ML12278A305
| ML12278A305 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/12/2012 |
| From: | Moore R Constellation Energy Nuclear Group |
| To: | D'Antonio J Operations Branch I |
| Jackson D | |
| Shared Package | |
| ML12067A049 | List: |
| References | |
| TAC U01849 | |
| Download: ML12278A305 (27) | |
Text
ES-401 PWR Examination Outline Fonn ES-401-2 II Facility: Calvert Cliffs Nuclear Power Plant Date ofExam: 08/13/20121 RO Category KIA Points SRO Only Points Tier Group K
K K
K K
K A
A A
Total A2 G
Total
~iG 1
2 3
4 5
6 I
2 3
- 1. Emergency &
1 3
3 3
3 3
3 18 3
3 6
Abnormal Plant 2
2 2
1 N/A 2
I N/A 1
9 2
2 4
Evolutions Tier Totals 5
5 4
5 4
4 27 5
5 10
')
')
1 3
3 2
2 3
2 3
3 28 3
2 5
- 2. Plant Systems 2
I 1
1 1
I 1
1 1
10 I
1 1
3 Tier Totals 3
3
=1 ! 4 3
3 K= 3 4
4 38 5
3 8
I 2
3 4
1 2
3 4
10 7
- 3. Generic Knowledge & Abilities Categories 3
.J" 2
2 2
2 I
2 Note:
- 1.
Ensure that at least two topics from every applicable KIA category are sampled within each tier ofthe RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA category shall not be less than two).
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/- I 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 assoeiated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.l.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.
- 4.
Select topics from as many systems and evolutions as possible; sar:1ple every system or evolution in the group before selecting a second topie for any system or evolution.
- 5.
Absent a plant-specific priority, only those KlAs having an importance rating (lR) of2.5 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 I and 2 from the shaded systems and KJA categories.
7.*
The generic (G) KlAs in Tiers I and 2 shall be seleeted from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.I.b ofES-401 for the applicable KlAs.
- 8.
On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side ofColumn 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 ofthe KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals
(#) on Form ES-401-3. Limit SRO selections to KlAs that are linked to 10 CFR 55.43.
ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions Tier 1 / Group 1 REACTOR OPERATOR EfAPE #lName/Safety Function 007 Reactor Trip - Stabilization Recovery II 008 Pressurizer Vapor Space Accident
/3 009 Small Break LOCA / 3 011 Large Break LOCA / 3 022 Loss ofRx Coolant Makeup / 2 025 Loss ofRHR System / 4 026 Loss of Component Cooling Water 18 027 Pressurizer Pressure Control System Malfunction / 3 029 ATWS 1\\
K x
KA Topic Imp Pts EKI - Knowledge of the operational implications of the following concepts as they apply to the reactor trip:
3.6 EK I.04 - Decrease in reactor power following reactor trip (prompt drop and subsequent decay)
AK2 - Knowledge of the interrelations between the Pressurizer Vapor Space 2.7 Accident and the following:
AK2.01 - Valves EK3 - Knowledge of the reasons for the following responses as the apply to the small break LOCA:
4.1 EK3.24 ECCS throttl ing or termination criteria EA2 - Ability to determine or interpret the following as they apply to a Large Break LOCA:
3.7 EA2.13 - Difference between overcooling and LOCA indications AK3 - Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup:
3.5 AK3.02 - Actions contained in SOPs and EOPs for RCPs, loss of makeup, loss of charging, and abnormal charging AK2 - Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following:
2.7 AK2.03 Service water or closed cooling water pumps AK3 - Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water:
3.5 AK3.04 - Effect on the CCW flow header of a loss ofCCW AK2 - Knowledge of the interrelations between tbe Pressurizer Pressure Control 2.6 Malfunctions and the following:
AK2.03 Controllers and positioners 2.4 - Emergency Procedures I Plan 2.4.45 - Ability to prioritize and interpret the 4.1 significance ofeach annunciator / alarm.
3.1 ES-401 PWR Examination Outline Fonn ES-401-2 Emergency & Abnonnal Plant Evolutions Tier 1 / Group 1 REACTOR OPERATOR E/APE #/Name/Safety Function 038 Steam Gen. Tube Rupture / 3 040 Steam Line Rupture - Excessive Heat Transfer / 4 055 Station Blackout 16 056 Loss of Off-site Power 1 6 058 Loss of DC Power 1 6 062 Loss ofNuclear Svc Water 14 065 Loss of Instrument Air / 8 077 Generator Voltage and Electric Grid Disturbances / 6 CE/E06 Loss of Main Feedwater 14 KIA Category Totals:
KA Topic Imp Pts EA2 - Ability to determine or interpret the following as they apply to a SGTR:
EA2.10 - Flowpath for charging and letdown flows AKI - Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture:
3.8 AK 1.03 - RCS shrink and consequent depressurization EAl - Ability to operate and monitor the following as they apply to a 8BO:
3.7 EAl.Ol - In-core thermocouple temperatures 2.2 - Equipment Control 2.2.3 - Knowledge of the design, procedural, 3.8 and operational differences between units.
AAl - Ability to operate and I or monitor the following as they apply to the Loss of 3.1 DC Power:
AA 1.03 - Vital and battery bus components AA2 - Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:
2.6 AA2.03 - The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition AAl-Ability to operate and I or monitor the following as they apply to the Loss of 3.5 Instrument Air:
AA 1.04-Emergency Air Compressor 2.1-Conduct of Operations 2.1.23 - Ability to perform specific system 4.3 and integrated plant procedures during all modes of plant operation.
EKI. Knowledge ofthe operational implications of the following concepts as they apply to the (Loss of Feedwater) 3.2 EK 1.2 - Normal, abnormal and emergency operating procedures associated with (Loss of Feedwater)
Group Point Total:
18
4.2 ES-401 PWR Examination Outline Fonn ES-401-2 Emergency & Abnonnal Plant Evolutions - Tier 1 / Group 2 - REACTOR OPERATOR E/APE #/Name/Safety Function 001 Continuous Rod Withdrawal / I 003 Dropped Control Rod / I 059 Accidental Liquid RadWaste ReI.
/9 061 ARM System Alarms 1 7 067 Plant Fire On-site 19 069 Loss ofCTMT Integrity I 5 076 High Reactor Coolant Activity / 9 CE/AI6 Excess RCS Leakage / 2 CE/E09 Functional Recovery KIA Category Totals:
KA Topic Imp Pts AA2 - Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:
AA2.04 - Reactor power and its trend AKI - Knowledge of the operational implications of the following concepts as 2.9 they apply to Dropped Control Rod:
AKJ.l6 - MTC AK2 - Knowledge of the interrelations between the Accidental Liquid Radwaste 2.7 Release and the following:
AK2.01 - Radioactive-liquid monitors AAI - Ability to operate and I or monitor the following as they apply to the Area Radiation Monitoring (ARM)System
3.6 Alarms
AAI.OI - Automatic actuation AK3. Knowledge of the reasons for the following responses as they apply to the Plant Fire on Site:
3.3 AK3.04 - Actions contained in EOP for plant fire on site AAl. Ability to operate and I or monitor the following as they apply to the Loss of Containment Integrity:
2.8 AA 1.03 - Flui,d systems penetrating containment AK2. Knowledge of the interrelations between the High Reactor Coolant 2.6 Activity and the following:
AK2.01 - Process radiation monitors
- 2. t Conduct of Operations 2.1.20 - Ability to interpret and execute 4.6 procedure steps.
EK1. Knowledge of the operational implications of the following concepts as they apply to the (Functional Recovery) 3.2 EK1.2 - Normal, abnormal and emergency operating procedures associated with (Functional Recovery)
Group Point Total:
9
ES-401 PWR Examination Outline Forrn ES-40 1 Plant Systems Tier 2 / Group 1 - REACTOR OPERATOR SystemlEvolution #lName 003 Reactor Coolant Pump 003 Reactor Coolant Pump 004 Chemical and Volume Control 004 Chemical and Volume Control 005 Residual Heat Removal 005 Residual Heat Rcmoval 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank KKK K 1 234 K
5 K
6 KA Topic Imp Pts x
K6 - Knowledge of the effect of a loss or malfunction on the following will have on the RCPS:
2.6 K6. 14 - Starting requirements A4 - Ability to manually operate and/or monitor in the Control Room:
3.1 A4.04 - RCP seal differential pressure instrumentation x
K6 - Knowledge of the effect of a loss or malfunction on the following CVCS components:
3.1 K6. J3 - Purpose and function ofthe boration/dilution batch controller 2.4 - Emergency Procedures /
Plan 2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
4.6 x
K5 - Knowledge of the operational implications of the following concepts as they apply the RHRS:
2.9 K5.03 - Reactivity effects of RHR fill water A4 - Ability to manually operate and/or monitor in the control room:
3.4 A4.02 - Heat exchanger bypass flow control x
K5 - Knowledge of the operational implications of the following concepts as they apply to ECCS:
K5.07 - Expected temperature levels in various locations ofthe 2.7 RCS due to various plant conditions x
Kl - Knowledge of the physical connections and/or cause/effect relationships between the PRTS and the following systems:
3.0 KI.03 - ReS
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 21 Group 1 - REACTOR OPERATOR System/Evolution #/Name KA Topic Imp Pts 007 Pressurizer Relief/Quench Tank A4 - Ability to manually operate and/or monitor in the control room:
2.7 A4.01 - PRT spray supply valve 008 Component Cooling Water K4 - Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following:
K4.09 - The "standby" feature for the CCW pumps 2.7 008 Component Cooling Water A2 - Ability to (a) predict the impacts ofthe 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 3.5 A2.03 - High/low CCW temperature 010 Pressurizer Pressure Control X
K2 - Knowledge of bus power supplies to the following:
3.0 K2.0J Pzr Heaters 012 Reactor Protection K3 - Knowledge of the effect that a loss or malfunction of the RPS will have on the following:
3.9 K3.01 -CRDS 2.4 - Emergency Procedures /
Plan 012 Reactor Protection 2.4.31 - Knowledge ofannunciator alarms, indications, or response procedures.
4.2 013 ESFAS K5 - Knowledge of the operational implications of the following concepts as they apply to the ESF AS:
2.9 K5.02 - Safety system logic and reliability 013 ESFAS A3 - Ability to monitor automatic operation ofthe ESFAS including:
3.7 A3.0J - Input channels and logic
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 I Group 1 - REACTOR OPERATOR System/Evolution #lName 022 Containment Cooling 026 Containment Spray 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution K
1 K
2 KA Topic ts At - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating 3.6 the CCS controls including:
A 1.0 I - Containment temperature A3 - Ability to monitor automatic operation of the CSS, including:
A3.02 - Verification that cooling 3.9 water is supplied to the cntmt spray heat exchanger 2.1 - Conduct of Operations 2.1.28 - Knowledge of the purpose 4.1 and function of major system components and controls.
K3 Knowledge of the effect that a loss or malfunction of the MRSS will have on the following:
2.8 K3.06 - SDS A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to 2.7 correct, control, or mitigate the consequences of those malfunctions or operations:
A2.03 - Overfeeding event A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to 3.4 correct, control, or mitigate the consequences of those malfunctions or operations:
A2.07 - Air or MOV failure K3 - Knowledge of the effect that a loss or malfunction of the ac distribution system will have on 3.5 the following:
K3.01 - Major system loads At - Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls
2.5 including
A1.0 I - Battery capacity as it is affected by discharge rate
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2 / Group 1 - REACTOR OPERATOR System/Evolution #lName 064 Emergency Diesel Generator 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air KIA Category Totals:
KKK 123 KA Topic Imp Pts x
Kl - Knowledge of the physical connections and/or cause/effect relationships between the ED/G system and the foJlowing systems:
K1.04 - DC distribution system 3.6 K4 - Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following:
4.0 K4.01 - Release termination when radiation exceeds setpoint x
K2 - Knowledge of bus power supplies to the following:
K2.01 - Service water 2.7 K4 - Knowledge of lAS design feature(s) and/or interlock(s) which provide for the following:
3.1 K4.03 - Securing of SAS upon loss ofcooling water 223 Group Point Total:
28
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2 / Group 2 - REACTOR OPERATOR System/Evolution #lName 001 Control Rod Drive 002 Reactor Coolant oII Pressurizer Level Control System 015 Nuclear Instrumentation OJ7 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 035 Steam Generator System 045 Main Turbine Generator KKK 1 234 x
x x
x KA Topic K4 - Knowledge of CRDS design feature(s) and/or interlock(s) which provide for the following:
K4.23 - Rod motion inhibit Kl - Knowledge of the physical connections and/or cause-effect relationships between the RCS and the following systems:
K1.06 - CVCS Al - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including:
A1.01 PZR level and pressure A3 - Ability to monitor automatic operation of the NIS, including:
A3.02 - Annunciator and alarm signals K6 - Knowledge of the effect of a loss or malfunction of the following ITM system components:
K6.0J - Sensors and detectors A4 - Ability to manually operate and/or monitor in the control room:
A4.01 CIRS controls K2 - Knowledge of bus power supplies to the following:
K2.0 1 - Hydrogen recombiners K3 - Knowledge of the effect that a loss or malfunction of the S/GS will have on the following:
K3.01 - RCS K5 - Knowledge of the operational implications of the following concepts as the apply to the MT/B System:
K5.17 - Relationship between moderator temperature coefficient and boron concentration in RCS as T/G load increases Imp Pts 3.4 3.7 3.5 3.7 2.7 3.3 2.5 4.4 2.5
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/ Group 2 REACTOR OPERATOR System/Evolution #lName KA Topic Imp Pts 2.1 - Conduct of Operations 075 Circulating Water 2.1.32 - Ability to explain and apply system limits and 3.8 precautions.
KIA Category Totals:
Group Point Total:
10
4.6 ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions Tier 1 / Group 1 - Senior Reactor Operator E/APE #/Name/Safety KA Topic Imp Pts Function 2.4 - Emergency Procedures / Plan 2.4.21 - Knowledge ofthe parameters and logic used to assess the status ofsafety 015/017 RCP functions, such as reactivity control, core Malfunctions / 4 cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
2.1 - Conduct of Operations 038 Steam Gen. Tube 3.8 2.1.19 - Ability to use plant computers to evaluate system or component status.
Rupture 13 AA2 - Ability to determine and interpret the following as they apply to the Loss of 054 Loss of Main Main Feedwater (MFW):
4.2 Feedwater / 4 AA2.03 - Conditions and reasons for AFW pump startup AA2 - Ability to determine and interpret the following as they apply to the Loss of 057 Loss of Vital AC Vital AC Instrument Bus:
3.9 Inst. Bus 16 AA2.03 - RPS panel alann annunciators and trip indicators 2.2 - Equipment Control CE/E02 Reactor Trip 2.2.42 - Ability to recognize system Stabilization - Recovery I 4.6 parameters that are entry-level conditions 1
for Technical Specifications.
EA2 - Ability to determine and interpret the following as they apply to the (Excess CE/E05 Steam Line Steam Demand)
Rupture - Excessive Heat 4.0 Transfer /4 EA2.1 - Facility conditions and selection of appropriate procedures during abnonnal and emergency operations KIA Category Totals:
Group Point Total:
6
ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions - Tier 1 I Group 2 Senior Reactor Operator E/APE #/Name/Safety Function 005 Inoperable/Stuck Control Rod / 1 024 Emergency Boration 11 028 Pressurizer Level Malfunction / 2 037 Steam Generator Tube Leak 13 KIA Category Totals:
KA Topic 2.4 - Emergency Procedures I Plan 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.
AA2 - Ability to determine and interpret the following as they apply to the EmergEmcy Boration:
AA2.05 - Amount ofboron to add to achieve required SDM AA2 - Ability to determine and interpret the following as they apply to the Pressurizer Level Control AA2.08 - PZR level as a function of 2.4 - Emergency Procedures I Plan 2.4.18 - Knowledge ofthe specific bases for EOPs.
roup Point Total:
Imp Pts 4.2 3.9 3.5 4.0 4
ES-401 PWR Examination Outline Form ES-401 Plant Systwms - Tier 2 / Group 1 Senior Reactor Operator System/Evolution #lName 003 Reactor Coolant Pump System 004CVCS 006 ECCS 059 Main Feedwater 063 DC Electrical Distribution KIA Category Totals:
KA Topic Imp Pts A2 - - Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to 3.9 correct, control, or mitigate the consequences of those malfunctions or operations:
A2.01 - Problems with RCP seals, especially rates of seal leak-off A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use 3.7 procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
A2.17 - Low PZR pressure A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to 4.4 correct, control, or mitigate the consequences of those malfunctions or operations:
A2.11 - Rupture of ECCS header 2.4 - Emergency Procedures /
Plan 2.4.4 - Ability to recognize 4.6 abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
2.2 - Equipment Control 2.2.42 - Ability to recognize system parameters that are entry-level 4.6 conditions for Technical Specifications.
Group Point Total:
4.6 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2 / Group 2 - Senior Reactor Operator System/Evolution #lName 029 Containment Purge 034 Fuel Handling Equipment 041 Steam Dump/Turbine Bypass Control KIA Category Totals:
KA Topic 2.2 - Equipment Control 2.2.37 - Ability to determine operability and/or availability of safety related equipment.
Kl - Knowledge of the physical connections and/or cause-effect relationships between the Fuel Handling System and the following systems:
Kl.04 - NJS A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations:
A2.03 - Loss of lAS Group Point Total:
Imp Pts 3.5 3.1 3
ES-401 PWR Examination Outline Form ES-401-3 Tier 3 Generic Knowledge & Abilities Outline - RO & SRO II Facility: Calvert Cliffs Nuclear Power Plant Date ofExam: 08/13/2012 Category KIA #
2.1.3 2.1.23 Conduct 2.1.43 of Operations 2.1.20 2.1.35 Subtotal 2.2.4 2.2.42 Equipment Control 2.2.43 2.2.18 Subtotal 2.3.4 2.3.5 Radiation Control 2.3.6 2.3.11 Subtotal 2.4.14 2.4.29 Emergency 2.4.1 ProcedureslPlan 2.4.11 Subtotal ITier 3 Total(s)
RO SRO Topic IR IR Knowledge ofshift or short-term reliefturnover practices.
3.7 1
Ability to perform specific; system and integrated plant procedures 4.3 1
during all modes ofplant operation.
Ability to use procedures to determine the effects on reactivity ofplant changes, such as reactor coolant system temperature, secondary plant, 4.1 1
fuel depletion, etc.
Ability to interpret and execute procedure steps.
4.6 1
Knowledge ofthe fuel-handling responsibilities ofSROs.
3.9 1
3 2
(multi-unit license) Ability to explain the variations in control board/control room layouts, systems, instrumentation and procedural 3.6 1
actions between units at a facility.
Ability to recognize system parameters that are entry-level conditions 3.9 1
for Technical Specifications.
Knowledge ofthe process used to track inoperable alarms.
3.0 1
Knowledge ofthe process for managing maintenance activities during 3.9 1
shutdown operations, such as risk assessments, work prioritization, etc.
3 1
Knowledge ofradiation exposure limits under normal or emergency 3.2 1
conditions.
Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personal monitoring 2.9 1
equipment, etc.
Ability to approve release permits.
3.8 1
Ability to control radiation releases.
4.3 1
2 2
Knowledge of general guidelines for EOP usage.
3.8 I
Knowledge ofthe emergency plan.
3.1 I
Knowledge ofEOP entry conditions and immediate action steps.
4.8 1
Knowledge ofabnormal condition procedures.
4.2 1
2 2
I 10 I I
I 7 I
I ES-401 Tier I Group RO-1I1 RO-1I1 RO - 111 RO-1I1 RO-1I1 RO -III RO - 112 RO - 211 RO 2/1 RO - 2/1 RO - 2/1 Randomly Selected KIA 008 Pressurizer Vapor Space Accident AK2.02 022 Loss of Reactor Coolant Makeup AK3.03 056 Loss ofOff-site Pwr, KIA 2.2.2 058 Loss of DC Power AAI.OI 065 Loss of Instrument Air KIA AA1.05 077 Generator Voltage &
Grid Disturbances, KIA 2.1.26 CE/AI6 Excess RCS Lkg, KIA 2.1.26 010 Pressurizer Pressure Control K2.04 012 Reactor Protection, KIA 2.4.42 026 Containment Spray, KIA 2.1.21 103 Containment A2.03 Record of Rejected KIAs Form ES-401-4 Reason for Rejection During exam review, noted question developed for this KIA provided answer to another question. There is not enough subject matter to author another question using original KIA. Kept system and replaced with KIA AK2.01 which was randomly selected, using numbered poker chips.
During NRC review, question does not match selected KIA. This KIA is not applicable to plant. Kept system and replaced with KIA AK3.02 which was randomly selected, using numbered poker chips.
ES-40 1 contains guidance, in the form of a list, on generic KIAs for use with Tiers I & 2. The randomly selected KIA is not on the ES-401 list.
Replaced with KIA 2.2.3, which was randomly drawn, using numbered poker chips.
During NRC review question selected did not match KIA. Determined KIA not applicable to plant. Kept system and replaced with KIA AA 1.03 which was randomly selected, using numbered poker chips.
During exam review, noted question developed for this KIA provided answer to another test question. Kept system and replaced with KIA AA 1.04 which was randomly selected, using numbered poker chips.
ES-40 I contains guidance, in the form ofa list, on generic KIAs for use with Tiers 1 & 2. The randomly selected KIA is not on the ES-401Iist.
Replaced with KIA 2.1.23, which was randomly drawn, using numbered poker chips.
ES-40 1 contains guidance, in the form of a list, on generic KIAs for use with Tiers 1 & 2. The randomly selected KIA is not on the ES-40l list.
Replaced with KIA 2.1.20, which was randomly drawn, using numbered poker chips.
Spent several unsuccessful hours attempting to develop a question for this combination of systemlKlA. Kept system 010 and replaced KIA with one that had not been sampled. KIA K2.01 was randomly selected, using numbered poker chips.
ES-40 I contains guidance, in the form of a list, on generic K/As for use with Tiers 1 & 2. The randomly selected KIA is not on the ES-401 list.
Replaced with KIA 2.4.31, which was randomly drawn, using numbered poker chips.
ES-40 1 contains guidance, in the form of a list, on generic KIAs for use with Tiers 1 & 2. The randomly selected KIA is not on the ES-401 list.
Replaced with KIA 2.1.28, which was randomly drawn, using numbered poker chips.
Spent several unsuccessful hours attempting to develop a question for this combination ofsystem/K/A. Kept KIA A2.03 and replaced system with one that had only been sampled once. System 008 - Component Cooling Water System was randomly selected, using numbered poker chips.
ES-401 Record of Rejected KIAs Form ES-401-4 Tier I Group RO - 211 RO - 212 RO - 212 RO - 2/2 RO - 2/2 SRO - 111 SRO - 111 SRO-211 SRO - 211 SRO - 2/2 Randomly Selected KiA 073 Process Radiation Monitoring, KiA 2.4.49 015 Nuclear Instrumentation KiA A3.0S 072 - Area Radiation Monitoring 075 Circulating Water, KiA 2.1.1 086 Fire Protection, KiA 3.01 0] 5/0] 7 RCP Malfs, KIA 2.4.5 057 Loss of Vital AC Inst.
Bus, KiA AA2.01 059 Main Feedwater, KiA 2.4.44 103 Containment System A2.01 029 Containment Purge, KIA 2.2.18 Reason for Rejection Spent several unsuccessful hours attempting to develop a question for this combination of systemlKiA. Kept KiA 2.4.49 and replaced system with one that had only been sampled once. System 004 - cves was randomly selected, using numbered poker chips.
Spent several unsuccessful hours attempting to develop a question for this combination ofsystemlKiA. Kept system and replaced KIA with another one in the same system. KIA 3.02 was randomly selected, using numbered poker chips.
Spent several unsuccessful hours attempting to develop a question for this combination ofsystemlKiA. Kept KIA 1.01 and replaced system with one that had not been sampled. System 011 - Pressurizer Level Control System, was randomly selected, using numbered poker chips.
ES-40 I contains guidance, in the form of a list, on generic Ki As for use with Tiers I & 2. The randomly selected KIA is not on the ES-401 list.
Replaced with KIA 2.1.32, which was randomly drawn, using numbered poker chips.
Spent several unsuccessful hours attempting to develop a question for this combination ofsystemlKiA. Kept KIA 3.01 and replaced system with one that had not been sampled. System 035 - Steam Generator, was randomly selected, using numbered poker chips.
ES-40 I contains guidance, in the form of a list, on generic Ki As for use with Tiers I & 2. The randomly selected KIA is not on the ES-401 list.
Replaced with KIA 2.4.21, which was randomly drawn, using numbered poker chips.
Site procedures do NOT specifically address this condition. Spent approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> trying to formulate question.
Replaced with KIA AA2.03, which was randomly drawn, using numbered poker chips.
ES-40 I contains guidance, in the form of a list, on generic Ki As for use with Tiers I & 2. The randomly selected KiA is not on the ES-401 list.
Replaced with KIA 2.4.4, which was randomly drawn, using numbered poker chips.
Spent several unsuccessful hours attempting to develop a question for this combination ofsystemlKiA. Kept KiA 2.01 and replaced system with one that had not been sampled. System 003 - Reactor Coolant Pump System was randomly selected, using numbered poker chips.
ES-40 I contains guidance, in the form of a list, on generic Ki As for use with Tiers 1 & 2. The randomly selected KiA is not on the ES-401 list.
Replaced with KIA 2.2.37. which was randomly drawn, using numbered poker chips.
ES-401 Record of Rejected KIAs Fonn ES-401-4 Tier /Group Randomly Selected KIA Reason for Rejection 3IRO Generic 2.3.5 Spent several unsuccessful hours attempting to develop a question for this KIA. KiA 2.2.43 was randomly selected, using numbered poker chips.
3/RO Generic 2.4.5 Spent several unsuccessful hours attempting to develop a question for this KIA. KIA 2.4.14 was randomly selected, using numbered poker chips.
Spent several unsuccessful hours attempting to develop a question that 3/SRO Generic 2.3.6 has not appeared on last two NRC exams for this KIA. KIA 2.2.36 was randomly selected, using numbered poker chips.
I I
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/3/12 thru 8/23112 Exam Level: RO / SRO-I / SRO-U Operating T,est #: 2012 Administrative Topic Type Describe activity to be performed (see Note)
Code*
Determine Status of Safety Functions for the FRP Conduct of Operations N,S (SRO-ADMIN-l) 2.1.20 - Ability to interpret and execute procedure steps (4.6, 4.6)
Evaluate the need for plant cooldown.
(SRO-ADMIN-2)
Conduct of Operations R,M 2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc. (3.9,4.2).
Monitor Azimuthal Power Tilt (Tq) using Excore Nuclear Instrumentation (SRO-ADMIN-3)
Equipment Control D. P, R 2.2.42 - Ability to recognize system parameters that arc entry-level conditions for Technical Specifications (3.9, 4.6)
Respond to a contaminated injured person (SRO-ADMIN-4)
Radiation Control M.R 2.3.14 - Knowledge of radiation or contamination hazards that may arise during normaL abnormaL or emergency conditions or activities. (3.4. 3.8)
Determine the appropriate emergency response actions per the ERPIP while Emergency Procedures /
maintaining an overview of plant conditions. (SRO-ADMIN-5)
D,R Plan 2.4.41 - Knowledge ofthe emergency action level thresholds and classifications (2.9, 4.6).
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (::; 3 for ROs; ::; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (2': 1)
(P)revious 2 exams (::; 1; randomly selected)
ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/3/12 thru 8/23/12 Exam Level: RO / SRO-I 1SRO-U Operating Test #: 2012 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
Safety System 1JPM Title Type Code*
Function
- a.
SIM-3, Respond to a FRV or FRV controller malfunction D,E,S 4 (secondary)
A, E, EN, L, P,
- b.
SIM-4, Verify Recirculation Actuation Signal 2
S
- c.
SIM-5, Verify the Containment Environment Safety Function is satisfied.
A,E,L, M, S 5
In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- a.
PL T -1, Obtain Safe Shutdown Equipment then Strip M CC-114 R D,E,L,R 6
- b.
PLT-2, Isolate Di Water And Condensate Makeup To The Service Water E,L,N 8
And Component Cooling Head Tanks All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO 1SRO-I 1SRO-U (A)ltemate path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank
-:;9/-:;8/-:;4 (E)mergency or abnormal in-plant 21121121 (EN)gineered safety feature
- / - 1 21 (control room system)
(L)ow-Power / Shutdown 21121121 (N)ew or (M)odified from bank including leA) 22/22/21 (P)revious 2 exams
-:; 3 1-:; 3 / -:; 2 (randomly selected)
(R)CA 21/21121 (S)imulator
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/3/12 thru 8/23/12 Exam Level: RO I SRO-I / SRO-U Operating Test #: 2012 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Detennill(~ Status of Safety Functions for the FRP Conduct of Operations N,S (SRO-ADMIN-l) 2.1.20 - Ability to interpret and execute procedure steps (4.6, 4.6)
Evaluate the need for plant cooldown.
Conduct ofOperations R,M (SRO-ADMIN-2) 2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc. (3.9, 4.2).
Monitor Azimuthal Power Tilt (Tq) using Excore Nuclear Instrumentation Equipment Control D,P,R (SRO-ADMIN-3) 2.2,42 - Ability to recognize system parameters that are entry-level conditiom for Technical Specifications (3.9, 4.6)
Respond to a contaminated injured person (SRO-ADM1N-4)
Radiation Control M,R 2.3.14 - Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (3.4, 3.8)
Detennine the appropriate emergency response actions per the ERPIP Emergency Procedures /
Plan D,R while maintaining an overview of plant conditions. (SRO-ADMIN-5) 2.4.41 - Knowledge ofthe emergency action level thresholds and classifications (2.9,4.6).
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:'S 3 for ROs; :'S 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (~ 1)
(P)revious 2 exams (:'S 1; randomly selected)
ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2
~Clear Power Plant Date of Examination: 8/3/12 thru O/"l/n I: RO / SRO-J / SRO-U Operating Test #: 2012 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
Safety System / JPM Title Type Code*
Function I a.
SIM-I, Respond to a Loss ofSDC with the RCS Open A,L,P,S 4 (primary)
- b.
SIM-2, Respond to a Pressurizer Spray Valve Failure A,P, S 3
- c.
SIM-3, Respond to a FRV or FRV controller malfunction D,S 4 (secondary)
- d.
SIM-4, Verify Recirculation Actuation Signal A,EN, L, P, S 2
- e.
SIM-5, Verify the Containment Environment Saiety Function is satisfied.
A,M,S 5
- f.
SIM-6, Null NI Pots to DeltaT Pots D, S 7
- g.
SIM-8, Unload and shutdown the OC DG D,S 6
In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- a.
PL T -1, Obtain Safe Shutdown Equipment then Strip MCC-114R D,E,L,R 6
- b.
PLT-2, Isolate Di Water And Condensate Makeup To The Service Water E,L,N 8
And Component Cooling Head Tanks
- c.
PLT-3, Align AFW Pump Speed Control to lC43 A,E,L,M 4 (secondary)
All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-J / SRO-U (A)iternate path 4-6/4-6/2-3 (C)ontrol room (D)irect from bank 59/58/54 (E)mergency or abnormal in-plant 2:1I2:1/;~1 (EN)gineered safety feature
- I - 1 (control room system)
(L)ow-Power I Shutdown 2:1/2:1/:~I (N)ew or (M)odified from bank including leA) 2:2/2:2/;~1 (P)revious 2 exams 5 3 15 3 152 (randomly selected)
(R)CA 2:112:1/2:1 (S)imulator
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date ofExamination: 8/3/12 thru 8/23112 Exam Level: RO / SRO-I / SRO-U Operating Test #: 2012 Type Administrative Topic (see Note)
Describe activity to be performed Code*
Estimate Time to Boiling and Core Uncovery (RO-Admin-l)
Conduct ofOperations R,M 2.1.20 - Ability to interpret and execute procedure steps (4.6,4.6).
Calculate BAST volume required to raise R WT to refueling boron concentration (RO-Admin-2)
Conduct ofOperations R,N 2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc. (3.9,4.2)
Equipment Control Determine protective clothing and limits associated with performance ofa task in the RCA (RO-Admin-3)
Radiation Control M,R 2.3.7 - Ability to comply with radiation work permit requirements during normal or abnormal conditions. (3.5, 3.6)
Perform an Independent Assessment ofan Event Using the EOP-O Diagnostic Flowchart and Recommend the Correct Recovery Procedure. (RO-ADMJN-4)
Emergency Procedures I Plan D,S 2.4.21 - Knowledge ofthe parameters and logic used to assess the status ofsafety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (4.0,4.6)
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank 3 for ROs; :s. 4 for SROs & RO retakes)
(N)ew or (M)odifil~d from bank (?:: 1)
(P)revious 2 exams (:s. 1; randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Date ofExamination: 813/12 thru 8/23/12 Exam Level: RO / SRO-I / SRO*U Operating Test #: 2012
~
om Systems: (8 for RO); (7 for SRO-J); (2 or 3 for SRO-U, including 1 ESF)
Safety System / JPM Title Type Code*
Function
- a.
SIM-l, Respond to a Loss of SDC with the RCS Open A,L,P,S 4 (primary)
- b.
SIM-2, Respond to a Pressurizer Spray Valve Failure A, P, S 3
- c.
SIM-3, Respond to a FRV or FRV controller malfimction D, S 4 (secondary)
- d.
SIM-4, VerifY Recirculation Actuation Signal A, EN, L, P, S 2
- e.
SIM-5, VerifY the Containment Environment Safety Function is satisfied.
A,M,S 5
- f.
SIM-6, Null NI Pots to DeitaT Pots D, S 7
- g.
SIM-7, Respond to Inadvertent Dilution During Reactor Startup D,L,S 1
- h.
SIM-8, Unload and shutdown the OC DG D, S 6
In-Plant Systems@ (3 for RO); (3 for SRO-J); (3 or 2 for SRO-U)
- a.
PLT-1, Obtain Safe Shutdown Equipment then Strip MCC-114R D,E,L,R 6
- b.
PLT-2, Isolate Di Water And Condensate Makeup To The Service Water E,L,N 8
And Component Cooling Head Tanks
- c.
PLT-3, Align AFW Pump Speed Control to 1 C43 A, E, L, M 4 (secondary)
All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those testt:d in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)ltemate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank
$9/$8/$4 (E)mergency or abnormal in-plant
- 1/2
- 112:1 (EN)gineered safety feature
- I - 1 2:1 (control room system)
(L)ow-Power / Shutdown
- 1/2
- 112:1 (N)ew or (M)odified from bank including I(A)
- 2/;
- ::2/2: 1 (P)revious 2 exams
$ 3 I :s 3 / :s 2 (randomly selected)
(R)CA 2:112:1/'21 (S)imulator
Appendix D Scenario Outline Form ES-D-l Facility: Calvert Cliffs Nuclear Power Plant Scenario #: 1 OP-Test #: CCNPP 2012 Examiners:
Operators:
Initial Conditions: Unit-l is at 65% Power with a core burnup of 10,885 MWD/MTU. Unit-2 is in Mode 5. 12 CS pump is OOS for repairs. 23 Auxiliary Feedwater Pump is out of service for overhaul of the coupling (expected back in three shifts).
Turnover: Place 12 SGFP in parallel operation with 11 SGFP and return power to 100%.
Event Malfunction #
Event Type*
Event Description 1
None N - BOP/SRO Aligns 12 SGFP for parallel operation will SGFP I -ATC 2
rcs026 01 Failure ofLT-llOX T-SRO R-ATC 3
None Commences power escalation N - BOP/SRO C - BOP/SRO 4
srw003 01 11 Service Water Pump Bkr Failure T-SRO 5
cd005 01 C - BOP/SRO Condensate Booster Pump trip Condensate Booster Pump discharge header rupture
- cd008, (25%). Reactor will not trip automatically or with 6
M-ALL rps005, rps006 Rx Trip pushbuttons. CEDM MG Sets must be de-energized to trip the reactor.
II 7
Various C-ALL Loss of All Feedwater I Once-Thru-Cooling (N)ormal (R )eacti vity (I)nstrument (C)omponent (M)ajor (T)ech Spec Critical Tasks: (shaded)
- 1. Deenergizes CEDM MG sets within 1 minute ofan existing Reactor Trip Condition.
Reports Reactivity Control complete (report not critical). N/A ifRPS setpoint not reached.
- 3. Initiates OTCC when both S/G levels are < -350" or TcoLDrises > 5°F uncontrollably (must be commenced prior to CET temperatures reaching 560°F)
Appendix D Scenario Outline Form ES-D-l Facility: Calvert Cliffs Nuclear Power Plaut Scenario #: 2 OP-Test CCNPP2012 Examiners:
Operators:
Initial Conditions: Uuit-l is at 100% power. Core Burnup is 17,885 MWD/MTU. Unit-2 is in Mode 5.
Turnover: 12 AFW Pump is OOS due to a governor problem (has been OOS for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and is due back by end of shift). The SMECO tie breaker is open. 14 CAR is OOS due to a failed motor bearing. The 2A DG was removed from service yesterday for scheduled maintenance. 11 ADV is wisping a small amount of steam. Instructions to the crew are to maintain full power.
.Ll T.'.
~
~
,~** TT n........
.,. Y,".lU.I.'y 1:"
Event Description 1
'v~:_01 C - ATC/SRO 11 Charging Pump coupling failure 2
rcs023 01 I -ATC PT-I00X fails high C - BOP/SRO 3
swOOI 02 Saltwater Leak T-SRO R-ATC 4
tg024 01 C - BOP/SRO MTCV-1 Fails closed T-SRO 5
swyd002 M-ALL Loss Of Offsite Power resulting in a reactor trip 6
I dg002_02 C-ALL No 4KV SR Busses resulting in Station Blackout afw004 01 7
C-ALL AFAS "A" & "B" failure afw004-02 (N)ormal (R )eacti vity (I)nstrument (C)omponcnt (M)ajor (T)ech Spec Critical Tasks: (shaded)
- 1. Stops Salt Water leak by securing 12 SW Pump or by isolating Salt Water to 12A and 12B SRW HXs prior to tripping the SRW pumps on high room level.
- 2. Establishes an RCS heat sink
Appendix D Scenario Outline Form ES-D-t Facility: Calvert Cliffs Nuclear Power Plant Scenario #: 3 OP-Test #: CCNPP 2012 Examiners:
Operators:
Initial Conditions: Unit-l is at 100% with a core burnup of 10,885 MWD/MTU. Unit-2 is in Mode 5.
Turnover: 11 4 KV Bus alternate feed is tagged out for breaker PMs. The crew is instructed to maintain 100% power.
Event #
Malfunction 1
rcs027 01 2
cntmOOI 01 3
ms015 msOOl 02 4
ms002 02 6
afwOOl 01
- T,,~~*
C-ATC T-SRO C - BOP/SRO T-SRO I-ALL R-ATC C-BOP/SRO T-SRO M-ALL C - BOP/SRO Event Description PORV-402leakby 11 Containment Air Cooler (CAC) trips ADV Controller, I-HIC-4056, fails in automatic 11 S/G Tube leak (0 65 GPM over 2 minutes)
Reduce TAvE to less than 53TF 11 S/G Tube Rupture - one tube 11 AFW Pump triplll 4KV Bus loss (N)orrnal (R )eactivi ty (I)nstrument (C)omponent (M)ajor (T)ech Spec Critical Tasks: (shaded)
- 1. Maintains RCP trip strategy. Trips 11 A and 12B or 11 B and 12A RCPs when ReS pressure < 1725 PSIA.
- 1. Isolates 12 S/G when it has been identified as the most affected S/G and THOT is < 515 0 F.