ML18080A015
| ML18080A015 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/07/2018 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | Entergy Operations |
| References | |
| Download: ML18080A015 (54) | |
Text
ES-401 PWR Examination Outline Form ES-401-2 Rev. 11 Facility: ANO Unit 1 (Rev. 3)
Date of Exam: 3/7/2018 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total 1.
Emergency and Abnormal Plant Evolutions 1
2 3
4 N/A 3
3 N/A 3
18 6
2 2
2 2
1 1
1 9
4 Tier Totals 4
5 6
4 4
4 27 10 2.
Plant Systems 1
3 2
2 3
2 3
3 3
2 3
2 28 5
2 1
0 1
1 1
1 1
1 1
1 1
10 3
Tier Totals 4
2 3
4 3
4 4
4 3
4 3
38 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
3 3
2 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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it 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 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 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 These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000007 (EPE 7; BW E02&E10; CE E02)
Reactor Trip, Stabilization, Recovery / 1 X
Ability to operate and / or monitor the following as they apply to the (Vital System Status Verification):
(CFR: 41.7 / 45.5 / 45.6)
EA1.2 Operating behavior characteristics of the facility.
3.2 1
Bank 604 2010 000008 (APE 8) Pressurizer Vapor Space Accident / 3 X
Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: (CFR: 41.10 / 43.5 / 45.13)
AA2.05 PORV isolation (block) valve switches and indicators.
3.9 2
New 1214 000009 (EPE 9) Small Break LOCA / 3 X
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 system integrity, containment conditions, radioactivity release control, etc. (CFR:
41.7 / 43.5 / 45.12) 4.0 3
Mod 506 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 X
Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): (CFR 41.5,41.10 / 45.6 / 45.13)
AK3.03 Sequence of events for manually tripping reactor and RCP as a result of an RCP malfunction 3.7 4
New 1215 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X
Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup: (CFR 41.5, 41.10 / 45.6 / 45.13)
AK3.02 Actions contained in SOPs and EOPs for RCPs, loss of makeup, loss of charging, and abnormal charging 3.5 5
Bank 549 2011 000025 (APE 25) Loss of Residual Heat Removal System / 4 X
Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following:
(CFR 41.7 / 45.7)
AK2.02 LPI or Decay Heat Removal/RHR pumps 3.2*
6 Bank 033 2007 000026 (APE 26) Loss of Component Cooling Water / 8 X
Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: (CFR 41.5,41.10 / 45.6 / 45.13)
AK3.01 The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the CCWS coolers.
3.2 7
New 1240 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X
Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: (CFR: 41.7 / 43.5 / 45.13)
AA2.02 Normal values for RCS pressure 3.8 8
New 1170 000029 (EPE 29) Anticipated Transient Without Scram / 1 X
Knowledge of the interrelationships between the ATWS and the following: (CFR: 41.7 / 45.7)
EK2.06 Breakers, relays, and disconnects.
2.9 9
Mod 509
ES-401 3
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO) 000038 (EPE 38) Steam Generator Tube Rupture / 3 X
Knowledge of the operational implications of the following concepts as they apply to the SGTR:
(CFR 41.8 / 41.10 / 45.3)
EK1.03 Natural circulation 3.9 10 New 1092 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 X
Ability to operate and / or monitor the following as they apply to the (Excessive Heat Transfer) (CFR:
41.7 / 45.5 / 45.6)
EA1.3 Desired operating results during abnormal and emergency situations.
3.8 11 New 1171 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X
Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): (CFR 41.8 / 41.10 / 45.3)
AK1.01 MFW line break depressurizes the S/G (similar to a steam line break) 4.1 12 Mod 334 000055 (EPE 55) Station Blackout / 6 X
Knowledge of the reasons for the following responses as the apply to the Station Blackout:
(CFR 41.5 / 41.10 / 45.6 / 45.13)
EK3.01 Length of time for which battery capacity is designed 2.7 13 New 1209 000056 (APE 56) Loss of Offsite Power / 6 X
Ability to operate and / or monitor the following as they apply to the Loss of Offsite Power: (CFR 41.7
/ 45.5 / 45.6)
AA1.11 HPI system 3.7*
14 Bank 689 2008 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X
Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:
(CFR: 41.7 / 43.5 / 45.13)
AA2.05 S/G pressure and level meters.
3.5 15 Bank 624 2010 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water / 4 X
2.4.1 Knowledge of EOP entry conditions and immediate action steps. (CFR: 41.10 / 43.5 / 45.13 4.6 16 New 1172 000065 (APE 65) Loss of Instrument Air / 8 X
2.4.11 Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 / 45.13) 4.0 17 Bank 691 2008 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X
Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)
AK2.02 Breakers, relays 3.1 18 New 1173 (W E04) LOCA Outside Containment / 3 (W E11) Loss of Emergency Coolant Recirculation / 4 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 K/A Category Totals:
2 3
4 3
3 3
Group Point Total:
18
ES-401 4
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000001 (APE 1) Continuous Rod Withdrawal / 1 X
Ability to operate and / or monitor the following as they apply to the Continuous Rod Withdrawal:
(CFR 41.7 / 45.5 / 45.6)
AA1.02 Rod in-out-hold switch 3.6 19 Bank 422 2002 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 X
Knowledge of the reasons for the following responses as they apply to the Inoperable / Stuck Control Rod: (CFR 41.5,41.10 / 45.6 /
45.13)
AK3.06 Actions contained in EOP for inoperable/stuck control rod 3.9 20 Bank 001 2008 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) FuelHandling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak / 3 X
Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak:
(CFR: 41.3 / 43.5 / 45.13)
AA2.04 Comparison of RCS fluid inputs and outputs, to detect leaks 3.4 21 Mod 1217 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 X
Knowledge of the interrelations between the Accidental Liquid Radwaste Release and the following: (CFR 41.7 / 45.7)
AK2.01 Radioactive-liquid monitors 2.7 22 Bank 951 2013 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 X
Knowledge of the interrelations between the Area Radiation Monitoring (ARM) System Alarms and the following: (CFR 41.7 /
45.7)
AK2.01 Detectors at each ARM system location 2.5*
23 New 1175 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
4 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1
ES-401 5
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)
(BW A02 & A03) Loss of NNIX/Y/7 X
Knowledge of the operational implications of the following concepts as they apply to the (Loss of NNI-X) (CFR: 41.8 /
41.10 / 45.3)
AK1.2 Normal, abnormal and emergency operating procedures associated with (Loss of NNI-X).
3.4 24 New 1212 (BW A04) Turbine Trip / 4 X
2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (CFR:
41.10 / 43.5 / 45.13) 3.8 25 New 1176 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 X
Knowledge of the reasons for the following responses as they apply to the (Flooding) (CFR: 41.5 /
41.10, 45.6, 45.13)
AK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations.
3.6 26 Mod 1211 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures X
Knowledge of the operational implications of the following concepts as they apply to the (EOP Rules) (CFR: 41.8 / 41.10 /
45.3)
EK1.3 Annunciators and conditions indicating signals, and remedial actions associated with the (EOP Rules).
3.0 27 New 1208 (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:
2 2
2 1
1 1
Group Point Total:
9
ES-401 6
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 003 (SF4P RCP) Reactor Coolant Pump X
Knowledge of the effect of a loss or malfunction on the following will have on the RCPS: (CFR: 41.7 / 45/5)
K6.04 Containment isolation valves affecting RCP operation 2.8 28 Bank 53 2011 004 (SF1; SF2 CVCS) Chemical and Volume Control X
Ability to manually operate and/or monitor in the control room: (CFR: 41/7 / 45.5 to 45.8)
A4.05 Letdown pressure and temperature control valves.
3.6 29 Bank 699 2008 004 (SF1; SF2 CVCS) Chemical and Volume Control X
Knowledge of the operational implications of the following concepts as they apply to the CVCS: (CFR: 41.5/45.7)
K5.40 Response of PRT during bubble formation in PZR: increase in Quench Tank pressure when cycling PORV shows that complete steam bubble does not exist, that significant noncondensable gas is still present.
3.0 30 Bank 963 2013 005 (SF4P RHR) Residual Heat Removal X
Knowledge of RHRS design feature(s) and/or interlock(s) which provide or the following:
(CFR: 41.7)
K4.08 Lineup for "piggy-back" mode with high-pressure injection 3.1*
31 New 1178 005 (SF4P RHR) Residual Heat Removal X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: (CFR: 41.5 / 45.5)
A1.01 Heatup/cooldown rates 3.5 32 Mod 91 006 (SF2; SF3 ECCS) Emergency Core Cooling X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: (CFR: 41.5 / 45.5)
A1.13 Accumulator pressure (level, boron concentration) 3.5 33 New 1228 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X
Ability to (a) predict the impacts of the following 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: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
A2.01 Stuck-open PORV or code safety 3.9 34 Bank 463 2002 008 (SF8 CCW) Component Cooling Water X
Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
A2.08 Effects of shutting (automatically or otherwise) the isolation valves of the letdown cooler 2.5 35 New 1227
ES-401 7
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO) 008 (SF8 CCW) Component Cooling Water X
Knowledge of bus power supplies to the following: (CFR: 41.7)
K2.02 CCW pump, including emergency backup 3.0*
36 Repe at 562 2017 010 (SF3 PZR PCS) Pressurizer Pressure Control X
Ability to monitor automatic operation of the PZR PCS, including: (CFR: 41.7 / 45.5)
A3.02 PZR pressure 3.6 37 New 1213 012 (SF7 RPS) Reactor Protection X
Knowledge of the physical connections and/or cause-effect relationships between the RPS and the following systems: (CFR: 41.2 to 41.9
/ 45.7 to 45.8)
K1.07 SDS (Steam Dump System) 2.8*
38 New 1230 012 (SF7 RPS) Reactor Protection X
2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7
/ 43.2) 3.2 39 New 1231 013 (SF2 ESFAS) Engineered Safety Features Actuation X
Knowledge of bus power supplies to the following: (CFR: 41.7)
K2.01 ESFAS/safeguards equipment control 3.6*
40 Bank 192 2013 022 (SF5 CCS) Containment Cooling X
Knowledge of the physical connections and/or cause-effect relationships between the CCS and the following systems: (CFR: 41.2 to 41.9
/ 45.7 to 45.8)
K1.01 SWS/cooling system 3.5 41 Bank 274 2013 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray X
Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following:
(CFR: 41.7)
K4.01 Source of water for CSS, including recirculation phase after LOCA 4.2 42 New 1221 026 (SF5 CSS) Containment Spray X
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
A4.01 CSS controls.
4.5 54 New 1241 039 (SF4S MSS) Main and Reheat Steam X
Knowledge of the operational implications of the following concepts as the apply to the MRSS: (CFR: 41.5 / 45.7)
K5.08 Effect of steam removal on reactivity 3.6 43 Repe at 1155 2017
ES-401 8
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO) 059 (SF4S MFW) Main Feedwater X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW controls including: (CFR: 41.5 / 45.5)
A1.03 Power level restrictions for operation of MFW pumps and valves 2.7*
44 Bank 537 2005 061 (SF4S AFW)
Auxiliary/Emergency Feedwater X
Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: (CFR: 41.7 / 45.7)
K6.01 Controllers and positioners 2.5 45 New 1232 062 (SF6 ED AC) AC Electrical Distribution X
2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2
/ 45.2) 4.0 46 New 1229 063 (SF6 ED DC) DC Electrical Distribution X
Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 /
45.3 / 45.13)
A2.01 Grounds 2.5 47 Bank 384 2008 064 (SF6 EDG) Emergency Diesel Generator X
Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: (CFR: 41.7 / 45.7)
K6.08 Fuel oil storage tanks 3.2 48 Bank 849 2011 064 (SF6 EDG) Emergency Diesel Generator X
Knowledge of ED/G system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
K4.01 Trips while loading the ED/G (frequency, voltage, speed) 3.8 49 New 1235 073 (SF7 PRM) Process Radiation Monitoring X
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
A4.03 Check source for operability demonstration 3.1 50 New 1226 073 (SF7 PRM) Process Radiation Monitoring X
Knowledge of the effect that a loss or malfunction of the PRM system will have on the following: (CFR: 41.7 / 45.6)
K3.01 Radioactive effluent releases 3.6 51 Bank 271 2005 076 (SF4S SW) Service Water X
Knowledge of the effect that a loss or malfunction of the SWS will have on the following: (CFR: 41.7 / 45.6)
K3.05 RHR components, controls, sensors, indicators, and alarms, including rad monitors 3.0*
52 New 1234 078 (SF8 IAS) Instrument Air X
Ability to monitor automatic operation of the IAS, including: (CFR: 41.7 / 45.5)
A3.01 Air pressure 3.1 53 Bank 673 2007
ES-401 9
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO) 078 (SF8 IAS) Instrument Air Replaced by 026 Containment Spray A4.01 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
A4.01 Pressure gauges (54) 103 (SF5 CNT) Containment X
Knowledge of the physical connections and/or cause-effect relationships between the containment system and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.05 Personnel access hatch and emergency access hatch 2.8*
55 Bank 158 2007 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:
3 2
2 3
2 3
3 3
2 3
2 Group Point Total:
28
ES-401 10 Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 001 (SF1 CRDS) Control Rod Drive X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CRDS controls including: (CFR: 41.5/45.5)
A1.06 Reactor power.
4.1 56 New 1179 002 (SF2; SF4P RCS) Reactor Coolant X
Knowledge of RCS design feature(s) and/or interlock(s) which provide for the following:
(CFR: 41.7)
K4.10 Overpressure protection.
4.2 60 New 1219 011 (SF2 PZR LCS) Pressurizer Level Control X
Knowledge of the operational implications of the following concepts as they apply to the PZR LCS: (CFR: 41.5 / 45.7)
K5.06 Indicated charging flow: seal flow plus actual charging flow.
2.9 57 Mod 1180 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation X
Ability to monitor automatic operation of the NNIS, including: (CFR: 41.7 / 45.5)
A3.02 Relationship between meter readings and actual parameter value 2.9*
58 Mod 1218 017 (SF7 ITM) InCore Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control X
Knowledge of the effect that a loss or malfunction of the HRPS will have on the following: (CFR: 41.7 / 45.6)
K3.01 Hydrogen concentration in containment 3.3 59 Bank 210 2002 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) FuelHandling Equipment Rejected system -see system 002 above 035 (SF 4P SG) Steam Generator X
Ability to (a) predict the impacts of the following malfunctions or operations on the SG; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
(CFR: 41.5 / 43.5 / 45.3 / 45.5)
A2.04 Steam flow/feed mismatch 3.6 61 New 1181 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X
Knowledge of the effect of a loss or malfunction on the following will have on the SDS: (CFR: 41.7 / 45.7)
K6.03 Controller and positioners, including ICS, S/G, CRDS 2.7 62 New 1182 045 (SF 4S MTG) Main Turbine Generator X
2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.
(CFR: 41.10 / 43.5 / 45.12) 3.9 63 Mod 1237 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate X
Knowledge of the physical connections and/or cause-effect relationships between the Condensate System and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.03 MFW 2.6*
64 Bank 261 2002
ES-401 11 Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO/SRO) 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal X
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.14 WGDS status alarms 2.8 65 New 1220 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:
1 0
1 1
1 1
1 1
1 1
1 Group Point Total:
10
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Rev. 11 Facility: ANO Unit 1 Date of Exam: 3/7/2018 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 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. (CFR: 41.10 /
43.2) 3.3 66 Mod 1184 2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. (CFR: 41.10 / 45.1 /
45.12) 4.1 67 Mod 1185 2.1.39 Knowledge of the station's requirements for verbal communications when implementing procedures. (CFR:
41.10 / 45.13) 3.7 68 Bank 991 2013 Subtotal
- 2. Equipment Control 2.2.14 Knowledge of the process for controlling equipment configuration or status. (CFR: 41.10 / 43.3 / 45.13) 3.9 69 New 1186 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 /
45.13) 3.7 70 Bank 118 2005 2.2.35 Ability to determine Technical Specification Mode of Operation. (CFR: 41.7 / 41.10 / 43.2 / 45.13) 3.6 71 New 1239 Subtotal
- 3. Radiation Control 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12 / 43.4 / 45.10) 3.4 72 Repeat 1144 2017 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
(CFR: 41.12 / 43.4 / 45.9) 2.9 73 Bank 995 2013 Subtotal
- 4. Emergency Procedures/Plan 2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. (CFR: 41.10 / 43.5 / 45.13) 3.5 74 Repeat 1143 2017 2.4.32 Knowledge of operator response to loss of all annunciators. (CFR: 41.10 / 43.5 / 45.13) 3.3 75 Bank 393 2001 Subtotal Tier 3 Point Total 10
ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /
Group (Original)
Randomly Selected K/A (New)
Reason for Rejection RO T1/G1 008 AA2.29 (3.9)
(Q#1) 008 AA2.27 (2.9) 008 - Pressurizer Vapor Space Accident. Original K/A concerned effect of a bubble in the reactor vessel. The effects of a bubble in the reactor vessel upper head would be pressurizer level making a sudden rise when pressurizer spray was used. Pressurizer level increases while RCS pressure decreases during a vapor space accident so it would be difficult to differentiate between the effects of the vapor space leak with that of the head bubble (pressurizer level would be constantly rising) and thus realistic conditions could not be established for a question based on this K/A. Replaced with K/A AA2.27 concerning the effects on indicated pressurizer level due to sensing line leakage.
RO T1/G1 022 AK3.03 (3.1)
(Q#5) 022 AK3.02 (3.5) 022 - Loss of Reactor Coolant Makeup. Original K/A concerned excess letdown which is Westinghouse specific equipment. Replaced with AK3.02 which is knowledge of actions in EOP/SOPs.
RO T1/G1 026 AK3.03 (4.0)
(Q#7) 026 AK3.01 (3.2) 026-Loss of Component Cooling Water. Original K/A concerned actions in EOP for Loss of CCW. Original question developed for this K/A conflicted with a simulator JPM. Following a review of other actions in the Loss of ICW (CCW) AOP, it was determined another question could not be developed for this AOP. Replaced with AK3.01 which concerns automatic closing of the SW isolation valves for the ICW coolers.
RO T1/G1 029 EK1.03 (3.6)
(Q#9) 029 EK2.06 (2.9) 029 - Anticipated Transient Without Scram. Original K/A concerned operational implications of the effects of boron on reactivity. A search of all available banks found one question for this K/A. This question was psychometrically flawed with two implausible distractors. Additionally this K/A could only result in a GFE question. In fact, of the five EK1 K/As for this system, all are GFE items (one with importance less than 2.5, so the available total is really four). Replaced with EK2.06 which concerns interrelationship between ATWS and breakers, relays, and disconnects.
RO T1/G1 057 AA2.03 (3.7)
(Q#15) 057 AA2.05 (3.5) 057 - Loss of Vital AC Instrument Bus. Original K/A concerned effects on RPS panel alarms, annunciators, and indicating lights. This conflicts with a simulator JPM involving RPS alarms and indicating lights.
Replaced with AA2.05 which involves S/G pressure and level instruments.
ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /
Group (Original)
Randomly Selected K/A (New)
Reason for Rejection RO T1/G1 065 2.2.39 (3.9)
(Q#17) 065 2.4.11 (4.0) 065 - Loss of Instrument Air. Original K/A concerned less than one hour Tech Specs. ANO-1 Tech Specs do not contain ANY specifications for Instrument Air. Replaced with 2.4.11which concerns knowledge of abnormal condition procedures.
RO T1/G2 001 AA1.03 (3.4)
(Q#19) 001 AA1.02 (3.6) 001 - Continuous Rod Withdrawal. Original K/A concerned use of boric acid pump control switch. ANO-1 procedure Control Rod Drive Malfunction Action (1203.003) section 9 - Continuous Rod Withdrawal, does not direct use of boric acid pumps. Replaced with K/A AA1.02 which is about use of rod control in-out handswitch.
RO T1/G2 BW A02 AK1.1 (3.2)
(Q#24)
BW A02 AK1.2 (3.4)
BW A02 - Loss of NNI-X. Original K/A concerned components, capacity, and function of emergency systems. NNI does not connect or input into any emergency systems. Replaced with K/A AK1.2 which concerns normal, abnormal, and emergency procedures related to a loss of NNI X.
RO T2/G1 004 A4.04 (3.2)
(Q#29) 004 A4.05 (3.6) 004 - Chemical and Volume Control System. Original K/A is about the ability to perform a calculation of boron concentration changes. This ability is best evaluated via a JPM. It would require too much time and too many references on the RO written exam. Replaced with K/A A4.05 which concerns Letdown pressure and valves.
RO T2/G1 004 K5.20 (3.6)
(Q#30) 004 K5.40 (3.0) 004 - Chemical and Volume Control System. Original K/A is about reactivity effects of xenon but this could only result in a question testing GFE knowledge. Replaced with K/A K5.40 which concerns PRT response when drawing a Pressurizer steam bubble.
RO T2/G1 006 A1.09 (2.8)
(Q#33) 006 A1.13 (3.5) 006 - Emergency Core Cooling System. Original K/A concerns operating the ECCS controls to prevent exceeding pump amperage. There are no indications of pump amperage for any pump in the control room.
Replaced with K/A A1.13 which is based on accumulator pressure, level, or boron concentration.
ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /
Group (Original)
Randomly Selected K/A (New)
Reason for Rejection RO T2/G1 078 A4.01 (3.1)
(Q#54) 026 A4.01 (4.5) 078 - Instrument Air. This K/A concerned pressure gauges. At NRC Chief Examiners request (due to oversampling of IA topics), this K/A and system was replaced by 026 - Containment Spray, K/A A4.01 which involves CSS controls.
RO T2/G2 001 A1.04 (3.7)
(Q#56) 001 A1.06 (4.1) 001 - Control Rod Drive. Original K/A concerned changes in PZR level and pressures to prevent exceeding Control Rod Drive (CRD) limits but there are no PZR level limits associated with CRD operation. Replaced with K/A A1.06 which concerns reactor power limits.
RO T2/G2 011 K5.04 (2.5)
(Q#57) 011 K5.06 (2.9) 011 -Pressurizer Level Control. Original K/A concerns the reasons for not allowing coolant to flash into steam in the letdown piping. A suitable question could not be developed for this K/A due to not being able to construct plausible distractors. Replaced with K/A K5.06 which concerns indicated charging flow.
RO T2/G2 034 K4.03 (2.6)
(Q#60) 002 K4.10 (4.2) 034 - Fuel Handling Equipment. Original K/A and system is an SRO level system. Operators at ANO-1 do not qualify as a fuel handler unless requested to join the refueling team. Only SROs receive training on fuel handling equipment as part of their licensing process. Replaced with system 002, Reactor Coolant, since it was not selected, and K/A K4.10 which concerns overpressure protection.
RO T2/G2 045 2.4.35 (3.8)
(Q#63) 045 2.1.25 (3.9) 045 - Main Turbine Generator. Original K/A concerned knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. The only local auxiliary operator task at ANO-1 during an emergency with the Main Turbine Generator is vent ing the generator of hydrogen during a blackout. This would not produce a discriminating RO level question. Replaced with K/A 2.1.25 which involves the interpretation of reference materials.
RO T2/G2 071 A4.11 (2.5)
(Q#65) 071 A4.14 (2.8) 071 - Waste Gas Disposal. Original K/A concerned manually operating or monitoring WGDS startup and shutdown from the control room. All WGDS operations are local only, none can be accomplished or monitored from the control room. Please note that of 40 WGDS A4 K/As only three apply to ANO-1 due to the local only actions. Replaced with K/A A4.14 which concerns monitoring WGDS status alarms from the control room.
ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /
Group (Original)
Randomly Selected K/A (New)
Reason for Rejection RO T3 2.1.39 (3.6)
(Q#68) 2.1.38 (3.7) 2.1.39 K/A: Knowledge of conservative decision making practices.
The Entergy procedure with guidance for conservative decision making (EN-OP-115) contains rather vague guidance in regards to conservative decision making. This did not result in development of a discriminating question. Replaced with K/A 2.1.38 which concerns requirements for verbal communications when implementing procedures.
RO T3 2.2.20 (2.6)
(Q#70) 2.2.12 (3.7) 2.2.20 K/A: Knowledge of the process for managing troubleshooting activities.
The above K/A is by nature an SRO Only K/A since the RO does not manage troubleshooting activities. The RO does not select which troubleshooting activities will be performed or when they will be performed, the RO simply responds to direction from SROs and requests from maintenance personnel performing the troubleshooting.
Replaced with 2.2.12 which concerns knowledge of surveillance procedures.
RO T3 2.2.42 (3.9)
(Q#71) 2.2.35 (3.6) 2.2.42 K/A: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
The above K/A could only result in a system specific question instead of a question testing a generic knowledge or ability.
Replaced with 2.2.35 which concerns ability to determine Tech Spec mode of operation.
RO T3 2.4.45 (4.1)
(Q#75) 2.4.32 (3.3) 2.4.45 K/A: Ability to prioritize and interpret the significance of each annunciator or alarm.
The above K/A could only result in a system specific question instead of a question testing a generic knowledge or ability.
Replaced with 2.4.32 which concerns operator knowledge of response to loss of all annunciators. This topic is similar to the rejected K/A and this condition is not addressed elsewhere in the K/A catalog.
ES-401 PWR Examination Outline Form ES-401-2 Rev. 11 Facility: ANO Unit 1 (Rev. 3)
Date of Exam: 3/7/2018 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total 1.
Emergency and Abnormal Plant Evolutions 1
N/A N/A 18 3
3 6
2 9
2 2
4 Tier Totals 27 5
5 10 2.
Plant Systems 1
28 2
3 5
2 10 1
1 1
3 Tier Totals 38 4
4 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it 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 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 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 These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO Only)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000007 (EPE 7; BW E02&E10; CE E02)
Reactor Trip, Stabilization, Recovery / 1 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 X
2.4.6 Knowledge of EOP mitigation strategies.
(CFR: 41.10 / 43.5 / 45.13) 4.7 76 New 1205 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 Rejected system - see 009 above 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient Without Scram / 1 000038 (EPE 38) Steam Generator Tube Rupture / 3 X
Ability to determine or interpret the following as they apply to a SGTR: (CFR 43.5 / 45.13)
EA2.17 RCP restart criteria 4.4 77 New 1189 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 000054 (APE 54; CE E06) Loss of Main Feedwater /4 000055 (EPE 55) Station Blackout / 6 X
2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR:
41.10 / 43.5 / 45.3 / 45.12) 4.3 78 Mod 1206 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 000058 (APE 58) Loss of DC Power / 6 X
2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5 / 45.3) 4.7 79 Bank 1003 2013 000062 (APE 62) Loss of Nuclear Service Water / 4 X
Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:
(CFR: 43.5 / 45.13)
AA2.03 The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition.
2.9 80 Bank 757 2009 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 (W E04) LOCA Outside Containment / 3 (W E11) Loss of Emergency Coolant Recirculation / 4
ES-401 3
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO Only)
(BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 X
Ability to determine and interpret the following as they apply to the (Inadequate Heat Transfer) (CFR:
43.5 / 45.13)
EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
4.4 81 Mod 1004 K/A Category Totals:
3 3
Group Point Total:
6
ES-401 4
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO Only)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 X
2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 /
43.5 / 45.2 / 45.6) 4.4 82 New 1190 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) FuelHandling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 X
Ability to determine and interpret the following as they apply to the Control Room Evacuation: (CFR:
43.5 / 45.13)
AA2.04 SG pressure.
4.0 83 Mod 1201 000069 (APE 69; W E14) Loss of Containment Integrity / 5 X
Ability to determine and interpret the following as they apply to the Loss of Containment Integrity:
(CFR: 43.5 / 45.13)
AA2.02 Verification of automatic and manual means of restoring integrity 4.4 84 New 1191 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
4 000076 (APE 76) High Reactor Coolant Activity / 9 X
2.1.20 Ability to interpret and execute procedure steps. (CFR:
41.10 / 43.5 / 45.12) 4.6 85 Mod 736 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNIX/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 CooldownDepressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures
ES-401 5
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO Only)
(CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:
2 2
Group Point Total:
4
ES-401 6
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (SRO Only)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 003 (SF4P RCP) Reactor Coolant Pump 004 (SF1; SF2 CVCS) Chemical and Volume Control X
2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 /41.7 /
43.2) 4.2 86 New 1204 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection 013 (SF2 ESFAS) Engineered Safety Features Actuation 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray X
2.4.46 Ability to verify that the alarms are consistent with the plant conditions. (CFR:
41.10 / 43.5 / 45.3 / 45.12) 4.2 87 Mod 1199 039 (SF4S MSS) Main and Reheat Steam 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW)
Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical Distribution 063 (SF6 ED DC) DC Electrical Distribution X
Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 /
45.3 / 45.13)
A2.01 Loss of ventilation during battery charging.
3.1*
88 Bank 649 2009 064 (SF6 EDG) Emergency Diesel Generator X
Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 /
45.3 / 45.13)
A2.13 Consequences of opening auxiliary feeder bus (ED/G sub supply) 2.8*
89 New 1200 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air
ES-401 7
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (SRO Only) 103 (SF5 CNT) Containment X
2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2
/ 45.2) 4.7 90 New 1202 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:
2 3
Group Point Total:
5
ES-401 8
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (SRO Only)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) InCore Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge Rejected system. See 034 below 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) FuelHandling Equipment X
Knowledge of design feature(s) and/or interlock(s) which provide for the following:
(CFR: 41.7 / 43.7)
K4.03 Overload protection 3.3 91 Bank 455 2002 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water X
2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 /
45.12 / 45.13) 4.7 92 New 1192 079 (SF8 SAS**) Station Air 086 Fire Protection X
Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 /
45.3 / 45.13)
A2.03 Inadvertent actuation of the FPS due to circuit failure or welding.
2.9 93 New 1193 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:
1 1
1 Group Point Total:
3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Rev. 11 Facility: ANO Unit 1 Date of Exam: 3/7/2018 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.40 Knowledge of refueling administrative requirements.
(CFR: 41.10 / 43.5 / 45.13) 3.9 94 New 1194 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.
(CFR: 41.10 / 43.5 / 45.12) 3.9 95 New 1207 Subtotal
- 2. Equipment Control 2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12) 4.6 96 New 1197 2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. (CFR: 41.10 / 43.3 /
45.13) 3.8 97 Bank 852 2011 Subtotal
- 3. Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10) 3.7 98 New 1242 Subtotal
- 4. Emergency Procedures/Plan 2.4.44 Knowledge of emergency plan protective action recommendations. (CFR: 41.10 / 41.12 / 43.5 / 45.11) 4.4 99 New 1198 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation. (CFR: 41.10 / 43.5 / 45.11) 4.5 100 Bank 998 2013 Subtotal Tier 3 Point Total 7
ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /
Group (Original)
Randomly Selected K/A (New)
Reason for Rejection SRO T1/G1 022 2.4.50 (4.0)
(Q#76) 011 2.4.6 (4.7) 022 - Loss of Reactor Coolant. Original K/A concerns the ability to verify system alarm setpoints and to operate controls in the alarm response manual. This K/A could only result in an RO level question since to recall setpoints and operation of controls is at the RO level. Additionally the Loss of Reactor Makeup AOP (1203.026) has two sections: loss of HPI pump, and Large MUP Leak. The actions for both sections are very similar, i.e. stop HPI pump, isolate letdown, etc. This topic was used on the last exam and another SRO level question could not be developed without it being a duplicate of that last exams question. Replaced with system 011, Small Break LOCA, and K/A 2.4.6 which concerns knowledge of EOP mitigation strategies.
SRO T1/G1 058 2.4.4 (4.7)
(Q#79) 058 2.2.40 (4.7) 058 - Loss of DC Power. Original K/A concerns indications for system parameters that are entry-level conditions for EOPs and AOPs. Entry conditions for EOPs and AOPs are RO level so a question for this K/A could only be at the RO level. Replaced with K/A 2.2.40 which concerns application of Technical Specifications.
SRO T1/G2 068 AA2.10 (4.4)
(Q#83) 068 AA2.04 (4.0) 068 - Control Room Evacuation. Original K/A concerns ability to interpret source range count rate as it applies to Control Room Evacuation.
ANO-1 procedure 1203.002, Alternate Shutdown, did not contain any steps specific to monitoring source range indication so a question could not be developed for this K/A. A condition report (CR-ANO-1-2017-02151) was immediately initiated since it was determined that this was a significant procedure flaw. Replaced with K/A AA2.04 which concerns ability to determine SG pressure.
SRO T2/G1 004 2.4.6 (4.7)
(Q#86) 004 2.2.25 (4.2) 004 - Chemical and Volume Control. Original K/A concerned knowledge of EOP mitigation strategies but no strategy could be found that would not also be common knowledge to ROs, without being a question based on minutiae. Replaced with K/A 2.2.25 concerning bases for Tech Spec LCOs.
SRO T2/G1 063 A2.01 (3.2)
(Q#88) 063 AA2.02 (3.1) 063 - DC Electrical Distribution. Original K/A concerns ability to interpret impacts of DC grounds and use procedures to mitigate the consequences of the grounds. This K/A duplicates the same system and K/A selected for the RO exam. Replaced with K/A A2.02which concerns loss of ventilation to the battery chargers.
SRO T2/G2 081 A2.04 (3.9)
(Q#93) 081 A2.03 (2.9) 061 - Fire Protection. Original K/A concerns failure to actuate the FPS when required, resulting in fire damage. This K/A is extremely difficult to develop a question for since the basic premise of the K/A involves an error by an operator and an SRO question would involve the fire procedure which would result in a direct lookup answer since the mitigating actions would have to be for a specific area. Replaced with K/A A2.03 which concerns spurious actuations due to welding or circuit failure.
ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /
Group (Original)
Randomly Selected K/A (New)
Reason for Rejection SRO T2/G2 029 A2.02 (2.9)
(Q#91) 034 K4.03 (3.3) 029 - Containment Purge. Original K/A concerns continuance of outdoor temperature inversion. ANO-1 procedure 1104.033, Reactor Building Ventilation, contains no actions for an outdoor temperature inversion.
Upon further consideration a decision was made to reject this system altogether due to low probability of development of an SRO level question for this system. Replaced with system 034 - Fuel Handling and K/A K4.03 which had been randomly selected for the RO outline but was rejected there due to being SRO level.
SRO T3 2.1.45 (4.3)
(Q#95) 2.1.5 (3.9)
Original K/A concerns identification and interpretation of diverse indications to validate the response of another indication. This K/A could only result in a system type question at the RO level. Replaced with K/A 2.1.5 which concerns procedures related to shift staffing or overtime limitations.
SRO T3 2.2.38 (4.5)
(Q#97) 2.2.17 (3.8)
Original K/A concerns conditions and limitations in the facility license.
Due to a lack of Tier 3 SRO level questions, a new question was developed but this question had a 100% miss rate despite several modifications. A TEAR will be generated following the exam to address this knowledge deficiency. Replaced with K/A 2.2.17 which concerns managing maintenance activities during power operations.
SRO T3 2.3.15 (3.1)
(Q#98) 2.3.4 (3.7)
Original K/A duplicated the Tier 3 section 3 K/A in the RO outline.
Replaced with K/A 2.3.4 which concerns normal and emergency radiation exposure limits.
SRO T3 2.4.11 (4.2)
(Q#99) 2.4.44 (4.4)
Original K/A concerned knowledge of abnormal condition procedures which could only result in a system level question. Replaced with 2.4.44 which concerns emergency plan protective action recommendations.
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
ANO Unit 1 Date of Examination:
2/26/2018 Examination Level: RO SRO Operating Test Number:
1 Administrative Topic (see Note)
Type Code*
Describe activity to be performed A1 Conduct of Operations K/A - 2.1.19, Importance Rating 3.9 M,S Calculate a Makeup No Concentration Change addition to the RCS, A1JPM-RO-PMS1 A2 Conduct of Operations K/A - 2.1.20, Importance Rating 4.6 P,R Perform Spent Fuel Pool Makeup Calculation A1JPM-RO-SFPMU A3 Equipment Control K/A - 2.2.13, Importance Rating 4.1 N,R Determine Tagging Boundaries for maintenance on P-10B, Domestic Water Pump Discharge Check Valve DW-6B A1JPM-RO-HCRD7 A4 Radiation Control K/A - 2.3.7, Importance Rating 3.5 M,R Calculate Dose received for described power entry.
A1JPM-RO-RAD3 Emergency Plan NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1, randomly selected)
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
ANO Unit 1 Date of Examination:
2/26/2018 Examination Level: RO SRO Operating Test Number:
1 Administrative Topic (see Note)
Type Code*
Describe activity to be performed A5 Conduct of Operations K/A - 2.1.25, Importance Rating 4.2 N,R Ability to Interpret Rod Insertion Limits per COLR for OPERABILITY A1JPM-SRO-ADMINRIL A6 Conduct of Operations K/A - 2.1.18, Importance Rating 3.8 N,R Determine Immediate Notification Requirement for a given plant condition.
A1JPM-SRO-ADMINREPORT A7 Equipment Control K/A - 2.2.37, Importance Rating 4.6 N,R Determine OPERABILITY of MSSV and apply Technical Specifications.
A1JPM-SRO-ADMINMSSV A8 Radiation Control K/A - 2.3.14, Importance Rating 3.8 P,R Determine which individuals are eligible to make the described power entry.
A1JPM-SRO-RAD2 A9 Emergency Plan K/A - 2.4.41, Importance Rating 4.6 D,R Emergency Action Level Classification A1JPM-SRO-EAL16 NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
ES-301 Administrative Topics Outline Form ES-301-1
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1, randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
ANO Unit 1 Date of Examination:
2/26/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:
1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function S1. Transfer Group 6 Rods to Auxiliary Power Supply 001 A4.07 (RO 3.3)
RO D/S 1
S2. Perform actions required for an ESAS actuation (Step 18) 006 A4.05 (RO 3.9 / SRO 3.8)
S3. Verify Proper ESAS Actuation 013 A4.01 (RO 4.5 / SRO 4.8)
S4. Shutdown P-32A at power with Reverse Rotation occurring 003 A2.02 (RO 3.7)
RO D/A/P/S 4A S5. Place Hydrogen Recombiner in service 028 A4.01 (RO 4.0)
RO D/L/S 5
S6. Synchronize and load #1 EDG with a failure of the load switch 064 A2.05 (RO 3.1)
RO D/A/S 6
S7. Reset A RPS Channel after a high RB Pressure signal 012 A4.04 (RO 3.3)
RO N/S/EN 7
S8. Loss of two ICW Pumps 008 A2.01 (RO 3.3 / SRO 3.6)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1. Take manual control of P-7A EFW Pump at turbine 061 A2.05 (RO 3.1)
RO D/E/L/R 4B P2. No DC Start of #2 EDG 064 A4.01 (RO 4.0 / SRO 4.3)
P3. Align T-16A (Treated Waste Monitoring Tank) for Recirc/Sample 068 A2.02 (RO 2.7 / SRO 2.8)
All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6 (5) / 2-3 (3) 9 (9) / 4 (4) 1 (2) / 1 (1) 1 (3) / 1 (2) (control room system) 1 (4)/ 1 (2) 2 (2) / 1 (1) 3 (3) / 2 (2) (randomly selected) 1 (2) / 1 (1)
Appendix D Scenario 1 Outline Page 1 of 45 Facility: ANO-1 Scenario No.: 1 R1 Op-Test No.: 2018-1 Examiners: _____DEGRADED POWER _____ Operators:
__IC-219_____________________
Initial Conditions:
- 99.5% Power.
- EFIC Failed and will not auto-actuate
- C-28A IA Compressor is out of service for overhaul
- CV-3807 SW to #2 EDG will not automatically open
- ULD will not respond to demand changes
- #1 EDG will not Auto-Start (IMF DG175)
- P-36C MU Pump in service
- Ensure Thundersound running Turnover:
- Day shift - normal working day.
- ICS in manual per 1105.004 Step 7.0 for NI Calibration
- Thunderstorms entering the area
- NI Calibration completed, return ICS to automatic per OP-1105.004 Section 8.0.
Event No.
Malf. No.
Event Type*
Event Description 1
N/A N-(ATC)
Return ICS to automatic per OP-1105.004 Section 8.0 2
DI_1413M I-(BOP)
(SRO)
TS BWST temperature controller fails, causing heaters to remain energized and BWST temperature to exceed 110oF.
3 CV063 C-(BOP)
C-(SRO)
Failure of the running Makeup Pump (P-36C) 4 Lightning DI_DG2S IMF K01A3 CV3807 to 0
C-(BOP)
(SRO)
TS Lightning strike causes #2 EDG to start with a failure of the Service Water Cooling Valve for the EDG (CV-3807).
5 N/A R-(ATC)
R(SRO)
Dispatcher directs a power reduction to 700 MW in the next 10 minutes.
ULDMAN OVR to True ULD will not respond to demand changes.
6 Lightning ED451 I-(BOP)
I-(SRO)
Second lightning strike results in the loss of the NNI-Y power supply.
Appendix D Scenario 1 Outline Page 2 of 45 Event No.
Malf. No.
Event Type*
Event Description 7
LOOP ED183 M
(ALL)
Loss of offsite power will occur resulting in an automatic reactor trip.
8 DG175 C-(BOP)
CT The #1 EDG will not auto start requiring the operators to manually start the EDG.
CT EFIC fails to Auto-actuate.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario 1 Outline Page 3 of 45 SCENARIO 1 OBJECTIVES
- 1)
Evaluate individual ability to perform routine ICS Control Station Operations.
- 2)
Evaluate individual ability to recognize when conditions require the entry into technical specifications conditions.
- 3)
Evaluate individual ability to recognize and respond to the loss of a makeup pump.
- 4)
Evaluate individual ability to recognize and respond to the false auto start of emergency diesel generator.
- 5)
Evaluate individual ability to recognize and respond to the failure to open of the service water supply valve on an operating emergency diesel generator.
- 6)
Evaluate individual ability to reduce plant power.
- 7)
Evaluate individual ability to recognize and respond to ICS system malfunctions.
- 8)
Evaluate individual ability to recognize and respond to the loss of power to the Y Non-Nuclear Instrumentation system.
- 9)
Evaluate individual ability to recognize and respond to a loss of offsite power.
- 10) Evaluate individual ability to recognize and respond to failure of the emergency diesel generator to auto start.
- 11) Evaluate individual ability to recognize and respond to the failure of the Emergency Feedwater Initiation and Control system to automatically actuate emergency feedwater.
- 12) Evaluate individual ability to perform the energizing of a vital bus from the Alternate AC generator. (RSOA - Risk Significant Operator Action)
Appendix D Scenario 1 Outline Page 4 of 45 Procedures and alarms Event One: Normal operation to return ICS to automatic OP-1105.004 When finished, K07-A1, UNIT MATER IN TRACK will clear Event Two: BWST Temperature exceeds TS Limit 1203.012H, K09 Annunciator Corrective Actions (ACA)
K09-C6 BWST TEMP HI/LO.
T.S. 3.5.4 Condition C.
Event Three: Failure of the running Makeup Pump (P-36C) 1203.026, Loss of Reactor Coolant Makeup RT-13, Restore Letdown K02-E8, PROGRAMMABLE ALARM No. 2 K08-A7, RCP SEAL INJ FLOW LO K10-A6, HPI PUMP TRIP K10-D8, HPI PUMP P36C OIL PRESSURE LO Event Four: Lightning strike causes #2 EDG to start with a failure of the Service Water Cooling Valve for the EDG.
1203.012A Annunciator K01 Corrective Actions.
K01-A3, EDG2 AUTO START COMMAND K02-C4, H1 NEG SEQ OVERVOLTAGE K02-C5, H2 NEG SEQ OVERVOLTAGE K02-E6, A1 NEG SEQ OVERVOLATGE K02-E7, A2 NEG SEQ OVERVOLATGE K04-A5, TURB ROTOR VIBRATION HI K01-C4, EDG2 CRITICAL TROUBLE K01-D3, EDG2 NOT AVAILABLE TS 3.8.1. Condition B. (CRS-TS)
Event Five: Dispatcher directs power reduction due to grid instability 1203.045, Rapid Plant Shutdown K02-E8, PROGRAMMABLE ALARM No.1 Event Six: ULD fails to respond to demanded power reduction K07-A1, UNIT MASTER IN TRACK Event Seven: Loss of NNI-Y power due to lightning strike 1203.047, Loss of NNI Power.
1203.012F, K07 ACA K02-E8, PROGRAMMABLE ALARM No.1 K07-A4, ICS/AUX POWER SUPPLY TROUBLE K07-B4, SASS MISMATCH K08-D7, RCP SEAL CAVITY PRESS HI/LO K09-B5, NAOH TANK TEMP HI/LO K09-C6, BWST TEMP HI/LO K10-A5, CFT A PRESS HI/LO K10-B5, CFT B PRESS HI/LO
Appendix D Scenario 1 Outline Page 5 of 45 Event Eight: Loss of offsite power 1202.001 Rx Trip. After the immediate actions are complete the CRS will transition to either 1202.008 Blackout 1202.007 Degraded Power Event Nine: Failure of #1 EDG to auto start K10-D2, EDG 1 NON-CRITICAL TROUBLE K01-A5, RS1 INVERTER TROUBLE K01-B5, RS3 INVERTER TROUBLE Event Ten: EFIC fails to actuate automatically K12-C7, EFIC SYSTEM TROUBLE Repetitive Tasks / Primary Performer RT-5 Verify Proper EFW Actuation and Control / ATC RT-6 Verify Proper MSLI and EFW Actuation and Control / ATC RT-13 Restore Letdown / BOP RT-21 Check EDG Operation / BOP
Appendix D Scenario 1 Outline Page 6 of 45 SCENARIO 1 NARRATIVE Event One: Normal operation to return ICS to automatic The crew will assume plant responsibility at 99.5% power. The SRO will direct returning ICS to automatic per OP-1105.004, Integrated Control System, Section 8.0, Transferring Major ICS Control Stations to AUTO. (ATC-N)
Event Two: BWST Temperature exceeds TS Limit BWST temperature controller fails, causing heaters to remain energized and BWST temperature to exceed 110oF. The CRS will enter the ACA 1203.012H for annunciator K09-C6 BWST TEMP HI/LO. He will direct the BOP to place the heater handswitch into OFF to secure the heaters and enter T.S. 3.5.4 Condition C. (CRS-TS) (BOP-I)
Event Three: Failure of the running Makeup Pump (P-36C)
Once the Technical Specification entries are completed, the running Makeup Pump (P-36C) will trip due to an instantaneous overcurrent condition as reported from the field operator upon request to investigate. The CRS will enter AOP 1203.026, Loss of Reactor Coolant Makeup. (BOP-C) (CRS-C)
Event Four: Lightning strike causes #2 EDG to start with a failure of the Service Water Cooling Valve for the EDG.
Once the Standby Makeup Pump is running with normal Seal Injection and Makeup, and Letdown is restored, a lightning strike will cause the #2 EDG to have an erroneous automatic start actuation. In addition to the automatic start, the cooling water valve to the EDG (CV-3807) will fail in the closed position. The crew should attempt to open the valve from the control room and send an operator to locally open the valve. Both attempts will fail.
The CRS should direct operations per 1203.012A Annunciator K01 Corrective Actions. The crew should determine it is a spurious actuation. With no cooling flow available, the crew should take the #2 EDG to lockout to prevent an automatic trip of the EDG.
(BOP-C) (SRO-C)
The CRS will enter TS 3.8.1. Condition B. (CRS-TS)
Event Five: Dispatcher directs power reduction due to grid instability The dispatcher will call and direct U1 to reduce net generator output to 700MWe in the next 10 minutes (SRO-R) (ATC-R).
Event Six: ULD fails to respond to demanded power reduction The Unit Load Demand in ICS will not respond to the demanded change in power, the ATC will manually control the power reduction using the SG/RX Master and will reduce power at a rate of > 2%/minute to achieve the direction of the dispatcher.
Appendix D Scenario 1 Outline Page 7 of 45 Event Seven: Loss of NNI-Y power due to lightning strike A second lightning strike will result in a loss of the NNI-Y power supply (ATC-I) (BOP-I)
(SRO-I). The crew should recognize the loss of NNI-Y power. The CRS will enter 1203.047, Loss of NNI Power. The power supply breakers on RS-4 BKR 9 and Y01 BKR 39 will not reset. Breakers S-1 and S-2 on the NNI-Y power supply will not be tripped. Both board operators will have actions to place all SASS (Smart Automatic Signal Selection) switches into the NNI-X position.
Letdown flow and pressure indication will be lost on C04. The letdown orifice bypass valve will fail to 50% reducing letdown flow. CFT pressure instrumentation will be lost along with NaOH tank temperature. All NNI-Y inputs to PMS/PDS will be lost or fail to mid scale.
Event Eight: Loss of offsite power A loss of offsite power will occur due to storm related grid instabilities (SRO-M) (ATC-M)
(BOP-M). The reactor will trip automatically. The CRS will direct operations per 1202.001 Rx Trip. After the immediate actions are complete the CRS will transition to either 1202.008 Blackout or 1202.007 Degraded Power depending on when the BOP manually starts the #1 EDG.
Event Nine: Failure of #1 EDG to auto start The #1 EDG will not auto start requiring the BOP to manually start the EDG (BOP-C)
(BOP-CT).
CT Justification:
Safety significance - Degraded emergency power capacity Initiating Cue - Loss of Offsite Power Measurable Performance Standard - Energized A3 within 15 minutes of LOOP. (Criteria to declare an SAE for a station blackout)
Performance Feedback - Status indicating lights, voltage, frequency and Kw loading for the diesel and A3 voltage indicated.
The Alternate AC (AAC) Generator initially will not come up to speed. Unit 2 operations will dispatch maintenance personnel to make repairs.
Event Ten: EFIC fails to actuate automatically The EFIC system is failed and will not automatically actuate EFW on the Loss of Offsite Power (ATC-C) (SRO-C). The ATC should manually actuate EFW from the remote switch matrix (ATC-CT).
CT Justification:
Safety significance - Failure of an automatic actuation of an ESF system Initiating Cue - Loss of all Reactor Coolant Pumps Measurable Performance Standard - Manually actuated EFW before the ERV opens in automatic. (Opening the ERV is an unnecessary challenge to the RCS fission product barrier)
Performance Feedback - Status indicating lights on the Remote Switch Matrix, EFW Pumps, and EFW valves.
The ATC will perform 1202.012 Repetitive Tasks RT-5 Verify proper EFW actuation and control. EFW may be manually actuated before the step in the EOP that directs verifying EFW actuated.
Appendix D Scenario 1 Outline Page 8 of 45 Critical Task Measurable Performance Standards (CT) The BOP energized A3 within 15 minutes of LOOP.
(CT) EFW manually actuated before the ERV opens in automatic.
The AAC Diesel Generator will be made available to the crew as directed by the lead examiner and the scenario can be terminated when the AAC Generator is powering the A4 vital bus.
Scenario Recapitulation Total Malfunctions:
8 Abnormal Events:
4 Major Transients:
1 EOPs Entered:
2 EOP Contingencies:
1 Malfunctions after EOP 2 Critical Tasks:
2 Causing an unnecessary plant trip or ESAS actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario 2 Outline Page 1 of 58 Facility: ANO-1 Scenario No.: 2 R2 Op-Test No.: 2018-1 Examiners: _____OVERHEATING _________ Operators: ___IC-229 ______________
Initial Conditions:
- 95% Power.
- RPS failed - will not cause a reactor trip
- MFW Crosstie valve will not automatically open
- C-5A in service
- P-33B ICW Pump OOS for maintenance
- P-3A, B, and D in service
- P-36C in service
- C-5B Vacuum pump will not automatically start Turnover:
- Day shift - normal working day.
- P-33B ICW Pump OOS for maintenance Event No.
Malf. No.
Event Type*
Event Description 1
N/A N-(BOP)
Swap Makeup Pumps, Place P-36B in service and remove P-36C from service.
2 CO_P3A to OFF C-(BOP)
P-3A, Circulating Water Pump, trips 3
TR049 to 0 in 10 I-(ATC)
(SRO)
TS LT-1001, Pressurizer Level Transmitter fails low.
C-(BOP)
C-(SRO)
Degraded Vacuum 5
N/A (SRO)
TS P-7B, Motor Driven EFW Pump, Inoperable due to oil leak.
6 FW074 FW059 C-(ATC)
P-1A, Main Feedwater Pump, trips with a failure of the crosstie valve (CV-2827) to open automatically.
7 CV1008 to 0.15 C-(ATC)
Pressurizer Spray Valve will not fully close following the MFWP trip pressure rise.
8 FW075 M-(ALL)
CT Loss of all feed water with a failure of RPS to automatically trip the reactor.
9 CV6601A C-(ATC)
CT P-7A, Turbine Drive EFW Pump, trips.
Transition to Overheating EOP (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (CT)
Critical Task
Appendix D Scenario 2 Outline Page 2 of 58 SCENARIO 2 OBJECTIVES
- 1)
Evaluate individual ability to perform Makeup Pump Operations.
- 2)
Evaluate individual ability to recognize and Respond to Circulating Water System Annunciators.
- 3)
Evaluate individual ability to recognize when conditions require the entry into Technical Specifications conditions.
- 4)
Evaluate individual ability to recognize and Respond to Pressurizer Level Indication Malfunction.
- 5)
Evaluate individual ability to recognize and Respond to Loss of Condenser Vacuum.
- 6)
Evaluate individual ability to reduce plant power.
- 7)
Evaluate individual ability to recognize when conditions require the entry into Technical Specifications conditions.
- 8)
Evaluate individual ability to recognize and respond to Loss of Steam Generator Feed.
- 9)
Evaluate individual ability to recognize and Respond to Pressurizer Spray Valve Failure.
- 10) Evaluate individual ability to recognize and Respond to Reactor Trip.
- 11) Evaluate individual ability to recognize and Respond to Overheating.
- 12) Evaluate individual ability to Perform EOP Repetitive Tasks.
Appendix D Scenario 2 Outline Page 3 of 58 Procedures and alarms Event One: Shift operating Makeup Pumps OP-1104.002 Event Two: P-3A Circulating Water Pump Trip 1203.012D, K05 Annunciator Corrective Actions (ACA)
K05-A1, CIRC WATER PUMP TRIP K05-B1, CW PUMP DISCH PRESS HI K05-B2, CONDENSER VACUUM LO Event Three: Pressurizer Level fails low 1203.012H, K09 Annunciator Corrective Actions (ACA)
K09-A3, PZR LEVEL LO LO K09-C3, PZR LEVEL LO T.S. 3.3.15 Condition A (PAM)
Event Four: Operating vacuum pump failure / Loss of Vacuum 1203.012D, K05 Annunciator Corrective Actions (ACA)
AOP 1203.016, Loss of Condenser Vacuum AOP 1203.045, Rapid Plant Shutdown K05-B2, CONDENSER VACUUM LO K05-A3, VACUUM PUMP TRIP K07-A1, UNIT MATER IN TRACK Event Five: P-7B (EFW Pump) oil leak OP-1107.001, Electrical System Operation T.S. 3.7.5.B Event Six: Main Feedwater Pump trip OP-1203.027 Loss of SG Feed OP-1203.045 Rapid Plant Shutdown.
K06-A5, MSR/T40 CRITICAL TROUBLE K06-A8, P8A/P8B FLOW LOW K06-C1, A MFP TURB CONTROL SYS TROUBLE K07-A1, UNIT MASTER IN TRACK K07-A7, A MFP TURBINE TRIP K07-B2, LOSS OF MFP RUNBACK IN EFFECT K07-C1, REACTOR IS FEEDWATER LIMITED K07-C2, FEEDWATER IS REACTOR LIMITED K07-C3, HI LOAD LIMIT IN EFFECT K07-E2, A OTSG BTU LIMIT K07-E7, A MFP FLOW LO K08-E5, AMSAC BYPASS K08-F3, TRIP ON LOSS OF TURB BYPASSED K09-E2, LOOP DTc >5oF K12-F1, K15 NON-CRITICAL TROUBLE K15-C1, REACTOR TRIP ON MAIN TURBINE TRIP BYPASS TROUBLE
Appendix D Scenario 2 Outline Page 4 of 58 Event Seven: Pressurizer Spray Valve failure AOP 1203.015, Pressurizer System Failure OR AOP 1203.027, Loss of Steam Generator Feed Event Eight: Second Main Feedwater Pump trips with a failure of RPS 1202.001, Reactor Trip K07-A8, B MFP TURBINE TRIP K08-A3, REACTOR TRIP K04-A3, TURBINE TRIP K12-A5, EFW ACTUATION SIGNAL Event Nine: Turbine Driven EFW Pump, P-7A trips 1202.004, Overheating K12-B6, P7B TRIP/AUTO START FAILURE K12-B5, P7A TURBINE TRIP Repetitive Tasks / Primary Performer RT-4 Initiate HPI Cooling / BOP RT-5 Verify Proper EFW Actuation and Control / ATC RT-14 Control RCS Pressure / ATC & BOP will both perform actions RT-16 Feeding Intact SG / BOP RT-19 Check Proper Electrical Response / BOP RT-20 Check NNI and ICS Power Available / ATC
Appendix D Scenario 2 Outline Page 5 of 58 SCENARIO 2 NARRATIVE Event One: Shift operating Makeup Pumps The crew will assume plant responsibility at 95% power. The SRO will direct placing P-36B Makeup Pump in service and securing P-36C per OP-1104.002.
(BOP-N)
Event Two: P-3A Circulating Water Pump Trip The crew will receive alarm K05-A1 for A Circulating Water Pump trip. Vacuum will degrade slightly and the ATC should maintain power less than 100%. Once the discharge valve for P-3A is fully closed the BOP will start P-3C Circulating Water Pump using OP-1104.008, Circulating Water and Water Box Vacuum System Operations.
(BOP-C)
Event Three: Pressurizer Level fails low The controlling pressurizer level transmitter will fail low causing annunciators on K09 and a loss of pressurizer heater due to a low level interlock. (ATC-I) (SRO-TS) The crew should utilize AOP 1203.015, Pressurizer System Failures to determine and select the good instrument and return the proportional pressurizer heater controls to automatic. The CRS should enter T.S. 3.3.15 Condition A (PAM) for the failed pressurizer level transmitter (LT-1001).
Event Four: Operating vacuum pump failure / Loss of Vacuum The operating condenser vacuum pump (C-5A) will have a broken sight glass which will cause a loss of sealing water and subsequently a loss of condenser vacuum. The standby pump (C-5B) auto-start feature is failed as an initial condition. The crew will recognize the lowering condenser vacuum trend and take action per AOP 1203.016, Loss of Condenser Vacuum. The AOP will require lowering plant power to stabilize vacuum. Once the ATC has lowered plant power at least 5%, the IAO will report the broken sight glass on C-5A. The BOP will be required to manually start C-5B and stop C-5A to regain condenser vacuum (ATC-R) (BOP-C) (SRO-C).
Event Five: P-7B (EFW Pump) oil leak While condenser vacuum is recovering, the WCO will call to report a large amount of oil on the floor below P-7B EFW Pump and no oil visible in the Motor Outboard Bearing bulls eye level indicator. The CRS should recognize that the given conditions result in making P-7B inoperable and enter T.S. 3.7.5.B for One EFW Train. (SRO-TS) The crew should either direct the field operators to rack down A-311 breaker for P-7B OR place P-7B in Pull-To-Lock to prevent an automatic start of the pump. If one or the other is not performed, then P-7B will trip when EFW is actuated later in the scenario.
Event Six: Main Feedwater Pump trip The A Main Feedwater Pump will trip with a failure of the cross-tie valve (CV2827) to automatically open. The ATC should recognize the failure and manually open CV-2827 in order to prevent a reactor trip. (ATC-C) (CRS-C) ICS will run back the plant automatically. The CRS will direct operations per OP-1203.027 Loss of SG Feed and OP-1203.045 Rapid Plant Shutdown.
Appendix D Scenario 2 Outline Page 6 of 58 Event Seven: Pressurizer Spray Valve failure Due to the loss of a MFWP, a pressure surge will cause the Pressurizer Spray Valve (CV-1008) to automatically open. When CV-1008 closes it will stick open ~15% and be indicating dual position, the unusually low RCS pressure along with additional heaters automatically running should alert the crew to the condition. (ATC-C) The ATC will isolate CV-1008 by closing CV-1009. The CRS will utilize guidance contained in either AOP 1203.015, Pressurizer System Failure, Section 6 or AOP 1203.027, Loss of Steam Generator Feed, Steps 8 and 9.
Event Eight: Second Main Feedwater Pump trips with a failure of RPS The second Main Feedwater Pump will trip, this will require a manual reactor trip by the ATC due to the RPS failure. (CT)
CT Justification:
Safety significance - Failure of Reactor Protection System (RPS)
Initiating Cue - Main Feedwater Pump Trip alarm Measurable Performance Standard - Manually trip the reactor within 1 minute of exceeding the RPS trip setpoint. This criterion is based on OP-1015.050, Time Critical Operator Actions Program, which describes the following: On failure of RPS trip and failure of Main Reactor Trip pushbuttons. Within 1 minute, trip CRD backup trip breakers (A-501 and B-631) from C03. While this is not exactly the same situation, one minute provides adequate time for the crew to recognize the condition and trip the plant as required.
Performance Feedback - Reactor tripped, all rods inserted, and associated alarms Event Nine: Turbine Driven EFW Pump, P-7A trips During Reactor Trip follow-up actions, P-7B will trip if the crew did not rack down the breaker or take the pump to Pull-To-Lock earlier in the scenario and then P-7A will trip resulting in a loss of all feedwater. The CRS will transition to the Overheating EOP.
The Auxiliary Feedwater Pump P-75 will be used for feeding the steam generators. The ATC restored FW flow to both Steam Generators using RT-16 (BOP) prior to SCM reaching 5 degrees. (CT)
CT Justification:
Safety significance - Failure of an ESF component Initiating Cue - EFW Pump P-7A trip Measurable Performance Standard - Restore FW prior to SCM reaching 5oF.
Performance Feedback - P-75 in service with FW flow indicated to both Steam Generators
Appendix D Scenario 2 Outline Page 7 of 58 Critical Task Measurable Performance Standards (CT) The ATC should trip the reactor within one minute of the second MFWP tripping.
(CT) FW should be restored to establish PSHT prior to SCM reaching 5oF.
The scenario can be terminated when P-75 is feeding the steam generators per RT-16.
Scenario Recapitulation Total Malfunctions:
7 Abnormal Events:
3 Major Transients:
1 EOPs Entered:
2 EOP Contingencies:
1 Malfunctions after EOP 1 Critical Tasks:
2 Causing an unnecessary plant trip or ESAS actuation may constitute a CT failure.
Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario 3 Outline Page 1 of 56 Facility: ANO-1 Scenario No.: 3 R1 Op-Test No.: 2018-1 Examiners: _____OVERCOOLING _________ Operators:
___IC-221____________________
Initial Conditions:
- 95% Power.
- P-40B in service
- P-36C in service
- P-4B aligned to A3
- B55/56 aligned to B6 Turnover:
- Day shift - normal working day.
- Y-41 in service (normal power source)
- Align B55/56 to B5
- VUC-9 secured due to high bearing vibrations Event No.
Malf. No.
Event Type*
Event Description 1
N/A N-(BOP)
Align B55/56 to B5.
2 N/A (SRO)
TS Inside AO reports low electrolyte level on D06 battery bank, Cell 30. T.S. 3.8.6 Condition C 3
TR449 to 620 in 30 I-(ATC)
I-(BOP)
Loop B (TT-1048, NNI-Y) Tcold instrument fails high.
4 CO_P4A C-(BOP)
C-(SRO)
TS P-4A Service Water Pump trips, momentary inoperability on Loop I Service Water. T.S. 3.7.7 Condition A 5
Several R-(ATC)
P-32A, Reactor Coolant Pump, high vibrations and bearing temperatures.
6 RX150 C-(BOP)
C-(ATC)
Turbine will stop responding during power reduction at ~85% power.
7 RC031 C-(ATC)
C-(BOP)
P-32A, RCP Sheared Shaft 8
MS131 to 0.3 in 180 M-(ALL)
Steam line break inside containment on A SG.
9 ES263 CV6203 to 1.00 I-(ATC)
C-(BOP)
CT ESAS Channel 5 fails to automatically actuate on high Reactor Building Pressure.
10 RB Isolation Valve for Chilled Water CV-6203 (Channel 6) fails to isolate.
11 CV2627 to 1.00 DI_HIC2645O to TRUE C-(ATC)
CT Vector fails to isolate EFW Flow through Flow Control Valve CV-2645 and Flow Isolation Valve CV-2627.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario 3 Outline Page 2 of 56 SCENARIO 3 OBJECTIVES
- 1)
Evaluate individual ability to perform MCC B55 & B56 operation.
- 2)
Evaluate individual ability to recognize when conditions require the entry into Technical Specifications conditions.
- 3)
Evaluate individual ability to perform SASS operation with failed input.
- 4)
Evaluate individual ability to respond to ICS Annunciators on K-07.
- 5)
Evaluate individual ability to respond to a loss of Service Water flow.
- 6)
Evaluate individual ability to recognize when conditions require the entry into Technical Specifications conditions.
- 7)
Evaluate individual ability to respond to RCP motor / bearing trouble.
- 8)
Evaluate individual ability to reduce plant power.
- 9)
Evaluate individual ability to respond to an RCP sheared shaft.
- 10) Evaluate individual ability to respond to Reactor Trip.
- 11) Evaluate individual ability to respond to Overcooling.
- 12) Evaluate individual ability to perform EOP Repetitive Tasks.
Appendix D Scenario 3 Outline Page 3 of 56 Procedures and alarms Event One: Align B55/56 to B5 OP-1107.002, ES ELECTRICAL SYSTEM OPERATION (BOP-N)
Event Two: D06 Battery Cell leaking T.S. 3.8.6 Condition C for the low electrolyte level.
Event Three: Loop B Tcold input to ICS fails high AOP 1203.001, ICS ABNORMAL OPERATION K02-D8, PLANT COMPUTER CRITICAL ALARM K07-A1, UNIT MASTER IN TRACK K07-B4, SASS MISMATCH K07-C1, REACTOR IS FEEDWATER LIMITED K07-C2, FEEDWATER IS REACTOR LIMITED K09-E2, LOOP dTc >5oF Event Four: P-4A Service Water Pump trips ACA 1203.012I, K10 Annunciator Corrective Action K10-A3, SERVICE WATER PUMP TRIP K11-B6, SW PUMP P4A ES FAILURE Event Five: P-32A RCP develops high vibrations and temperatures 1203.031, Reactor Coolant Pump and Motor Emergency K08-A8, RCP MOTOR BRG TEMP HI K08-B6, RCP VIBRATION HI K02-D8, PLANT COMPUTER CRITICAL ALARM Event Six: Turbine stops responding during power reduction 1203.045, Rapid Plant Shutdown K07-A1, UNIT MASTER IN TRACK Event Seven: P-32A RCP Sheared Shaft 1203.031, Reactor Coolant Pump and Motor Emergency K07-B4, SASS MISMATCH K07-C1, REACTOR IS FEEDWATER LIMITED K07-E3, B OTSG BTU LIMIT K09-D2, RCS FLOW LO K09-E2, LOOP dTc >5oF Event Eight: Steam Leak in containment 1202.003, Overcooling EOP. (ALL-M)
K12-F2, MAIN CHILLER TROUBLE K12-A7, MSLI OTSG A K12-A5, EFW ACTUATION SIGNAL K11-A2, HPI CHANNEL 1 K11-A3, HPI CHANNEL 3 K11-B2, LPI CHANNEL 2 K11-B3, LPI CHANNEL 4
Appendix D Scenario 3 Outline Page 4 of 56 Event Nine and Ten: Containment Isolation RT-10 K11-C2, REACTORBUILDING ISOLATION CHANNEL 5 K11-C3, REACTORBUILDING ISOLATION CHANNEL 6 Event Eleven: EFW Flow Control Valves / Vector Isolation RT-5, VERIFY PROPER EFW ACTUATION AND CONTROL Repetitive Tasks / Primary Performer RT-5 Verify Proper EFW Actuation and Control / ATC RT-6 Verify Proper MSLI and EFW Actuation and Control / ATC RT-10 Verify Proper ESAS Actuation / BOP RT-14 Control RCS Pressure / ATC and BOP both have actions RT-19 Check Proper Electrical Response / BOP RT-20 Check NNI and ICS Power Available / ATC
Appendix D Scenario 3 Outline Page 5 of 56 SCENARIO 3 NARRATIVE Event One: Align B55/56 to B5 The crew will assume plant responsibility at 95% power. The SRO will direct aligning B55/56 to B5 per OP-1107.002, in preparations for #2 EDG maintenance later this shift.
(BOP-N)
Event Two: D06 Battery Cell leaking A report from the Inside Auxiliary Operator (IAO) indicates that the electrolyte on D06 Cell 30 is below the top of the plates. There is no indication of leakage. CRS will enter T.S. 3.8.6 Condition C for the low electrolyte level. (CRS-TS)
Event Three: Loop B Tcold input to ICS fails high Loop B Tcold instrument (TT-1048) fails high requiring the ATC to take ICS stations to hand per AOP 1203.001 and stabilize the plant. After determining and selecting a good instrument the BOP will return ICS to automatic. (ATC-I) (BOP-I)
Event Four: P-4A Service Water Pump trips P-4A Service Water Pump will trip, resulting in a momentary loss of Loop I Service Water until the BOP manually starts P-4B Service Water Pump. CRS will enter T.S.
3.7.7 Condition A for the duration of the loss of Loop I SW. (BOP-C) (CRS-TS)
Event Five: P-32A RCP develops high vibrations and temperatures P-32A Reactor Coolant Pump will develop a high motor bearing and shaft vibrations and high motor bearing temperatures, requiring the ATC to lower power to 60 - 65% power in preparation for stopping P-32A. (ATC-R)
Event Six: Turbine stops responding during power reduction At ~ 85% power the Turbine EHC System will stop responding during the power reduction. Once identified the ATC should take the SG/Rx Master station to hand to prevent ICS from raising power due to the resultant header pressure error built in during the turbine control issue. Once the BOP returns header pressure back to setpoint, the ATC will return SG.RX Master to automatic and the BOP should continue the power reduction with the Turbine in Operator Auto at the rate directed by the CRS. (ATC-C)
(BOP-C)
Event Seven: P-32A RCP Sheared Shaft As cued by the lead examiner or after the Heater Drain Pump is stopped, P-32A will experience a sheared shaft. This will require the ATC to trip the reactor and the BOP will trip the affected RCP (P-32A). (ATC-C) (BOP-C)
Appendix D Scenario 3 Outline Page 6 of 56 Event Eight: Steam Leak in containment Post trip a steam leak on the A Steam Generator will develop inside the Reactor Building. The leak will cause Reactor Building pressure to exceed the ESAS setpoint for building pressure resulting in a transition into the Overcooling EOP. (ALL-M)
Event Nine and Ten: Containment Isolation The building pressure should result in ESAS Channels 1-6 actuating, however ESAS Channel 5 will fail to actuate and CV-6203 (RB Isolation Valve for Chilled Water) will also fail to close from the ESAS Channel 6 actuation signal. The critical task of establishing containment closure can be met by either manually actuating ESAS Channel 5 and/or manually closing CV-6203. (CT)
CT justification:
Safety Significance - Degradation of any fission product barrier Initiating Cue - ESAS actuation and alarms Measurable Performance Standard - Manually actuate Channel 5 or close CV-6203 prior to the BOP reporting that RT-10, Verify Proper ESAS Actuation, is completed.
Performing this action prior to announcement of RT-10 complete provides reasonable time for the crew to utilize procedural guidance to determine that the channel is failed and to implement actions to correct.
Performance Feedback - Status light indication of proper containment isolation Event Eleven: EFW Flow Control Valves / Vector Isolation The ESAS actuation will result in an EFW actuation. Two valves in series to the faulted Steam Generator will not receive the Vector Isolation signal (CV-2645 and CV-2627) requiring the ATC to either take manual control and close CV-2645 or stop the EFW pump, P-7A, providing flow to them in order to stop the Overcooling condition. (CT)
CT justification:
Safety Significance - Failed ESF components resulting in exceeding cool down limits Initiating Cue - EFIC actuation and indicated flow rate Measurable Performance Standard - Isolating faulted flow path prior to T-cold reaching 465 oF ensures excessive overcooling does not occur before corrective actions are taken. Additional actions are required due to concerns with hydraulic lifting of fuel assemblies at temperatures below 465 oF.
Performance Feedback - EFW flow, SG level, and RCS temperature Critical Task Measurable Performance Standards (CT) Actuate ESAS Channel 5 or close CV-6203 prior to completing RT-10.
(CT) Take manual control of CV-2645 and isolate flow to the faulted SG or stop P-7A EFW pump providing flow through the failed flow path to the faulted SG prior to T-cold reaching 465 oF.
The scenario can be terminated when overcooling is terminated and the crew has control of RCS temperature.
Appendix D Scenario 3 Outline Page 7 of 56 Scenario Recapitulation Total Malfunctions:
8 Abnormal Events:
4 Major Transients:
1 EOPs Entered:
2 EOP Contingencies:
1 Malfunctions after EOP 2 Critical Tasks:
2 Causing an unnecessary plant trip or ESAS actuation may constitute a CT failure.
Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.