ML12165A601

Final Outlines (Folder 3)

ML12165A601
Person / Time
Site:	Oyster Creek
Issue date:	06/06/2012
From:	D'Antonio J M Operations Branch I
To:	Exelon Nuclear
Jackson D E
Shared Package
ML120230007	List: IR 05000219/2012301 ML120250519 ML12128A224 ML12159A244 ML12159A252 ML12159A258 ML12159A391 ML12165A601 ML12165A609 ML12165A635 ML12172A069
References
TAC U01848
Download: ML12165A601 (25)

Text

Administrative Outline Form ES-301-1 Facility:

Creek Date of Examination:

Examination Level: RO [8J SRO 0 Operating Test Number: 11-1 Administrative Topic Type Describe activity to be performed (See Note) Code" Calculate Identified Leak Rate lAW 351.2; 2.1.20 (4.6) Conduct of M,R [NRC RO Admin JPM 1:1 Perform Core Thermal Limit Verification; 2.1.7 (4.4) [NRC Conduct of P,R RO Admin 2] Determine Vortex and NPSH Impacts on the Core Spray Equipment D,R System; 2.2.44 (4.2) [NFiC RO Admin ,JPM 3] Radiation Control Review a Completed State/Local Notification Form; 2.4.39 Emergency M,R (3.9) [NRC RO Admin JPM 4] All items (5 total) are required for SROs. RO applicants require only 4 items unless they retaking only the administrative topics, when 5 are " Type Codes & (C)ontrol room, (S)imulator, or (D)irect from bank (:s. 3 for ROs;.=:; 4 for SROs & RO (N)ew or (M)odified from bank (P)revious 2 exams Gs. 1; randomly ES 301, Page 22 of 27 JPM# RO Admin JPM 1 Conduct of Ops RO Admin JPM 2 Conduct of Ops RO Admin JPM Equipment RO Admin JPM ILT 11-1 NRC RO ADMIN "IPM Summary The applicant will be given conditions of the Drywell Equipment Drain Tank integrator being out of service. The must manually calculate the Primary Containment leak rate lAW 351.2, High Purity Waste System, and determine that it exceeds Technical Specification limits. The applicant will pertorm Core Thermal Limits Verification lAW 202.1-3 section 1.0, Perform Shiftly Core Thermal Limits Verification.

After performing the attachment and reviewing a printout of the Reactor Core State Parameters from the PPC, the applicant must determine that MAPRAT and FLLLP are unacceptable.

The applicant must evaluate plant parameters and determine Core Spray System Vortex and NPSH limits lAW SP-4, Operation of the Core Spray System, and state what actions are required per the Support Procedure.

The candidate will review a completed EP-MA-114-1 03, State / Local Notification Form. There will be several errors and incomplete items on the form. The candidate will document those items and also state that the form is NOT ready to be faxed.

Administrative Topics Outline Form ES-301-1 Facility:

Oyster Creek Date of Examination:

Examination Level: RO D SRO [gI Operating Test Number: 11-1 Administrative Topic (See Note) Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan Type Code* Describe aGtivity to be performed D,R Review / Approve a Completed Reactor Heat Balance; 2.1.7 (4.7) [NRC SRO Admin JPM 1] D,R Review Request to Allow LPRM (input into APRM) Bypass lAW 403; (4.5)

SRO Admin JPM 2] D,R Review Completed Surveillance Procedure 610.3.105 (Core Spray Sys 1 Inst Cal and Operability);

2.2.12 (4.1) [NRC SRO Admin JPM 3] M,R Authorize Emergency Exposures lAW EP-AA-113; 2.3.4 (3.7) [NRC SRO Admin JPM 4] M,R Determine Primary Containment Water Level lAW SP28 and Determine Required Action; 2.4.21 (4.6) [NRC SRO Admin JPM 5] All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)irnulator, or Class(R)oom (D)irect from bank Gs. 3 for ROs;.::: 4 for SROs & RO retakes) (N)ew or (M)odified from bank 1) (P)revious 2 exams (.:5. 1; randomly selected)

ES 301, Page 22 of 27 JPM# SRO Admin JPM Conduct of SRO Admin JPM Conduct of SROAdmin ..IPM Equipment SRO Admin JPM Radiation SRO Admin JPM ILT 11-1 NRC Exam SRO ADMIN JPM

SUMMARY

Summary The applicant will review/approve a completed manual heat balance lAW 1001.6. The applicant will discover an error, which when corrected, will place the thermal heat balance above the licensed limit. The applicant will then direct that actual reactor power be lowered to less than the licensed limit. The applicant will review a work package requesting the bypass of an APRM. Attachment 2 of procedure 403 will show that the requested APRM cannot be bypassed due to inoperability of an APRM in the same RPS channel (2 LPRMs in the same string will be inoperable/bypassed from the APRM). The applicant will review a completed surveillance test, 610.3.015, Core Spray System -I Instrument Calibration and Operability.

The data sheets will show that both the Drywell high pressure instruments which input into Core Spray System 1, will not meet the procedural requirements and will be declared inoperable.

The applicant will review/apply Tech Table 3.1.1 and 3.4 for the impact of the instrument inoperability.

The applicant will approve or not approve the issuance of KI to emergency workers lAW procedure EP-AA-113.

The applicant will evaluate plant parameters and calculate the Primary Containment Water Level lAW EMG-SP28, Determining Primary Containment Water Level. The applicant must correctly calculate the Primary Containment Water Level (allowing tolerance for minor rounding errors). The applicant must also state that Drywell Sprays must be terminated lAW the Primary Containment Control EOP due to Primary Containment Water Level being greater than 348 in.

Control Room/ln*Plant Systems Outline Form ES-301*2 Facility:

Oyster Creek Date of Examination:

05/14/2012 Exam Level: RO D SRO-I D SRO-U IZI Operating Test Number: 11-1 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) Safety System / JPM Type Code* Function a. Place a second RWCU Pump in service with a high temperature alarm M,A,S 2 and isolation failure lAW 303 (Alternate Path); 204000 A4.01 (3.1/3 .. 0) [NRC Sim JPM 21 c. d. e. Restore 4160VAC Bus 1C to normal with EDG-1 supplying power D,A, L. EN, 6 (Alternate Path); 264000 A4.04 (3.?/3.?)

rNRC Sim JPM 61 S g. In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for Control CRD in the plant Post-Scram lAW SP-3; 201001 A1.03 (2.9/2.8)

D,L,R,E 1 [NRC Plant JPM 1] Line up to vent the Torus through the Hardened Vent lAW SP-35; 295024 D,E 5 EA 1.14 (3.4/3.5)

[NRC Plant JPM 21 Bypass the Air Dryers and the Pre/Post Filters; 300000 A2.01 (2.9/2.8)

D.R 8 [NRC Plant JPM 3] I

@ 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 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator I Criteria for RO 1 SRO-II SRO-U 4-61 4-6 I .::.8 1 1 1 !1 I -I -/ ! 1 (control room !1/ .::. 1 / ! !2 I ! 31 I 2 (randomly ! 1 I I ES-301, Page 23 of Control Room/ln*Plant Systems Outline Form ES*301*2 Facility:

Creek Date of Examination:

05/14/2012 Exam Level: RO 0 SRO-I [gI SRO-U 0 Operating Test Number: 11-1 NRC Control Room Systems@ (B for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / ,IPM Title Perform Recirculation Pump Trip Circuitry Test lAW 603.4.001 with Multiple Recirculation Pumps Trip (Alternate Path): 202001 A2.04 (3.7/3.8)

[NRC Sim JPM 1] Place a second RWCU Pump in service with a high temperature alarm and isolation failure lAW 303 (Alternate Path); 204000 A4.01 (3.1/3.0)

[NRC Sim JPM 2] Shutdown of the Automatic Depressurization System lAW 30B; 21BOOO A4.03 (4.2/4.2)

[NRC Sim JPM 31 d. Perform Core Spray Surveillance with faulted Core Spray Pump lAW 610.4.002 (Alternate Path); 209001 A4.01 (3.B/3.6)

[NRC Sim JPM 4] Purging the Primary Containment with Elevated Stack Radiation (Alternate Path); 223001 A4.07 (4.2/4.1)

[NRC Sim JPM 5] Restore 4160VAC Bus 1C to normal with EDG-1 supplying power (Alternate Path); 264000 A4.04 (3.7/3.7)

[NRC Sim JPM 6] Swap Instrument Air Compressors (Alternate Path); 300000 K4.04 (2.8/2.9)

[NRC Sim JPM 7] h. In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) Control CRD in the plant Post-Scram lAW SP-3; 201001 A1.03 (2.9/2.8)

[NRC Plant ,IPM 1] Line up to vent the Torus through the Hardened Vent lAW SP-35; 295024 EA 1.14 (3.4/3.5)

[NRC Plant JPM 2] Bypass the Air Dryers and the Pre/Post Filters; 300000 A2.01 (2.9/2.8)

[NRC Plant JPM 3] Type Code* Safety Function P,A,S 1 M,A,S 2 D,EN,S 3 P,A,S 4 P,A,EN,S 5 D, A, L, EN, 6 S N,A,S 8 D,L,R,E 1 D,E 5 D,R 8 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 Criteria for RO I SRO-II SRO-U (A)lternate path 4-6/ 4-6 I 2-3 (C)ontrol room (D)irect from bank I (E)mergency or abnormal in-plant 1 I ::::1 I (EN)gineered safety feature -I -I :::: 1 (control room system (L)ow-Power I Shutdown :::: 1 / ::::1 I ::::1 (N)ew or (M)odified from bank including 1 (A) ::::2 I ::::1 (P)revious 2 exams I 2 (randomly selected) (R)CA :::: 1 / ::::1 I :::: 1 {S)imulator ES-301, Page of 27 Control Room/ln*Plant Systems Outline Form ES*301*2 Facility:

Creek Date of Examination:

Exam Level: RO [gI SRO-I 0 SRO-U 0 Operating Test Number: 11-1 Control Room Systems@ (B for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Perform Recirculation Pump Trip Circuitry Test lAW 603.4.001 Multiple Recirculation Pumps Trip (Alternate Path); 202001 (3.7/3.B)

[NRC Sim JPM Place a second RWCU Pump in service with a high and isolation failure lAW 303 (Alternate Path); 204000 A4.01

[NRC Sim JPM Shutdown of the Automatic Depressurization System lAW 308; 218000 A4.03 (4.2/4.2)

[NRC Sim JPM 3] . d. Perform Core Spray Surveillance with faulted Core Spray Pump lAW 610.4.002 (Alternate Path); 209001 A4.01 (3.B/3.6)

[NRC Sim JPM 4] Purging the Primary Containment with Elevated Stack (Alternate Path); 223001 A4.07 (4.214.1)

[NRC Sim JPM Restore 4160VAC Bus 1C to normal with EDG-1 supplying (Alternate Path); 264000 A4.04 (3.7/3.7)

[NRC Sim JPM g. Swap Instrument Air Compressors (Alternate Path); 300000 (2.B/2.9)

[NRC Sim JPM Re-establishing Off-Gas System Flow after an Off-Gas Explosion; 271000 A2.06 3.5/3.9 [NRC Sim JPM In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. Control CRD in the plant Post-Scram lAW SP-3; 201001 A1.03 (2.9/2.8) [NRC Plant JPM 1] Line up to vent the Torus through the Hardened Vent lAW SP-35; 295024 EA1.14 (3.4/3.5)

[NRC Plant JPM 2] Bypass the Air Dryers and the PrelPost Filters; 300000 A2.01

[NRC Plant JPM Type Code* Safety Function P,A,S 1 M,A,S 2 D,EN,S 3 P,A,S 4 P,A,EN,S 5 0, A, L, EN, 6 S N,A,S B 0, L, S 9 D,L,R,E 1 D,E 5 D,R 8 All RO and SRO-l 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 Criteria for RO I SRO-II SRO-U (A)lternate path 4-6/ 4-6 I 2-3 (C)ontrol room (D)irect from bank :::8 I (E)mergency or abnormal in-plant :::1 I 1 (EN)gineered safety feature -I -I 1 (control room system (L)ow-Power I Shutdown I (N )ew or (M)odified from bank including 1 (A) :::2 I :::1 (P)revious 2 exams ::3 I :: 2 (randomly selected) (R)CA ::: 1 I :::1 I :::1 (5)imulator ES-301, Page 23 of 27 Written Examination Outline Form ES-40 1-1 Facility:

Oyster Creek Date of Exam: 05/14112 RO KiA Category Points SRO-Only Points Tier Group K 1 K 2 K 3 K 4 A 1 A 2 A 3 A 4 G

• Total A2 G* Total 1. 1 3 4 3 3 4 I 3 20 3 4 7 Emergency

& 2 1 1 1 2 1 I 7 2 1 3 Plant Evolutions Tier Totals 4 5 4 5 5 4 27 5 5 10 I 2 3 3 2 3 2 2 2 2 3 2 26 2 3 5 2. 2 2 1 1 1 1 1 1 1 1 1 1 12 0 2 1 3 Plant Systems Tier 4 4 4 3 4 3 3 4 3 38 4 4 8Totals I 2 3 4 1 2 3 4 3. Generic Knowledge

& Abilities 7 Categories 3 2 3 2 2 2 2 ) Note: Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA category shall not be less than two). 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 +/-l from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section D.I.b of ES-401, for guidance regarding elimination of inappropriate KIA statements. 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. Absent a plant specific priority, only those KAs 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. Select SRO topics for Tiers I and 2 from the shaded systems and KIA categories. The generic (G) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D .l.b of ES-40 1 for the applicable KIA's On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (lR) for the applicable license level, and the point totals (#) for each system and category.

Enter the group and tier totals for each category in the table above. If fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams. For Tier 3, select topics from Section 2 of the KIA Catalog, and enter the K/A numbers, descriptions, IRs, and Eoint totals (#) on Form ES-401-3.

Limit SRO selections to KlAs that are linked to lOCFR55.43 2 Form ES-401-1 ILT 11-1 NRC Written Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE # / Name Safety KIA Topic(s) Imp. Q# AA2.05 -Ability to determine and/or 295023 Refueling Acc Cooling interpret the following as they apply to 4.6 I Mode REFUELING ACCIDENTS:

Entry conditions of emerl!;ency plan EA2.04 -Ability to determine and/or 295031 Reactor Low Water interpret the following as they apply to 4.8 2REACTOR I_OW WATER Il:VEL : Adequate core coolinl!;

AA2.02 -Ability to dcterminc and/or 295021 Loss of Shutdown interpret the following as they apply to 3.4 3LOSS OF SHUTDOWN COOLING: RHRlshutdown coolinl!;

system flow 2.2.25 -Equipment Control: Knowledge of 295026 Suppression Pool High X bases in technical specifications for limiting 4.2 4 Water Temp. 15 conditions for operations and safety limits. 2.1.23 -Conduct of Operations:

Ability to 295018 Partial or Total Loss of perform spcdfic system and integrated plant 4.4 5 CCW procedures during all modes of plant operation.

2.2.38 -Equipment Control: Knowledge of 295004 Partial or Total Loss of DC X conditions and limitations in the facility 4.5 6 Pwr/6 license. 2.4.29 -Knowledge of the emergency plan. 600000 Plant Fire On-site I 8 4.4 7 AKl.03 -Knowledge of the 295021 Loss of Shutdown Cooling implications of the following concepts 3.9 39 14 they apply to LOSS OF SHUTDOWN COOLING: Adequate core cooling EKl.02 -Knowledge of the operational 295030 Low Suppression Pool implications of the following concepts as 3.5 40 Water Levell 5 they apply to LOW SUPPRESSION POOL WATER LE VEL: Pump NPSH AKl.OI -Knowledge ofthe operational 295023 Refueling Acc Cooling implications of the following concepts as 3.6 41 Mode/8 they apply to REFUELING ACCIDENTS:

Radiation exposure hazards AK2.03 -Knowledge of the interrelations 600000 Plant Fire On-site I 8 between PLANT FIRE ON SITE and the 2.5 42 following:

Motors EK2.04 -Knowledge ofthe interrelations between HIGH REACTOR PRESSURE and 295025 High Reactor Pressure I 3 3.9 43 the following:

ARIIRPT/ATWS:

Plant-Specific AK2.02 -Knowledge of the interrelations 295006 SCRAM I 1 between SCRAM and the following:

3.8 44 Reactor water level control system AK3.01 -Knowledge ofthe reasons for the following responses as they apply to 700000 Generator Voltage and GENERATOR VOLTAGE AND 3.9 45 Electric Grid Disturbances ELECTRIC GRID DISTURBANCES:

Reactor and turbine trip criteria EK3.01 -Knowledge of the reasons for the following responses as they apply to 295037 SCRAM Conditions SCRAM CONDITION PRESENT AND Present and Reactor Power Above X REACTOR POWER ABOVE APRM 4.1 46 APRM Downscale or Unknown II DOWNSCALE OR UNKNOWN : Recirculation pump trip/fUnback:

295005 Main Turbine Generator AK3.04 -Knowledge of the reasons for 3.2 47 Trip following responses as they apply to MAIN 2 Form ES-401-1 IL T 11-1 NRC Written Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE # / Name Safety G KIA Topic(s) Imp. Q# TURBINE GENERATOR TRIP: Main generator trip AAI.03 -Ability to operate and/or monitor 295016 Control Room X the following as they apply to CONTROL 3.0 48 Abandonment / 7 ROOM ABANDONMENT:

RPIS EAI.17 -Ability to operate and/or monitor the following as they apply to HIGH 295024 High Drywell Pressure / 5 3.9 49 DRYWELL PRESSURE:

Containment spray: Plant-*Specific EAI.03 -Ability to operate and/or monitor 295028 High Drywell the following as they apply to HIGH3.9 50DRYWELL TEMPERATURE:

Drywell coolin" system AA2.01 -Ability to determine and/or 295001 Partial or Complete Loss interpret the following as they apply to of Forced Core Flow Circnlation / I X PARTIAL OR COMPLETE LOSS OF 3.5 51 &4 FORCED CORE FLOW CIRCULATION:

Powerlflow AA2.03 -Ability to determine 295004 Partial or Total Loss of interpret the following as they apply 2.8 52 DC PARTIAL OR COMPLETE LOSS OF D.C. POWER: Battery voltage EA2.01 -Ability to determine and/or 295031 Reactor Low Water interpret the following as they apply to 4.6 53/2 REACTOR LOW WATER LEVEL: Reactor water level 295026 Suppression Pool High 2.4.18 -Emergency Procedures

/ Plan: 3.3 54 Water Temp. Knowled"e of the specific bases for EOPs. 2.4.50 -Emergency Procedures

/ 295019 Partial or Total Loss of Ability to verify system alarm setpoints 4.2 55 Inst. Air / operate controls identified in the alarm response manual. 2.1.23 -Conduct of Operations:

Ability 295018 Partial or Total Loss of perform specific system and integrated 4.4 56 CCW procedures during all modes of plant operation.

AA2.02 -Ability to determine and/or interpret the following as they apply to 295003 Partial or Complete Loss X PARTIAL OR COMPLETE LOSS OF AC. 4.2 57 of AC /6 POWER: Reactor power. pressure, and level EK2.06 -Krowledge ofthe interrelations 295038 High Off-site Release between HIGH OFF-SITE RELEASE 3.4 58RATE and the following:

Process liquid radiation monitoring system KIA Category Totals: 3 4 3 3 4/3 3/4 Group Point Total: I 2017 3 Form ES-401-1 ILT 11-1 NRC Written Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE # I Name Safety Function K1 K2 K3 A1 A2 G KIA Topic(s) Imp. Q# AA2.03 Ability to determine 295009 Low Reactor Water Levell interpret the following as they apply 2.9 LOW REACTOR WATER Reactor water cleanup blowdown 2.4.47 -EIrergency Procedures 1295036 Secondary Containment Ability to diagnose and recognize trends 4.2 9 High SumplArea Water Level an accurate a nd timely manner utilizing appropriate control room reference AA2.04 Ab ility to determine 295014 inadvLTtent Reacti interpret the following as they apply to 4.4 10 Addition II INADVERTENT ADDITION:

Violation of fuel thermal AK1.03 -Knowledge of the implications of the following concepts as 295015 Incomplete SCRAM I I 59 they apply to INCOMPLETE SCRAM : 8 Reactivity effects AK2.08 Knowledge of tbe interrelations between INADVERTENT 295020 Inadvertent 2.5CONTAINMENT ISOLATION and the 60 Isolation 1 5 & 7 following:

Traversing in-core probes: Plant* Specific EK3.01 -Knowledge oflhe reasons forthe following responses as they apply to HIGH 295032 High 3.SECONDARY CONT AREA 61 Containment Area Temperature 5 TEMPERATURE:

EA 1.03 -Ability to operate and/or 295033 High Secondary the following as they apply to 3.Containment Area X SECONDARY CONTAINMENT AREA 62 8 Levels 19 RADIATION LEVELS:

containment AA2.01 Ability to determine interpret the lollowing as they apply to 295022 Loss of CRD Pumps I 1 63 LOSS OF CRD PUMPS: Accumulator 5 pressure 2.2.36 -Equipment Control: Ability 295034 Secondary Containment analyze the effect of maintenance activities, 64 Ventilation High Radiation 1 9 such as degraded power sources, on the I status of Iimiling conditions for operations.

AAl.lO -Ability to operate and/or monitor I Z High Off-site Release 3.X the following as they apply to HIGH OFF-65 6 SITE RELEASE RATE: RPS KIA Category Totals: I I 1 2 In Group Point Total: I 7/3 ES-401 4 Form ES-401-1 ILT 11-1 NRC Written Written Examination Plant Systems -Tier 2 Group KKK Imp System # I Name G 0#123 A2.06 -Ability to (a) predict the impacts of the following on the AC. ELECTRICAL 262001 AC Electrical Distribution X DISTRIBUTION; and (b) based on those predictions.

use procedures to 2.9 II correct. control. or mitigate tbe consequences of those abnormal conditions Of operations:

Deenergizing a plant bus A2.09 -Ahility to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (h) based on those 212000RPS X predictions, use procedures to 4.3 12 com:ct, control, or mitigate the consequences of those abnormal 207000 Isolation (Emergency)

Condenser 400000 Component Cooling Water X X coneiitions or operations:

High containmentldrywell pressure 2.2.40 -Equipment Control: Ability to apply Technical Specifications 4.7 13 for a system. 2.4." -Koow'o)" of ..ff conelition procedures.

2.2.22 -Equipment Control: 215005 APRM I LPRM X Knowledge of Limiting Conditions 15 for operations and satc-ty limits. KI.03 -Knowledge of the physical connections andlor cause-effect 259002 Reactor Water Level Control X relationships between REACTOR 3.8 I WATER LEVEL CONTROL SYSTEM and the following:

Reactor water level KI.OS -Knowledge of the physical connections and/or cause-effect relmionships between 205000 Shutdown Cooling X SHUTDOWN COOLING 3.1 2 SYSTEM (RHR SHUTDOWN COOLING MODE) and the following:

Component cooling watcr systems 2620() I AC Electrical Distribution X K2.01 -Knowledge of electrical power supplies to the fl'lll'm;",,'

3.3 3 K2.01 -Knowledge of eJ 215003 IRM X power supplies to the follow 4 IRM channel&/detectors K3.04 -Knowledge of the effect that a loss or malfunction of the 212000 RPS X REACTOR PROTECTION 3.3 5 SYSTEM will have on following:

Average power range monitoring system: Plant-Specific K3.01 -Knowledge of the effect that a loss or malfunction of tbe 300000 Instrument Air (INSTRUMENT AIR SYSTEM) 2.7 6 will have on the following:

Containment air system ES-401 4 Form ES-401-1 ILT 11-1 NRC Written Exam Written Examination Outline Plant Systems 2 Group 1 System # / Name K 1 K 2 K 3 K K K A A2 A 4 5 6 1 3 K4.CIl -Knowledge ofCCWS 400000 Component Cooling Water X design feature(s) and or interlocks which provide for the following:

3.4 7 Automatic start of standby pump K4.06 -Knowledge of EMERGENCY GENERATORS 264000EDGs X (DIESEL/JET) design feature(s) 2.6 8 and/or interlocks which provide for the following:

Governor control K5.01 -Knowledge of the operational implications of the 215004 Source Range Monitor X following concepts as they apply to SOURCE RANGE MONITOR 2.6 9 (SRlvl) SYSTEM: Detector operation K5.CIl -Knowledge of the operational implications of the 218000 ADS X following concepts as they apply to AUTOMATIC 3.8 10 DEPRESSURIZATION SYSTEM : ADS logic operation K6.01 -Knowledge of the effect that a loss or malfunction of the 262002 UPS (AC/DC) X following will have on the UNINTERRUPTABLE POWER 2.7 11 SUPPLY (A.C.ID.C.)

A.c. electrical power K6.07 -Knowledge of the effect that a loss or malfunction of the 207000 Isolation (Emergency)

Condenser X following will have on the ISOLATION (EMERGENCY) 3.0 12 CONDENSER:

A.C. power: BWIR-2,3 Al.01 -Ability to predict and/or monitor changes in parameters 263000 DC Electrical Distribution X associated with operating the D.C. ELECTRICAL DISTRIBUTION 2.5 13 controls including:

Battery charging/discharging rate Al.05 -Ability to predict and/or monitor changes in parameters associated with operating the 215005 APRM / LPRM X AVERAGE POWER RANGE MO:"IITORILOCAL POWER 3.3 14 RANGE MONITOR SYSTEM controls including:

Lights and alarms A2.05 -Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and 239002 SRVs X (b) based on those predictions, use procedures to correct, control, or 3.2 15 mitigate the consequences of those abnormal conditions or operations:

Low reactor pressure ES-401 4 Form ES-401-1 ILT 11-1 NRC Written Written Examination Plant Systems -Tier 2 Group

1. I Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A2 A 3 A 4 G I Imp I Q# I A2.06 -Ability to (a) predict the of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEMINUCLEAR 223002 PCISlNuclear Steam Supply Shutoff X STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, 3.0 16 or mitigate the consequences of those abnormal conditions or operations:

Containment instrumentation failures A3'(14 -Ability to monitor automatic operations of the 261000SGTS X STANDBY GAS TREATMENT 3.0 17 SYSTEM including:

System temperature A3JI2 -Ability to monitor 209001 LPCS X automatic operations of the LOW PRESSURE CORE SPRAY 3.8 18 SYSTEM including:

Pump start 211000 SLC X M.O! -Ability to

,000oc owll'OC Tank level A4.02 -Ability to 215004 Source Range Monitor X andlor monitor in SRI\! recorder 2.2.22 -Equipment Control: 205000 Shutdown Cooling X Know ledge of limiting conditions 4.0 21 for operations and sa 2.1.28 -Conduct of Operations:

239002 SRVs X Know ledge of the purpose and function of major system 4.1 22 components and controls.

KS.Ol Knowledge of the operational implications of the 263000 DC Electrical Distribution X following concepts as they apply to D.C ELECTRICAL 2.6 23 DISTRIBUTION

Hydrogen generation during battery charging.

A4.05 -Ability to manually operate 218000 ADS X andlor monitor in the control room: 4.2 24 ADS timer reset K2.02 -Knowledge of electrical 211000SLC power supplies to the following:

3.1 25 Explosive valves K3.05 -Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE 215005 APRM 1 LPRtvI X MONITOruLOCALPOWER 3.8 26 RANGE MONITOR SYSTEM will have on following:

Reactor power indication KIA Category Totals: 2 3 3 2 3 2 2 212 2 3 213 Group Point Total: I 26/5 5 Form ES-401-1 System # I Name 223001 Primary CTMT and Aux. 214000 RP[S 219000 RHR/LPCI:

Torus/Pool Cooling Mode 216000 Nuclear Boiler lnst. 256000 Reactor Condensate 239001 Main and Reheat Steam 272000 Radiation Monitoring 20l006RWM 286000 Fire Protection 290001 Secondary CTMT ILT 11-1 NRC Written Written Examination Plant Systems -Tier 2 Group 2 K K K K K K A A2 G 1 2 3 4 5 6 3 I A2.10 Ability to (a) predict tbe impacts of tbe foJlowing on !be PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES; and (b) based on those predictions, X use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

High drywell temperature 2.2.40 Equipment Control: Ability to apply Technical Specifications for a system. A2.l3 -AbilityLO (a) predict the impacts of the following on tbe RHRlI.PCI:

TORUS/SUPPRESSION POOL COOLING MODE; and (b) based X on those predictions, use proc.;dures to correct, control, Of mitigate !be consequences of those abnormal conditions or operations:

Hi"p suppression pool temperature Kl.l5 Knowledge of the physical connections and/or cause-effect relationships between NUCLEAR X BOILER INSTRUMENTATION and the following:

Isolation condenser:

Plant-Specific K2.01 Knowledge of electrical X power supplies to the following:

Systl!m pumps K3.16 Know ledge of tbe effect that a loss or malfunction of the MAlN AND REHEAT STEAM SYSTEM will have on following:

Relief/safety valves K4.01 Knowledge of RADIATION MONITORING design feature(s) and/or interlocks which provide for the fOllowing:

Redundancy K5.12 Knowledge of ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) design featurc(s) and/or interlocks which provide for the following:

Withdtaw block:

BWR6) K6.(11 Know ledge of the effect thai a loss or malfunction of the following will have on the FIRE X PROTECTION SYSTEM: A. C. electrical distribution:

Plant-Spedfic A1.01 Ability to predict and/of monitor changes in parameters associated with operating

!be SECONDARY CONTAINMENT controls including:

System lineups I Imp. II 3.8 16 4.7 17 3.7 18 3.9 27 2.7 28 3.6 29 2.7 30 3.5 31 3.1 32 3.1 33 ES-401 5 Form ES-40 1-1 ILT 11-1 NRC Written Written Examination Plant Systems -TiElr 2 Group System it I Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A2 A 3 A 4 G A2.03 -Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT 234000 Fuel Handling Equipment X ; and (b) based on those predictions, use procedures to correct, control, or mitigate thc cons':xtttcnces of those abnormal conditions or operations:

Loss of electrical power A3.02 -Ability to monitor automatic operations of the 201002 RMCS X REACTOR MANUAL CONTROL SYSTEM including:

Rod mowment seQttcnce lights A4.m Ability to manually operate 271000 Off-gas X and/or monitor in the control room: Reset system isolations 2.1.23 Conduct of Operations:

215001 Traversing In-core Probe X Abibty to perform specific system and integrated plant procedures during all modes of plant operation.

Kl.(}8 Know ledge of the physical connections and/or eause-effect 245000 Main Turbine Gen. / Aux. X relationships hetween MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS and the following:

Reactorlturbine pressure control system: Plant-Specific KIA Category Totals: 2 1 1 1 1 1 1 112 I 1 III Group Point Total: I Imp. II 2.8 34 2.8 35 2.8 36 4.3 37 3.4 38 12/3 I ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility:

IL T 11-1 NRC Written Exam Date: 05/14112 Category KIA # Topic RO IR Q# SRO-Only IR Q# 2.1.42 Knowledge of new and spent fuel moveme:nt procedures.

3.4 19 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. 4.2 24 1. Conduct of Operations 2.1.1 2.1.36 Knowledge of conduct of operations requirements.

Knowledge of procedures and limitations involved in core alterations.

3.8 3.0 66 67 2.1.38 Knowledge of the station's requirements for verbal communications when implementing rrocedures.

3.7 74 Subtotal 3 2 Ability to analyze the effect of maintenance 2.2.36 activities, such as degraded power sources, on the 4.2 20 status of limiting conditions for orerations.

2.2.19 Knowledge of maintenance work order requirements.

3.4 23 2. Equipment Control 2.2.13 Knowledge of tagging and clearance procedures.

4.1 68 2.2.37 Ability to determine operability and / or availability of safety related eguirment.

3.6 69 Subtotal 2 2 2.3.11 Ability to control radiation releases.

4.3 21 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

3.2 25 Knowledge of radiation monitoring systems, such 3. 2.3.15 as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring 2.9 70 Radiation equipment, etc. Control Ability to use radiation monitoring systems, such 2.3.5 as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring 2.9 71 equipment, etc. Knowledge of radiation or containment hazards 2.3.14 that may arise during normal, abnormal, or 3.4 75 emergency conditions or activities.

Subtotal 3 2 ES-401 Generic Knowledge and Abilities OutlinE:! (Tier 3) Form ES-401-3 Knowledge of the bases for prioritizing safety 2.4.22 4.4 22 functions during abnormal/emergency operations. Emergency Knowledge of low power / shutdown Procedures I 2.4.9 in accident (e.g., loss of coolant accident or loss of 3.8 Plan residual heat removal) mitigation I 2.4.42 Knowledge of emergency response facilities.

2.6 73 Subtotal 2 1 Tier 3 Point Total 10 7 ES-401 Record of Rejected KIA's Form ES-40 1-4 Tier I Group Randomly Selected KIA III SRO 295023 AA2.05 111 SRO 600000 2.4.29 I II RO 295018 2.1.23 2/1 SRO 207000 2.2.40 211 SRO 400000 2.4.11 211 SRO 215005 2.2.22 2/2 SRO 214000 2.1.36 2/2RO 215001 2.L23 3/RO 2.2.13 3/SRO 2.3.11 3/SRO 2.3.4 111 SRO 295021 AA2.02 III SRO 295026 2.2.25 111 SRO 295015 2.1.23 2/1SRO 262001 A2.06 2/1 RO 212000 K3.04 2/1 RO 264000 K4.06 21 I RO 239002 A2.05 211 RO 215004 A4.02 2/2RO 286000 K6.01 2/2 SRO 214000 2.2.40 3 SRO 2.l.25 295033 EAI.03 211 RO 259002 K1.03 2/2 RO 290001 A1.01 1/2 RO 295022 AA2.0J III RO 295034 2.2.36 Reason for Rejection 295023 AA2.02 Unable to develop 3 credible dislractors.

Rejected KIA and randomly selected a new KIA. 600000 2.4.18 Unable to devclop a quesdon linked to lOCtR55.43b.

A new KiA was randomly selected.

295018 2.2.38 KiA rejected due to CCW not being referenced in the Facility license. A new KiA was randomly selected.

207000 2.2.3 KIA rej<:cted due to Oyster Creek not being a "multi-unit" site. A new KiA was randomly selected.

400000 2.4.41 Unable to develop a question linked to IOCFR55.43b.

A new KiA was randomly selected.

215003 2.4.41 Unable to develop 3 credible disLractors.

Rejected KIA and randomly selected a new KiA. 214000 2.1.31 -KIA supports testing at the RO level, but not the SRO-Only level due to job responsibilities.

A new KIA was randomly selected.

215001 2.4.6 -There are no EOP actions associated with the TIP system therefore a question could not be written. A new KiA was randomly selected.

2.2.3 -KiA rejected due to Oyster Creek nDt being a "multi-unit" site. A new KIA was randomly selected.

2.3.5 -KiA rejected due to overlap with NRC question 71. A new KIA was randomly selected.

2.3.15 -KiA rejected due to overlap with NRC question 70. A uew KiA was randomly selected.

295021 AA2.05 Unable to develop an operationally relevant question to this KiA. A new KiA was randomly selected.

295026 2.2.37 Unable to develop an operationally relevant question to this KiA. A uew KiA was randomly selected.

295018 2.2.22 -KIA reJected since there are no Tech Spec l.COs associated with CCW. A new KiA was randomly selected.

262001 A2.08 -Unable 1.0 develop an operationally relevant question to this KIA. A new KIA was randomly selected.

212000 K3.03 -Unable to develop an operationally relevant question to this KIA. A new KIA was randomly selected.

264000 K4.03 Unable to develop an operationally relevant question to this KIA. A new KIA was randomly selected.

239002 A2.03 -KIA rejected dne to overlap with NRC Scenario 1 event 6 and Scenario 3 event 8. A new KIA wa.<, randomly selected.

209001 A4.02 Unable to develop 3 credible distractors.

Rejected KiA and randomly selected a new 286000 K6.04 -KiA rejected since Oystel Creek fire diesels do not have a Fuel Transfer system. Both fire diesels have their own independent fuel oil tank. A new was randomly 214000 2.1.36 -Unable to develop an operationally relevant question at the SRO-Only level for this KIA. A new KiA was randomly selected.

2.1.14-COUld not develop an operationally relevant question at the SRO-Only leveL A new KiA was randomly selected.

295012 AAL02 KIA rejected due to overlap with RO question #50. A uew KiA was randomly 259002 Kl.15 KIA rejected since the Recirc Flow Control System does not connect to the Feedwater Control System at Oyster Creek. A new KiA was 268000 A1.02 Unable to develop an operationally relevant question to this KIA A new KiA was randomly selected.

295008 AA2.04 Unable 10 develop an operationally relevant question to this KiA. A new KJA was randomly selected.

295007 2.2.12 -Unable to develop an operationally relevant question to this KiA Anew KiA was randomly selected.

ILT 11-1 NRC Scenario 1 (NEW) Scenario Outline Facility:

Oyster Scenario No.: 1 Op Test No.: 11-1 NRC Ol)erators:

Initial Conditions: 15% power with mode switch in RUN (IC 153) RWM is inoperable and bypassed

• Control Room HV AC System A is inoperable Turnover: Continue with rod withdrawal.

Complete step 24 Group 5-1. When rod pulls are complete wait for further direction from Reactor Engineering.

Event No. Malf. No. Event TypeO' 1 N/A N BOP Swap Service Water Pumps. 2 N/A R ATC Withdraw control rods to raise reactor power. 3 CRDOO8_ C ATC Respond to an uncoupled control rod >10% power. 3451 MAL-C BOPRespond to the loss of VMCC 182.EDSOO4B TS SRO RCPOO3D C BOP Respond to Recirculation Pump 0 inner seal failure, then 5 MAL-TS SRO outer seal Respond to the E EMRV lifting leading the crew to aC ATCmanual scram. 7 CAEP M Crew Respond to an Electric A TWS.ATWS.CAE C Crew Respond to Standby Liquid Control Pump shaft break. * (N)ormal, (R)eactlvity, (I)nstrument. (C)omponent. (M)ajor Transient, (T5) Tech Specs IL T 11-1 NRC Scenario Page 1 of 28 ILT 11-1 NRC Scenario 3 (Modified)

Scenario Outline Facility:

Oyster Scenario No.: 1 Op Test No.: 11*1 NRC Operators:

Initial Conditions: 85% power * 'B' RWCU Pump is OOS Turnover: Lower power to 80% using recirculation flow lAW 1001.22-3, Care Maneuvering Daily Instruction Sheet Backwash Main Condenser Half B South Event No. Malf. No. Event Type*

1 NA R ATC Lower reactor power to 80% using recirculation flow. 2 NA N BOP Continue backwashing Main Condenser Half B South. I BOP Respond to the EPR setpoint failing high. TCS010 BKR-C ATCRespond to CRD Pump A trip. CRDOO2 TS SRO TBCOO1A Respond to the trip of TBCCW Pump 1-3 and auto start C BOP BKR-failure ofTBCCW Pump 1-2. TBCOO3 MAL-I ATC Respond to a reference leg leak in the A & C GEMAC NSS012E TS SRO RPV level indicators ID13A and ID13C MAl-M Crew Respond to a multiple rod drift. CRD006 MAL-M Crew Respond to a Safety Valve lifting post scram NSS016A Respond to a trip of the operating Containment Spray CNSOO4A-C Crew Pump D (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs ILT 11-1 NRC Scenario Page 1 of 30 ILT 11-1 NRC Scenario 4 (NEW) (Backup Scenario)

Scenario Outline Facility:

Oyster Creek Scenario No.: Op Test No.: 11*1 NRC Examiners:

Operators:

Initial Conditions:

• The plant is at 95% power

• Dilution Pump 2 is tagged out of service

• Air Compressor

1. 3 is tagged out of service in PTL

• The RWM is inoperable and bypassed Turnover:

• Perform Turbine Valve testing lAW 625.4.002 Event No. Malf. No. Event Type*

1 N/A N BOP Tests MPR lAW 625.4.002 2 MAL* C ATC Respond to a CRD Flow Control Valve failed closed. CRD001A CRD011_1 C 3 415 Respond to trip of RPS MG Set 1 and a single rod scram. TS SRO SWI* C BOP 4 TBS027C Respond to a trip of Control Room Vent Fan B TS SRO ANN*L4f 5 I PSW-R ATC Respond to a major oil leak on 'B' Feed Pump requiring a CFW015A C BOP rapid power reduction.

MAL* CFWOO6C M Respond to a trip of the 'C' Feed Pump requiring a reactor 6 Crew MAL-C scram and a failure of all control rods to insert. CRD022 MAL-Respond to a Torus Leak requiring entry into Primary 7 PCNOO? M Crew Containment Control. VLV-Respond to Core Spray system suction valves being 8 CSS001, C Crew 009 mechanically seized when lining up the CST to the Torus. MAL* Respond to a Torus leak increase requiring the crew to 9 M Crew PCNOO? Emergency Depressurize.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs ILT 11-1 NRC Scenario 4 Page 1 of 31