an outage.

Outage management should ensure that the current Switch Yard Administrative Tagout and changes to the Switch Yard Administrative Tagout are appropriately communicated to site personnel.

Surveillance Procedure OP-903-001 Technical Specification Surveillance Logs Revision 066 1.0 PURPOSE 1.1 This procedure and accompanying attachments provide the means for compliance with Operations surveillance requirements that have intervals of one week as listed in the Reference Section. This procedure contains surveillance logs which are completed once per shift, logs that are completed once per day, logs which cover longer periods up to 7 days, and logs to track various component cumulative hours.

3

Surveillance Procedure OP-903-001 Technical Specification Surveillance Logs Revision 066 7.0 PROCEDURE 7.1 MODE 1-4 LOGS 7.1.1 Record data specified on Attachment 11.1, Mode 1-4 Technical Specification Surveillance Log. The taking of data for each set of Technical Specification Surveillance Logs should begin no earlier than 1100 and 2300. This data should be completed between the hours of 1200-1400, and 0000-0200 to allow timely review by the Shift Manager (SM) or Control Room Supervisor (CRS).

7.2 MODE 5-6 LOGS 7.2.1 Record data specified on Attachment 11.2, Mode 5-6 Technical Specification Surveillance Log. The taking of data for each set of Technical Specification Surveillance Logs should begin no earlier than 1100, and 2300. This data should be completed between the hours of 1200-1400, and 0000-0200 to allow timely review by the Shift Manager (SM) or Control Room Supervisor (CRS).

7.3 CUMULATIVE HOURS CALCULATIONS 7.3.1 Calculate the cumulative hours for Containment Purge, Regulating Group CEA and Group P Insertion limits in accordance with the applicable Attachment.

Copies of Attachments and data sheets shall be maintained in the Nuclear Operations Cumulative Tracking Log. Verify and update cumulative hours at the frequencies stated in Attachment 11.1 and 11.2. Record data on applicable data sheets as required.

7.3.2 Cumulative Hours Calculations may be kept by computer using Typical Forms. In this case the legal record for satisfying Technical Specifications will be a completed copy of the computer generated form maintained in Nuclear Operations Cumulative Tracking Log.

9

(DRN 02-216) 3/4.3 INSTRUMENTATION (See note below)

(DRN 02-216) 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.10 RADIOACTIVE LIQUID EFFLUENT LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Requirement 3.11.1.1 are not exceeded during releases to the environment. The alarm/trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the Offsite Dose Calculation Manual (ODCM).

APPLICABILITY: At all times.

ACTION:

a. With radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Requirement, immediately suspend release to the environment of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within 30 days if release to the environment are in progress or, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report, pursuant to Technical Specification 6.9.1.8, why this inoperability was not corrected within the time specified. Releases need not be terminated after 30 days provided the specified ACTIONS are continued.

SURVEILLANCE REQUIREMENT 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3-8.

(DRN 02-216)

NOTE: TRM Specifications 3.3.3.10 and 4.3.3.10 are part of the Offsite Dose Calculation Manual (ODCM), reference UNT-005-014. Revision of these TRM Specifications requires the approval of the General Manager Plant Operations (GMPO) in accordance with Technical Specification 6.14.

(DRN 02-216) 3/4 3-23 AMENDMENT NO. 28, 51

(DRN 02-216)

TABLE 3.3-12 (See note below)

(DRN 02-216)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS RELEASE INSTRUMENT OPERABLE INFORMATION ACTION

1. BORON WASTE MANAGEMENT SYSTEM (BWMS):

a. Radioactivity Monitor Providing Alarm and Automatic 1 Batch 1 Termination of Release(PRM-IRE-0627) Release from Boric Acid

b. Waste (Process) Flow Rate Measurement Device 1 Condensate 2 (BM-IFT-0627) Tanks

2. LIQUID WASTE MANAGEMENT SYSTEM DISCHARGE (LWMS):

a. Radioactivity Monitor Providing Alarm and Automatic 1 Batch 1 Termination of Release (PRM-IRE-0647) Release from Liquid Waste

b. Waste (Process) Flow Rate Measurement Device 1 Management 2 (LWM-IFT-0647) Tanks (DRN 02-216)

NOTE: TRM Table 3.3-12 is part of the Offsite Dose Calculation Manual (ODCM), reference UNT-005-014. Revision of this TRM Table requires the approval of the General Manager Plant Operations (GMPO) in accordance with Technical Specification 6.14.

(DRN 02-216) 3/4 3-24 AMENDMENT NO. 04, 28, 51,

(DRN 02-216)

TABLE 3.3-12 (Continued, See note below)

(DRN 02-216)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS RELEASE INSTRUMENT OPERABLE INFORMATION ACTION

3. DRY COOLING TOWER SUMPS (DCTS):

a. Radioactivity Monitor Providing Alarm and Automatic 1/sump Release 3 Termination of Release Path is

[PRM-IRE-6775 (DCTS#1) and PRM-IRE-6776(DCTS#2)] NOT Aligned to LWMS

b. Waste (Process) Flow Rate Measurement Device N/A (see Note #2) N/A (See Table Note. #1)

4. INDUSTRIAL WASTE SUMP TURBINE BUILDING (TBIWS):

a. Radioactivity Monitor Providing Alarm and Automatic 1 Release 3 Termination of Release (PRM-IRE-6778) Path is NOT Aligned to LWMS

b. Waste (Process) Flow Rate Measurement Device N/A (see Note #2) N/A (See Table Note. #1)

(DRN 02-216)

NOTE: TRM Table 3.3-12 is part of the Offsite Dose Calculation Manual (ODCM), reference UNT-005-014. Revision of this TRM Table requires the approval of the General Manager Plant Operations (GMPO) in accordance with Technical Specification 6.14.

(DRN 02-216) 3/4 3-25 AMENDMENT NO. 04, 51

(DRN 02-216)

TABLE 3.3-12 (Continued, See note below)

(DRN 02-216)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS RELEASE INSTRUMENT OPERABLE INFORMATION ACTION

5. CIRCULATING WATER DISCHARGE (CWD) - BLOWDOWN AND 1. Detectable Activity BLOWDOWN HEAT EXCHANGER DISCHARGES AND AUXILIARY in Secondary Plant COMPONENT COOLING WATER PUMPS: 2. During Blowdown of Steam Generators

a. Radioactivity Monitor Providing Alarm and initiate 1 to CW System. 4 Automatic Closure of Blowdown Valve BD-303 3. During Discharge (PRM-IRE-1900) of ACCW Basins

b. Waste (Process) Flow Rate Measurement Device N/A to CW System N/A (See Table Note. #1)

6. STEAM GENERATOR BLOWDOWN (SGB) EFFLUENT LINE During Blowdown Of S / Gs

a. Continuous Composite Sampler 1 to CW System or 4 Metal Waste Ponds (see Note #3)

(DRN 02-216)

NOTE: TRM Table 3.3-12 is part of the Offsite Dose Calculation Manual (ODCM), reference UNT-005-014. Revision of this TRM Table requires the approval of the General Manager Plant Operations (GMPO) in accordance with Technical Specification 6.14.

(DRN 02-216) 3/4 3-25a AMENDMENT NO. 51

(DRN 02-216)

TABLE 3.3-12 (Continued, See note below)

(DRN 02-216)

TABLE NOTATIONS NOTE #1 Waste (process) Flow Measurement Devices are not installed on the release paths for the DCTS, TBIWS or CWD monitors. For these release paths, pump performance curves generated in place or some form of volumetric estimate or measurement device may be used for effluent flow rate estimates.

NOTE #2 DCTS and TBIWS monitor operation should be maximized during releases to the environment, even when detectable activity is not present in the CCW/ACCW or secondary systems, to provide capability for release termination in the event that Primary to Secondary or Primary to CCW leakage occurs.

NOTE #3 The Steam Generator Blowdown Composite Sampler is capable of sampling blowdown discharge to either the CW System or Waste Ponds.

Blowdown to the Waste Ponds is not allowed unless radiation monitoring capable of release termination is added to the release path.

ACTION STATEMENTS ACTION 1 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement effluent releases via this pathway may continue provided best efforts are made to repair the instrument and that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Requirement 4.11.1.1.1 and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup.

ACTION 2 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided best efforts are made to repair the instrument and that the flow rate is estimated at least once per FOUR hours during actual releases. Pump performance curves generated in place may be used to estimate flow.

(DRN 02-216)

NOTE: TRM Table 3.3-12 is part of the Offsite Dose Calculation Manual (ODCM), reference UNT-005-014. Revision of this TRM Table requires the approval of the General Manager Plant Operations (GMPO) in accordance with Technical Specification 6.14.

(DRN 02-216) 3/4 3-26 AMENDMENT NO. 02, 51

Operating Instruction OI-038-000 Emergency Operating Procedures Revision 013 Operations Expectations / Guidance 5.4.32 Identify Success Paths The Resource Assessment Trees (RATs) are used to verify available success paths. The SFSC criteria listed on the lower portion of the RATs is used in conjunction with the success path to determine priorities.

The Resource Assessment Trees (RATs) are a tool to aid the CRS in choosing the correct path based on equipment availability. The CRS should use judgment as to whether a path can be used if no AC is available, since it may be possible to meet the success path criteria and not have any of the equipment listed. A brief should be held to confirm this choice.

When selecting Reactivity Control Success Path:

If two or more CEAs are stuck out, then RC-2 should be chosen even if RC-1 criteria is met. This is due to RC-1 having no guidance for emergency boration initiation or termination.

RC-3 should only be chosen if Charging Pumps are not available and the SM/CRS determines that emergency boration is required to meet SDM.

This path will require the RCS to be depressurized.

When selecting Inventory Control Success Path:

If SIAS has occurred, then IC-2 should be chosen.

When selecting Containment Isolation Success Path:

Success Path CI-1 should be considered not met if secondary activity exists and all EOP steps for isolating the affected SG have not been completed. MSIS is not adequate for meeting this step.

5.4.33 Implement Placekeeping If the EOPs are entered from Mode 3 or 4, then the time of entry into the procedure should be documented in the place for Time of Reactor Trip.

5.4.34 Implement Success Paths None 5.4.35 Insert CEAs The substeps for this step are listed in a preferred order. If one of the actions results in inserting the CEAs, it is not necessary or desired to perform the remaining substeps.

23

WATERFORD 3 SES OP-902-008 Revision 026 Page 22 of 284 FUNCTIONAL RECOVERY SFSC Page 1 of 15 6.0 SAFETY FUNCTION STATUS CHECK SAFETY FUNCTION:

1. Reactivity Control Tree A RC-1: CEA Insertion PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. CEAs < 2 NOT fully inserted

b. Reactor Power Dropping

c. SUR Negative Condition 2

a. Reactor Power < 10-4%

b. Reactor Power Stable or dropping

WATERFORD 3 SES OP-902-008 Revision 026 Page 23 of 284 FUNCTIONAL RECOVERY SFSC Page 2 of 15 SAFETY FUNCTION:

1. Reactivity Control Tree A RC-2: Boration Using CVCS PARAMETER CRITERIA CRITERIA SATISFIED Condition 1

a. Emergency boration 40 gpm

b. Reactor Power Dropping

c. SUR Negative Condition 2

a. Reactor Power < 10-4%

b. Reactor Power Stable or dropping Condition 3

a. Shutdown Margin Verified