ML20022A259

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Final Written Examination and Operating Test Outlines (Folder 3)
ML20022A259
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 01/16/2020
From:
Exelon Generation Co
To: Todd Fish
Operations Branch I
Shared Package
ML19151A455 List:
References
CAC 000500
Download: ML20022A259 (27)
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Text

ES-301

  • Administrative Topics Outline Form ES-301-1 Facility:

Nine Mile Point Unit 2 Date of Examination:

December 2019 Examination Level: RO Operating Test Number:

LC2 18-1 NRG Administrative Topic (see Note)

Type Describe activity to be performed Code*

Determine Containment Water Level Conduct of Operations D,R N2-EOP-6.23, KA 2.1.25 (3.9)

Determine Heatup Rate During Startup Conduct of Operations D,R N2-0SP-RCS-@001, KA 2.1.43 (4.1)

Perform Off-Site AC Breaker Alignment Verification Equipment Control N,S N2-0SP-LOG-W001, KA 2.2.15 (3.9)

Radiological and Heat Stress Requirements P,R Related to Operator Work In High Radiation Radiation Control Areas - Valve leak in RWCU Pump Room (2015 NRC)

RP-AA-460, RP-AA-203, KA 2.3.7 (3.5)

Em~rg~ncy Plan NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

  • Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:5 3 for ROs; :5 4 for SROs and RO retakes)

(N)ew or (M)odified from bank (~ 1)

(P)revious 2 exams (:5 1, randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

Nine Mile Point Unit 2 Date of Examination:

December 2019 Examination Level: SRO Operating Test Number:

LC2 18-1 NRC Administrative Topic (see Note)

Type Describe activity to be performed Code*

Determine the Significance of a Reactivity Conduct of Operations D,R Event and Actions Required OP-AA-300, N2-0P-96, KIA 2.1.37 (4.6)

Reactivate SRO Licenses Conduct of Operations D,R OP-AA-105-102, KA 2.1.4 (3.8)

Review Daily Logs - Jet Pump Flows Equipment Control M,S N2-0SP-LOG-D001, Technical Specifications, KA 2.2.12 (3. 7)

Radiological and Heat Stress Requirements P,R Related to Operator Work In High Radiation Radiation Control Areas - Valve leak in RWCU Pump Room (2015 NRG)

RP-AA-460, RP-AA-203, KA 2.3. 7 (3.6)

Security Event Re-Classification Notification Emergency Plan D,R EP-CE-111, EP-AA-1013, KA 2.4.41 (4.6)

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: Nine Mile Point Unit 2 Date of Examination: December 2019 Exam Level: RO/SR0-1/SRO-U Operating Test No.: LC2 18-1 NRG Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U)

System I JPM Title Type Code*

Safety Function

a. Start 2 FWS-P1 A and Transfer Feedwater Level Control to FWS-LV55A D,L,S 2

KIA 259001 A4.05 (4.0/3.9) N2-0P-3

b. Manual Initiation of Control Building Special Filter Train D,EN,S 9

KIA 290003 A4.01 (3.2/3.2) N2-0P-53A C.

Main Steam Line Warmup Operation (Alternate Path) (RO Only)

A,D,L,S 3

KIA 239001 A4.02 (3.2/3.2) N2-0P-1 & N2-SOP-83

d. Restore SOC to Service (Alternate Path)

A,D,L,S 4

KIA 205000 A4.01 (3.7/3.7) N2-0P-31

e. Suppression Pool Fill Utilizing CSH Pump M,S 5

KIA 295030 EA1.03 (3.4/3.4) N2-0P-33

f.

Parallel and Load 2EGS*EG1 KIA 264000 A4.02 (3.4/3.4) and A4.04 (3.7/3.7) N2-0SP-EGS-D,S 6

M@001

g. Enter a Substitute Rod Position in the RWM D,P,S 7

KIA 201006 A4.06 (3.2/3.2) N2-0P-95A (NRG 2015)

h. Temper SW Using Circ Water N2-0P-11 (Alternate Path)

A,D,S 8

KIA 400000 A4.01 (3.1 / 3.0) N2-0P-11 In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Vent Control Rod Overpiston D,E,R 1

KIA295015 AA.1.01 (3.8/3.9) N2-EOP-6.14

j. Place Battery Charger 2BYS-CHGR1 C1 in Service. (Alternate Path)

A,N 6

KIA 263000 A1.01 (2.5/2.8) N2-0P-73A

k. Reset a Reactor Protection System Electrical Protection Assembly (EPA) (Alternate Path)

A,D 7

KIA 212000 A4.14 (3.8/3.8) N2-SOP-97 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 I

Criteria for RO/ SRO-I / SRO-U

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 2 Scenario No.: NRC-1 Op-Test No.: LC2 18-1 Examiners:

Operators:

Initial Conditions: Reactor power is being reduced in preparation for a rod line adjustment. The plant is currently operating at 95% power with service water pump SWP*P1 E out of service for pump bearing replacement.

Turnover: The crew will perform N2-0SP-CSL-Q@002, LPCS Pump and Valve Operability and System Integrity Test (section 6.2 only), then lower reactor power to 90% using reactor recirculation flow.

Critical Tasks: See page 2 Event Malf.

Event Event No.

No.

Type*

Description N/A N-BOP, Perform N2-0SP-CSL-Q@002, LPCS Pump and Valve Operability

SRO, and System Integrity Test (section 6.2 only) with failure of CSL 1

TS-SRO suction valve to re-open.

N2-0SP-CSL-Q@002, T.S. 3.6.1.3 N/A R-ATC Lower reactor power to 90% using reactor recirculation flow.

2 SRO N2-0P-101D CUOB I-BOP, RWCU fails to automatically isolate on RWCU flow mismatch 3

SRO, caused by cleanup RWCU non-regen heat exchanger tube leak.

TS-SRO ARPs, T.S. 3.3.6.1 PC28A C-BOP, Loss of Drywell Cooling.

4 SRO TS-SRO N2-S0P-60, T.S. 3.3.6.1 FW13 I-ATC, Feedwater Master Controller Failure - High.

5 SRO N2-S0P-06 TU02 C-ATC Rising Main Turbine vibrations require scram.

6 SRO ARP's, N2-S0P-21, N2-S0P-101C MS04 M-AII Main Steam Line Break in Primary Containment with loss of 7

condensate and feedwater system.

N2-EOP-RPV, N2-EOP-PC RR27 I-All All RPV level Instruments fail upscale.

8 N2-EOP-C4 RH22A C-AII RHR 'A' injection valve (2RHS*MOV24A) loss of power.

9 N2-EOP-C4 RH02B C-AII 2RHS*MOV24B will fail to open, requiring manual line up and inject 10 with 2RHS*MOV40B.

N2-EOP-C4

Facility: Nine Mile Point Unit 2 Scenario No.: NRC-1 Op-Test No.: LC2 18-1

1. Malfunctions after EOP entry (1-2) 3 Event8,9,10
2. Abnormal events (2-4) 4 Events 3, 4, 5, 6
3. Major transients (1-2) 1 Event 7
4. EOPs entered/requiring substantive actions (1-2) 2 N2-EOP-RPV, N2-EOP-PC
5. Entry into a contingency EOP with substantive actions (?.1 per scenario set) 1 N2-EOP-C4
6. Pre-identified Critical Tasks (> 2) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0, Given the plant with RPV water level unknown, open 7 SRVs to in Critical Task 1.0 is identified as critical accordance with N2-EOP-C4.

because with Reactor water level unknown, the status of core cooling is unknown. N2-EOP-C4 requires that MS/V's, MSL drain isolations and RCIC steam isolations be closed if at least 6 SRV's can be opened. In this case 7 SRV's can be opened. Opening 7 SRV's provides a flowpath for injected water to flow through the SRV's, thereby providing indications that RPV level is at the level of the main steam lines using the SRV acoustic monitors.

CT-2.0, Given the plant with RPV water level unknown, establish injection Critical Task 2.0 is identified as critical and flood the RPV in accordance with N2-EOP-C4.

because with Reactor water level unknown, the status of core cooling is unknown. RPV flooding is required to establish conditions to cool the core.

This protects the fuel cladding integrity.

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 2 Scenario No.: NRC-2 Op-Test No.: LC2 18-1 Examiners:

Operators:

Initial Conditions: Reactor power is at 95% with 21AS-C3C out of service for unloader valve replacement. Reactor power is being returned to rated following a feedwater pump swap. Power ascension is on hold awaiting reactor engineering update of the fuel preconditioning ReMA.

Turnover: The crew will perform a Live Bus Transfer of 2NNS-SWG013 to 2NNS-SWG012 followed by performance of N2-0SP-RMC-W @001, Control Rod Movement and Position Verification Test.

Critical Tasks: See page 2 Event Malf.

Event Event No.

No.

Type*

Description N/A N-BOP, Live Bus Transfer of 2NNS-SWG013 to 2NNS-SWG012.

1 SRO N2-0P-718 RD11 I-ATC, Perform N2-0SP-RMC-W@001, Control Rod Movement and

SRO, Position Verification Test with Rod Position Indication Failure.

2 TS-SRO N2-0SP-RMC-W@001, N2-0P-96, T.S. 3.1.3 RD18 C-ATC CRD Pump Trip on Low Suction Pressure.

3 SRO ARP's, N2-SOP-30 ED05B C-BOP, Loss of 2ENS*SWG102 (Electrical Fault).

4 SRO TS-SRO ARPs, T.S. 3.5.1, 3.8.4, 3.8.8, 3.6.4.3 FW22B1 R-ATC, First Point Feed Water Heater (1 B) Tube Leak.

5 C-BOP SRO N2-SOP-08, N2-SOP-101 D MT01 M-AII Seismic Event Causes a Break in the Suppression Pool Wall and 6

subsequent RPV Blowdown.

N2-EOP-RPV, N2-EOP-PC, N2-EOP-C2 RP03 I-All When manual scram attempted, Control rods fail to insert using RPS, RRCS initiation required to insert the control rods.

7 N2-SO P-101 C

TC15A, C-AII Trip of running EHC Pump with a Failure of Standby to Start.

8 TC16B N2-EOP-RPV (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 2 Scenario No.: NRC-2 Op-Test No.: LC218-1

1. Malfunctions after EOP entry (1-2) 2 Event7,8
2. Abnormal events (2-4) 4 Events 2, 3, 4, 5
3. Major transients (1-2) 1 Event6
4. EOPs entered/requiring substantive actions (1-2) 2 N2-EOP-RPV, N2-EOP-PC
5. Entry into a contingency EOP with substantive actions (?_ 1 per scenario set) 1 N2-EOP-C2
6. Pre-identified Critical Tasks (> 2) 3 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0, Given the plant operating in the "Exit Region" of the power to flow Critical Task 1.0 is identified as critical map due to a RCS-FCV runback, the crew will insert the first four CRAM rods because without operator action the in accordance with N2-SOP-29.

reactor would be operating in a high power (rodline) low core flow condition which is a condition that could cause core power oscillations which is a precursor to fuel damage.

CT-2.0, Given the failure of RPS to initiate a successful reactor scram, the Critical Task 2.0 is identified as critical crew will manually initiate RRCS in accordance with N2-SOP-101 C.

because the reactor a reactor scram is required "before" the blowdown is initiated to shut down the reactor and reduce the steam generation rate. With the failure of RPS to function, manual action is required to shutdown the reactor.

CT-3.0, Given the plant with suppression pool water level that cannot be Critical Task 3.0 is identified as critical maintained above the 192' elevation, the crew will commence a RPV because 192' is the minimum indicated blowdown before suppression pool level reaches the 192' elevation in suppression pool level. An on-scale accordance with N2-EOP-PC and N2-EOP-C2.

indication is required to ensure that the actual suppression pool level is above the top of the SRV discharge devices. If the SRVs were opened with the discharge devices exposed, steam would pass directly into the suppression chamber airspace, bypassing the suppression pool. The resulting pressure increase could exceed the maximum pressure capability of the primary containment.

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 2 Scenario No.: NRC-4 Op-Test No.: LC2 18-1 Examiners:

Operators:

Initial Conditions: The plant is operating at rated power with 'C' Narrow Range Level Transmitter failed high.

Turnover: The crew will perform N2-0SP-RHS-Q@006, RHR System Loop C Pump and Valve Operability Test and System Integrity Test.

Critical Tasks: See page 2 Event Malf.

Event Event No.

No.

Type*

Description N/A N-BOP, Perform N2-0SP-RHS-Q@006 Surveillance Test.

1 SRO N2-0SP-RHS-Q@006 NM19A I-ATC, RBM "A" lnop requires bypassing.

2

SRO, TS-SRO ARP's, N2-0P-92 CS01A I-BOP, Inadvertent HPCS Initiation.

3

SRO, TS-SRO ARP's, N2-0P-33, T.S. 3.5.1, T.S. 3.6.1.3 RD05 R-ATC, Control Rod Drift Out.

4 C-BOP

SRO, ARP's, N2-S0P-8, T,S. 3.1.3 TS-SRO Remote I-BOP, CCS-TIK104 Auto Setpoint Failure.

5 CW27 SRO N2-0P-14, N2-S0P-14 NM12B I-ATC, APRM #2 Failure Downscale.

6

SRO, ARP's, N2-0P-92
RD17Z, M-AII Loss of 2NNS-SWG011 & Remaining Condensate Pumps, Scram,
ED04A, ATWS, RCIC Trip, RPV Slowdown, Re-inject with Preferred ATWS 7
FW01B, Injection Systems.

RC06 N2-EOP-RPV, N2-EOP-PC, N2-EOP-C2, N2-EOP-C5 SL03 C-AII SLC pump suction valve fails to open (2SLS-MOV1 A).

8 N2-0P-36, Attachment 1 RH01 C-AII RHR Pump trip during Reflood.

9 N2-EOP-RPV (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 2 Scenario No.: NRC-4 Op-Test No.: LC218-1

1. Malfunctions after EOP entry (1-2) 2 Event8,9
2. Abnormal events (2-4) 5 Events 2, 3, 4, 5, 6
3. Major transients (1-2) 1 Event 7
4. EOPs entered/requiring substantive actions (1-2) 2 N2-EOP-RPV, N2-EOP-PC
5. Entry into a contingency EOP with substantive actions (~1 per scenario set) 2 N2-EOP-C2, N2-EOP-C5
6. Pre-identified Critical Tasks (> 2) 5 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0, Given the plant at rated power with a control rod drifting out, the Critical Task 1.0 is identified as.critical crew will reduce reactor power to approximately 85% in accordance with N2-because without a power reduction, SOP-8 and N2-SOP-101D.

APRM power will rise above the licensed limit and present a challenge to thermal limits and be a precursor to fuel damage.

N2-SOP-08 requires a power reduction to approximately 85%, however the safety significance is to only reduce power below the licensed limit of 3988 MWth.

CT-2.0, Given the plant with a high power ATWS and degraded high pressure Critical Task 2.0 is identified as critical preferred injection sources, the crew will inhibit ADS in accordance with N2-because with lowering RPV level; the EDP-CS.

ADS System, if not disabled, would automatically open all 7 ADS valves and allow the low pressure EGGS pumps to inject if not terminated and prevented.

With a high power A TWS in progress, the pressure transient and resultant uncontrolled injection of relatively cold water would result in fuel damage.

CT-3.0, Given a failure of the reactor to SCRAM and RPV Slowdown Critical Task 3.0 is identified as critical required, the crew will terminate and prevent all injection sources except because without operator action, the boron, CRD, and RCIC in accordance with N2-EOP-C2.

manual RPV blowdown combined with a high power A TWS in progress would cause the uncontrolled injection of relatively cold water which would result in fuel damage.

CT-4.0, Given a failure of the reactor to SCRAM with an RPV blowdown Critical Task 4.0 is identified as critical required, the crew will open all 7 ADS valves in accordance with N2-EOP-C2.

because without operator action, reactor pressure would remain too high to facilitate the only remaining preferred injection source to inject into the vessel.

This would prevent RPV water level from being restored and therefore prevent adequate core cooling from being assured. The intent is to get at least 7 SRV's (ADS or non-ADS) open.

CT-5.0, Given a failure of the reactor to SCRAM and the RPV has been blown Critical Task 5.0 is identified as critical down per N2-EOP-C2, the crew will resume injection when RPV pressure because without operator action, no lowers below the MSCP to restore and maintain RPV water level above the injection will occur. The sources that are MSCWL in accordance with N2-EOP-C5.

available for injection must be manually lined up and therefore failure to perform this step would cause RPV water level to continue to lower below the level at which adequate core cooling is assured.

ES-401 1

Form ES-401-1 IL..

Mile Point Unit 2 Date of Exam:

Tier Group RO KIA Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

1 3

3 4

4 3

3 20 4

3 7

Emergency and 2

1 2

1 N/A 1

1 N/A 1

7 2

1 3

Abnormal Plant Evolutions Tier Totals 4

5 5

5 4

4 27 6

4 10

2.

1 3

1 3

2 3

3 2

3 2

2 2

26 3

2 5

Plant 2

1 1

1 1

1 1

1 1

1 1

2 12 0

2 1

3 Systems Tier Totals 4

2 4

3 4

4 3

4 3

3 4

38 5

3 8

3. Generic Knowledge and Abilities 1

2 3

4 10 1

2 3

4 7

Categories 3

2 2

3 1

2 2

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 NRG 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 0.1.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 K/As having an importance rating (IR) of 2.5 or higher. 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 0.1.b of ES-401 for the applicable Kl As.
8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' I Rs for the applicable license level, and the point totals (#) for each system and category. Enter the group arid 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 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

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

E/APE #/Name/Safety Function K1 K2 K3 A1 A2 G*

KIA Topic(s)

IA 295001 (APE 1) Partial or Complete Loss of 3

Knowledge of the operational 3.6 1

Forced Core Flow Circulation/ 1 & 4 implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW.

CIRCULATION: Thermal Limits 295003 (APE 3) Partial or Complete Loss of

. 2.1.28 Knowledge and the purpose of major 4.1 2

AC Power /6 system components and controls.

295004 (APE 4) Partial or Total Loss of DC 1

Knowledge of the reasons for the 2.6 3

Power/6 following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Load shedding: Plant-Specific 295005 (APE 5) Main Turbine Generator Trip/

4 Knowledge of the interrelations between 3.3 4

3 MAIN TURBINE GENERATOR TRIP and the followinq: Main qenerator protection 295006 (APE 6) Scram / 1 6

Ability to determine and/or interpret the 3.5 5

following as they apply to SCRAM:

Cause of reactor SCRAM 295016 (APE 16) Control Room Abandonment 1

Knowledge of the interrelations between 4.4 6

17 CONTROL ROOM ABANDONMENT and the following: Remote shutdown panel:

Plant-Specific 295018 (APE 18) Partial or Complete Loss of 7

Knowledge of the reasons for the 3.1 7

ccw /8 following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

Cross-connectinq with backup systems 295019 (APE 19) Partial or Complete Loss of 2

Ability to operate and/or monitor the 3.3 8

Instrument Air/ 8 following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Instrument air system valves: Plant-Specific 295021 (APE 21) Loss of Shutdown Cooling/

1 Knowledge of the reasons for the 3.3 9

4 following responses as they apply to LOSS OF SHUTDOWN COOLING:

Raisinq reactor water level 295023 (APE 23) Refueling Accidents I 8 2

Knowledge of the reasons for the 3.4 10 following responses as they apply to REFUELING ACCIDENTS: Interlocks associated with fuel handlina eauioment 295024 High Drywell Pressure / 5 4

Ability to determine and/or interpret the 3.9 11 following as they apply to HIGH DRYWELL PRESSURE: Suppression chamber pressure: Plant-Specific 295025 (EPE 2) High Reactor Pressure/ 3 1

Knowledge of the interrelations between 4.1 12 HIGH REACTOR PRESSURE and the following: RPS 295026 (EPE 3) Suppression Pool High Water 1

Ability to operate and/or monitor the 4.1 13 Temperature I 5 following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool coolinq

ES-401 3

Form ES-401-1 295028 (EPE 5) High Drywell Temperature 2

Ability to operate and/or monitor the 3.9 14 (Mark I and Mark II only)/ 5 following as they apply to HIGH DRYWELL TEMPERATURE: Drywell ventilation system 295030 (EPE 7) Low Suppression Pool Water 2

Knowledge of the operational 3.5 15 Level/ 5 implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Pump NPSH 295031 (EPE 8) Reactor Low Water Level/ 2 2.4.18 Knowledge of the specific bases for 3.3 16 EOPs.

295037 (EPE 14) Scram Condition Present 2.4.35 Knowledge of local auxiliary operator 3.8 17 tasks during emergency and the resultant and Reactor Power Above APRM Downscale operational effects or Unknown / 1 295038 (EPE 15) High Offsite Radioactivity 3

Ability to determine and/or interpret the 3.5 18 Release Rate/ 9 following as they apply to HIGH OFF-SITE RELEASE RATE: Radiation levels 600000 (APE 24) Plant Fire On Site/ 8 2

Knowledge of the operational 2.9 19 implications of the following concepts as they apply to Plant Fire On Site: Fire Fiqhtinq 700000 (APE 25) Generator Voltage and 5

Ability to operate and/or monitor the 3.9 20 Electric Grid Disturbances/ 6 following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Engineered safety features 295003 (APE 3) Partial or Complete Loss of 4

Ability to determine and/or interpret the 3.7 76 AC Power/6 following as they apply to PARTIAL OR COMPLETE LOSS OF A.G. POWER:

Svstem lineups 295016 (APE 16) Control Room Abandonment 2.4.34 Knowledge of RO tasks performed 4.1 77 17 outside the main control room during an emergency and the resultant operational effects.

295018 (APE 18) Partial or Complete Loss of 2.1.23 Ability to perform specific system and 4.4 78 CCW/8 integrated plant procedures during all modes of plant operation 295019 (APE 19) Partial or Complete Loss of 2.4.4 Ability to recognize abnormal indications 4.7 79 Instrument Air/ 8 for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

295025 (EPE 2) High Reactor Pressure / 3 5

Ability to determine and/or interpret the 3.6 80 following as they apply to HIGH REACTOR PRESSURE: Decay heat generation 295031 (EPE 8) Reactor Low Water Level/ 2 3

Ability to determine and/or interpret the 4.2 81 following as they apply to REACTOR LOW WATER LEVEL: Reactor pressure 295037 (EPE 14) Scram Condition Present 7

Ability to determine and/or interpret the 4.2 82 and Reactor Power Above APRM Downscale following as they apply to SCRAM or Unknown / 1 CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Containment conditions/isolations KIA Category Totals:

3 3

4 4

3/4 3/3 Group Point Total:

2n'~

ES-401 4

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emeroencv and Abnormal Plant Evolutions-Tier 1/Group 2 (RO/SRO)

E/APE # / Name/ Safety Function K1 K2 K3 A1 A2 G*

KIA Topic(s)

IR 295002 (APE 2) Loss of Main Condenser 2.1.31 Ability to locate control room switches, 4.6 21 Vacuum/ 3 controls and indications and to determine that they are correctly reflectina the desired plant lineup 295009 (APE 9) Low Reactor Water Level/ 2 1

Knowledge of the reasons for the 3.2 22 following responses as they apply to LOW REACTOR WATER LEVEL:

Recirculation pump run back: Plant-Specific 295012 (APE 12) High Drywell Temperature/

2 Knowledge of the interrelations between 3.6 23 5

HIGH DRYWELL TEMPERATURE and the followina: Drywell cooling 295029 (EPE 6) High Suppression Pool Water 1

Knowledge of the operational 3.4 24 Level/ 5 implications of the following concepts as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Containment inteqrity 500000 (EPE 16) High Containment Hydrogen 3

Knowledge of the interrelations between 3.3 25 Concentrations I 5 HIGH CONTAINMENT HYDROGEN CONCENTRATIONS the following:

Containment Atmosphere Control System 295034 (EPE 11) Secondary Containment 2

Ability to determine and/or interpret the 3.7 26 Ventilation High Radiation/ 9 following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Cause of high radiation levels 295036 (EPE 13) Secondary Containment 4

Ability to operate and/or monitor the 3.1 27 High Sump/Area Water Level/ 5 following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Radiation monitoring:

Plant-Specific 295015 (APE 15) Incomplete Scram/ 1 2-Ability to determine and/or interpret the 4.2 83 following as they apply to INCOMPLETE SCRAM: Control rod IPOSition 295032 (EPE 9) High Secondary Containment 2

Ability to determine and/or interpret the 3.5 84 Area Temperature/ 5 following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Equipment operability 295034 (EPE 11) Secondary Containment 2.4.6 Knowledge of EOP mitigation 4.7 85 Ventilation High Radiation/ 9 strategies.

KIA Categorv Point Totals:

1 2

1 1

1/2 1/1 Group Point Total:

7/3

ES-401 5

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

System # I Name K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G*

KIA Topic(s)

IR 203000 (SF2, SF4 RHR/LPCI) 7 Knowledge of RHR/LPCI: INJECTION 3.7 28 RHR/LPCI: Injection Mode MODE (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following: Emergency generator load sequencing 205000 (SF4 SGS) Shutdown Cooling 6

Ability to (a) predict the impacts of the 3.4 29 following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

SDC/RHR pump trips 209001 (SF2, SF4 LPCS) 8 Knowledge of the physical connections 3.2 30 Low-Pressure Core Spray and/or cause-effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following:

A.G. electrical power 209002 (SF2, SF4 HPCS) 1 Knowledge of the effect that a loss or 3.9 31 High-Pressure Core Spray malfunction of the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) will have on following: Reactor water level:

BWR-5,6 211000 (SF1 SLCS) Standby Liquid 5

Ability to manually operate and/or 4.1 32 Control monitor in the control room: Flow indication: Plant-Specific 212000 (SF? RPS) Reactor Protection 2

Knowledge of the operational 3.3 33 implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM: Specific logic arrangements 215003 (SF? IRM) 2 Knowledge of the effect that a loss or 3.6 34 Intermediate-Range Monitor malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM: 24/48 volt D.C. power:

Plant-Specific 215004 (SF? SRMS) Source-Range 1

Knowledge of electrical power supplies 2.6 35 Monitor to the following: SRM channels/detectors 215005 (SF? PRMS) Average Power 4

Knowledge of the operational 2.9

36.

Range Monitor/Local Power Range implications of the following concepts as Monitor they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: LPRM detector location and core symmetry 215005 (SF? PRMS) Average Power 5

Knowledge of the operational 3.6 37 Range Monitor/Local Power Range implications of the following concepts as Monitor they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: Core flow effects on APRM trip setpoints

ES-401 6

Form ES-401-1 217000 (SF2, SF4 RCIC) Reactor 2

Ability to monitor automatic operations 3.6 38 Core Isolation Cooling of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) including:

Turbine startup 218000 (SF3 ADS) Automatic 3

Knowledge of the physical connections 3.7 39 Depressurization and/or cause-effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following: Nuclear boiler instrument system 209001 (SF2, SF4 LPCS) 5 Knowledge of the effect that a loss or 2.8 40 Low-Pressure Core Spray malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM: EGGS room cooler(s) 223002 (SF5 PCIS) Primary 1

Ability to monitor automatic operations 3.4 41 Containment Isolation/Nuclear Steam of the PRIMARY CONTAINMENT Supply Shutoff ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including:

System indicating lights and alarms 239002 (SF3 SRV) Safety Relief 3

Ability to predict and/or monitor changes 2.8 42 Valves in parameters associated with operating the RELIEF/SAFETY VALVES controls including: Air supply: Plant-Specific 259002 (SF2 RWLCS) Reactor Water 2.1.28 Knowledge of the purpose and function 4.1 43 Level Control of major system components and controls 261000 (SF9 SGTS) Standby Gas 3

Ability to (a) predict the impacts of the 2.9 44 Treatment following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High train temperature 262001 (SF6 AC) AC Electrical 1

Knowledge of the effect that a loss or 3.5 45 Distribution malfunction of the A.G. ELECTRICAL DISTRIBUTION will have on following:

Major system loads 262002 (SF6 UPS) Uninterruptable 1

Ability to (a) predict the impacts of the 2.6 46 Power Supply (AC/DC) following on the UNINTERRUPTABLE POWER SUPPLY (A.C./0.C.); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Under voltage 262002 (SF6 UPS) Uninterruptable 2.1.30 Ability to locate and operate 4.4 47 Power Supply (AC/DC) components, including local controls.

263000 (SF6 DC) DC Electrical 2

Ability to manually operate and/or 3.2 48 Distribution monitor in the control room: Battery voltage indicator: Plant-Specific 264000 (SF6 EGE) Emergency 5

Knowledge of EMERGENCY 3.2 49 Generators (Diesel/Jet) EOG GENERATORS (DIESEUJET) design feature(s) and/or interlocks which provide for the following: Load shedding and sequencing

ES-401 300000 (SF8 IA) Instrument Air 300000 (SF8 IA) Instrument Air 400000 (SF8 CCS) Component Cooling Water 400000 (SF8 CCS) Component Cooling Water 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray 215004 (SF7 SRMS) Source-Range Monitor 218000 (SF3 ADS) Automatic Depressurization 262001 (SF6 AC) AC Electrical Distribution 263000 (SF6 DC) DC Electrical Distribution KIA Category Point Totals:

7 2

13 2

4

.5.*

Form ES-401-1

  • *,,; ;
  • Knowledge of the effect that a loss or malfunction of the INSTRUMENT AIR
  • . *. SYSTEM will have on the following:

Systems having pneumatic valves and

. /* controls

      • Knowledge of the effect that a loss or
  • . malfunction of the following will have on 1:: ;*'..;

the INSTRUMENT AIR SYSTEM: Filters Ability to predict and I or monitor

.*. changes in parameters associated with operating the CCWS controls including:

,,,, CCW temperature

.. ':" Knowledge of the physical connections and I or cause-effect relationships

,. * ** *** between CCWS and the following:

Reactor coolant system in order to

  • determine source(s) of RCS leakage
  • :, :1 into CCWS Ability to (a) predict the impacts of the

. *., following on the HIGH PRESSURE

. CORE SPRAY SYSTEM (HPCS);and (b) based on those predictions, use procedures to correct, control, or

. mitigate the consequences of those

  • abnormal conditions or operations:

Inadequate system flow: BWR-5,6 Ability to (a) predict the impacts of the following on the SOURCE RANGE MONITOR (SRM) SYSTEM; and (b)

  • . * :* based on those predictions, use

., procedures to correct, control, or

.:..*1 mitigate the consequences of those abnormal conditions or operations:

. Faulty or erratic operation of

., detectors/system

.2.4.9 Knowledge of low power/ shutdown

  • ** implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

2.1.32 Ability to explain and apply all system

  • *
  • limits and precautions.

' **** Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL

\\* DISTRIBUTION; and (b) based on those

' predictions, use procedures to correct,

,. control, or mitigate the consequences of

  • ,. ; those abnormal conditions or

. operations: Grounds 3.3 50 2.8 51 2.8 52 2.9 53 3.2 86 3.5 87 4.2 88 4.0 89 3.2 90 3

1 3

2 3

3 2

3/3. 2 2

  • 212 Group Point Total:

26/5 II

ES-401 8

Form ES-401-1 IES-401 BWR Examination Outline Form ES-401-1 Plant Svstems Tier 2/Grouo 2 (RO/SROl Svstem # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4

.. G*~*:.,,

KIA Tooic(sl IR 201003 (SF1 CROM) Control Rod and Drive

.:,., '. Knowledge of annunciators 4.2 54 2.4.31 alarms, indications or response Mechanism

',,,"",*1

" orocedures 202002 (SF1 RSCTL) Recirculation Flow 6

  • Knowledge of the effect that a 2.9 55 Control loss or malfunction of the

' following will have on the

  • RECIRCULATION FLOW

,*,; CONTROL SYSTEM:

Reactor/turbine pressure

.,,. reaulatina system: Plant-Specific 215002 (SF7 RBMS) Rod Block Monitor 4

  • ' ?,

Ability to monitor automatic 3.6 56

  • ' operations of the ROD BLOCK
MONITOR SYSTEM including:

Verification or proper

' functioning/ operability: BWR-

,* Knowledge of the effect that a 3

' loss or malfunction of the 3.4 57 Auxiliaries

, PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES

'/ will have on following:

  • .e, "',

Containment/drywall pressure:

,,, Plant-Soecific 234000 (SF8 FH) Fuel-Handling Equipment 1

i Knowledge of the physical 3.2 58 1~.,

connections and/or cause-effect t

relationships between FUEL HANDLING EQUIPMENT and the followina: Fuel 241000 (SF3 RTPRS) Reactor/Turbine Ability to manually operate 2.9 59 13 Pressure Regulating and/or monitor in the control

.., room: Turbine inlet oressure 245000 (SF4 MTGEN) Main Turbine 10

')

. Knowledge of MAIN TURBINE 2.6 60 Generator/Auxiliary GENERATOR AND AUXILIARY

\\

SYSTEMS design feature(s) and/or interlocks which provide for the following: Extraction steam 259001 (SF2 FWS) Feedwater 6

' Ability to predict and/or monitor 2.7 61

  • changes in parameters

" associated with operating the REACTOR FEEDWATER SYSTEM controls including:

Feedwater heater level 268000 (SF9 RW) Radwaste

1
  • Ability to (a) predict the impacts 2.9 62

'1, of the following on the

  • .: RADWASTE; and (b) based on
./
    • those predictions, use

,. procedures to correct, control, or mitigate the consequences of those abnormal conditions or I,

ooerations: System rupture

.-... '* Knowledge of the operational 272000 (SF7, SF9 RMS) Radiation Monitoring 1

implications of the following 3.2 63

,. concepts as they apply to RADIATION MONITORING

  • , SYSTEM: Hydrogen injection operations effect on process radiation indications: Plant-

,,,, 1,'-

  • Specific

ES-401 9

Form ES-401-1 286000 (SF8 FPS) Fire Protection 2

Knowledge of electrical power 2.9 64 suoolies to the followina: Pumps 288000 (SF9 PVS) Plant Ventilation 2.1.7 Ability to evaluate plant 4.4 65 performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

201006 (SF? RWMS) Rod Worth Minimizer 4

  • Ability to (a) predict the impacts 3.3 91 of the following on the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC); and (b) based on those prepictions, use procedures to correct, control, or mitigate the consequences of those
  • a abnormal conditions or
a operations: Stuck rod: Plant Specific (Not-BWR6) 230000 (SF5 AHR SPS) RHR/LPCI:

10 Ability to (a) predict the impacts 3.0 92 Torus/Suppression Pool Spray Mode of the following on the RHR/LPCI:

TORUS/SUPPRESSION POOL SPRAY MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Nuclear boiler instrument failures 245000 (SF4 MTGEN) Main Turbine 2.1.32 Ability to explain and apply all 4.0 93 Generator/Auxiliary system limits and precautions.

KIA Cateaorv Point Totals:

1 1

1 1

1 1

1 1/2 1 1

2/1 Grouo Point Total:

12/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility: Nine Mile Point Unit 2 Date of Exam:

Category KIA#

Topic RO SRO-only IR IR 2.1.26 Knowledge of industrial safety procedures (such as rotating 3.4 66 equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydroqen).

2.1.36 Knowledge of procedures and limitations involved in core 3.0 67 alterations

1. Conduct of 2.1.43 Ability to use procedures to determine the effects on reactivity 4.1 68 of plant chanoes Operations 2.1.26 Knowledge of industrial safety procedures (such as rotating 3.6 94 equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydroqen).

Subtotal 3

1 2.2.14 Knowledge of the process for controlling equipment 3.9 69 confiauration or status.

2.2.44 Ability to interpret control room indications to verify the status 4.2 70 and operation of a system, and understand how operator actions and directives affect plant and system conditions

2. Equipment Control Knowledge of the process for managing maintenance activities 2.2.18 durinq shutdown operations.

3.9 95 2.2.25 Knowledge of the bases in Technical Specifications for limiting 4.2 96 conditions for operations and safety limits Subtotal 2

2 2.3.12 Knowledge of radiological safety principles pertaining to 3.2 71 licensed operator duties 2.3.15 Knowledge of radiation monitoring systems 2.9 72

3. Radiation 2.3.5 Ability to use radiation monitoring systems, such as fixed 2.9 97 Control radiation monitors and alarms, portable survey instruments, personnel monitorinq equipment, etc.

2.3.14 Knowledge of radiation or contamination hazards that may 3.8 98 arise during normal, abnormal, or emergency conditions or activities.

Subtotal 2

2 2.4.20 Knowledge of operational implications of EOP warnings, 3.8 73 cautions and notes.

2.4.27 Knowledge of "fire in the plant" procedures.

3.4 74 2.4.39 Knowledge of the RO's responsibilities in emergency plan 3.9 75 implementation.

4. Emergency Procedures/Plan 2.4.11 Knowledge of abnormal condition procedures 4.2 99 2.4.44 Knowledge of emergency plan protective action 4.4 100 recommendations.

Subtotal 3

2 Tier 3 Point Total 10 7

ES-401 Tier/

Group Randomly Selected KIA Record of Rejected K/ As Form ES-401-4 Reason for Rejection

~.. **.

The following topics/ K/As were excl.ud~d from the systematic and *random sampling process:

1/1 1/2 2/1 2/1 2/2 2/2 Rejected Emergency and Abnormal Plant Evolution 295027 (EPE 4) High Containment Temperature (Mark Ill Containment Only). Nine Mile Point 2 containment design is Mark II.

Rejected Emergency and Abnormal Plant Evolution 295011 (APE 11) High Containment Temperature (Mark Ill Containment Only). Nine Mile Point 2 containment design is Mark II.

Rejected Plant System 207000 (SF4 IC) Isolation (Emergency) Condenser. Nine Mile Point 2 has no isolation condenser.

Rejected Plant System 206000 (SF2, SF4 HPCIS) High Pressure Coolant Injection. Nine Mile Point 2 is a BWR 5 and has no HPCI. HPCS instead (209002) was included in the random drawing.

Rejected Plant System 201004 (SF7 RSCS) Rod Sequence Control. Nine Mile Point 2 has no RSCS.

Rejected Plant System 201005 (SF1, SF7 RCIS) Rod Control and Information. Nine Mile Point 2 has no RCIS system.

The following Kl As wete resampled. l;)y,the.NRC during the initial draftin~:i00he outline: *.

1/1 1/2 2/1 295025 EK2.06 RO Question #12 295034 G2.2.4 SRO Question #85 262002 A2.04 RO Question #46 Rejected 295025 EK2.06 - Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:

HPCI Plant Specific. NMP2 does not have HPCI.

Randomly reselected 295025 EK2.01 - Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: RPS.

Rejected this KIA as this KIA is for a multi-unit site. NMP2 is considered a single unit site. Randomly reselected G2.2.36 Rejected this KIA as it is specific to a BWR 1. NMP2 is a BWR5.

Randomly reselected 262002 A2.01 - Under voltage

ES-401 2/2 3

3 3

234000 K1.07 RO Question #58 G2.3.12 SRO Question #97 G2.3.15 SRO Question #98 G2.4.20 SRO Question #99 Record of Rejected K/ As Form ES-401-4 Rejected this KIA as it is specific to a Mark Ill containment.

NMP2 is a Mark II containment design.

Randomly reselected 234000 K1.01 - Fuel Rejected KIA due to oversampling. This is the same KIA as RO question #71 Randomly reselected G2.3.5 Rejected KIA due to oversampling. This is the same KIA as RO question #72 Randomly reselected G2.3.14 Rejected KIA due to oversampling. This is the same KIA as RO question #73 Randomly reselected G2.4.11

.The toilowin*g K/As wer~.resampled ;dur:ing corisfr:u'ction* cifthe written"exafo: * *...

1 / 1 1 / 1 Question #6 295016 Control Room Abandonment AK2.03 - Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:

Control room HVAC Question #7 295018 Partial or Complete Loss of ccw AK3.04 - Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

Starting standby pump An acceptable question could not be developed without testing minutia due to limited interrelation between Control Room HVAC and Control Room Abandonment.

Randomly resampled KIA 295016 Control Room Abandonment AK2.01 - Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: Remote shutdown panel: Plant-Specific.

An acceptable question could not be developed at a discriminating RO level due to the simplicity of the KIA and limited plant-specific guidance.

Randomly resampled KIA 295018 Partial or Complete Loss of CCW AK3.07 - Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

Cross-connecting with backup systems.

ES-401 Record of Rejected K/As Form ES-401-4 Question #11 An acceptable question could not be developed due to lack 295024 High Drywell of operationally relevant use of Drywell radiation levels related to high Drywell pressure at the RO job level. Also Pressure would have potentially overlapped with Question #97.

EA2.08 - Ability to Randomly resampled KIA 295024 High Drywell Pressure determine and/or EA2.04 - Ability to determine and/or interpret the following 1 / 1 interpret the as they apply to HIGH DRYWELL PRESSURE:

following as they Suppression chamber pressure: Plant-Specific.

apply to HIGH DRYWELL PRESSURE:

Drywell radiation levels Question #16 An acceptable question could not be developed due to 295031 Reactor Low mismatch between the generic KIA and the given Water Level evolution. Also, this generic KIA is already used on Question #85.

2.2.36 - Ability to Randomly resampled KIA 295031 Reactor Low Water analyze the effect of Level 2.4.18 - Knowledge of the specific bases for EOPs.

1 / 1 maintenance activities, such as degraded power sources, on the status of limiting conditions of operations Question #25 An acceptable and discriminating question could not be 295032 High developed without testing minutia due to limited operationally relevant tie between Fire Protection and the Secondary given evolution. Additionally, N2-EOP-SC concepts are

  • Containment Area oversampled on the exam (Questions #26, #27, #84, #85),

Temperature preventing development of a question without overlap.

EK2.03 - Knowledge Randomly resampled KIA 500000 High Containment of the interrelations Hydrogen Concentration EK2.03 - Knowledge of the 1/2 between HIGH interrelations between HIGH CONTAINMENT HYDROGEN SECONDARY CONCENTRATIONS the following: Containment CONTAINMENT AREA Atmosphere Control System.

TEMPERATURE and the following:

Fire protection system

ES-401 Record of Rejected K/ As Form ES-401-4 Question #40 ADS I SRV concepts are oversampled (Questions #39, 40, 218000 ADS 42, 88), with 218000 triple sampled. The Kl As for #40 and

  1. 42 in particular were causing direct overlap issues.

K6.05 - Knowledge Randomly resampled KIA 209001 Low Pressure Core of the effect that a Spray K6.05 - Knowledge of the effect that a loss or loss or malfunction 2 I 1 of the following will malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM: ECCS room have on the cooler(s).

AUTOMATIC DEPRESSURIZATI ON SYSTEM: A.C.

power: Plant-Specific Question #44 An acceptable question could not be developed because 261000 Standby "high fuel pool ventilation radiation" does not have an Gas Treatment impact at NMP2.

A2.12 -Ability to (a)

Randomly*resampled KIA 261000 Standby Gas Treatment A2.03 - Ability to (a) predict the impacts of the following on predict the impacts the STANDBY GAS TREATMENT SYSTEM; and (b) of the following on based on those predictions, use procedures to correct, the STANDBY GAS control, or mitigate the consequences of those abnormal TREATMENT SYSTEM; and (b) conditions or operations: High train temperature.

based on those 2 I 1 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal r

conditions or operations: High fuel pool ventilation radiation: Plant-Specific

ES-401 Record of Rejected K/ As Form ES-401-4 Question #47 An acceptable question could not be developed because 262002 there are no Immediate Actions related to this system at Uninterruptable NMP2.

Power Supply Randomly resampled KIA 262002 Uninterruptable Power (AC/DC)

Supply (AC/DC) 2.1.30 - Ability to locate and operate 2.4.49 - Ability to components, including local controls.

2 / 1 perform without reference to procedures those actions that require immediate operation of system components and controls Question #57 An acceptable question could not be developed without 223001 Primary overlapping/oversampling concepts tested in Questions Containment and

  1. 42 and #50.

Auxiliaries Randomly resampled KIA 223001 Primary Containment K3.08 - Knowledge and Auxiliaries K3.03 - Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT SYSTEM of the effect that a AND AUXILIARIES will have on following:

loss or malfunction Containment/drywell pressure: Plant-Specific.

of the PRIMARY 2/2 CONTAINMENT SYSTEM AND AUXILIARIES will have on following:

Pneumatically operated valves internal to containment/drywell:

Plant-Specific Question #65 An acceptable question could not be developed due to lack 288000 Plant of applicable reference materials (graphs, curves, tables, Ventilation etc.) related to Plant Ventilation systems.

2.1.25 - Ability to Randomly resampled KIA 288000 Plant Ventilation 2.1.7 -

2/2 Ability to evaluate plant performance and make operational interpret reference judgments based on operating characteristics, reactor materials, such as behavior, and instrument interpretation.

graphs, curves, tables, etc.

ES-401 Record of Rejected Kl As Form ES-401-4 Question #66 The given generic KIA is already used on the SRO exam 2.1 ;32 - Ability to and is not "generic" in nature, as preferred for a Tier 3 explain and apply all question (would lead to a Tier 2 question).

3 system limits and Randomly resampled KIA 2.1.26 - Knowledge of industrial precautions.

safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

Question #69 An acceptable question could not be developed that was 2.2.23 - Ability to both discriminating and at the RO level.

track Technical Randomly resampled KIA 2.2.14 - Knowledge of the 3

Specification limiting process for controlling equipment configuration or status.

conditions for operations.

Question #77 An acceptable question could not be developed for this 295016 Control evolution with the given generic KIA without testing minutia Room Abandonment due to limited relevant surveillance procedures.

1 / 1 2.2.12 - Knowledge Randomly resampled KIA 295016 Control Room Abandonment 2.4.34 - Knowledge of RO tasks performed of surveillance outside the main control room during an emergency and procedures the resultant operational effects.

Question #79 An acceptable question could not be developed for this 295019 Partial or evolution with the given generic KIA due to lack of related Complete Loss of Immediate Actions.

Instrument Air Randomly resampled KIA 295019 Partial or Complete Loss 2.4.49 - Ability to of Instrument Air 2.4.4 - Ability to recognize abnormal indications for system operating parameters that are entry-perform without level conditions for emergency and abnormal operating 1 / 1 reference to procedures.

procedures those actions that require immediate operation of system components and controls

ES-401 Record of Rejected K/ As Form ES-401-4 Question #85 An acceptable question could not be developed for this 295034 Secondary evolution with the given generic KIA without being overly similar to Question #84 (both related to N2-EOP-SC and Containment Technical Specifications).

Ventilation High Radiation Randomly resampled KIA 295034 Secondary Containment 2.2.36 - Ability to Ventilation High Radiation 2.4.6 - Knowledge of EOP analyze the effect of mitigation strategies.

1/2 maintenance activities, such as degraded power sources, on the status of limiting conditions for operations Question #87 A discriminating question could not be developed for the 215004 Source original KIA due to lack of SRO level material to test for just Range Monitor a failed recorder (other instruments still satisfy all requirements, just need to get maintenance to fix recorder).

A2.06 - Ability to (a)

Randomly resampled KIA 215004 Source Range Monitor predict the impacts A2.05 - Ability to (a) predict the impacts of the following on of the following on the SOURCE RANGE MONITOR (SRM) SYSTEM; and (b) the SOURCE based on those predictions, use procedures to correct, RANGE MONITOR (SRM) SYSTEM; control, or mitigate the consequences of those abnormal conditions or operations: Faulty or erratic operation of 2 I 1 and (b) based on detectors/system.

those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failed recorder

ES-401 Record of Rejected K/ As Form ES-401-4 Question #59 An acceptable question could not be developed that was 241000 both discriminating and not minutia.

Reactor/Turbine Randomly resampled KIA 241000 Reactor/Turbine Pressure Regulating Pressure Regulating A4.13 - Ability to manually operate 2/2 A4.12 - Ability to and/or monitor in the control room: Turbine inlet pressure.

manually operate and/or monitor in the control room:

Turbine acceleration

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