ML19227A013
| ML19227A013 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/26/2019 |
| From: | Exelon Generation Co |
| To: | Peter Presby Operations Branch I |
| Shared Package | |
| ML18151A269 | List: |
| References | |
| CAC00500, EPID: L-2019-OLL-0014 | |
| Download: ML19227A013 (23) | |
Text
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Three Mile Island Date of Examination:
06/10/19 Examination Level: RO IZI SRO D Operating Test Number:
TMl2019 Administrative Topic (see Note)
Type Describe activity to be performed Code*
Perform a Reactivity Balance at Power Conduct of Operations N,R KIA: 2.1.25 (3.9)
Complete RB Average Air Temperature Conduct of Operations D,R Calculation K/A: 2.1.7 (4.4)
Station Print Reading - Isolate Instrument Air Equipment Control Leak N,R KIA: 2.2.41 (3.9)
Radiation Control Emergency Plan Perform State and Local Event Notification D,S KIA: 2.4.43 (3.2)
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 (s 3 for ROs; s 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:
Three Mile Island Date of Examination:
06/10/19 *-*
Examination Level: RO D SRO IZI Operating Test Number:
TMl2019 Administrative Topic (see Note)
Type Describe activity to be performed Code*
Issue a Controlled Key Conduct of Operations M,R KIA: 2.1.13 (3.2)
Calculate and Approve an SOM Conduct of Operations D,R KIA: 2.1.37 (4.6)
Evaluate completed surveillance and perform Equipment Control D,R actions KIA: 2.2.37 (4.6)
Authorize emergency personnel exposure in Radiation Control D,R excess of 5 REM KIA: 2.3.4 (3.7)
EAL and PAR Emergency Plan N,R KIA: 2.4.44 (4.4)
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 (s 3 for ROs; s 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (,;: 1)
(P)revious 2 exams (S 1, randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Three Mile Island Date of Examination:
06/10/19 Exam Level: RO
[Z]
SRO-I D SRO-U D Operating Test Number:
TMl2019 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. 001 / Respond to a dropped control rod - ICS fails to complete D,S 1
run back 003AA 1.02 - Control Rod System
system L
- c. 006 I Lower CFT level and pressure from the Control Room - 006 D,S 3
A4.02 - Core Flood System
- e. 003 I Restore SI with a loss of ICCW 003A3.01 - Reactor Coolant D,A,S,P 4P Pumo
- f. 007 I Pump RCDT to MWST 007A1.01 - Pressurizer Relief Tank D,S 5
064A4.01 - Emergency Diesel Generators
- h. 072 / Respond IAW OP-TM-MAP-C0101 Fuel Handling Incident in N,S 7
the Spent Fuel Pool - Radiation Monitors In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U i.008 I Loss of Instrument Air - 008A2.05 - Intermediate Closed D,E,R 8
Cooling Water System
- j. 071 / Purge of the Waste Gas System Radiation Monitor (RM-A-7)
D,R 9
071A4.09-Waste Gas Disposal System
- k. 061 / Respond to a failure of EF-P-2A and EF-V-30D 061 A2.04 -
D,E 4S Emergency Feedwater 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 I
Criteria for R /SR0-1/SRO-U
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Three Mile Island Date of Examination:
06/10/19 Exam Level: RO D SRO-I
~ SRO-U D Operating Test Number:
TMl2019 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. 001 / Respond to a dropped control rod - ICS fails to complete D,S 1
runback 003AA1.02 - Control Rod System
system L
- c. 006 I Lower CFT level and pressure from the Control Room - 006 D,S 3
A4.02 - Core Flood System
- d. 061 / Respond to Emergency Feedwater Actuation -ALT Path D,A,S 4S 061 A2.05 - Emerqencv Feedwater
- e. 003 I Restore SI with a loss of ICCW 003A3.01 - Reactor Coolant D,A,S,P 4P Pump
- f. N/A
064A4.01 - Emergency Diesel Generators
- h. 072 / Respond IAW OP-TM-MAP-C0101 Fuel Handling Incident in N,S 7
the Spent Fuel Pool - Radiation Monitors In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U i.008 I Loss of Instrument Air- 008A2.05 - Intermediate Closed D,E,R 8
Cooling Water System
- j. 071 / Purge of the Waste Gas System Radiation Monitor (RM-A-7)
D,R 9
071 A4.09 - Waste Gas Disposal System
- k. 061 / Respond to a failure of EF-P-2A and EF-V-30D 061A2.04 -
D, E 4S Emergency Feedwater 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.
I
- Type Codes I
Criteria for R /SR0-1/SRO-U I
Appendix D Scenario Outline Form ES-D-1 ILT 18-01 NRC EXAM MATERIAL Facility:
Three Mile Island Scenario No.:
1 Op Test No.:
TMl2019 Examiners:
Operators:
Initial Conditions:
85% power, MOL as ordered by the load dispatcher.
AH-E-18B is running for a surveillance (1303-5.5B)
EF-P-1 is OOS for the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Main Feedwater Pumps in Manual Turnover:
Maintain 85% power Critical Tasks:
CT-1 Shutdown reactor-ATWS CT-2 Restore feed to a dry OTSG CT-3 Establish a cooldown rate less than Guide 11 limits (if necessary)
Event Malf. No.
Event Event Description No.
Type*
1 CH630TCRC TSCRS AH-E-18 trip CARO (ARO: Re-aligns ventilation, CRS: TS call).
2 RD10B ICRS Uncontrolled inward rod motion, entry into OP-TM-AOP-070 IURO (URO/ARO: Manual control of ICS)
IARO 3
MU06 TSCRS MU-V-18 fails partially closed CURO (URO: Controls pzr level with HPI) 4 FW16A CCRS
'A' MFP Trips, manual runback required RURO CARO (URO/ARO: manual runback) 5 FW15B MCRS
'B' MFP trips, Reactor Trip with an ATWS RD28 MURO RD32 MARO 6
FW18A CCRS Sequential loss of all EFW pumps. Entry into OP-TM-EOP-004, FW18B CURO Lack of Heat Transfer.
CARO (URO: Secures RCP, ARO: Condensate Booster pump cooling)
CCRS (If required) HPI-PORV cooling, entry into OP-TM-EOP-009, HPI CURO Cooling (URO: Opens PORV)
(N)ormal,
( R )eactivity, (l)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline I
Form ES-0-1 Facility:
Three Mile Island Scenario No.:
2 Op Test No.:
TMl2019 Examiners:
Operators:
Initial Conditions:
28% power RC-P-1 B ready to start Nl-8 OOS due to a failed power supply RPS Channel 'D' is in manual bypass, RPS logic is 2 out of 3 to trip Turnover:
Start RC-P-18 Critical Tasks:
Control HPI (CT-5)
Event Malf. No.
Event Type*
Event Description No.
1 NCRS Start RC-P-1 B IAW OP-TM-226-102 NARO (ARO: Start RCP) 2 RCR42 ICRS Pressurizer Spray Valve Failure RCR43 IURO (URO: Closes spray block valve) 3 Nl15B TSCRS Nl-6 failure (fails low)
CARO (ARO: Places RPS channel 'B' in tripped state) 4 RD0117 TSCRS Dropped rod group 7 CURO (URO: Recovers dropped rod) 5 MS02A CCRS Steam leak in RB entry into OP-TM-AOP-051 and 1102-4 RURO CARO (URO: Lower power, ARO: RB Emergency Cooling) 6 MS02A MCRS Steam line rupture in RB, Reactor Trip, OP-TM-EOP-003, XHT MURO entry.
MARO 7
FW19A CCRS EF-V-30A fails open, entry into OP-TM-424-901 CARO (ARO: Closes EF-V-2A, secures EF-P-2A)
(N)ormal, (R)eactivity,
{l)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Facility:
Three Mile Island Scenario No.:
3 Op Test No.:
TMl2019 Examiners:
Operators:
Initial Conditions:
2% power, MOL, ICS is in manual with reactivity control at the diamond Turbine Reset and all 6 Circulating Water Pumps are running for a PMT.
FW-P-1 B is operating with control on the air speed changer Engineers are doing systems walkdowns in the control tower and turbine building.
Fuel is not deconditioned Turnover:
Raise reactor power to 10%, initiate a bleed to the 'B' RCBT Critical Tasks:
Establish and Maintain Reactor Shutdown Requirements (CT-1)
Control HPI (CT-2)
Event No.
Malt. No.
Event Type*
Event Description 1
NCRS Raise reactor power from 3% to 10%
RURO (URO: Power ascension with ICS in Manual, ARO: Bleeds to 'B' NARO RCBT) 2 RM0323 TSCRS Reactor Building Hi Range Radiation Monitor, RM-G-23, Failure 3
RC04A ICRS Pressurizer Level Transmitter fails, entry into OP-TM-MAP-G0105, IURO OP-TM-MAP-G0205 (URO: Controls MU-V-17 in HAND) 4 ED40A TSCRS Loss of the 'D' 4kv Bus, EG-Y-1A fails to auto start EG21A CARO (ARO: Starts EG-Y-1A) 5 CCRS Cavitating Gire Water Pump CARO (ARO: Secure cavitiating circ water pump) 6 MU07 ICRS Seal Flow Instrument Fails, RCP Seal flow High IURO (URO: Normalizes Seal Injection) 7 PLA-4-9 MCRS Gire Water Rupture, Loss of Vacuum, Reactor Trip, Entry into EOP-PLB-8-3 MURO 001, Stuck Rods MARO 8
TH06 CCRS RCS leak, PZR Level Cannot be maintained without HPI, Entry into CURO EOP-006 CARO (URO: Initiate HPI, ARO: Initiate EFW)
(N)ormal, (R)eactivity,
( I )nstrument, (C)omponent, (M)ajor ES-401 PWR Examination Outline Form ES-401-2 Facility: TMI Date of Exam: June 2019 RO KIA Category Points SRO-Only Points Tier Group K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
1 3
1 4
4 4
2 18 2
4 6
Emergency and Abnormal Plant 2
1 1
1 N/A 2
3 N/A 1
9 2
2 4
Evolutions Tier Totals 4
2 5
6 7
3 27 4
6 10 1
3 2
2 3
2 2
2 3
4 3
3 29 2
3 5
- 2.
Plant 2
1 0
1 2
0 2
2 0
0 1
0 9
2 0
1 3
Systems Tier Totals 4
2 3
5 2
4 4
3 4
4 3
38 4
4 8
- 3. Generic Knowledge and Abilities 1
2 3
4 10 1
2 3
4 7
Categories 3
3 1
3 2
2 1
2 Note: 1.
Ensure that at least two topics from every applicable KIA 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 KIA category shall not be less than two). (One Tier 3 radiation control KIA is allowed if it is replaced by a KIA 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 KIA 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 KIAs 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 KIA categories.
- 7.
The generic (G) Kl As in Tiers 1 and 2 shall be selected from Section 2 of the KIA 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 KIAs.
- 8.
On the following pages, enter the KIA 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 mimbers, descriptions, I Rs, and point totals(#) on Form ES-401-3. Limit SRO selections to KIAs 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 KIA catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the KIA catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 IIES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO/SRO)
E/APE #/Name I Safety Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR 000007 (EPE 7; BW E02&E10; CE E02)
X K3.01 Knowledge of the reasons for the following as 4.0 1
Reactor Trio, Stabilization, Recovery/ 1 the apply to a reactor trip:
000008 (APE 8) Pressurizer Vapor Space X
AA2.2s Aeility te Eletern1iAe aAEl iAteFpFet U1e 2.8 2
Accident/ 3 follewiAg as they apply to the PrnssurizeF VapeF SpaGe AGGiEleAt: expeGtoEl leak rate freFA opeA PORV eF Gede-safety X
AA2.04 Ability to determine and interpret the 3.2 2
following as they apply to the Pressurizer Vapor Space Accident: High-temperature computer alarm and alarm tvoe 000009 (EPE 9) Small Break LOCA / 3 X
2.4.18 Knowledge of the specific bases for EOPs.
3 000011 (EPE 11) Large Break LOCA / 3 X
EK1.01 Knowledge of the operational implications 4.1 4
of the following concepts as they apply to the Large Break LOCA : Natural circulation and cooling, including reflux boiling 000015 (APE 15) Reactor Coolant Pump X
AK3.07 Knowledge of the reasons for the following 4.1 5
Malfunctions / 4 responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow) : Ensuring that SIG levels are controlled properly for natural circulation enhancement 000022 (APE 22) Loss of Reactor Coolant X
AA2.02 Ability to determine and interpret the 3.2 6
following as they apply to the Loss of Reactor Makeup/ 2 Coolant Makeup: Charging pump problems 000025 (APE 25) Loss of Residual Heat X
AK3.G1 KAewleElge ef the FeaseAs for the follewiAg J4 +
Removal System / 4 respeAses as they apply to the boss ef Residual Heat ReFAeval SysteFA: Shift te alternate flewpath X
2.4.34 KAowledge ef RO tasl~s peFfoFFAed outsiEle 4.- 7--e the FAaiA G8Atrel F80FA EluFiAg aA eFAergeAGy aAd the rnsultaAt eperatioAal effeGts.
X 2.1.25 Ability to interpret reference materials, such 4.2 76 as qraphs, curves, tables, etc.
000026 (APE 26) Loss of Component X
AA 1.03 Ability to operate and / or monitor the 3.6 8
Cooling Water/ 8 following as they apply to the Loss of Component Cooling Water: SWS as a backup to the CCWS 000027 (APE 27) Pressurizer Pressure X
AK2.03 Knowledge of the interrelations between the 2.6 9
Control System Malfunction / 3 Pressurizer Pressure Control Malfunctions and the followina: Controllers and positioners 000029 (EPE 29) Anticipated Transient X
EK3.02 Knowledge of the reasons for the following 3.1 10 Without Scram / 1 responses as the apply to the ATWS: Starting a specific charging pump X
2.4.41 Knowledge of the emergency action level 4.6 77 thresholds and classifications.
000038 (EPE 38) Steam Generator Tube X
EA 1.36 Ability to operate and monitor the following 4.3 11 Rupture/ 3 as they apply to a SGTR: Cooldown of RCS to specified temperature
ES-401 3
Form ES-401-2 IIES-401 PWR Examination Outline Form E_ -
Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO/SRO)
E/APE #/Name I Safety Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR 000040 (APE 40; BW E05; CE E05; W E12) X X
AK1.04 Knowledge of the operational implications 3.2 12 Steam Line Rupture-Excessive Heat of the following concepts as they apply to Steam Transfer/ 4 Line Rupture: Nil ductility temperature X ~
le aJ3J3ly +86ARiGal aJ38GifiGatieRS feF a 4.--7
~
syslem--c X
2.4.21 Knowledge of the parameters and logic used 4.6 78 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.
000054 (APE 54; CE E06) Loss of Main X
AA2.02 Ability to determine and interpret the 4.1 13 Feedwater /4 following as they apply to the Loss of Main Feedwater (MFW): Differentiation between loss of all MFW and trip of one MFW pump 000055 (EPE 55) Station Blackout/ 6 X
EA2.01Ability to determine or interpret the following 3.7 79 as they apply to a Station Blackout: Existing valve I positioning on a loss of instrument air system.
000056 (APE 56) Loss of Offsite Power I 6 X
AK~.Q::l KRewleEl§e ef U1e eJ3eFatieRal imJ3liGatieRs
~ 44 ef U1e fellewiR§ GeRsei:its as ttiey~-te--bes&m G#site f2eweF: QefiRitieR ef sbll:lseeliR§: b!Se ef steam tal:lles te deteFmiRe it X
AK1.01 Knowledge of the operational implications 3.7 14 of the following concepts as they apply to Loss of Offsite Power: Principle of cooling by natural convection 000057 (APE 57) Loss of Vital AC X
AA2.20 Ability to determine and interpret the 3.6 15 Instrument Bus/ 6 following as they apply to the Loss of Vital AC Instrument Bus: Interlocks in effect on loss of ac vital electrical instrument bus that must be bypassed to restore normal equipment operation 000058 (APE 58) Loss of DC Power I 6 X
.l..1.44 Al:lilily le iRl8Ff)F8l seRIFel Feem iRdisatieRS te 4A w veFify ttie statbls aRd ef)eFatieR ef a system, aRd b!RdeFstaRd tiew ef)eFateF astieRs aRd diFestives a#est i:ilaRI aRd system seRdilieRs.
X 2.2.22 Knowledge of limiting conditions for 4.7 80 operations and safety limits.
000062 (APE 62) Loss of Nuclear Service X
AK3.01 Knowledge of the reasons for the following 3.2 16 Water/ 4 responses as they apply to the Loss of Nuclear Service Water: The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the nuclear service water coolers 000065 (APE 65) Loss of Instrument Air/ 8 X
AA 1.05 Ability to operate and / or monitor the 3.3 17 following as they apply to the Loss of Instrument Air: RPS
ES-401 4
Form ES-401-2 ES-401 PWR Examination Outline F_..,
Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO/SRO)
E/APE # / Name/ Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000077 (APE 77) Generator Voltage and X
AA 1.04 Ability to operate and/or monitor the 4.1 7
Electric Grid Disturbances / 6 following as they apply to Generator Voltage and Electric Grid Disturbances: Reactor controls X
~interpret the
~ g.,i fellewin§ as they apply te Generator Velta§e-aoo Electric Grid Disturbances: Operatin§ point en the
§eflerater capability curve X
AA2.09 Ability to determine and interpret the 4.3 81 following as they apply to Generator Voltage and Electric Grid Disturbances: Operational status of emergency diesel generators (W E04) LOCA Outside Containment I 3 (W E 11) Loss of Emergency Coolant Recirculation / 4 (BW E04; W E05) Inadequate Heat X
- l.4.49 Ability te perforrn without reference te Ml 4,8 Transfer-Loss of Secondary Heat Sink/ 4 procedures these actions that reEJuire I irnrnediate operation ef systern cernpenents and controls.
X 2.1.31 Ability to locate control room switches, 4.6 18 controls, and indications, and to determine that they correctly reflect the desired plant lineup.
KIA Cateaorv Totals:
3 1
4 4
4/2 2/4 Group Point Total:
18/6
ES-401 5
Form ES-401-2 ES-401 PWR Examination Outline For11, cc.:,- *u -~ II Emergency and Abnormal Plant Evolutions-Tier 1 /Group 2 (RO/SRO)
E/APE # / Name I Safety Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR 000001 (APE 1) Continuous Rod Withdrawal/ 1 X
AA2.04 Ability to determine and 4.2 19 interpret the following as they apply to the Continuous Rod Withdrawal : Reactor power and its trend 000003 (APE 3) Dropped Control Rod / 1 X
AA 1.01 Ability to operate and I or 2.9 20 monitor the following as they apply to the Dropped Control Rod: Demand position counter and pulse/analog converter 000005 (APE 5) Inoperable/Stuck Control Rod / 1 X
AK1.06 Knowledge of the 2.9 21 operational implications of the following concepts as they apply to Inoperable / Stuck Control Rod: Bases for power limit, for rod misaliqnment 000024 (APE 24) Emergency Boration / 1 X
2.4.3~ KRewleEl§e ef aRRlJRGialeF 4.-4
~
alarms, iREiiealieRs, er rnspeRse preGeSlJFeS.
X 2.4.40 Knowledge of events 4.1 82 related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system ooerator.
000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear X
AK2.01 Knowledge of the 2.7 22 Instrumentation/ 7 interrelations between the Loss of Source Range Nuclear Instrumentation and the following:
Power supplies, including proper switch positions 000033 (APE 33) Loss of Intermediate Range Nuclear X
AA2.Q3 Al3ilily le ElelermiRe aREi
~ ~
Instrumentation / 7 iRlerprnl !Re fellewiR§ as !Rey apply le !Re bass ef IRler-mBElia-le RaR§e NlJGlear IRSIFlJmeRlaliBlt IREiiealieR ef 131ewR flJse X
AA2.12 Ability to determine and 2.5 23 interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:
Maximum allowable channel disagreement 000036 (APE 36; BW/A08) Fuel-Handling Incidents/ 8 X
AA2.02 Ability to determine and 4.1 83 interpret the following as they apply to the Fuel Handling Incidents: Occurrence of a fuel handlina incident 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 17 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation I 8
ES-401 6
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions-Tier 1/Group 2 (RO/SRO)
E/APE #/Name I Safety Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR 000069 (APE 69; W E14) Loss of Containment Integrity/ 5 X
AK-4~4-Km=iwleEl§e ef !Re u
~
er:ieFalieRal imr:ilisalieRs ef !Re follewiR§ seRser:its as !Rey ar:ir:ily le bess ef GeRlaiRmeRI IRle§Fily:
Effesl ef r:iressurn eR leak rate X
AA2.02 Ability to determine and 3.9 24 interpret the following as they apply to the Loss of Containment Integrity: Verification of automatic and manual means of restoring inteoritv 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling I X
2',:l.de Al:Jility le aRaly2e !Re effest 4.-:lc 84 4
ef maiR!eRaRse astivities, susR as Ele§raEleEl r:iewer seurnes, eA
!Re status ef limiliR§ seRElilieRs for er:ieratieRs.
X 2.4.49 Ability to perform without 4.4 84 reference to procedures those actions that require immediate operation of system components and controls.
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 E 16) High Containment Radiation /9 (BW A01) Plant Runback / 1 X
AA2.1 Ability to determine and 3.7 85 interpret the following as they apply to the (Plant Runback):
Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
(BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip/ 4 (BW A05) Emergency Diesel Actuation / 6 X
AA 1.3 Ability to operate and I or 3.7 25 monitor the following as they apply to the (Emergency Diesel Actuation): Desired operating results during abnormal and emeroencv situations.
(BW A07) Flooding / 8 X
2.2.22 Knowledge of limiting 4.0 26 conditions for operations and safety limits.
(BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA Cooldown-Depressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures X
EK3.02 Knowledge of the reasons 3.2 27 for the following responses as they apply to the (EOP Rules):
Normal, abnormal and emergency operating procedures associated with (EOP Rules).
(CE A 11 **; W E08) RCS Overcooling-Pressurized Thermal Shock/ 4
ES-401 7
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emerqencv and Abnormal Plant Evolutions-Tier 1/Group 2 (RO/SRO)
E/APE # / Name I Safety Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR (CE A 16) Excess RCS Leakage/ 2 (CE E09) Functional Recovery (CE E 13*) Loss of Forced Circulation/LOOP/Blackout/ 4 IK/A Category Point Totals:
11 11 11 12 13/2 I 112 I Group Point Total:
I 19/4 I
ES-401 8
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S,stems-Tier 2/Group 1 (RO/SRO)
Svstem # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
KIA Topic(s)
IR 003 (SF4P RCP) Reactor Coolant X
K4.03 Knowledge of RCPS design feature(s) 2.5 28 Pump and/or interlock(s) which provide for the following: Adequate lubrication of the RCP 2.1.31 Ability to locate control room switches, X
controls, and indications, and to determine that 4.6 54 they correctly reflect the desired plant lineup.
004 (SF1; SF2 CVCS) Chemical and X
K2.07 Knowledge of bus power supplies lo the 'b+
29 Volume Control following~eal~
X K2.05 Knowledge of bus power supplies to the 2.7 29 following: MOVs X
2-c+.30 Ability to locate and operate 4-0 86 components, including-leGal-coo-trols.
X 2.2.40 Knowledge of SRO responsibilities in 4.5 86 emergency plan implementation.
005 (SF4P RHR) Residual Heat X
K6.03 Knowledge of the effect of a loss or 2.5 30 Removal malfunction on the following will have on the RHRS: RHR heat exchanger 006 (SF2; SF3 ECCS) Emergency X
A 1.09 Ability to predict and/or monitor changes 2.8 31 Core Cooling in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: Pump amperage, including start, normal and locked X
A4.02 Ability to manually operate and/or 4.0 51 monitor in the control room: Valves 007 (SF5 PRTS) Pressurizer X
K3.01 Knowledge of the effect that a loss or 3.3 32 Relief/Quench Tank malfunction of the PRTS will have on the following:
008 (SF8 CCW) Component Cooling X
A2.04 Ability to (a) predict the impacts of the 3.3 33 Water 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: PRMS alarm X
A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the 3.2 87 CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Hiqh/low CCW temperature 010 (SF3 PZR PCS) Pressurizer X
K1.08 Knowledge of the physical connections 3.2 34 Pressure Control and/or cause-effect relationships between the PZR PCS and the following systems: PZR LCS X
K5.02 Knowledge of the operational 2.6 49 implications of the following concepts as the apply to the PZR PCS: Constant enthalpy expansion through a valve
ES-401 9
Form ES-401-2 ES-401 PWR Examination Outline Fo
-~
Plant S stems-Tier 2/Group 1 (RO/SRO)
Svstem # I Name K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G*
KIA Topic(s)
IR 012 (SF7 RPS) Reactor Protection X
A2.G7-Aeilily te fat-f*E*HBI tt:ie impasts ef tt:ie
~ Je felle*NiA§J malflclAGlieAs eF epeFatieAS eA !Ae
~12~; 8A9 fe) 98589 eA !Aese pFe9iG!ieAS, lclS8 pF96891clF8S le GElFFeGI, 6ElAIF91, ElF FRili§Jale !Ae 6ElAS8Ejlcl8A68S ef !Aese malflclAGlieAS-Of eperntieAs: Less ef 96 seAlrnl pewm X
A2.02 Ability to (a) predict the impacts of the 3.6 35 following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of instrument power X
K1.02 Knowledge of the physical connections 3.4 50 and/or cause effect relationships between the RPS and the following systems: 125V de system 013 (SF2 ESFAS) Engineered X
K3.01 Knowledge of the effect that a loss or 4.4 36 Safety Features Actuation malfunction of the ESFAS will have on the following: Fuel 022 (SF5 CCS) Containment Cooling X
A3.01 Ability to monitor automatic operation of 4.1 37 the CCS, including: Initiation of safeguards mode of operation X
2.2.12 Knowledqe of surveillance orocedures.
4.1 88 025 /SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray X
K2.01 Knowledge of bus power supplies to the 3.4 38 followinq: Containment sprav pumps 039 (SF4S MSS) Main and Reheat X
A 1.05 Ability to predict and/or monitor changes 3.2 39 Steam in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: RCS T-ave X
A2.03 Ability to (a) predict the impacts of the 3.7 89 following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Indications and alarms for main steam and area radiation monitors (during SGTR)
I 059 (SF4S MFW) Main Feedwater X
K1.07 Knowledge of the physical connections 3.2 40 and/or cause effect relationships between the MFW and the following systems: JCS X
A3.03 Ability to monitor automatic operation of 2.5 53 the MFW, including: Feedwater pump suction flow pressure
ES-401 10 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S stems-Tier 2/Grouo 1 (RO/SRO)
Svstem # / Name K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G*
KIA Topic(s)
IR 061 (SF4S AFW)
X KA.~;3 KnewleEJ§e ef AF-VV. EJesi§n-feallifet&)
'b+
44 Auxiliary/Emergency Feedwater anEJ/eF inteFlesk(s) whish previEJe feF !Re fellewin§: lnitiatien ef seelin§ wateF ans lube eil X
K4.02 Knowledge of AFW design feature(s) 4.5 41 and/or interlock(s) which provide for the following: AFW automatic start upon loss of MFW pump, SIG level, blackout, or safety injection X
A3.02 Ability to monitor automatic operation of 4.0 52 the AFW, including: RCS cooldown during AFW ooerations 062 (SF6 ED AC) AC Electrical X
K4.03 Knowledge of ac distribution system 2.8 42 Distribution design feature(s) and/or interlock(s) which provide for the following: Interlocks between automatic bus transfer and breakers X
M.Q:3 Ability le pFeEiist anEJleF menileF shan§es a ea in paFameteFs (te pFevent exseeEJin§ EJesi§n liffiHSJ assesiateEJ with eperatin§ the as EJislFibutien system senlrels insluEiin§: e#est--eR ins!Fumentatien anEI sentrels ef switshin§ pe*.veFSU~
X A.4.03 Ability to manually operate and/or 2.8 55 monitor in the control room: Synchroscope, including an understanding of running and incominq voltaqes 063 (SF6 ED DC) DC Electrical X
A3. 01 Ability to monitor automatic operation of 2.7 43 Distribution the DC electrical system, including: Meters, annunciators, dials, recorders, and indicating lights 064 (SF6 EOG) Emergency Diesel X
K6.08 Knowledge of the effect of a loss or 3.2 44 Generator malfunction of the following will have on the ED/G system: Fuel oil storage tanks 073 (SF7 PRM) Process Radiation X
2.1.7 Ability to evaluate plant performance and 4.4 45 Monitoring make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
X K5.01 Knowledge of the operational 2.5 64 implications as they apply to concepts as they apply to the PRM system: Radiation theory, includinq sources, types, units, and effects 076 (SF4S SW) Service Water X
2.2.12 Knowledge of surveillance procedures.
3.7 46 078 (SF8 IAS) Instrument Air X
A4.01 Ability to manually operate and/or 3.1 47 monitor in the control room: Pressure gauges X
~--Km=,wlefl§B--ef-the bases in Teshnisal
~ 9G
~BaOOR&4ef~ senEJitiens fer 9fIBl"-atioos--aRG-safety--limit&
X 2.2.37 Ability to determine operability and/or 4.6 90 availability of safety related equipment.
ES-401 11 Form ES-401-2 ES-401 PWR Examination Outline Fv, -- - -
L-v-""tU 1-L.
Plant S stems-Tier 2/Group 1 (RO/SRO)
System # I Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
KIA Tooic(s)
IR 103 (SF5 CNT) Containment X
A2.03 Ability to (a) predict the impacts of the 48 following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation 053 (SF1; SF4P ICS*) Integrated Control KIA Category Point Totals:
3 2
2 3
2 2
2 3/ 4 3
3/ Group Point Total:
29/5 2
3
ES-401 12 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S stems-Tier 2/Group 2 (RO/SRO)
System # I Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
KIA Topic(s)
IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer X
A<!.Qe Asility to (a) pFedist tile ifflpasts of tile J-c9 94 Level Control fellewiFl§ FflalftmstiORS OF opeFatiORS OR tile PZR bGS; aRd (13) eased oR !nose pFedistioRs, lJSe prneedlJFes to GOFFeGt, GOR!FOI, OF Ffliti§ate the-GORSeE11cJeRGes of !nose FflalfllRGtiORS--ef operntioRs:: IRadveFleRl PZR spFay aGtllatioR X
A2.03 Ability to (a) predict the impacts of the 3.9 91 following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of PZR level 014 (SF1 RPI) Rod Position X
K3.02 Knowledge of the effect that a loss or 2.5 56 Indication malfunction of the RPIS will have on the followino: Plant computer 015 (SF7 NI) Nuclear X
K6.01 Knowledge of the effect of a loss or 2.9 57 Instrumentation malfunction on the following will have on the NIS: Sensors, detectors, and indicators 016 (SF7 NNI) Nonnuclear X
K4.03 Knowledge of NNIS design feature(s) 2.8 58 Instrumentation and/or interlock(s) which provide for the following: Input to control systems 017 (SF7 ITM) In-Core Temperature X
K4.01 Knowledge of ITM system design 3.4 61 Monitor feature(s) and/or interlock(s) which provide for the followino: Input to subcooling monitors 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen X
A4-,0~ Aeility to pFediG! aRdfOF FflORitof GRafl§eS J-4 w Recombiner and Purge Control iR parnffleteF (to pFeveAI e*eeediR§--GBSi§R liffli!s) assoeiated witn operntiR§ tile HRPS
..J:--. u
.J __ *-
029 (SF8 CPS) Containment Purge X
A4.01 Ability to manually operate and/or 2.5 60 monitor in the control room: Containment I puroe flow rate 033 (SF8 SFPCS) Spent Fuel Pool X
A 1.01 Ability to predict and/or monitor changes 2.7 59 Cooling in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including: Spent fuel pool water level 034 (SF8 FHS) Fuel-Handling Eauipment 035 (SF 4P SG) Steam Generator 041 (SF4S SOS) Steam X
K4.1e KAowled§e of SOS desi§R feat1cJFe(s)
&.-9 G Dump/Turbine Bypass Control aRd/or iRterloek(s)--wfliBl:l prnvide foF tile 1,_,,_ :--* I -
':-'I"
~
045 (SF 4S MTG) Main Turbine X
A 1.05 Ability to predict and/or monitor changes 3.8 62 Generator in parameters (to prevent exceeding design limits) associated with operating the MT/G system controls including: Expected response of primary plant parameters (temperature and pressure) following T/G trip 055 (SF4S CARS) Condenser Air X 2.4.8 Knowledge of how abnormal operating 4.5 92 Removal I procedures are used in conjunction with EOPs.
ES-401 13 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S,stems-Tier 2/Group 2 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
KIA Topic(s)
IR 056 (SF4S CDS) Condensate X
K1.03 Knowledge of the physical connections 2.6 63 and/or cause-effect relationships between the Condensate System and the following systems: MFW 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas X
.A.'2.071>,eility-te (a) pFeElict !Re impacts ef !Re
~ 9J Disposal fellewiR§ malflclRclieRs eF epeFatieRs eR-#IB Waste Gas Dispesal System~~
e!Hflese pFeElictieRs,-usei*"-eseEiu-Fe&-te Geffe61,-BORtfBl, eF miti§ale !Re ceRSeEjlcleRces ef IRese malflclRctieRs eF eperntieRs: Less-Of meteemle§ical tewm X
A2.02 Ability to (a) predict the impacts of the 3.6 93 following malfunctions or operations on the Waste Gas Disposal System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Use of waste gas release monitors, radiation, gas flow rate, and totalizer 072 (SF? ARM) Area Radiation X
K5.01 KRewleEl§e ef !Re epeFatieRal
&.+
e4 Monitoring imj:w6atieRs-4the--felkw,iR§ ceRcepts as !Rey apply le !Re ARM system: RaEliatiGfl--theBf.y, i-RGltle: - -
,......,,.I
.LL 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection X
K6. Knowledge of the effect of a loss or 2.6 65 malfunction on the Fire Protection System following will have on the : Fire, smoke, and heat detectors 050 (SF 9 CRV*) Control Room Ventilation Jf-
=-- ry Point Totals:
1 0
1 2
0 2
2 Of 0 1
Of Group Point Total:
9/3 2
1
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility: TMI Date of Exam: June 2019 Category KIA#
Topic RO SRO-only IR IR 2.1.34 Knowledoe of primary and secondary plant chemistry limits.
2.7 66 2.1.36 Knowledoe of procedures and limitations involved in core alterations.
3.0 67 2.1.43 Ability to use procedures to determine the effects on reactivity of 4.1 68
- 1. Conduct of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.
Operations 2.1.5 Ability to use procedures related to shift staffing, such as minimum 3.9 94 crew complement, overtime limitations, etc.
2.1.8 Ability to coordinate personnel activities outside the control room.
4.1 95 Subtotal 3
2
~
Ability to determine operability and/or availability of safety related
~ e9 equipment.
2.2.17 Knowledge of the process for managing maintenance activities 2.6 69 during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.
~
Ability to r0Go§ni2,o system parameters U:iat are entry level Gonditions
~ +G
- 2. Equipment for TeGhniGal SpeGifiGations.
Control 2.2.13 Knowledoe of taoaing and clearance procedures.
4.1 70 2.2.14 Knowledge of the process for controlling equipment configuration or 3.9 71 status.
2.2.6 Knowledqe of the process for makinq chanqes to procedures.
3.6 96 2.2.19 KnowledQe of maintenance work order requirements.
3.4 97 Subtotal 3
2 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation 2.9 72 monitors and alanns, portable survey instruments, personnel monitorinQ equipment, etc.
- 3. Radiation 2.3.12 Knowledge of radiological safety principles pertaining to licensed 3.7 98 Control operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, alionino filters, etc.
Subtotal 1
1 2.4.3 Ability to identify post-accident instrumentation.
3.7 73 2.4.50 Ability to verify system alann setpoints and operate controls identified 4.2 74 in I the alann response manual.
2.4.22 Knowledge of the bases for prioritizing safety functions during 3.6 75
- 4. Emergency abnonnal/emergency operations.
Procedures/
2.4.37 Knowledge of the lines of authority during implementation of the 4.1 99 Plan emerQency plan.
2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss 4.2 10 of coolant accident or loss of residual heat removal) mitigation 0
strateqies.
Subtotal 3
2 Tier 3 Point Total 10 7
ES-401 Record of Rejected K/As Form ES-401-4 Tier/
Randomly Reason for Rejection Group Selected K/ A 1/ 1 008 / AA2.25 Q2: Scenario #4 overlap with failed open PORV. Replaced with 008 AA2.04.
1/ 1 025 I AK3.0l Q7: System/AOP is oversampled with two SRO questions. Replace with 077 AAl.04 1 /1 056 I AK 1.03 Q 14: Calculation of SCM does not change with or without off site power based on safety grade instrument usage. No requirement for the use of steam tables exists. Cannot make an operationally valid question to meet the KA. Replaced with 056 AKl.01 1 /1 BW E04 I 2.4.49 Q18: No immediate actions are required on a Lack of Heat Transfer.
Cannot make an operationally valid question to meet the KA. Replaced with KA BW E04 2.1.31.
1/2 033 I AA2.03 Q23: No fuse exists in the Intermediate Range Nuclear Instrument strings.
Cannot make technically accurate question and meet the KA. Replaced with 033 AA2.12.
1/2 069 I AKl.01 Q24: Effects of pressure on leak rate is covered in generic fundamentals.
Replaced with KA 069 AA2.02.
2 /1 004 I K2.07 Q29: Power supply to heat tracing is minutia. Replaced with 004 KA 2.05 2 /1 012 / A2.07 Q35: No DC control power exists to the Reactor Protection System.
Replaced with 012 A2.02.
2 I l 061 /K4.13 Q41: Cooling water and lube oil to the EFW pumps at Three Mile Island are not active systems. There is no start signal to initiate cooling water or lube oil to the pumps. Replaced with 061 K4.06.
2 I I 061 / K4.06 Q41: Emergency Feedwater at Three Mile Island does not have startup permissives that other plants may have. Replaced with 061 K4.02.
2/2 062 I Al.03 Q55: Topic is oversampled. Changing AC power supplies is covered in question #38 (transfer to an Emergency Diesel) and #42 (transfer to another bus). The only other Al KIA on Emergency Diesel Generator limits is covered in question 25. Replaced with 062 A4.03.
2/2 028 / Al.01 Q59: Hydrogen Recombiner and Purge System have no design basis function at Three Mile Island. Not appropriate for an Initial License Exam. Replace with 033 Al.01 2/2 041 / K4.16 Q61: The Steam Dump/Turbine Bypass system does not have a low main steam pressure function at Three Mile Island. The system protects against high main steam pressure. Other aspects of the system are tested in various questions. Topic oversampled. Replaced with O 17 K4.01 2/2 072 I K5.0l Q64: Replaced KIA based on discretion of Chief Examiner (Presby).
Replaced with 073 K5.01 3
2.2.37 Q69: Operability determinations are SRO only knowledge. Replaced with 2.2.17.
3 2.2.42 Q70: Could not make a Tier 3 KA with 3 plausible non-system topics for this KA. Replaced with 2.2.13
ES-401 Record of Rejected K/As Form ES-401-4 3
2.4.50 Q74: Could not make a question with 3 plausible distractors for this topic.
Replaced with 2.1.20 3
2.4.22 Q75: Safety functions are SRO only knowledge. Replaced with 2.4.2.
3 2.4.2 Q75: Could not make a Tier 3 question to meet this KA. Replaced with 2.4.14 1 / 1 025 I 2.4.34 Q76: No RO task outside of the control room exists that is exclusively SRO knowledge on a loss ofDHR. Replaced with 025 2.1.25.
l/ l 040 I 2.4.40 Q78: Could not make a question with operational validity that applies system technical specifications during a steam line rupture. Replaced with 040 KA 2.4.21 l/ l 058 / 2.2.44 Q80: Could not craft a question without trivial knowledge required to determine correct answer. Replaced with KA 058 2.2.22 l/ l 077 / AA2.0l Q8 l: Could not make a question with appropriate level of difficulty for an initial license test. Replaced with 077 AA2.09.
l/2 024 I 2.4.31 Q82: Could not make an SRO-only question based on this generic KIA.
Replaced with 024 KA 2.4.30 1/2 074 / 2.2.36 Q84: Could not make a question with operation validity that requires analyzing maintenance activities and Inadequate Core Cooling. Replaced with 074 2.4.49.
2 I l 078 / 2.2.25 Q90: No technical specification exists for the Instrument Air System, therefore knowledge of the basis of a Technical Specification is not testable. Replaced with KIA 078 2.2.37 2/2 004 I 2.l.20 Q86: Could not make a question of adequate difficulty for an initial license exam. Replaced with 004 2.2.40 2/2 011 / A2.06 Q91: Overlap with Scenario 2 event. Replaced with O 11 A2.03 2/2 071 / A2.07 Q93: Could not find an adequate tie between a waste gas release and a loss of the meteorological tower. Replaced with 071 A2.02