ML21237A507

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Written - Draft Examination and Operating Test Outlines (Folder 2)
ML21237A507
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 03/26/2021
From:
Exelon Generation Co
To: Todd Fish
Operations Branch I
Shared Package
ML20259A337 List:
References
EPID L-2021-OLL-002
Download: ML21237A507 (26)
v • d • e


Text

Appendix D Scenario Outline ES-D-1

Simulation Facility Peach Bottom Scenario No. #1 Op Test No. 2021 NRG

Examiners Operator ______ CRS (SRO)

______ URO (ATC)

______ PRO (BOP)

Scenario The scenario begins with the reactor at approximately 4% power during a reactor Summary startup.

Following shift turnover, the PRO will secure the Mechanical Vacuum Pump in accordance with GP-2, Reactor Startup, and SO 8.1.A, Off-Gas System Startup for Normal Operations.

The URO will continue the startup by raising reactor power to > 4% by withdrawing control rods in accordance with the approved startup sequence until 2 main turbine bypass valves are open with reactor pressure at 910 psig using procedure GP-2-2, "Normal Plant Startup".

A control rod will drift into the core. The URO will respond per ON-121, Drifting Control Rod, and fully insert the control rod. CRS will declare the control rod inoperable per T.S. 3.1.3.C.

The PRO will then respond to a failed drywell pressure instrument and half scram.

The crew will respond in accordance with the alarm response procedures. The crew will bypass the failed instrument and the URO and PRO will reset the half scram. The SRO will declare the instrument inoperable and will enter/apply Technical Specifications 3.3.1.1.

The PRO will then respond to the inadvertent start of a core spray subsystem. The PRO will secure the core spray system and the SRO will declare the core spray subsystem inoperable. The SRO will enter/apply Technical Specifications 3.5.1.A.

Then, the Main Turbine pressure regulator will fail, indicating a rising reactor pressure and causing the Turbine Control Valves to open. The crew will be required to scram the reactor prior to pressure reaching 850 psig to prevent the MSIV's from closing. The crew will enter T-101, RPV Control, and control plant parameters. Pressure will continue to lower; requiring the crew to close the MSIV's to prevent exceeding the cooldown rate limit. This will require the crew to transition level control to HPCI and RCIC.

Once the MSIV's are closed, a water leak will occur in the containment. HPCI will fail to start, and RCIC will not be able to maintain RPV water level. When RPV water level cannot be maintained above -172", the crew will perform an Emergency Slowdown, IAWT-112.

The scenario may be terminated following the Emergency Slowdown.

2021 NRG Scenario #1 D-1 Rev 0

I Appendix D Scenario Outline ES-D-1

Initial IC-71 Approximately 3% power Conditions

Turnover

  • Unit 2 startup is in progress.
  • Reactor Power is approximately 3% with direction to continue to raise Reactor power with control rod withdrawal using GP-2-2.

Critical Critical Task 1: Close the MSIVs prior to exceeding 450 psig.

Tasks Critical Task 2: Inhibit ADS before an automatic depressurization occurs.

Critical Task 3: Perform an Emergency Slowdown when RPV Level cannot be restored and maintained above -172".

2021 NRC Scenario #1 D-1 Rev 0

, I i

Appendix D Scenario Outline ES-D-1

Event Malfunction Event Event

No. No. Type* Description

1 See Scenario N PRO Secure the Mechanical Vacuum Pump Guide CRS

2 See Scenario R URO. Raise reactor power by withdrawing control rods.

Guide CRS

3 See Scenario C URO Control Rod Drift/URO inserts rod Guide TS/C CRS

4 See Scenario TS CRS DW Pressure Instrument Failure Guide I

. I I

5 See Scenario C PRO "A" Core Spray loop spuriously starts / PRO secures pump~

Guide CITS CRS and CRS enters Tech Specs. I

6 See Scenario C ALL Pressure Regulator Failure/Scram/T-101 Entry Guide

7 See Scenario M ALL Recirc Water Leak in PC Guide

8 See Scenario C URO HPCI Fails to Start Guide CRS

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec

2021 NRC Scenario #1 D-1 Rev 0

'i I

Appendix D Scenario Outline ES-D-1

Simulation Facility Peach Bottom Scenario No. #2 Op Test No. 2021 NRC

Examiners Operator _______ CRS (SRO)

______ URO (ATC)

PRO (BOP)

Scenario The scenario begins with the reactor at 100% power with no equipment out of Summary service.

After taking the shift, the PRO will place a cooling tower in service IAW SO 288.1.A, Cooling Tower Startup for Normal Operation.

Next, the A CRD pump will trip. The URO will enter ON-107, Loss of CRD Regulating Function and start the B CRD pump.

Following the CRD event, a high vibration condition for the main turbine will occur.

The URO will lower reactor power with recirculation flow by lowering recirculation pump speed of both recirculation pumps, and by inserting control rods IAW GP 2. Turbine vibrations will stop once reactor power has been lowered by 10%.

After turbine vibrations have stopped, the A SRV will fail open. The crew will enter and execute OT-114, Inadvertent Opening of a Relief Valve, and close the SRV. The SRV will close when the 885 psig button is depressed. The SRO will declare the SRV inoperable and enter/apply Technical Specification 3.5.1.E.

After the SRV is closed, the E-4 Diesel Generator will inadvertently start and one minute later 005 F-1 "E-4 Diesel Gen Differential and Ground" alarm will annunciate. The crew will take action in the alarm response to shutdown the E-4 diesel generator and place in Pull-to-Lock. The CRS will apply Technical Specifications for an inoperable Diesel Generator.

Once the DG has been secured and the tech spec determination has been made, a loss of TBCCW will occur. The PRO will attempt to start the standby pump, but it will not start. The URO will reduce reactor power per GP-9 to lower generator loading to less than 20 Kamps.

While reactor power is being reduced, a loss of the #2 bus will occur. The crew will scram the reactor due to the loss of 2 Condensate pumps. The crew will enter T-101, RPV Control, and control RPVwater level.

When the reactor is scrammed, a steam leak from RWCU will occur. Cleanup will fail to automatically and manually isolate. The crew will enter T-103, Secondary Containment Control. Secondary containment temperatures will continue to rise and the crew will perform an RPV Slowdown per T-112 when 2 areas exceed the maximum safe temperatures.

The scenario may be terminated when the RPV Slowdown is in progress.

2021 NRC Scenario #2 D-1 Rev 0 Appendix D Scenario Outline ES-D-1

Initial IC-14, 100% power Conditions

Turnover

  • Reactor power is 100% power.

Critical Critical Task #1: Scram the reactor or restore charging header pressure Tasks above 940 psig within 20 minutes of charging header pressure lowering below 940 psig and 2 or more accumulator alarms in.

Critical Task #2: Perform an RPV Slowdown when the second Reactor Building area temperature exceeds an Action Level.

2021 NRC Scenario #2 D-1 Rev 0 I

Appendix D Scenario Outline ES-D-1 ! I I

I I

I Event Malfunction Event Event

No. No. Type* Description

1 See Scenario Guide N PRO Perform SO 28B.1.A to start a cooling tower CRS

2 See Scenario Guide C URO CRD Pump Trip CRS

3 See Scenario Guide R URO Main turbine high vibrations require lowering power to addr~ss C CRS the vibrations I I

4 See Scenario Guide C ALL ADS Valve Fails Open. CRS enters Tech Specs.

TS CRS

I 5 See Scenario Guide C PRO I Shutdown E-4 diesel generator following inadvertent start. ORS TS/C CRS enters Tech Specs.

6 See Scenario Guide C ALL Loss of TBCCW

7 See Scenario Guide C ALL Loss of #2 Bus/T-101 Entry

8 See Scenario Guide M ALL RWCU Steam Leak/T-103 Entry

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec

2021 NRC Scenario #2 D-1 Rev 0 I

I I

I Appendix D Scenario Outline ES-D-1

Simulation Facility Peach Bottom Scenario No. #4 Op Test No. 2021 NRC

Examiners Operator _______ CRS (SRO)

______ URO (ATC)

______ PRO (BOP)

Scenario The scenario begins with the reactor at approximately 97% power.

Summary After taking the shift, the URO will withdraw control rod 58-31 to position 48.

When verifying control rod coupling, the control rod will over-travel. The URO will enter ON-105, "Control Rod Uncoupled" and will attempt to re-couple the control rod. Recoupling will fail and the CRS will declare the control rod inoperable per T.S. 3.1.3.C.

Following the control rod movement, the URO will raise reactor power with recirculation flow back to 100% power.

With the reactor at full power, the PRO will then perform a swap of the RBCCW pumps IAW SO 35.6.A, Reactor Building Closed Cooling Water System Startup and Normal Operations.

Next, a loss of DC power to one of two (in series) 'B' RPS MG set output breakers will occur. A subsequent operability concern with the other (in series) 'B' RPS MG set output breaker will require the Crew to transfer the 'B' RPS bus to the alternate power supply IAW SO 60F.6.A-2 "Transferring Reactor Protection System Power Supplies". The resultant half scram and containment isolations will have to be reset by the PRO, IAW GP-11.E "Reset RPS - Scram Reset" and GP

8. D "Group 1, 2, 3 Outboard Half isolation". The CRS will declare one electric power monitoring assembly inoperable and will enter/apply Technical Specifications 3.3,8.2.

Following the transfer of RPS B to alternate, a loss of RBCCW will occur. The crew will isolate RWCU and reduce power per GP-9 to control component temperatures.

Next, a loss of both recirc pumps will occur. The crew will immediately scram the reactor. An electrical A TWS will occur and the crew will enter T-101 and T-117 to lower RPV water level to control reactor power.

The first SBLC Pump will trip, requiring the URO to start the alternate SLC pump.

The URO will then insert control rods per T-220. After the first control rod is inserted, a trip of the running CRD pump will occur. The URO will start the alternate CRD pump and continue inserting control rods.

The scenario will be terminated when RPV water level is being maintained above TAF and control rods are being inserted.

2020 NRC Scenario #4 D-1 Rev 1 Appendix D Scenario Outline ES-D-1

Initial IC-14, 100% power (Reactor power is lowered to 97% power)

Conditions

Turnover

  • Reactor power is approximately 97%. Reactor power was lowered to allow scram valve repairs on control rod 58-31. The repairs are complete, the control rod has been scram-timed, and is at position 00. The URO may withdraw the control back to its target position of 48.
  • Following the control rod withdrawal, the URO may raise reactor power back to 100% with recirculation flow.

Critical Critical Task #1: Attempt to shutdown the reactor by performing one of the Tasks following:

  • T-213, Scram Soleno,id De-energization
  • Injecting Standby Liquid before Torus Temperature exceeds 110°F.

Critical Task #2: Perform T-240, Termination and Prevention of Injection into the RPV to minimize thermal-hydraulic instabilities (THI) until RPV water level is below -60".

2020 NRC Scenario #4 D-1 Rev 1 Appendix D Scenario Outline ES-D-1

Event Malfunction Event Event I

No. No. Type* Description

1 See Scenario Guide C URO Control rod becomes uncoupled / attempt to recouple TS/C CRS using ON-105

2 See Scenario Guide R URO Raise reactor power with recirculation flow.

CRS

3 See Scenario Guide N PRO Swap RBCCW Pumps CRS

I 4 C PRO Loss of power to RPS breaker and transfer of RPS to See Scenario Guide TS/C CRS alternate supply. CRS enter/apply Technical Specifications. I

i 5 C ALL Loss of RBCCW See Scenario Guide

6 See Scenario Guide C URO Both Recirc Pumps Trip CRS

I i

7 See Scenario Guide M ALL ATWS I I

8 See Scenario Guide C URO SBLC Pump Trip i CRS

9 See Scenario Guide C URO CRD Pump Trip CRS I

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec

2020 NRC Scenario #4 D-1 Rev 1 ES-301 Administrative Topics Outline Form ES-301-1

Facility: Peach Bottom Atomic Power Station Date of Examination: NRC

Examination Level: RO [X] SRO Operating Test Number:

Administrative Topic (see Note) Type Describe activity to be performed Code*

G2.1.32 (3.8), Evaluation of High CRD Conduct of Operations R/D Temperatures on Control Rod Scram Time

(PLOR-348C)

R/M G2.1.5 (2.9), Application of Work Hour Rules

Conduct of Operations (PLOR-391C)

G2.2.6 (3.0), Initiating a Temporary Procedure

Equipment Control R/D Change (PLOR-245C)

Not Required Radiation Control

G2.4.39 (3.9), Perform State/Local Notifications Emergency Plan S/N for a Declared Emergency (PLOR-418C)

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) 2 (N)ew or (M)odified from bank (;:: 1) 2 (P)revious 2 exams (:;; 1, randomly selected) 1 ES-301 Administrative Topics Outline Form ES-301-1

Facility: Peach Bottom Atomic Power Station Date of Examination: NRC Examination Level: RO SRO ~ Operating Test Number:

Administrative Topic (see Note) Type Describe activity to be performed Code*

G2.1.25 (4.2), Perform SRO Review of Conduct of Operations RID Completed Surveillance (PLOR-393C)

G2.1. 7 ( 4. 7), Resolution of Thermal Limit Conduct of Operations R/M Violation (PLOR-218C)

G2.2.6 (3.6), Approve a Partial Procedure

Equipment Control R/N (PLOR-416C)

G2.3.14 (3.8), Review and Authorize Issuance

Radiation Control R/N of Thyroid Blocking Agent (Kl} (PLOR-41 ?C)

G2.4.41 (4.6), EAL Classification and Emergency Plan ~ S/M State/Local Notifications for SAE - Control

Room Evacuation (PLOR-180C)

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) 1 (N)ew or (M)odified from bank(.?: 1) 4 (P)revious 2 exams (S 1, randomly selected) 0 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

Facility: PBAPS Date of Examination: NRC Exam Level: RO IZl SRO-I SRO-U Operating Test Number:

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. 295037 EA1.01 4.6/4.6, Insert Control Rods using Individual A,N, S 1 Scram Test Switches (PLOR-412CA)
b. 295001 A4.02 3.9/3.7, Start the "C" Reactor Feedwater Pump with Vessel Level Control Through AO-8091 (PLOR-M, L, S 2 012C)

C. 239001 A4.01 4.2/4.0, Perform a Slow Closure and N,S 3 Restoration of a Main Steam Isolation Valve (PLOR-413C)

d. 202001 A2.05 3.8/4.0, Reset Recirc Pump Speed Hold A,N,S 4 (Alternate Path - Recirc Speed Oscillates) (PLOR-414CA)
e. 295024 EA 1.11 4.2/4.2, Spray the Containment using HPSW M, EN, S 5 per T-205 (PLOR-079C)
f. 262001 A4.04 3.6, Transfer House Loads to the Unit D,S 6 Auxiliary Transformer (PLOR-039C)
g. 400000 A2.01 3.3/3.4, Placing the Standby TBCCW Pump In-Service (Alternate Path - Standby Pump Trips) (PLOR-A,N,S 8 415CA)
h. 261000 A2.05 3.0/3.1, Manually Place SBGT on the Equipment Cell Exhaust (Alternate Path -1sr Fan Fails to A,EN,D,S 9 Align) (PLOR-265CA)

In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U.

i. 295029 EA2.01 3.9/3.9, Lowering Torus Level using the N,E,R 5 Torus Water Filter Pump (PLOR-409P)
j. 239002 A2.03 4.1/4.2, Remove Fuses per OT-114 for Stuck M,E,R 3 Open SRV (PLOR-191P)
k. 201001 A2.06 2.9/2.9, Loss of CRD Regulating Function D,E, R 1 (Outside Control Room Actions) (PLOR-073P)
  • 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

(A)lternate path 4-6/4-6 /2-3 4 (C)ontrol room (D)irect from bank S 9/S 8/S4 3 (E)mergency or abnormal in-plant ~1k1k1 3 (EN)gineered safety feature ~ 1/~ 1/~ 1 (control room system) 2 (L)ow-Power/Shutdown ~1k1k1 1 (N)ew or (M)odified from bank including 1 (A) ~ 2/~ 2/~ 1 8(3)

(P)revious 2 exams s 3/s 3/S 2 (randomly selected) 0 (R)CA ~1,~1,~1 3 (S)imulator ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

Facility: PBAPS Date of Examination: NRC Exam Level: RO SRO-I ~ SRO-U Operating Test Number:

Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U.

System/JPM Title Type Code* Safety

I Function

a. 295037 EA 1.01 4.6/4.6, Insert Control Rods using Individual A,N, S 1 Scram Test Switches (PLOR-412CA)
b. 295001 A4.02 3.9/3.7, Start the "C" Reactor Feedwater Pump with Vessel Level Control Through AO-8091 (PLOR-M, L, S 2 012C)
c. 239001 A4.01 4.2/4.0, Perform a Slow Closure and N,S 3 Restoration of a Main Steam Isolation Valve (PLOR-413C)
d. 202001 A2.05 3.8/4.0, Reset Recirc Pump Speed Hold A,N,S 4 (Alternate Path - Recirc Speed Oscillates) (PLOR-414CA)
e. 295024 EA 1.11 4.2/4.2, Spray the Containment using HPSW M, EN, S 5 per T-205 (PLOR-079C)

f.

g. 400000 A2.01 3.3/3.4, Placing the Standby TBCCW Pump In-Service (Alternate Path - Standby Pump Trips) (PLOR-A,N,S 8 415CA)
h. 261000 A2.05 3.0/3.1, Manually Place SBGT on the Equipment Cell Exhaust (Alternate Path - 1sT Fan Fails to A,EN,D,S 9 Align) (PLOR-265CA)

In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U.

i. 295029 EA2.01 3.9/3.9, Lowering Torus Level using the N,E, R 5 Torus Water Filter Pump (PLOR-409P)
j. 239002 A2.03 4.1/4.2, Remove Fuses per OT-114 for Stuck M,E,R 3 Open SRV (PLOR-191 P)
k. 201001 A2.06 2.9/2.9, Los.s of CRD Regulating Function D,E,R 1 (Outside Control Room Actions) (PLOR-073P)
  • 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 /SRO-1/SRO-U ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

(A)lternate path 4-6/4-6 /2-3 4 (C)ontrol room (D)irect from bank S 9/S 8/S4 2 (E)mergency or abnormal in-plant ~1k1k1 3 (EN)gineered safety feature ~ 1/~ 1/~ 1 (control room system) 2 (L)ow-Power/Shutdown ~1k1k1 1 (N)ew or (M)odified from bank including 1 (A) ~ 2/~ 2/~ 1 8(3)

(P)revious 2 exams s 3/S 3/s 2 (randomly selected) O (R)CA ~1,~1,~1 3 (S)imulator ES-401 BWR Examination Outline FORM E$-401-1

Facility Name: Date of Exam: I I I

RO KIA Category Points SRO-Only Points I I Tier Group K K K K K K A A A A G G* i 1 2 3 4 5 6 1 2 3 4

  • Total A2 Tptal
1. 1 4 3 4 3 3 3 20 4 3 \\7 Emergency & I Abnormal 2 1 1 1 N/A 2 1 N/A 1 7 2 1 13 '

Plant. I I Evolutions Tier Totals 5 4 5 5 4 4 27 6 4 i10 I

1 3 2 3 2 2 2 3 3 2 2 2 26 3 2 Is 2.

Plant 2 1 1 1 1 2 1 2 1 0 1 1 12 0 2 1 !3 Systems I Tier Totals 4 3 4 3 4 3 5 4 2 3 3 38 5 3 I Is

3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 Categories 10 7 3 3 2 2 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 a~d 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). (One Tier 3 Radiation Control KIA is allowed if the KIA is replaced by a KIA froril I

another Tier 3 Category). I I

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 revision{ The final RO exam must total 75 points and the SRO-only exam must total 25 points. !
3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do' not I

apply at the facility should be deleted and justified; operationally important, site-specific systems that are no,t included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimiiiation of inappropriate KIA statements. I

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the gr~>up before selecting a second topic for any system or evolution. i
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. i 7.* The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics mu:st be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIAs. I I
8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance rati~gs (I Rs) for the applicable license level, and the point totals(#) for each system and category. Enter the group land tier totals for each category in the table above; if fuel handling equipment is sampled in other than CategorylA2 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 KIA catalog, and enter the KIA numbers, descriptions, I Rs, and point totals (#) on Form ES-401-3. Limit SRO selections to Kl As that are linked to 10 CFR 55.43.

G* Generic KIAs ES-401 2 Form ESi401-1

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO)

E/APE # / Name I Safety Function K K K A KIA Topic(s) IR # 1 2 3 1

295001 Partial or Complete Loss of Forced 0 Ability to operate and/or monitor the following as they apply to Partial or Complete 3.3 Core Flow Circulation / 1 & 4 2 Loss of Forced Core Flow Circulation: RPS

295003 Partial or Complete Loss of AC / 6 3 Partial 0 Knowledge of the operational implications of the following concepts as they apply to or Complete Loss of AC: Under voltage/degraded voltage effects on electrical 2.9

loads

295004 Partial or Total Loss of DC Pwr / 6 0 Knowledge of the interrelations between Partial or Total Loss of DC Pwr and the 3.1 following: Battery charger

295005 Main Turbine Generator Trip / 3 2.8 ! 0 !11! Knowledge of the reasons for the following responses as they apply to Main Turbine I 3 Generator Trip: Feedwater temperature decrease i

295006 SCRAM / 1 Ability to use plant computers to evaluate system or component status. 3.9

295016 Control Room Abandonment I 7 0 Ability to operate and/or monitor the following as they apply to Control Room 4.2 7 Abandonment: Control roomnocal control transfer mechanisms

295018 Partial or Total Loss of CCW / 8 0 Knowledge of the interrelations between Partial or Total Loss of CCW and the 3.3 1 following: System loads

295019 Partial or Total Loss of Inst. Air/ 8 0 Knowledge of the reasons for the following responses as they apply to Partial or Total 3.2 3 Loss of Inst. Air: Service air isolations: Plant-Specific

295021 Loss of Shutdown Cooling/ 4 0 Knowledge of the reasons for the following responses as they apply to Loss of 3.3 1 Shutdown Cooling: Raising reactor water level

295023 Refueling Ace / 8 0 Knowledge of the interrelations between Refueling Accidents and the following: 3.4 6 Containment ventilation: Mark-Ill

295024 High Drywell Pressure / 5 Ability to operate and/or monitor the following as they apply to High Drywall Pressure: 3.4 4 Drywall ventilation system

295025 High Reactor Pressure / 3 Ability to determine and/or interpret the following as they apply to High Reactor 4.3 Pressure: Reactor pressure

295026 Suppression Pool High Water 0 Knowledge of the operational implications of the following concepts as they apply to 3.5 Temp./ 5 2 Suppression Pool High Water Temp.: Steam condensation

295027 High Containment Temperature/ 5 0

295028 High Drywell Temperature/ 5 Ability to determine and/or interpret the following as they apply to High Drywall 3.7 Temperature: Reactor water level

295030 Low Suppression Pool Wtr Lvl / 5 Knowledge of the operational implications of EOP warnings, cautions, and notes. 3.8

295031 Reactor Low Water Level/ 2 0 Knowledge of the reasons for the following responses as they apply to Reactor Low 4.4 2 Water Level: Core coverage

295037 SCRAM Condition Present and 0 Knowledge of the operational implications of the following concepts as they apply to Reactor Power Above APRM Downscale or 5 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown: 3.4 Unknown/ 1 Cold shutdown boron weight: Plant-Specific

295038 High Off-site Release Rate I 9 0 Knowledge of the operational implications of the following concepts as they apply to 4.2 2 High Off-site Release Rate: Protection of the general public

600000 Plant Fire On Site I 8 Ability to determine and/or interpret the following as they apply to Plant Fire On Site: 2.8 Damper position

700000 Generator Voltage and Electric Grid Ability to analyze the effect of maintenance activities, such as degraded power 3.1 Disturbances / 6 sources, on the status of limiting conditions for operations.

KIA Category Totals: 4 3 4 3 Group Point Total: 20

ES-401, Page 34 of 50 ES-401 3 Form ES}401-1

ES-401 BWR Examination Outline Form ES-401-1 11-----------------~------- ---------------------------...... Emergency and Abnormal Plant Evolutions - Tier 1 /Group 2 (RO) ' I ---11

E/APE # / Name I Safety Function K K K A KIA T. ( ) IR I # 1 2 3 1 op1c s

295002 Loss of Main Condenser Vac / 3 : I 0

295007 High Reactor Pressure / 3 I 0

295008 High Reactor Water Level / 2 0 Ability to operate and/or monitor the following as they apply to High Reactor Water 8 Level: Feedwater system,-

295009 Low Reactor Water Level I 2 I 0

295010 High Drywell Pressure / 5 I 0

295011 High Containment Temp / 5 0

295012 High Drywell Temperature/ 5 0

295013 High Suppression Pool Temp./ 5 0 Ability to operate and/or monitor the following as they apply to High Suppression Pool 3.9 1 Temp.: Suppression pool cooling

295014 Inadvertent Reactivity Addition/ 1 0

295015 Incomplete SCRAM/ 1 0 Knowledge of the reasons for the following responses as they apply to Incomplete 1 SCRAM: Bypassing rod insertion blocks

295017 High Off-site Release Rate/ 9 0

295020 Inadvertent Cont. Isolation / 5 & 7 0

295022 Loss of CRD Pumps / 1 Ability to determine and/or interpret the following as they apply to Loss of CRD Pumps:

CRD system status

295029 High Suppression Pool Wtr Lvl / 5 0

295032 High Secondary Containment Area 0 Temperature I 5

295033 High Secondary Containment Area 0 Radiation Levels/ 9

295034 Secondary Containment Ventilation 0 Knowledge of the interrelations between Secondary Containment Ventilation High High Radiation /*9 3 Radiation and the following: SBGT/FRVS: Plant-Specific

295035 Secondary Containment High O Knowledge of the operational implications of the following concepts as they apply to Differential Pressure / 5 2 Secondary Containment High Differential Pressure: Radiation release

295036 Secondary Containment High 0 Sump/Area Water Level / 5

500000 High CTMT Hydrogen Cone. / 5 Knowledge of EOP entry conditions and immediate action steps.

KIA Category Totals: 2 Group Point Total: 7

ES-401, Page 35 of 50 I

ES-401 4 Form ES-401! I ES-401 BWR Examination Outline I Form ES-401-1 1

Plant Systems - Tier 2/Group 1 (RO) I I

System # I Name K K K K K K A A A KIA Topic(s) IR # 2 3 4 5 6 1 4 I

203000 RHR/LPCI: Injection Mode 0 Knowledge of the operational implications of the following concepts as they 11 1 apply to RHR/LPCI: Injection Mode: Testable check valve operation 2.7

Knowledge of the effect that a loss or malfunction of the Shutdown Cooling '

205000 Shutdown Cooling 0 0 ill have on following: Reactor water level: Plant-Specific; Knowledge of the 3.2; i 2 2 8 effect that a loss or malfunction of the following will have on the Shutdown 3.5 I Coolin : RHR service water: Plant-S ecific I

206000 HPCI Knowledge of annunciator alarms, indications, or response procedures. 4.2 11

207000 Isolation (Emergency) I Condenser 0

209001 LPCS bility to monitor automatic operations of the LPCS including: System 3.5 11

I I

209002 HPCS 0 I

I 211000 SLC Ability to locate control room switches, controls, and indications, and to 4.6 1I determine that they correctly reflect the desired plant lineup. i

Knowledge of the physical connections and/or cause-effect relationships 212000 RPS 2 between RPS and the following: Reactor/turbine pressure control system: 3.4 11 Plant-Specific I '

215003 IRM 2.5 1 0 Knowledge of electrical power supplies to the following: IRM !

channels/detectors I

215004 Source Range Monitor 0 Knowledge of the effect that a loss or malfunction of the Source Range 3.7 1i 4 Monitor will have on following: Reactor power and indication I

215005 APRM / LPRM 3 between APRM / LPRM and the following: RBM: Plant-Specific 3.4 1, 0 Knowledge of the physical connections and/or cause-effect relationships I

I 217000 RCIC 3.5 1 0 Knowledge of the effect that a loss or malfunction of the following will have on !

3 the RClC: Suppression pool water supply I Ability to (a) predict the impacts of the following on the ADS; and (b) based on I those predictions, use procedures to correct, control, or mitigate the 4.2; i 218000 ADS consequences of those abnormal conditions or operations: ADS initiation 2 signals present; Ability to monitor automatic operations of the ADS including: 3.7 I

~ ADS valve acoustical monitor noise: Plant-S ecific i Ability to predict and/or monitor changes in parameters associated with I 223002 PCIS/Nuclear Steam Supply operating the PClS/Nuclear Steam Supply Shutoff controls including: System 3.5; 2 Shutoff indicating lights and alarms; Ability to manually operate and/or monitor in the 3.6 I control room: Valve closures Ability to (a) predict the impacts of the following on the SRVs; and (b) based I 239002 SRVs 3.0 1 on those predictions, use procedures to correct, control, or mitigate the I consequences of those abnormal conditions or operations: Stuck open I I

259002 Reactor Water Level Control 3.1 1 0 Knowledge of Reactor Water Level Control design feature(s) and/or interlocks I 6 which provide for the following: Control signal failure I

to manually operate and/or monitor in the control room: Suction valves 3.1 1 I 261000 SGTS Ability I

262001 AC Electrical Distribution Ability to predict and/or monitor changes in parameters associated with 2.7 11

§ll I operating the AC Electrical Distribution controls including: Load currents I

Knowledge of the effect that a loss or malfunction of the UPS (AC/DC) will I 262002 UPS (AC/DC) 7 1 (AC/DC) design feature(s) and/or interlocks which provide for the following: 3.1 ~ 1 0 have on following: Process monitoring: Plant-Specific; Knowledge of UPS 2.9;

Transfer from preferred power to alternate power supplies Ability to predict and/or monitor changes in parameters associated with I 263000 DC Electrical Distribution *ca1 Distribution controls including: Battery 2.5 1 te I

Knowledge of the physical connections and/or cause-effect relationships i 264000 EDGs 0 0 between EDGs and the following: Emergency generator cooling water system; 3.2; 2 4 5 Knowledge of the operational implications of the following concepts as they 3.4 I apply to EDGs: Paralleling A.C. power sources I

300000 Instrument Air 2.8 1 0 '" Knowledge of electrical power supplies to the following: Instrument air I 1 I Ability to (a) predict the impacts of the following on the Component Cooling i 400000 Component Cooling Water Water; and (b) based on those predictions, use procedures to correct, control, 3.3 1 or mitigate the consequences of those abnormal conditions or operations: !

Loss of CCW um

3 2 3 2 2 2 26 I I

ES-401, Page 36 of 50 ES-401-1 5 Form ES-4()1-1

ES-401 BWR Examination Outline Form ES401-1 Plant Systems - Tier 2/Group 2 (RO)

System # I Name K K K K K K KIA Topic(s) IR # 2 3 4 5 6 0 Knowledge of the physical connections and/or cause-effect relationships 201001 CRD Hydraulic 3 between CRD Hydraulic System and the following: Recirculation pumps (seal 3.1 purge): Plant-Specific 201002 RMCS 0 Knowledge of the effect that a loss or malfunction of the following will have on 2.5 the RMCS: Select matrix power

201003 Control Rod and Drive Mechanism 2.8

201004 RSCS 0

201005 RCIS 0

201006 RWM 0

202001 Recirculation 0

202002 Recirculation Flow Control 0

204000 RWCU 0 Knowledge of the effect that a loss or malfunction of the RWCU will have on 2.6 6 following: Area radiation levels

214000 RPIS 0

215001 Traversing In-core Probe 0

215002 RBM 0

216000 Nuclear Boiler Inst. 0 Knowledge of the operational implications of the following concepts as they 3.1 8 apply to Nuclear Boiler Inst.: Steam flow effect on reactor water level 219000 RHR/LPCI: Torus/Pool Cooling 0 Mode Knowledge of the operational implications of the following concepts as they 223001 Primary CTMT and Aux. 3 apply to Primary CTMT and Aux.: Oxygen concentration measurement: Plant-2.7 Specific

226001 RHR/LPCI: CTMT Spray Mode 0

230000 RHR/LPCI: Torus/Pool Spray Mode 0

233000 Fuel Pool Cooling/Cleanup 0 Knowledge of Fuel Pool Cooling/Cleanup design feature(s) and/or interlocks 2.8 which provide for the following: Maintenance of adequate pool temperature

234000 Fuel Handling Equipment Knowledge of less than or equal to one hour Technical Specification action 3.9

239001 Main and Reheat Steam 0

239003 MSIV Leakage Control 0

Ability to (a) predict the impacts of the following on the Reactor/Turtline 241 ODO Reactor/Turbine Pressure Regulator Pressure Regulator; and (b) based on those predictions, use procedures to 3.3 correct, control, or mitigate the consequences of those abnormal conditions or operations: Main turbine overspeed

245000 Main Turbine Gen. / Aux. 0

256000 Reactor Condensate 0

259001 Reactor Feedwater Ability to manually operate and/or monitor in the control room: System valves 3.1

268000 Radwaste 0

271 ODO Off gas 0

Ability to predict and/or monitor changes in parameters associated with 272000 Radiation Monitoring operating the Radiation Monitoring controls including: Lights, alarms, and 2.9 indications associated with surveillance testing

286000 Fire Protection 0 Knowledge of electrical 2 power supplies to the following: Pumps 2.9

288000 Plant Ventilation 0

290001 Secondary CTMT 0

290003 Control Room HVAC 0

290002 Reactor Vessel Internals 0

KIA Category Totals: 2 12

ES-401, Page 37 of 50 ES-401 2 Form ESt401-1

, I

ES-401 BWR Examination Outline Form ES-401-1 11-----------------r--..--.----.-- ------------------------.-----"-r---~I Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)

  • 1 E/APE # / Name I Safety Function K K K A KIA Topic(s) IR #

2 3 1

295001 Partial or Complete Loss of Forced 0 Core Flow Circulation I 1 & 4

Ability to determine and/or interpret the following as they apply to Partial 295003 Partial or Complete Loss of AC/ 6 or Complete Loss of AC: Whether a partial or complete loss of AC. power 4.2 has occurred

295004 Partial or Total Loss of DC Pwr / 6 Ability to determine and/or interpret the following as they apply to Partial 2.9 or Total Loss of DC Pwr: Battery voltage

295005 Main Turbine Generator Trip/ 3 Ability to determine and/or interpret the following as they apply to Main 3.9 '

Turbine Generator Trip: Reactor power

295006 SCRAM / 1 0

295016 Control Room Abandonment 17 0

295018 Partial or Total Loss of CCW / 8 0

Ability to determine and/or interpret the following as they apply to Partial 295019 Partial or Total Loss of Inst. Air/ 8 or Total Loss of Inst. Air: Status of safety-related instrument air system loads (see AK2.1-AK2.19)

295021 Loss of Shutdown Cooling I 4

  • I 0

295023 Refueling Ace / 8 0

295024 High Drywell Pressure / 5 0

I

295025 High Reactor Pressure / 3 0

295026 Suppression Pool High Water Temp./ 5.1 0

295027 High Containment Temperature/ 5 0

295028 High Drywell Temperature/ 5 0

295030 Low Suppression Pool Wtr Lvl / 5 Knowledge of EOP mitigation strategies.

295031 Reactor Low Water Level/ 2 Knowledge of the specific bases for EOPs.

295037 SCRAM Condition Present and Reactor Power Above APRM 0 Downscale or Unknown / 1

295038 High Off-site Release Rate / 9 0

600000 Plant Fire On Site I 8 0

700000 Generator Voltage and Electric Grid Knowledge of the emergency action level thresholds and classifications. 4.6 Disturbances / 6

KIA Category Totals: 0 0 0 0 Group Point Total: 7

ES-401, Page 34 of 50 ES-401 3 Form ES 401-1

  • 1 ES-401 BWR Examination Outline Form ES-401-1 11------------------.:---,--,----,-- ------------------------,----+-r---*11 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO) ' I

E/APE # / Name/ Safety Function K K K A IR: I # 1 2 3 1 KIA Topic(s)

295002 Loss of Main Condenser Vac / 3 Knowledge of abnormal condition procedures. 4.~ I

295007 High Reactor Pressure / 3 : I 0

295008 High Reactor Water Level / 2 : I 0

295009 Low Reactor Water Level / 2 : I 0

295010 High Drywell Pressure/ 5 0

295011 High Containment Temp/ 5 0

295012 High Drywell Temperature/ 5 0

295013 High Suppression Pool Temp./ 5 0

295014 Inadvertent Reactivity Addition/ 1 0

295015 Incomplete SCRAM/ 1 0

295017 High Off-site Release Rate/ 9 0

295020 Inadvertent Cont. Isolation / 5 & 7 0

295022 Loss of CRD Pumps / 1 0

295029 High Suppression Pool Wtr Lvl / 5 0

295032 High Secondary Containment Area 0 Temperature/ 5

295033 High Secondary Containment Area Ability to determine and/or interpret the following as they apply to High Secondary Radiation Levels I 9 Containment Area Radiation Levels: Cause of high area radiation

295034 Secondary Containment Ventilation 0 High Radiation / 9

295035 Secondary Containment High 0 Differential Pressure / 5 295036 Secondary Containment High Ability to determine and/or interpret the following as they apply to Secondary Sump/Area Water Level / 5 Containment High Sump/ Area Water Level: Operability of components within the 3.2 affected area

500000 High CTMT Hydrogen Cone. / 5 0

KIA Category Totals: 0 0 0 0 Group Point Total: I 3

ES-401, Page 35 of 50 ES-401 4 Form ES.~401-1

ES-401 BWR Examination Outline FormltS-401-1 11----------------.---.---,.---.---.---,.---.-- ------------------------r---i-T----11 Plant Systems - Tier 2/Group 1 (SRO) I System # I Name K K K K K K A A A K/AT. () IR I #

2 3 4 5 6 1 opIc s

203000 RHR/LPCI: Injection I 0

205000 Shutdown Cooling Mode. I 0

206000 HPCI I 0

207000 Isolation (Emergency)

Condenser I 0

209001 LPCS Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

209002 HPCS I 0

211000 SLC 0

212000 RPS, Ability to determine operability and/or availability of safety related equipment. 4.6

215003 IRM 0

215004 Source Range Monitor 0

215005 APRM / LPRM 0

Ability to (a) predict the impacts of the following on the RCIC; and (b) based on 217000 RCIC those predictions, use procedures to correct, control, or mitigate the 3.0 consequences of those abnormal conditions or operations: Loss of vacuum um

218000 ADS 0

223002 PCIS/Nuclear Steam Supply 0 Shutoff Ability to (a) predict the impacts of the following on the SRVs; and (b) based on 239002 SRVs those predictions, use procedures to correct, control, or mitigate the 4.3 consequences of those abnormal conditions or operations: Reactor high ressure

259002 Reactor Water Level Control 0

Ability to (a) predict the impacts of the following on the SGTS; and (b) based on 261000 SGTS those predictions, use procedures to correct, control, or mitigate the 3.1 consequences of those abnormal conditions or operations: Fan trips

262001 AC Electrical Distribution 0

262002 UPS (AC/DC) 0

263000 DC Electrical Distribution 0

264000 EDGs 0

300000 Instrument Air 0

00000 Component Cooling Water 0

Category Totals: 0 0 0 0 0 0 0 Group Point Total: 5

ES-401, Page 36 of 50 ES-401 5 Form Es-4:01-1 I

ES-401. BWR Examination Outline Form ES-401-1

Plant Systems - Tier 2/Group 2 (SRO) I

System # / Name K K K K K K KIA Topic(s) IR; # *1 2 3 4 5 6

201001 CRD Hydraulic 0

201002 RMCS : I 0

01003 Control Rod and Drive Mechanism 0

01004 RSCS 0

01005 RCIS 0

201006 RWM 0

02001 Recirculation Ability to apply Technical Specifications for a system. 4.7

02002 Recirculation Flow Control 0

04000 RWCU I 0

14000 RPIS : I 0

15001 Traversing In-core Probe I 0

15002 RBM I 0

16000 Nuclear Boiler Inst. I 0 Ability to (a) predict the impacts of the following on the RHR/LPCI: Torus/Pool 19000 RHR/LPCI: Torus/Pool Cooling Cooling Mode; and (b) based on those predictions. use procedures to correct. 3.2:

Mode control, or mitigate the consequences of those abnormal conditions or o erations: Valve o enin s 23001 Primary CTMT and Aux. 0

226001 RHR/LPCI: CTMT Spray Mode 0

230000 RHR/LPCI: Torus/Pool Spray Mode 0

33000 Fuel Pool Cooling/Cleanup 0

234000 Fuel Handling Equipment I I 0 Ability to (a) predict the impacts of the following on the Main and Reheat 39001 Main and Reheat Steam Steam; and (b) based on those predictions, use procedures to correct, control, 1 I or mitigate the consequences of those abnormal conditions or operations: Main 4.2 steam line hi h radiation 39003 MSIV Leakage Control 0

41000 Reactor/Turbine Pressure Regulator 0

45000 Main Turbine Gen./ Aux. 0

56000 Reactor Condensate 0

59001 Reactor Feedwater 0

68000 Radwaste 0

271000 Offgas 0

272000 Radiation Monitoring 0

286000 Fire Protection 0

0

0

90003 Control Room HVAC 0

0

Category Totals: 0 0 0 0 0 0 Group Point Total: 3

ES-401, Page 37 of 50 I

ES-401 Generic Knowledse and Abilities Outline (Tier 3) Form ES-401-3 *

!Facility Name: Date of Exam: I Category KIA# Topic KU ~KU-Only IR # IR #

2.1. 20 Ability to interpret and execute procedure steps. 4.6 1 4.6

2.1. 30 Ability to locate anci operate components, including local controls. 4.4 1 4.0

Knowledge of RO duties in the control room during fuel handling such as responding to

1. operated from the control room in support of fueling operations, and supporting 2.1. 44 alarms from the fuel handling area, communication with the fuel storage facility, systems 3.9 1 3.8

instrumentation.

Conduct of Operations 2.1. 13 Knowledge of facility requirements for controlling vital/controlled access. 2.5 3.2 1

2.1. 41 Knowledge of the refueling process. 2.8 3.7 1

2.1.

Subtotal 3 2

to perform pre-startup procedures for the facility, including operating those controls 4.5 1 4.4 2.2. 01 Ability associated with plant equipment that could affect reactivity.

2.2. 03 Knowledge of the design, procedural, and operational differences between units. 3.8 1 3.9 2.2. 44 Ability to interpret control room indications to verify the status and operation of a system, 4.2 1 4.4

2. and understand how operator actions and directives affect plant and system conditions.

Equipment 2.2. 25 Knowledge of the bases in Technical Specifications for limiting conditions for operations 3.2 4.2 1 Control and safety limits.

2.2. 18 Knowledge of the process for managing maintenance activities during shutdown 2.6 3.9 1 operations, such as risk assessments, work prioritization, etc.

2.2.

Subtotal -3 2 2.3.. 15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, 2.9 1 3.1 portable survey instruments, personnel monttoring equipment, etc.

2.3. 04 Knowledge of radiation exposure limits under normal or emergency conditions. 3.2 1 3.7

3. portable survey instruments, personnel monitoring equipment, etc. 2.3. 05 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, 2.9 2.9 1

Radiation 2.3.

Control 2.3.

2.3.

Subtotal 2 --=c Knowledge of the parameters and logic used to assess the status of safety functions, such 2.4. 21 as reactivity control, core cooling and heat removal, reactor coolant system integrity: 4.0 1 4.6 containment conditions radioactivitv release control etc.

2.4. 35 Knowledge of local auxiliary operator tasks during an emergency and the resultant 3.8 1 4.0 operational effects.

4. 2.4. 43 Knowledge of emergency communications systems and techniques. 3.2 3.8 1 Emergency Knowledge of the organization of the operating procedures network for normal, abnormal, Procedures 2.4. 05 and emergency evolutions. 3.7 4.3 1

/ Plan 2.4.

2.4.

Subtotal 2 2 Tier 3 Point Total 10 7

ES-401, Page 43 of 50 ES-401 Record of Rejected K/As Form ES-401-4

Tier/ Randomly Reason for Rejection Group Selected KIA

SRO 1/1 295031 G2.4.02 Supports testing the RO level, but not the SRO level. Replaced with (79) 295031 G2.4.18

SRO 1/2 295002 G2.4.04 Overlap with NRC item 295031 G2.4.04. Replaced with 295002 G2.4. l 1 (85)

RO 1/2 295033EKI.03 Overlap with NRC item 295033 A2.03. Replaced with 295034 K2.03 (56)

RO 2/2 (1) 239003 K4.02 System not at Peach Bottom. Replaced with 233000 K4.03 ROI/I 295037 EKI.07 Unable to write adequate question. Replaced with 295037 EKI.05 (40)

RO 1/1 295016 AAI.03 Unable to develop adequate question. Replaced with 295016 AAI.07 (53)

SRO 1/1 295005 AA2.03 Unable to develop adequate question. Replaced with 295006 AA2.05 (82)

SRO 1/2 295036 AA2.03 Unable to develop SRO level question. Replaced with 295036 AA2.01 (86)

SRO2/2 233000 G2.42 Oversample issue. Replaced with 202001 G2.40 (94)

SRO2/2 234000 K4.02 Oversample issue. Replaced with 239001 A2.05 (88) 3 (22) G2.3.7 Unable to develop adequate question. Replaced with G2.3.04

SRO 2/1 209001 G2.4.50 Unable to develop adequate question. Replaced with G2.2.42 (91) 3 (92) G2.2.4 Unable to develop SRO level question. Replaced with G2.2.25 RO 1/2 (43) 500000 G2.4.03 Unable to develop adequate question. Replaced with G2.4.01 RO2/1 218000 A3.09 Unable to develop adequate question. Replaced with 218000 A3.03 (26)

RO 1/1 295030 G2.4.30 Unable to develop adequate question due to overlap. Replaced with (13) G2.4.20 RO2/2 201003 A3.01 Unable to develop adequate question. Replaced with 201003 Al.02 (65)

RO 1/2 295035 EK2.03 Unable to develop adequate question. Replaced with 295035 Kl.02 (30)

SRO 2/1 239002 A2.04 Overlap issue. Replaced with 239002 A2.06 (99)

3 (6) 2.1.19 Unable to develop adequate question. Replaced with 2.1.20 3 (97) 2.1.39 Unable to develop adequate question. Replaced with 2.1.41 3 (48) 2.2.4 Unable to develop adequate question. Replaced with 2.2.3 3 (89) 2.3.7 Too many G2.3 topic questions. Replaced with 2.2.18

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